ML20203M555

Requests That Proprietary AP600 Rept on Hydrodynamic Loads in Irswt Be Withheld Per 10CFR2.790

ML20203M555
Person / Time
Site:	05200003
Issue date:	02/27/1998
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19317C947	List: ML20203M555 NSD-NRC-98-5588, Forwards Proprietary & non-proprietary W Responses to FSER Open Items on AP600.Summary of Encl Responses Provided in Table 1.Proprietary Response to Part F of FSER 440.796F, Withheld,Per 10CFR2.790
References
AW-98-1212, NUDOCS 9803090073
Download: ML20203M555 (130)

Text

{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Westinghouse Energy Systems Ba 355 Electric Corporation Pmsburgh Pennsytvania 15230 0355 AW-98-1212 February 27,1998 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: MR. T. R. QUAY APPLICATION FOR WITilllOI. DING PROPRIETARY INFORMATION FROM PUBL!C DISCLOSURE

SUBJECT:

AP600 REPORT ON IIYDRODYNAMIC LOADS IN Tile IRWST

Dear Mr. Quay:

The application for withholding is submitted by Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), pursuant to the provisions of paragraph (b)(1) of Section 2.790 of the Commission's regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence. The proprietary material for which withholding is being requested is identified in the proprietary version of the subject report, in conformance with 10CFR Section 2.790, Affidavit AW-981212 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure. Accordingly, it is respectfully requested that the subject information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations. Correspondence with respect to this application for withholding or the accompanying affidavit should reference AW-98-1212 and should be addressed to the undersigned. i Very truly you, h-h f. h O Brian A. McIntyre, Manager Advance Plant Safety and Licensing jml cc: Kew. Hohrer NRC OWFN - MS 12E20 9803090073 990227 PDR ADUCK 05200003 E PDR

_ __-.._-._._ _ __. _ _ _. ~._._ _ _ _ _._ _ __..__ __ _ ___ __..__ _. _ _ _. AW-98-1212 4 i 4 AFFIDAVIT i i COMMONWEALTil OF PENNSYLVANIA: i r 1 l ss i COUNTY OF ALLEGilENY: Before me, the undersigned authority, pe sonally appeared Brian A. McIntyre, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief: X Af ,9 Brian A. McIntyre, Manager - Advance Plant Safety and Licensing Sworn to and subscribed before me this d 7T day of YM_ ,1998 0 &R. Shla-. Notary Pubiic NotarialSeal Lorralne M. PWica, Notary Public Monroevihe Boro. Allegheny County My Cornrnission Expires Dec. 14,1999 Marnber, Pennsylvania Assoca00n of Notvies 6e A wr'

AW 98-1212 (1) I am anager, Advance Plant Safety and Licensing, in he New Plants Projects Division, of the Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withhel from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit. (2) I am making this AfGdavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying thia AfGdavit. (3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as j confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the fo!!owing is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. I (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to ti.. public. Westinghouse has a rationai basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: 16H A apf l I

AW-981212 (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, iactuding test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor wouh reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensirg a similar product. J l (d) It reveals cost c-nformation, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. l (e) It reveals aspects of past, pre ent, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. 4 (f) It contains patentable ideas, for which patent protection may be desirable. There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors it is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) !t is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. Mil A wpf

AW.98-1212 (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense. (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the positior of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. l (f) The Westinghouse capacity to invest corporate assets in research and 1 devel ment depends upon the success in obtaining and maintaining a v competitive advantage. (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.799, it is to be received in confidence by the l Commission. (iv) The information sought to be protected is not available in public sources or available informatior. has not been previously employed in the same original manner or method to the best of our know! edge and belief. (v) Enclosed is Letter DCP/NRCl271 (NSD-NRC-98-5588), February 27,1998, being transmitted by Westinghouse Electric Company (W), a division of CBS Corporation (" Westinghouse"), letter and Application for Withholding Proprietary Information from Public Disclosure, Brian A. McIntyre (W), to Mr. T. R. Quay, Office of NRR. The proprietary information as submitted for use by Westinghouse Electric Company is in response to questions concerning the AP600 plant and the associated design certification application and is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of licensing advanced nuclear power plant designs. htt A opf

AW 981212 This information is part of that which will enable Westinghouse to: (a) Demonstrate the design and safety of the AP600 Passive Safety Systems. 1 (b) Establish applicable verification testing methods. (c) Design Advanced Nuclear Power Plants that meet NRC requirements. (d) Establish technical and licensing approaches for the AP600 that will ultimately result in a certified design. (e). Assist customers in obtaining NRC approval for future plants, Further this information has substantial commercial value as follows: (a) Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for advanced plant licenses. (b) Westinghouse can sell support and defense of the technology to its customers in the licensing process. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar advanced nuclear power designs and licensing defense - services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. ana.g l

AW.981212 The development of the technology described in part by the information is the result of applying the results of many years of experien:e in an intensive Westinghouse effort and the expenditure of a considerable sum of money, in order for competitors of Westinghouse to duplicate this information, similar technica; programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods. Further the deponent sayeth not. M11 A wpf

, to V.'estinghouse Letter Di'P/NRCl271 February 27,1998 .h m M114 epf

NRC FSER OPEN ITEM Revision 1 f '4 1 Question 440,788F (OITS #6445) The staff recognizes the importance of establishing a procesa for ensuring that the performance of the actual ADS valves in an AP600 plant meets functional requirements consistent with those determined from the design certification test program and reflected in design-basis analyses performed for the plant. Accordingly, the staff has determined that the ADS " road map" documented in Westinghouse letter NSD-NRC-976-5100, dated April 30,1997, should be incorporated into the AP600 SSAR. In addition, the steps in the " road map" leading from the design certification test program to the qualification of the actual AP600 valves should be incorporated into the inspections, tests, analyses, and acceptance criteria (ITAAC) for the AP600, and cross-referencing between the ITAAC, SSAR, and other appropriate documentation should be included, to ensure that the process is properly and consistently implemented. This is an Open item.

Response

The AP600 SSAR identifies the type of valves that will be used in the ADS. The portions of the ADS roadmap that are important to ensure that the ADS valves will be reliable include: ADS valve equipment qualification 1 ADS valve production operational verification Pre-operational valve operational verification in-service valve operational verification The ADS valves are discussed in several SSAR subsectionc; the primary subsection is 5,4.6 (ADS Valves) which is.n the RCS component and subsystem section (5.4). The important aspects of the ADS valve roadmap have been added to a new subsection 5.4.6.3 (Design Verification). The current SSAR subsection 5.4.6.3 (Inspection and Testing Requirements) is re-numbered 5.4.6.4. This new ! bubsection provides references for: - ADS valve qualification in SSAR subsection 5.4.8.1.2 (MOVs) and 5.4.8.1.3 (Other Power-Operated Valves including Squibs). - ADS valve production operational venfication will be performed. These tests are vendor specific and are not included in the SSAR. - ADS valve pre-operational valve operational venfication is addressed in SSAR subsection 14.2.9.1. - ADS valve i5 service valve operational venfication is addressed in SSAR subsection 3.9.6.2.2 and SSAR Table 3.9-16. The ADS valves are included in ITAAC 2.1.2 (RCS). Design requirement 12.a and its associated ITAAC in Table 2.1.2-4, have been revised to indicate that the design requirements applies to ADS valve W westingh0use masm L

NRC FSER OPEN ITEM operability and their flow capacity. This ITAAC requires type tests for MOVs and a venfication that the installed valve is bounded by the type tests. Revision 3 of this ITAAC did not include requirements for type tests for the ADS squib valves. An ITAAC has been added to require a similar type test for the ADS stage 4 squib valves. ADS valve flow capacity is included by reference to other RCS ITAAC items. SSAR Change: ( l Attachert is a revision to SSAR subsection 5.4.6. ITAAC Change: Attached is a revision to the ITAAC Design Description 2.1.2 and ITAAC Table 2.1.2-4, item 12.a. l k 440.788F(RI)-2

NRC FSER OPEN ITEM Revision 1 1 l Revision to SSAR Subsection 5.4.6 5.4.6 Automatic Depressurization System Valves The automatic depressurization system (ADS) valves are part of the reactor coolant system and interface with the passive core cooling system (PXS). Twenty valves are divided into four depressurization stages. These stages connect to the reactor coolant system at three differen* locations. The automatic depressurization system first, second, and third stage valves are included as part of the pressurizer safety and relief valve (PSARV) module and are connected to nazzles on top of the pressurizer. The fourth stage valves connect to the hot leg of each reactor coolant loop. The reactor coolant system P&lD, Figure 5.1-5, shows the arrangement of the valves. Opening of the automatic depressurization system valves is required for the passive core cooling system to function as required to provide emergency core cooling following postulated accident conditions. Operation of the passive core cooling system, including setpoints for the opening of the automatic depressurization system valves is discussed in Section 6.3. The first stage valves may also be used, as required following an accident, to remove noncondensable gases from the steam space of the pressurizer. (See subsection 5.4.11.) 5.4.6.1 Design Bases Subsection 5.4.8 discusses the general design basis, design evaluation, and testing and inspection for reactor coolant system valves, including the automatic depressurization system valves. The automatic depressurization system valves are seismic Category 1, AP600 equipment Class A components. (See subsection 3.2.2.) The fourth stage valves are interlocked so that they can not t's opened until reactor coolant system pressure has been substantially reduced. The design cntena and bases, functional requirements, mechanical design, and testing and inspection of the passive core cooling system are included in Section 6.3. The design requirements for the passive core cooling system also apply to automatic depressunzation valves except where the requirements for reactor coolant system valves are more restnctive. 5.4.6.2 Design Description The first stage automatic u urization system valves are motor-operateo 4 inch valves. The second and third stage au'ork. ; depressunzation system valves are motor-operated 8-inch valves The fourth stage automate depressurization system valves are 10 inch squib valves arranged in senes with nprmally-open, de powered motor-operator valves. See Section 6.3 for a discussion of the sizing of the automitic depressurization system valves. The control system for the opening of the automatic copressuitzation system valves, as part of the passive core cooling sys;em, has an appropnate level of diverse and redundant features to minimize the inadvertent opening of the valves. 440.788F(R1)-3 F

NRC FSER OPEN ITEM Revision 1 7 IS! 1 For each discharge path a pair of valves are placed in series to minimize the potential for an inadvertent discharge of the automatic depressurization system valves. The fourth stage valves are interlocked so that they cannot be opened until reactor coolant system pressure has been substantially reduced. The first. second, and third stage valves are located on the pressurizer safety and relief valve module clustered into two groups. Each group has one pair of valves for each stage. The two groups are on different elevations and are separated by a steel plate. Vacuum breakers are provided in the AP600 ADS discharge lines to help prevent water hantmer following ADS operation. The vacuum breakers limit the pressure reduction that could be caused by steam condensation in the discharge line and thus limit the potential for liquid backflow from the in containment refueling water storage tank following ADS operation. A bypass test line is connected to the inlet and outlet of the first, second, and third stage upstream isolation valves. This bypass line can control the differential pressure across the upstream valves dunng inservice testing. The bypass test solenoid valves do not have a safety-related function to open. S.4.6.3 Design Verification The automatic deoressurization system valves are venfied to meet their safety-related functional requirements by the following: Valve equipment qualification Pre operational valve operational venfication In service valve operational venfication Automatic depressurization system valve qualification is addressed in SSAR subsection 5.4.8.1.2 for the stage 1/2/3 motor operated valves and in subsection 5.4.8.1.3 for the stage 4 squib valves. The equipment qualification i'iciudes type testing which venfies the automatic depressunzation system valve operability and flow capacity. Automatic depressurization system valve pre-operational valve operational verification is addressed in SSAR subsection 14.2.9.1. Automatic depressunzation system valve in-service valve operational venfication is addressed in SSAR subsection 3.9.6.2.2 and SSAR 1hle 3.916. l S.4.6.4 Inspection and Testing Reqv.m;nents The requiren)ents for tests and inspections for reactor coolant system valves is found in subsection 5.4.8.4. In addition, tests for the automatic depressunzation system valves and piping are conducted dunng preoperational testing of the passive core cooling system, as discussed in Sections 6.3 and 14.2. T wesungnouse " #88"""

NRC FSER OPEN ITEM Revision 1 itit; t::: n l l 5.4.6.4.1 Flow Testing Initial verification of the resistance of the automatic depressurization system piping and valves is performed during the piant initial test program. A low pressure flow test and associated analysis is conducted to determine the total piping flow resistance of each automatic depressurization system valve group connected to the pressurizer (i.e. stages 13) from the pressurizer through the outlet of the downstream valve. The reactor t.colant system shall be at cold conditions with the pressurizer full of water. The normal residual heat removal pumps will be used to provide injection flow into the reactor coolant system, discharging through the ADS valves. Inspections and associated analysis of the piping flow paths from the discharge of the automatic depressurization system valve groups connected to the pressurizer (i.e. stages 1-3) to the spargers shall be conducted to verify the line routings are consistent with the line routings used for design flow resistance calculations. The calculated piping flow resistance from the pressurizer through the sparger, with valves of each group open is shown to be less than that used in the design basis LOCA analyses presented in Section 15.6. l [ Westiflgh0USS 440.788F(R1)-5

NRC FSER OPEN ITEM Revision 1 [ E[ Revision to ITAAC 2.1.2 Design Description 8. The RCS provides the following safety related functions: a) The pressurizer safety valves provide overpressure protection in accordance with Section ill of the ASME Boiler and Pressure Vessel Code, b) The reactor coolant pumps (RCPs) have a rotating inertia to provide RCS flow coastdown on loss of power to the pumps. c) The RCS provides automatic depressurization during design basis events. 9. The RCS provides the following nonsafety related functions: a) The RCS provides circulation of coolant to remove heat from the core, b) The RCS provides the means to control system pressure.

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

11. a) Controls exist in the MCR to cause the remotely operated valves identified in Table 2.1.2-1 to perform active functions.

I l L) The valves identified in Table 2.1.2-1 as having protection anc' safety monitonng system (PMS) control perform an active safety function after receiving a signal from the PMS. c) The va'ves identified in Table 2.1.21 as having diverse actuation system (DAS) control perform an active safety function after receiving a signal from DAS. l

12. a) The automatic depressurization valves identified in Table 2.1.21 perform an active safety-related function to change position as indicated in the table.

b) After loss of motive power, the remotely operated valves identified in Table 2.1.2-1 assume the indicated loss of motive power position. y,,, 440.788r<ni>-6

NRC FSER OPEN ITEM Revision 1 ys H j 1 Table 2.1.2-4 (Cont.) Inspections. Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 11.c) The valves i) Testing will be performed on the

1) The semib valles receive a identified in Table 2.191 squib valves identified in Table 2.1.2-signal at the valve electrical as having DAS control 1 using real or simulated signals into leads that is capable of actuating perform an active safety the DAS without stroking the valve, the squib valve, function after receiving a signal from DAS.

ii) Testing will be performed on the other remotely operated valves li) The other remotely operated identified in Table 2.1.2-1 using real valves identified in Table 2.1.21 or simulated signals into the DAS. as naving DAS control perform i the active !Jnction identified in the table after receiving a signal l from DAS. 12.a) The automatic i) Tests or type tests of motor.

1) A test report exists and depressurization valvet.

operated valves will be performed concludes that each motor-identified in Table 2.1.21 that demonstrate the capability of the operated valve changes position perform an active safety-valve to operate uno~er its design as indicated in Table 2.1.21 t related function to change conditions. under design conditions. position as indicated in the table. ii) Inspection will be performed for ii) A report exists and concludes the existence of a report verifying that the as installed motor-that the as-installed motor operated operated valves are bounded by valves are bounded by the tests or the tests or type tests. type tests, iii) Tests or type tests of squib iii) A test report exists and valves will be performed that concludes that each squib valve demonstrate the capability of the changes position as indicated in valve to operate under its design Table 2.1.2-1 under design conditions. conditions. iv) Inspection will be performed for iv) A report exists and concludes the existence of a report verifying that the as installed squib valves that the as-installed squib valves are are bounded by the tests or type bounded by the tests or type tests. tests. h

NRC FSER OPEN ITEM Revision 1 7$ l 1 v) See item 8.c.i in this table. v) See item 8.c.I in this table. The ADS stage 13 valve flow resistances are verified to be consistent with the ADS stage 1 3 path flow resistances. vi) See item 8.c.li in this table. vi) See item 8.c.ii in this table. 3 The ADS stage 4 valve flow resistances are ventied to be consistent with the ADS stage 4 I path flow resistances. vii) See item 8.c.iii in this table. vii) See item 8.c.iii in this table. [E viii) See item 8.c.iv in this table, viii) See item 8.c.iv in this table. 8

NRC FSER OPEN ITEM Question: 440.789F (OITS #6446) The ADS test program was discussed in a meeting between Westinghouse, the staff, and the ACRS Subcommittee on Thermal Hydraulic Phenomena on December 9 10, 1997, in that meeting, questions were raised regarding Westinghouse's evaluation of the data from the test program, particularly the calculation of key thermal-hydraulic parameters. To respond to the ACRS concerns, Westinghouse must revise the Test Analysis Report (TAR) for the ADS test program to more completely discuss how the data were evaluated, including assumptions made with respect to thermal hydraulic conditions in the test facility (e.g., assumed negligible or unimportant effects), and how these assumptions affect the inferred performance characteris-tics of the ADS valves in the AP600 plant. These revisions should also provide complete justification as to why the range of thermal hydraulic conditions covered by the ADS l test program, and the data acquired therefrom, comprise an adequate basis for validation of code models for ADS performance analysis. Revision of the ADS TAR is an Open item.

Response

The Rev,2 issue of WCAP-14305; AP600 Test Program, ADS Phase BI Test Analysis Report has been transmitted via letter DCP/NRCl270, <tated February 27,1998. SSAR Revision: None 440,789F-1

M=

ii J NRC FSER OPEN ITEM Question 440,795F (OITS. 6440)

Westinghouse stated during the December 10,1997, ACRS Thermal / Hydraulic Subcommittee meeting that changes had been made to the NOTRUMP code numerics. The code numerics must be descnbed in full detail, including derivation of all diffe ence forms of the equations being solved In addition, every equation altered or changed from the form that exists in the onginal, approved NOTRUMP code must be provided with denvation complete ta the level of the difference form in the code.

Response

] Re original NOTRUMP central numenes, which are documented in Reference 440.795F 1, were modified for application to AP600. The main modification to the numerics was the change from the net mass now based to the net volumetric flow based momentum conservation equation. Additional modifications to the numerics were the implicit treatment of bubble rise and droplet fall, and the implicit treatment of gravitational head. Rese modifications are desenbed in Reference 440.795F 2 (in Sections 2.4,2.9, and 2.11). To address the above comments on documemation, a new Section 2.20 is being added to Revision 4 of Reference 440.795F-2. With this new section, changes. to the NOTRUMP central numerics, which were made for AP600, are documented in one place, from beginning to end. He derivations, which are required to modify the startmg differential equations of Reference 440.795F 1, and then place them into finite-difference form for solution, are presented. The changes to the NOTRUMP central numencs for AP600 affect the momentum conservation equation for each non-critical now link (i.e., Equation 2 33 in Reference 440.795F-1), and indirectly affect the energy and mass conservation equations for the mixture and vapor regions of each interior Guid node (i e., Equations 21 through 2-4 in Reference 440.795F-1). He modifications for the volumetric Dow-based momentum formulation affect the structure of the starting momentum conservation equation, as well as the finite-differencing of all of the aforementioned conservation equations. De modifications for the implicit treatment of bubble rise and droplet fall only affect cenain implicit terms in the finite-differencing of the energy and mass conservation equations for the mixture and vapor regions of each interior fluid node. De modifications for the implicit treatment of grautational head only affect certain implicit terms in the finite-differencing of the momentum conservation equation for each non-entical flow link. In the new Section 2.20 in Revision 4 of Reference 440.795F 2, the modifications to the momentum conservation equation for non-cntical flow links are documented. This includes the changes to the structure of the equation for the volumetric How-based formulation, followed by the finite differencing denvations. Next, the finite-differencing denvations for the energy and mass conservation equations of the mixture and vapor regions of the interior fluid nodes are documented De section concludes with a presentation of the modifications to the NOTRUMP central matnx and its elements, which is analogous to what is docurrented in Appendix E in Reference 440.795F 1. 3 West lI1gh00S8 1

NRC FSER OPEN ITEM

References:

440.795F-1 Meyer, P. E., "NOTRUMP: A Nodal Transient Small Break And General Network Code." Westinghouse Electric Corporation, WCAP 10079 P-A, August 1985. 440 795F 2 Fittante, R. L., et. al., "NOTRUMP Final Validation Report for AP600," Westinghouse Electnc Corporation, WCAP 14807 Revision 3, November 1997. 4 SSAR Revision: None 440.795F 2 g 3, -__A

NRC FSER OPEN ITEM Question 440.7%F Part a (OITS 6441) The following commitments were made by Westinghouse at the conclusion of the December 10. 1997 ACRS T/H Subcommittee meeting and must be fulfilled. Momentum flux - Deficiencies (in the NOTRUMP model) are to be benchmarked against a. additional detailed calculations using actual two phase flow aquations that include the effects of compressibility, including the condition of constant entropy. l

