ML17309A502

Proposed Tech Specs Reflecting Removal of Table of Containment Isolation Valves

ML17309A502
Person / Time
Site:	Ginna
Issue date:	11/30/1992
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17262B100	List: ML17262B099 ML17309A502 ML17311A016
References
NUDOCS 9212140159
Download: ML17309A502 (203)

Text

ATTACHMENT A Proposed Technical Specification Changes 9212140159 921130 PDR ADOCK 05000244.'

PDR

ATTACHMENT A Revise the Technical Specification pages as follows:

Remove Insert 3.6-1 3.6-1

-3.6-2 3.6-2 3.6-3 3.6-3 3.6-3a 3.6-4 3.6-5 3.6-6 3.6-7 3.6-7A 3.6-8 3.6-9 3.6-10 3.6-11 3.8-1 3.8-1 3.8-3 3.8-3 3.8-5 3.8-5 3.8-6 4 ' 4 4 '-4 4.4-6 4.4-6 4.4-7 4.4-7 4.4-8 4.4-8 4.4-11 4.4-11 4.4-13 4.4-13 4.4-14 4.4-14 4.4-17 4.4-17

Containment S stem A licabilit Applies to the integrity of reactor containment.

To define the operating status of the reactor containment for plant operation.

S ecification:

3.6.1 Containment Inte rit a~ Except as allowed by 3.6.3, containment integrity shall not be violated unless the reactor is in the cold shutdown condition. Closed valves may be opened on an intermittent basis under administrative control.

b. The containment integrity shall not be violated when the reactor vessel head is removed unless the boron concentration is greater than 2000 ppm.

c ~ Positive reactivity changes shall not be made by rod drive motion or boron dilution whenever the containment integrity is not intact unless the boron concentration is greater than 2000 ppm.

3.6.2 Internal Pressure If the internal pressure exceeds 1 psig or the internal vacuum exceeds 2.0 psig, the condition shall be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the reactor rendered subcritical.

Amendment No. AS 3.6-1 Proposed

3.6.3 Containment

~

Isolation Boundaries ~

3.6.3.1

~ With a containment isolation boundary inoperable for one

~

or more containment penetrations, either: ~

a. Restore each inoperable boundary to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, one closed manual valve, or a'blind flange, or c ~ Verify the operability of a closed system for the affected penetrations within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and either restore the inoperable boundary to OPERABLE status or isolate the penetration as provided in 3.6.3.1.b within 30 days, or

d. Be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.4 Combustible Gas Control 3.6.4.1 When the reactor is critical, at least two independent containment hydrogen monitors shall be operable. One of the monitors may be the Post Accident Sampling System.

3.6.4.2 With only one hydrogen monitor operable, restore a second monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.6.4.3 With no hydrogen monitors operable, restore at least one monitor to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.6.5 Containment Mini-Pur e Whenever the containment integrity is required, emphasis will be placed on limiting all purging and venting valves times to as low as achievable. The mini-purge isolation will remain closed to the maximum extent practicable but may be open for pressure control, for ALARA, for respirable air quality considerations for personnel entry, for surveillance tests that may require the valve to be open or other safety related reasons.

Amendment No. 9,18 3.6-2 Proposed

Basis:

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.

The shutdown margins are selected based on the type of activities that are being carried out. The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances. When the reactor head is not to be removed, a cold shutdown margin of 1%~k/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig.~'> The containment is designed to withstand an internal vacuum of 2.5 psig.<'> The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.

In order to minimize containment leakage during a design basis accident involving a significant, fission product release, penetrations not required for accident mitigation are provided with isolation boundaries. These isolation boundaries consist of either passive devices or active automatic valves and are listed in UFSAR Table 6.2-15. Closed manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges and closed systems are considered passive devices. Automatic isolation valves designed to close following an accident without operator action, are considered active devices. Two isolation devices are provided for each mechanical penetration, such that no single credible failure or malfunction of an active component can cause a loss of isolation, or result in a leakage rate that exceeds limits assumed in the safety analyses.

In the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure. Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange. A closed system also meets this criterion, however, a 30 day period to either fix the inoperable boundary or provide additional isolation is conservatively applied. Verification of the operability of the closed system can be accomplished through normal system operation, containment leakage detection systems, surveillance testing, or normal operator walkdowns.

The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the containment.

