Text

Rev 3 to AP600 Design Control Document
	ML20216J076
Person / Time
Site:	05200003
Issue date:	09/22/1999
From:	Mcintyre B; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20216J072	List: ML20216J068; ML20216J076;
References
	GW-GL-700, GW-GL-700-R03, NUDOCS 9910040175
	Download: ML20216J076 (108)
	v • d • e

AP600 DOCUMENT COVER SHEET Form 58202G(3/99) AP600 CENTRAL FILE USE ONLY: TDC: IDS: 1 S RFS#: RFS ITEM #:

Q AP600 DOCUMENT NO.

GW-GL-700 REVISION NO. ASSlGNED TO 3 Page 1 of 2 --

ALTERNATE DOCUMENT NUMBER: WCAP-15150 WORK BREAKDOWN #: 3.4.1 DESIGN AGENTORGANIZATION: Westinghouse TITLE: AP600 Design Control Document ATTACHMENTS: DCP #/REV. INCORPORATED 'N THIS DOCUMENT None REVISION:

Incorporates consistency corrections into 3/99 Revision (2)

CALCULATION / ANALYSIS

REFERENCE:

None ELECTRONIC FILENAME ELECTRONIC FILE FORMAT ELECTRONIC FILE DESCRIPTION Wordperfect (C) WESTINGHOUSE ELECTRIC COMPANY LLC 1999 WESTINGHOUSE PROPRIETARY CLASS 2 This document contains information proprietary to Westinghouse Electric Company LLC; it is submitted in confidence and is to be used solely for the purpose for which it is fumished and retumed upon request. This document and such information is not to be reproduced, transmitted, disclosed Jr used otherwise in whole or in part without prior written authortzation of Westinghouse Electric Company LLC, subject to the legends contained hereof.

WESTINGHOUSE PROPRIETARY CLASS 2C

/n\ This document is the property of and contains Proprietary information owned by Westinghouse Electric Company LLC and/or its subcontractors y/ and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditions of the agreement under which it was provided to you.

X WESTINGHOUSE CLASS 3 (NON PROPRIETARY)

COMPLETE 1 IF WORK PERFORMED UNDER DESIGN CERTIFICATION ~OR COMPLETE 2 IF WORK PERFORMED UNDER FOAKE.

1 DOE DESIGN CERTIFICATION PROGRAM - GOVERNMENT LIMITED RIGHTS STATEMENT [See page 2)

Copyright statement: A ticense is raserved to the U.S. Govemment under contract DE AC03-90SF18495.

L DOE CONTRACT DEllVERABLES (DELIVERED DATA)

Subject to specified exceptions, disclosure of this data is restricted until September 30,1995 or Design Certification under DOE contract DE-AC03-90SF18495, whichever is later.

EPRI CONFIDENTIAL: NOTICE: 1X 2 3 4 5 CATEGORY: AX B C D E F 2 ARC FOAKE PROGRAM - ARC LIMITED RIGHTS STATEMENT [See page 2)

Copyright statement: A license is reserved to the U.S. Govemment under contract DE-FC02-NE34267 and subcontract ARC 93 3-SC-001.

,_, ARC CONTRACT DELIVERABLES (CONTRACT DATA)

Subject to specified exceptions, disclosure of this data is restricted under ARC Subcontract ARC-93-3-SC-001.

AP600 RESPONSIBLE MANAGER SIGNATURE

f APPROVAL D TE Brian A. McIntyre __

gL F fy J

' Approval of the responsible manager signifies that document is complete, all regt.peTrewews are complete, electronic file is attached and document is released for use.

9910040175 990929

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AP600 DOCUMENT COVER SHEET Page 2 Form S8202G(S/94) LIMITED RIGHTS STATEMENTS DOE GOVERNMENT LIMITED RIGHTS STATEMENT (A) These data are submitted with limited rights under govemment contract No. DE AC03-90SF18495. These data may be reproduced and used by the government with the express limitat on that they will not, without wntten permission of the contractor, be used for purposes of manufacturer nor disclosed outside the govemment; except that the govemment rnay disclose these data outside the govemment for the following purposes, if any, provided that the govemment makes such disclosure subject to prohibition against further use and disclosure:

(1) This "Propnetary Data' may be disclosed for evaluaton purposes under the restnchons above.

(11) The " Proprietary Data

may be disclosed to the Electnc Power Research Institute (EPRI), electnc utility representatives and their direct consultants, excluding direct commercial competitors, and the DOE National Laboratones under the prohibitions and restnctions above.

(B) This notice shall be marked on any reproduchon of these data,in whole or in part.

ARC LIMITED RIGHTS STATEMENT:

This proprietary data, fumished under Subcontract Number ARC-93-3-SC-001 with ARC may be duplicated and used by the govemment and ARC, sub}ect to the limitations of Article H-17.F. of that subcontract, with the express limitat ons that the propnetary data may not be disclosed outside the govemment or ARC, or ARC's Class 1 & 3 members or EPRI or be used for purposes of manufacture without prior permission of the Subcontractor, except that further disclosure or use may be made solely for the following purposes:

This propnetary data may be disclosed to other than commercial compettors of Subcontractor for evaluation purposes of this subcontract under the restncton that the proprietary data be retained in confidence and not be further disclosed, and subject to the terms of a non-disclosure agreement between the Subcontractor and that organization, excluding DOE and its contractors. I DEFINITIONS CONTRACT / DELIVERED DATA - Consists of documents (e.g. specifications, drawings, reports) which are generated under the DOE or ARC contracts which contain no background proprietary data.

1 EPRI CONFIDENTIALITY / OBLIGATIONNOTICES NOTICE 1: The data in this document is subject to no confidentiality obligations.

NOTICE 2: The data in this document is propnetary and confidential to Westinghouse Electric Company LLC and/or its Contractors. It is forwarded to recipient under an obligation of Confidence and Trust for limited purposes only. Any use, disclosure to unauthorized persons, or ,

copying of this document or parts thereof is prohibited except as agreed to in advance by the Electnc Power Restarch Institute (EPRI) and Westinghouse Electric Company LLC. Recipient of this data has a duty to inquire of EPRI and/or Westinghouse as to the uses of the informahon contained herein that are permitted.

NOTICE 3: The data in this document is proprietary and confidential to Westinghouse Electric Company LLC and/or its Contractors. It is forwarded to recipient under an obligation of Confidence and Trust for use only in evaluation tasks specifically authorized by the Electne Power Research Institute (EPRI). Any use, disclosure to unauthorized persons, or copying this document or parts thereof is prohibited except as agreed to in advance by EPRI and Westinghouse Electric Company LLC. Recipient of this data has a duty to inquire of EPRI and/or Westinghouse as to the uses of the information contained herein that are permitted. This document and any copies or excerpts thereof that may have been generated are to be retumed to Westinghouse, directly or through EPRI, when requested to do so.

NOTICE 4: The data in this document is proprietary and confidential to Westinghouse Electnc Company LLC and/or its Contractors. It is being revealed in confidence and trust only to Employees of EPRI and to certain contractors of EPRI for limited evaluation tasks authonzed by EPRI.

Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited. This Document and any copies or excerpts thereof that may have been generated are to be returned to Westinghouse, directly n through EPRI, when requested to do so.

NOTICE 5: The data in this document is proprietary and confidential to Westinghouse Electric Company LLC and/or its Contractors. Access to this data is given in Confidence and Trust only at Westinghouse facilities for limited evaluation tasks assigned by EPRI. Any use, disclosure to unauthorized persons, or copying of this document or parts thereof is prohibited. Neither this document nor any excerpts therefrom are to be removed from Westinghouse facilites.

EPRI CONFIDENTIALITY / OBLIGATION CATEGORIES CATEGORY "A"-(See Delivered Data) Consists of CONTRACTOR Foreground Data that is contained in an issued reported.

CATEGORY "B" -(See Delivered Data) Consists of CONTRACTOR Foreground Data that is not contained in an issued report, except for computer programs.

CATEGORY *C"- Consists of CONTRACTOR Background Data except for computer programs.

CATEGORY "D"- Consists of computer programs developed in the course of performing the Work.

CATEGORY *E"- Consists of computer programs developed pnor to the Effective Date or after the Effectve Date but outside the scope of the Work.

CATEGORY *F"- Consists of administrative plans and administrative reports.

