ML20212A247
| ML20212A247 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 09/15/1999 |
| From: | Joshua Wilson NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9909160127 | |
| Download: ML20212A247 (47) | |
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,j UNITED STATES NUCLEAR REGULATORY COMMISSION i
% '2 WASHINGTON, D.C. 20066 4001 o* September 15, 1999 Y . . ...D APPLICANT: Westinghouse Electric Company FACILITY: AP600 Standard Plant Design
SUBJECT:
MEETING
SUMMARY
ON DESIGN CONTROL DOCUMENT CHANGES The subject meeting was held on September 2,1999, in the Rockville, Maryland, office of the U.S. Nuclear Regulatory Commission (NRC) between representatives of Westinghouse and the NRC staff. Enclosure 1 is a list of meeting attendees and Enclosures 2,3,4, and 5 are the handouts provided by Westinghouse during the meeting. The purpose of the meeting was for Westinghouse to discuss the changes that they plan to make in Revision 3 of the AP600 Design Control Document (DCD). Westinghouse identified the need for these changes as a part of the process to ensure that design changes and commitments were included in the AP600 DCD. Westinghouse's review consisted of checking the consistency of the implementation of approved design change proposals; ongoing detailed design work on systems, structures, and components outside the scope of design certification; and commitments made in the final stages of the design certification review. As a result of this review, several documentation inconsistencies were noted by Westinghouse and corrected. In addition, one change to the description of the plates above the recirculation pump screens was necessitated by the location of other nearby hardware. Westinghouse believes that none of the items will change the conclusions of the staff's design certification review. Mr. McIntyre began the meeting by presenting a list of the changes to the DCD (enclosure 2) and the revised DCD pages (enclosure 3) that correspond to the list of changes. Mr. McIntyre explained that most of the changes were made to correct typographical errors and inconsistencies in the AP600 DCD. Mr. Schulz made a presentation on the change to the size of the plates that protect the containment recirculation screens (enclosure 4). Enclosure 5 consists of large size drawings that showed some of the changes. These drawings were not distributed with this meeting summary. 1Cr'009 DGO3{ t 9909160127 990915 PDR ADOCK 05200003 A PDR _ _ _ _ _ _ _ _ _ _3
\ "s o September 15, 1999 The NRC staff advised Westinghouse that the proposed changes described in the meeting were acceptable for the AP600 standard plant design. The staff will determine if it is necessary to revise the AP600 Final Safety Evaluation Report (NUREG-1512) because of these changes.
original signed by: Jerry N. Wilson, Senior Policy Analyst License Renewal and Standardization Branch
. Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation Docket No. 52-003
Enclosures:
As stated l cc w/ enclosures 1-4: See next page Distribution: Docket File (enclosures 1-5) PUBLIC (enclosures 1-4) RLSB (enclosures 1-4) ACRS (enclosures 1-4) JNWilson (enclosures 1-4) Distribution w/o enclosures: SCollins DMatthews BSheron CGrimes WKane GMizuno YGHsii TCheng i RLobel MSnodderly EHylton ! i DOCUMENT NAME: G:\RLSB\ WILSON \MTG-9299.WPD OFFICE LA /, SPA:RLSB q PD:RLSB _ NAME EH[ JNWilso CGrimes h DATE 09/ /Y /99 09//Y/99 h 09/16 /99 OFFICIAL RECORD COPY
f -n , ,i ( Westinghouse Electric Corporation Docket No. 52-003 l l Mr Brian A. McIntyre, Manager l' Advanced P! ant Safety & Licensing
- Westinghouse Electric Company P.O. Box 355 l
Pittsburgh, PA 15230 Mr. H. A. Sepp
- Advanced Plant Safety & Licensing Westinghouse Electric Company P.O. Box 355 Pittsburgh, PA 15230 Ms. Lynn Connor Doc- Search Associates Post Office Box 34 Cabin John, MD 20818 .