Amendment No. 45 3. 6-3 Proposed

References:

(1), Westinghouse Analysis, "Report for the BAST Concentration Reduction for R. E. Ginna", August 1985, submitted via Application for Amendment to the Operating License in a letter from R.W. Kober, RGGE to H.A. Denton, NRC, dated October 16, 1985 (2) UFSAR Section 3.8.1.2.2 (3) UFSAR Table 6.2-15 3.6-3a Proposed

3.8 REFUELING A licabilit Applies to operating limitations during refueling operations.

To ensure that no incident could occur during refueling operations that would affect public health and safety S ecification 3.8.1 During refueling operations the following conditions shall be satisfied.

a ~ Containment penetrations shall be in the following status:

i. The equipment hatch shall be in place with at least one access door closed, or the closure plate that restricts air flow from containment shall be in place, ii. At least one access door in the personnel air lock shall be closed, and iii. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. Closed by an isolation valve, blind flange, or manual valve, or

2. Be capable of being closed by an OPERABLE automatic Shutdown Purge or Mini Purge valve.

b. Radiation levels in the containment shall be monitored continuously.

C ~ Core subcritical neutron flux shall be continuously monitored by at least two source range neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment and control room available whenever core geometry is being changed. When core geometry is not being changed at Amendment No. g,gg 3.8-1 Proposed

flange. If this condition is not met, all operations involving movement of fuel or control rods in the reactor vessel shall be suspended.

3.8.2 If any of the specified limiting conditions for refueling is not met, refueling of the reactor shall cease; work shall be initiated to correct the violated conditions so that the specified limits are met; no operations which may increase the reactivity of the core shall be made.

3.8.3 If the conditions of 3.8.1.d are not met, then in addition to the requirements of 3.8.2, isolate the Shutdown Purge and Mini Purge penetrations within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Basis:

The equipment and general procedures to be utilized during refueling are discussed in the UFSAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard 3.8-3 Proposed

provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at, a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over stored racks containing spent fuel.

The operability requirements for residual heat removal loops will ensure adequate heat removal while in the refueling mode. The requirement for 23 feet of water above the reactor vessel flange while handling fuel and fuel components in containment, is consistent with the assumptions of the fuel handling accident analysis.

The analysis<'> for a fuel handling accident inside containment establishes acceptable offsite limiting doses following rupture of all rods of an assembly operated at peak power. No credit is taken for containment isolation or effluent filtration prior to release.

Requiring closure of penetrations which provide direct access from containment atmosphere to the outside atmosphere establishes additional margin for the fuel handling accident and establishes a seismic envelope to protect against seismic events during refueling. Isolation of these'penetrations may be achieved by an OPERABLE shutdown purge or mini-purge valve, blind flange, or isolation valve. An OPERABLE shutdown purge or mini-purge valve is capable of being automatically isolated by Rll or R12.

Penetrations which do not provide direct access from containment atmosphere to the outside atmosphere support containment integrity by either a closed system within containment, necessary isolation valves, or a material which can provide a temporary ventilation barrier, at atmospheric pressure, for the containment penetrations during fuel movement.

Amendment No. 3.8-5 Proposed

References (1) UFSAR Sections 9.1.4.4 and 9.1.4.5 (2) Reload Transient Safety Report, Cycle 14 (3) UFSAR Section 15.7.3.3 3.8-6 Proposed

4.4.1.4

~ ~ ~ Acce tance Criteria

a. ~ The leaka g e rate Ltm shall be <0.75 Lt at Pt. Pt is defined as the containment vessel reduced test pressure which is greater than or equal to 35 psig.

Ltm is defined as the total measured containment leakage rate at pressure Pt. Lt is defined as the maximum allowable leakage rate at pressure Pt.

I Pt 1~I~

b. Lt shall be determined as Lt = LalsaJ which equals

.1528 percent weight per day at 35 psig. Pa is defined as the calculated peak containment internal pressure related to design basis accidents which is greater than or equal to 60 psig. La is defined as the maximum allowable leakage rate at Pa which equals .2 percent weight per day.

c. The leakage rate at Pa (Lam) shall be <0.75 La.

Lam is defined as the total measured containment leakage rate at pressure Pa.

4.4.1.5 Test Fre uenc

'a ~ A set of three integrated leak rate tests shall be performed at approximately equal intervals during each'0-year service period. The third test of each set shall be conducted in the final year of the 10-year service period or one year before or after the final year of the 10-year service period provided:

the interval between any two Type A tests does not exceed four years, following each in-service inspection, the containment airlocks, the steam generator inspection/maintenance penetration, and the equipment hatch are leak tested prior to returning the plant to operation, and 3.3. 1 ~ any repair, replacement,, or modification of a containment barrier resulting from the inservice inspections shall be followed by the appropriate leakage test.

4 4 4 Proposed

b. The local leakage rate shall be measured for each of the following components:

Containment penetrations that employ resilient seals, gaskets, or sealant compounds, piping penetrations with expansion bellows and electrical penetrations with flexible metal seal assemblies.

ii. Air lock and equipment door seals.

iiio Fuel transfer tube.

lvo Isolation valves on the testable fluid systems lines penetrating the containment.

vo Other containment components, which require leak repair in order to meet the acceptance criterion for any integrated leakage rate test.