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1 AP600 Design Control Document AP600 List of Effective Pages Volume 2 Page Revision Page Revision Page Revision Page Revision 1.9-23 original 1.9-66 original 1.9-109 (2/99) 1.9-152 original 1,9-24 (2/99) 1.9-67 (2/99) 1.9-110 original 1.9-153 original 1.9 25 (2/99) 1.9-68 original 1.9-111 original 1.9-154 original 1.9-26 (2/99) 1.9-69 original 1.9-112 original 1.9 155 original 1.9-27 (2/99) 1.9-70 original 1.9-113 (2/99) 1.9-156 original 1.9-28 (2/99) 1.9-71 original 1.9-114 (2/99) 1.9-157 original 1.9-29 (2/99) 1.9-72 original 1.9-115 original 1.9-158 original 1.9-30 (2/99) 1.9-73 original 1.9-116 original 1.9-159 original 1.9-31 (2/99) 1.9-74 original 1.9-117 original 1.9-160 original 1.9-32 original 1.9-75 original 1.9-118 original 1.9-161 original 1.9-33 original 1.9-76 original 1.9-119 original 1.9-162 original 1.9-34 original 1.9-77 original 1.9-120 original 1.9 163 original 1.9-35 original 1.9-78 (2/99) 1.9-121 original 1.9-164 original 1.9-36 original 1.9-79 (2/99) 1.9-122 original 1.9-165 original 1.9-37 original 1.9-80 (2/99) 1.9-123 original 1.9-166 (2/99) 1.9-38 original 1.9-81 (2/99) 1.9-124 (2/99) 1.9-167 original 1.9-39 original 1.9-82 (2/99) 1.9-125 original 1.9-168 original 1.9-40 original 1.9-83 (2/99) 1.9-126 original 1.9-169 original 1.9-41 original 1.9-84 (2/99) 1.9-127 original 1.9-170 original 1.9-42 original 1.9-85 (2/99) 1.9-128 original 1.9-171 original 1.9-43 original 1.9-86 (2/99) 1.9-129 original 1.9-172 original 1.9-44 original 1.9-87 (2/99) 1.9-130 original 1.9-173 original 1.9-45 original 1.9-88 (2/99) 1.9-131 original 1.9-174 original 1.9-46 original 1.9-89 original 1.9-132 original 1.9-175 original 1.9-47 original 1.9-90 original 1.9-133 original 1.9-176 original 1.9-48 original 1.9-91 original 1.9-134 original 1.9-177 original 1.9-49 original 1.9-92 original 1.9-135 original 1.9-178 original 1.9-50 original 1.9-93 original 1.9-136 original 1.9-179 original 1.9-51 original 1.9-94 (2/99) 1.9-137 original 1.9-180 original 1.9-52 original 1.9-95 (2/99) 1.9-138 original 1.9-181 original 1.9-53 original 1.9-96 (2/99) 1.9-139 original 1.9-182 original 1.9-54 original 1.9-97 (2/99) 1.9-140 original 1.9-183 original 1.9-55 original 1.9-98 (2/99) 1.9-141 original 1.9-184 original 1.9-56 original 1.9-99 original 1.9 142 original 1.9-185 original 1.9-57 original 1.9-100 original 1.9-143 original 1.9-58 original 1.9-101 original 1.9-144 original I A-1 (2/99) 1.9-59 original 1.9-102 original 1.9-145 original l A-2 (2/99) 1.9-60 original 1.9-103 original 1.9-146 original 1 A-3 original 1.9-61 original 1.9-104 (2/99) 1.9-147 original l A-4 original 1.9-62 original 1.9-105 (2/99) 1.9-148 original l A-5 criginal 1.9-63 original 1.9-106 (2/99) 1.9-149 original I A-6 original 1.9-64 original 1.9-107 (2/99) 1.9-150 original 1 A-7 original 1.9-65 original 1.9-108 original 1.9-151 original IA 8 original O

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AP600 Design Control Document AP600 List of Effective Pages Volume 3 Page Revision Page Revision Page Revision Page Revision 2-i original 2A-10 original 2A-54 original 2B-8 original 2-ii original 2A-11 original 2A-55 original 2B-9 original 2-iii original 2A-12 original 2A-56 original 2B-10 original 2-iv original 2A-13 original 2A-57 original 2B-11 original 2-v original 2A-14 original 2A-$8 original 2B-12 original 2-vi original 2A-15 , original 2A-59 original 28-13 original 2-vii original 2A-16 original 2A-60 original 2B-14 original 2A-17 original 2A-61 original 2B-15 original 2-1 original 2A-18 original 2A-62 original 2B-16 original 2-2 original 2A-19 original 2A-63 original 2B-17 original 2-3 original 2A-20 original 2A-64 original 2B-18 original 2-4 original 2A-21 original 2A-65 original 2B-19 original 2-5 original 2A-22 original 2A-66 original 2B-20 original 2-6 original 2A-23 original 2A-67 original 2B-21 original 2-7 original 2A-24 original 2A-68 original 2B-22 ori ginal 2-8 original 2A-25 original 2A-69 original 2B-23 original 2-9 original 2A-26 original 2A-70 original 2B-24 original 2-10 original 2A-27 original 2A-71 original 2B-25 original 2-11 original 2A-28 original 2A-72 original 2B-26 original 2-12 original 2A-29 original 2A-73 original 2B-27 original 2-13 original 2A-30 original 2A-74 original 2B-28 original 2-14 original 2A-31 original 2A-75 original 2B-29 original 2-15 original 2A-32 original 2A-76 original 2B-30 original 2-16 original 2A-33 original 2A-77 originr1 2B-31 original 2-17 original 2A-34 original 2A-78 original 2B-32 original 2-18 original 2A-35 original 2A-79 original 2B-33 original 2-19 original 2A-36 original 2A-80 criginal 2B-34 original 2-20 original 2A-37 original 2A-81 original 2B-35 original 2-21 (2/99) 2A-38 original 2A-82 original 2B-36 original 2-22 original 2A-39 original 2A-83 original 2B-37 original 2-23 original 2A-40 original 2A-84 original 2B-38 original 2-24 original 2A-41 original 2A-85 original 2B-39 original 2-25 original 2A-42 original 2A-86 original 2B-40 original 2-26 original 2A-43 original 2A-87 original 2B-41 original 2A-44 original 2A-88 original 2B-42 original 2A-1 original 2A-45 original 2A-89 original 2B-43 original 2A-2 original 2A-46 original 2B-44 original 2A-3 original 2A-47 original 2B-1 original 2B-45 original 2A-4 original 2A-48 original 2B-2 original 2B-46 original 2A-5 original 2A-49 original 2B-3 original 2B-47 original 2A-6 original 2A-50 original 2B-4 original 2B-48 original 2A-7 original 2A-51 original 2B-5 original 2B-49 original 2A-8 original 2A-52 original 2B-6 original 2B-50 original 2A-9 original 2A-53 original 2B-7 original 2B-51 original O

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V List of Effective Pages (9/99) Page 9

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Q AP600 List of Effective Pages Volume 4

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V List of Effectiw Jages (9/99) Page 11 1

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AP600 Design Control Document AP600 List of Effective Pages Volume S Page Revision Page Revision Page Revision Page Revision 4.4-9 original 4.6-1 original 5.2-11 original 5.3-13 original 4.4-10 original 4.6-2 original 5.2-12 original 5.3-14 original 4.4-11 original 4.6-3 original 5.2-13 original 5.3-15 original 4.4-12 original 4.6-4 original 5.2-14 original 5.3-16 original 4.4-13 original 5.2-15 original 5.3-17 original 4.4-14 original 5-i original 5.2-16 original 5.3-18 original 4.4-15 origir.al 5-ii original 5.2-17 original 5.3-19 original 4.4-16 original 5-iii original 5.2-18 original 5.3-20 original 4.4-17 original 5-iv original 5.2-19 original 5.3-21 original 4.4-18 original 5-v original 5.2-20 original 5.3-22 original 4.4-19 original 5-vi original 5.2-21 original 5.3-23 original 4.4-20 original 5-vii original 5.2-22 original 5.3-24 original 4.4-21 original 5.2-23 original 5.3-25 original 4.4-22 original 5.1-1 original 5.2-24 original 5.3-26 original 4.4-23 original 5.1-2 original 5.2-25 original 5.3-27 original 4.4-24 original 5.1-3 original 5.2-26 original 5.3-28 original 4.4-25 original 5.1.

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b List of Effective Pages (9/99) Page 21

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List of Effective Pages (A99) Page 27

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AP600 Design Control Document p-1 %.)

Table 2.2.3 4 (cont.)

