ML20069E653

Forwards Revised Final Draft Responses to Selected Commission 820825 & 1005 Questions Re Fire Protection,Leak Rate Testing & ESF Sys.Proposed Resolution Discussion Items Encl

ML20069E653
Person / Time
Site:	05000447
Issue date:	03/18/1983
From:	Sherwood G GENERAL ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
JHF-018-83, JHF-18-83, MFN-059-83, MFN-59-83, NUDOCS 8303220138
Download: ML20069E653 (58)

Text

GENERAL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN 059-83 MC 682 (408) 925-5040 JNF 018-83 March 18, 1983 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Mr. D.G. Eisenhut Division of Licensing Gentlemen :

SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)

DOCKET N0. STN 50-447 REVISED DRAFT RESPONSES, RESPONSES TO DISCUSSION ITEMS AND TEXT CLARIFICATIONS Attached please find revised final draft responses to selected questions of the Commission's August 25, 1982 and October 5,1982 information requests.

Only modifications (new or revised) to the responses of the referenced letters are provided. Also attached are proposed resolution discussion i tems . The following are provided:

Attachment Number

, 1 Proposed Resolution of Chemical Engineering Branch Discussion Items on Fire Protection

and Appendix 9A Text Clarifications 2 Proposed Resolution of Containment Systems Branch Discussion Items 3 Draft Responses to Instrumentation and Control Systems Questions 4 Draft Responses to Structural and Geotechnical Engineering Branch Questions Sincerely,

. )

{@[

GTenn G. Sherwood, Manager Nuclear Safety & Licensing Operation cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o attachments)

D.C. Scaletti L.S. Gifford (w/o attachments) 8303220138 830318 PDR ADOCK 05000447 A PDR

ATTACHMENT NO. 1 PROPOSED RESOLUTION OF CHEMICAL ENGINEERING BRANCH DISCUSSION ITEMS ON FIRE PROTECTION AND APPENDIX 9A TEXT CLARIFICATIONS

The purpose of this attachment is to resolve the GE/NRC discussion items listed below pertaining to fire protection:

1. Qualifications of fire rated barriers

2. Qualifications of fire rated penetration seals

3. Qualifications of fire rated doors

4. Lack of 3-hour fire rated dampers

5. Safe Shutdown Capability

6. Alternate Shutdown Capability

7. Ventilation Systems

8. Separation of the Control Room

9. Lack of smoke detectors in the control room outside air intakes

10. Separation of the cable rooms

11. Separation of the remote shutdown panel If the resolution to these items described herein are accepted by the NRC, the detailed changes to the GESSAR II design will be provided to the NRC prior to the first Applicant referencing GESSAR II. Any GESSAR II/ Applicant interface requirements pertaining to the resolution of these items will be included in Section 1.9, Interface requirements in the next amendment.

Also included in this attachment are Appendix 9A text clarifications requested by the NRC during the fire protection review.

PROPOSED RESOLUTION OF DISCUSSION ITEMS Discussion Items 1,5,8,10 and 11 For the type 1,2 and 3 wall assemblies, the GESSAR II design will provide completely equivalent construction to tested wall assemblies or testing will be required. All three assemblies will be required to have a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rating. In addition, a wall and fire door rated 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> will be added in the corridor at the (-) 16'10" elevation of the auxiliary building. The combination of these two actions should resolve discussion items 1,3,8,10 and 11.

Discussion Item 2 _

GESSAR II will require qualification of all penetrations by test if possible or by analysis if testing is impractical or impossible. The penetrations are already required to have a fire rating equivalent to the barrier which they penetrate.

Discussion Item 3 GESSAR II will require that, with the exception of the fuel building railroad door, all door assemblies be tested to prove their ability to provide the required fire rating. The exterior railroad door for the fuel building is too large to be tested in a furnace. Also, for the GESSAR II designi a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating for the railroad door is not required to meet the requirements

of BTP ASB 9.5-1. The plant design objective was to provide a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire l rating for external walls. The construction of this door is equivalent or better than that which would be required to provide a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating. On this basis, the requirements to resolve discussion item 3 should be met.

, Discussion Item 9 Smoke detectors will be added to the air intakes for the control building.

This will resolve discussion item 9.

Discussion Item 6 .

Discussion item 6 is an NRC staff action item concerning the shutdown capabili ty. Since the GESSAR II design will have redundant remote shutdown capability which meets the requirements for fire separation, no future concerns are expected.

Discussion Items 4 and 7 These two items concern fire dampers in ventilation ducts used for smoke venting. Some of these ventilation ducts are shared systems in that they also provide normal ventilation. Other ducts are for smoke venting only.

Based on the discussion below the present GESSAR II design should be adequate and should be acceptable to the NRC, so that items 4 and 7 should be resolved.

The auxiliary building smoke removal system is shown on Figure 9.4-4 and described in Section 9.4.3.2.1.11. Each set of duct work serves and traverses only fire areas of one safety division. There is a smoke vent intake in each fire area with a remote manually operated fire damper which is normally closed.

There is a fusible link from the air operator to the vanes so that the damper will close on high temperature. The fire rating of the dampers is lh hours.

, The duct is heavy gage, welded construction which exceeds the requirements for 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated construction. Hence, the design is considered completely adequate for the service.

One of the design objectives of GESSAR II is to avoid fire dampers in smoke vents, as their automatic closure would render the smoke vent inoperative at the very time it was needed. With two exceptions, smoke vents pass through safety areas only of the same division as the vented area. The two exceptions are the Division 2 cable tunnel vent and the primary containment vent.

The Division 2 cable tunnel located in the corridor of (-)6'-3'! elevation of the auxiliary building has a dedicated smoke removal system. Which passes .

through the division 1 area. The inlet to the duct is fitted with a standard sprinkler head. Any heat or smoke that exceeds 1650 F will fuse the link allowing fire water to flow through the head. The duct opening is 2.5 sq ft.

This deluge spray will be sufficient to cool inlet gases from either direction below a temperature that could cause duct failure which could allow migration of heat to other fire resistence areas. This is consistent with NFPA 13 using sprinklers to protect openings in fire resistence walls where dampers cannot be fitted for other overriding criteria. The calculated flow rate from the cable tunnel during smoke venting is 3000 cfm, a relatively low flow rate.

The sprinkler is designed to flow .15 gpm per 100 square foot of floor area or a minimum of 15 gpm, therefore, 3000 cfm will be cooled by a minimum of 15 gpm water. This is sufficient flow to cool gases or smoke below the temperature that would weaken or collapse even a duct of standard gage construction. The duct has a thick wall and is all welded construction, which adds a redundant degree of protection.

i l

The other exception is the vent for primary containment. It has two l inboard (1 manual) isolation valves and one outboard isolation valve.

If a fire occurs, either the inboard valves or the outboard valve would be located out of the fire area and could be closed. The valve within the fuel building is located in a room with 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rated walls. The room is directly accessible from the fuel building or the stair tower between the fuel and auxiliary building. All return registers except for th,e pool sweep are located high in the containment so that bulk mixing, aided by the dome mixing system, would occur before any combustion gases enter the ventilation duct. The containment is more sensitive to bulk air temperature than the ventilation duct. If a fire raised the bulk temperature excessively, containment spray would be initiated to protect the containment at a temperature well below the threshold of damage to the ventilation duct. For these reasons, the current GESSAR II design for the containment ventilation is considered proper and adequate.

The remaining smoke vents which do not have fire dampers are the two in the control building. Each one of these smoke vents serves and traverses one division. Since it is impossible for these smoke vents to allow the fire in the area of one division to spread to another division, the current GESSAR II design is considered to be adequate and proper.

uma>nsu 7RMIA%U 238 NUCLEAR ISLAND Rev. O TE X r Ct ARI F t c.M) ON 5' -

9A.5.6 Carbon Dioxide Storage (Continued)

After initial discharge, a second discharge for the largest single hazard area must be maintained in the storage tank. Therefore, the Applicant must maintain a minimum of 11,200 lb of CO 2 fr Diesel-Generator Building fire protection.

In the event of malfunction of the automatic sequencing for CO 2 discharge to a hazard area, manual activation of the discharge sequence is provided in the control room.

9A.5.7 HVAC Systems The majority of the HVAC systems are provided with fire dampers where the duct penetrates a fire-resistive wall; however, there are some exceptions. There are some cases where divisional control valves are in the same fire area. These cases are presented, and the justification and/or effect on the plant operation relative to reactor safe shutdown is presented.

9A.5.7.1 Control Building The smoke removal systems for the cable rooms and control room are

• a function of damper arrangement, utilizing the existing air con-ditioning system. The cable room tunnel exhaust ducts are not I

provided with fire dampers The cabic rooms are provided with automatic wet pipe spr' lers and POC detectors. The cable trays are solid bottom, vered metal trays. A postulated electrically initiated XLP R cable insulation fire in a closed tray or PGCC would evo little smoke or heat. The anticipated transitory combu ble load, a function of the Applicant's fire safety p gram, is expected to be negligible. Inlet ducts are equipped ith fire dampers to prevent hot smoke or gases from entering the areas from fire sources exterior to the areas.

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s*MainIB 7EASAMF 238 NUCLEAR ISLAND Rev. 0 9A.S.7.2 Division 2 Cable Tunnel - Auxiliary Building The Division 2 cable tunnel is provided with a dedicated smoke removal system. This system is not fitted with a fire damper.

The duct inlet is fitted with4sMec-- ' ' ' -

a .PM )jrtR_ M spray nozzle fed from the cable tunnel automatic wet pipe sprinkler system.b P detection is provided. The cable trays are solid bottom, covered metal trays. A postulated electrically initiated XLPE-FR cable insulation fire in a closed tray would involve little smoke or heat. The anticipated transitory combustible load, a function of the Applicant's fire safety program, is expected to be negligible.

Inlet and exhaust ducts from the normal ventilation system are fitted with fire dampers to prevent hot smoke or gases from entering the tunnel via fire sources exterior to the tunnel. .

COO \ % f O Q % W Uth N 9%5D 9A.S.7.3 SGTS Exhaust Stack - Fuel Building The SGTS exhaust stack begins at the (-)5 ft 3 in. level of the Fuel Building and extends through the roof of the building. There )

are no fire dampers in this stack; however, fire stops are pro-vided where the stack penetrates a fire-resistive floor. The stack is fabricated of 3/8-in. steel plate and is 18 in. in diameter. Since the exhaust gases that enter the stack pass through a charcoal filter bed equipped with water sprays that pre-clude a high temperature condition, and the stack must function to maintain safe plant conditions, fire dampers are not necessary or desirable.The k nchono.M CY h 5GNMY5EE no CNCk Oh i: k ado N 4douw. . e. EME4 Shr % doeb Mi fJetW%A(_,

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.7.4 Reactor Building HVAC Penetrations There are two Reactor Building HVAC penetrations. These ducts are 42 in. diameter and are manufactured from 3/8-in. seamless SA106 grade B pipe with the divisional isolation valves welded to the pipe. Transition to ductwork is provided downstream of each valve.

l The valves are positioned on either side of the Reactor Building ,

! wall. Fire dampers are not provided for these penetrations. Y CC G N f6 cJr uccMg M k. 4ktst_ Obe.5, Skew \& 4 ks.\ %

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GESSAR II 22A7007 238 NUCLEAR ' ISLAND Rev. 0

- 9A.5.7.5 Reactor Building HVAC divisional isolation valves within the Reactor Building are located in opposite quadrants of the building. Since the Reactor Building has no fire separations and is considered one fire area, fire dampers are not provided. There is no HVAC penetration between the Reactor Building and the drywell portion of the Reactor Building. 5kk_ C4 hM h% opc A or OlDMS, Ca.nn6b pres >et4 Gok_. 6 k h t\. 6r N NCLC.46e.