4.4.2.2 Acce tance Criterion Containment isolation boundaries are inoperable from a leakage standpoint when the demonstrated leakage of a single boundary or cumulative total leakage of all boundaries is greater than 0.60 La.

4.4.2.3 Corrective Action a 0 If at any time it is determined that the total leakage from all penetrations and isolation boundaries exceeds 0.60 La, repairs shall be initiated immediately.

4.4-6 Proposed

~ . >>K, ~

b. If repairs are not completed and conformance to the acceptance criterion of 4.4.2.2 is not demonstrated within 48 hours, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion.

c~ If it is determined that the leakage through a II mini-purge supply and exhaust line is greater than 0.05 La an engineering evaluation shall be performed and plans for corrective action developed.

4.4.2.4 Test Fre uenc a~ Except as specified in b. and c. below, individual penetrations and containment isolation valves shall be tested during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years.

b. The containment equipment hatch, fuel transfer tube, steam generator inspection/maintenance penetration, and shutdown purge system flanges shall be tested at each refueling shutdown or after each use, if that be sooner.

Amendment No. i8 4.4-7 Proposed

c ~ The containment air locks shall be tested at intervals of no more than six months by pressurizing the space between the air lock doors. In addition, following opening of the air lock door during the interval, a test shall be performed by pressurizing between the dual seals of each door ,opened, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the opening, unless the reactor was in the cold shutdown condition at the time of the opening or has been subsequently brought to the cold shutdown condition. A test shall also be performed by pressurizing between the dual seals of each door within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of leaving the cold shutdown condition, unless the doors have not been open since the last test performed either by pressurizing the space between the air lock doors or by pressurizing between the dual door seals.

Amendment No. gg Proposed

f ' N

• W I

the tendon containing 6 broken wires) shall be inspected.

The accepted criterion then shall be no more than 4 broken wires in any of the additional 4 tendons. If this criterion is not satisfied, all of the tendons shall be inspected and if more than 5% of the total wires are broken, the reactor shall be shut down and depressurized.

4.4.4.2 Pre-Stress Confirmation Test a ~ Lift-offtests shall be performed on the 14 tendons identified in 4.4.4.1a above, at the intervals specified in 4.4.4.1b. If the average stress in the 14 tendons checked is less than 144,000 psi (60% of ultimate stress), all tendons shall be checked for stress and retensioned, of 144,000 psi.

if necessary, to a stress

b. Before reseating a tendon, additional stress (6%)

shall be imposed to verify the ability of the tendon to sustain the added stress applied during accident conditions.

4.4.5 Containment Isolation Valves 4.4.5.1 Each containment isolation valve shall be demonstrated to be OPERABLE in accordance with the Ginna Station Pump and Valve Test program submitted in accordance with 10 CFR 50.55a.

4.4.6 Containment Isolation Res onse 4.4.6.1 Each containment isolation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.1-1.

4.4.6.2 The response time of each containment isolation valve shall be demonstrated to be within its limit at, least once per 18 months. The response time includes only the valve travel time for those valves which the safety analysis assumptions take credit. for a change in valve position in response to a containment isolation signal.

Amendment No. 9,1Z 4.4-11 Proposed

The Specification also allows for possible deterioration of the leakage rate between tests, by requiring that the total measured leakage rate be only 75% of the maximum allowable leakage rate.

The duration and methods for the integrated leakage rate test established by ANSI N45.4-1972 provide a minimum level of accuracy and allow for daily cyclic variation in temperature and thermal radiation. The frequency of the integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns. Refueling shutdowns are scheduled at approximately one year intervals.

The specified frequency of integrated leakage rate tests is based on three major considerations. First is the low probability of leaks in the liner, because of (a) the use of weld channels to test the leaktightness of the welds during erection, (b) conformance of the complete containment to a 0.1% per day leak rate at 60 psig during preoperational testing, and (c) absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60 La) of the total leakage that is specified as acceptable. Third is the tendon stress surveillance program, which provides assurance than an important part of the structural integrity of the containment is maintained.

4.4-13 Proposed

. The basis for specification of a total leakage of 0.60 La from penetrations and isolation boundaries is that only a portion of the allowable integrated leakage rate should be from those sources in order to provide assurance that the integrated leakage rate would remain within the specified limits during the intervals between integrated leakage rate tests. Because most leakage during an integrated leak rate test occurs though penetrations and isolation valves, and because for most penetrations and isolation valves a smaller leakage rate would result from an integrated leak test than from a local test, adequate assurance of maintaining the integrated leakage rate within the specified limits is provided. The limiting leakage rates from the Recirculation Heat Removal Systems are judgement values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test 4.4-14 Proposed

d 4 The pre-stress confirmation test provides a direct measure of the load-carrying capability of the tendon.