Iaspections, Tests, Analyses, and Acceptance Criteria Design Conanaltenent Inspections, Tests, Analy3es Acceptance Criteria vii) Inspection of the as-built vii) Plates located above cach components will be conducted containment recirculation screen i' I for plates located above the are no more than 1 ft above the containment recirculation top of the screen and extend out at I screens. least 10 ft perpendicular to and at i

least 7 ft to the side of the trash 1- '

rack portion of the screen.

viii) Inspections of the IRWST viii) The screen surface area and containment recirculation (width x height) of each screen is screens will be conducted. 2 70 ft'. The bottom of the containment recirculation screens is ,

2 2 ft above the loop compartment i floor.

ix) Inspections will be ix) The type ofinsulation used on conducted of the insulation used these lines and equipment is a inside the containment on ASME metal reflective type or a suitable Class I lines and on the reactor equivalent.

vessel, reactor coolant pumps, pressurizer and steam generators.

x) Inspections will be conducted x) A report exists and concludes of the as-built nonsafety-related that the coatings used on these coatings or of plant records of surfaces has a dry film density of '

the nonsafety-related coatings 2100 lb/ft'.

used inside containment on walls, floors, ceilings, structural steel which is part of the building structure and on the polar crane.

xi) Inspection of the as-built xi) The CMT inlei diffuser has a CMT inlet diffuser will be flow area 2165 in'.

conducted.

xii) Inspections will be xii) The centerline of each upper conducted of the CMT tevel level tap line at the tee for each sensors (PSX-11 A/B/D/C, - level sensor is located I" i 1" 12A/B/C/D, - 13A/B/C/D, - below the centeiline of the upper 14A/B/C/D) upper level tap level tap connection to the CMT.

lines.

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Tab! a 2.2.3-4 (cont.)

Inspections, Te.,ts, f.nalyses, and Acceptance Criteria Design Commitment Inspections, Tests, Analyses Acceptance Criteria xiii) Inspections will be xiii) A report exists and concludes conducted of the materials used that these surfaces are stainless in the vicinity of the steel or that coatings used on these containment recirculation screens surfaces are qualified to remain or analysis, type tests, or a attached to these surfaces during combination of analysis and type design basis events.

tests will be performed on coatings used on surfaces in the vicinity of the containment recirculation screens. In the vicinity of the containment recirculation screens includes surfaces located above the bottom of the recirculation screens up to and including the bottom surface of the plate discussed in Table 2.2.3-4, I item 8.c.vii, out at least 10 feet I perpendicular to and at least 7 I feet to the side of the trash rack I portion of the screen.

8.d) The PXS provides pH Inspections of the pH adjustmen* Two pH adjustment baskets exist, adjustment of water flooding the baskets will be conducted. each with a calculated volume containment following design 2107 ft'.

basis accidents.

The pH baskets are located below plant elevation 107 ft,2 in.

9.a) The PXS provides a i) A flow test and analysis for i) The calculated now resistance function to cool the outside of the each IRWST drain line to the for each IRWST drain line reactor vessel during a severe containment will be conducted. between the IRWST and the accident. The test is initiated by opening containment is s 1.38 x 10 5 isolation valves in each line, ft/gpm .

Test fixtures may be used to simulate squib valves.

ii) Inspections of the as-built ii) The combined total flow area reactor vessel insulation will be of the water inlets is not less than performed. 6 ft 2. The combined total Cow area of the steam outlet (s) is not less than 7.5 ft 2, Tier 7 Material (9/99) Page 2.2.3-22 O

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AP600 Design Control Document

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(2)(vii) Plant Radiation Shielding (NUREG 0737 Item II.B.2)

" Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID-14844 source term radioactive materials, and design as necessary to permit viequate access to important areas and to protect safety equipment fmm the radiation environment."

AP600 Response:

Post accident radiation sources, used in the shield design and assessment of post-accident access to vital areas, are addressed in subsection 12.2.1.3. The post-LOCA instantaneous and

' integrcted source strengths as a function of time are also included as Tables 12.2-20 a.-

12.2-21, respectively. The sources are based on the core activity release model from NUREG-1465, which supersedes the TID-14844 source term assumptions as reDected in Regulatory Guide 1.4. The source term assumptions and differences from the guidance of Regulatory Guide 1.4 are described in subsection 15.6.5.3.

Vital areas for post-accident personnel access are addressed in Section 12.3, including radiation zone maps that show projected dose rates in these areas and access routes for the various post-accident actions in vital areas. Time estimates have been made for ingress, egress, and performance of actions at the vital area locations and have been used in

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demonstrating total individual radiation doses are limited to less than 5 rem and that s

item II.B.2 of NUREG-0737 and GDC 19 requirements are met.

Environmental quali6 cation of safety-related equipment is addressed in Section 3.11. The determination of the radiation environments during postulated accident situations considers the activity releau model based on NUREG-1465, which supersedes the source te m dennition of Parts 1 and 4 of Item II.B.2 of NUREG-0737.

As noted in subsection 12.2.3, the Combined License applicant will address any additional contaired radiation sources not identified in 12.2.1. Thus, appropriate source terms have been identined and used in establishing that the requirements of item II.B.2 of NUREG-0737 and GDC 19 are met and the issues are resolved.

(2)(viii) Post Acuaent Sampling (NUREG-0737 Item II.B.3)

" Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID-14844 source term radioactive materials without I

radiation exposures to any individual exceeding 5 rem to the who:e-body or 50 rem to the extremities. Materials to be analyzed and quantined incicJe certain radionuclides that are indicators of the degree of core damage (e.g. noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and f boron concentrations."

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Tier 2 2*aterial-Introduction and GeneralDescription of Plant (9/99) Page 1.913

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AP600 Response:

The AP600 primary sampling system is designed to perform the functions required by NUREG-0737 and Regulatory Guide 1.97 for post-accident sampling. The primary sampling system provides the capability to obtain post-accident samples from the reactor coolant system, the containment sump, and the containment atmosphere.

Radiation exposures to any individual obtaining and analyzing samples do not exceed five rem I to the whole-body or 50 rem to the extremities, when operating procedures are followed. See Chapter 9 for additional information pertaining to the primary sampling system.

(2)(ix) Ilydrogen Control (NUREG-0660 Item II.B.8)

" Provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal-water reaction. Preliminary design information on the tentatively preferred system option of those being evaluated in paragraph (1)(xii) of this section (50.34)is sufficient at the construction permit stage. The hydrogen control system and associated symms shall provide, with reasonable assurance, that:

(A) Uniformly distributed hydrogen concentrations in the containment do not exceed 10 percent during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100 percent fuel-clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.

(B) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features. -

(C) Equipment necessary for achieving and maintaining safe shu'down of the plant and maintaining containment integriiy will perform its safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the equivalent of a 100 percent fuel-clad metal-water reaction includ.ng the environmental conditions created by activation of the hydrogen control system.

(D) If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation."

i AP600 Response:

f l See the response provided for issue (1)(xii).

O Tier 2 Material-Introduction and General Description of Plant (9/99) Page 1.9-14 l

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(s width of the plate. The evaluations are summarized in Table 3.8.3-7. Design loads and load combinations are shown on sheet 1. Sheet 2 shows the ratio of the desig:t stresses to the allowable stresses. When thermal effects result in stresses above yield, the evaluation is in accordance with the supplemental criteria}* as described in subsection 3.83.53.4.

3.83.5.83 Column Supporting Operating Floor (This subsection summarizes the design of the most heavily loaded column in the containment internal structures. The column extends from elevation 107'-2" to the underside of the operatingfloor at elevation 135'-3'. In addition to supporting the operatingfloor, . also supports a steel gratingfloor as elevation 118'-0" The load combinations in Table 3.8.41 were used to assess the adequacy of the column. For load combination 1 in the table, the interaction factor due to biaxial bending and axial load is 0.38. For load combination 6 without thermalloads, the interactionfactor is 0.42 and with thermalloads the interactionfactor is 0.61. Since the interactionfactors are less than 1, the column is adequatefor all the applied loads.J*

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3.83.6 Materials, Quality Control, and Special Construction Techniques Subsection 3.8.4.6 describes the materials and quality control program used in the construction e

of the containment internal structures. The structural steel modules are constructed using A36 O

b plates and shapes. Nitronic 33 (American Society for Testing and Materials 240, designation S24000, Type XM-29) stainless steel plates are used on the surfaces of the modules in contact with water during normal operation or refueling. The structural wall and floor modules are fabricated and erected in accordance with AISC-N690. Loads during fabrication and erection due to handling and shipping are considered as normal loads as described in subsection 3.8.43.1.1. Packaging, shipping, receiving, t.torage and handling of structural modules are in accordance with NQA-2, Part 2.2 (formerly ANSI /ASME N45.2.2 as speciGed in AISC N690).

3.8.3.6.1 Fabrication, Erection, and Construction of Structural Modules Modular construction techniques are used extensively in the containment internal structures (Figure 3.83-1). Subassemblies, sized for commercial rail shipment, are assembled offsite and transported to' the site. Onsite fabrication consists of combining the subassemblies in structural modules, which are then installed in the plant. A typical modular construction technique is described in the following paragraphs for Module M1, which is the main structural module in the containment internal structures.