9A.5.7.6 Auxiliary Building: (-) 32 ft 0 in., Zone 1, Col. E-ll There are two air-operated divisional valves located 3 ft 6 in.

apart. Failure of these valves as a result of an area fire would result in a loss of corridor ventilation at this level of the building. Rooms containing ECCS equipment would not be affected, since they are separately exhausted from this system. The dedi-cated smoke removal system would provide ventilation from this area; safe reactor shutdown would not be prevented. The area is provided with POC detection.

9A.S.7.7 Auxiliary Building: (-) 32 ft 0 in., Zone 1, Col. D-ll The normal ventilation system for the ECCS areas is supplied by divisional valves located 4 ft 0 in, apart in this area. Loss of both of these valves as a result of an area fire would result in loss of ECCS room pressure co b he SGTS can provide pressure control for these a if normal exhaust is lost as a result of fire or o b ystem failures. The area is provided with POC .

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9A.5.7.B Auxiliary Building: (-) 6 ft 10 in. N MC-t Two divisional SGTS valves are on 5-ft centers in this area.

Failure of both of these valves as a result of fire or other 9A.5-25

erar2ns NR7AWTF 238 NUCLEAR ISLAND R v. 0 9A.S.7.8 Auxiliary Building: (-) 6 ft 10 in. (Continued) occurrence would result in failure of the SGTS to exhaust the ECCS areas. Since these valves and the area are protected with wet pipe sprinklers and POC detection, it is unlikely that the. valves would be affected by fire. The normal valve mode is fail: closed

~

upon loss of the air-operated motor. Loss of either or both valves would not prevent safe reactor shutdown.

9A.S.7.9 Auxiliary Building: (-) 6 ft 10 in.

Two divisional valves that provide ventilation to the RWCU area are located on 3-ft centers in the corridor. Loss of these air motor-operated valves, as a result of fire or other causes, would result in loss of ventilation to the RWCU area. These valves provide ventilation only if the area is entered for sur-veillance or maintenance activities. The valves are protected by automatic wet pipe sprinklers and POC detection.h MO w nob V Ne.d br 45ct(e. 6ba.4 dour.

9A.S.7.10 Auxiliary Building: 28 ft 6 in. HVAC Equipment Room, Zone 1

, The pressurizing air supply to the Reactor Building has two divisional valves located in this room. These valves are sep-arated by 12 ft. The loss of both valves as a result of fire or other causes would result in loss of pressurizing air to contain-ment. ea e of air from the reactor to the annulus, as a result of this loss, uld be handled by.the annular ventilating system -

and would be rou d to either the plant exhaust or SGT System if high radiation oc urs. The HVAC room is provided with POC M.e detection systems. g g oggp biW

. e4 h 6p nv dtCS no r i 9A.5.7.ll Fuel Building: (-) 17 ft 0 in.

There are two divisional valves located on 8-ft centers in this area that control room ventilation to the divisional SGT system.

9A.5-26

GESSAR.II 22A7007 l 238 NUCLEAR ISLAND Rav. O  !

9A.5.7 ll Fuel Building: (-) 17 ft 0 in. (Continued)

Loss of these air motor-operated valves, from fire or other accidental causes, will result in the loss of comfort ventilation in the SGTS rooms. The SGTS is separately cooled; therefore, there would be no loss of SGTS availability nor would safe shutdown be prevented. The area is provided with POC detection.

9A.S.7.12 Fuel Building: (-) 17 f t 0 in.

There are two divisional valves located on 8-ft centers, separated from the valves discussed in Subsection 9A.5.7.ll by about 20 f t.

These air-operated valves are part of the outside air system that supplies cooling air to the SGTS charcoal filter beds. One valve is normally closed except when needed for system operation. The loss of these valves, from fire or other accidental causes, will result in lack cf ability to provide air cooling to the SGTS. The SGTS is provided with water sprays in the charcoal filters that can be initiated manually upon high temperature alarm should the outside air system be inoperative. POC detection is provided.

SGTS ( A) is located in a separate fire rated compartment from SGTS(B), so that a single failure would not fail the entire system. ,

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9A.5.7.13 Fuel Building: (-) 5 ft 3 in.

There are two divisional valves located on 8-ft centers, that are part of the fuel pool air sweep system. This system is used only -

when low level radioactivity is present in the fuel pool. Loss of

, the system would result in loss of sweep air. High radiation level in the Fuel Building exhaust is monitored and, if detected, the air is sent to the SGTS. The valves are protected by an automatic wet pipe sprinkler system and POC detection is provided. N42- Poo\ 6toceP b f\ \6 htk E6 \ted M 64ht 6hOu3O ,

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9A.5-27

mennwru 238 NUCLEAR ISLAND Rev. 0

]

l 9A.5.7.14 Fuel Building: (-) 5 ft 3 in.

There are two divisional valves, located in an identical configuration as described in Subsection 9A.S.7.12, and which per-form the same function. This building level has wet pipe sprinkler provided as well as POC detection. The analysis and justification is the same. The fire dampers in the common duct, which feeds

.both divisions of SGTS, could close as a result of fire in an arca which feeds SGTS. However, since fire and LOCA are not required to be considered concurrently, the SGTS is not required for safe shutdown under fire conditions.

9A.S.7.15 Fuel Building: EI 28 ft 6 in.

There are sever 51 sets of divisional valves, located on this mez-zanine floor level, associated with the fuel and reactor building ventilation system. These valves are located in pairs, about 4 to 12 ft on centers, and there are three pairs spatially.sep-arated from each other around the building walls. The loss of any pair of valves, from any cause, would result in using the'SGTS at a reduced flow rate to pick up the affected system. The area has POC detection. No plant operations are performed on this level and the possibility of~ Iire is remote.h operWonct\ cerd. Mons Ck

%.se., $ms acusncV precenk <cctQe. hWor\ ONier 9ste Gof\AMon5-9A.S.7.16 Control Building: El 28 ft 6 in.

• l l

The HVAC equipmenb systems for the Control Building are housed in two 3-hr fire-resistive rooms. The systems are divisional and

. 100% redundant. There are two valves of the opposite division in i

I each area. The loss of either division, for any cause, will result l

in transferring the HVAC load to the standby division. Both divisional areas have POC detection systems.

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Table 9A.5-2 NONFIRE RATED DOORS Door No. Building and Elevation Location - Function Design Criteria R0 1 ReacLes ( -) 31 fL 7 in. Exteries Wall A.u.uluo F ussume Omisaic

-R-0 0 Reacter (-) 5 f L 3 in. Eatei-ics Wall A.n.ulus Fiessere Ceisaic P 11 7 Eccctor 11 ft 0 in. Extcrics ilall Pcrconnci Oulkhcad Occrc in Pressure Vessel F-Si-S Eccctcr Sd ft 7 in. Extcricr t';11 Pcrcc.nci Sulhtcad Occrc ir -

O'\N- Prc,scurc "cccel u F-7-4 Fuel 0") @4 v  %

fL 7 in. @ecendcy;a.qhhinmenY

.Lcracr J Q .

<e Pcrcc.nci TulkEe'a9 r ,

oors in w

CD Pressure Vessel z e F-27-12 Fuel (-) 32 ft 0 in. CO Secondary Containment Seismic, Pressure, Watertight O$

Nm Y F-7-25 Fuel (-) 5 ft 3 in.

w Secondary Containment Seismic, Pressure ""

F-27-31 Fuel (-) 5 ft 3 in. Secondary Containment Seismic, Pressure ws mM F-29-3 b Fuel (-) 5 ft 3 in. Exterior Wall - Exit Tornado Pressure,

1. Seismic D F-25-33 Fuel (-) 9 ft 0 in. Exterior Wall - Cb Car Tornado Pressure, Mis-sile, Seismic F-7-43 Fuel 11 ft 0 in. Secondary Containment Seismic, Pressure P 0 44 Tucl 11 ft 0 in. Fuci Trancfcr nocr Chicld Shielding, Scir-ic, thuey Prcccurc F-27-45 Fuel 11 ft 0 in. Secondary Containment Seismic, Pressure F-7-53 Fuel 28 ft 6 in. Secondary Containment Seismic, Pressure F-27-57 Fuel 28 ft 6 in. Secondary Containment l

F-4-6 [ Fuel Seismic, Pressure zU' m>

51 ft 7 1/2 in. Exterior Wall, Exit Tornado Pressure, <

,g

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SNo.d(nI A. Ehriec c\eoc, nny regwed % b 9we, ccdesh No rnee k CPeaq.5-1

Table 9A.5-2 NONFIRE RATED DOORS (Continued)

Door No. Building and Elevation Location - Function Design Criteria F-5-68 Fuel 84 ft 7 in. Secondary Containment

? seismic, Pressure F-4-72 Fuel 101 ft 5 1/2 in. Exterior Wall, Exit Tornado Pressure, Seismic A- -49 Auxiliary (-) 32 ft 0 in. Secondary containment Watertight, Pressure, Seismic A-10-3 Aux ry (-) 12 ft 0 in. Secondary Containment Watertight, P ure, w Seismic $

A-ll-4 Auxiliary (-) t 0 in. Secondary Containment Wa ight, Pressure, 2

. A-14-6 Auxiliary (-) 32 ft 0 in.

eismic $$

condary Contai Watertight, Pressure, i

Seismic hh-wz.

U A-13-5 Auxiliary (-) 32 ft 0 in. Second ntainment Watertight, Pressure, Seismic Q[

A-15-7 Auxiliary (-) 32 ft 0 in. g Secondary Containm. Watertight, Pressure, Seismic $

A-17-9 Auxiliary (-) 32 0 in. Secondary Containment tertight, Pressure,

. Set ic A-19-10 Auxilia (-) 32 ft 0 in. Secondary Containment Waterti t, Pressure, Seismic A-20-11 uxiliary (-) 32 ft 0 in. Secondary Containment Watertight, Pr ure,

, Seismic

(_ 2-12 Auxiliary (-) 32 ft 0 in. Secondary Containment Watertight, Pressure, 2 N eismic w A-16-25 Auxiliary (-) 6 ft 10 in. xw, Exterior Wall - Exit Tornado Pressure, @[,*

Watertight, Seismic - o oa

Table 9A.5-2

~

NONFIRE RATED DOORS (Continued)

Door No. Building and Elevation Location - Function Design Criteria A-10-27 Auxiliary (-) 6 ft 10 in. Secondary Containment Pressure, Seismic A-19-28 Auxiliary (-) 6 ft 10 in. Secondary Containment Pressure, Seismic A-14-29 Auxiliary (-) 6 ft 10 in. Secondary Containment Pressure, Seismic A-19-37 Auxiliary (-) 0 ft 10 in. RHR Blowout Panel Pressure, Seismic A-9-42 Auxiliary 11 ft 0 in. Vestibule - Exit Tornado Pressure, Seismic

! A-9-52# ' Auxiliary 28 ft 6 in. Exterior Wall - Exit Tornado Pressure, Seismic c

> n ssu

]Iro g Auxiliary 29 ft 10 1/2 in. Steam Tunnel, Eight Blowout Panels h((gtPanes '

o P s-

"3 E A-10-62 sure Relief to Steam

" Tunnel y C-19-14 Control (-) 6 ft 10 in. Exterior Wall, Equipment Tornado Pressure, Mis- E Doors sile, Seismic $