If the surveillance program indicates by extensive wire breakage or tendon stress relation that the pre-stressing tendons are not behaving as expected, the situation will be evaluated immediately.

The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible. Thus the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down the reactor. The containment, is provided with two readily removable tendons that might be useful to such a study. In addition, there are 40 tendons, each containing a removable wire which will be used to monitor for possible corrosion effects.

Operability of the containment isolation boundaries ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

Performance of cycling tests and verification of isolation times associated with automatic containment isolation valves is covered by the Pump and Valve Test Program. Compliance with Appendix J to 10 CFR 50 is addressed under local leak testing requirements.

References:

(1) UFSAR Section 3.1.2.2.7 (2) UFSAR Section 6.2.6.1 (3) UFSAR Section 15.6.4.3 (4) UFSAR Section 6.3.3.8 (5) UFSAR Table 15.6-9 (6) FSAR Page 5.1.2-28 (7) North-American-Rockwell Report 550-x-32, Autonetics Reliability Handbook, February 1963.

(8) FSAR Page 5.1-28 4.4-17 Proposed

ATTACHMENT B Safety Evaluation

Attachment B Pago 1 of 4 The primary purpose of this amendment is to remove Table 3.6-1, "Containment Isolation Valves", from the R.E. Ginna Technical Specifications. The reference to Table 3.6-1 in Technical Specification sections 3.6.3.1, 4.4.5.1, and 4.4.6.2 will be deleted with a reference to UFSAR Table 6.2-13 being added to the bases for Technical Specification 3.6. In addition, the inoperability definition and action required statement for Technical Specifications 3.6.1 and 3.6.3.1 will be clarified. The Specifications and Bases for containment integrity during refueling operations (3.8.1 section a and 3.8.3) will be revised to make them more consistent with industry standards. Technical Specifications 4.4.1.5, section a (ii) and 4.4.2.4, section b, will be revised to include the modified steam generator inspection/ maintenance penetration. Technical Specification 4.4.1.5, section a (ii) and the Bases for section 4.4 will also be clarified. The temporary notes associated with the purge system and mini-purge valves (Technical Specifications 3.6.5 and 4.4.2.4 section a and d) will be removed since the valves have been installed. Also, the acceptance criteria for containment leakage criteria as listed in Technical Specification 4.4.1.4 and 4.4.2.2 will be clarified.

The 1988 Inservice Test (IST) Program provided a complete review of the containment isolation valves for Ginna and their testing requirements. The information obtained during this review was submitted to the NRC to define the IST requirements for the third ten-year interval at Ginna. This submittal was subsequently approved by the NRC. As a result of this submittal and approval, numerous clarifications were required of Technical Specification Table 3.6-1 and UFSAR Table 6.2-15 (formerly 6.2-13). However, this amendment will remove Technical Specification Table 3.6-1. The necessary changes to UFSAR Table 6.2-15 have been completed.

Attachment E contains the revised UFSAR table and associated figures for your information.

Generic Letter 91-08 provides guidance on removing component lists from technical specifications, including the table of containment isolation valves, since their removal would not alter the requirements that are applied to these components. Removing Table 3.6-1 from the Technical Specifications and incorporating the required information into Ginna UFSAR Table 6.2-15 will maintain the listing of the containment isolation boundaries within a licensee controlled document. Changes to this document can only be performed under the criteria of 10 CFR 50.59 to ensure that no unreviewed safety questions are related to the change. Any future changes to UFSAR Table 6.2-15 will be submitted as part of the required UFSAR update. In addition, a report summary of the changes to the Ginna UFSAR are furnished to the NRC on a required basis. A reference to UFSAR Table 6.2-15 has also been provided in the bases for Technical Specification 3.6 consistent, with Generic Letter 91-08.

Generic Letter 91-08 also provided instructions to add a note to the containment isolation valve LCO with respect to opening locked or sealed closed containment isolation valves under administrative control. This note was added to Technical Specification 3.6.1 and a discussion of the necessary administrative controls required for performing this action was added to the bases.

Attachment B Page 2 of 4 Technical Specification 3.6.3.1 is revised to include the use of a closed system as an allowable means to isolate a containment penetration that has a inoperable containment isolation boundary.