RESPONSE

In section 1.7.5 of the Final Validation Report for NOTRUMP, an assessment was performed of the effect of ignoring the momentum flux terms. 'Ihis initial assessment indicated that while the ADSl 3 valves and piping would experience a small effect due to fluid acceleration, in the ADS 4 piping the effect could be significant. To further evaluate whether the lack of momentum flux terms for this component in NOTRUMP could lead to erroneous results, a detailed pipe model was developed. The modelintegrates the momentum and energy equations along a detailed mesh representing the ADS 4 piping from the hot leg to the squib valves, where the minimum area occurs. First, the model of the ADS 4 piping will be described. A comparison will then be made with flows calculated by NOTRUMP. Detailed model of the ADS 4 oicinn Figures 440.796f l and 440.796f 2 show two views of the ADS 4 valves and their piping. A pipe of inner diameter 10.125 inches (0.56 sq. ft area) is connected to the top of the hot leg. An cibow - turns the pipe to a horizontal configuration. About 7 feet downstream, a horizontal tee diverts some of the flow into the 8.5 inch (0.39 sq. ft area) piping leading to one of the two valve packages (the pipe from the tee to the valves is designated " branch 2" in this response). Downstream of the tee in the main pipe, a reducer leads to the 8.5 inch diameter piping which will lead to the other valve package (the main piping and :his valve package are designated " branch 1 "). The flow resistance (irreceverable losses due to friction and form loss) through this piping network has been conservatively established for incompressible flow. Bounding tssumptions have been used for pipe length (about 46 feet total, in contrast to the typical configuration shown in Figure 440.796f 1), and fittings (a total of 6 elbows are assumed in branch 1, and 7 elbows in branch 2, compared with the smaller number in the typical configuration). The total irrecoverable loss coefficient for this conservative configuration was estimated as 4.2, based on the nominal flow area through both branches (2*0.39 sq. ft), assuming complete turbulence (constant) friction factors. T Westinghouse i

NRC FSER OPEN ITEM This piping network was simulated with a total of 442 cells, as illustrated in Figure 440.796f 3. In this figure, each "+" represents one cell boundary or node. The cell length is 0.25 feet, with smaller increments taken at area changes in the reducer and gate valve (there are also area changes at the tees from the hot leg into branch 1, and from branch 1 to branch 2, but these are treated with special models as discussed below). Momentum and Enerev Eauations / ] The momentum and energy equations to be integrated along the piping network are simplified equations in which steady state, equilibrium, homogeneous, adiabatic conditions have been assumed. The assumption of homogeneity (zero slip) results in a high estimate of the effect of acceleration on the pressure gradient, as pointed out in Section 1.7.4 of the NOTRUMP Final Validation report. The momentum and energy equations in this form are: dP fvt (W '$ h 2 W du -= 3 dz 2D < A > A di r 23 1(a,b) d h+u =0 dz 2, y where h and u are the mixture enthalpy and velocity, W is the mixture flow rate, v is the liquid r specific volume, @2 is the two phase multiplier, f is the friction factor and D and A are the pipe diaineter and area. Since: W = uA 2 v where v is the mixture specific volume, and W is constant, this substitution can be made into equations I and 2. In addition, since for homogeneous flow: 440.796F part a-2 3 Westinghouse

y NRC FSER OPEN ITEM h = h + Xh, f f 3(a,b) y = v, + xv, 7 where x is the flow quality WJW, equations 1 and 2 can be set up in terms of pressure and quality After some manipulation, 1 dA 'dh dv'dP 22 2 Gv +Gv 3 L A dz <dP BP>dz dz h, + G vv, 4 f f a 1dA b dP c A dz c dz fvt 2 2 a h 1 dA G @, + g dy + G L v - v, c,A dz f dP _ 2D v dz 5 dz ' dv b> 2 1+G <dP- - vf, c, where dy/dz is the elevation gradient. Friction and form losses Friction and form losses are calculated using two phase multipliers developed by Collier and 2 others. R: two phase multiplier for losses in both pipes and fittings takes the basic form: 2 $, = (1 - x)2$,2 @, = l + C 1 6(a,b) 7 + X X* where C is a value which depends on the fitting or pipe, and on the fluid conditions, and where X is the Leckhart Martinalli parameter, defined by: 3 Westinghouse

-{ l l' NRC FSER OPEN ITEM r gp> # dz n 1-x 'vf X= m 7 d[ ) v, x dz>,, g in equations 6 and 7 above,it has been assumed that the single phase loss coefficient and/or friction factors are independent of Reynolds number (mass velocities are sufficiently high such that this assumption is reasonable). Ilts Tees require special treatment because flow splitting and phase se,paration will occur, Methods summarized in Lahey (1984)' were used to calculate pressure le sses, Rese methods attribute s pressure changes in the main pipe due to momentum chtnge (modified by a pressure recovery term K.2), defined by (equations (19] snd [20] of Reference 2): i K -' (V G * - V[G * ) h2 M-2= 2 i 5.0 K -2 = 1 I + 8(a,b) i f yi7 GD i i sM/> where 1 denotes the main pipe upstretm of the tee,2 denotes the main pipe downstream of the tee, and (assuming homogeneous conditions): v;=vf+xviy 9(a,b) v=v7+xv2y 2 where xi and x2 are the flow qualities upstream and downstream of the tee (see below). The pressure change into the branch (denoted as 1) consists of an acceleration change:nd a form loss. The form loss is calculated as described in the previous section. The acceleration change is 440.796F port a-4 3 Westinghouse

= NRC FSER OPEN ITEM given by (equation [13] of Reference 3 for homogeneous flow): 23 v G _ v;2gi 1 2 AP _3 = - i 3 3 10

• " A I,3

/ The assumption is made that the flow is locally incompressible at the tee junction. The flow split at the tee is calculated using correlations recommended by Seeger (1986)' These correlations describe the quality into the branen (3) in terms of: S = f 11 G, x, g The functional form depends on the tee orientation. For the vertical tee (from the hot leg), equation [2] of Reference 4 is applied, while for the horizontal tee, equation [8] of Reference 4 is applied. Critical flow at the squib valve By careful integration of the momentum and energy equation, it is possible to find the maximum flowrate (choked flow ) in a pipe, by finding the point at which the denominator in the momentum equation approaches zero. Because the last valve in each branch is the minimum flow area (a design requirement), choking is likely to occur at this valve (this was confirmed by later calculations). The HEM was applied at the last cell in each branch, using as reservoir conditions (for branch 1, for example): 2u' kt = hn+ g 301 = Si * *> t It where jl is the next to last node in branch 1. The flow was assumed to be adiabatic and frictionless from this point to the squib valve minimum area. A similar calculation was performed for branch 2. Calculated resuits The equations above were implemented in a small computer program; the flowrate through ADS 4 was calculated for a range of pressures (20 to 80 psia in the hot leg,14.7 psia at the exit) and T Westinghouse

NRC FSER OPEN ITEM qualities (20% to 100% in the hot leg). The model was benchmarked to the incompressible tota! loss coefficient by running a case with a hot leg pressure of 15 psia and 100% quality. At this low pressure difference, effects of compressibility and acceleration are minimal and the predicted loss should agree with the incompressible value. An adjustrrent of approAimately 10 percent in the overall resistance was required to achieve good agreement with the loss coefficient of 4.2. It was also assumed that all the hot leg now entered the ADS 4, for maximum acceleration. Figures 440.796f 4 and 5 show the static pressure and Guld velocity in the piping for a hot leg pressure of 50 psia and a flow quality of 100%. Here is an immediate 5 psi pressure drop at the entrance, then a pressure loss followed by a pressure ruovery at the tee, then additional losses along the pipe and at each cibow. At the first valve, there is a pressure loss, then recovery, followed by a pressure loss (the irrecoverable loss due to the valve is applied at the valve, exit). Figure 440.796f 5 shows that the Guld velocity within the pipe is highest at the hot leg entrance, reaching nearly 800 ft/s. Most of the acceleration occtirs, however, at tht: squib valve. Figures 440.796f 6 to 8 show conditions when the now quality is 20%, In this case, phase ~ separation occurs at the tee, with a higher quality mixture Dowing into branch 2 (Figure 440.796f 8). Calculations assuming no separation occurs show that this phenomenon has a negligible effect on the amount of vapor whlen can be vented. Vapor flow versus hot leg pressure for a range of now qualities prMicted by the model are shown in Figure 440.7%f 9. ne 100 percent quality data show good agreement with points estimated from a handbook, shown in Figure 1.7 9 of Reference 1. At approximately 40 psia, critical flow is calculated to occur at the squib valve, and the model and handbook data begin to diverge. The model calculated data were used to generate a response surface which were then used to calculate ADS 4 vapor Dow, given hot leg pressure and quality from NOTRUMP. This comparison is shown in Figure 440.796f.10 for the time period between ADS 4 opening and IRWST injection. nese figures show good agreernent between NOTRUMP and the detailed model, as long as critical now conditions exist pnor to 3000 seconds. De good agreement during choked now indicates that the now resistance upstream of the squib valve has a minor impact on the now, even with the relatively high fluid velocities noted. When the flow becomes sub critical, NOTRUMP predicts a higher vapar flow of about 20 percent (cornsponding roughly to a 35 percent lower now resistance). Overall, the total vapor vented is underpredicted by NOTRUMP soon after opening, then is overpredicted, as seen in Figure 440.796f l!. However, the NOTRUMP total vapor released is only about 5 percent higher at the time the IRWST comes on. Compadson of NOTRUMP and model details indicates that the difference du@g sub-critical now can be attributed to: 440.796F port a 6 gp

O NRC FSER OPEN ITEM a) Underestimation of the two phase pressure drop through fittings. The large number of elbows assumed in the ADS 4 piping, fo; example, contributes to a 20 percent increase in now resistance, b) Underestimation of the acceleration terr,s. Even if the flow is no longer critical, fluid acceleration and expansion at low quality will contribute '.o increased pressure drop. Conclusion The over prediction by NOTRUMP of ADS 4 vapor now near the end of the transient is not considered to be a signjricant problem because of the bounding nature of the ADS 4 pip!ng which was modelled. As shown in Figure 440,796f 1, the number of cibows and lengths of piping in actual designs will be substantially less than what was assumed in the NOTRUMP calculation. In addition, pressure losses due to acceleration were maximited by assuming homogeneous fluid conditions.

REFERENCES:

' WCAP 14807, Revision 3. Collier, J. G., Convective Boiline and Condensation,3"' Edition, Oxford Clarendon press, 1998 3 Lahey, R. T., Nematollah, S.,"The Analysis of Phase Separation Phenomena in Branching Conduits", int. J. Multiphase Flow, Vol.10, No.1,1984.

• Seeger, W., et al.,"Two Phase Flow in a T Junction with a Horizontal Inlet", Int. J. Multiphase Flow, Vol.12, No. 4,1986.

440.796F port o 7

Ft [IL-J NRC FSER OPEN litM Figure 440.796f 1. Typical AP600 ADS 4 piping layout. View 1. be oa m s'.0"_ l'* 6" 6'a t" l l n r/ $0o s / f r i ) n/ M.L'. ~. w.!.L!.i_. - h / a /' / / r y ii . fat / w' a _.+W A L . / // // A g.g/ . _/_ 3; c ') .+ h 3 et.s.. E'..;as " r $,r ao L' io s" 3l 440.796F port a*8 Westingtmuse

II "{ NRC FSER OPEN ITEM f figure 440.796f 2. Typical AP600 ADS 4 piping layout. View 2. / l f / / / "CS :008C u248 i "C3 18C8C L1283 A / Y { acs :p yeg4ef acs,8 <tede l l 1 i $, L. i. m.cir. I 1 l 7 l I ACS 18 v8t*0 8 5 :8 votes l

• CS testa (i338 l

p / .Cs a8r. u3r8 e M 8

NRC FSER OPEN ITEM Figure 440.796fd, ADS 4 piping flow area distnbution ADS 4 PIPING FLOW AREA AREA (SQ. FT) 2-18 16 1.4 - - 12- - 1 08 BRANCH 1 REDUCER 06- - 04 + GATE V 02 BRANCH 2 TEE gogig Y8 0 0 5 10 15 20 25 30 35 40 45 50 DISTANCE FROM HOT LEG (FT) 440.796F part a 10 W Westinghouse

lIl NRC FSER OPEN ITEM Figure 440.796f 4. Static pressure in ADS 4 piping for hot leg pressure = 50 psia, quality = 100 %. STATIC PRESSURE IN ADS 4 PIPING (P=50,X=100%) PRESSURE (PSIA) 50 BRANCHi REDUCER 45 -y 40 -- g W 1, 35 GiTE VALVES 30 25 - SQUlBVALVES ++ 20 0 5 10 15 20 25 30 35 40 45 50 DISTANCE FROM HOT LEO (FT) 440.796F part o 11

NRC FSER OPEN ITEM Figure 440.796f.$. Fluid.elocity in ADS 4 piping for hot leg pressure = $0 psia, quality = 100%. FLUlO VELOCITY IN ADS 4 PIPING (P0=50,X=100%) VELOCITY (FT/S) 1600 1400

• +

I 1200-- 1000 - - 600 - -

:t 600 p_r----',

400 -- + BRANCH 1 200 - - P 0 t 3 5 10 15 20 25 30 35 40 45 50 OISTANOE FROM HOT LEO (FT) 440.796F part a.12 T Westinghouse

NRC FSER OPEN ITEM Figure 440.796f 6. Static pressure in ADS 4 piping for hot leg pressure o 50 psia, quality = 20%. STATIC PRESSURE IN ADS 4 PIPING (P=50,X=20%) PRESSURE (PSIA) 50 BRANCH 1 45 "Y 40 -- + g "\\. N 35 - BRANCH 2 ~a 30 .4 v. 25 -- + 20 + 15 - - 10 O 5 10 15 20 25 30 35 40 45 50 OlSTANCE FROM HOT LLG (FT) 440.796F part a 13 l

NRC FSER OPEN ITEM Figure 440.796f.7. Fluid velocity in ADS 4 piping for hot leg pressure = 50 psia, quality = 20%. FLUID VELOCITY IN ADS 4 PIPING (P0=50,X=20%) VELOCITY (FT/S) 900 + 800- - 700 - - 600 500 + BRANCH 2 400- - .n ~ 300 r 200-r- 100 BRANCH 1 0 O 5 10 15 20 25 30 35 40 45 50 OlhTANCE FROM HOT LEG (FT) 440.796F port a 14 T Westinghouse

r NRC FSER OPEN ITEM Figure 440.796f 8. Flow quality.n ADS 4 piping for hot leg pressure = 50 psia, quality = 20%. QUALITY IN ADS 4 PIPilNG (P0=50, X0=20%) QUAllrr 04 0.35 - - 03-. BRANCH 2 0 25 -- 02 0 15 - + Oi BRANCH 1 0 05 - - 0 1 O 5 10 15 20 25 30 35 40 45 50 OlSTANCE FROM HOT LEO (FT) 440.796F part o 15 T Westinghouse

1 l-NRC FSER OPEN litM Figure 440.796f.9. Vapor flow vs hot leg pressure and quality predicted by pipe model, ADS 4 VAPOR FLOW PREDICTED BY MODEL MASS FLOWRATE (L8/S) 80 '. ' 0 - - eMODEL aCRANE 60 - + e e FLOW QUALITY RANGE. 0 2 TO 10 50 - a e s 9 40 - - g e 9 4 e 30 - - e e e 4 g e s e 20 -- I9 9 e 4 10 O O 10 20 30 40 50 60 70 80 HOT LEO PRESSURE (PSIA) 440,796F port a 16 g

NRC FSER OPEN ITEM Figure 440.796f 10. Comparison of NOTRUMP and pipe model vapor flows ADS 4 VAPOR FLOW FLOWRATE (LB/S) 70 4 NOTRUMP 60 - MODEL \\ k 50 - - f 40 - - 'I 30 - t ll 'l j' [;.pgkY' Y If 10 - - 0 2000 2200 2400 2600 2800 3000 3200 2400 3600 TIME (S) 440.796, part a -i7 y ming, p M

i NRC FSER OPEN litM Figure 440.796f l1. Comparison of NOTRUMP and pipe model vapor flows (integral) INTEGRAL OF VAPOR FLOW I Mass (ts) 25000 NO' RUMP - - - MODEL ~' 20000 -- 15000- - MODEL 10000 - - $000 0 2000 2200 2400 2600 2800 3000 3200 3400 36 % TIME (S) 440.796F part a la Westinghouse

NRC FSER OPEN ITEM Question 440.796F Part b (OITS 6441) The following coramitments were made by Westinghouse at the conclusion of the December 10.1997 ACRS T/H Subcommittee meeting and must be fulfilled, b. ADS l 3 the test data analysis r: port is to be reviewed to assue that the data reduction was performed correctly.

Response

Westinghouse has reviewed the methodclogy used in the data reduction and has incorporated several modifications discussed at the December 10,1997 ACRS T/H Subcommittee meeting. Consideration of the source stagnation enthalpy used to determine the local Guid qualities and the velocity head of the fluid upstream of components at choked now ct nditions are the most significant items. These, along with other items, will be documented in the revis d ADS Test Analysis Report (WCAP 14305. Rev. 2), expected to be submitted to the Staff by February 27,1998. SSAR Revisiont None 440,796F Part b 1

y 2 NRC FSER OPEN ITEM Question 440.796F Part c (OITS. 6441) The following commitments were made by Westinghouse at the conclusion of t e December 10,1997 ACRS T/H Subcommittee meeting and must be futulled.

c. Entrainment consider as part of the overall scaling and level penalty development

Response

Pha.w separation assc,ciated with the Hot Leg (HL)/ ADS 4 T junctions is ranked as a High/ Medium phenomena during IRWST injection in the AP600 SBLOCA PIRT. While a general analytical method does not exist for predicting this complex two phase now phenomena, experimental work by Seeger, et. al. (Reference 440.796F c 1) end Mudde, et. al. (Reference 440.7%F-c 2) with vertically-oriented branch geometries at different scales (similar to AP600 HUADS 4 T junction) suggests that this phenomena can be correlated with diameter and massnow ratios between le main run of a T junction and its branch. Based upon preservation of massuow ratio, bottom up (i.e. diameter ratio) scaling indicates that both SPES 2 and OSU Hot leg / ADS T junction diarneter ratios are very well scaled to the full scale AP600 plant. Therefore, phase separation phenomena in the HUADS 4 T junctions should also be well-scaled. Further details on this subject can be found in References 440.7%F-c 3. Since this phenomenon is well scaled in OSU, the level penalty which is based on OSU data (see RAI 440.721 (g) contained in Appendix A of Reference 440.796F c 4), can be applied to AP600. References 440 796F c l. W. Seeger, J. Reimann and U. Muller, "Two phase Flow in a T Junction with a Honzontal Inlet", Int. J. Multiphase Flow, vol.12 No.4, pp. 575 585,1986. 440.796F c 2. R. Mudde J. Groen H. van den Akker, "Two phase Flow Redistnbution Phenomena in a Large T junction", Int. J. Multiphase Flow, vol.19, No. 4, pp. 563 573,1993. 440.796F-c.3. AP600 Scaling and PIRT Closure Report, WCAP 14727 Rev. 2. 440 796F c-4. NOTRUMP V&V Report, WCAP-14807, Rev. 4 SSAR Revision:' None 440.796F part c 1

NRC FSER OPEN ITEM Question 440,796F Part d (OITS. 6441) The following commitments were made by Westinghouse at the conclusion of the December 10,1997 ACRS T/H Subcommittee meeting and must be fulfilled d. Level Penalty a multiloop scaling analysis is to be performed for the time penod of ADS 4 and IRWST draining. Justify the basis for ADS How, ADS 4 flow affected by entrainment of liquid and the corresponding effect on the pressure loss due to two phase now. This is to be desenbed in sufGeient detail that a step towards scaling to AP600 can be made.

Response

A rnultiloop scaling analysis has been performed for the time frame between ADS 4 and IRWST draining. The technical areas desenbed in the above FSER Open item hase been addressed and will be documented in the revised "AP600 Scaling and FIRT Closure Report"(WCAP 14727) expected to be issued by February 27.1998. SSAR Revision: None 440.796F Part d 1

g NRC FSER OP5N ITEM Question 440.796F Part e (OITS 6441) The following commitments were made by Westinghouse at the conclusion of the December 10.1997 ACRS T/II Subcommittee meeting and must be ful0lled.

c. SurFe Line Flooding an efr t similar to that applied to the level penalty is to be made.

or Responset

1. Introduction The AP600 pressurizer surge line is composed of various sections of vertical pipe, inclined straight pipes, and inclined helical elbows, as shown in Figures 1 1 and 12. An attempt was made to find the limiting section for the CCFL and the scaling effect so that the liquid downtlow rate from the pressurizer can be determined.