C-19-13 Control (-) 6 ft 10 in. Vestibule, Exit Tornado Pressure, Seismic C-19-5 ' Control (-) 6 ft 10 in. EL: ;rior Wall, Exit Tornado Pressure, Seismic C-19-17 Control (-) 6 ft 10 in. Corridor, Equipment Doors Oversize Doors for Equipment Removal C-15-18 Control (-) 6 ft 10 in. Cable Tunnel Bullet Resistive, Seismic C-15-20 Control (-) 6 ft 10 in. Control Room Bullet Resistive, z Seismic $

C-22-21 Control (-) 6 ft 10 in. Computer Room Bullet Resistive, Seismic

Table 9A.5-2 NONFIRE RATED DOORS (Continued)

Door No. Building and Elevation Location - Function Design Criteria C-15-22 Control (-) 6 ft 10 in. Cable Tunnel Bullet Resistive, C-25-44 Seismic Control 11 ft 0 in. Exterior Wall, Exit Tornado Pressure, RW 43-4 Seismic Radwaste (-) 6 ft 10 in. Exterior Wall, Truck Dock Shielding, seismic 00 1.- Cic;cl Ccacratcr ; '

C ft Extcricr "all, Equirent 10 .. Civisica 1 Oc.T.c'ic1, Exit Terr E^4" i" fc Pr re"r . w DG-1-6 c"o

> Diesel Generator (-) 28 ft

. . . . Vestibule, Exit Tornado Pressure, 2 w 6 In. Division 1 .

co

• Seismic OM

, OC-5-1 Diccci Ccncratcr (-) 5 ft Extcricr h'c11, Equip;cnt cm us w

i 10 in., Divi;ica 2 R;acyul, Exit

"'crncdc Prcccurr ,-

Oci;.7.ic py DG-5-3# Diesel Generator (-) 6 ft :c :o Exterior Wall, Exit Tornado Pressure, 10 in. Division 2 Seismic y[.

DG-5-7 Diesel Generator (-) 6 ft g Vestibule, Exit Tornado Pressure, 10 in. @

Seismic Oc ^ 1 Diccc1 Cencrater (-) E ft Extericr Mcll, Equip 0- t Ter-de Pr^

10 ... Divicics 2 Rc.-.cval, Exit rr".

Scic=ic DG-10-9 Diesel Generator 28 ft Vestibule, Exit 6 in. Division 3 Tornado Pressure, Seismic UM i

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. O l O l O -J l

ATTACHMENT NO. 2 PROPOSED RESOLUTION OF CONTAINMENT SYSTEMS BRANCH DISCUSSION ITEMS l

1 i

Item 3a i

ThefollowdgchangewillbemadeinGESSARSection s e. e . 6.2.1.6 1.4, Page 6.2-70.

The first Htte of that section will read: "The drywell i<,ssujectedtogc.

g preoperational and periodic low pressure integrated leak rate tests to confirm continuing adequate leak-tightness."

~~

,- ~~

~~

GzzsAR II 238 NUCLEAR ISLAND 22A7007 6 Rev. 0 6.2.1.6.1.3.2 High-Pressure Leak-Rate Test

(

Inmediately following the high-pressure structural proof test, the k_I drywell pressurization source is shut off and the change in dry-well pressure and temperature is monitored for the next 30 minutes.

The drywell pressure and temperature decay information is used to establish that the drywell leak rate is less than the allowable value. The drywell air-flow rate from the 1-hr structural test holding period is used as a gross check on the drywell leak rate.

Figure 6.2-37 shows the expected pressure decay rate for the drywell from the 30-psig starting point, the possible effect of temperature, and the calculated allowable and technical specifi-cation limits.

The figure demonstrates that adequate accuracy in the drywell leak rate can be obtained by a 30-min test.

The acceptance criterion for the high-pressure leak-rate test is demonstration that the drywell has a bypass-A/VE"of less than 10%

of the A/YTI value for bypass capability under DBA conditions' (i.e. ,

less than 10% of 4.3 ft2 or 0.43 ft 2),

6.2.1.6.1.4 Post-Construction Drywell Test (4 g.

+ * % s- e y ., J A &

The drywell isssubjected togperiodic low pressure integrated leak p rate tests to confirm continuing adequate leak-tightness. The

( frequency of these tests will be identified in the technical speci-fications.

The differential pressure selected for the periodic tests is sufficient to simulate controlling SBE conditions, but slightly less than the differential pressure required to clear the top row of horizontal vents.

That is, the head of suppression pool water above the top row of horizontal vents, under test conditions, is sufficient to seal the vents without having to install temporary closures.

(

f 0.2-70

- -----a --

,,, ,o,, , 9 l

0 9 e

o - a e aw $m-- . . GESSk& W G. 2. l.G.2, f e. 2 -72 J G 2.I.7, fy G . 2 -7+ Ak #M$

w m M f M p: -

& } ~ _ A ~ - A ~ , A ..

t I

L

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0

'~~'

UEN 3 b gp The acceptance criteria for the bypass A/VR~for the drywell at 3 psig is less than 10% of the A/VE value of 1.45 ft2, as calculated in Subsection 6.2.1.1.5.5. Figure 6.2-39 shows the expected pres-sure decay for the drywell, assuming several leak rates and rates-of-temperature changes. The figure demonstrates that the low pressure leak rate test can be completed within the 30-min period and the gross effects of temperature change can be accounted for.

6.2.1.6.2 Post-Operational Leakage Rate Tests The containment vacuum relief valves will be tested once a year.

The leaktightness of the valves will not be tested separately but will be tested along with the entire containment, during the con-

,tainment leak rate tests. Operability of the vacuum relief valves willbeverifiedbyposition-limitswitchesonthgvalves,after ve thevafM"gasbeenactivatedlocallyorremotely. Accessidu retrashJ an% va c ao m u ue s .rvs 1%svets su n oda s wru s nrut"*nstr""~

a A soutan sue.m s ys sr s aufvnsomr sa nt< vro Ro n s t a m ak r.s e r r m * " *** * * *

• Fordescriptionsofthecontainmentintedratedleakratetest 7;'i'f, , I (ILRT) and other post-operational leakage rate tests (10CFR50 ** AcMtwd6 Appendix J tests Type A and B) see Subsection 6.2.6. ^ '

6.2.1.6.3 Design Provisions for Periodic Pressurization In order to assure the capability of the containment to withstand the application of peak accident pressure at any time during plant life, for the purpose of performing integrated leakage rate tests, -

close attention has been given to certain design and maintenance provisions. Specifically, the effects of corrosion on the struc-tural integrity of the containment have been minimized by the use of stainless steel cladding in the suppression pool area. Other design features which have the potential to deteriorate with age, such as flexible seals, will be carefully inspected and tested as outlined above. In this manner, the structural and leak integrity of the containment will remain essentially the same as originally accepted. '

6.2-72

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

{M 3(o The acceptance criteria for the bypass A/fK- for the drywell at 3 psig is less than 10% of the A/YK valon nf 1 49 ft_2, as calculated

(. in Subsection 6.2.1.1.5.5. Figure 6.2-39 shows the expected pres-sure decay for the drywell, assuming several leak rates and rates-of-temperature changes. The figure demonstrates that the low pressure leak rate test can be completed within the 30-min period and the gross effects of temperature change can be accounted for.

v i s usury r a ana.r N f e h* s u m *

• rY s s

• rtw A 'L PJ Mo<. 1 MLeuss.uc w.uam e e. pmtas c av isc w a s.c o s va m a a os s.uu st. [hi.Ju s,c, f. w 4 ) . ,

6.2.1.6.2 Post-Operational Leakage Rate Tests Twen.c.r A -,emmesa otr6J ts-um.u sw 444 *g ** % Oo s sa e.6.s.4 s4dce

(.

pha pe49 e 6 4 M4F4A*ML' The containment vacuum relief valves will be tested once a year.

The leaktightness of the valves will not be tested separately but will be tested along with the entire containment, during the con-tainment leak rate tests. Operability of the vacuum relief valves will be verified by position-limit switches on the valves, after the valve has been activated locally or remotely.

For descriptions of the containment integrated leak rate test (ILRT) and other post-operational leakage rate tests (10CFR50 Appendix J tests Type A and B) see Subsection 6.2.6.

6.2.1.6.3 Design Provisions for Periodic Pressurization In order to assure the capability of the containment to withstand the application of peak accident pressure at any time during plant life, for the purpose of performing integrated leakage rate tests,

( close attention has been given to certain design and maintenance provisions. Specifically, the effects of corrosion on the struc-tural integrity of the containment have been minimized by the use of stainless steel cladding in the suppression pool area. Other desi'gn features which have the potential to deteriorate with age, such as flexible seals, will be carefully inspected and tested as

, outlined above. In this manner, the structural and leak integrity i of the containment will remain essentially the same as originally accepted.

6.2-72

6.2.1.7 Instrumentation Requirements (Continued) b

-(

. containment. Similar transmitters, which sense containment-to-

,,j

' shield-annulus differential pressure, are initiating , inputs to 4

( the Containment Vacuum Relief System. hwun tr.g us.5 VMS y r I (w ww .v.a.

Au_i s s d t t

.t % e *r w w = *

• m4r+. 1 h M 5-*E F v 4 a V t. .s g 74.m 2. h h 4 t.

+

L vssa w f /aY8*Po27*

In3 N E funNt o f *'-w'MCd or5 Y M eratures Ure TnN ts to the Leak Detection system. Four thermoCouples are mounted at wg appropriate elevations of the drywell space, and 12 thermo- ' # t, couples monitor drywell HVAC differential temperatures. Six-(

teen thermocouples are mounted in the containment RWCU rooms.

Four suppression pool-level sensors are immersed in the sup- '

pression pool water, and the assocaited level trar}sducers are mounted above the water level. The level signals are transmitted ,

to SPMU System logic in the conti ol room. Eighteen thermo-

- couples are immersed in the suppression pool wate'r. Suppression i pool temperature readouts and alarms are located in the -

control room.

Two hydrogen analyzers are mounted in the drywell, and two tr. n.,. : M in the containment. Each analyzer draws a sample from an appropriate area of the drywell or containment.

] g, Hydrogen concentration alarms and recorders are located in the control room.

t Radiation cetectors are mounted in the containment ventilation exhaust ducts. Radiation monitors and containment isolation trip circuitry is located in the control room.

Refer to Section 7.2 for a description of drywell pressure as an 0

input to the Reactor Protection System, and Section 7.3 for a description of containment and drywell pressure, containment-to-6.2-74

4 a

9 g g g U M ,u .

0

• GESSAR II 22A7007 238 NUCLEAR ISLAND R:v. 0 6.2.3.3.1.2.5 Residual Heat Removal (RHR-C) Compartment (Continued)

There are no high energy lines in the RHR-C compartment; therefore, the DBA for this compartment is the moderate energy line crack of the steam condensing line in the RHR-B compartment.

6.2.3.3.1.2.6 Main Steam Tunnel The Auxiliary Building main steam tunnel is located in between the k 48-in. concrete walls separating it from the RHR-A and B compartments. The steam tunnel houses the high energy and highly radioactive main steam and feedwater lines along with some por-tions of the high energy RCIC steam bypass lines, RHR steam con-densing lines, and RWCU piping. ~

The DBA for the Auxiliary Building steam tunnel is the double-4 ended break of one of the 26-in. main steamlines which route from the vessel, into the tunnel and through, into the Turbine Building.

2 There is 658 ft of open vent space into the Turbine Building in the event of the postulated DBA or any other high energy line break occurring in the Auxiliary Building compartments.