The M1 module is a multicompanmented structure which, in its fm' al form, comprises the central walls of the containment internal structures. The venical walls of the module house the refueling cavity, the reactor vessel compantaent, and the two steam generator I

compartments. The module (Figure 3.83-14) is in the form of a "T" and is approximately 89 I feet long,95 feet wide and 77 feet 6 inches high. The module is assembled from about 40

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T' AP600 Design Control Document O

prefabricated wall sections called structural submodules (Figure 3.83-15). The submodules are designed for railroad transport from the fabricator's shop to the plant site with sizes up to 12 feet by 12 feet by 80 feet long, weighing up to 30 tons. A typical submodule weighs between 9 and 11 tons. The submodules are assembled outside the nu: lear island with full penetration welds between the faceplates of adjacent subunits. The completed M1 module is lifted to its final location within the containment vessel by the heavy lift construction crane.

Following placement of the M1 module within the containment building, the hollow wall structures are filled with concrete, forming a portion of the structural walls of the containment internal structures.

Tolerances for fabrication, assembly and erection of the structural modules conform to the requirements of section 4 of ACI-l17, sections 33 and 3.4 of AWS DI.1, and sections Ql.23 and Ql.25 of AISC-N690.

3.83.6.2 Nondestructive Examination Nondestmetive examination of the submodules and module is performed according to AISC-N690 and AWS D 1.1, Welds are visually examined for 100 percent of their Imgth.

Full penetration welds are inspected by ultrasonic or radiograpide examination for 10 percent of their length. Partial penetration welds are inspected by magnetic particle or liquid penetrant examination for 10 percent of their length.

3.83.63 Concrete Placement After installation of the M1 module in the containment, the hollow walls are filled with concrete. Concrete is placed in each wall continuously from bottom to top. The concrete is placed through multiple delivery trunks located along the top of the wall. It is placed in incremental layers with the placement rate based on the pressure of the wet concrete and its setting time. During concrete. placement, workers and inspectors have access to the inside of the modules. The arrangement of the module internal tmsses provides communication to aid in the free flow of concrete and movement of personnel.

3.83.7 In Service Testing and Inspection Requirements There are no in-service testing or inspection requirements for the containment intemal structures.

3.8.3.8 Construction Inspection Construction inspection is conduct:d to verify the concrete wall thickness and the surface plate thickness. Inspections will be measured at applicable sections er.:luding designed openings or penetrations. Inspections will confirm that each section provides the minimum required steel and concrete thicknesses as shown it. Table 3.83-3. The minimum required steel and concrete thicknesses represent thJ minimum values to meet the design basis loads.

Table 3.83-3 also indicates the steel plate thickness provided which may exceed the minimum required value for the following reasons:

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l AP600 Design Control Document d(~ : of reactor coolant, gap and early in-vessel releases. Funher, based on industry test data and catalyst poison literature, the hydrogen recombination subsystem performance for a DBA will be evaluated with the reduction in performance anticipated as a result of the effects of poisons and inhibitors as determined in Reference 28c The fractions of core inventory released through reactor coolant, gap and early in-vessel stages of core damage event are identified in l NUREG 1465. Reference 28 considers the releases of NUREG 1465 and addresses the
. potential effects of poisons and inhibitors on the performance of the PAR's. The approach in Reference 28 is to compile existing information and data as a basis for establishing a generic bounding value of a deactivation reduction factor for design and qualification of ; PAR's. The approach combines qualitative information based on established chemical and physical principles with quantitative information from testing of catalysts systems subjected to a wide range ofinhibitors and poisons. The sources of test data include (1) tests on PAR's conducted by two suppliers over the past several years, (2) tests on the same two types of PAR's conducted recently in France by EPRI/EdF/CEA and (3) tests on catalyst pellet-bed filters conducted in a laboratory. about 25 years ago and additional testing described in Reference 28. De report concludes that "Even if the accident were to progress to beyond a DBA to substantial in-vessel damage, PAR recombination capacity would be reduced by no - more than 25%.J" To illustrate the margin available and the tolerance to catalytic poisons and inhibitors, Figure 6.2.4-2 demonstrates the effects of the presence of elevated concentrations of poisons and inhibitors in containment. The curve demonstrates the effects of the conservative j depletion rate discussed above in combination with a 25 percent penalty due to poisons and U l inhibitors. The curve remains below 2.2 volume percent hydrogen concentration. The margin of difference between the regulatory limit of 4 percent and the projected concentration . may be considered to irpresent the maximum potential gradient between global hydrogen concentration and an isolated postulated hypothetical volume of containment atmosphere.
He environments in which safety-related components of the hydrogen control system are _ designed and qualified to function are discussed in Section 3.1I and subsection 6.2.4.1.2. The I pressure, temperature, and chemical environment conditions for which components are j designed to function have been based on analysis of the design basis event and the systems l response .with additional considerations identified in Reference 27. The radiation environments have, in contrast, been the result of a deterministic application of the accident j source term. As specified in NUREG 1465, to determine the accident source term for j regulatory purposes, the staff examined a range of severe accidents that have been analyzed ; for light water reactors.. The environmental qualification guidance and practice is l conservatively based on the effects of radiation due to a severe accident source term. The passive autocatalytic recombiner testing and reporting of test data, conducted under the NIS quality assurance program has demonstrated proof of principle as appropriate for design Lcenification.~ An evaluation and summary of the quality assurance program for the Battelle tests is provided in Reference 21.' Procurement will be in accordance with the COL applicants l
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Tier 2 Material - Engineered Safety Features (9/99) Page 6.2-47
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AP600 Design Control Document O; 6.2.4.2.3 Ilydrogen Ignition Subsystem The hydrogen ignition subsystem is provided to address the possibility of an event that results in a rapid production of large amounts of hydrogen such that the rate of production exceeds the capacity of the recombiners. Consequently, the containment hydrogen concentration will exceed the flammability limits. This massive hydrogen production is postulated to occur as the result of a degraded core or core melt accident (severe accident scenario)in which up to , 100 percent of the zirconium fuel cladding reacts with steam to produce hydrogen. l The hydrogen ignition subsystem consists of 64 hydrogen igniters strategically distributed throughout the containment. Since the igniters are incorporated in the design to address a low-probability severe accident, the hydrogen ignition system is not Class IE. Although not class 1E, the igniter coverage, distribution and power supply has been designed to minimize the potential loss of igniter protection globally for containment and locally for individual companments. The igniters have been divided into two power groups. Power to each group i will be normally provided by offsite power, however should offsite power be unavailable, then I each of the power groups is powered by one of the onsite non-essential diesels and finally should the diesels fail to provide power then approximately 4 hours of igniter operation is supponed by the non-Class IE batteries for each group. Assignment of igniters to each group is based on providing coverage for each companment or area by at least one igniter from each group. The locations of the igniters are based on evaluation of hydrogen transport in the containment and the hydrogen combustion characteristics. Locations include compartmented areas in the containment and various locations throughout the free volume, including the upper dome. For enclosed areas of the containment at least two igniters are installed. The separation between igniter locations is selected to prevent the velocity of a flame front initiated by one igniter from becoming significant before being extinguished by a similar flame front propagating from another igniter. De number of hydrogen igniters and their locations are selected considering the behavior of hydrogen in the containment during severe accidents. The likely hydrogen transport paths in the containment and hydrogen burn physics are th: two imponant aspects influencing the choice of igniter location. The primary objective of installing an igniter system is to promote hydrogen burning at a low concentration and, to the extent possible, to burn hydrogen more or less continuously so that the hydrogen concentration does not build up in the containment. To achieve this goal, igniters are placed in the major regions of the containment where hydrogen may be released, through which it may flow, or where it may accumulate. The criteria utilized in the evaluation and the application of the criteria to specific compartments is provided in Table 6.2.4-6. The location of igniters throughout containment is provided in Figures 6.2.4-5 through 6.2.4-11. The location of igniters is also summarized in Table 6.2.4-7 identifying subcompartment/ regions and which igniters by power group provide protection. The locations identified are considered approximations ( 2.5 feet) with the final locations governed by the installation details. O Tier 2 Material- Engineered Safety Features Page 6.2-48 l t
o AP600 Design Control Document o O I The results of Westinghouse and Oak' Ridge National Laboratory studies indicate maximum hydrogen yields of 0.44 molecules per 100 eV for core radiolysis and 0.3 molecules per 100 eV for solution radiolysis. The results of these studies are published in References 10,
- 11, and 12. ~ The analysis performed for the AP600 assumes a hydrogen yield of 0.5 molecules per 100 eV for both the core and the solution radiolysis cases. This value is conservative relative to the referenced studies and is consistent with the guidance of Regulatory Guide 1.7.
t In a design basis loss of coolant accident there is expected to be no damage to the core and .
. thus no release of activity from -the core to the sump solution. The source term used for determining radiolysis production of hydrogen is conservatively based on guidance of Regulatory Guide 1.7 which states that 100 percent of noble gases,50 percent ofiodines, and . I percent of other nuclides are ' assumed to be released from the core even though it is inconsistent with the limited amount of fuel cladding reaction that is determined to take place.
Appendix 15A provides the core fission product inventory at shutdown. Table 6.2.4-4 contains a summary of the assumptions used in the analysis of hydrogen produced from radiolysis. Production rate of hydrogen as a function of time is shown graphically in Figure 6.2.4-3 and the production of hydrogen is shown in Figure 6.2.4-4. 6.2.4.3.1.3~ Corrosion of Metals In the environment that would exist inside the containment following a postulated loss of coolant accident, aluminum and zinc corrode to form hydrogen gas. Table 6.2.4-4 lists the inventory of aluminum and zine inside the containment. Aluminum corrosion may be described by the overall reaction: 2 Al + 3 Hg - +Al2O3 + 3 H2 About 21.4 standard cubic feet of hydrogen gas is produced for each pound of aluminum corroded. The corrosion of zinc is described by the folbwing reaction: Zn + 2 H2O --+Zn(OH)2 +H 2 i'
. About 5.9 standard cubic feet of hydrogen gas is produced for each pound of zine corroded. ' The corrosion . rates for both aluminum and zinc are dependent on the post accident - temperature and pH conditions that the materials are subjected to. Table 6.2.4-5 provides the time temperature cycle considered in the analysis of aluminum and zine corrosion and also the corrosion rates for the metals at these temperatures. -y: ^ Tier 2 Material- Engineered Sehrty Features _ Page 6.2-51
AP600 Design Control Document O Production rate of hydrogen as a function of time is shown graphically in Figure 6.2.4-3 and the production of hydrogen is shown in Figure 6.2.4-4. 6.2.4.3.1.4 Initial Reactor Coolant flydrogen Inventory During normal operation of the plant, hydrogen is dissolved in the reactor coolant and is also contained in the pressurizer vapor space. Following a loss of coolant accident, this hydrogen ' is assumed to be immediately released to the containment atmosphere. Table 6.2.4-4 lists the assumptions used for determining the amount of hydrogen from this source. The total hydrogen released to the containment as a result of this source is i171 standard cubic feet. 6.2.4.3.2 flydrogen Mixing The AP600 is designed to prevent the accumulation of hydrogen in compartments. If there is the possibility of accumulation in compartments, venting is provided to allow the hydrogen to escape to the larger containment volume. Mixing of the containment air mass is accomplished through natural processes as a result of the passive cooling of the containment (see subsection 6.2.2) that induces a recirculating air flow in the containment. The release rate for a design basis accident is sufficiently slow that mixing is effectively assured. I Additional details are provided in subsection 6.2.4.1.1 and Appendix 6A. 6.2.4.3.3 Ilydrogen Recombination l Assuming no hydrogen removal, the concentration of hydrogen in the containment atmosphere increases with time as shown in Figure 6.2.4-1. The curve shows that the flammability limit (g of 4 volume percent is not reached until after 12 days. Hydrogen recombination begins prior to reaching this limit. The passive autocatalytic r-combiners are brought into service by the presence of the reactants. 'Ihe available passive autocatalytic recombiner test data as discussed in Reference 19 supports passive autocatalytic recombiner startup within 7 hours 1 of reaching 1 volume percent hydrogen concentration in containment. Subsequent to passive l autocatJytic recombiner startup the conservative lower bound equation for depletion rates provided in Reference 19 has been used to predict containment concentrations considering only the PARS between the 150 and 175 foot elevations. Figure 6.2.4-1 shows the impact of operation of one of the two recombiners on containment hydrogen concentration. The hydrogen concentration never exceeds 1.5 percent which indicates ample margin in the hydrogen recombiner capacity. Figure 6.2.4-2 evaluates the containment hydrogen concentration using the same lower bound equation for a single PAR's depletion rate but with additional conservatism to account for catalyst poisons or evaluation of margin. The curve identified as " Worst Case PAR Depletion Rate with Catalyst Poisons" has been reduced as a result of the effects of poisons and inhibitors which could be released to containment following a core melt scenario progressing through reactor coolant, gap, and early in-vessel releases. In accordance with Reference 28 the lower bound depletion rate has been reduced by 25 percent to conservatively account for the effects of potential poisons and inhibitors. A further demonstration of the passive autocatalytic recombiner's available capacity margin is provided by calculation of containment concentrations with artificially reduced depletion rates. Figure 6.2.4-2 provides the impact of one of two available passive autocatalytic recombiners operating at 20,10 and I percent of the conservative lower bound capacity. The O Tier 2 Material- Engineered Safety Features (9/99) Page 6.2-52
AP6% Design Control Document Table 6.2.4-5 POST-ACCIDENT CONTAINMENT TEMPERATURE ) AND ASSOCIATED CORROSION RATES FOR ALUMINUM AND ZINC Interval Temperature Al Corrosion Zn Corrosion (sec) ('F) (Ib/ft2-hr) (ib/ft2-hr) l 0 2,500 267 0.024 0.00030 l 2,500-5,000 260 l 0.019 0.00027 ' l 5,000-10,000 245 0.012 0.00021 l 10,000-20,000 230 0.0070 0.00016 l 20,000-60,000 228 0.0065 0.00015 1 60,000-80,000 220 0.0049 0.00013 i l 80,000-108,000 218 0.0045 0.00013 l 108,000-173,000 225 0.0058 0.00034 l >173,000 220 0.0049 0.00013 .r (,- i O Tier 2 Material Engineered Safety Features (9/99) Page 6.2-129 l J
r AP600 Design Control Document O Table 6.2.4-6 (Sheet I of 3) IGNITER LOCATION i Criteria
A sufficient number of igniters are placed in the major transport paths (including dominant natural circulation pathways) of hydrogen so that hydrogen can be burned continuously close to the release point. This prevents hydrogen from preferentially accumulating in a certain region of the containment.