Geometrical characteristics of the surge lines for AP600, OSU, and SPES test facilities are sum-marized in Table 1 1. Fluid properties in Table 12 are for saturated steam and liquid at 35 psia. This pressure is assumed when liquid draining from the pressurizer becomes imponant to regulat-ing IWRST liquid injection rate to the primary system. CCFL in inclined Pipe vs. Vertical CCFL ll Wallis' model for CCFL in a vertical pipe is given in terms of dimensionless volumetric tluxes by . t/: . i/: (j,) + m(-j, ) =C (1 1) wherej = j,./p,/(gDap) with steam superficial velocity,j,, and similarly ji* is dc6ned. C ranges from 0.75 to 1.0. This equation is for a small pipe; scaling effects will be discussed later. The flooding in an inclined straight pipe is estimated by applying Taitel Dukler's flow regime transition from a co-current strati 6ed flow 23 to a counter-current strati 6ed flow. It is composed of the Kelvin Helmholtz critical flow condition for instabili'y of stratified flow (Section 1.1) and the ste.idy state flow balance (Section 1.2), i 440.796F part e 1 l l n

i NRC FSER OPEN ITEM l.1 Kelvin Helmhyltz Critical Condition _ The Kelvin Helmhcitz critical condition can be expressed in terms of Wallis' dimensionlest sclu-metric dux.j

• g "t/:

a e 50

• j,2 C, (g,3) da

-D7g. g where D is the pipe diameter, h is the liquid level of the stratified flow, at is the void fraction, and i 0 is the pipe inclination. Various values are assigned to C by investigators but Taitel Dukler pro-2 posed C, = 1 h /D. Eq. (1 2) is a funttion of the liquid level or the void fraction. Fig.1 3 for cos0 i - 1. It should be noted that this condition applies for both co-current and counter current flows. 1.2 Force Balance Relatiom Steady state force balance for counter current flow is JP' ~ 4 ~ 4 team b <sp, M - A, y + tw,S,-t,S, + p, A,g sin 0 = 0 (1-4) where tw and t, are the wall shear and interfacial shear and S, S and S are the wall (for each i s i phase 1 and interfacial surface areas per unit length. From the difference of these two equations the interfacial force balance is obtained. S e S S S, 8 - t w, i + t w, + + + apgsin0 = 0 (1 5) w here the approximation tw, a t employed by Taitel Dukler has been used. In addition, it is 3 assumed that: SPt(lI)I B cpg (lg)l.8 twt = 11 and Twg ' l's 2 440,796F part e 2 Westinghouse

NRC FSER OPEN ITEM S 5 wherefi andfg are the liquid and steam smooth wall friction factors except for the velocity in the Reynolds number, which is combined with j2. Eq. (15)is, therefore, a function of the superti-cial selocities ji,j, and the void fraction a. Elimination of the void fraction from Eqs. (I 2) and t 1 Si gives a critical curve in the coordinate system of the superficial velocities. The obtained curve is translated to the dimensionless coordinate system ([, [,) and compared with the sertical CCFL in Figures 1-4 and 15 for OSU and AP600 surge lines, respectively. it is seen that the CCFL in the vertical sezion is more limiting than the inclined pipe section aiid controls the liquid downflow rate from the pressurizer, provided that centrifugal effects are small (Section 3.2).

2. Scaling for CCFL in A Vertical Pipe In terms of dimensionless pipe.ameter, D = DIN (2 1)

\\ Rc0a the following dimensionless flux was defincoT I,aj, D' # (2 2) r K', where -- = min K6, ? K ,K (2 3) D D. where Ko a 0.645 and K = 3.2 Wallis' vertical CCFL equation is generalized to i . iz: . v: (I,) + m(-l, ) =1 (2 4) such that the critical Kutateladze number for liquid hold up is satisfied, in fact, for smail pipes of D' < 25 [D <-2.5") Eq. (2 4) becomes . v: . v: (j,) + m(-j, ) = 0.803 (2 5) w hile it becomes [ Westlnghouse 440.799F part e 3

O NRC FSER OPEN ITEM ( k,' ) ' ' + m(-k, ) = 1.79 (2 6) for large pipes, D* > 25. When k * = 0, Eq. (2 6) becomes k * = 3.2, which is the Kutateladze liq. i g uid hold up condition, where Kutateladze number is P:'

• k,aj,4 (2*71

1. 1. ccop Eq.(2 6) becomes

. ie2 . i/: 3 *T (),) + m(-j, ) 2<0.8 (2 8) = This means that Eq. (2 6)is more restrictive than the Wallis flooding curve, Eq. (2 5). For example, at.15 psia: Tla ( Wallis D* = D (126. (1/ft)) c = 35.9 for OSU U = 151.6 for AP600. 3 Figure 21 Scaling Effect in Vertical CCFL [ In the previous section, it is concluded that the Wallis' CCFL is more limiting than the inclined pipe. Since larger vertical pipe is more restrictive, the conclusion that the CCFL in the vertical section is limiting still holds.

3. CCFL in An Elbow batween the Inclined Pipes Assummg that the vertical section is controlling the CCFL in the pressurizer surge line, the steam riow rate required to hold up the liquid is estimated. With this magnitude of steam tiew, what would happen in the elbows due to centrifugal force is studied in this section.

3 i The Minimum Steam Flow Rate for Liouid Hold Un For SPES, the pipe is small so that Wallis* CCFL shows dj['*) = 0.75, where j ** is the steam g 440.796F part e 4 i )

p: Aij{ NMC FSER OPEN ITF.M flow at Ji

0. On the other hand, the pipes of OSU and AP600 are large pipes so that the hold up poir. is detined as k *'"" = 3.2. These values are converted to the superficial velocities; g

j,'"" = 28.17 ft/see for SPES, j,'"" = 42.65 ft/sec for OSU and AP600. (3 1) Correspondingly, the dimensionless volumetric fluxes are: ))["" = 0.75 for SPES = 0.731 for OSU = 0.510 for AP600 (3 2) With this steam flow assumed in the elbow,it can be determined if a flow regime transition from stratified flow could take place inside the elbow, which would cause liquid hold up in the inclined cibow sections. 3 2 Centrifugal Force vs. Gravity (Reference 4) Banerjee (Reference 4) has shown that the liquid phase may liow on the inside of the elbow as illustrated below (flow inversion). pipe cross section Centrifugal Force = pR($)' = p(u,) g Steam Location of Liquid R 3 o tan 6 = '"' ~ E '"': (3 3) Rapg 6 -.- G ui Figure 31 Centrifugal Force vs. Gravity if the elbow were the limiting CCFL, the critical situation would take place at high soid fraction, w hen cz a -0.8. Therefore, the super 6cial selocities, Eq. (3 l), are approximated by j, = u, t3 4) When the liquid is being held up in the sertical section, the centnfugal force on the steam is esti-mated assunung the liquid is motionless. 440.796F part e 5

I NRC FSER OPEN ITEM ( m > ': (m> 2 (4.46 x 10") <3 5) tan 6 = = Rapg R = 0.092 (6 = 5.3 deg)....... OSU = 0.023 (6 = 1.3 deg).... AP600 Thus, the angle of flow inversion is small and the gravity effect is much more important than the centrifugal force if the liquid is still. 33 Liouid in Motion When the liquid moves, the liquid phase may move on the outside of the elbow (6 < 0) as indi-cated below. 3.3 i Balanced Centrifugal Force Centrifugal forces exened by steam and liquid can cancel each other resulting in 6 = 0. The liquid velocity at 6 = 0 is estimated while the steam flow is holding up liquid in the vertical section, b j ' (3 6) u, = 9 = (0.0379)*(42.65) = 1.61 ft/sec...... OSU & AP600 312 Natu al Fall of Liquid Now esamine the natural tiow of liquid in an inclined straight pipe 2 L9t t u (3 7) p,Lgsine = f3 with f = 0.015. Assume the smallest inclination: AP600 D = 1.2 ft and 0 = 2.5 deg............ ui = 15.0 ft/sec OSU D = 0.285 ft and 0 = 2.0 deg.. ui = 6.5 ft/sec The corresponding effect of centnfugal force is tan 6 = -(15.0)2 / (42/12)(32.2) = 2.0 ( 6 = -63 deg) .. AP600 s -(6.5)2 /(10.5/12)(32.2) = 1.5 ( 6 = 56 deg) ..OSU The draming liquid is therefore on the outside of the elbow for zero steam flow, as expected. 440.796F part e 6 3 Westingh0USS

NRC FSER OPEN ITEM 3.33 Effect of Steam Motion on Liquid Flow When steam tiow rate given by Eq. (3 2), liquid can flow through the inclined straight pipe at the rate of [ = 1.125 for AP600 from Fig.15 = 0.625 for OSU from Fig.1-4. Correspondingly, ji = 7.88 fusec for AP600 ji=1.183 fusec for OSU Applying Eq. (3 2) to Figure 1 1, stratified flow vold fractions are: a 2 0.63 for AP600 l n 2 0.80 for OSU l Therefore, the liquid velocity in the elbow becomes: ui = 21.3 fusec for AP600 ui a 5.9 fusec for OSU at least. These values are approximately the same as the netural flow rates obtained in the previous-l section and these are substantially larger than the equilibnum velocity for 6 = 0. Thus, the centrif-l ugal force on the liquid phase is large for OSU and AP600 and the liqitid phase will remain on the j outside of the elbow, when the liquid is being held up in the vertical section of the surge line. When the liquid flow is on the outside of the elbow, the body forces exerted upon ti.2 liquid (vec. torial sum as shown in Figure 3 2) are stronger than in the straight pipe section. So, the Kelvin. Helmholtz condition in the elbow section is less likely to be reached than the straight section. Thus, the flooding condition in the straight inclined section is to be more limiting, and the conclu. sion that flooding in the vertical sections controls pressurizer draining remains valid. 440,796F part e 7

NRC FSER OPEN ITEM R R \\ V 6<0 5>0 V Figure 3 2 Flow inversion Effect to Centrifugal Force and Gravity 34 Vertical Elbow vs. Vertical CCFL The vertical pipe section of the surge line is connected to the inclined straight pipe section via a sertical elbow, I I l PRIZER I ~7Y \\ Curvature l Di ter AP600 42" 14.44" R I OSU 10.5" 3.55" "~ Vertical Elbow The CCFL in this section of the surge line is estirnated by regarding the elbow as a series of inclined straight pipes. As the inclination angle 0 increases, Kelvin Helmholtz condition is pro-440.796F part e 8 Westingh00S8

NRC FSER OPEN ITEM portional to square root of cose so that the condition is nearly independent of the inclination for 0 < 60 degrees or so. On the other hand, the gravity effect in the flow force balance. makes liquid flow rate proportional to square root of sin 0 so that the effect becomes very large with increasing inclination. So. a !arge amount of liquid can flow without disturbing the stabili*y of the stratified flow, which can be seen in Figure 14 by the increasing limit lines. The gravity body force acts to resist the instability caused by local depressurization due to steam flow. This effect is quickly reduced around 0 = 90 degrees and it is replaced by the surface tension as evident from the Kutateladze number. However, the effect of the surface tension is substantially weaker than the gravity so that the steam flow rate required for liquid hold up is observed to be small(see Figure 21). This should 50e the reason why the surfacc tension effects can be ignored for the flow regime transition in the nearly horhontal pipes. This implies that the tiow instability and CCFL is the most limiting at 0 = 90 degrees when the gravity effect becomes zero. !n addition, the centrifugal force and gravity force make the liquid flow take place on the outside of the elbow and this third force also acts to suppress the surface wave disturbance. Once again,it is concluded that the vertical section of the surge line is the most limiting CCFL.

4. Conclusions it is concluded in Section 3 that the CCFL in the straight inclined pipe section of the pressurizer surge line is more restrictive than the elbow section. In Section 1, the CCFL for the vertical sec-tion is more restrictive than the inclined pipe section. Scaling effect of the vertical CCFL is meluded in Eq,(2 4). It should be noted that Eq. (2 4)is mere limiting than Wallis CCFL for larger pipes such as OSU and AP600. Therefore, it is concluded that the CCFL in the pressurizer surge line is limited by Eq. (2 4), and controls the liquid draining rate from the pressurizer.

References:

1. G. B. Wallis,"One dimensional Two phase Flow," McGraw Hill, trie., N. Y. (1969) 440.796F part e 9 T Westinghouse

NRC FSER OPEN ITEM 2. Y. Taitel and A. E. Dukler,"A Model for Predicting Flow Regime Transitions in Horizon-tal and Near Horizontal Gas Liquid Flow" AIChE J.,22,47 (1976) 3. K. Takeuchi, M. Y. Young, and L. E. Hochreater," Generalized Drift Flux Correlation for vertical Flow," Nucl. Sci. & Eng.,112,170 (1992) 4. S. Banerjee. E. Rhodes, and D. S. Scott," Film inversion of Co Current Two Phase Film in Helical Coils," AICHE J., 13, 189 (1967). 440.796F part e 10 Westinghouse

NRC FSER OPEN ITEM Table 1 1 Surge Line Characteristics of Test Facilities inclination. Curvatute I.D. D/R liq. How 0 R D AP600 13 deg inf. 54" *3 14.438" 0.267 2.5 deg*5 42" *2 1.203 ft 0.34 OSU 12.5 deg in6nity 4.8 deg 13.5" *3 3.548" 0.26 3 deg 10.5" *i 0.285 ft 0.34 1.8 deg SPES 24 deg ~6" 1.338" 0.223 13 deg 0.112 ft 2.5 deg Taitel-1.5. 03, in6nity 1.97"(5cm) 0 down Dukler .1 deg 0.164 ft Grolman -0.5

l. deg in6nity 5.1. 2.6 cm 0

up Whalley 6 deg 19.7" (.5 m) 0.80"(2cm) 0.041 up 0.066 ft Study 13 deg 1.203 ft down 7.7 deg 0.285 ft 2.5 deg Shown m the above Table are ti.. scalings ofpressurizer surge lines of AP600. OSU. and SPES totfacilitv. Flow regime transitsonsfrom horizontal stratifiedflow to intermittent or annular despersedflow were investiga;ed by Taitel Dukler. Grolman, and Whalley. The scalings of their test sections are also shown in the table. From these scalings, cases to be studied are selected; that is, the pipe mclinations are 13. 7.7. and 2.3 degrees andpipe diameters are 1.203 and 0.285 ft. The positive inclination means that the liquidflows down by that angle ofinclination. For etample. Whalley's test section is a helically inclined pipe at 6 degrees and liquid is injected from the bottom of the test section (therefore the angle is given as negctive). R and D are the cunature of elbows and the pipe diameter, respectively. D/R is a scalingfor the helical effect. etc. It can be seen that Whalley's test section has a smaller D/R compared to OSU and AP600 elbows. 440.796F part e 11

NRC FSER OPEN ITEM Table 12 Saturated Fluid properties at 35 psia Density ilb /ft ): p = 0.08406 pi = $8.54 i 3 m g i Es = 8.687 x 10 6 p = 1.533 x 10 '4 l Viscosity (Iby/ft sec) : 2 Kinematic Viscosity (ft /sec): ", = 10.334 x 10'8 vi = 0.2619 x 10'8 Surface Tension (Ib/ft) : o = 3.680 x 10 3 s NOMENCLATURE: wbscripts: 1 Liquid phase g Steam phase wperscnpts: am) Steam tiow at liquid hold up (ji = 0). Dimensionless quantity 2 A Steam flow area (ft ) g 2 Ai Liquid flow area (ft ) C Flooding constant. Eq. (l 1) 440.796F part e 12 4

NRC FSER OPEN ITEM l l C2 Multiplier for Kelvin Helmholtz. Eq. (12) = 1 h /D i 3 D Hydraulic diameter (ft) D' Dimensionless hydraulic diameter, Eq. (2 1) f,s Wall steam friction factor (except for phase velocity if Reynolds number)((fusec)0 2) f,5 Wall liquid friction factor ((fusec)o.2) 2 g Acceleration of gravity (fusec 3 ] ge Conversion factor (ib ft /lbr ec) s m h Liquid level in the pipe (ft) i j Superficial velocity (volumetric flux)(fusec) j' Dimensionless volumetric flux K generalized critical Kutateladze number, Eq. (2 3) Ko coefficient to the critical Kutateladze number for a small pipe Ki critical Kutateladze number for a large pipe k* Kutateladze dimensionless volumetric flux, Eq. (2 7) l' Generalized dimensionless voiametric flux, Eq. (2-2) m Flooding coefficient, Eq. (1 1) P Pressure (psia) R Radius of co vature (ft) S Wall liquid surface area per unit length (ft) i S Wall steam surface area per unit length (ft) g i interfacial surface area per unit length (ft) S ui Liquid velocity (fusec) u, Steam velecity (fusec) Coordinate along the pipe length (ft) s Greek: a Void fraction S Pipe cross sectional angle for location of liquid film Eq. (3 U 9 Angular velocity (rad /sec) 2 v Kinematic viscosity (ft /sec) 0 Angle of pipe inclination k 440 796F part e -13

NRC FSER OPEN ITEM 3 p Density (lb /ft ) m SP Pi Pg o Surface tension (lbr/ft) 2 twi Wallliquid shear (Ib /ft sec ) m 2 t; Interfacial shear (Ib /ft sec ) m s 8 Definition of friction factorsfi andfg : The original definition of friction factors: + Tw:

• ft Twg
• fy 2

f, = 0.046 ,up-0 2 <D <D u /t = 0.046 W> -0 2 < vg > g y, 4Ag 4A* D=S D = g 8 S, + Sg In terms of Superficial Velocity:_ .18 .l.8 Twt = 1 Tw =l 2 r s ! A >2 ! 4 A ' ~0 2 3 f A *2 ( 4A

• ~0 2 f, =

Ap (0.046) S y fs = 1< A,,(0.046) (S, + S;)v,, <gp as a result of: ls ' A ' ' * -0 t I}, "I

• ll 3

1. 0 8 '

f, = f A > l 8 0.2(0.046) 11 u, \\Ap <vi g < A,2 ' 4 A ' ~0 2 = - (0.046) S y <Ap <gp 440.796F part e -14 W Westingh00S8

NRC FSER OPEN ITEM O e \\ ez e. st / m Y / f /\\ / Q '\\R, / w; gg v e- \\ .u 4'* ., \\ g o 4g/ 1 1 e, 4 / pigurel'1 050 W er Sut'1' 4403969 9 #

NRC FSER OPEN ITEM l'fas eteeseette e. ./.. 7 'g 7.. . s,, N ' s, / I (p.. tT r s 4........,~..

QBf,

/< .x. s ,;;p.

• • ~

/ q; "Jg'ln x y/

• C.....

/A e E g p.g Ishap. a. F l y -Q r e.IT. i n. Figure 12 AP4 '.'nssurizer Surge Line 440.796F part e -16 W85 tin @US8

NRC FSER OPEN ITEM i i i i 2 15 F (d ( i .534 05 .26 02 04 06 08 ai Figure 1-3 Kelvin-Helmholtz Crideal Condidon as Modified Sy Taitel Dukler 440.796F part e -17

NRC FSER OPEN ITEM 3 i i i 4 ss p SQRTjg5 4.0 SQRTjs3i,; I SQRT;g3 i,2 14) tI( al I ..,"*.,44) ( ) g3 (2> l l Wallis 0 O O5 I l.3 2 SQRTjf5 i.0,5QRTjf5 g,g,SQRTjf5 i. 2 * * ' "I (1) Vertical CCFL, sat steam-water 3 35 psia (2) Taital-Dukler, 0 = 2.5 deg, sat steam-water 3 35 psia (3) Taitel-Dukler, 0 = 7.7 deg, sat steam-water 3 35 psia 1 (4) Taitel-Dukler, 0 = 13. deg, set steam-water 3 35 psia Figure 1-4 Wallis Vertical CCFL vs Taitel Dukler in An Inclined Pipe with Diameter 3.548" [OSU] 440.796F part e 18 3 Westinghouse

NRC FSER OPEN ITEM I i i

-t Gg f 3

= 5QRTjg5 4.0 SQRTjs3,,, 5QRTjg5 3 i.2 i.U > '.....,c - ,o o i .',[4 v..., 08 i33 (2) 1) Wallis N 0 O O5 l.3 2 SQRTjG i. 0 ' 39I ' '"' i. t. 5QRT]5 i. 2 '" ' Al l# (Ij (1) Vertical CCFL, sat steam-water a 35 psia (2) Taital-Dukler, 9 = 2.5 deg, sat steam-water a 35 psia (3) Taitel-Dukler, 9 = 7.7 dog, sat steam-water 4 35 psia (4) Taitel-Dukler, 9 = 13. dog, sat steam-water G 35 psia Figure 1-5 Wallis Vertica'. CCFL vs Taitel Dukler in An Inclined Pipe with Diameter 14.438" [AP600] W Westinghouse 440.796F part e 19

NRC FSER OPEN ITEM SSAR Revision: None 7 4 3 i i t t 440.796F part e -20 W Westinghouse 1 i

r.; :m:ML NRC FSER OPEN ITEM Question 440.7%F Part f (OITS 6441) The following commitments were made by Westinghouse at the conclusion of the December 10,1997 ACRS T/H Subcommittee meeting and must be fulfilled. f. Noding provide more justification for the basis used whict. "irfers from the accepted approach deseloped under the CSAU work, especially for the PRHR ano downcomer.