6.2.3.3.1.3 Design Evaluation Blowout panels are used in place of open vent pathways when the environmental conditions of one compartment must be isolated from the environment in another compartment, for the benefit of person-nel during maintenance periods. The RWCU pump and valve room, and the RHR-A and B compartments utilize one-way blowout panels for this purpose. The panels are designed to open upon a differential pressure of 0.25 psid. The panels are assumed to be fully opened after 0.1 sec following their release. fe</orm*4e* fos /! .rh a //

p s. fe. r 4vm s.d. k a u s ta.. c y +ks% +%- 9 tssai s will opt a a ge cofon d t 'eswc

( ao. oc c a s o . +,

• c mag c a.o o c a o. m = 4 6, 's s m s s /4u ,

The RELAP4 computer program was used to calculate the mass and energy release rates and the resultant compartment pressures and 6.2-93 6

e Item 5a G

The following note is to be added to Table 6.2-21 on Page 6.2-185 of GESSAR. This note is to be attached to the last column of the table [

th:t pYwhich indicates the status of opening and if they are displayed in the control room. The note is to read as follows: "The applicant will provide that openings not indicated as having status lights must be under administrative control with alarm indications in the control room."

w

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O 4

hem b Table 6.2-21 SECONDARY CONTAINMENT PENETRATIONS - ARCHITECTURAL -

AUXILIARY BUILDING Opening Leakage Opening Status Type (cfm) Elevation Room / Area Number Lights Doors 0 (-)32 ft 0 in. LPCS A-22-12 No RCIC A-20-ll No RHR "C" A-19-10 No HPCS A-17-9 No 0 RHR "A" A-14-29 Yes

(- (-) 6 f t 10 in.

RWCU A-19-28 Yes RHR "B" A-10-27 Yes Equipment 0 (-)32 ft 0 in. LPCS A-15-7 No Hatches RHR "A" A-14-6 No PCIC A-13-5 No RCR "C" A-ll-4 No RHK "B" A-10-3 No HPCS A-9-2 No O (+) 51 f t 11 in. RHR "A' A-50-E No (Roof) A-50-F No RHR "B" A-50-B No A-50-C No Blowout 0 (+) 28 f t 6 in. RHR "A" A-14-59, No Panels A-14-60 No A-14-61 No A-14-62 No RHR "B" A-10-55 No A-10-56 No A-10-57 No A-10-58 No FUEL BUILDING

{ .

Doors 30 (-)32 ft 0 in. Access From F-7-4 Yes Elevator Tower Door 50 (-) 32 f t 0 in. Access From' F-27-12 Yes Stair Tower Door 30 (-) 5 f t 3 in. Access From F-7-25 Yes Elevator Tower Door 40 (-) 5 f t 3 in. Access From F-27-31 Yes Stair Tower

(

Door 30 (+)l1 f t 0 in. Access From F-7-43 Yes Elevator Tower 6.2-185 t

e

1. GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O Men. S A. - - -

Table 6.2-21 SECONDARY CONTAINMENT PENETRATIONS - ARCHITECTURAL (Continued)

• FUEL BUILDING (Continued)

Opening Leakage Opening StatusDI Type (cfm) Elevation Room / Area Number Lights Door 30 (+)l1 f t 0 in. Access From F-27-45 Yes Stair Tower Door 30 (+)28 ft 6 in. Access From F-7-53 Yes Elevator Tower

( Door 30 (+) 28 f t 6 in. Access From Stair Tower F-27-57 Yes Door 50 (+) 84 f t 7 in. Access to F-5-68 No Railroad 50 (-) 5 f t 3 in. Personnel Air- F-25-33 No lock Cask Car Space Equipment 0 (+)51 ft 11 in. Roof . F-51-E No Hatch Equipment 0 (+) 51 f t 11 in. Roof F-51-F No Hatch REACTOR BUILDING Door 50 (-) 32 f t 0 in. FB to Annulus R-8-1 & No F-8-1 Manhole 50 (+)35 ft 3 in. Steam Tunnel 85 No (AB) to Annulus N . 4. t:

( g) T he a p r u e c.n o T to*bb (AoV80e k'k ofE480CJ d' #d O 8 04 ILO f As h e.v . 3 4 1 4ws lis k i .r wws 4 b e. mod. u d s, , , ,4 ,,,, (, l, y c.ow V o \. wi+b <J c ~ . 4 i <_4 h .', u i w ha ce4 e. t v co .

(

6.2-186 1

'[ Ac.(i)

And %Q 9

0

, f k i b GESSkR. Tolku C.2-20 A G2'W , d  ? i pnu C. 2-42 h 5-$+, *

^4#"

1) The 4" SPCU from demineralizer line will be added to secondary containment Table 6.2-20.

H/C-5 < d

2) E4ght-4ech RCIChom condensate stora will be added to Table 6.2-24 as a footnote. His-feetnote wi+1-be-ettached-to^

penetration-No.-16C-end-HC8

3) A note will be added to Figure 6.2-48 on GESSAR Page 6.2-279 indicating that the Type 4 penetration has a sealing system on the inner set of i

valves. l TL

~ Gesska _,p. ,

l l

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O Men SC[i)

Table 6.2-20 SECONDARY CONTAINMENT PENETRATIONS AUXILIARY BUILDING PIPING Isolation Sleeve Seal Line Number Scheme (1) Penetration Type (2) Type (2) 8-in. COND4-AAB Type 4 107 AP 2 F 6-in. MS206-AEC Type 3(3) 118 AP 7 F 1-in. IA7-ADC Type 2 108 AP 1 or 2 F 1-in. SA109-ADC Type 3 109 AP 1 or 2 F 1-in. COND84-AAC Type 3 260 AP 1 or 2 F 1-in. SAll4-AAC Type 3 288 AP 1 or 2 F 3-in. CRWil-AEC Types 5 & 7 457 AP 1 F 6-in. DRW20-AEC Types 5 & 7 456 AP 1 F 1-in. RCIC27-AHC Type 4 (4) 120 AP 7 S 6-in. RCIC7-EAB Type 4 (4) 119 AP 1 or 2 P 4-in. COND21-AAC Type 6 117 AP 1 or 2 F 16-in. ESW28-ADC Type 8 106 AP 1 F 3-in. DMW21-ABC Type 1 113 AP 1 or 2 F 16-in. ESW26-ADC Type 8 105 AP 1 F 1-1/2-in. ESW34-ATC Type 8 111 AP 1 or 2 G 4-in. ESW30-ADC Type 8 114 AP 1 or 2 F 1-1/2-in. ESW35-ADC Type 8 112 AP 1 or 2 S

'( 4-in. ESW29-ADC Type 8 115 AP 1 or 2 F 4-in. ESW21-ADC Type 8 253 AP 1 or 2 F 4-in. ESW22-ADC Type 8 252 AP 1 or 2 F 3/4-in. ADS 24-ADC Type 8 300 AP 1 or 2 F 3-in. COND22-ADC Type 3 22 AP 1 or 2 F 6-in. ESW352-ADC Type 8 62 AP 1 or 2 F 6-in. ESWl7-ADC Type 8 61 AP 1 or 2 F 3-in. IA6-ADC Type 2 60 AP 1 or 2 F 4-in. SA62-ADC Type 1 59 AP 1 or 2 F 10-in. RHR51-BAC Type 8 58 AP 1 or 2 F 12-in. FPCC94-AAC Type 8 57 AP 1 F 3-in. COND153-AAC Type 1 302 AP 1 F 1-1/2-in. SA150-ADC Type 1 301 AP 1 F 6-in. RHR63-AEC Type 4 Embedded - -

8-in. SPCU7-ABC Type 1 788 AP 1 F 2-in. WPS14-ABB Type 4 795 AP 1 F 3-in. DRW5-AEC Type 1 960 AP 1 F 1-1/2-in. ESW-ATC Type 8 783 AP 1 F 1-1/2-in. ESW196-ATC Type 8 782 AP 1 F 2-in. DMW18-ABC Type 1 749 AP 1 F 2-in. RWCU262-AGC Type 3 Embedded - -

2-in. RWCU277-AGC Type 3 Embedded - -

3-in. DRW47-AEC Type 3 Embedded - -

3-in. CRWl-AEC Type 3 Embedded - -

3/4-in. APS20-ADB Type 8 800 AP 1 F 3/4-in. WPS26-ABB Type 8 799 AP 1 F 6.2-179

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O h w 5c(0 -

Table 6.2-20 SECONDARY CONTAINMENT PENETRATIONS AUXILIARY BUILDING PIPING (Continued)

Isolation Sleeve Seal Line Number Scheme (l) Penetration Type (2) Type (2) 4-in. SA62-ADC Type 1 752 AP 1 F 3-in. IA6-ADC Type 1 753 AP 1 F 3-in. CW306-ADC Type 8 750 AP 1 F 3-in. CW304-ADC Type 8 751 AP 1 F 6-in. ESW17-ADC Type 8 754 AP 1 F 6-in. ESW352-ADC Type 8 755 AP 1 F 2-in. WPS14-ABB Type 8 950 AP 1 S 3/4-in. APS7-ADB Type 8 951 AP 1 S 1-1/2" CW205-ADC Type 5 744 AP 1 S 1-1/2-in. CW204-ADC Type 5 745 AP 1 S 2-in. PLCS8-ECB Type 8 952 AP 1 S 1-in. PLCS9-ECB Type 3 955 AP 1 S 4-in. FPW52-ADC Type 1 793 AP 1 S 4-in. FPW52-ADC Type 1 791 AP 1 S 3-in. DRW20-AED Types 5 & 190 AP 1 S 4-in. RWCU96-AEC Type 4 (4)7 305 AP 1 S 3-in. DRW5-AEC Type 3 960 AP 1 S 2-in. COND160-AAC Type 3 315 AP 1 F 18-in. COND3-AAB Type a 98 AP 1 F 4-in. ESW63-ADC Type 8 130 AP 1 *S 4-in. ESW64-ADC Type 8 129 AP 1 S 16-in. ESW60-ADC Type 8 127 AP 1 F 3-in. ESW70-ADC Type 8 126 AP 1 F 3-in. ESW69-ADC Type 8 125 AP 1 F 3-in. ESW69-ADC Type 8 125 AP 1 F 1-in. SA98-ADC Type 1 774 AP 1 S 3/4-in. APS7-ADB Type 8 961 AP 1 S 3/4-in. APS7-ADB Type 8 966 AP 1 F 4-in. FPW2-ADC Type 1 792 AP 1 S 4-in. FPW2-ADC Type 1 786 AP 1 S, 3/4-in. WPS23-ABB Type 8 962 AP 1 S 3/4" WPS23-ABB Type 8 965 AP 1 F 3-in. ESW54-ATC Type 1 725 AP 1 S 3-in. ESW53-ATC Type 1 724 AP 1 S 1-in. SA98-ADC Type 1 963 AP 1 F 3/4-in. ADS 37-ADB Type 8 963 AP Later F 3/4-in. CRW72-ECB Type 3 535 FAP 3 or 4 Later 3/4-in. CRW73-ECB Type 3 535 FAP 3 or 4 Later 3/4-in. CRW74-ECB Type 3 535 FAP 3 or 4 Later 3/4-in. CRW75-ECB Type 3 535 FAP 3 or 4 Later 6-in. RWCU135-EAC Type 1 105 FAP 3 or 4 Later 6-in. RWCU285-EAC Type 1 171 FAB 3 or 4 Later 6-in. RHR26-BAB Type 8 522 FAP 3 or 4 Later 2-in. PLCS2-ECB Type 8 954 AP 1 S 6.2-18C