Igniters (minimum of 2) are located in major regions or compartments where h)drogen may be released, through which it may flow, or where it may accumulate.

It is preferable to ignite a hydrogen-air mixture at the bottom so that upward flame propagation can be promoted at lean hydrogen concentrations. Igniters within each subcompartment are located in the vicinity of, and above, the highest potential release location within the subcompartment.

In compartments with relatively small openings in the ceiling, the potential may exist for the hydrogen-air mixture to rise and to collect near the ceiling. Therefore, one or more igniters are placed near the ceiling of such compartments. Igniter coverage is provided within the upper 10 percent of the vertical height subcompartments or 10 feet from the ceiling whichever is less. In cases where the highest potential release point is low in the compartment, both this and the previous criteria are considered.

To the extent possible, igniters are placed away from walls and other large surfaces so that a flame front created by ignition at the bottom of a compartment can travel unimpeded up to the top.

A sufficient number of igniters are installed in long, narrow compartments (corridors) so that the flame fronts created by the igniters need to travel only a limited distance before they merge. This limits the potential for significant flame acceleration.

Igniter coverge are provided to contol combustion in areas where oxygen rich air may enter into an inerted region with combustible hydrogen levels during an accident scenario, a Igniters are located above the flood level, if possible. Those which may be flooded have redundant fuses to protect the power supply.

In locations where the potential hydrogen release location can be defined, i.e. above the IRWST spargers, at IRWST vents, etc igniter coverage is provided as close to the source as feasible.