Response

A description of the noding utilized for the AP600 analyses using hdTRUMP is included as subsection 1.16 in Revision 4 of WCAP 14808, NOTRUMP Final Vahdation Report for AP600. Other subsection numbers of Section 1.0 will be changed appropriately. It should be noted that based upon this study, there was a sensistivity discovered for [ ]" in the downcomer for AP600; OSU results were insensitive to the number of downcomer nodes. For AP600, this sensitivity occurs for DEDVI and larEer breaks. Use of the [ 1" node downcomer model results in a possible increase of PCT: however, this PCT is well below regulatory limits. Westinghouse will update the SBLOCA section of the AP600 SSAR for those breaks sensitive to the number of downcomer nodes. SSAR Revision: Westinghouse will update the SBLOCA section of the AP600 SSAR for those breaks sensitive to the number of downcomer nodes. A 440.796F part f -1

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION Revision 2 V $ l Short Term Availability Control Questions Question 440.804F (OITS #6478) Section A of SECY 94 084, " Policy and Technical Issues Associated with the Regulatory Treatment of Non Safety Systems (RTNSS) in Passive Plant Designs," March 28,1994, discusses the process used (a) to develop insights regarding the importance of non-safety-related systems to the overall safety of the AP600 design, and (b) to determine what, if any, additional regulatory controls should be implemented for those non safety related systems determined to be important to safety. Chapter 22 of the FSER discusses the RTNSS process in detail. Westinghouse's original evaluation of RTNSS implementation is discussed in WCAP 13856, "AP600 Implementation of the Regulatory Treatment of Nonsafety Related Systems Process," in addition, the focused PRA sensitivity study that forms a major part of the RTNSS process is contained in Chapter 52 of the AP600 PRA. The origincl evaluation in WCAP 13856 identified only two conditions requinng regulatory controls on non-safety related systems: the reactor trip function of the Diverse Actuation System (DAS) for mitigation of anticipated transients without scram (ATWS), and the normal residual heat removal system (RNS) and supporting fluid and ac electrical and systems for operations during midloop conditions. However, after extensive discussions with the staff, Westinghouse has agreed to expand the number of SSCs covered by RTNSS and to expand the MODES during which RTNSS controls apply, as discussed in the attachment to Westinghouse letter NSD NRC 97 5485, dated December 12,1997. RTNSS oversight is accomplished through administrative controls on the identified SSCs, which specify operability requirements, required actions and the time to accomplish those actions if the operabihty requirements are not met, surveil lance requirements, t nd the bases for the control3. However, there are no limiting conditions for operation associated with these RTNSS controls. The staff has reviewed the administrative controls related to shutdown operations (MODES 5 and 6), and has identified a concern related to the allowed completion time for required actions dunng periods of reduced inventory. The proposed adminCrative control for RNS during MODES S and 6 specify that both RNS pumps should be available prior to entry into MODE 5 with the pressure boundary open or MODE 6 with upper internals in place and the cavity level less than full. If one RNS pump subsequently fails. the operator is permitt9d up to 72 hours to remove the plant fro n the MODE in which these controls are apphcable. The staff has concluded that this time is Ncessive when the plant is operating in a reducedanventory condition. The short refueling schedules proposed for the AP600 mean the plant could be in reduced-inventory conditior's for a relatively short time. Thus, it could be possible to enter reduced inventory operations, tnen have the RNS or one of its supporting SSCs become inoperable, but with the 72 hour action completion i,.ne, necessary work could be completed and the plant could exit the MODE within the time specified for operator action. (The same action times are specified for RNS support systems, such as component cooling water. service water, and on-site ac power.) Thus, for reduced inventory operations in the applicable MODES, the 72-hour completion time effectively serves no safety purpose. The staff thus concludes that the administrative controls on RNS and supporting SSCs for reduced inventory operations during MODES 5 and 6 are not conservative and are not consistent with the safe shutdown objective and that action completion times when operability requirements are not met should be more restrictive and consistent with the length of time the plant is 440.804F(R2)-1

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION Revision 2 expected to be in reduced inventory operations in the applicable MODES Resolution of this issue is an Open item, 4

Response

An additional action has been added to these short term availability controls (for short-term availability controls 2.2, 2.3,2.4 and 3.2) which requires the operators to initiate actions within 12 hours to increase I the water inventory above the core in the RCS or the refueling cavity as appropriate. The Dasis were l revised to indicate that the RCS water level should be increased to 20% pressunzer level or to a full i refuehng cavity. The twtJve hours is considered appropriate for the following reasons: Initiating actions within 12 hours is a reasonable balance between taking rapid actions which might lead to mistakes and doing nothing for most of the 72 hours (available to take the plant out of the applicable MODE). Increasing the water inventory, vill increase the time before steaming would occur in case the second RNS, CCS, or SWS pump is lost. The RNS only provides a nonsafety related means of removing decay heat at reduced inventory conditions. The PXS provides the safety related means to remove decay teat in these conditions through the use of passive feed and bleed cooling (IRWST injection and ADS venting). From a PRA perspective the PXS capability is more reliable than the RNS / CCS / SWS because of its redundancy, simplicity, and passive nature. The PXS features are required to be available by Technical Specification LCOs 3.5.8 (IRWST) and 3 4.14 (ADS). In cases where one of these redundant components is inoperable, the Technical Specifications allow 72 hours to restore the component.- Durt iat time the plant does not have to take action to exit the reduced inventory condition. If the component is not restored within the 72 hour Action Time, the plant must immediately initiate actions to remove the plant from the recuced inventory condition. No time limit is specified for lening the reduced inventory condition. 3 The plant is only expected to be in ieduced inventory for limited times during a refueling outage. However, in the shutdown PRA the time that the plant was assumed to be in reduced inventory conditions is 120 hr / year. This longer time was selected to cover unexpected delays dunng refueling operations and forced outages for maintenance repairs. In addition, the PRA assumed that the reduced inventory condition associated with retum to power had the same risk as the one assoctated with shutting down. As a result, the PRA over estimates the nsk importance of the reduced inventory condition. The AP600 s'hutdown PRA takes no credit for the reduced inventory RNS/CCS/SWS short-term . availability controls. The plant is assumed to stay in reduced inventory for the w. tale 120 hr / year even when one RNS / CCS / SWS pump fails. Pequiring the plant to make sudden changes in plant operations can lead to increased risk. This philosophy is also used in selecting Technical Specification Action Times. 040.804F(R2)-2 1

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION hvision 2 PE l SSAR Change: .)= Revise short term availability enntrols 2.2, 2.3,2.4 and 3 2 in SSAR section 16.3. ITAAC Change: None R 4 440.804F(R2) 3 s

f!!I $N![+

16. Technical Specifications i

Table 163 2 (Cont.) INVESTMENT I4.0TECYlON SHORT TERM AVAll.AllII,lTY CONTROLS 2.0 Phtnt Systems 2.2-Normal Residual lleat Removal System (RNS) RCS Open - OPER ABILITY: Both RNS pumps should be operable for RCS cooling APPLICABILITY: MODE 5 with RCS pressure boundary open. MODE 6 with upper intemals in place or cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump not operable. A.! Initiate actions to increase the 12 hours water inventory above the core. AND A.2 Remove plant from applicable 72 hours MODES B. Required Action and B.I Submit repon to [ chief nuclear iday associated Completion officerl or lon-call attematel Time not met. detailing interim compensatory measures, cause for inoperability, and schedule for restoration to OPERABLE. AND B.2 Document in plaat records the i month justification for the actions t. Ken to restore the func% a to OPERABLE. Resision:-21

77. IM 3 Westinghous?

163 18 so.8o4F(R2) 'l

16. Technical Specifications

( ) Tabb 16.3-2 (Cont.) INVESTMENT PROTECTION SHORT. TERM AVAILABILITY CONTROLS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R 2.2.1 Verify that one RNS pump is in operation and that each Within I day prior to RNS pump operating individually circulates reactor entering the MODES coolant at a flow > [9001 gpm of applicability OR Verify that both RNS pumps are in operation and circulating reactor coolant at a flow > (1800] gpm 7 Revision: 21 T Westinghouse in.3. i9 ??.I m 440,selv/Ad*5

I!!F O!!! Hi ~

16. Technical Specifications n,,

Table 16.3 2 (Cont.) - INVESTMENT PROTECTION SHORT TERI-l AVAII.ABII.ITY CONTROLS -2.0 Plant Systems 2.2 RNS - RCS Open !!ASES: The RNS cooling function provides a nonsafety-related means to normally cool the RCS during shutdown operations (h10 DES 4. 5,6). This RNS cooling function is importat during conditions when the RCS pressure boundary is open and the refueling cavity is not ikxxied because it reduces the probability of an initiating event due to loss of RNS cooling and because it pmvides margin in the PRA sensitivity performed assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. De RCS is considered open when its pressure boundary is not intact. The RCS is also considered open if there is no visible level in the pressurizer, The margin provided in the PRA study hssumes a minimum availability of 90% for this function during the MODES of applicability, considering both maintenance unavailability and failures to operate. He RNS cooling of the RCS involves the RNS suction line from the RCS HL, the two RNS pumps and the RNS discharge line returning to the RCS through the DVI lines. The valves located in these lines should be open prior to the plant entering reduced inventory conditions. One of the RNS pumps has to be operating; the other pump inay be operating or may be in standby. Standby includes the capability of being able to be placed into operation from the main control nxim. Refer to SSAR - section 5.4.7 for additional infonnation on the RNS. Both RNS pumps should be available during the MODES of applicability when the loss of RNS-cooling is risk imponant. If both RNS pumps are not available, the plant should not enter these conditions. If the plant has entered reduced inventory conditions, then the plant should take action to I restore full system operation or leave the MODES of applicability. If the plant has not restoredf4dl l system operation or left the A10 DES of applicability within 12 hours, then actions need to be initiated-I to increase the RCS water level to either 209 pressuri:er level or to a fidl refueling cavirv. Pl.umed maintenance affecting this RNS cooling function should be perfomied in MODES 1. 2,3 when the RNS is not nonnally operating. The bases for this recommendation is that the RNS is more risk important during shutdown MODES, especially during the MODES of applicability conditions than during other MODES when it only provides a backup to PXS injection Revision: 21 ??,1998 3 Westinghouse th3 2n 410. 80 M Y

16. Technied Sptcincatio s y

] Table 16.3 2 (Cont.) INVESTMENT PROTECTION SHORT. TERM AVAILAllit.ITY CONTROL.S 2.0 Plant Systems 2.3 Comp < ment Cooling Water System (CCS)- RCS Open OPERABILITY: Both CCS pumps should be operable for RNS cooling APPLICABILITY: MODE 5 with RCS pressure boundary open, MODE 6 with upper internals in place or cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump not operable. A.! Inh. - ms to increase the 12 hours water mventory above the core, AND A.2 Remove plant from applicable 72 hours MODES B. Required Action and B.I_ Submit report to [ chief nuclear iday associated Completion of ficerl or [on-call attemate] _. Time not met. -detailing interim te:npensatory measures, cause for inoperability, and schedule for restoration to OPERABLE. AND B.2 Document in plant records the I month justification for the actions taken to restore the function to OPERABLE. Resision: 21 3 WestingtI00$8 t h.L21 ??,I M Wo so'/F /'4 2) - 7

T!!

16. Technical Specifications

/$ # Table 16.3 2 (Cont.) INVESTMENT PR(rir.CTION SHORT TERM AVAILAHil.lTY CONTROLS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR. 23.'s Verify that one CCS pump is in operation and each CCS Within i day prior to pump operating individually pmvides a CCS tiow thniugh entering the MODES one RNS heat exchanger > [2520] ppm of applicability OR Verify that both CCS pumps are in operation and the CCS liow through each RNS heat exchanger is > [2520] gpm e Resision: 21

77. 1998

[ Westingh0USS 16.3-22 ygo,604 F/R 2) - 6

lli" d!E It. Technical Specincations Table 16.3 2 (Cont.) INVESTNIENT PROTECTION SHORT TERN 1 AVAll.AllIIlTY CONTROI.S 2.0 Plant Systems 2.3 CCS - RCS Open BASES: %c CCS cooling of the RNS HXs provides a nonsafety-related means to nonnally cool the RCS during shutdown operations (N10 DES 4. 5. 6). This RNS cooling function is important because it reduces the probability of an initiating event due to loss of RNS cooling and because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety-related SSCs to mitigate at-pimer and shutdown events. De RCS is considered open when its pressure boundary is not intact. The RCS is also considered open if there is no visible level in the pressurizer. The margin pnwided in the PR A study assumes a minimum availanility of 909 for this function during the N10 DES of applicability. considenng both maintenance unavailability and failures to operate. The CCS cooling of the RNS involves two CCS pumps and HXs and the CCS line to the RNS HXs. The valves around the CCS pumps and ilXs and in the lines to the RNS HXs should be open prior to the plant entering these conditions. One of the CCS pumps and its HX has to be operating. One of the lines to a RNS HX also has to be open. The other CCS pump and HX may be operating or may be in standby. Standby includes the capability of being able to be placed into operation from the main control niom. Refer to SS AR section 9.2.2 tor additional infonnation on the CCS. Both CCS pumps should be available dunng the N10 DES of applicability v, hen the loss of RNS cooling is risk important. If both CCS pumps are not available. the plant should not enter these conditions. If the phmt has entered these conditions, then the plant should take action to restore both 1 CCS pumps or to leave these conditions. If d.e plant has not restoredfull system operation or left the i AIODES of applicahdity within 12 hours, then actions need to be initiated to increase the RCS water I les el to either 20'R pressuri:er level or to a full refueline cavity. Phumed maintenance af fecting this CCS cooling function should be perfonned in N10 DES 1. 2. 3 when the CCS is not supporting RNS operation. The bases for this recommendation is that the CCS is more risk important during shutdown N10 DES especially dunng the N10 DES of applicability conditions than during other N10 DES. Revisiem: 21 W Westinghouse in.+ 23 ?? 1998 qqc. ht;lF/2 L) - 9

"'{'ilf .i ~

16. Technical Specifications

! n. Table 16.3 2 (Cont.) i INVESTMENT-PROTECTION SHORT. TERM AVAll.AlllLITY CONTROL.S 2.0 Plant Systems 2.4 Service Water System (SWS) RCS Open OPERABILITY: Both SWS pumps and cooling tower fans should be operable for CCS cooling APPLICABILITY: htODE 5 with RCS pressure boundary open. N10DE 6 with upper internals in place or cavity level less than full ACTIONS CONDITION REQUIRED ACTION CONtPLETION TlhtE A. One pump or fan not A.1 Initiate actions to increase the 12 hours

operable, water inventory above the core.

AND A.2 Remove plant from applicable 72 hours N10 DES B. Required Action and B.!- Submit report to [ chief nuclear i day awwiated Completion otticerl or [on call attemate] Time not met. detailing interim compensetory measures, cause for inoperability, and schedule for restoration to OPERABLE. AND B '. Document in plant records the i month justification for the actions taken to restore the function to OPERABLE. m Revision: 21 ??,1998 3 Westinghouse 16.3-28 yyo.bc 6 M U 4 B

rmn.mm w

16. Techtlegl Specifications

~ EAM - Table 16.3 2 (Cont.) INVESTMENT PROTECTION SHORT TERM AVAILABILITY CONTROLS SURVEILLANCE REQUIREMENTS SURViiiLL.tNCE FREQUENCY S R 2.4.1 Verify that one SWS pump is operating arid that each Within I day prior to SWS pump operating individually provides a SWS Ilow entering the MODES > [6200] gpm of applicability SR 2.4.2 Operate each cooling tower fan for > 15 min Within i day prior to entering the MODES of applicability i 4 m Revision: 11 [ W85tingil00Se 16.3-25 ??.I M 440. ScM F(R.8 -l /

]lj7 Mi

16. Technical Specifications Table 16.3 2 (Cont.)

INVESTN1ENT PROTECTION SilORT TERN 1 AVAll.Allit.ITY CONTROL.S 2.0 Plant Systems 2.4 SWS - RCS Open BASES: The SWS cooling of the CCS IlXs provides a nonsafety related means to normally cool the RNS flX which cool the RCS during shutdown operations (N10 DES 4,5,6). This RNS cooling function is imponant because it reduces the pnibability of an initiating event due to loss of RNS cooling and because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety-related SSCs to mitigate at-power and shutdown events. The RCS is considered open when its pressure boundary is not intact. The RCS is also considered open if there is no visible level in the pressurizer. The margin provided in the PRA study assumes a minimum availability of 90% for this function during the N10 DES of applicability, considering both maintenance unavailability and failures to operate. The SWS cooling of the CCS tlXs involves two SWS pumps and cooling tower fans and the SWS line to the RNS IlXs. De valves in the SWS lines should be open prior to the plant entering these conditiors. One of the SWS pumps and its cooling tower fan has to be operating. De other SWS pump tu.d cooling tower fan may be operating or may be in standby. Standby includes the capability of I':ir.g able to be placed into operation from the main control room. Refer to SSAR section 9.2.1 for additional infonnation on he CCS. Both SWS pumps and cooling tower fans should be available during the N10 DES of applicability when the loss of RNS cooling is risk important. If both SWS pumps and cooling tower fans are not asailable. the plant should not enter these conditions. If the plant has entered these conditions then I the plant should take action to restore both SWS pumps / fans or to leave these conditions. If the 1 plant has not restoredfull system operation or left the MODES of applicability within 12 hours, then I at tions need to be initiated to increase the RCS water level to either 209 pressuri:er level or to a fitti i refueling cavity. Plarmed maintenance affecting this SWS cooling fur. tion should be perfonned in N10 DES when the SWS is not supporting RNS operation, ie during N'.JDES 1. 2. 3. The bases for this recommendation is that the SWS is more risk important during shutdown N10 DES. especially during the N10 DES of applicability constitions than during other N10 DES. Revision: 21 ??.199M 16.3 26 W W25tingh0USB w.semRd - R

p_. -

16. Technical Specificctions FA1 -

Table 163-2 (Cont.) INVESTMENT PROTECTION SilORT TERM AVAll.AHil.ITY CONTROI.S 3.0 Electrical Power Systems 3.2 AC Power Supplies - RCS Open OPERABILITY: Two AC power supplies should be operable to support RNS operation APPLICABILITY: MODE 5 with RCS pressure boundary open, MODE 6 with upper intemals in place or cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required AC power A.I Initiate actions to increase the 12 hours supply not operable, water inventory above the core. AND A.2 Remove plant from applicable 72 hours MODES B. Required Action and B.I Submit repon to [ chief nuclear iday associated Completion. officerl or [on-call attematel Time not met, detailing interim compensatory measures, cause for inoperability, and schedule for restoiation to - OPERABLE. AND B.2 Document in plant records the 1 month justincation for the actions taken to restore the function to OPERABLE. Resision: 21 [ Westinghouse 163-4 A ??,1998 %O.eC-t an -a

J g;l- !n-i.

16. Technical Specifications c,

I Table 163 2 (Cont.) INVESTMENT PROTECTION SHORT. TERM AVAILAllit.ITY CONTROL.S l SURVEILLANCE REQUIREMENTS SURVEILLANCE - FREQUENCY l SR 3.2.1 Verify that the required number of AC power supplies are Within I day prior to operable entering the MODES- ) of applicability f J l 1 a f 4 ? r i f e i l 1 I Resision: 21 ??, IWM y Wegi g ge 163-44 f qqo.scMr(RIN

si itil i E

16. Technic l Specific tisms f

Table 16.3 2 (Cont.) INVESTMENT PR(FIECTION SilORT TERM AVAll.Allit.lTY CONTROL.S 3.0 Elecincal Power Systems 3.2 AC Power Supplies - RCS Open flASES: AC power is required to power the RNS and its required support systems (CCS & SWS); the RNS pnwides a nonsafety related means to nonnally cool the RCS during shutdown operations. This RNS cooling function is important when the RCS pressure boundary is open and the refueling cavity is not thxxted because it reduces the probability of an initiating event due to loss of RNS cooling during these conditions and because it provides margin in the PR A sensitivity performed assuming no credit for nonsafety-related SSCs to mitigate at power and shutdown events. The RCS is considered open when its pressure boundary is not intact. The RCS is also considered open if there is no visible level in the pressuriier. The margin provided in the pRA study assumes a minimum availability of 90% for this function during the N10 DES of applicability, considering both maintenance unavailability and failures to operate. Two AC power supplies, one offsite and one onsite supply, should be available as follows: l a) Of fsite power through the transmission switchyant and either the main step-up transfonner / unit auxiliary transfonner or the reserve auxiliary transfonner supply from the transmission switchyard, and b) Onsite power from one of the two standby dicsci generators. Refer to SS AR section 8.3.1 for additional infonnation on the stamiby diesel generators. Refer to SS AR section 8.2 for infomiation on the offsite AC power supply. One offsite amt one onsite AC power supply should be available during the N10 DES of applicability when the loss of RNS cooling is important. If both of these AC power supplies are no: available, the plant should not enter these conditions. If the plant has already entered these conditions, then the I plant should take action to restore this AC power supply function or to leave these conditions. If the l plant has not restoredfull system operation or left the A10 DES of applicability within 12 hours. then i actions need to be initiated to increase the RCS water level to either 20% pressuri:er level or to a full l refueline < avity.. Planned maintenance should not be pertonned on required AC power supply SSCs. Planned maintenance atfecting the standby dicsci generators should be performed in N10 DES 1,2,3 when the RNS is not nonnally in operation. Planned maintenance of the other AC power supply should be perfonned in N10 DES 2,3. or N10DE 6 with the refueling cavity full. The bases for this Revision: 21 W Westinghouse 16.3-45 ?? I " 440,bcHFIR8-6

w ..w NRC FSER OPEN ITFM Question: 480.ll36F (OITS 6560) SSAR Section 9.4.7.2.1 states that the containment air filtration system supply and exhaust air lines penetrations are 36 inches in diameter with 16 inch branch lines containing containment isolation s v'.lves. SSAR Section 6.2.1.5.3 states that two 10-inch diameter flow paths were modeled in the m9:imum containment pressure analyses to simulate containment purge lines. Westinghouse should explain why the analyses did not model the actual plant design and why this analysis is conservative.

Response

Westinghouse has performed emergency core cooling analyses with minimum containment backpressure that reflects the AP600 plant design which includes 15 inch diameter (16 inch Schedule 40 pipe) containment purge supply and exhaust flow paths with purge isolation valves having a closure delay of 22 seconda follow.. g receipt of a high containment pressure signal. The results of the minimum containment pressure calculation are shown in Figure 480.!!36F 1 (SSAR Figure 6.2.1.51). He impact of the containment pressure transient shown in Figure 480.ll36F 1 on the large break LOCA ECCS performance of AP600 has been shown to be minor. The limiting C =0.8 DECLG o break with bounding initial and boundary conditions case presented in SSAR subsection 15.6.5.4A.3.3 was executed again with WCOBRAfrRAC, using the Figure 480.ll36F-1 containment pressure as l input. He AP600 large break LOCA transient exhibits its calculated peak cladding temperature (PCT) early in blowdown, while the break now is choked; the PCT is therefore independent of the containment pressure employed. A second cladding temperature peak occurs during the re0ood phase of the DECLG break. The temperature of this second peak increases by less than 20*F in the reanalyzed subsection 15.6.5.4A.3.3 case, so the renood phase peak cladding temperature value t remains well below the blowdown PCT. De AP600 95th percentile PCT limiting value of 1676*F shown in the SSAR Table 15.6.5 9 is unchanged due to the revised containment pressure transient. SSAR subsection 6.2.1.5.2 and 6.2.5.3 are marked up to reflect the current AP600 design which includes a 15-inch diameter (16-inch. Sch.40 pipe) containment purge supply now path and exhaust flow path and the current minimum containment pressure calculation. SSAR Revision: SSAR subsections 6.2.1.5.2 and 6.2.1.5.3 and Figure 6.2.1.5-1. gg 480.1136F-1

5 NRC FSER OPEN ITEM _ 40 o m

a. 3 5 30 N

25 x 20 N E c 's 15 c O " 10 ~ 0 60 120 180 240 300 Ilme (SeC) Figure 480.ll36F.I Predicted AP600 Containment Minimum Pressure Following DECL Breaks 480.1136F-2 T Westingtwuse

6. Engineered S:fety Fe:t:res in the analyses are based on input for the AP600 steam generator and ' main feed system design.