GESSAR II 22A 007 238 NUCLEAR ISLAND Re c. O M tm Ic h Table 6.2-20

SECONDARY CONTAINMENT PENETRATIONS AUXILIARY BUILDING PIPING (Continued)

Line Number Isolatgp Scheme J Penetration Sleev Type (g)

Seal Type (2 )

{

1 2-in. PLCS12-ECB Type 8 972 AP 1 5 3-in. ESW170-ADC Type 8 447 AP 1 F

3-in. ESW169-ADC Type 8 446 AP 1 F 1-1/2-in. SA153-ADC Type 1 357 AP 1 or 2 I l 3-in. CW304-ADC Type 9 356 AP 1 or 2 5 3-in. CW306-ADC Type 9 355 AP 1 or 2 5 14-in. RHR32-BAB Type 8 309 AP 7 7 10-in. MS201-ECB Type 8 310 AP 7 7 l-in. RHR136-AAB Type 8 524 FAP 3 or 4 Later 6-in. MS202-ECB Type 8 104 FAP 3 or 4 Later 4-in. COND21-AAC Type 6 103 FAP 3 or 4 Later 1-in. IA7-ADC Type 2 536 FAP 3 or 4 Later 10-in. MS204-ACB Type 8 102 FAP 3 or 4 Later 4-in. RWCU96-AEC Type 3 106 FAP 3 or 4 Later l 6-in. RCIC2-EAB Type 8 523 FAP 3 or 4 Later i 3/4-in. CRW46-EAC Type 3 534 FAP 3 or 4 Later

! 3/4-in. CRW47-EAC Type 3 534 FAP 3 or 4 Later

, 3/4-in. CRW48-EAC Type 3 534 FAP 3 or 4 Later

, 3/4-in. CRW49-EAC Type 3 534 FAP 3 or 4 Later 20-in. RHR20-BAB Type 8 170 FAP 3 or 4 Later 6-in. RWCU4-EAC Type 1 297 AP 1 5 6-in. RWCUS-EAC Type 1 295 AP 1 5 3-in. ESW39-ATC Type 1 746 AP 1 5 3-in. ESWil-ATC Type 1 747 AP 1 5 3-in. ESW53-ATC Type l' 772 AP 1 5 3-in. ESW54-ADD Tfpe 1 773 AP 1 5 12-in. FPCC95-AAC Type 8 6 AP 1  ?

10-in. RHR52-BAC Type 8 7 AP 1 or 2 I 3/4-in. ADS 54-ADC Type 3 312 AP 1 or 2  ?

4-in. CSSW4-ADC Type 8 8 AP 1 or 2 7 4-in. CSSW3-ADC Type 8 9 AP 1 or 2 7 -

1-in. COND76-AA Type 3 255 AP 1 or 2  ?

10-in'."HPCS7-EAB Type 4 (3) 99 AP 1  ?

1-in. COND79-AAC Type 3 132 AP 1 or 2 T l-in. SAlll-ADC Type 3 131 AP 1 or 2  ?

l-in. COND80-AAC Type 3 257 AP 1 or 2  ?

l-in. SA69-ADC Type 1 121 AP 1 or 2  ?

3-in. COND27-AAC Type 3 316 AP 1 or 2 5 10-in. MS201-ECB Type 8 442 AP 1 5 l

14-in. RHR33-BAB Type 8 441 AP 1 5 3-in. DMW22-ABC Type 3 128 AP 1 7 3-in. COND31-AAC Type 3 124 AP 1 T 2-in. CCW21-ADC Type 5 123 AP 1 7 2-in. CCW22-ADC Type 5 122 AP O

1 I l-in. SAll2-ADC Type 3 10 AP 1 5 6.2-181

GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0

$en NC.h Table 6.2-20 SECONDARY CONTAINMENT PENETRATIONS AUXILIARY BUILDING PIPING (Continued)

Line Number Isolatyp Scheme 1 Penetration Sleev Type (g)

Seal Type (2)

.2-in. ESW48-ADC Type 5 597 FP 7 F

.2-in. ESW47-ADC Type 5 598 FP 8 F i-in. CW3-ADC Type 5 544 FP 8 F i-in. CW8-ADC Type 5 545 FP 8 F

-in. HWD17-AAC Type 5 546 FP 8 F

-in. HWDl-AAC Type 5 547 FP 8 F

.-in. SDW215-AEC Type 5 Roof Drain - -

,-in. SDW214-AEC Type 5 Roof Drain - -

i-in. SPCU7-AAC Type 5 578 FP 1 or 2 F

.'4-in. APS19-ADB Type 8 504 FP 1 or 2 F l'4-in. APS19-ADB Type 8 560 FP 1 or 2 F

-in. ESWl87-ADC Type 8 614 FP 1 or 2 F 1-in. ESW192-ADC Type 8 615 FP - 1 or 2 F

-in. CCWl-ABC Analyzed 585 FP 1 or 2 F for leakage I-in. DRW56-AEC Type 3 Embedded - -

i-in. FPWil-ADC Type 5 Embedded - -

I /4-in. APS4-ADB Type 8 621 FP 1 or 2 F V 4 - i n . s fCV 12.- AA C. rype.3 (1) tnbelled _

4- In. srev sr-AAc ryge 3 (c) c ,g,jj,/ ,_ ,,

.3TES:

(1) See Figure 6.2-9247 (2) See Figure 6.2-47.5 2 (3),Lb b

-~

& [

CM -WW 4 T y .

. c, s -r;hpt A ,G c.24 .

(c) n ~~& /~L*L a .

. (c) cgy G O

6.2-l83/6.2-184 L

v s Table 6.2-24 EVALUATION OF POTENTIAL CONTAINMENT BYPASS LEAKAGE PATHS Line Size Penetrating Termination Bypass Leakage Potential Primary Containment Penetration Containment Region (l) Barrier (2) Bypass Path 2C. Equipment Hatch NA S NA No 3C. Personnel Lock NA S NA No 4C Personnel Lock NA S NA No BC. Fuel Transfer Tube NA S NA No 9C-12C. Main Steamlines 26-in. E C, (3), (4) No

! 13C. Steam Isolation Valve Drain 3-in. E C, A, (3) No 16C, 17C. Feedwater Lines 20-in. E C, (3), (4) No 25C CRD Pump Discharge 2-in. E (6) C, (3), (6) No N l 27C. RHR System (LPCI Mode) A - Div 1 14-in. S (5), (6) C (5), (6) No $

28C. RHR System (LPCI Mode) B - Div 2 14-in. S (5), (6) C, (5), (6) No g

< cn 29C. RHR System (LPCI Mode) C - Div 2 14-in. S (5), (6) C, (5), (6) No c: O y 30C. RHR "A" Pump Suction - Div 1 24-in. S (5), (6) (5), (6) No h$

31C. RHR "B" Pump Suction - Div 2 Qy 4 32C. RHR "C" Pump Suction - Div 2 24-in. S (5), (6) (5), (6) No

! co 24-in. S (5), (6) (% , (6) No :U :0

• 33C. RHR Pump "A" Disch to Suppression Pool 14-in. , S (5). (6) (5), (6) No wH j 34C. RHR Pump "B" Disch to Suppression Pool 14-in. S (5) . (6) (5), (6) No "H

35C. RHR Pump "C" Disch to Suppression Pool 14-in. S (5), (6) (5), (6) No 1

36C. Demineralizer Water to G33-2020 2-in. E C, (3) No o 37C. RHR SRV to Suppression Pool 12-in. S NA No i

39C. RHR SRV to Suppression Pool 12-in. S NA No 40C. RHR SRV to Suppression Pool 6-in. S NA No 41C. RHR SRV to Suppression Pool 4-in. S NA No 42C. RHR SRV to Suppression Pool 3-in. S NA No 4

9 43C. RHR SRV to Suppression Pool 3-in. S NA No 44C. RHR Suction FM Recirc 20-in. S C No

}

47C. Steam to RCIC Turbine 10-in. S C, (3) No f//

48C. RCIC pump Discharge Head Spray 6-in. S (5) C, A, (5), (4) No 6) P 49C.

50C.

RCIC Pump Suction from Suppression Pool RCIC Turbine Discharge to Suppression 8-in. S (5) (5) No D v ,, gN Pool ld-in. S V No e :P 52C. RCIC Pump Minimum Flow Bypass 3-in. S (5) (5) No [ $'

O 53C. LPCS Pump Discharge 12-in. S C No o

Table 6.2-24

• EVALUATION OF POTENTIAL CONTAINMENT BYPASS LEAKAGE PATHS (Continued)

Line Size Penetrating Termination Bypass Leakage Potential Primary Containment Penetration Containment Region (II Ba rrier (2) Bypass Path 54C. LPCS Pump Discharge 12-in. S NA No 55C. LPCS Pump Test Line 12-in. S NA No 56C. LPCS SRV Discharge to Suppression Pool 2-in. S NA No 57C. Air Positive Seal to Air System 3/4-in. S NA No 58C. IIPCS Pump Discharge 12-in. S (5), (6) C (5) No ( 7) 59C. IIPCS Pump Suction 24-in. S (5), (6) (5), (6) No 60C. HPCS SRV Discharge 12-in. S (5), (6) (5), (6) No 63C. RWCU Pump Suction From Recirc Pump 6-in. S C No N 64C. RWCU Return to Feedwater Line 6-in. S C No <n 65C. RWCU Discharge to Main Condenser 4-in. E C, (6), (3) No m 68C.

z Containment Supply Purge (IIVAC) 42-in. E C, (3) No g 69C. Containment Exhaust (IIVAC) 42-in. E C, (3) No QQ t* fA 4 70C. Containment Vacuum Relief Outlet 24-in. S C No h$

to 72C. Containment Vacuum Relief Outlet 24-in. S C No WN O

78C. Skimmer Drain to FPCC 10-in. E C, (3) No HH 79C. Demineralizer to FPCC Pool 10-in. 3 L No 83C. 24-in. Pipe Spare 24-in. S NA No 7

84C1 Instrument Line 3/4-in. S NA No O 84C2 -84C4 Spares -

S NA No ll4C. Drywell CRW Sump to CRW 3-in. E C, (3) No 1 115C. Drywell DRW Sump to DRW 3-in. E C, (3) No ll6C, ll7C. 12-in. Pipe Spares 12-in. S NA No ll8C. 24-in. Pipe Spare p

24-in. S NA No (t 119C. RWCU Backwash Drain 2-in. E C, (3), (4) No 3 120C. CCW To Containment 10-in. E C, (3) No 1.' I r . rcW Peturn from Containment 10-in. E C, ( 'll No 124C. 12-in. Pipe Spare 12-in. S NA No p

I ?'ic. NI chiller! Water to containment 6-in. E c, T. No g 126C. N1 Chilled Water from Containment 6-in. E C, (3) No r- M 127C. w V o :p Condensate Dist to Containment 6-in. E C, L No 128C1 3/4-in. Pressure Sensing Line for ILRT Spare 3/4-in. S NA No 4$O 128C2 -

S NA No o O O O

GESSAR II 22A7007 238 NUCLEAR ISLAND R::v. O MM S C Table 6.2-24 EVALUATION OF POTENTIAL CONTAINMENT BYPASS LEAKAGE PATHS (Continued)

NOTES:

(1) Termination Region S = Secondary containment (ECCS Rooms or Fuel Building) . Lines terminating within the secondary containment are not potential throughline leakage path.

E = Environmental, beyond secondary containment. Such lines either pass directly through the secondary containment to the environment, or are connected to branch lines which pass through the secondary containment to the environment.