A closed system can be considered equal, or in many cases preferable, to the remaining alternatives (e.g., a closed manual valve), since the closed system by definition must be missile protected, seismically designed and leak tested. The use of a closed system is also consistent with the intent of the bases for containment isolation in NUREG-1430 which states:

In the event one containment isolation valve in one or more penetration flow paths is inoperable [except for purge valve leakage not within limits], the affected penetration must, be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

Since a closed system is not affected by any single active failure, it provides an equivalent barrier to a blind flange, a closed manual valve, or a deactivated containment isolation valve. However, a 30 day limit was conservatively assigned to the use of only the closed system before additional isolation must be provided or the inoperable boundary repaired. The 30 day limit is also consistent with Standard Technical Specifications which require that the flow path for penetrations with inoperable containment isolation valves be verified isolated once every 31 days. The Bases for Technical Specification 3.6 were also updated to provide necessary supporting information with respect to using a closed system to isolate an inoperable isolation boundary.

The remaining changes with respect to the required actions of Technical Specification 3.6.3.1 allow consistency with Standard Technical Specifications. However, "isolation boundary" was used in place of "isolation valve" since not all penetrations have two containment isolation valves. For example, penetrations under the specifications for General Design Criteria 57 only require a single isolation valve; the piping provides an additional boundary. The use of "isolation boundary" is also consistent with the column headings of the current Containment Isolation Valve Table 3.6-1.

Information on what qualifies as an "isolation boundary" is provided in the bases for Technical Specification 3.6. These criteria are consistent with the necessary General Design Criteria, or exemption, as appropriate. "Isolation boundary" was also used in place of "isolation valve" in Technical Specifications 4.4.2.2, 4.4.2.3, and the Bases for section 4.4.

The inoperability definition based on leakage for containment isolation boundaries was also removed from Technical Specification 3.6.3.1. This definition is found in Technical Specification 4.4.2.3 which was subsequently updated to make it more consistent with 10 CFR 50 Appendix J. This change eliminates duplication within the Technical Specifications and is consistent with Standard Technical Specifications.

Attachment B Page 3 of 4 The action statement associated with Technical Specification 3.8.1 section a was modified to make it more nearly consistent with Standard Technical Specifications. The most significant change was with respect to removing the requirement of having all automatic containment isolation valves operable during refueling operations.

The proposed specification now only requires that all penetrations providing direct access from the containment atmosphere to the outside atmosphere be either isolated or capable of being isolated by an automatic purge valve. This change is considered acceptable since a fuel handling accident will not significantly pressurize the containment. In addition, the fuel handling accident analyzed for Ginna do'es not take credit for isolation of containment while remaining well within 10 CFR 100 guidelines (UFSAR Section 15.7.3.3.1.1). Therefore, the removal of this requirement does not affect the consequences of a fuel handling accident.

The changes to Technical Specification 3.8.3 now specifically identify which penetrations must be closed if there is no residual heat removal loop in service (i.e., Shutdown Purge and Mini-Purge).

The remaining penetrations that provide direct access from the containment atmosphere to the outside atmosphere are already required to be isolated during refueling operations per new Technical Specification 3.8.1 section a (iii). The changes to the bases are consistent with Standard Technical Specifications.

Consequently, these are not technical changes.

The changes with respect to containment leakage criteria in Technical Specification 4.4.1.4 are clarifications only. All terms contained in the definition for Lt is specified in the Technical Specifications consistent with 10 CFR 50 Appendix J.

The addition of the steam generator inspection/maintenance penetration to both the UFSAR Table and the necessary Technical Specification surveillance requirements is the result of a modification to enhance containment closure during mid-loop operation (Generic Letter 88-17). No new containment isolation valves were added as a result of this modification. The addition of this penetration to the UFSAR Table and Technical Specifications 4.4.1.5, section a (ii) and 4.4.2.4, section b, results in the new penetration to be treated consistent with respect to the Personnel and Equipment Hatches, and the fuel transfer tube (see letter from R.C. Mecredy, RGEE, to A.R. Johnson, NRC, dated March 13, 1990).

The first line of Technical Specification 4.4.1.5, section a (ii) is also modified to state "following each in-service inspection..."

The hyphenation of "in-service" is to correct a typographical error only. The replacement of "one" with "each" provides greater understanding of the test frequency requirements. These changes are a minor clarification only and do not involve a technical change.

The temporary notes associated with the purge and mini-purge valves in Technical Specifications 3.6.5, 4.4.2.4 section a and d are removed since these valves have been installed. This is not a technical change since the notes were only intended to be applicable until the completion of the necessary modifications.

Attachment B Page 4 of 4 Technical Specifications 4.4.5.1 and 4.4.6.2 were revised to remove the reference to Table 3.6-1 since this is being deleted. These specifications were also changed to make them consistent with Standard Technical Specifications.