6.2.1.4.3.3 Containment Pressure. Temperature Results De results of the containment pressure-temperature analyses for the postulated secondary system pipe ruptures that produce the highest peak containment pressure and temperature are presented in subsection 6.2.1.1.3. 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of Emergency Core Cooling System (PWR) The containment backpressure used for the AP600 cold leg guillotine and split breaks for the cmergency core ccv> ling system (ECCS) analysis presented in subsection 15.6.5 is described. De minimum containment backpressure for emergency core cooling system performance during a loss-of-coolant accident is computed using the EGOTHIC computer code. Subsection 6.2.1.1 demonstrates that the AP600 containment pressurizes during large break LOCA events. An analysis is performed to establish a containment pressure boundary condition applied to the ECOBRA/ TRAC code (Reference 8). A single node containment model is used to assess containment pressure response. Containment internal heat sinks used heat transfer correlations of 4 times Tagami during the blowdown phase followed by 1.2 times Uchida for the post-blowdown phase. De calculated contenment backpressure provided in Figure 6.2,1.51. Results of the ECOBRA/ TRAC analyses demonstrate that ;he AP600 meets 10 CFR 50.46 requirements (Reference 7). 6.2.1.5.1 Mass and Energy Release Data The mass and energy releases to the containment during the blowdown portion only of the double-ended cold leg guillotine break (DECLG) transient are presented in Table 6.2.1.51, as computed by the ECOBRA/ TRAC code. The mathematical models which calculate the mass and energy releases to the containment are described in subsection 15.6.5. A break spectrum analysis is performed (see references in subsection 15.6.5) that considers various break sizes and Moody discharge coefficients for the double-ended cold leg guillotines and splits. Mixing of steam and accumulator water injected into the vessel reduces the available energy released to the containment vapor space, thereby minimizing calculated containment pressure. Note that the mass / energy releases during the reflood phase of the subject break are not considered. His produces a conervatively low containment pressure result for use as a boundary condition in the ECOBRA/ TRAC large break. LOCA analysis. 6.2.1.5.2 Initial Containment Internal Conditions b g\\tt.t d Initial containmerit conditions were biased for the emergency cere cooling sys backpressure analysis to predict a conservatively low containment backpressure. Initial containment conditions include an initial pressure of 14.7 psia, initial temperature of 90*F, and L J Revision: 20 [ W8stingh0US8 6.2 19 February 6,1998 Wo.it % D 3

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6. Engineered S:fety Fe:tures

) a relative humidity of 99 percent. An air annulus temperature of O'F is assumed and a linear temperature profile between O'F and 90'F is used in the containment shell, which separates i annulus frem the containment volume. 6.2.1,5.3 Other Parameters Containment parameters, such as containnie'nt ~ volume ar'd passive heat sinks, are biased predict a conservative low containment backpressure. Tir containment volume used in the geS k calculation is conservatively set to 1.05 times the cold volume. Passive heat sink surface areas were approximately doubled from the heat sinks, presented in Tables 6.2.1. 6.2.1.1-6. Material properties were biased high (density, conductivity, and heat capacity) as 7 indicated in CSB 61 (Reference 8). To further minimize containment pressure, containment, purge was assumed to be in operation at time zero. Two 10-inch diameter flow paths were provided in the containment model to simulate containment pun;e lines. Valves within these l Qines were closed 5 seconds after 8 psig was reached. 6.2.1.6 Testing and Inspection This section describes the functional testing of the containment vessel. Testing and in service inspection of the containment vessel are described in subsection 3.8.2.6. Isolation testing and leak testing are described in subsection 6.2.5. Testing and inspection are consistent with regulatory requirements and guidelines. De valves of the passive containment cooling system are stroke tested periodically. Subsection 6.2.2 provides a description of testing and inspection. The baffle between the containment vessel and the shield building is equipped with removable panels and clear ooservation panels to allow for inspection of the containment surface. See subsection 3.8.2 for the requirements for in service inspection of the steel containment vessel. Subsection 6.2.2 provides a description of testing and inspection to be perfonned. Testing is not required on any subcompartment vent or on the collectiori of condensation from the containment shell. The collection of condensate from the containment shell and its use in leakage detection are discussed in subsection 5.2.5. 6.2.1,7 Instrumentation Requirements Instrumentation is provided to monitor the conditions inside the containment and to actuate the appropriate engineered safety features, should those conditions exceed the predetermined levels. De instruments measure the containment pressure, containment atmosphere radioactivity, and containment hydrogen concentration. Instrumentation to monitor reactor coolant system leakage into containment is described in subsection 5.2.5. He containment pressure is measured by four independent pressure transmitters. De signals are fed into the engineered safety features actuation system, as descr; bed in subsection 7.3.1. Upon detection of high pressure inside the containment, the appropriate safety actuation Revision: 20 February 6,1998 6.2 20 3 WSStingh00S8 9Ec. // 3cP/

6.2.1.5.2 initial Containment Internal Conditions initial containment conditions were biat d for the emergency core cooling system backpressure analysis to predict a conservatively low containment backpressure. Initial containment conditions include an initial pressure of 14.7 psia, initial containment temperature of 90"F and a relatise humidi y of 99 percent. Ai. t air annulus temperature of O'F is a,sumed. De initi,-l through thickness metal temperature of the containment shell is assumed to also be O'F. 6.2.1.5.3 Other Parameters Ccatainment parameters, such as containment volt've and passive heat sinks, are biased to predict a conservative low containment backpressure. De containment volume used in the calculation is conservatively set to 1.1 times the free volume of the AP600 containment Evaluation Model. Passive heat sink surface areas were increased by a factor of 2.1 times the values presented in Reference 20. Material properties we:e biased high (density, conductivity, and heat capacity) as indicated in CSB 61 (Reference 8). No air gap was modeled between the steel liner and base concrete of jacketed concrete heat sinks. De outside surface of the containment shell was maintained at O'F throughout the calculation. To further minimire containment pressure, containment purge was assumed to be in operation at time z.ero and air is vented through both the 15. inch ditweter (16 inch, Sch.40 piping) containment purge supply and exhaust lines until the isolatiori valves I a.ve fully closed. These valves were modeled to close 22 seconds after the 8 psig closure setpoint was reached. = //f-(, // 3 t.. F 'l

6. Engineered S:fety Features Ar A

y %Y 'b Y W 45 \\ /J m J o ~ 40 ~ sn a v 35 k E 30 K 25 E c 20 o C 15 o C) 10 i 0 50 100 150 200 200 300 Time (sec) l l Figure 6.2.1.51 EGOTHIC Predicted AP600 Containment Response to DECL Breaks Revision: 13 May 30,1997 6.2 158 Westinghouse M c.i/30 F -G.

) Figure 6 2 1 5-1 Predicled AP600 Contoinment Minimum Pressure Following DECL Breaks _ 40 O [ a 35 E 23 N l 2 N 25 .c 1 x ~ Y m 20 ~ g i C o 15 C O " 10 O 60 120 180 240 300 h_ Time (Sec) I k h ~ i

RESPnNSES TO NRC REQUEST FOR ADDITIONAL INFORMATION Question 640.127 (OITS 5881) l l Re: Section 33.7 Normal Residual llent Remosal System The following design features addressing mid loop operation should be senfied in the ITAAC: l (a) Loop piping effect, reactor coolant system hot legs and cold legs are vertically offset. (b) Step norile connection, system employ s a step.norile connection to the reactor coolant system hot leg. (c) Self senting suction line, pump suction line is sloped continuously upward from the pump to the re *or coolant system hot leg with no local high points. Response (ResIsion 2): Based on discussions with the NRC at the ITAAC Task Group meetings held on Nosember 24 & 25, 1997 ITAAC Table 2 3 6 4 has been revised as shown to address this RAI. I Resision 2 of this response is provided based on feedback at the ITAAC Task Group mecting held on l January 15, 1998. The acceptance criteria that the norile inside diametei be 16.8 inches is based on the I design of the step noule presented in the SSAR that specifies that the pipe sire for the step nozzle is 20-l inch, Schedule 160. The acceptance entena for the step nozzle is revised to specify that the step.nonle I be constructed of 20 inch, Schedule 160 pipe. The pipe sire and schedule is found on SSAR Figure 51-5 l and proudes the basis for the ITAAC acceptance enteria-SSAR Resision: None m.1271 W Westinghouse Revision 2

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION ITAAC Resisiona: Table 2 3.6 4 9 b) The RNS prosides heat

) Inspection will be i) A report esists and remosal from the reactor performed for the existence concludes that the product vf coolant during shutdown of a report that determines the oserall heat transfer operations.

the heat remos al capability of coefficient and the effectise the RNS her.t exchangers heat transfer area. UA, of each RNS heat exchanger is greater than or equal to 2 0 milhon B tu/h r.'F. ii) Testing will be performed ii) Each RNS pump prosides to confirm that the RNS can at least 900 gpm net flow to the l provide flow through the RCS when the hot leg water l RNS heat exchangers when levelis at an eteration 15.5 l the pump suction is aligned inches i 2 inches abore the I to the RCS hot leg and the bottom of the hot /cg. discharge is aligned to both PXS DVI lines. l l lii) Inspectwn will be lis,t The reactor coolant syste n i performed of the reactor cold legs piping centerline is coolant loop piping. I7.5 inches t 2 inches above the hot legs piping.rentse ne, n j iv) Inspection will be ivl The RNS pump suction t i performed of the RNS pump pipingfrom the hot leg to the i suction piping. pump suction piping low point i does notform a local high point (defined as an upward slope with a vertical rise greater than 3 inche.;). v) Inspection will be v) Ike-neul+4nside liameter performed of the RNS pump inwo-lentkan-lM-indesc-suction noule connection to The RNS suction line the RCS hot leg. connected to the RCS is I constructedfrom 20 inch Schedule 160 pipe. 640.127 2 [ W85tingh00$8 Revision 2 l

RESPONSES TO NRC REQUEST FOR ADDITl0NAL INFORMATION 9 c) The RNS piosides low Testing w ' be perfo*med to Each RNS pump proudes at pressure makeup flow from confirm ti.ut the RNS can least 92$ gpm net flow to the l the IRWST to the RCS for provide low pressure makeup RCS when the water /crci l scenarios following actuation flow from the IRWST to the above the bottom of the l of the ADS. RCS when the pump suction /RirSTis 4fcci 1/2 inches. is aligned to the IRWST and the discharge is aligned to both PXS DVI lines with RCS at atmospheric pressure. 640.127 3 3 Westinghouse Revision 2

NRC FSER OPEN ITEM FSER Open item 640.',76 (OITS #6609) The Tier 1.1 formation for the passive containment cooling system (PCS) must include a design commitment and ITAAC for the coatings that will be applied to the internal and external surfaces of the containment vessel, including the surfaces of structures and equipment inside of containment, as necessary, The acceptance criteria for the PCS ITAAC must include the thickness of the coatings and i Other materialinside of containment and thi transport analysis must be included in the AP600 SSAR. i Westinghouse Response: i i Th9 AP600 SSAR (subsection 6,1.2), as contained in our response to OITS 6590 revision 1, specifies that the coatings apolled to the containn ent shollinside containment (greater than 7' above operating deck) and outside containment (above the operating deck) are classihed as safety-related. The SSAR also specifies tha applicability of 10 CFR 50, Appendix B, to these coatings, including the necessary inspections. Refer to the response to OITS #6590, revision 1, for discussion of the transport analysis and the proposed ITAAC associated with it. SSAR Revisions: None ITAAC Revisions: None ( #estingh0use

NRC FSER OPEN GTEM Question 640,180F (OITS 6613) In sour response to RAI 440.742F, you committed to type testing of the pressuriier safety salses (refer to SSAR section 5.4.9). Verincation of this test must be included in ITAAC The ITA AC should specify the range of temperature and pressure for the type test and a minimum now rate through the salses.

Response

Westinghouse has provided the requested ITAAC committment in Revision I to the Response to RAI 440.742F. SSAR Resition: Wone 640.180F 1 W Westingh0ese

NRC FSER OPEN ITEM Ouestion: 720.439F (OITS #6177) [of NRC letter dated November 19,1997] contains the staff's insi hts as a result of the review of the F Lesel 2 PRA. Enclosure 2 contains additional insights frorn those contained in Chapter 59 of Westinghouse's PRA. Incorporation of the additional insights that exist in Enclosure 2 to the Westinghouse insights is an open item. In addition, the staff belieses that certain insights are so important that they need to be incorporated into the Technical Specifications or into the inspections, tests, analyses, and acceptance criteria (ITAAC). In the cases where the staff believes the disposition of the insight should be Technical Speci0 cations or ITAAC a separate FSER open item number has been astigned.

Response

Westinghouse has reviewed the staffs insights presented in Enclosure 2 of NRC's November 19,1997 letter for technical accuracy and appropriateness. Attached is the Westinghouse feedback on the staff insights. As noted in the staff's question above, several items were provided with a separate FSER open item iumber. Those items are not being addressed by this FSER open item response, because they are addressed by their assigned FSEP open item number. Changes to the PRA report insights table provided in Chapter 59 are described below. PRA Revision: Revision 2 The R AP section in the SSAR has been moved to section 17.4, per request of NRC. As a result, when "SSAR 16.2" is e amed in the disposition column of PRA Table 59 29, it will be changed to "SSAR 17.4" The following changes will be made to PRA Table 59 29: Under item Id (IRWST): The operator action to dood the reactor cavity is determined in Emergency Response Guideline I'R C.1, which instructs the operator to Ocod the reactor cavity if injection to the RCS cannot be recovered or containment radiation reaches a lev-l that indicates fission product releases as determined by a core damage assessment guideline. [ disposition = Emergency Response Guidelines] Under item 26: + The redecuve insulation panels and support members can withstand pressure differential loading due to the IVR boiling phenomena. RPV/ insulation panel clearances, water entrance and steam eut flow areas, and loss coefficients are based on scale tgst data from the ULPU facility. Water inlets aad steam vents are provided at the entrance and eut of the insulation boundary. Reactor vessel insulation is an important SSC. The COL will maintain the reliability of the insulation. (disposition = SSAR 17.4] 720.439F(R2) 1

NRC FSER OPEN ITEM + Under item 30: Operability of the hydrogen igniters is addressed by short term availability controls during modes I,2,5 (with RCS pressure boundary open), and 6 (with upper internals in place and cavity lesel less than full). [6sposition = SSAR 16.3) Under item 36: The COL will develop and implement severe accident management guidance for operation of the nonsafety. related containment spray system using the suggested framework provided in WCAP 13914. (disposition = PRA Chapter 59 (subsection 59.10.6)] Under item 39: e l Operability of DAS for selected containment isolation actuations is addresstJ by short term availability l controls. [ disposition = SSAR 16.3) Under item 41: ne reactor cavity design incorporates features that extend the time to basemat melt through in the event of RPV failure. The cavity design includes: A minimum floor area of 48 m' available for spreading of the molten core debris, A minimum thickness of concrete above the embedded containment liner of 0,85 m, l 1 There is no piping buried in the concrete beneath the reactor cavity; sump drain lines are not i enclosed in either of the reactor cavity floor or reactor cavity sump concrete. Thus, there is no I direct pathway from the reactor cavity to outside the containment in the event of core concrete Interactions. The openings between the reactor cevity and cavity sump are small diameter openings in which core debris in the casity will solidify. Thus, there is no direct pathway for core debris to enter the sump except in the case where it might spill over the sump curbing.

hete-eee-*o-interconnectenp-1 p:! ::; ::nbedded-in-the-eemerete-new-eneer-wmp-evebetheveby 4

peeventenplebew4 rem-pewapn:e he ;;mp

.;mp rueb44Seffici

" 'righ: :,J UJ:S :ptevent r^:: are41ebrevirom4 veeGoweep+eblaung theowt -thewh [ disposition = PRA Append.x B) h New item (#43L Capabihty exists to vent the containment via the RNS suction lines to the spent fuel pool, with the RCS depressurized and opea to the containment atmosphere via cither the ADS or the sessel failure. [ disposition

= PRA Appendit D] i l The COL will develop and implement severe accident management guidance for senung containment using the suFgested framework provided in WCAP 13914. idisposition = PRA Chapter 59 (subsection 59.10.6)] NRC Follow-on Question (for response revision 1): Dunng a telecon on February 3,1998, the staff (R. Palla) requested that the following items be added or modified to the PRA insights table (Table 59 29): 720.439F(R2) 2

NRC FSER OPEN ITEM 1. Containment spray modify insight #36 to state APu00 has a nonsafety related containment spray system, and direct the disposition to the SSAR. The rest of what was written in #36 was acceptable. 2. Reactor vessel insulation need disposition to be included in the FSER 01 response under *PRA Revision" discussion. 3. Protection from diffusion names (related to 720.444F, which requested an ITAAC) Some diffusion flame vent information was added to SSAR and ITAAC, please include disposition in PRA insights table. Response to follow-on question: 1. The following sentence will be added to insight #36: AP600 has a nonsafety related containment tpray system. [.skposition = SSAR subsection 6.5.2]. 2. For insight #26, "SSAR 5.3.5, and Certified Desigr. Material" will be added to the dispesition column. Some of the infonnation is within the SSAR and some is within ITAAC, thus the reason to mention both within the disposition column. 3. Per the revised response to 720.444F (W letter DCP/NRCl240, dated I/30/98), the third item of insight #31 is now covered by SSAR 6.2.4.5 and Certified Design Material (includes ITAAC). He disposition column in the insights table will be noted as " Certified Design Material." The ch.inges noted above for this follow on question are illustrated on the Table 720.439F l. i NRC Follow on Question (for response revision 2): i The NRC issued a letter to Westinghouse dated February 19,1998. Item 5 e'the enclosure requests Westinghouse modify Lesel 2 insight number 41 regarding embedded piping in the concrete batemat. As stated in the staff's letter: "The concern is twofold i.e., there is no embedded piping that can provide a pathway (1).nto the sump (already included by WEC), i.nd t2) from the sump to the outside of containment (not presently included by WEC)." Response to followon question: Item 41 of Table 59 29 will be revised as shown by the "PRA Revision" section of this FSER open item *esponse. The changes address the staffs concerns. 720.w2N W Westinghouse 1

NRC FSER OPEN ITEM I Table 720.439F.I Modincations to AP600 PRA. based insights INSIGIIT l DISPOSITION l

26. The retlectne reactor sessel insulation provides an engineered now path to allow PRA Chapter 39, the ingression of water and venting of steam for externally cooling the sessel in the SSAR 5.J.5, and event of a severe accident involving core relocation to the lower plenum.

Cerryicd Design hiaterial l The reDe:tive insulation panels and support members can withstand pressure differential loading due to the IVR boiling phenomena. l l PRVhnsulation panel clearances, water entrance and steam exit now areas, and loss I coef0cients are based on scale test data from the ULPU facility. l Water inlets and steam vents are provided at the entrance and exit of the insulation I boundary. I l No coatings are applied to the outside surface of the reactor vessel which will l inhibit the wettal,ility of the surface. Re. ctor sessel insulation is an important SSC. The COL will maintain the SSAR 17.4 reliability of the insulation.

31. The containment layout presents the formation of diffusion names that can SSAR I.2, challenge the integrity of the containment shell.