Provisions for installation, maintenance, and testing is be considered.
O Tier 2 Material- Engineered Safety Features Page 6.2-130
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f 1 V an identical structural configuration to the inlet channel head. Both channel head tubesheets are similar to the steam generator tubesheets and they have manways for inspection and l maintenance access. 1 i The passive residual heat removal heat exchanger is designed to remove sufficient heat so that l its operation, in conjunction with available inventory in the steam generators, provide reactor ! coolant system cooling and prevents water relief through the pressurizer safety valves during loss of main feedwater or main feedline break events. Passive residual heat removal heat exchanger flow and inlet and outlet line temperatures are { monitored by indicators and alarms. The operator can take action, as required, to meet the technical specification requirements or follow emergency operating procedures for control of the passive residual heat removal heat exchanger operation. 63.2.2.6 Depressurization Spargers Two reactor coolant depressurization spargers are prov;ded. Each one is connected to an i automatic depressurization system discharge header (shared by three automatic depressurization system stages) and submerged in the in-containment refueling water storage tank. Each sparger has four branch arms inclined downward. The connection of the sparger branch arms to the sparger hub are submerged below the in-containment refueling water storage tank overflow level by 511.5 feet. The component data for the spargers is shown in
/G Table 6.3-4. he spargers are AP600 Equipment Class C and are designed to meet seismic O Category I requirements.
The spargers perform a nonsafety-related function -- minimizing plant cleanup and recovery actions following automatic depressurize. tion. They are designed to distribute steam into the ; in-containment refueling water storage tank, thereby promoting more effective steam condensxtion. The first three stages of automatic depressurization system valves discharge through the spargers and are designed to pass sufficient depressurization venting flow, with an acceptable pressure drop, to support the' depressurization system performance requirements. The installation of the spargers prevents undesirable and/or excessive dynamic loads on the !n- , containment refueling water storage tank and other structures. ! Each sparger is sized to discharge at a f'ow rate that supports automatic depressurization system performance, which in tum, allow cdequate passive core cooling system injection. 6.3.2.2.7 ' IRWST and Containment Recirculation Screens The passive core cooling systems has two different sets of screens that are used following a LOCA;IRWST screens and containment recirculation screens. These screens prevent debris from entering the reactor and blocking core cooling passages during a LOCA. These screens are designed to comply with applicable licensing regulations including:
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(T) Yler 2 Material - Engineered Safety Features Page 6.3-17
AP600 Design Control Document O GDC 35 of 10 CFR 50 Appendix A Regulatory Guide 1.82 NUREG-0897 The operation of the passive core cooling system following a LOCA is described in subsection 6.3.2.1.3. Proper screen design, plant layout, and other factors prevent clogging of these screens by debris during accident operations. 6.3.2.2.7.1 General Screen Design Criteria
1) Screens are designed to Regulatory Guide 1.82, including:
Redundant screens are provided for each function Separate locations are used for redundant screens Screens are located well below containment floodup level. Each screen has a coarse and a fine screen, and a debris curb Floors slope away from screens (not required for AP600) Drains do not impinge on screens Screens can withstand accident loads and credible missiles Screens have conservative flow areas to account for plugging. Operation of the non-safety-related normal residual heat removal pumps with suction from the IRWST and the containment recirculation lines is considered in sizing screens. System and screen performance are evaluated Screens have solid top cover. Containment recirculation screens have protective I plates that are located no more than 1 feet above the top of the screens and i extend at least 10 feet in front and 7 feet to the side of the screens, The plate I dimensions are relative to the ponion of the screens where water flows through l the trash rack. Screens are seismically qualified Screen openings are sized to prevent blockage of core cooling Screens are designed for adequate pump performance. AP600 has no safety-relateo pumps. Corrosion resistant materials are used for screens O Tier 2 Material- Eq!neered Safety Features (9/99) Page 6.318 i t L_ 2
E i o ' i I L i l AP600 Design Control Document m i U f 6.3.2.2.7.3 Containment Recirculation Screens l The containment recirculation screens are oriented vcitically along walls above the loop companment floor (elevation 83 feet). Figure 6.3-8 shows a plan view and Figure 6.3-9 shows a section view of these screens. Two separate screens arc provided as shown in Figure 6.3-3. The loop compartment floor elevation is significantly above (11.5 feet) the lowest level in the containment, the reactor vessel cavity. The t>ottom of the recirculation screen is .wo foot above the floor, providing a curb function. During a LOCA, the reactor coolant system blowdown will tend to carry debris created by l the accident (pipe whip / jets) into.the cavity under the reactor vessel which is located away i from and below the containment recirculation s..eens. As the accumulators, core inakeup ! tanks and IRWST inject, the containment water level will slowly rise up to the 108 foot ! elevation over at least 5 hours. The containment recirculation line opens when the water level ! in the IRWST drops to a low level setpoint a few feet above the final containment floodup ! level. When the recirculation lines initiaUy open, the water level in the IRWST is higher than l the containment water level and water flows from tb IRWST backwards through the containment recirculation screen. This back flow tends to flush debris located close to the recirculation screens away from the screens. The water level in the containment when recirculation begins is well above (~ 10 feet) the top of the recirculation screens. During the long containment floodup time (>5 hours), floating _[ ( d4ris does not move toward the screens and heavy materials settle to the floors of the loop compartments or the reactor vessel cavity. During recirculation operation the containment water level will not change significantly nor will it drop below the top of the screen The amount of debris that may exist following an accident is limited. Reflective insulation is used to preclude fibrous debris that can be generated by a loss of coolant accideat and be postulated to reach the screens during recirculation. The nonsafety-related coatings used in the containment are designed to withstand the post accident environment. The containment recirculation screens are protected by plates located above them. These plates prevent debris from the failure of nonsafety-related coatings from getting into the water close to the screens such that the recirculation flow can cause the debris to be swept to the screens before it settles I to the floor. Stainless steel or safety-related coatings are used on the underside of these plates and on surfaces located below the plates, above the bottom of the screens,10 foot in front and I. 7 foot to the side of the screens to prevent coating debris from reaching the screens. A COL cleanliness program (refer to subsection 6.3.8.1) controls foreign debris introduced into the containment during maintenance and inspection operations. The Technical Specifications require visual inspections of the screens during every refueling outage. The design of the containment recirculation screens reduces the chance of debris reaching the screens. The screens are orientated vertically such that debris settling out of the water will not fall on the screens. The protective plates described above provide additional protection to the screens from debris. The bottom of the screens are located 2 feet above the floor.
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L Toer 2 Material- Engineered Safety Features (9/99) Page 6.3-21 i ! I l l
i i AP600 Design Control Document
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gi instead of using a debris curb, to prevent high density debris from being swept along the floor , by water flow to the containment recirculation screens. The containment recirculation screens are made up of a trash rack and a fine screen. The trash rack prevents larger debris from reaching the finer screen. The fine screen prevents debris larger than 0.125" from being injected into the reactor coolant system and blocking fuel cooling passages. Tne screen flow area is conservatively designed, considering the operation of the normal residual heat removal system pumps, which produce a higher flow than the gravity driven IRWST injection / recirculation flows. As a result, when the normal residual heat removal system pumps are not operating there is even more margin in screen clogging. 6.3.2.2.8 Valves Design features used to minimize leakage for valves in the passive core cooling system include: Packless valves are used for manual isolation valves that are 2 inches or smaller. Valves which are normally open, except check valves and those which perform control function, are provided with back seats to limit stem leakage. 6.3.2.2.8.1 Manual Globe, Gate, and Check Valves Gate valves have backseats and extemal screw and yoke assemblies. O Globe valves, both "T" and "Y" styles, are full-ported with external screw and yoke construction. Check valves are spring-loaded lift piston types for sizes 2 inches and smaller, and swing-type for sizes 2.5 inches and larger. Stainless steel check valves have no penetration welds other than the intet, outlet, and bonnet. The cneck valve hinge is serviced through the bonnet. 1 The gasket of the stainless steel manual globe and gate valves is similar to those described , in subsection 6.3.2.2.8.3 for motor-operated valves. 1 6.3.2.2.8.2 Manual Valves Manual valves are generally used as maintenance isolation valves. When used for this function they are under administrative cor, rol. They are located so that no single valve can i isolate redundant passive core cooling system equipment or they are provided with alarms in ! the main control room to indicate mispositioning. l To help preclude the possibility of passive core cooling system degradation due to valve , mispositioning, line connections such as vent and drain lines, test connections, pressure points, l flow element test points, ilush connections, local sample points, and bypass lines are provided O Tier 2 Material- Engineered Safety Features Page 6.3-22 L
I l 1^ l AP600 Design Control Document i g ! G l 63.7.6.23 Passive Residual Heat Removal Heat Exchanger Inlet Motor-Operated Valve Control l I The motor-operated valve in the passive residual heat removal heat exchanger inlet line is I normally open during normal plant operation. Power to this valve is locked out. Redundant I valve position indications and alarms are provided to alert the operator if the valve is open. l This valve also receives an actuation signal to confirm that it is open in the event of an I accident.
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53.7.7 Automatic Depressurization System Actuation at 24 Hours A timer is used to automatically actuate the automatic depressurization system if offsite and onsite power are lost for about 24 hours. This prevents discharging the Class lE de power sources such that they are no longer able to operate the automatic depressurization system valves. If power becomes available to the de batteries and they are no longer discharging prior to activation of the timer, Uen the automatic depressurization system actuation would be delayed. If the plant does not need actuation of the automatic depressurization system based on having stable pressurizer level, full core makeup tanks, and high and stable in-containment refueling water storage tank levels, the operators are directed to de-energize all loads on the 24-hour battuies. This action will block actuation of the automatic , depressurization system . rad allow for its actuation later should the plant conditions ) unexpectedly deg ade. (' 63.8 Combined License Information 6.3.8.1 Containment Cleanliness Program The Combined License applicants referencing the AP600 will address preparation of a program to limit the amount of debris that might be left in the containment following refaeling and maintenance outages. 63.9 References
1. WCAP-8966, " Evaluation of Mispositioned ECCS Valves." September 1977.