' Ms. Susan Fanto Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit P.O. Box 355 Pittsburgh, PA 15230 Mr. Jack Bastin Westinghouse Electric Company 11921 Rockville Pike Suite 107 Rockville, MD 20852 Barton Z. Cowan, Esq.
Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Char!ss Thompson, Program Manager AP600 Certification NE-50 19901 Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design Certification Electric Power Research Institute , 3412 Hillview Avenue l Palo Alto, CA 94303 , l l w -. . _ _
AP600 DESIGN CONTROL DOCUMENT ATTENDANCE LIST SEPTEMBER 2,1999 NAME ORGANIZATION JERRY N. WILSON NRC\NRRORIP\RLSB BRIAN MCINTYRE WESTINGHOUSE TERRY SCHULZ WESTINGHOUSE NARENDRA PRASAD WESTINGHOUSE Y. GENE HSil NRC\NRROSSA\SRXB THOMAS CHENG NRC\NRROE\EMEB RICHARD LOBEL NRC\NRR\DSSA\SPLB MICHAEL SNODDERLY NRC\NRROSSA\SPSD l i i Enclosure 1 C _ _ . _ .
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- A'P600 De:Ign Control Docum:nt
. i Table 2.2.3-4 (cont.)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria vii) Inspection of the as built vii) Plates located above each components will be conducted containment recirculation screen i for plates located above the are no more than I 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 ! I 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 i I screens will be conducted. 2 70 ft'. The bottom of the containment recirculation screens is 2 2 ft above the loop compartment floor, ix) Inspections will be in) 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 nonsafery 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 inlet diffuser has a CMT inlet diffuser will be flow area 2165 in'. conducted. xii) Inspections will be ali) 1he centerline of each upper conducted of the CMT level level tap line at the tee for each sensors (PSX ll A/B/D/C, - level sensor is located I" !" 12A/B/C/D, 13A/BC/D, - below the centerline of the upper 14A/B/C/D) upper level tap level tap connection to the CMT. lines. Tier i Afsterial (9/99) Page 2.2.3 21 Enclosure 3 m __ _
[ . P600 Design Control Docurnent i L (2)(vil) Plant Radiation Shielding (NUREG 0737 Item D.B.2)
Perform radiation and shielding design reviews of spaces around systems that may, as a result l of an accident,'contain TID-14844 source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the l
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. De post-LOCA instantaneous and integrated source strengths as a function of time are also included as Tables 12.2 20 and 12.2 21, respectively. De sources are based on the core activity release model from NUREG-1465, which supersedes the TID-14844 source term assumptions as reflected in Regulatory Guide 1.4. De 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, inclu' ding radiation zone maps that show projected dose rates in these areas and access routes for the various pos'.-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 demonsw.ing total individual radiation doses are limited to less than.5 rem and that Item IIs 2 of NUREG-0737 and GDC 19 requirements are met. Environmer, tl qualification of safety related equipment is addressed in Section 3.11. De
~ detennination d$e radiation environments during postulated accident situations considers the activity release most based on NUREG 1465, which supersedes the source term definition i
l of Parts I and 4 of Itea 11.B.2 of NUREG-0737. r As noted in subsection 12.2.a. the Combined License applicant will address any additional contained radiation sources not Ventified in 12.2.1. Dus, appropriate source terms have been - identified and used in establishing that the requirements of item II.B.2 of NUREG-0737 and GDC 19 are mer and the issues are resolved. i (2)(viii) Post Accident Sampling (NUREG 0737 Item II.B.3)
' Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844 source term radioactive materials without l 1 radiation caposures to any individual eaceeding 5 rem to the whole-body or 50 rem to the e emities. Materials to be analyzed and quantified include certain radionuclides that are inoicators of the degree of core damage (e.g. noble gases, iodines and cesiums, and non volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations.*
nor 2 Assesrial- hidroduction aruf General Osscr$ption of Piant (949) Page 1.913
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- AP600 De:Ign Control Document 1
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 atmosphe:e. l 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 prim'ay sampling system. (2)(ix) Hydrogen 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. 'Ihe hydrogen control system and associated systems shall provide, with reasonable assurance, that:
(A) Uniformly distributed hydrogen concentations 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 shutdown of the plant and maintaining containment integrity 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 including l the environmental conditions created by activation of the hydrogen control system. I (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." AP600 Response: See the response provided for issue (1)(xii). Tier 2 Material- introduction and General Description of Plant (9/99) Page 1.9-14 i g p am- _-- -
IJ P600 Design Control Document 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 design stresses to the allowable stresses. When thermal efects result in stresses above yield, the evaluation is in accordance with the supplemental criteria)* as described in subsection 3.8.3.5.3.4. 3.8.3.5.8.3 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 operating floor at elevation 135'3". In addition to supporting the operating floor, it also supports a steel gratingfloor at elevation 118'-O'.