General Arrangement Vents from compartments where hydrogen releases can be postulated area away Drawings from the containment wall and pene' ations or are hatched and locked closed. IRWST sents near the containment wall are turned to direct releases away from the Certyicd Design containment shell. Afartrial

36. AP600 has a nonsafety related containment spray system.

SSAR 6.5.2 Containment spray is not credited in the PRA. Failure of the nonsafety related PRA Chapter 43 containment spray does not prevent the plant schieving the safety goals. The COL will develop and implement sescre acident management r. dance for PRA Chapter 59 operation of the nonsafety related containment spray system using the suggested (59.10.6) framework provided in WCAP-13914. 720.439F(R2) 4 g,g

NRC FSER OPEN ITEM t NRC Staff insights of the AP600 Level 2 & 3 PRA and Westinghouse Feedback hur Containment Coolitm System The first item of NRC's November 19,1997 Enclosure 2 was FSER open item 720.440F, which pertains to the passive containment cooling system. De response to 720.440F was provided in Westinghouse letter DCP/NRCI 179, dated December 11.199L There is no change to the PRA insights table presented in PRA Chapter 59 as a rewit of this FSER open item. De staffs wording is covered by item 37 of PRA Table 59-20 Ructor Casity Floodine System A safety related reactor cavity flooding system is included in the AP600 design to prevent reactor vessel breach and es sessel phenomena in the event of a severe accident. He system is comprised of the following design features: ' vo 6-i, ch di'. meter recirculation lines that provide a path for gravity draining the IRWST to the reactor cavity, 4 a squib valve and a motor operated valve in each recirculation line, each powered from the Class IE de power supply, and actuated from the control room. tnd a reactor vessel thermal insulation system designed specifically to enhance RPV cooling, as described in FSER Section 19.2.3.3.1. E' Respcmse: The information provided in the stafs two bullets (above) are covered by item Id of PRA Table 59 29. The reactor vessel insulation is described by item 26 of PRA Table 59 29. The IRWST injection squib valves are diserse from the containment ecirculation squib valves. [ Diversity between these valses is specified in SSAR Section 6.3.2.2.8.9, but the criteria for confirming that diversity has been ach;eved is not provided. His needs to be addressed by ITAAC. This is Open item 720A41F.1 W Response: The stafsfirst sentence is cosered by item Id of PRA Table 39 29. The stafs statement which has been placed in braclets is addressed in the response to FSER open item 720.44ff. The response was provided by Westinghouse letter DCP/NRCil68, dated December 4,1997. The containmant recirculation squib valves and isolation MOVs. and containment recirculatio.i screens are included as risk significant SSCs within D-RAP. E' Resp <mse: The information providc1in the stafs statement is covered by item Id of PRA Table 59-29, Note the PRA inssghts table states the reliability of the IRWST subsystem is important, and then directs the reader to SSAR 16 2 (now 17.4) which is the section on RAP, 720.439F(R2) S

NRC FSER OPEN ITEM A Surveillance and maintenance requirements on the related piping and valves are provided in the In Service inspection and Testing Programs. W Response: The information on va.'ves is specified in ite.n Id of PRA Table 39 29, and directs the reader sia the disposition column to SSAR subsection 3.9.6. nhere ISTprogram is discussed. The operator acton to Hood the rea. tor cavity is provided in Emergency Response Guideline IR.C 1, w hich instructs the operator to Good the reactor cavity if injection to the RCS cannot be recovered or containment radiation reaches levels that indicate fission product releases as determined by a core damage assessment guideline. W Response: This is a correct statement. An insight will be added to item Id of PRA Table 39-29. [ Key aspects of the reactor cavity Hooding system and the containment layout need to be confirmed by ITAAC to assure that the teactor cavity will flood and the RPV will renood rs modelled in the PRA (by gravity draining and by manual actuation of the cavity Gooding system). The ITAAC should include confirmation of internal volumes, elevations, and inter compartment vent and drain paths of the subcompartments containing RCS piping components and impacting reactor cavity Gooding and RPV renooding. WEC needs to provide this ITAAC. This is open item 120.442F. The response to FSER open item 720.442F was provided by Westinghouse letter DCP/NRCll80, dated December i2. I997.) RPY Thermal Insulation System The AP600 design includes a tenectise reactor vessel insulation system that provides an enE neered now path to i allow the ingression of water and senting of steam for externally cooling the vesselin the event of a severe accident involving core relocation to the lower plenum. W Response: This.'s a corr ct statems..t and is consistent with stem 26 of PRA Table 39 29. Key attnbutes of the insulation system are: RPV/ insulation panel clearances, water entrance and steam exit now areas, and loss coefficients based on scale tests in the ULPU facility, ball and cage check valves and steam sent dampers at the entrance and exit of the insulation boundary that open due to buoyant forces during cavity Good up, and insulation panels and support members designed to withstand the pressure differential loading due to the IVR bmling phenomena. W Resp <mse: The first and third bullets above are accurate statements. The second bullet should be reworded asfollows to be consiste 2r with design information provided in the SSAR: " water inlets and steam sents are provided at the entrance and exit of the insulation boundary that open due to buoyant forces during cavity flood-up." The information provided in the staff's statements will be added to item 26 of PRA Table 39 29 with the changes noted. 720.439F(R2b6

NRC FSEP OPEN ITEM 1 No coatings are applied to the out ide surface of the reactor vessel which will inhibit the wettability of the surface. W Respon?e: This is a correct statement and is consistem with stem 26 of PRA Table 39 29. [The reactor vessel insulation system should be included as a risk significant SSC in the reliability assurance program, and reliabihty/ availability controls and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained. ITAAC and availability controls are also needed to assure that the RPV insulation system will perform as designed. WEC needs to provide these coramitments and ITAAC. This is open item 720.443F. The response to FSER open item 720.443F was provided by Westinghouse letter DCP/NRCil80, dated December 12, 1997. Note per the responn to FSER open item 720.443F, the reactor vessel insulation is included as a risk.signtficant SSC in the RAP. The tie to RAP will be provided as an insight with item 26 of PRA Table 39 29.] Protection of Galaingient From h ffusion Samu i Th containment layout prevents the formation of diffusion Games that can challenge the integrity of the containment shch. Specifically: the openings from the accumulator rooms and CVS compartments that can vent hydrosen to the CMT room are either located away from the containment wall and electrical penetration junction boses, or are covered by a sect re hatch, and IRWST sents near the containment wall are osiented to direct releases away from the containment shell. W Response: The above utformatwn is covered by items 31 and 38 of PRA Table 39 29. [These provisions need to be confirmed by ITAAC. WEC has not provided this ITAAC. This is Open item 120 444F. The revised response to FSER open item 720.444F was provided by Westinghouse fetter DCPINRCl240, dated January 30.1998. Per the revised response, the third item of insight #31 will have ' Certified Design Material" meluded in the disposition.] Operatioti af ADS stage 4 pravides a vent path for the severe accident hydrogen to the steam generator compartments, bypassing the IRWST, and mitigating the conditions required to produce a diffusion Hame near the containment wall. W Resp <mse: The above information is covered by item 38 of PRA Table 39-29. "4 W wealnghouse 72a e 2N

NRC FSER OPEN ITEM { !!!MWi mm-Containment Isolation System Containment isolation valves in lines that represent risk significant release paths are u.itrolled by DAS in eddition to PMS to further limit offsite releases following core melt accidents. These lines are: cont inment air filter supply and exhaust, RCDT out, and normal containment sump. The containment isolation valves controlled by D AS are included as risk significant SSCs within D RAP. The operability of DAS actuation of these isolation valves is address:d by short term availability controls for DAS. i W Respcnse: The above information is covered by item 39 of PRA Table $9 29. A sentence will be adde 2 to item 39 to address the availability controls on this DASfunction. 4 Reactor Cavity Desien for Direct Containmtat Heatinn ) The res tor cavity and RPV arrangement provides no direct now path for the transport of particulated molten debris from the reactor cavity to the upper containment regioes. l W Response: The above information is covered by item 29 of PRA Table 59 29. ) Reador Cavity Desien for Ex Vessel Fuel Coolant Interactions The design can withstand a best estimate ex vessel steam explosion without loss of containment integnty. f i W Response: The above information is covered by item 28 of PRA Table 59 29. Eractor Cavity Design for Core Concrete Interactions The reactor cavity design incorporates features that protect against basemat mell through in the event of RPV failure. The cavity design includes: 8 a minimum Door area of 48 m available for spreading of the molten core debris, 3 l layout, elevations, and now areas of the reactor cavity and RCDT subcompartments and interconnectmg sentilation duct consistent with Figure B 3 of Appendix B of the PRA and the supporting ANL analysis, a minimum thickness of concrete above the embedded containment liner of 0.85 m, provisions to prevent core debris from passing into the sump via interconnecting pipelines embedded in the concrete Door and/or sump curb, and a sump curb of sufficient height and width to prevent molten core debris from over0owing or ablating through the curb. (WEC still needs to confirm these items as part of Open item 19.2.3.3.3 1 (Open item 720.418F). Note the response to FSER open item 720.418F was provided by Westinghouse letter DCP/NRCI171, dated December 9,1997.) 720.439F(R2) 8

NRC FSER OPEN ITEM W Response: It is not technically accurate to say the reactor cavity design incorporates features that protect against basemat melt through, but rather the plant is designedfor m sessel retention of molten core debris and the reactor cavity incorporates features that extend the time to basemat melt-through on the event of s esselfailure. Thefirst. third. andfifth sub bullet above are accurate. The fourth subbullet is accurate if scritten as "there are no interccmnecting pipelines embedded in the concretefloor and/or sump curb. thereby pres enting debrisfrom passing into the sump. " Theflrst, third fourth, andfifth sub bullets identyr charact< ristics of the reactor cavity design; u hereas the second sub bullet conta.ns secondary irformation and is written too broadly, thus Westinghouse does not consider that it is an insight. Item 42 of PRA Table 39-29 aillinclude the appropriate infermation per this Westinghouse response. A specific type of concrete is not specified for use in the basemat. W Response: The above information is covered by item 41 a'?RA Table 39 29. Ib dwattLigmtttP, item De AP609 design includes a hydrogen igniter system to limit the concentration of hydrogen in the contamment during severe accidents. R-features of the system are: 66 glow plug igniters distributed throughout the containment powered from the non safety related onsite ac power system, but also capable of being powered by offsite ac power, ensite non essential diesel generators, or non Class IE batteries via de to ac inverters. manually actuated from the control room when core exit temperatu.e exceeds 1200F, as the first step in ERG FR.C 1 to ensure that the igniter activation occurs prior to rapid cladding oxidation. The igniter system is non safety related but is subject to insestment protection short term availabilny controls. The AP600 design also includes four passise autocatalytic recombiners (PARS) strategically located within the containment. De PARS are provided primarily to cope with hydtcgen production during design basis accidents, but are expected to function to reduce combustible Eas concentrations during severe accidents. W Response: Aluch of the above information is covered by item 30 of PRA Table 39 29. As part ofitem 30. the igniters and PARS are introduced. and then the reader is directed to the Certified liesign Afaterial forfurther design information fi e., numbn ofigniers, etcl The operation of the igniters via the ERGS is also presented in item 30. A -.ement a tti be added to item 30 to cover the igniter short-term availability control ahich is. sered by SSAR section 16.3. 720.439F(R2) 9 W Westinghouse

NRC FSER OPEN ITEM A Nonaately Containment Spray A non safety grade containment spray system is included in the AP600 design with the sapability to supply water to the containment spray header from an esternal source in the event of a severe accident. Loss of ac power does not contribute significantly to the core damage frequency, therefore, non safety related containment spray does not need to be ac independent. The spray system comprises the following design features: two containment spray ring headers equipped with a total of 66 spray nonles and providi.ng approsimately 80 percent containment coserage. a 6 inch diameter supply pipe connecting the spiny ring-headers to the fire protection system header inside containment, containing one normally closed, air operated valve with remote actuation from the control room (5701), and one normally open, manual valve (V700), 6-inch diameter piping connecting the fire header inside containment to the fire main header outside containment, and capable of being supplied from both the diesel driven fire putap and the motor-driven fire pump. De detailed design and location of all associated valves and connections will take into ac ount espected radiation levels and shielding requirements for any required local operator actions. De COL applicant will develop and implement guidance and procedures for use of the non safety containment spray system as part of the COL Action item regarding accident management program. h' Rest onse: As stated in item 36 of PRA Table $9 29, "Contai'iment spray is not credited in the PRA. Failure of the nonsafety related containment spray does not prevent the plant achieving the safety goals." Since the wnsafety related containment spray system is not credited in the PRA, it is not appropriate to provide the staff's above design configuration statements in the PRA insights. Hourver, a statement sill be added to item 36 regarding the COL development of SAAfG using the suggested framework provided in WCAP 139/4. (Note WCAP 139/4 is being revised to incorporate this aspect.] Cimtainment V_tn! The following will be completed after W submits design descciption. This is identified as an Open item in FSER Section 19.2.5 (Open Item 720.421F). In the esent of a severe accident that results in gradual containment pressurization, the AP600 containment can be sented via the _ line to present over pressure failure. I:isuon product reicases from the,__ line are touted to the stack. Valves in the _ line are qualified to operate at containment pressures corresponding to Service Level C. The __ kne is capable of withstanding the pressures associated with vent actuation at Service C. 720,439F(R2)-10

NRC FSER OPEN ITEM E Detaded procedures for use of the containment vent system will be developed by the COL applicant, as part of the COL Action item regardinF accident management. This section contingent on WEC's response to staff request concerning providing a sent pursuant to 10 CFR $0.34(O(3)(iv). }E Response: The responte to FSER open item 720.421F was provided by Westinghouse Irtter DCP/NRCil94, dated December i8,1997. Based on the response to FSER open item 720.421F, an statement well be added to Table 39 29 to address the containment vem issue. The statement will be as follows: Capability esists to vent the containment via the RNS suction lines to the spentfuelpool, with the RCS depressurized and open to the containment atmosphere via either the ADS or the vesselfailure. (disposstion a Appendiz D]. In addition, a statement well also be added to Table $9 29 to address the COL develo,> ment of SAMGfor venting containment using the suggestedframework provided in WCAP 139/4. Accident Management The COL will develop and implement severe accident management guidance and procedures using t$e framework provided in WCAP 13914, Revision 2. }Y Response: The abose information is covered by a COL item described in PRA subsection 39.10.6. 720.439F(R2) 11

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION Revision 3 ijiE E I NRC Letter Dated 10G/97 Question (6) It is not clear how the minimum availability value relates to the ignitor system operability. For instance,is the ignitt r system considered unavailable if one or more ignitors are inoperable or can multiple ignitors be inoperable as long as 751 of them are available? To avoid confusion in this area Westinghouse sl ould provide availability goals for the power supplies, each group of ignitors. and each coverage ione. De staf f eywcts availability goals for the ignitor system to be consistent with ignitor systems in operating plants. Ahhough ignitors in operating plants were installed to meet dif ferent regulations, both ignitor systems, APMW and operating plants, are designed to promote hydrogen burning soon af ter the lower llammability limit is reached. The stall estiinates the availc.bility of ignitor systems at operating plants to tw greater than 901.

Response

Eacit of the area's of the containment listed in Table 2.8 l contain redundant sets of ignitors. An additional colum1 has been added to this table to define the minimum number of ignitors that should be available. De minimum number of ignitors for a given area is N 1 as long as that leaves at least one ignitor trom each group m each area.11 'here are 2 ignitors in a given area (I from each group), then both should be available. !! there are 4 ignitors in an area (2 from each group), then 3 out of 4 should be available, in the upper containment I area, ahme the lua rr tragion, one half of the ignitors should be available; the N I approach is related in this i area because it is relatively open, n ell mixed and ignitors lot ated in this area are dtJJicidt to at cess for l maintenance l De availability goal has been claintied that it applies to the minimum number of ignitor.; necessun for sun es3 I in the l'RA and the value has been increased to 901. Refer to the response to question 2. SN Al Change: Reuse short tenn asailability controls 2.8 in SSAR section 16.3. Il' A AC Change: None NRC 10'2/97(Rf) /l E =m.

N R

16. Technical Specifications

.s l Table 16.3 2 (Cont.) J INVESTMENT PC .l TION SHORT. TERM AVAILAHILITY CONTROLS i 2.0 Plant Systems 2.8 flydrogen Ignitors OPERABILITY: The hydrogen ignitors should be operable in accordance with Table 2.N 1 APPLICAlllLITY: M ODE 1, 2, MODE 5 with RCS pressure boundary open, MODE 6 with upper internals in place or cavity level less than full ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.I Notify [ chief nuclear officer] or 72 hours hydrogen ignitor lon-call attemate), inoperable. AND A.2 Restore required ignitors to 14 days operable status. B. Required Action and B.I Submil report to [ chief nuclear I day associated Completion ollicer) or [on-call alternate] q Time of Condition A not detailing interim compensatory m et. measures, cause for inoperability,- and schedule for restoration to OPERABLE. AND B.2 Document in plant records the i month justification for the actions taken to restore the function to OPERABLE. Resision: 21 ?? l'M 16.3 36 y Westinghouse u n,q @ ( & z.

y! ![

16. Technic:1 Specific:tions i

Table 16.3 2 (Cont.) INVESTMENT l'ItOTECTION SilOltT.'IERN1 AVAI.AlllLIT' CONTitol.S SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 2.8.1 Energi/c cach required hydrogen ignitor arul verify the surf ace Eacli relucting outage temperature is > [17(10] F. Resision: 21 [ Westiligh0USS Ih.3 37 77' I998 N&c. io/ t /e il4 b - 3 t

j {i: U

16. Technical Specifications 1

Table 1h.3 2 (Cont.) INVESTMENT l'ROTECTION SilORT TERM AVAILAlHl.lTY CONTifOLS Table 2.N 1, ll3 rogen Ignitors d 1.ocatiori lltdrogen Ignitors Number (iroup I (iroup 2 Available (1) Reactor Cavity note 2 note 2 na Loop Companment 01 12,13 11,14 3 of 4 Loop Companment 02 5,8 6,7 3 of 4 l'ressuriier Companment 49,60 50,59 3 of 4 Tunnel connecting Loop 1,3,31 2,4,30 5 of 6 Companments Southeast Valve Room & 21 20 2 of 2 Southeast Accumulator Room East Valve Room. Nonheast 18 17,19 note 3 Accumulator Room, & Nonheast Valve Room - Nonh CVS Equipment Room 34 33 2 of 2 Lower Companment Area 22,27,28,29,31, 23,24,25,26,30 10 of i1 (CMT and Valve Area) 32 IRWST 9,35,37 10,36,38 5 ol 6 IRWST inlet 16 15 2 of 2 Refueling Cavity $5,58 56,57 3 of 4 Upper Companment I Lov cr Region 39,42,43,44,47 40,41,45,46,47 9 of 10 Mid Region 51,54 52,53 2 of 4 Upper Region 61,63 62,64 2 of 4 Notes:

1) In each location, the minimum number of ignitors that should be available are delined in this colum!).

2) Ignitors in this location are shared with other locations,

3) Ignitor 18 and either 17 or 19 should be available, 3

f Resision: 21 77.199N lh.3 3M T Westinghouse y,u.o/zhi/ 0) '/

16. Technical Specifications Table 16.3 2 (Cont.)

INVESTh1ENT PROTECTION SilORT. TERN 1 AVAll.AlllLITY CONTitOI.S 2.0 Plant Systems 2.x Ilydrogen Ignitors liASES: 1he l'ydrogen ignitors should be available to provide the capability of buming hydrogen generated during severe accidents in order to prevent failure of the containment due to hydnigen detonation. These hydrogen ignitors are required by 10 CFR 50.34 to IPuit the buihtup of hydrogen to less than 107, assuming that 1(XF7, of the active rircaloy fuel cladding is oxidized 1his function is also imponant because it provides margin in the PRA sensitivity perfonned assuming no credit for nonsafety related SSCs to mitigate at power and shutdown events. The margin provided in the PRA study assumes a minimum availability of 907c for this function during the N10 DES of applicability, considering toth maintenance unavailability and failures to operate. The ignitors are distributed in the contairunent to limit the buildup of hydrogen in local areas. Two groups of ignitors are provided in each area; one of which is sufficient to limit the buildup of hydrogen. When an ignitor is energlied, the ignitor surface heats up to 2[1700l'F. This temperature is sullicient to ignite hydrogen in the vicinity of the ignitor when the lower llammability limit is reached. SSAR section 6.2.4 provides additional infonnation. The hydrogen ignitor function should be available dunng h10 DES I and 2 when core decay heat is high and during N10DE 5 when the RCS pressure boundary is open and in N10DE 6 when the . ling cavity is not full. Piarmed maintenance should be perfonned on hydrogen ignitors when they I are not required to meet ',his availability control. T.Ible 2.S / indicates the minimum number of I hulrocen ienitors that should be available. 4 Resision: 21 W Westinghouse 16.3 39 ??*1998 m wi,wn s

l 1 NRC FSER OPEN ITEM Revision 1 f@ l NRC Fax Dated 1/27/98 (OITS #6590) Staff Comment: Westinghouse has not provided a quantitative or experimental bases which demonstrate why failure of the containment Coatings will r)t prevent functioning of the engineering safety features. Calculations are required to determine if failure of ine coatings wi!I result in blockage of strainers. The analysis must be reviewed by the NRC staff since there is uncertainty in the calculations. Even if Westinghouse were to perform a transport analysis, the staff is not likely to accept it without additional experimental validation.