In accordance with 10 CFR 50.91, these changes to the Technical Specifications have been evaluated to determine if the operation of the facility in accordance with the proposed amendment would:

involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. create the possibility of a new or different kind of accident previously evaluated; or

3. involve a significant reduction in a margin of safety.

These proposed changes do not increase the probability or consequences of a previously evaluated accident or create a new or different type of accident. Furthermore, there is no reduction in the margin of safety for any particular Technical Specification. The detailed changes are described in Attachment F.

Therefore, Rochester Gas and Electric submits that the issues associated with this Amendment request are outside the criteria of 10 CFR 50.91; and a no significant hazards finding is warranted.

ATTACHMENT C 10 CFR 50 Appendix J Relief Requests

Attachment C Page 1 of 4 In support of preparing this amendment request, RGGE has performed an extensive review of the containment isolation valves (CIVs) and boundaries (CIBs) for Ginna Station. Included with this review was an assessment of the test procedures that are used for 10 CFR 50, Appendix J testing. These procedures were replaced in their entirety with the new procedures being used for necessary Appendix J testing during the recent 1992 refueling outage. However, as a result of preparing and using these new test procedures, RGEE determined that relief is necessary from certain provisions of Appendix J for several containment isolation valves and boundaries.

These relief requests are directly related to this application for amendment since relief is necessary in'rder .to eliminate the need for potential station modifications and revision of the isolation valves and boundaries currently identified on UFSAR Table 6.2-15.

The relief requests, and their basis, are provided below. If granted, these requests will be added to the 1990-1999 Inservice Pump and Valve Test Program for Ginna Station as necessary.

(1) Penetrations 105 and 109 contain the Containment Spray injection lines to the ring headers. Both penetrations have test and drain lines located outside containment that are not used for 10 CFR 50 Appendix J testing. These 3/4 inch lines have the necessary containment isolation valves and boundaries; however, these components cannot be leak tested since there are no available test connections. The Containment Spray lines are normally filled with water to a level at least 45 feet above the test and drain lines in order to facilitate faster response of the system during an accident. RGGE has performed an analysis of this line and concluded that the water would not boil off during a LOCA.

Since the test and drain lines are constantly exposed to this head of water during power operations, any leakage would be noticed either by normal operator walkdowns (i.e., indication of water on valve or floor), or during monthly tests of the containment spray pumps which require confirmation of the head of water. Consequently, a verifiable water barrier between the containment atmosphere and the valves will always be in place such that leak testing with air should not be required.

RGGE estimates that it would cost approximately $ 40,000 to install the necessary test connections for these lines. As such, RGGE proposes to fill the Containment Spray injection lines using the RWST each refueling outage to a minimum level of 66.9 feet (or 29 psig). This is the maximum height of water that can be used without creating the potential for flooding the containment charcoal filter units. Each test and drain line containment isolation valve or boundary would then be evaluated for any, observed leakage either through visual inspection or the use of local pressure indication. RG6E believes that this test meets the underlying purpose of Appendix J without creating undue hardships'n the licensee.

tea Attachment C Pago 2 of 4 for penetration 111 provides a containment isolation II (2) AOV 959 boundary by isolating the non-closed portion of the Residual Heat Removal (RHR) system. Based on 10 CFR 50, Appendix J, AOV 959 would be required to be leak tested once every refueling outage since it, is an automatic containment isolation valve. However, leak testing of this valve cannot, be accomplished since there are no available test connections.

AOV 959 is normally closed 'at power with its fuses removed and a boundary control tag in place. Following an accident,, the valve is continuously pressurized above the peak containment accident pressure by the head of the RHR pumps acting in the safety injection mode. This pressure head is available throughout the post accident period regardless of any single active failure. Consequently, AOV 959 should not require testing since it does not perform a containment isolation function as defined by 10 CFR 50, Appendix J, Section II.B.

The manual valve downstream of 959 (957) is also maintained closed at power in order to provide additional redundancy. It should be noted that this position was accepted by the NRC for MOVs 720 and 721 which are also CIVs for penetration 111 (see letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E, subject: Completion of Appendix J Review, dated May 6, 1981).

(3) Penetrations 130 and 131 contain the Component Cooling Water (CCW) return and supply lines respectively, for the Reactor Support Coolers. These penetrations take credit for a closed system inside containment (CLIC) and two MOVs (813 and 814) as the containment isolation boundaries. The two MOVs are currently tested with air in accordance with 10 CFR 50, Appendix J; however, RG&E proposes to test these valves with water. The CCW system provides a 30 day water seal for the two MOVS since the system is required to support the Residual Heat Removal Coolers post-LOCA. The only time that CCW would not be operating during this 30 day period is for the injection phase of the accident. However, a failure of the CLIC does not need to be assumed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident since it is a passive component. At this time, the recirculation phase would be initiated and the CCW system operating.