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l The load combinations in Table 3.8.41 were used to assess the adequacy ofthe 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 adequate for all the opplied loads.}* i 3.8.3.6 Materials, Quality Control, and Special Construction Techniques Subsection 3.8.4.6 describes the materials and quality control program used in the construction of the containment intemal structures. De structural steel modules are constructed using A36 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. De 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 3.8.4.3.1.1. Packaging, shipping, receiving, storage and handling of structural l modules are in accordance with NQA 2 Part 2.2 (formerly ANSI /ASME N45.2.2 as specified 1 in AISC N690). i 3.8.3.6.1 Fabrication, Erection, and Construction of Structural Modules Modular construction techniques are used extensively in the containment internal structures (Figure 3.8.3 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 lechnique is described in the following paragraphs for Module MI, which is the main structural module in the containment intemal structures. T. . M1 module is a multicompartmented structure which, in its final form, comprises the central walls of the containment intemal structures. De vertical walls of the module house the refueling cavity, the reactor vessel compartment, and the two steam generator
- l. l compartments. De module (Figure 3.8.314) is in the form of a "T" and is approximately 89 1 feet long,95 feet wide and 77 feet 6 inches high. De module is assembled from about 40
*NRC staff approval is required prior to +====f a change is this sasannel; see DCD Introduccan secnon 31 l
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. AP600 De:Ign Control Documsnt GDC 35 of 10 CFR 50 Appendix A { Regulatory Guide 1.82 ' NUREG-0897 De 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. l 63.2.2.7.1 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 beat removal pumps with suction from the i IRWST and the containment recirculation lines is considered in sizing screens. 1 System and screen performance are evaluated
)
Screens have solid top cover. Containment recirculation screens have protective l plates that are located no more than I 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. He plate i dimensions are relative to the portion of the screens where water flows through I 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 related pumps. Corrosion resistant materials are used for screens Tier 2 Material- Eng(neered Sekty Features (9/99) Page 6.318
AP600 l Design Control Docurnent j] 6.3.2.2.7.3 Containment Recirculation Screens De containment recirculation screens are oriented vertically along walls above the loop compartment floor (elevation g3 feet). Figure 6.3-8 shows a plan view and Figure 6.3-9 . shows a section view of these screens. Two separate screens are provided as shown in Figure 6.3 3. De loop compartment floor elevation is significantly above (11.5 feet) the lowest level in the containment, the reactor vessel cavity. De bottom of the recirculation screen is two foot above the floor, providing a curb function. During a LOCA, the reactor coolant system blowdown will tend to carry debris created by the accident (pipe whip / jets) into the cavity under the reactor vessel which is located away from and below the containmerit recirculation screens. As the accumulators, core makeup tanks and IRWST inject, the containment water level will slowly rise up to the 108 foot elevation over at least 5 hours. De 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 initially open, the water level in the IRWST is higher than the containment water level and water flows from the IRWST backwards through the containment recirculation screen. Dis back flow tends to flush debris located close to the !