Response

l The response to this open item is in four parts: Part 1 A commitment is added to the SSAR to c!assify the coatings used on the inside and the outside of containment shell above the operating deck as safety related. Information is added to the SSAR to define tne quality assurance provisions (under 10 CFR 50, Appendix B) that are applied to these coatings. l Part 2 - A commitment is added to the SSAR to procure the majority of the nonsafety related coatings used inside com ; nment to 10 CFR 50, Appendix B quality assurance requirements. Information is added to the SSAR to define the quality assurance provisions (under 10 CFR 50. Appendix B) that are applied to th6se coatings. Part 3 Results of calculations are presented which demonstrate that failure of nonsafety related coatings inside containment does not block any portion of the AP600 containment recirculation screens { Part 4 - An ITAAC is added to ensure that the majority of the as procured nonsafety related coatings j used inside containment are consistent with the Coating debris transport Calculation I discussed in Part 3. Part 1: } The Coating on the outs!de of the containment shell above the operating deck elevation supports I passive cooling of the containment. The coating effectively conducts heat and promotes the establishment of a water film which enhances water evaporation. The coating on the ;nside of the containment shew above the operating deck supports passive cooling of the containment. The coating effectively conducts heat. Formation of a water film is not required on the inside surface, nowever testing and analys!s have been conducted with a coating that promotes the condensate to form a film, [ Westinghouse

NRC FSER OPEN ITEM Revision 1 mi tin i 9 I The coating used on both of these surfaces is an inorganic zinc coating which is safety related in conformance with the intent of Regulatory Guide 1.54. A new SSAR Table (6.12) is attached to this response which identifies the quality assurance provisions (under 10 CFR 50, Appendix B) that are applied to these coatings. Pertinent portions of Appendix B quality assurance provisions are applied to the procurement of the inorganic zinc coating. These portions include DBA testing, manufacture, and traceability. Appendix B provisions also apply to the initialinspection that is performed after application of the coating. This inspection includes a non destructive dry film thickness test (ASTM D1186) and a MEK rub test (ASTM D4751) Long term monitoring of the coating is provided by visualinspections performed during refueling outages. It is not necessary to apply Appendix B to the application of this coating because the post application inspections detect improper application. Tl e coatings used in areas close to the recirculation screens are safety related in conformance with the intent of Regulatory Guide 1.54. SSAR subsection 6.3.2.2.7.3 describes where safety related coatings are used inside the containment. The standards referenced in Regulatory Guide 1.54 are not current. The draft revision to Standard Review Plan identifies that the quality assurance standards of ASTM D3842 and ASTM D3911 should be followed. The discussion of Regulatory Guide 1.54 in Appendix 1 A of the SSAR will be changed to reflect this The quality assurance requirements for safety related coating procurement, application, inspection and surveillance are subject to the pertinent provisions of Appendix B to 10 CFR Part 50 as shown in new SSAR Table 6.12 attached to this response. Part 2: Plant design features and the type of coatings selected for the majority of other coatings used inside containment permit the use of nonsafety related coatings. Debris resulting from failure of these nonsafety relca coatings will not have a negative impact on the performance of safety related, post-accident cooling systems. To avoid the use of inappropriate types of coatings inside conta;nment and to provide coating material traceability, the procurement of the manrity of nonsafety related coatings used inside containment will be subject to pertinent portions of 10 CFH 50, Appendix B. in conformance with the intent of Regulatory Guide 1.54. Application, inspection and surveillance of these coatings is not treated as safety related. Coatings procured as safety related will be used inside containment on walls, floors, ceilings, structural steel which is part of the builddg structure and on the polar crane as identified in Table 1. j New SSAR Table 6.12 attached to this response indicates the quality assurance provisions (portions of 10 CFR 50, Appendix B) applicable to these coatings. Procunng coatings that are subject to 10 CFR I Part 50, Appendix B quality assurance requirements, provides confidence that the coating material information used in settling calculations is valid. This change also increases the probability that the coatings will not fail even through their application, inspection, and surveillance are not safety-related A revision to SSAR subsection 6.1.2.1.1 is attached tnat implements this change, it is not necessary to apply Appendix B to the coatings on engineered components based on the following: m m2 W westingh0use

NRC FSER OPEN ITEM Revision 1 / H I e The total surface area of coatings applied to engineered components is a small percentage of the total area of coatings inside containment. The coatings applied to engineered components are less subject to failure during accidents because their dimensions are smaller and their shapes are more complex. Their shapes are complex involving many corners, angleb, nuts, bolts, protrusions, holes, etc. For engineered components, temperature changes cause smaller relative expansions and their complex shapes tend to prevent relative movement so that failure of the coating bond is less likely. In addition, even if the coating bond does f ail, it tends to retain the coating. Coatings applied to engineered ecmponents are done so in controlled factory conditions which include application of coatings in a timely fashion after manufacture, ;asier control of surf ace conditions, automated application of coatings and use of personnel that are highly trained. This results in high quality, durable coatings. In order to reduce pollution, manufacturers have switched to the use of dry powder coatings (polyesters) and water reduced coatings (acrylics). For components located inside containment vendors use dry powder coatings because water reduced coatings are not suitable for use in the harsh containment environment. Dry powder coatings tend to be very tough and defects in application tend to be noticeable. They also have relatively high densities, greater than epoxys, so that even if they did fall they would settle out before reaching the recirculation screens. Engineered components are located throughout the containment such that the majority are located where coating debris would settle out well away from the rectrCulation screens. Even in the unlikely event that some of these coatings failed, delaminated and did not settle out because of their location and coating characteristics, ta PXS recirculation screens have significant flow area margin for PXS recirculation. Part 3: The AP600 has several unique characteristics that allow the plant to tolerate the failure of nonsafety-related coating used inside containment. Table 1 attached to this response provides a lis! of these characteristics. These characteristics include long settling times between the end of RCS blowdown during a LOCA and the beginning of recirculation; the large water volumes provided by the PXS and the shape of the Containment lower volumes provides for high flood up levels. These high flood up levels allow the containment recirculation scr00n3 to be locateo relatively high. The bottom of the screens are located well above the lowest elevations of the containment which allows coating debris to settle out without challenging the bottom of the screens. The screens are very tall, which further reduces the Chance that coatog debris can reach the screens. The AP600 screens also have a unique feature (screen plates) that have been added to specifically prevent Coating debris %m entering the post accident containment water close to the screens and potentially blocking the screens. These screen plates are located above each recirculation screen and extend well cut in front and tu the sides of the screens. W westinghouse

NRC FSER OPEN ITEM Revision 1 V Y Ant".her AP600 characteristic that reduces the potential for coating debris blocking screens is that fibrous insulation is not used where it may be damaged by a LOCA. This eliminated the potential adverse interaction where fibrous debns act like a fine filter and collects small particles of dust or coating debris and leads to unacceptable screen pressure drops. As discussed in Part 1 of this response, the nonsafety related coating material used in the containment will be procured with 10 CFR Part 50, Appendix B quality assurance requirements, As a result, the nonsafety related coatings are not expected to fail. However, to provide a robust design, the failure of the nonsafety related coatings is assumed. A calculation was penormed to demonstrate that even with the failure of the nonsafety related coatings inside containment that the recirculation screens are not blocked. The application of 10 CFR Part 50, Appendix B quality assurance to the manufacture and procurement of nonsafety related coating materials allows the characteristics of the paint to be identified in terms of density and failure mechanisms. This information makes it possible to bound the size and density of the debris that might be generated by failure of these coatings and have the potential for blocking the screens. The key inputs to this calculation are shown in Table 2. Figure 1 shows the results of these calculations. TKs figure shows that coating debns entering the post accident flood up water at the edge of the protective plate will settle to an elevation that is below the l bottom of the screens after drifting about 7 feet, which is at about 3 feet away from the screen. This calculation includes significant conservatisms, including the lightest / smallest coating debris. It assumes the maximum recirculation flow of 1600 gpm, consistent with the maximum flow from one RNS pump operating unthrottled with suction taken from one screen. Another area where significant margin has been applied is in the debris settling rates. AP600 uses debris settling rates that contain a factor of 2 margin compared to the reference settling data (Reference 1) which is sufficient to cccount for l uncertainties. Without this added margin, the coating debris settles out very quickly in about 3 feet, as shown in Figure 1. Soveral sensitivity studies have been performed to demonstrate the robustness of the AP600 design with respect to uncertainty in coating debris settling. The margin applied to settling rates was l increased from a f actor of 2 (base AP600) to a factor of 2.3 so that debns just started to reach the bothm of the screen. The margin factor has to increase to 3.2 before debris could block 50% of the screen and about 4.7 before debns could block 90% of the screen. Even with such extreme margins, the PXS recirculation wou'd continue to function because the recirculation scre,an can tolerate significant 5 blockage and still support RNS pump operation. Note that if screen blockage ever reached the point where the RNS pumps cavitated and stopped operating, the PXS would revert to gravity recirculation at its lower flow rate (< 700 gpm). The debns settlirig calculation performed for the AP600 provides confidence that the plant can sustain complete failure of the nonsafety related coatings located inside containment without blockage of the receculation screens. [ Westingh00S8

NRC FSER OPEN ITEM Revision 1 y Part 4: An ITAAC has been added to the PXF ITAAC (Tab le 2.2.3-4) which ensures that the nonsafety related coatings used in the AP600 containment are consistent with the coatings used in the AP600 coating debris settling calculation, The key coa'ing characteristic used in this calculation is the coating density. A density of 100 lb/ft3 is expected to bound the coatings that will be used in the containment. Other coating characteristics such as debris particle diameter are bounded by the analysis. The smaller the particle diameter the slower the settling rate. The AP600 settling calculation assumes the smallest debris particle size that could block the screen openings. Larger particles settle faster and smaller particles would nct block the screens.

References:

1, Gibbs and Hill Report, " Evaluation of Paint and insulatior Debris Effects on Containment Emergency Sump Performance", Revision 1 September 1994. SSAR Change: Revised SSAR sections are attached (6.1.2.1.1,6.1.2.1.6, Appendix 1 A conformance 1.54) ITAAC Change: I Added new ITAAC to PXS, section 2.2.3-4 Vj Westinghouse

NRC FSER OPEN ITEM Revision 1 $!E % l Table 1 - AP600 Post LOCA Recirculation Conditions Time of Initiation of Recirculation - Large LOCA 5 hr DVI (8") LOCA 5 hr DVI (8") LOCA with max RNS operation 3 hr Flood up level (water level) above RV cavity floor 35 ft above 1000 compartment floor 24 ft Screen elevation Bottom screen above RV nearby f',or 2 ft Bottom screen above RV cavity floor 15 ft Heigh; screen 13 ft Recirculation flow rates - Maximum total (no failures, all pumps, both screens) 2600 gpm - Expected total (no failures, all pumps, both screens) 2000 gpm Maximum per screen (one RNS pump, one screen) 1600 gpm . Maximum per scraen (PXS, no failure, one screen) 700 gpm 3 Westirighouse

NRC FSER OPEN ITEM Revision 1 p mj; l l Table 2 -Inputs to Coating Debris Settling Calculation Coating debris: Shape circular disk Diameter 2 200 mils - Thickness 2 5 mils - Density 2100 lb/ft3 Screen geometry: Height 13 feet - Width 5.5 feet - Distance off floor 2 feet Protective plate: Height above top screen 0 to 10 feet - Distance plate extends out from screan 10 feet Screen flow rates: Maximum RNS flow per screen 1600 gpm (1) Maximum PXS tiow per screen 700 gpm (2) Notes: (1) The following con::ervative assumptions are made in calculating the maximum RNS pump driven flow rates. One RNS pump is assumed to take suction from one recirculation screen. The RCS pressure is assumed to be equal to the containment p essure. The RNS pump is assumed to have a conservatively high head vs flow charactenstics. Cavitation of the pump due to inadequate NPSHa is conservatively ignored. The piping and equipment flow resistances are assumed to be low. (2) The following cc iservative assumptions are made in calculating the maximum PXS gravity dnven flow rates. All PXS IRWST injection valves are assumed to open. Only one screen is assumed to operate. The one operable screen feeds both DVI lines. The pipe and equipment flow resistances are assumed to be conservatively low. The containment water level is assumed to be conservatively high. The RCS pressure is assumed to equal to the containment pressure. "*7 W westingn0use

NRC FSER OPEN ITEM Revision 1 Table 3 - AP600 Coated Surfaces, Containment Shell and Surfaces inside Containment Surface Boundary Surface Coating Coating Functions I Safety Safety-Related Coating Requirevnents (a) Motorial Classifications Procure Applicat inspect Surveilla ment (b) ion (c) ion (d) nce (e) Containment Shell, Shell surfaces Carbon Inorganc Zinc 1 Promote wettabihty 1 Safety yes no yas' yes Outside Surface above elevation Steel 2 Heat conduction 2 Safety 135'3* 3 Nondetatchable 3 Safety 4 inhbd corrosion 4 Nonsafety Containment Shell, Sheft surfaces Carbon inorganc Z nc 1 Promote wettabilny 1 Safety (1) yes no yes yes inside Surface above 7 feet Steel 2 Heat conduction 2 C afety above operating 3 Nondetatchable 3 Safety deck 4 Inhiba corrosion 4 Nonsafety inside Containment Areas surround og Carbon Inorganc Zinc 1 Nondetachable 1 Safety yes yes yes yes the containment steel with Epoxy 2 inhiba corrosion 2 Nonsafety recirculation Topcoat 3 Enhance twiioactive 3 Nonsafety screens (2) decontaminatior: Concrete walls. Concrete Epoxy Sealer 1 Ensure settling 1 S afety ves (3) no no no ceilings and floors wth Epoxy 2 Prevent dusting 2 Nonsafety Topcoat 3 Protect from 3 Nonsafety chemical attack 4 Ennance radioactive 4 Nonsafety decontamination Steel walls, Carbon inorganc Zinc 1 Ensure settling 1 Safety y is (3) no no no ceihngs, floors, Steel 2 inhibn corrosion 2 Nonsafety columns, beams, braces, plates Steel walls, Carbon inorganic Zinc I Ensure settling 1 Safety yes (3) no no no ceshngs, floors. Steef wdh Epoxy 2 inbtit corrosion 2 Nonsafety columns, beams, Topcoat 3 Enhance radioactive 3 Nonsafety braces, plates decontamination 6590(R1)-8 W westinghouse

NRC FSER OPEN ITEM hs / Revision 1 fT Table 3 - AP600 Coated Surfacus, Containment Shell ar.L Surfaces inside Containment Surface Boundary Surface Coating Coating Functiona / Safety Safety-Related Coating Requirements (a) Mate;ial Classifications Procure Applicat inspect Surveilla ment (b) lon (c) ion (d) nee (e) Miscellaneous Carbon Galvanized 1 Ensure setthng 1 (4) no no no no steel - ceilings, Steel 2 inhba cortesion 2 Nonsafety stairs, gratings, conduit, ducts Major components Carbon (5) 1 inhbit corrosion prior 1 (5) no no no no Steel to initial operaten l Engineered Varies Per 1 inhbrt conosion 1 Nonsafety no no no no components manufacturer M. weous pepe Carbon Inorganic Zinc 1 Ensure setthng 1 (4) no no no no /v Steel 2 Inhed costosion 2 Nonsafety i l i 6590(RI)-9 W Westinghouse l

P8RC FSER OPEN ITEM Revision 1 Table 3 - AP600 Costed Surfaces, Containment Shell and Surfaces inside Containtrent Alpha Notes: Safety-related asnects of coatings are separated into four categories; procurement, appleaton, inspection, arxi survedlance as descriwd in the following Alpha Notes a. (b. c, d, e). 10 CFR 50 Appendix B, quality assurance requirements apply when the aspect is safety-related. b. Procurement of coatings as safety related includes qualifcation for apptcable post accident conditions, control of materiats used to manufacture the coating, manufacture of the coating, and control of the coating after manufacture up to but not including application. Appicaten of coatings as safety-related includes qualifcatbn of personnel, preparation of surfaces and verdcaton that coating procedures were fo8 owed. c. d. Inspection of coatings as safety-related includes the inspections that verdy that the coating has been applied properly and include qualdrcation of personnel and inspections performed (vs coating type). Survedlance of coatings as safety-related includes the inspections that vedy that the f.oatings have not degraded during plant operation and include qualdcaton of e. personnel, inspections performed (vs coating type) and inspecten frequency.- Notes: 1. Coating on inside of containment shell is not required to promote the formation of a than water fdm. however it has been included an the PCS testog and analysis. 2. Areas around FXS recirculation screens that require safety-related coatings are defined in SSAR subsection 6.322.7.3. 3. Appendix B manufacture not required for caatings in the CVS room inside containment because this room is isolated from the recirculation screens. 4. Density of inorganic zinc and galvanizing is signifcantly greater than the density used in the AP600 settling analysis. 5. Coating used to prevent corroston following manufacture untd operation. i 6590(RI)-10 W wesonghouse s

NRC FSER OPEN ITEM Revision 1 QiB 54 1 Figure 1 - AP600 Coating Debris Settling This figure shows that the AP600 has significant margin to accommodate uncertainty in coating settling rates. The heavy solid line shows the AP600 design case, which includes a margin factor of 2.00 times the reference settling data (Reference 1). The light dashed lines represent sensitivity studies with greater margins. The margins applied were arbitrarily chosen to force the debris to settle out at the bottom of the screen (factor of 2.3), covering half of the screen (facto. of 3.2) and covering 90% of the r ieen (factor of 4.7). Even with 90% of the screen blocked, the PXS would still be able to operate. The light solid line shows the reference settling data. All of these cases assume the rcaximum possible RNS flow assuming one RNS takes suction from one screen and is unthrottled (1600 gpm). 25 i Plate 4 l

o

' g_L,..__ _ _ _.._.._-.. _. _... _.... _... _ _ _ _ _ l g \\ \\, k m + w --A,-en or.~. e - ~,. ~w- -.e 5 's, j ~ j d I go S s ~ 3 ~ Cont. ~ Rectre. Screens AP600 o L i: b Floor 18 .lu 9 4 J 6 .4 ..I .1 2 .I o i 2 Dutance Frim Screen (ft) T westirigh0ese

NRC FSER OPEN ITEM Revision 1 f 1 Revised SSAR Subsection 6.12 6.1.2 Organic Materials 6,1.2.1 Protective Coatings 6.1.2.1.1 General The AP600 is divided into four areas with respect to the use of protective coatings. These four areas are: inside containment e Exterior surfaces of the containment vessel Radiologically controlled sieas outside containment e Remainder of plant. The cor.siderations for protective coatings differ for these four breas and the coatings selection process accounts for these differing considerations. The AP600 design considers the function of the coatings, their potential failure modes, and their requirements for maintenance. Table 6.12 lists different areas and surfaces inside containment and on the containment shell that have coat'ngs, their functions and to what extent their coatings are safety related. C' 9 tings used outside containment do not provide safety-related functions except for the coating on t s outside of the containment shell. The coating on the outside of the containment above elevation 135' 3" shell supports passive containment cooling system heat transfer and is classified as safety related. The coating used on the inside surface of the containment shell, greater than 7' above the operating deck, is not required to support passive containment cooling system heat transfer. However, passive containment cooling system testing and analysis have been performed with a coating. This 3 coating is classified as safety-related. Coatings used in the vicinity of the containment recirculation screens are classified as safety 4 elated in order to prevent their failure from producing debris that may be transported to the screens. j Subsection 6.3.2.2.7.3 defines the area where safety-related coatings are used in the vicinity of the recirculation screens. Other coatings used inside containment, except for the containment shell, are classified as nonsafety related because their failure does not prevent functioning of the engineered safety I features. If the nonsafety related coatings delamiinate, the solid debris they may form will not have a negative impact on the performPnce of safety related post-accident cooling systems. See subsection 6.1.2.1.5 for a discussion of the factors including plant design features and Icw water flows that perrh.t the use 01 nonsafety4 elated paint inside conte nment. Protective coatings are 6590(RI)-12 I

NRC FSER OPEN ITEM Revision 1 jj l maintained to provide corrosion protection for the containment pressure boundary and for other sysiem components inside containment. l The corrosion protection, good housekeeping and decontamination functions of the coatings are nonsafety related functions. l For information on costing design features, quality assurance, material and application requirements, and performance monitoring requirements, see subsection 6.1.2.1.6. 6.1.2.1.2 Inside Containiaent Carbon Steel l Inorganic zinc is the basic coating applied to the containment vessel and structural carbon steel inside containment that need coating. Below the operating floor, the inorganic zinc coating is top l coated with epoxy where enhanced decontamination is desired. The epoxy top coat also extends above the operating floor on structural modules and to a wainscot height of 7 feet above the operat-ing floor on the containment vessel. Where practical, miscellaneous carbon steel items (such as l ceilings, stairs, gratings, ladders, railings, conduit, duct, and cable tray) are hot-dip galvanized. Steel surfaces subject to immersion dunng normal plant operation (such as suinps and gutters) are l stainless steel or are coated with epoxy or epoxy phenolic applied directly to the carbon steel without an inorganic zinc primer. Caibon steel structures and equipment are assembled in modules l and the modules are coated in the fabrication shop under controlled conditions. Concrete l Concrete surfaces inside containment are coated to prevent concrete from dusting, to protect it from chemical attack and to enhance decontaminability. In keeping with ALARA goals the exposed concrete surfaces are made as decontaminable as practicalin areas of frequent personnel access and areas subject to liquid spray, splash, spillage or immersion. l Exposed concrele surfaces inside containment are coated with an epoxy to help bind the concrete surface together and reduce dust that can become Contaminated and airborne. Concrete floors inside containment are coated with a self-leveling epoxy. Exposed concrete walls inside con-l tainment are coated to a minimum height of 7 feet with an epoxy applied as a top coat over an epoxy surf acer that has been struck flush. 6.1.2.1.3 Exterior of Containment Vessel The exterior'of the containment vesselis coated with the same inorganic zinc as is used inside of tne containment. The inorganic zinc coating enhances heat transfer by providing good heat c,onduction and by enhancing surf ace wetting of the extenor surface of the containment vessel. T he Inorganic zinc also provides corrosion protection. p 6590(R1)-13

NRC FSER OPEN ITEM Revision 1 jin % 6.1.2.1.4 Radiologically Controlled Areas Outside Containment and Remainder of Plant The coatings used in the radiologically controlled areas outcide containment and in the remainder of the plant are nonsafety related. However, coatings are selected, specified and applied in a manner that optimizes performance and standardization. Wherever practical, the same coating systems are used in radiologically controlled areas outside containment as are used inside containment. The ALARA concept is carried through in areas subject to radiation exposure and possible radiological contamination. The rernainder of the plant coating systems are commercial grade materials that are selected and applied according to the expected conditions in the specific areas where the coatings are applied. The coatings used in radiologically controlled areas outside of containment are identified in the following. Carbor. Steel Surfaces Carbon steelis coated with inorganic zinc. An enoxy top coat is used in areas suoject to decontamination such as a 7 foot wainscot in high traffic areas or on surfaces subject to raciologically contaminated liquid spray, splash, or spills. Concrete Floors Floors subject to heavy traffic or contaminated hquid spills are coated with self-leveling epoxy. An Epoxy top coat is applied a minimum of 1 foot up the wall where liquid spills might splash. Floors subject to light traffic and not subject to contaminated liquid spills are coated with an epoxy top coat. The epoxys applied to concrete surfaces are the same epoxy used as a top coat for the inorganic zinc-coated steel. Concrete Walls A 7-foot wainscot on exposed concrete walls in high traffic areas and any surfaces of walls subject to spray. splash or spills of contaminated liquids are coated with an epoxy top coat applied over an epoxy surfacer that has been struck flush. The epoxys used on concrete surfaces are the same as that used as a top coal for the inorganic zinc coated steel. Remaining concrete walls are coated I with an epoxy sealer to reduce or eliminate dusting. Concrete Ceilings l Exposed concrete ceilings are coated with an epoxy sealer to reduce dusting. 6590(RI)-14