(4) Penetration 140 contains the Residual Heat Removal (RHR) suction line from Hot Leg A. The two main containment isolation barriers for this penetration are MOV 701 and a closed system outside containment (CLOC). MOV 701 does not require 10 CFR 50, Appendix J testing for the same reason as MOVs 720 and 721 (see t2 above). Instead, MOV 701 is hydrostatically tested every refueling outage. The drain and vent lines used in support of this test are located between MOV 701 and containment; consequently, they are required to have containment isolation valves and be tested in accordance with Appendix J. RG&E estimates that it, would cost approximately $ 100,000 to add the necessary test connections for these lines. In addition, there are significant ALARA concerns with respect to modifying this piping. Therefore, RG&E requests that relief from Appendix J be granted for the isolation valves on these lines consistent with MOV 701.

Attachment C Page 3 of 4

(~) Penetration 143 contains the Reactor Coolant Drain Tank Discharge Line. The isolation boundaries for this penetration consist, of three automatic air-operated valves and two manual valves. Manual isolation valve 1722 cannot currently be directly tested since there is no downstream vent; however, the valve is "inferred" tested (i.e., exposed to air test pressure through the testing of another isolation boundary whereby the leakage through 1722 could be inferred). In addition, it cannot be assured that all water has been drained from the valve seat prior to Appendix J testing since this is the drain line from the Fuel Transfer Canal. RG6E estimates that it would cost approximately $ 50,000 to install a vent line and isolation valve for 1722. There are also ALARA concerns since the piping normally contains radioactive fluid.

Consequently, RG&E proposes to continue to "infer" test 1722 after draining the line as much as possible. It should also be noted that 1722 will normally have a water seal against when containment integrity is required.

it (6) Penetrations 20la, 20lb, 209a, and 209b contain the Service Water (SW) supply and return lines for the Reactor Compartment Cooling Units. In addition, penetrations 312, 316, 319, and 320 contain the SW supply lines to the Containment Fan Coolers while penetrations 308, 311, 315, and 323 contain the return lines. These twelve penetrations all take credit for a closed system inside containment (CLIC) and a normally open manual containment isolation valve outside containment. These manual valves are only hydrostatically tested (i.e., not tested to 10 CFR 50, Appendix J criteria) as a result of a cost/benefit study performed during the Systematic Evaluation Program for Ginna. This study determined that the manual isolation valves would only be required if there was a significant breach of the CLIC following a design basis LOCA whereas installing new automatic valves and test connections would cost several million dollars. The NRC accepted the proposed hydrostatic testing approach since the CLIC is seismically designed and missile protected (NUREG-0821, Section 4.22.3). A further review of these lines has found various pressure indicators, flow and temperature transmitters, and drain valves located between the manual isolation valves and containment. However, these components cannot be leak tested since they do not have the necessary test connections. RGGE estimates that it would cost approximately $ 120,000 to install new test connections.

Consequently, RGRE proposes to continue to hydrostatically test these components, similar to that performed for the manual valves, in place of the required Appendix J testing.

Attachment C Page 4 of 4 (7) Penetrations 206B and 207B are the Steam Generator Sample lines while 321 and 322 are the Steam Generator Blowdown lines. Each of these four penetrations contain two containment isolation valves consisting of a normally open manual valve and an automatic air-operated valve. All eight valves are currently tested to 10 CFR 50, Appendix J.

However, these four lines originate from the steam generator secondary side; consequently, the steam generator tubes form one containment barrier as a closed system inside containment (CLIC). Other similar penetrations for Ginna (e.g., Main Steam, Main Feedwater) only have a single isolation valve outside containment that does not need to be tested to Appendix J (See Attachment D, Question 414). Consequently, RG&E proposes to only identify the automatic air-operated valves as CIVs and remove all Appendix J testing requirements.

Fire Service Water penetration 307 contains check valve 9229 which is located inside containment. However, it cannot be assured that all water has been drained from the valve seat prior to Appendix J testing due to its location with respect to available drain lines. RG&E estimates that it would cost approximately $ 20,000 to install the necessary drain line.

Since valve 9229 is outside the missile shield, it is highly unlikely that the Fire Service Water pipe would break in a location such that all water would be completely drained from the valve seat. Therefore, testing 9229 in its current configuration is representative of the conditions that the valve would most likely see during an accident. RG&E will continue to try and remove as much water as possible before the test, but does not believe that the addition of a drain valve is necessary.