2. WCAP-13662 (Nonproprietary), "FMEA of Advanced Passive Plant Protection System,"
Revision 1, June 1998. '
3. WCAP-14837, "AP600 Shutdown Evaluation Report," Revision 3, March 1998.
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Tier 2 Material- Engineered Safety Features (9/99) Page 6.3-53
AP600 Design Control Document O Table 6.3-1 (Sheet 1 of 12) AP600 PXS PROCESS FLOW TABLE (SAFETY INJECTION) Case 1: NonLOCA, CMT Operation in Water Circulation Mode Location Description Fluid Pressure Temp. Flow Volume (psig) ( F) (Ib/sec) (ft') 1 Pzr steam Steam 2,265 663 - 550 2 Pu liquid W ater . 2,265 663 - 1050 5 Cold leg W ater 2,265 545 Nat Cire - 6 Hot leg Water 2.265 600 Nat Cire - 11 CIJCMT line Water 2,265 545 47 - 12 CMT inlet line Water 2,265 545 47 - 15 CMT (top) Water 2,265 545 - 100 16 CMT W ater 2,265 50 - 1,900 18 CMT outlet line Water 2,265 50 62 - 20 Accum gas Nitrogen 700 50 - 300 21 Accum water W ater 700 50 - 1,700 22 Accum outlet line Water 2,265 50 0 - 25 IRWST vent Air 0 50 0 - 26 IRWST gas Air 0 50 - 5,000 28 IRWST water W ater 14 50 - 70,800 30 IRWST outlet line Water 14 50 0 - 32 Cont sump ck valve W ater 0 50 0 - 33 Cont sump MOV Air 0 50 0 - O ~~~' 2 Material- Engineered Safety Features Page 6.3-54
7: i 1-AM Design Control Document Legend for Tables 6.31 and 6.3 2 Pzr - Pressurizer Accum . - ~ Accumulator Cont - Containment CK - Check MOV -- Motor-operated valve Nat Cire - Natural circulation flow DVI - Direct vessel injection Disch - Discharge N/A - Not applicable SG - Steam generator IIX - Heat exchanger Forced - Forced circulation flow i AOV- -
. Air-operated valve N2 -
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Tier 2 Materiel- Engineered Safety Features Page 6.3-69
AP600 Design Control Document O Table 6.3-3 PASSIVE CORE COOLING SYSTEM - REMOTE ACTUATION VALVES Normal Actuation Failed Position Position Position Notes Core Makeup Tanks CMT inlet isolation MOV (V002A/B) Open Open As is (1) CMT outlet isolation AOV (V014A/B,V015A/B) Closed Open Open Accumulators Accumulator discharge MOV (V027A/B) Open Open As is (2,4) In-Containment Refueling Water Storage Tank 1RWST injection line MOV (V121 A/B) Open Open As is (2,4) IRWST injecticn line squib (V123A/B, V125A/B) Closed Open As is Containment Reciret.lation Sump Valves Recirculation line MOVs (V117A/B) Closed Open As is Recirculation line squib valves (V)18A/B,120A/B) Closed Open As is Passive Residual Heat Removal Heat Exchanger l PRHR HX inlet MOV (V101) Open Open As is (1,2) PRHR HX outlet AOVs (V108A/B) Closed Open Open IRWST gutter isolation AOVs (V130A/B) Open Closed Closed Automatic Depressurization System Valves ADS Stage 1 MOVs (V001 A/B, V011 A/B) Closed Open As is ADS Stege 2 MOVs (V002A/B, V012A/B) Closed Open As is ADS Stage 3 MOVs (V003A/B, V013A/B) Closed Open As is ADS Stage 4 MOVs (V014A/B/C/D) Open Open As is (3) ADS Stage 4 squib valves (V004A/B/C/D) Closed Open As is Notes: (1) These valves are normally in the correct post-accident position, but receive confirmatory actuation signals to redundant controllers. (2) These valves are normally in the correct post-accident position with their power locked out. They also receive confirmatory actuation signals. (3) These valves are normally in the correct post-accident position, but receive confirmatory actuation signais. (4) The operation of these valves is not safety-related. O Tier 2 Material- Engineered Safety Features (Cl99) Page 6.3 70
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j- l i AP600 Design Control Document l ,in x( ) i Fire Protection System Integrity
' An evaluation of the consequences ofinadvertent operation of an automatic suppression system !
I or of a break in a fim protection line in this fire area are considered in the evaluation ofintemal ! flooding in Section 3.4 I Safe Shutdown Evaluation There are no safe shutdown components in this area, so a fire in th;s area has no impact on safe l shutdown. The electrical equipment in this area is non-Class IE; however, some division A and ; C cables are routed through this area. In the event of a fire, the division A and C cabling in , this area can be damaged. This damage can result in loss of control of equipment serviced by l these cables. Other components in divisions A and C are not affected. Safe shutdown is ! possible from equipment in other fire areas. i Neither a fire nor fire suporession activities in this fire area affect the safe shutdown capability ; of components located in adjacent fire areas. ! No fim in this fire area can cause spurious actions which could cause a breach in the reactor coolant boundary or defeat safety-related decay heat removal capability or cause an increase in shutdown reactivity of the reactor. O 9A.3.1.3.1.3 Fire Area 1204 AF 01 I V This fire area is subdivided into the following fire zones: Fire Zone Epom No.
1205 AF 12365 12365 Waste monitor tank room B

1214 AF 12354 12354 Mid-annulus access room

1234 AF 12351 12351 Maintenance floor staging area
= 1234 AF 12352 12352 Personnel hatch = 1235 AF 12363 12363 Waste monitor tank room A
1244 AF 12452 12452 Containment air filtration system penetration room
= 1244 AF 12454 12454- Containment air filtration system / spent fuel pool cooling system / primary sampling system penetration room
1254 AF 12553 12553 Personnel occess area i
= 1244 AF 12554 12451 Security room . 1254 AF 12554 12554 Security room -= 1264 AF 12651 12651 Radiologically controlled area ventilation system equipment room = 1205 AF 12362 12362 Normal residual heat removal heat exchanger room = 1215 AF 12162 12162 Norms' residual heat removal pump room A
1220 AF 12256 12256 Containment isolation valve area
* - 1220 AF 12256 12269 Pipe chase .. 1220 AF 12256 12253 Pipe chase ; . ,rx, i .i : .
i (] Tier 2 nieterini- Auxiliary Systems (9/99) Page 9A 81
AP600 Design Control Document O l
1225 AF 12262 12262 Piping / valve room Some of the piping in this fire area normally contains radioactive material.
Fire Detection and Suppression Features
Fire detectors Hose station (s)

Portable fire extinguishers Smoke Control Features
%e radiologically controlled area ventilation system serves this fire area on a once-through basis. Some of the ventilation system equipment is also located within this fire area. For a fire that does not disable the ventilation system, the system continues to ventilate the fire area unless the operator decides to shut down the system, or until heat from the fire is sufficient to close the fire dampers. Fire dampers close automatically on high temperature to control the spread of fire and smoke. If the radiologically controlled area vc:ltilation system is not affected by the fire, smoke is removed from the fire area by reopening the fire damper (s) after a fire and exhausting to the atmosphere. If the radiologically controlled area ventilation system is unavailable, smoke is removed from the fire area by using portable exhaust fans and flexible ductwork.
Fire Protection Adequacy Evaluation A fire in this fire area is detected by a fire detector which produces an audible alarm locally and both visual and audible alarms in the main control room and the security central alarm station. The fire is extinguished manually using hose streams or portable extinguishers. Combustible materials in this large fire area are listed in Table 9A-3, and consist primarily of cable insulation for the cables associated with the mechanical equipment and instrumentation in this fire area. There are small concentrations of lubricants in the radiologically controlled area ventilation system equipment room fire zone. Concentrations of paper or plastic anti-contamination clothing may also be present in some fire zones. Concentrations of transient combustibles may be present in tlie maintenance floor staging area fire zone. There are small concentrations of cable in overhead cable trays. This is a light hazard fire area and the rate of fire growth is expected to be slow. Three-hour fire barriers provide adequate separation from adjacent fire areas and the fire is contained within the fire area. The ventilation system does not contribute to the spread of the fire or smoke as described in the Smoke Control Features section above. Fire Protection System Integrity An evaluation of the consequences ofinadvertent operation of an automatic suppression system is not required because there are no such systen:s in this fire area. The consequences of a break Tier 2 Afsterial- Auxillary Systems (9/99) Page 9A 82 \
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l in a fire protection line in this fire area were considered in the evaluation of internal flooding i in Section 3.4 Safe Shutdown Evaluation i Table 9A 2 lists the safe shutdown components located in this fire area. The normal residual heat removal, primary sampling system, spent fuel pool cooling system and containment air filtration system containment isolation valves are conservatively assumed to be disabled as a result of a fire in this fire area. The redundant normal residual heat removal, primary sampling system, spent fuel pool cooling system and containment air filtration system containment isolation valves located inside containment are outside of this fire area and are sufficient to perform the applicable functions to maintain containment integrity. Cable trays supplying these valves and other components are not required fot safe shutdown. Neither a fire nor fire suppression activities in this fire area affect the safe shutdown capability of components located in adjacent fire areas. No fire in this fire area can cause spurious actions which could cause a breach in the reactor coolant boundary or dehat safety-related decay heat removal capability or cause an increase in shutdown reactivity of the reactor. 9A.3.1.3.1.4 Fire Area 1220 AF 02 b 'Ihis fire area is compnsed of the following room (s): L Room No. 12244 Lower annulus valve area - Some of the piping in this fire area norraally contains radioactive material. Fire Detection and Suppression Features
Fire detector Hose station (s)