NRC FSER OPEN ITEM Revision 1 y% 3.1.2.1.5 Sc.foty Evaluation This subsection rfescribes the basis for the oxtent of safety related coatings and the basis for classifying coatings on other areas inside containment as nonsafety related. Table 6.12 identifies the extent to which coatings are classified as safety-related. The coating on the outside of the containment shell above elevation 135' 3" supports passive containment cooiing system heat transfer and is classif.ed as safety-related. The coating used on the inside surface of the containment shell, greater than 7' above the operating deck, is not required to support passive containment cooling system heat transfer, however, passive containment cooling system testing and analysis have been performed with a coating. This coating is classified as safety-related. The AP600 has a number of design features that facilitate the use of nonsafety-related coatings. These features include a passive safety injection system that provides a long delay time (more than 5 hours) between a LOCA and the time recirculation starts. This time delay provides time for settling of debris. These passive systems also flood the containment to a high level which allows the use of containment recirculation screens that are located well above the floor and are relatively tall. Significant volume is provided for the accumulation of coat;ag debris without affecting screen plugging. These screens are protected by plates located above the screens that extend out in front and to the side of the screens. Coatings used under these plates in the vicinity of the screens are classified as safety related. The protective plates, together with low recirculation flow, approach velocity and the screen size preclude postulated coating debris above the plates from reaching the screens. Refer to suosection 6.3.2.2.7.3 for additional discussion of these screens, their protective plates and the areas utilizing safety related coatings. The recirculation inlets are screened enclosures located near the northwest, a southwest corners of the east steam generator compartment (refer 'o the figures in Section 6.3.2.2.7.3). The enclosure bottoms are located above the surrounding floor which prevent ingress of heavy debns (specific gravity greater than 1.05). Additionally, the screens are onented vertically and are protected by large plates located above the screens further enhancing the capability of the screens to tunction with debns in the water. The screen mesh size and the surface area of the containment recirculation screens in the AP600, in conjunction with the large floor area for debris to settle on, can accommodate failure of coatings inside containment dunng a design basis accident even though the residue of such a failure is unlikely to be transported to the vicinity of the enclosures. The AP600 does not have a safety-related containment spray system. The containment spray system provided in the AP600 is only be used in beycnd design basis events. This reduces the chance that coatings will peel off surfaces irside containment because the thermal shock of cold spray water on hot surfaces combined with the rapid depressurization following spray initiation are recognized as contrioutors to coating failure. Parts of the containment below elevation 107'-2" are flooded and water is recirculated through the passive core cooling system. However, the volume of riter moved ir this manner is relatively small and the flow velocity is very low. [ Westingh00S8

NRC FSER OPEN ITEM Revision 1 if 1 The coating cystems used inside containment alsu include epoxy coatings. These are applied to concrete substrates, as top coats over the inorganic zine primer, and direct'y to steel, as noted in subsection 6.1.2.1.2. The failure modes of these systems could includa delamination or peeling if the epoxy coatings are not properly applied (References 6,7,8). The epoxys applied to concrete and carbon steel surfaces are sufficiently heavy (dry film density greater than 100 lb/ft3) so that transport wi'h the low water veloci;y in the AP600 containment is negligible. Inside containment, there are engineered components ceated wi'h various manufacturer's standard coating systems which are also classified as nonsafety related and may peel or delaminate under design basis accident conditions. The density of these coatings la not limited based on the following considerations: The total surface area of low density coatings applied to engir,eered compunents is a small percentage of the total area of coatings inside containment. The coatings applied to engineered componsnts are less subject to failure during accidents because their dimensions are smaller and their shapes are more complex. Their shapes a:a complex involving many corners, angles, nuts, bolts, protrusions, holas, etc. For engin:ared components, temperature changes cause smaller relative expansions and their complex shapes tend to prevent relative movement so that failure of the coating bond is less likely. In addition, even if the coating boad does fail, it is less likely to detach because the complex shapes tend to retain the coating. l Coatings applied to engineered components are done so in controlled factory conditions so that l the quality of application is better than that achieved in the field. Factors contributing to this higher quakty include application of coatings in a timely fashion after n,anufacture, easier control of surface conditions, automated application of coatings and use of personnel that are highly trained. Manufacturers have switched to the use of dry powder coatings (polyesters) and water reduced coatings (acrylics). Coatings used on components located 'nside cantainment are expected to be dry powder coatings because water reduced coatings are not suitable for use in tM harsh containment environment. Dry powder coatings tend to be very tough and defects in application i tend to be noticeable. They also have relatively high densmes, greater than epoxys, so that even if they did fail they would settle out before reaching the recirculation screens. l Engineered components are located throughout the contdinment so that the ma,ority are located where low derisity coating debris settle out well away from the recircule'.lon screens. Even in the unlikely event that some of these coatings fail, delaminate and do not settle out because of their location and low density, the PXS recirculation screens will prevent blockage of the PXS recirculation. h westinghouse 6mm6

NRC FSER OPEN ITEM Revision 1 7@ l Production of hydrogen as a result of zinc corrosion in detgn basis accident conditions, including the zinc in paints applied inside containment, is addressed in subsection 6.2.4.3.1. 6.1.2.1.6 Design Features A number of design features provide confidence that the coating systems inside the containment, on the exterior of he containment vessel and in potentially contaminated areas outside containment will l perform as ini.ided. Refer to Table 6.1-2 for definition of the safety-related aspects of coatings. These features enhance the ALARA program and enhance corrosion resistance. The features include: Specification of qualified coating materials Provision of coating specifications Provision of coating procedures Use of qualified painters Use of qualified coatings inspectors a Documentation of coatings work Performance of as much coating work as practical under controlled shop conditions Specification of coating performanct nonitoring Specification of coating inspection and maintenanco Safety related coatings l Saf aty related coatings meet the pertinent provisions of 10CFR Part 50 Appendix B. The safety-reiated coating systems arid their applications is concistent with the intent of Regulatory Guide 1.54, 'Ouality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants' and the quality assurance standards of References 4 and 5. Refer to Table 6.1-2 for identification of coating applications in the AP600 and the aspects covered by Appendix B including manuf acture, appl; cation, inspection and surveillance. Nonsafety Related Coatings The use of nonsafety-relL!ed coatings inside containment is based on the use of selected types of s coatings and the properties of the coating if it deteriorates. To preclude the use of inappropriate coatings, appropnate quality assuranca requirements are applied to the manufacture, procurement, handling, and storage of the nonsafety related coating used inside containment on internal structu es. Refer to Table 6.1-2 for a listing of coating applications in the AP600 and the aspects covered by Appendix B. The application, inspection and monitonng of nonsafety-related coatings used inside containment are not classified as safety-related as shown in Table 6.1-2. The application, inspection and monitanng of nonsafety-related coatings ce controlled by a program prepared by the Combined License applicant. This program is not subject to 10 CFR 50. At endix B, quality assurance W westinghouse _J

NRC FSER OPEN ITEM Revision 1 l requirements. The specified coatings used inside containment are tested for radiation tolerance per ASTM D4082 (Reference 1), for decontaminability per ASTM D42% (Reference 2) and for performance under design basis accident conditions per ASTM D3s11 (Reference 3). The coatings used in radiologically controlled areas outside containment are tested for radiation resistance ano decontaminebility out are not specified to be design basis accident tested. Whet, practical, the same coating materials are used in radiologically controlled areas outside containment as are used inside containment. This provides a high level of quality and optimizes maintenance painting over the life of the ple Appendix B to 10 CFR Part 50 applies to manufacture of coatings used inside conta!nment on internal structures, including walls, floor slabs, structural steel, and the polar crane. The coating manufacturer is required to manufacture the coatings under a suitable quality assurance program and to provide a product identity certification record. Coating specifications also require that the surfaces to be coated are oroperly prepared, coated, inspected and documented. For coatir'gs used inside containment and in radiologically controlled areas outside containment the coating applicator follows acceptable procedures to provide confidence that correct coating practices are used. Appendix B to 10 CFR 50 quality assurance i requirements do not app!y to the application and inspection of nonsafety related coatings. l Due to the.aodularized construction. a significant portion of the containment coatings are shop applied to the containment vessel and to piping, structural and equipment modules. This application of coatings under controlled shop conditions provides additional confidence that the coatings will perform as designed and as expected. 6.1.2.2 Other Organic Materials A listing of other organic materials in the containment is developed based on the specific type of equipment and the supplier selected to provide it. Materials are evaluated for potential interaction with engineered safety features to provide confidence that the performance of the engineered safety features is not unacceptably affected. 6.1.3 Combined License information items 6.1.3.1 Procedure Review The Combined License applicants referencing the AP600 will address review of vendor fabrication and welding ~ procedures or other quality assurance methods to judge conformance of austenitic stainlass steels with Regulatory Guides 1.31 and 1.44 T Westinghouse mma

NRC FSER OPEN ITEM Revision 1 F 6.1.3.2 Coating Program The Cocinea Ucense applicants referencing the AP600 will address preparation of a program to control testing, application, end monitoring of nonsafety related coatings. The program for the control of the use of coati igs applied to the containment vessel (inside and outside containment above the operating deck) and other coatings bsed inside containment will be consistent witn Table 6.1 -2. 6,1.4 References 1. ASTM-04082, " Test Method for Effects of Radiation on Coatinga Used in Ught-Water Nuclear Power Plants." 2. ASTM-D4256, " Test Method hr Determination of the Decontaminability of Coatings Used in Light Water Nuclear Power Plants." 3. ASTM-D3911. " Test Method for Evaluating Coatings used in Light-Water Nuclear Power Plants at Simulated Design Basis Accident (DBA) Conditions." 4. ASTM D3842, " Selection of Test Methods for Coatings for Use in Light-Water Nuclear Power i Plants," 5. ASTM D3911. " Evaluating Coatings Used in Ught-Water Nuclear Power Plants at Simulated Design Basis Accident (DBA) Conditions." 6. NUREG 0797," Safety Evaluation Report related to the operation of Comanche Paak Steam l Electric Station, Units 1 and 2." 7. Bolt R.O. and J.G. Carroll, " Radiation Effects on Organic Materiais", Academic Press, New York.1963 Chapter 12. l 8. Parkinson, W.W. and O. Sisman, "The Use of Plastics and Elastomers in Nuclear Radiation", Nuclear Engineering and Design 17 (1971), pp 247-280, North-Holland Publishing Co.. Amsterdam. T westingtiouse 6s a w

NRC FLER OPEN ITEM Revision 1 if is l hew SSAR Table 6 f-2 Table 6.1 APOO Coated Surfaces. Containment Shell and Surfaces inside Containment Surface Boundary Surface Coating Coating Functions / Safety Safety-Related Coating Requirements (a) Material Classifications Procure Applicat inspect Surveilla ment (b) ion (c) ion (d) nce(e) Containment Shell. Shell surfaces Carbon Inorganic Zinc 1 Promote wettability 1 Safety yes no yes (4) yes (4) Outside Surface above elevatsn Steel 2 Heat conduction 2 Safety 135'3' 3 Nondetatchable 3 Safety 4 inhbit corrosion 4 Noasafety Containment Shell, Shell surfaces Carbon Inorganic Zinc 1 Promote wettWiity 1 Safety (1) yes no yes (4) yes (4) Insde Surface above 7 feet Steel 2 Heat conduction 2 Safety above operating 3 Nondetatchable 3 Safety deck 4 inhibit corrosion 4 Nonsafety Insde Containment Areas surrounding Carbon Inorganic Zinc 1 Nondetachable 1 Safety yes yes yes yes ' the containment steel with Epoxy 2 inhibit corrosion 2 Nonsafety recirculation Topcoat 3 Enhance radioactive 3 Nonsafety screens (2) decontaminaton Concrete wa!Is, Concrete Epoxy Seoler 1 Ensure setthng 1 Safety yes (3) no no no ceilings and floors with Epoxy 2 Prevent dusting 2 Nonsafety Topcoat 3 Protect from 3 Nonsafety chemical attack 4 Enhance radioactive 4 Nonsafety decontamination Steel walls, Carbon Inorganic Zinc 1 Ensure settling 1 Safety yes (3) no no no ceihngs, floors. Steel 2 Inhibit corrosion 2 Nonsafety columns, beams, braces, plates 6590(R1)-20 W Westinghouse l i tr

NRC FSER OPEN ITEM i Revision 1 nu .I Tab e 6.1 AP600 Coated Surfaces, Containment Shell and Surfaces inside Containment Surface Bouridary Surface Coating Coating Functions / Safety Safety-Related Costing Requirements (a) Material Classifications Procure Applicat inspect Surveilla ment (b) ion (c) ion (d) nce (e) Steel walls, Carbon Inorganic Zinc 1 Ensure settling 1 Safety yes (3) no no no ceilings, floors, Steel w:th Epoxy 2 Inhibd corrosion 2 Nonsafety j columns, beams, Topcoat 3 Enhance radioactive 3 Nonsafety braces. plates decontamination Alpha Notes: a. Safety-related aspects of coatings are separated into iour categones; procurement, appik: ; ion, inspection, and surves!!ance as described en the fn!!owing Alpha Notes (b c, d, e) 10 CFR 50, Appendix B, qualdy assurance requirements apply when the as ect is safety-related. p b. Procmement of coatings as safety related includes qJafdcation for appleable post accident conditions, control of materials used to manufacture the coating. manufacture of the coating, and control of the coating after manufacture up to but not including appleation. Appleation of coatings as safety-related includes qualdcation of personnel, preparation of surfaces and verdcaten that coating procedures were fo!Iowed. c. d. Inspe'inon of coatings as safety-related includes the inspections that venfy that the coating has beert applied properly and inc'ude qualdcaton of personnel and inspections performed (vs coating type). Surveillance of coatings as safety-related includes the inspections that verdy that the coatings have not degraded during plant operaton and include qualdcation of e. personnel, inspectons performed (vs coating type) and inspecten frequency. Notes: 1. Coating on inside of containment shell is not required to,ntomote the formation of a thin water film, however d has been included in the PCS testing and analysis. 2. Areas a:ound PXS recirculation screens that require safety-related coatings are defined in SSAR subsection G.3.2.2.7.3. 3. Anpendix B manufacture not required for coatings in the CVS room inside containment, because this soom is isalated from the recirculation screens. 4. The inorgane zinc inspectron includes a non<festructive dry film thickness test (ASTM D1186) and a MEK rub test (ASTM D4752) performed after appication. It is not necessary to apply Appendix B to the appleation of this coating because the post appleation enspections detect improper appication. Long term surveillance of the coating is provided by visualinspectons performed during refueling outages. 6590(R1)-21 W Westinghouse

NRC FSER OPEN ITEM Revision 1 j!!E @ Reg. Guide 1.54, Rev. O,673 - Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants Criteria Referenced A P600 Section Criteria Position Clarification / Summary Description of thceptions l General ANSI N101.4 1972 Exception Some Gcoatings inside containment are nonsafety-related and satisfy appropriate ASTM Standards. See subsection 6.1.2 for additional infonnation. Application is controlled by procedures using qualified personnel to provide a high quality product. The paint materials for nonsafety related coatings inside the containment are subject to 10CFR Part 50 Appendix B Quality Assurance requirements. The standards referenced are not current. Appropnate current standards including the quality assurance stImdards of ASTM D3842 and ASTM D3911 are followed for safety related coatings and for procurement of nonsafety related coatings inside containment. An evaluation is made to show that failure of the nonsafety-related coatings does not prevent the function of safety-related systems. C. I ANSI N101.4 1972 Exception The standards referenced are not current. ANSI N45.2-1971 Appropriate current standards including ASMti NQA-1 are followed to develop the quality assurance program for safety related coatings and for procurement of nonsafety related l coatings inside centainment. ! C.2 ANSI N101.41972 Exception The standard referenced is not current. Appropriate current standards are followed. l l C.3 ANSI b"01.4 1972 Exception The standard referenced is not current. Appropriate current standards including the quality assurance standards of ASTM D3842 and ASTM D3911 are followed for safet'; related I coatings. l C.4 ANSI N101.4-1972 Conforms The standard referenced is not current. Guidance for the control of cleaning and other material used on stainless steel matenal is provided in Regulatory Guides 1.37 and 1.44 Subsections 5.2 3.4 and 6.1.1.2 address the control of processes used on stainless material for safety-related piping and components. I g 6590(RI)-22

NRC FSER OPEN ITEM Revision 1 Mt 1 New ITAAC, in PXS Section 2.2.3 Table 2.2.3 4 Inspections. Tests, Analyses, and Acceptance Criteria Design Commitment inspections Tests, Analyses Acceptance Criteria H.c) The PXS provides RCS v) Inspection of the as-built v) Plates located above each makeup, boration. and safety componcnts will be conducted for containment recirculation screen are injection Juring design basis events. plates located above the no more than 10 ft above the top of containment recirculation screens. the screen and extend out at least 10 ft from the trash rack portion of the screen, vi) Inspections of the IRWST and vi) he screen surface area (width x containment recirculation screens height) of each screen is 2 70 ft*, wi!! be conducted. The bottom of the containment recirculation screens is 2 2 ft above the klop compartment lioor. I vii) Inspections will be conducted vii) De type of insulation used on of the insulation used inside the these lines and equipment is not a containment on ASME class I lines fibrous type, t and on the reactor vessel, reactor coolant pumps, pressurizer and s.eam generators. viii)Irupections willbe conducted viii) A report e.tists ard concludes that of the as built nonsafety-related the coatings used on these swfaces coatings or ofplant records of the has a density of 2100 lbl)P. nonsafety related coatings used leide containment on walls, floors, ceilings, structural steel which ;s part of the btilding structwe and on the polar crane. e [ W8511rigil0'JSe

NRC FSER OPEN ITEM Revision 1 73 NRC Fax Dated 1/27/98 (OITS #6591) Staff Comment: Westinghouse provides some qualitative discussions on what it believes will occur upon failure of containment coatings (e.g., assumes only localized f ailure of coatings, assumes coating material will settle on the bottom of various compartments and will not get transpcrted in sufficient quantities which are undefined to the intake screens) but has not provided any evidence of this mechanism besides engineering judgment.

Response

3 As discussed in response to OITS #6590, the coatings used on the containment shell on the inside (7' above the operating deck) and on the outside (above the operating deck) are safety-related in conformance with the intent of Regulatory Guide 1,54. This approach eliminates the need to consider the failure of these coatings. As discussed in response to OITS #6590, the coatings used on surfaces located close to the PXS l recirculation screens are safety-refated in conformance with the intent of Regulatory Guide 1.54. This approach eliminates the need to consider the failure of these coatings. l l The coatings used inside the containment in the AP600 are greatly simplified as compared with operating plants, both in terms of number of different coatings and in terms of the ease of mixing and applying the coatings, This simplification reduces the chance of improper application. As discussed in response to OITS #6590, the majority of nonsafety-related coatings used inside containment are procured with 10 CFR 50, Appendix B, quality assurance requirements in conformance with the intent of Regulatory Guide 1.54. The application, inspection and surveillance of these nonsafety-related coatings is not performed with 10 CFR 50, Appendix 0, quality assurance requirements. This approach provides Coatings that are Consistent with coating charactenstics used in the debris settling calculations performed for the AP600. It also increases the probability that these coatings will not f ail during an accident. However, because the application, inspection cnd surveillance are not safety-related for these coatings, they are assumed to fail, as discussed in response to OITS #6590. The unique characteristics of the AP600 allows the plant to tolerate this failure. The coating debris settling calculation performed for the AP600 is not depend 6nt on the specific coating failure modes. The key coating characteristic of the coating with respect to settling is its density, which is controlled by the application of Appendix B to the procurement of these coatings. Other issues such as the size of the debris is addressed in the settling calculation by assuming the worst size, ie the smallest size that could block the screen opening. Larger debris sizes settle faster. The discussion of coating failure modes contained in the SSAR is based on an understanding of the characteristics of the coatings and on experience with their use. For example, inorganic zinc has a l relatively low tensilo strength and a chemically based bond which is not challenged by acciden; W85tl!)gh0US8

NRC FSER OPEN ITEM ( Revision 1 ff ' j i conditions. As a result, if it is not properly applied its failure would be powdering. Experience with inorganic zinc coatings confirms this. Ilure mode. On the other hand, epoxy has a high tensile strength and a mechanically based bond whicn may be challenged by accident conditions. As a resuh, if it is not properly applied its failure would be delamination or peeling. Experience with epoxy coatings confirms this failure mode. The SSAR subsection 6.1.2.1.5 and 6.1.4 have been revised to add references for the coating failure modes discussed there. SSAR Change: l Revised SSAR subsections 6.1.2.1.5 and 6.1.4 are attached to the response to OITS #6590(R1). ITAAC Change: None i T Westinghouse \\

NRC FSER OPEN ITEM Revie,lon 1 [% NRC Fax Dated 1/27/98 (OITS #6592) Staff Comment: In addition to the staff's concerns on debris transport to the containment sumps, the performance characteristics of PM passive containment cooling system (PCS) are based on an experimental test program with the mting. No testing is known to exist with either degraded coatings or without coatings. The justification for the mass and heat transfer correlations, the PCS f!!m model and the water coverage model are all based ori testing with the coating, In the Large-Scals Test (LST), the coating exists on both the exterior and the interior surface of the vessel. Failure of the coating will impact heat transfer, film formation and water coverage. Westinghouse has not provided any data or experimental evidence as to why deterioration of the containment coatings will not affect the design basis properties of the containment shell and the PCS performance. Westinghouse Response: 1 As discussed in response t - OITS #6530, the coatings used on the containment shell inside I containment (greater than '/' above the operating deck) and outside containment (above the operating i deck), are safety-related in conformai i with the intent of Regulatory Guios 1.54 This approach I eliminates the need to consider the failure of these coatings and the potential degradation of the passive I containment cooling system performance. SSAR Revisions: s None W westinghouse m2m i}}