For either case, potential throughline leakage is pre-cluded by a combination of leakage barrier.

(2) Bypass Leakage Barriers C = Redundant Primary Containment Isolation Valves A = Redundant Secondary Containment Isolation Valves L = Water Leg Seal V = Vented to Secondary Containment with CLOC (Closed Loop Outside Containment, see Subsection 6.5.3.2.1)

(3) Containment Seal Leakage Control System Provided.

(4) Third Isolation Valve (Remote Manual) Provided.

(5) The system generally operates in a closed-loop mode, within the secondary containment. However, there are several lines such as flushing water, etc, which penetrate the secondary containment and offer a potential leakage path from the pri-mary containment to environment. For such case, however, throughline leakage and bypass of the secondary containment is precluded by the following:

a. If the line provides a source of makeup water to the RPV, no isolation is necessary.

b. If the line does not provide makeup to the RPV, isolation is provided by redundant valves at the secondary contain-ment or a single valve with redundant solenoids.

(6) Secondary containment leakage control is provided. Type of protection is shown in Figure 6.2-52 for each individual case.

(7) MCS s. KC& YeEWU$ cdAwa- pp L f Jyq%b&. Lak&Qp A a Je bLe&fLa . 6y Q & M MaAe.cm. 6.2-192

, GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 Iw h(0

. (%.- i C DARY CONTAINMENT ""

QUALITY GROUP C TYPEO1 OUALITY CONTAINMENT me GROUP C q SEAL. S YSTEt1 ATTActtc)

To (N 6oh M VA LV C VACUUM ~ ~

BREAKER (TYP)

QUALITY

- OlV 2 ~

OlV 1

" GROUP ~~ ~~

C

]

Q n g00 E G >

18in. MIN OUALITY GROUP C

l' QUALITY GROUP C OUALITY GROUP C w

l 20 i O  :

4 WATER LEG q N--N e O SAFETY

[ CLASS 1 '

l LOCA LOCA SYSTEM OUALITY l PUMP MOTOR GROUP B OR C KR

~

KR M , [ lTY p

Div 1 olv 2 ,h 1 O '

o i Figure 6.2-48.

' Secondary Containment Penetration Types for Leakage Control l

l 6.2-279/6.2-280 l

Item Sd c.r.l.t.Iand It was agreed to revise is the GESSAR text in Sections A6.5.1.4.2>rinservice testing on Page 6.5- 5 to read as follows:

=.. 1 ,

. ..p.

e e m

e m .

~

3 GESSAR II 22A7007

, 238 NUCLEAR ISLAND R v. 0 J

(y

., 7 S.1.4.1 Preoperational Testing (Continued) MA.SekS)

'v i

,7

. /

( [f (6) adsorber bed residence time verification test (7) air-aerosol uniformity test (8) in-place HEPA test (9) in-place adsorber test (10) laboratory test of adsorbent (11) electric heater test.

(' After installation of the ESF filter trains, a performance test of system capabilities to meet the specified requirements is conducted. The test is to d monstrate the ability to maintain the prescribed ncgative pressure and the ability to respond to i # " M Ed

• Mb flow transients, 8 d.f required.

W[w~5 h +.*

% Fi to ^ app -W, .t %.ve-p.mm.m. 74 w ' -,-fqb p M gg we 6 N T.1 b D N I Ie cIEchn" T % .st .-

5

<^'"*5 u"DU g% p g .~ b h .h m q . -h x. N #

4 m u % e.e,n sw% *e h 4 % wt A ~f

( Jaftervice2 test-ing' of Lhe=ESF=fi'ltration-eysteMs-conducted- in accordance with the surveillance requirements given in the plant technical specifications,- Chapter 16.

6.5.1.5 Instrumentation Requirements Controls and instrumentation for CBOACS and SGTS charcoal filter trains are discussed in Section 7.3. Each system is designed to function automatically upon receipt of an applicable ESF

( actuation signal. ,

Differential pressure indicators are provided to measure the pressure drop across each filter and charcoal bed. Pressure drop across each filter train is measured and an indication of high differential pressure is alarmed in the Control Room.

L.S . I A .7-.

T>dk

% + 4 m tse & g b ~ a h A M M "

\l s.4...ru'.3 -_ .A M s a w J9 se A . Q M 4%

m n s

_ g s ma nw a %W. a#~ n

% a;<,_2). 6 . % q scrs e w

. g. sp % @ uy h' .

b 1s M -fo--r fm '- 2 "

,x.- qmm ww.y .m.

w~4Q 7mkh M & @ A- ~_ n M d

~m.

- l

.L u A.) ,

T.4& C 2-/$ k ceuot g-LocA

6) GesshK k y' s Auny-L I~L 5-d ,

PAhl Fj . c. S -/0.

c) cup,4 A

. s -9 d/o .

'>) t1% Fj . C.r -l J e

t W

~ '

3, .m Table 6.2-19 DESIGN AND PERFORMANCE DATA OF Tile SECONDARY CONTAINMENT STRUCTURES Shield Auxiliary Fuel Secondary Containment Design _ Annulus Building Building A. Free Volume (ft ) 433,000 sy0,000 400,000 (a)

B. Pressure (inches water) 30(,000 (b) .

1. Normal Operation (-) 5. 0 ( - ) C fo 2 T m a x. ( } },emy

2. Post-LOCA (-) 0.25 max ( ) e. goo m 9 (4) ( ; o.goe m9, (4)

C. Leak Rate at Post-LOCA > 100 (c)  ;>ioo (e) y/p (e) "

Pressure (%/ day)

• y D. Exhaust Fans (c) ba Om F

Number

~

h 1. 2-100% 2-100% 2-100% $s'o

2. Type Centrifugal Centrifugal xx Centrifugal ms E. Filters (in SGTS)* s 6 Prefilters Same as Shield Same as Shield I5 6 HEPA Annulus Annulus 1 Charcoal 6 IIEPA Total 19 per system

< N '

oc t.ocA N

• Flow diverted to SGTS on high radiation gsignal.

Notes: i (a) Above operating floor. (d) At (-) . 5 in 21I 0 normal operating ,,C (b) Below operatiniJ floor.. pressure. w> '

(c) Not including StandbysGas-Treatment System, (e) At (-)0.25 in H2 O Post-LOCA <$

u . a. % o u 4 r , ,".> . "'

pressure.

o$

(-4) o.xg .w the p eps b wedo t.

ei(cdts.

t s

3 5

• % \- k e

,. A I -+ *4 >% -

g , ->

' GESSAR II 22A7007 238 NUCLEAR ISLAND R*v. 0 6.5.1.3.2.2 Auxiliary Building ECCS and RWCU Pump Rooms Pressure Response Analysis

(, '

The Auxiliary Building pressure control exhaust system fan main-tains the ECCS and RWCU. pump rooms under a negative pressure differential of 0.625 in WG during normal plant operation by

' withdrawing the amount of air equal to the in-leakage. Ductwork connection and dunper switch control are provided between the ECCS pressure control exhaust system and the SGTS.

(- q p .. . .,

In the . event of a LOCA coincidental-with the loss of normal AC-M# I,c the AuxiliaI;y Building pressure control exhaust system fan An a e # s u-s e o m A -ru.e w f.

stops., The SGTS fan operates at its ratea flow 27 sec after a

+t.otP LOCA, and picks up the exhaust flow required be maintain the ECCS Jo M &A' and RWCU Pump Rooms under ay a negative mressure. The ECCS equipment

y. .

is also activated to provide4 coolingA to the core in the RPVM 2.7 h-h ,

R.CA c. 4 -A R. w c. %u w & e.w : - b =0 t

Activation of the ECCS equipment generates heat to the space and f consequently causes both pressure and temperature to rise. An exhaust flow rate of 2980 cfm to the SGTS is required. The ECCS Pump Room coolers aye assumed effective when space temperature 0

reaches 122 F. The chronological sequence associated with LOCA

/

signal heneration is as follows:

I .b a . , ts

! )

i

• /

. ~

'T = 0',sec. 'All dampers ~ assume their failure mode position and

• ~'

/ the#ECCS pressure control syctem fans stops.

[ ~

g.,

/

T = 10 sec. Diesel Generators start and provide emergency M NW.7 E" y Power. E4 % b M 9 1-4 * %,f,3 w"Y gy w,ty .

'/ y , p % .?p pM Sm ~L.km  % ~oc6. ) 0.ide " (J .G m % 4.j'C O Ao m --

-% ., </

d% ccff m o..

. 0 m.-2-7-eeo . SGTS' fans have sa tarted M .C k-_^=% 'and reach their rated

/ capacity, associated dampers are at least three-fourths s'

open allowing th,e SGTS fan to draw 2980 cfm from the ECCG

.l ., Wf&W e ~

A &+- %p e -~ _

+>- m-

,: p w,m e e-u~

l. .2 f.M-a ta -w a auss.

~. -

6'.5-9

> WE' N

GESSAR II 22A7007

. 238 NUCLEAR ISLAND Rev. 0 6.5.1.3.2.2 Auxiliary Building ECCS and RWCU Pump Rooms Pressure Response Analysis (Continued)

L' T = 27 see to 122 sec.

v. ee u r.nm P^**a=

e..e . . ~.

ccenng '"^6e , 2980 cfm

\

exhausted to SGTS. V hA-- M 4 '- %

Q ,

< t 3 e.ge o

• w 4 is t fe. 4 r. m E c c.s 4.u.s e T = 122 to 164 sec.

c4 sc.r.: Cio q Intermediate cooling mode.r:duced.

u sp. % p,ec3 % v t_ ., i n heat

c. p <(.3.m %:l generatien, 2025.- l L fa cxhcucted ' M TS.

% cL 7 /f b A 7mA p_:-> t_ n. .. o . A @ A b 3 ^ ^ * --

.A ni  % t 'f .%.m cce.s Lu m

(..

A T = AfL4 to 744 sec.

53 6b Laeracdiate cooling mode with-ECCS sc 3.H L.pcr < m m a t u.* n La cs con 1are, 946 cfm a.rr s w .vhausted p2.rectl=y _ to SGTS.

e e w e, o c.mc4 p- - P c._ ( . i o.s uo w.c. i v r., su % '7 4 A r,~ s cc s a.. % co u.a ma c e_ . 9 3 r q A sm n-..

T - Ecyond---7-74-sec . Long-+ arm _ cooling mnda 40-cfm.

A & I i. y 4 (L.Attc w m....L t2r 4 a.M.,c.~ c i b g / .

p <. c.. ) o.m a v.c.

The te "perat.ure-response-profile for the-ECCS-end-RWCU Pump Rooms is-shown~in Figure-6.5-10.-

q ;34 % F-s a ft. c.c.t A o% ca o I. 4 t.

,,7 ,.,9 % c .4 A.W) . A.mt Q 3 4mAlcfC m3Mwo 1124 _ZF o.S t E3 4mm.c L T16 1 F o Ac l ~- 1-cu N 4*k @

, 6.5.1.3.2.3 Fuel Building Pressure Response Analysis # - l- 2 0 8 '" ' '" A

\

i i

The Fuel Building pressure control exhaust system fan maintains i , ca the Fuel Building under a negative pressure dif ferential of 9r625-in WG during normal plant operation by withdrawing the amount of air equal to the in-leakage. Ductwork connection and damper switching control are provided between the. pressure control system and the SGTS. In the event of a LOCA coincidental with the loss of normal AC power, the Fuel Building pressure control system fan

( stops. The SGTS fan operates at its rated flow 27 sec after a LOCA and exhausts 1645 cfm continuously from the Fuel Building to maintain a subatmospheric condition. The pressure and temperature response profiles for the Fuel Building are shown in Figures 6.5-11 and 6.5-12, respectively.