Portab!e fire extinguishers Smoke Control Features "Ihe radiologically controlled area ventilation system ser"es this fire ama en a once-through basis. In the event of a fire the system continues to ventilate the fire area unless the operator decides to shut down the system, or until heat from the fire is sufficient to close the fire dampers. Fire damperts) close automatically on high temperature to control the spread of fire and smoke. A.ter the fire, smoke is removed from the fire area by reopening the fire dampers.
The radiologically controlled area ventilation system exhausts the smoke and hot gases to the atmosphere. A '( p
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1 Tier 2 Ataterial- Auxiliary Systems (9/99) Page 9A 83
f l AP600 Design Control Document O Fire Protection Adequacy Evaluation A fire in this fire area is detected by a fire detector which produces an audible alarm locally and both visual and audible alarms in the main control room and the security central alarm station. He fire is extinguished manually using hose streams or ponable extinguishers. Combustible materials in this fire area are listed in Table 9A-3, and consist primarily of cable insulation for cables associated with the few containment isolation valves in this fire area. There are no significant concentrations of combustible materials. This is a light hazard fire area and the rate of fire growth is expected to be slow. Three-hour fire barriers provide adequate separation from adjacent fire areas and the fire is contained within the fire area. He ventilation system does not contribute to the spread of the fire or smoke as described in the Smoke Control Features section above. Fire Protection System Integrity An evaluation af the consequences ofinadvertent operation of an automatic suppression system is not required because there are no such systems in this fire area. An eva.oaion of the consequences of a break in a fire protection line is not required because no such lines pass through or terminate in this fire area. Safe Shutdown Evaluation Table 9A-2 lists the safe shutdown components located in this fire area. The chemical and volume control system and liquid radwaste system containment isolation valves are conservatively assumed to be disabled as a result of a fire in this fire area. The redun<hnt chemical and volume control system and liquid radwaste system containment isolation 6.fes located inside containment are outside of this fire area and are sufficient to perform the applicable func* ions to achieve and maintain safe shutdown. Neither a fire nor fire suppression activities in this fire area affect the safe shutdown capability of components located in adjacent fire areas. No fire in this fire area can cause spurious actions which could cause a breach in the reactor coolant boundary or defeat safety-related decay heat removal capability or cause an increase in shutdown reactivity of the reactor. l 9A3.13.2 Miscellaneous Areas 9A3.13.2.1 Fire Area 1204 AF 02 l l This fire area is comprised of the following room (s): Room No. l S03 Stairwell O Tier 2 Material- Auxillary Systems Page 9A-84
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PRHR HX - Operating 3.5.4 p SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify the outlet manual isolation valve 12 hours is fully open. SR 3.5.4.2 Verify the inlet motor operated 12 hours isolation valve is open. SR 3.5.4.3 Verify the volume of noncondensible 24 hours gasesintpePRHRHXinletlineis s [0.4] ft SR 3.5.4.4 Verify that power is removed from the 31 days inlet motor operated isolation valve. SR 3.5.4.5 Verify both PRHR nh cperated outlet In accordance p isolation valves aN both IRWST gutter with the System Q isolation valves are OPERABLE by stroking open the valves. Level Inservice i Testing Program j SR 3.5.4.6 Verify PRHR HX heat transfer performance 10 years in accordance with the System Level Operability Testing Program. SR 3.5.4.7 Verify by visual inspection that the 24 months IRWST gutters are not restricted by i debris. ps L.) b AP600 3.5 11 (9/99) Amendment 0
1 PRHR HX - Shutdown, RCS Intact 3.5.5 l ! 3.5 PASSIVE CORE COOLING SYSTEM (PXS) 3.5.5 Passive Residual Heat Removal Heat Exchanger (PRHR HX) - Shutdown, RCS Intact I LC0 3.5.5 The PRHR HY shall be OPERABLE.
.......................... . NOTE- --- - - - - --
When any reactor coolant pumps (RCPs) are operating, at least one RCP must be operating in loop one. APPLICABILITY: H0DE 4 with the RCS cooling provided by the Normal Residual Heat Removal System (RNS), i MODE 5 with the RCS pressure boundary intact. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One air operated outlet isolation A.1 Restore air operated outlet valve to OPERABLE 72 hours g valve inoperable. status. B. One air operated B.1 Restore air operated 72 hours IRWST gutter IRWST gutter isolation isolation valve valve to OPERABLE inoperable. status. C. Presence of non- C.1 Vent noncondensible 24 hours condensible gases in gases. the high point vent. D. PRHR HX inoperable D.1 Restore PRHR HX to 8 hours for reasons other OPERABLE status. than A, B, or C.
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(continued) O b AP600 3.5 12 Amendment 0 1
PRHR HX - Operating B 3.5.4 BASE.5, SURVEILLANCE -SR 3.5.4.1 (continued) REQUIREMENTS Misalignment of this valve could render the heat exchanger inoperable. A 12 hour Frequency is reasonable considering that the valve is manually positioned and has control room position indication and alarm. SR 3.5.4.2 Verification that the motor operated inlet valve is fully open, as indicated in the main control room, ensures timely discovery if the valve is not fully open. The 12 hour Frequency is consistent with the ease of verification, confirmatory open signals, and redundant series valve controls that prevent spurious closure. SR 3.5.4.3 Verification that excessive amounts of noncondensible gases are not present in the inlet line is required every 24 hours. The inlet line of the PRHR HX has a vertical section of pipe which serves as a high point collection point for noncondensible gases. Control room indication of the water level in this high point collection aoint is available to verify that noncondensible gases lave not C. l collected to the extent that the water level is depressed below the allowable level. The 24 hour Frequency is based on the expected low rate of gas accumulation and the availability of control room indication. ' SR 3.5.4.4 Verification is required to confirm that power is removed from the motor operated inlet isolation valve every 31 days. Removal of power from this valve reduces the likelyhood that the valve will be inadvertently closed as a result of a fire. The 31 day Frequency is acceptable considering the frequent surveillance of valve position and that the valve has a confirmatory open signal SR 3.5.4.5 Verification that both air operated outlet valves and both IRWST gutter isolation valves are OPERABLE ensures that the PRHR HX will actuate on command, with return flow from the gutter to the IRWST, since all other components of the i (continued) O N AP600 B 3.5 23 (9/99) Amendment 0 E
PRHR HX - Operating B 3.5.4 BASES SURVEILLANCE SR 3.5.4.5 (continued) REQUIREMENTS (continued) system are normally in the OPERABLE configuration. Since these valves are redundant, if one valve is inoperable, the system can function at 100% capacity. Verification requiresthe actual operation of each valve through a full cycle to demonstrate operability. The Surveillance Frequency is provided in the Inservice Testing Program. SR 3.5.4.6 This SR requires performance of a system performance test of the PRHR HX to verify system heat transfer capabilities. The system performance test demonstrater that the PRHR HX heat transfer assumed in accident analyses is maintained. Although the likelihood that system performance would degrade with time is low, it is considered prudent to periodically verify system performance. The System Level Operability Testing Program provides specific test requirements and acceptance criteria. SR 3.5.4.7 This surveillance requires visual inspection of the IRWST gutters to verify that the return flow to the IRWST will not be restricted by debris. A Frequency of 24 months is adequate, since there are no known~ sources of debris with which the gutters could become restricted. REFERENCES 1. Section 6.3, " Passive Core Cooling System."
2. Chapter 15. " Safety Analysis."

3. AP600 PRA Thermal / Hydraulic Uncertainty Evaluation for Passive System Reliability." WCAP 14800, June 1997.
I O bAP600 B 3.5 24 (9/99) Amendment 0 l l _}}

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REFERENCE:

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