. recirculation screens away from the screens.
l De water level in the containment when recirculation begins is well above (- 10 feet) the top of the recirculation screens. During the long containment floodop time (>5 hours), floating debris 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 screens. De 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 accident and be postulated to reach the screens during recirculation. De nonsafety-related coatings used in the containment are designed to withstand the post accident environment. De containment recirculation screens are protected by plates located above them. Dese 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. Costings used on the underside of these plates and on surfaces located below the l plates and above the bottom of the screens are safety-related. A COL cleanliness program (refer to subsection 6.3.8.1) controls foreign debris introduced into the containment during maintenance and inspection operations. De Technical Specifications require visual inspections of the screens during every refueling outage. De design of the containment recirculation screens reduces the chance of debris reaching the screens. De screens are orientated vertically such that debns settling out of the water will not fall on the screens. De protnctive plates described above provide additional protection to the screens from debris. De bottom of the screens are located 2 feet above the floor, Der 2 Adatorist - Engissered sansfy Festures (W99) Page 6.3-21 M-- __ w , - - w
AP600 Design Control Document l0 l 6.3.7.6.2.3 Passive Residual Heat Removal Heat Exchanger Inlet Motor Operated Valve Control l l De 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 Dis valve also receives an actuation signal to confirm that it is open in the event of an l acci, lent. { 6.3.7.7 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 abc,ut 24 hours. Dis prevents discharging the Class IE 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, then 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<ontainment refueling water storage tank levels, the operators are directed to de-energize all loads on the 24-hour batteries. His action will block actuation of the automatic depressurization system and kilow for its actuation later should the plant conditions unexpectedly degrade.
l l 1 6.3.8 Combined License Information 6.3.8.1 Containment Cleanliness Program I l De 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 refueling and maintenance outages. 6.3.9 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.
Der 2 Afsterdaf - Krigirisered Sansty Features (W99) Page 6.3 53
r AP600 Design Control Document g Table 6.3-3 PASSIVE CORE COOLING SYSTEM REMOTE ACTUATION VALVES Normal Actuation Failed Position Position Eg3]Ilan FlPJn 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 IRWST injection line MOV (V121 A/B) Open Open As is (2,4) IRWST injection line squib (V123A/B, V125A/B) Closed Open As is Containment Recirculation Sump Valves Recirculation line MOVs (VI17A/B) Closed Open As is Recirculation line squib valves (VilBA/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 (Vl30A/B) Open Closed Closed Automatic Depressurization System Valves ADS Stage 1 MOVs (V00l A/B, V01I A/B) Closed Open As is ADS Stage 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 l Notes: (') Tne.:e valves are normally in the correct post-accident position, but receive confirmatory actuatic . signals to i redundant contro!! cts. l (2) These valves are normally in the correct post-accident position with their power locked out. Dey alsc receive ! confirmator" actuation signals. ! (3) Dese valvs are normally in the correct post-accident position, but receive confirmatory actuation sign 31s. (4) De operation of these valves is not safety-related. l I I l Tier 2 Afatorial- Engineered Sakty Features (9/99) Page 6.3-70
AP600 Design Control Document cB 1 CONTAINMENT RECIRCULATION SCREENS A i\ ,
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. AP600 Design Control Document l Fire Protection System Integrity 1 An evaluation of the consequences ofinadvertent operation of an automatic suppression system j l or of a break in a fue protection line in this fue ama are considered in the evaluation of intemal l flooding in Section 3.4 i Safe Shutdown Evaluation l
'Ihere are no safe shutdown components in this area, so a fire in this area has no impact on safe j shutdown. The electrical equipment in this crea is non-Class IE; however, some division A and C cables are routed through this area. In the event of a fue, the division A and C cabling in this area can be damaged. This damage can result in loss of control of equipment serviced by these cables. Other components in divisions A and C are not affected. Safe shutdown is possible from equipment in other fire areas.