I 6.5-10 t

ast>ln c HH NM>Jo

- a-NWc o Zc@t>:n D n HihzO 0 2o<* o 0

0 8

4 0

l 0

- 2

- 4 e

s 0 n 0 o I

6 p 3 s e

R e

r 0 u 0 s I

0 s 3 e r

- P g

n i

0 d 0 l I

4 i 2 u B

m l f

c ) e 5

c e

u 4 s F 0 (

6 1

I 0 E A 8 C T 1 W O A

l T -

L S -

T t G , s S o 0 P O I 0 T E 2 I

L .

E T ,

A 1 A C 1 R S W F

/

5 O O .

L 0 F E I 0 6 G 6 T N e S A U H r A

H C u g

X E i c F e T I 0

N <

? 6 A

!

T S

N O

C i

0

_ 3 e E e R Y s /

7 2

UA S  !

O EE SC RD T /

O P 0

e

l. 0 0

s. 4 0 os

+g  : 9 te

1. H ,

~ d4 e

o0

,o i

I .e-

- +m 1

_o$= w=3 ewe n.wi c

a*

9a) After review of the design, we concluded that we cannot meet single failure of the Division 1 power because we lose both the jockey pump and the main pump. Therefore, we must commit to an air test. We willimplementthefollowingontheRCIC,LPCI[AB&C,LPCSandHPCS discharge lines. We will add requirements to drain the discharge piping on these systems and perform Type A tests or Type C air tests and the resulting leakage shall be added to the Type A leakage for acceptance. f g C,2 2 1 A g 4, 9

I

l

.~ C l - ,

. Table 6.2-29 Nh9 tt ,

I CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) ,

i, i Inboard Isolation Outboard Isolation l

Barrier Barrier Penetration Bellows Test Barrier Description / Barrier Description /

Number Description Seal Type d Valve Number O"1) Notes Valva NumberDl) Notes 2c Equipment Hatch No B Double O-Ring 1 - -

3c Personnel I4ck - Iower Inner Door No B Double Gasket 1 - -

Outer Door No B - -

Double Gasket 1

, Barrel No B Inner Door 2 Outer Door 2 4c Personnel Lock - Upper Uco Inner Door No B Double Gasket 1 - -

Z

, Outer Door No B Double Gasket Barrel No B

. 1 NO i

Inner Door 2 Outer Door 2 Qy o" 8c Fuel Transfer Tube No B Double O-Ring 1 - - ##

A HH 9c Main Steamline D Yes C B21F022D 34 B21F028D 3 Yes C B21F067D 3 a Yes ,, B21F006D 3 U Main Steamline B loc Yes Yes C C B21F022B 3$ B21F028B B21F067B 3

3 k*

Yes C 4 B21F006 3 11c Main Steamlin'e A Yes P B21F022A 3[ B21F028A B21F067A 3

3

,k B21F086 3

.j A

12c Main Steamline C Yes C B21F022C y B21F028C 3 B21F067C B21F006 3 m" C 3

$[

. o O

O4 4 "T*<t e. A mW At so <ts!roevia w O

w g, cO " -

i g,) -

.m i UEd 9A - <

Table 6.2-29 MNM '

CbHTAINMI:NT PE.lETRAYII.N AND CONTAINMENT ISOLATION VAIVE LEAKAGE (Continued)RATE T Inboard Iselation Outboaril Isolation Barrier Barrier Penetration Bellows Test Barrier Description / '

Number Description Seal TypeT' Barrier Description /

_ Valve Number Notes Valve Ntsaber Notes 13c steamIsolaIlonvalve Yes C B21F016 '3 B21F019 Drain 3 C, (

B21F085 3 16e Feedwater Line Yes C B21F010A j K.

C '

B21F032A 5j f6 )

B21F065A 51 g 1I B21F102A 5

,N 17e Feedwater Line Yes C B21F010B Sj iG B21F0328 5 h.2

, C, J B21F0658 5 M c B21FF102B g

Jr 5 'r p b1 25c CRD Pump Discharge Yes v' C C11FF215 6 C11F083 6 C C11F122 6 s tn 27c RilR System (LPCI Mode) Yes C E12F042A jf3;S{G E12F027A Line A Division 1 4 f]iS,,7ihl ts 7

C#

7 2Rc Rl!R System (LPCI Mode) Yes '

E12F042B j l E12F027B 358 '7 Line n Division 2 1 7 29c RilR System (LPCI Mode) Yes Cg E12F041C p tIr Y E12F042C 6(

4 Line C Division 2 l r

l ,

30c RIIR A Pump Suction Division 1 No C d' Closed System Nl $ E12F004A

• [@

b

, I 31c i RiiR B Pump Suction Division 2 No CM Closed System ,

E. E12F0048 '

PA in I

32c R11R C Pump Suction No C M* Closed System $ f E12F105 I4 Division 2 1r V

1. Tec A w.w A.cso v.Eco ct.sp i t> r* LEM HoTE.9 C>mt.tt mm j

• • Rees %m vium on,ea Tw A rn,-

w ~,. se ~

f p ~

Tablo 6.2-29 T{e (

CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE RATE TEST LIST (Continued) .

Inboard Isolation Outboard Isolation Darrier Barrier Penetration Bellows Test Barrier Description /

Humber Barrier Description /

Description Seal Tippet Valve Number

• Notes Valve Number Notes 33c

• RilR Pump A Discharge No Closed System '

PIG to Suppression Pool stic- E12F024A q ;lIG I L

E12F0llA -

it I 4 i E12F064A 1i I 34c RIIR Pump B Discharge No O l D to Suppression Pool 1 Closed System

' l6 E12F024B E12FOllB.

11 h

f' I 9'

• / E12F064B I" I W ta 35c j os RifR Pump C Discharge No A Closed System

• ? 16 E12F021 l ,

I g

to Suppression Pool Closed System 7IN u E12F064C ll nM i

m J .Ls la w 36c Demin Water to G33-2020 No M O, RWCU Sample Panel C/ P46FF182 (. I, P46FF055 6 3' HH 37c IUIR SRV FOSSA & F055C Yes M Closed System lh E12F055A 9 0 is to Suppression Pool (, Closed System ~

E12F055C C 1i ~

O 39c RIIR SRV FOSSB & F055D Yes A 4- Closed System b E12F055B ll to Suppression Pool Closed System 4 P E12F055D l l 40c RIIR SRV F036 No Closed System '

p 5(- E12F036 I to Suppression Pool 4

W 41c RilR SRV F005, 5'017A Yes O Jh h- Closed System 7 E12F005 11 and FF236 to E12F017A Il Suppression Pool '

E12F236 ,

il .

la 42c RilR SRV F101, F025C C4k l

to Suppression Pool Yes Closed System '

r,; E12F101 , I wU i

l E11F025C ha

• p I Jg $$o e

O

• Tyry A Nw Ai-So Nier od orat cn IMO OTussA3M ME.

• • 7-EMoir4 WAxEsty1:M.t.Ep Ourt.ity K pc Tg,gi--

\.) )

y.. - - - - - -

.y -

% -n' Tcblo 6.2-29 Mem b -

CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEA Inboard Isolation Outboard Isolation Barrier

• n na e , ,s e t on Barrier n=1 town Taat tsat e ime tement tel inse/

rin.nt.. i Dancelastson H<e s e l e s Ilmour i pt loss /

saaL Ty 4 Valve Nussber Notes Valve Number Notes 43c RIIR SRV FO)78, F030 Yes U uppression S Pool I 15 E12F0178 to Suppression Pool 9 E12F030 ,13 44c RHR Suction from Recirc Yes C 04' E12F009 %d 47c Steam to RCIC Turbine Yes E12F008 p

E51F063 5, E51F064 '

ESWO76 5 40c RCIC Pump Discharge 5 U Yes C Head Spray i Ci g]g,i 1 51F065 i /ll f$

P

+**

(01: .,, . i i1 7

M ,

49c RCIC Pump Suction C NO- 7 w From Suppression Pool No Closed Systesa 7lG E51F031 401 o  ; 11 l G Wl

(* )

• j Soc Turbine Discharge to Yes C Closed System '

E51F068

Suppression Pool j i;

$l 52c RCIC Puenp Nini m Flow Sypass No CQ Closed System 'J[., E51F019 I l j_ U 53c LPCS Pusep Discharge Yes C E12F006 Il E21F005 11 l

54c LPCS Pump Suction No C dJP Closed System ~

lG E21 FOOL f fg 55c LPCS Pump Test No C *We Closed System ~

E21F012

)

/ I Closed System

• E21F011 P1 56c LPCS SPV Discharge to No Suppeension Pool C MI S Closed System '

{ E21F006 11 Closed System /

E21F031 11

o A 57c et 'N Air Positive Seal to No C Closed System <l 4

,T .- P61FF046 PSIFF0lO & P51FF040 j [ Y -

oa C

g g Ti m A W y- A s.5. ,c.gto v ,,, c o M S-E *4 P4ct Ep o pEtt,Wgsg 4

A

• R.%W v4Aren.1:'ss.t.r.o 90 tan % Y t'E Y A Tussy

n y.. -

,, = '

n 4

Table 6.2-29 T{en 9a I CONTAINMI-:NT PF:NETRAfl0N AND CONTAIllMENT ISOLATION VALVE LEAKAGE RATE TMST (Continued)

Inboard Isolation outboard Isolation narrier narrier Penetration Bellows Test Barrier Description / Barrier Description /

Number Description Seal Typed Valve Number Notes Valve Number Notes 58c HPCS Pump Discharge Yes 6 E22F005 I11 E22F004 fp 59c HPCS Pump Section No C #@ Closed System 'N E22F015 &

l(c 60c HPCS SRV Discharge to No C $4 Closed System E22F023, E22F012 F

4 Suppression Pool No Closed System g l E22F014, E22F035 Y C g g. y ,

63c RWCU Pump Suction from Recirc Yes C G33F001 h G33F004 f  !

m e 64c RWCU Return to Yes C G33F040 5,10 Feedwater Line G33F039 5,10 l

[ j

! o .

• 65c RWCU Discharge to Yes C G33F028 6 G33F034 6 l Main Condenser  !

m  !

68c Containment Supply No O T(1FF015 5 '

T41FF014 5 Purge INAC & T4iFFo47 E C- [T41FFott l 69c Containment Exhaust No O T41FF005 5 T41FF006 5 IwAC T4l FFo44 & T;41 FFoi6 '

3

(

70c Containment Vacuum No ,C T41FF034A g T41FF033A f,42

) Relief Outlet i

l 72c Containment Vacuum No ,C T41FF034B T41FF033B Rv11cf outlet f* ((*

l 70c Skimmer Drain No C G41F044 6 G41F029 6 $:

to FPCC

• t- Brs A ver Ai.5e twaum<a uru.ess v r* w:e o vwen.wi+c . p

,y.+

FCum , wr m e- n u. o osei % T m .A ,,, , &

i

,% q I

/M p' '

Table f.2-29 CONTAINMI.NT PE ETRATIf.'N AND CONTAIFt4ENT ISOLITION VAI VE LEAKAGE RATE TEST .IST (Continued)

Inboard I olation Outboa':d Isolation Barr..er Harrier Penetration tiumber Bellows Test Barrier Description / Barrier Description /

Description Seal Type + Valve Number Notes Valve Number Notes 143c Chilled Water from No E4N"NON nrywell coI1nra -

744 T'St9tT_ cw .;ro 43 hw.

~

-- ~ ~ -

C gui i4 E osta e 4 (4.44.+) do C."

• E 81 #U73k _ '" R st FoM /6 A 145c ESW Supply to H2 y

Mixing. Blower (Div 1)

Yes' C P41FF169 g -

145 ESW Return from If 2 Yes 6J 2 C P41FF172 vg cs er Div O

P41FF171

' [ y d e A(tuR-4)

(L tda. i+ w C- K R E UM E' Ib 7 146c 24-in. Pipe Spare ' Ho Cappel -

_ k @g M

b 147c 12-in. Pipe Spare O[0 No Capped - - -

h 148c ADS Pneumatic Supply No P53FF017A Division 1 Q231 P53FF015A "fufL $b 156c Spare No - -

157c -

158c ESW Supply to Il No C P41FF115 Mixing Blower (biv 2)  % P41FFil4 gI 160c Air to RCIC Turbine No -

5 E51F078, E51F077 Exhauat Line 5 164c RWCU Pump to Filter Yes C G33'F053 Domineralizer 'j\

l,tf G33F054 #

gy w

165c ESW Return from No C <: 4 P41FFil7 l P41FFil6 Mixing ulower (Div 2) [ *$

o4

. rp;rt A. wr w,o gamute b, u.4 s ss.> H ort 9 O Y1NE W tSF g ( R E W A. m s V96WJL- P t L-1.E 9 004lt4C T3ee A. res.i .

a. =.-..--...-. ..

GISSAR II 22A7007

, . . 238 NUCLEAR ISLAND Rev. 0

~

M<n b - ,

Table 6.2-29 CONTAINMENT PENETRATION AND CONTAINMENT ISOLATION VALVE LEAKAGE  !

3 RATE TEST LIST (Continued)

V ht ET11 o l

% (11) h t t,. f 1, m_= ' -- --- ' rted i- -~rrre dir;;ti;;,. 4 to.q (12)

W A.nn-- k.Ch Seas., W 'PL'8-oC

- , , _T.;.

7TF. %m C , tv.st 54% 0E' Thi- -:1 : x _ m* fr 2_r :: _.._ ,__._m :n go,4 cf 6"- *N L; 1:5 sec.1?cy h

s-eme.9 vfm1 f/ATtut. Ar yih.g,wurty .1.lC d go et y (13) This valve is located on a non-Safety Grade piping system ]

which is not a CLOC.

( 14) The primary seal for each of the electrical penetrations con-(?( sist of two concentric O-rings with a test connection to permit leak testing the space between the 0-ring seals.

Test volume is pressurized with dry nitrogen.

(15) Influent lines terminating in the suppression pool are dis-cussed in Subsection 6.2.4.3.2.2.1.1. up (jQ -um w3.ms penetrate the containment (tit =mt.

• are designed to remain intact following a LOCA.'----

C

~ 4.. re 1 ;; _;3 ; _ -t  : =rgre not specifically , f vented to.the containment at=osphyre or to the outside c>c.

atmosphere and may remain water filled during Type A +

test.ing.

IN ' '

i W Ta b J e b ~r O r p J ' ._; CaN eve Ou a d s UJN h

• k A c l i Ncd C le c ct\ G.w cl C ow bt w mc.w b mIC)'"* b' d Ie k v-ch % 4.,$ ch~lI,oc wwa. c.Am.,,A be. '

cas%.\ w ~1 sain+ L p e r < o k s a u u s ll w e r L l

\

CSSi>4 e -f h e a r ud D i ly . ,

l (.< '

}

b

)l I

I 1

/

( ',

e l

l l

i 1

6.2-218 l

l

6 fM9c) It was agreed that we would add the following words to GESSAR Section 6.2.6.3.2, " Leakage testing of the closed ESF systemsoutside containment is performed in accordance with Section XI of the ASME BtPV code,but will comply with the testing frequencies and leakage reporting requirements of Appendix J. of [oCFR&c.

GESSAR II 22A7007 a 238 NUCLEAR ISLAND Rev. O Nom b s j 6.2.6.4 Scheduling and Reporting of Periodic Tests (Continued)

(I results shall be submitted to the NRC in a summary report approxi-mately three months after each test.

6.2.6.5 Sp.ecial Testing Requirements The maximum allowable leakage rate into the secondary containment and the means to verify that the inleakage r~ ate has not been exceeded, as well as the bypass leakage rate, are discussed in Subsections 6.2.3 and 6.5.1.3.

L a a n s a s n.s t U s o n (n. c.les. J tsF cycl,~, ov is ,6e e.w r+.taed r sa red.M us s c.e s e.os.n e.t w .% s Ea.J. .a D a*. ASmt 6* ht code. , ba l wil ca~ r 19 .ea h tkc 6.2.7 Suppression Pool Makeup System s. g ;o,, tre re p ,,,-..cbo,..c,

. 4 Au 44a n .. , s .r.f.4 % (c.

+ u ten The Suppression Pool Makeup System provides additional water from the upper containment pool to the suppression pool by gravity flow following a LOCA. Th'e quantity of water is sufficient to account for all conceivable post-accident entrapment volumes (i.e., places where water can be stored while maintaining long-term drywell vent water coverage).

6.2.7.1 Design Basis The following criteria were used in the design of the Suppression Pool Makeup System:

(

(1) The system is redundant with two 100% capacity lines.

The redundant lines are physically separated and electrical controls are separated into two divisions in accordance with IEEE-279.

(2) The system is Safety Class 2, Seismic Category I, and Quality Group B.

[

(3) Minimum long-term post-accident suppression pool water coverage over the top of the top drywell vent is 2 ft.

6.2-149 t

o ATTACHMENT NO. 3 DRAFT RESPONSES TO INSTRUMENTATION AND CONTROL SYSTEMS BRANCH QUESTIONS

OraFf Dple~ "T 3-I8-83 421.32 Based on our review, it appears tnat the proposed logic for manual Guc5rud initiation for several ESF systems is interlocked with permissive e logic from various sensors. In some cases, it appears that the permissive logic is dependent on the same sensors as those used for automatic initiation of the system. Our position on this matter is that the capability to manually initiate each safety system should be independent of the permissive logic, the sensors and the circuitry used for automatic initiation of that system. (Refer to Section 4.17 of IEEE Std. 279). Identify each safety system which is interlocked in a manner similar to that described above. Provide proposed modifications or justification for the present design.

In this regard, manual control of actuated devices at the motor control center (MCC) has been typically provided in previous designs. Our review of drawings I-960 A through M indicates that this feature has not been provided for your proposed design. Provide your rationale for not providing local control at the MCC'.s.

Response Supplemen I The response submitted previously on the docket remains unchanged.

However, the following information is provided to supplement the justification supporting the design of the HPCS and the Containment Spray System interlocks.

HPCS The HPCS injection valve can be closed without restriction using the component level manual switch. However, the same switch (spring return to AUTO from either OPEN or CLOSE) will only open the valve if a two-out-of-two level 8 signal is not present.

This interlock is important in that it prevents inadvertent overfill of the vessel. Such overfill can escalate a relatively minor occurance (loss of feedwater) to a major event (Water in the main steam lines).

This could potentially expose the safety relief valves and turbine controlled safety equipment to high pressure water slugs.

In an emergency situation requiring the Standby Liquid Control System, a vessel overfill could also dilute the boron concentration in the vessel.

For these reasons, it is GE's position that the L8 signals should not be bypassed to force manual opening control. It is possible to provide separate hardware for the manual L8 control interlock. However, without a bypass, such redundant hardware could not open the valve anyway, if the automatic interlocks were forcing it closed.

It is recognized that a single failure could cause inadvertent closure of the HPCS injection valve. This loss of HPCS is acceptable from a safety standpoint because of ADS and the low pressure system which maintains water inventory in the vessel . These systems are powered from divisions 1 and 2 as opposed to discussion 3 for HPCS. Thus, 3 their functions and interlocks are totally independent from those of HPCS.

Inadvertant failure of HPCS is a risk most certainly preferable to inadvertant flooding of the main steam lines.

e 431.32 Response -Con ti nued-Containment Spray Mode of RHR As explained above, separate automatic vs. manual interlocks for single valves cannot function independently in both opening and closing modes of the valve. With the manual spring-return-to "AUT0" switch, an opposing automatic closure signal would immediately reclose~ the valve as soon as its opening command seal-in relays dropped out. Valve cycling would thus result from opposing signals unless one permanently bypassed the other. Such a bypass would preclude the need for separate interlocks.

A manual bypass for the Containment Spray would increase the risk of inadvertent operation. Int $efiarkIIIdesign,thesprayislocated over the refueling area. Porfe$${1bE'RHR is functioning in another mode to keep the vessel cool. The bypass would pose a safety hazzard to personnel within the containment.

Operator initiated inadvertent containment spray has occured several times in both foreign and domestic plants where containment spray systems are employed. (Note recent occurrence at Oyster Creek: LER-50-219NA. )

The interlock is a deliberate preventive measure for the Mark III. A single failure of the interlock could cause loss of one train of containment spray. However, such a failure is of little consequence because of the redundant train. Thus, the risk of failure of one train is most certainly preferable to inadvertent containment spray.

o

. I ATTACHMENT NO. 4 DRAFT RESPONSES TO STRUCTURAL AND GE0 TECHNICAL ENGINEERING BRANCH QUESTIONS

e 220.11 At the time of this review, Appendix 3H which describes the effect of (3.7.2) the concrete between the containment and the shield building on the seismic analysis, is not available. Indicate when this appendix will be provided. This infonnation should be made available prior to the forthcoming structural audit in December 1982.

Response

The effects of the concrete between the Containn, erit and the Shield building on the seismic analysis is discussed in Appendix 3H which will be submitted in April 1983.

Shell forces, shell moments and element stresses have been obtained for various structures and for individual soil cases. These results were then enveloped to arrive at a set of final responses. These are included in the appendix and are generally within the envelope used in the design. At a few locations, the unit forces are slightly higher, for instance in the range of 3% of the allowable stress. This is certainly within the structural design margin.

Response spectra for various structures have also been generated for individual soil cases. Envelopes were obtained for them and are included in Appendix 3H as well. The majority of the curves fall within the original envelope and are still used for conservatism.

A few curves show some exceedance in accelerations and some frequency shift in the range of 5-10%. However, the increased accelerations are within engineering design margin for equipment and systems.

220.13 It is not clear in the discussion provided in Sections 3.7.2.3 and (3.7.2) 3.7.2.5 of your FSAR how you have accounted for the vertical flexibility of floors in the generation of the vertical response spectra. Accordingly, provide the procedures you have used to account for this phenomenon.

Response

A confinnatory analysis which accqunts for the vertical flexibility of floors is in progress and non6Pic"ombleted and will be submitted in April 1983. A brief description of this analysis follows:

Three typical floor panels were modeled by Spring-Dashpot Oscillators; they were then added to the mass point at the floor of of interest in the mathematical model of the building. A time-history analysis was performed for a soil case. expected to provide the maximum response.

Vertical Response Spectra for the selected floor panels and the main building mass point were generated and a comparison with the current corresponding response spectrum curves at the selected floor will be made.

By views of previous seismic analysis work the selection of the building and floor in this building will produce the maximum vertical amplification due to floor flexibility was made and resulted in the selection of the floor, at elevation 28'6" in the Auxiliary Building. Preliminary results indicate that the existing vertical response spectrum curve (Fig. 3.10-38) will envelope the response spectra from the confirmatory analysis.

.