Neither a fire nor fue suppression activities in this fire area affect the safe shutdown capability of components inced in adjacent fue uras. No fire in this fue area can cause spurious actions which could cause a breach in the reactor i coolant boundary or defeat safety-related decay heat removal capability or cause an increase in j shutdown reactivity of the reactor. 1 9A.3.13.1.3 Fire Area 1204 AF 01 I This fire area is subdivided into the following fue zones: Fire Zone Room 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 access area e 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 Normal residual heat removal pump room A
- 1220 AF 12256 12256 Containment isolation valve area e 1220 AF 12256 12269 Pipe chase
. 1220 AF 12256 12253 Pipe chase Tier 2 Material- Aux / Nary Systems (9/99) Page 9A-81
, ' EP600 Design Control Document c
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- 1225 AF 12262 12262 Piping / valve room Some of the piping in this fue area normally contains radioactive material.
Fire Detaction and Suppression Features Fire detectors Hose station (s) Portable fue extinguishers Smoke Control Features
'Ihe radiologically controlled area ventilation system serves this fue area on a once-through basis. Some of the ventilation system equipment is also located within this fue 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 fue dampers. Fire dampers close automatically on high temperature to control the spread of fue and smoke. If the radiologically controlled area ventilation system is not affected by the fire, smoke is removed from the fue area by reopening the fue 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 fue in this fire area is detected by a fue detector which produces an audible alarm locally and both visual and audible alarms in the main control room and the security central alarm station.
"Ihe 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. 'Ihere are small concentrations of lubricants in the radiologically controlled area ventilation sysiem equipment room fire zone. Concentrations of paper or plastic anti-contamination clothing may also be present in some fue zones. Concentrations of transient combustibles may be present in the maintenance floor staging area fire zone. 'Ihere are small concentrations of cable in overhead cable trays. '!his is a light hazard fue area and the rate of fue growth is expected to be slow. 'Ihree-hour fire barners provide adequate separation from adjacent fue areas and the fire is contained within the fire area.
'Ihe ventilation system does not contribute to the spread of the fire or smoke as described in the Smoke Control Features section above.
Fire Protection Systesa Integrity An evaluation of the consequences of ir$advenent operation of an automatic suppression system is not required because there are no such systems in this fire area. "Ihe consequences of a break flier 2 Masorief- Amriliary Systems (iW99) Page 9A-82
AP600 De:Ign Control Document 2 l in a fire protection line in this fue area were considered in the evaluation of internal flooding in Section 3.4 Safe Shutdown Evaluation Table 9A 2 lists the safe shutdown components located in this fue 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 fue in this fue area. The redundant normal residual heat amoval, primary sampling system, spent fuel pool cooling system and containment air filtration system containment i isolation valves located inside containment are outside of this fue area and are sufficient to ! perform the applicable functions to maintain containnent integrity. Cable trays supplying these valves and other components are not required for safe shutdown. Neither a fire nor Src suppression activities in this fue area affect the safe shutdown capability of components lewd is 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-selated decay heat removal capability or cause an increase in shutdown reactivity of the reactor. 9AJ.13.1.4 Fire Area 1220 AF 02 This fire area is comprised of the following room (s): Room No. 12244 kwer annulus valve area Some of the piping in this fire area normally contains radioactive material. i Fist Detection and Suppression Features
- Fire detector Hose station (s)
- Portable fire extinguishers Smoke Control Features The radiologically controlled area ventilation system serves this fue area on a once-through be 's. In the event of a fue the system continues to ventilate the fire area unien the operator de xies to shut down the system, or until heat from the fire is sufficient to close the fire dampers. Fire damper (s) close automancally on high temperature to control the spread of fire and smoke. After the fire, smoke is removed from the fue area by recpening the fire dampers.
The radiologically controlled area ventilation system exhausts the smoke and hot gases to the atmosphere. s Tier 2 Afslerist- Auz#dery Systems (9/99) Page 9A-83
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** Q 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 gases in t PRHR HX inlet line is 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 air operated outlet In accordance isolation valves and both IRWST gutter with the System isolation valver are OPERABLE by Level Inservice stroking open the valves. Testing Program SR 3.5.4.6 Verify PRHR HX heat transfer performance 10 years in accordance with the System Level Operability Testing Program. 4 SR 3.5.4.7 Verify by visual inspection that the 24 months IRWST gutters are not restricted by debris.
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