(9) The containment isolation boundaries for Hydrogen Monitor Instrumentation penetration 332a are SOVs 922 and 924 and the nuclear sampling system (i.e., closed system outside containment). The two SOVs are required to be tested in accordance with 10 CFR 50, Appendix J since they are automatic CIVs; however, there is no available downstream vent. RG&E estimates that it the necessary vents.

would cost approximately $ 15,000 to install The second containment isolation boundary for this penetration is the Hydrogen Monitor Sampling System which is a closed system outside containment (CLOC).

This closed system is tested by pressurizing the Hydrogen Monitor piping up to the two SOVs. Consequently, SOVs 922 and 924 are "inferred" tested though in the opposite direction.

Therefore, RG&E proposes to continue to infer test the valves based on the cost related to adding a vent.

ATTACHMENT D Response To NRC Request For Additional Information Letter From A.R. Johnson, NRC, to R.C. Mecredy, RGREg dated September 26, 1991

'r Attachment D Page 1 of 12 As a result of reviewing RG&E's Application for Amendment to Operating License DPR-18 with respect to removing the list of containment isolation valves from Technical Specifications, the NRC responded with a Reque'st for Additional Information (see letter from A. R. Johnson, NRC, to R. C. Mecredy, RG&E,, dated September 26, 1991). The issues discussed in this RAI have already been addressed within the Amendment Request and associated UFSAR table and figures; however, a specific response to each of the twenty-nine comments and questions is provided below.

Table 6.2-13 identifies many valves, but does not distinguish between which valves are containment isolation valves and those that are not, other than by use of notes. For example some notes indicate that some valves are not considered containment i sol ati on valves. The use of the term "considered" does not clarify what the classification of the i

valve s and should (not J be used to describe valves.

there is any clarification to be noted because valves are If val ves, it i listed which are not cl assi fied as containment sol ation should be provided for accuracy. Additional comments on specific notes are provided in paragraphs below.

With respect to boundaries, the table, in general, is not clear on what constitutes a boundary particularly in cases where only one valve is classified as a containment isolation valve for a given penetration. A containment isolation boundary may be a blind flange or a closed system. However, the table does not make clear the boundaries of the second i

contai nment sol ation barrier. The figures note that some instruments constitute a containment isolation boundary.

Therefore, where a system or component is considered a second barrier, in addition to a single containment isolation valve, it should be so identified. Also, the location of that component would be identified under Tabl e 6. 2-13 column heading "Position Relative to Containment." Footnote 4 to the table would be appli cabl e where this boundary is a closed system outside containment, however, this note presently does not identify that closed system. Footnote 4 is poorly worded since it is appended to the line entry that identifies the containment isolation valve. This information is important since the TS requires an operabl e boundary, or second isolation valve in the case that one containment isolation valve is inoperable, and the TS Bases references this table for such information.

RESPONSE

RG&E has performed an extensive review of the containment isolation 'valves (CIVs) and boundaries (CIBs) for Ginna Station. The results of this review have b'een incorporated into the CIV/CIB testing program, UFSAR Table 6.2-15 (formerly 6.2-13), and,the associated UFSAR figures (see.,Attachment E).

Details concerning the specific changes which were made are provided in the answers to the questions which follow; however, a summary of the significant changes that were made is presented below.

Attachment D Page 2 of 12 a.) All components which provide a containment isolation boundary are identified on both UFSAR Table 6.2-15 and the associated figures. Closed systems that are used as an isolation boundary have been specifically identified on UFSAR Table 6.2-15 with either "CLOG Closed Loop Outside Containment" or "CLIC Closed Loop Inside Containment". Blind flanges, instruments, or other components which provide a passive containment isolation boundary have been identified with "CIB" on the figures.

b.) UFSAR Table notes have been clarified to provide the explicit basis for Appendix J relief where necessary.

c.) All CIV/CIB test procedures were reviewed, upgraded, and subsequently used for necessary 10 CFR 50, Appendix J testing during the recent 1992 refueling outage. UFSAR Table 6.2-15 was then revised to ensure that consistent with the Appendix J testing program. Most it was UFSAR figures are now taken directly from the CIV/CIB test procedures to ensure that they remain accurate in the future.

In a number of cases, more than one penetration is listed under a single penetration number in Table 6.2-13. This is contrary to the general practice of identifying each separate entry.

i penetration with ts associated val ves or boundary as a Each penetration should be listed and identified individually. This includes the following:

124a (Separate penetrations for supply and return.)

124b (Separate penetrati ons for ai r sample to "C" fan and common return.)

201 Top and 201 Bottom 202 (Separate penetrati ons for H2 "main " and "pilot "

burners. )

203b (Separate penetrati ons for air sample to "B" fan and common return.)

209 Top and 209 Bottom 305c (Appears to be three penetrations, but containment boundary is not shown on Figure 6.2-61) 332c (Three penetrati ons shown for the same penetration number. )

RESPONSE

A containment penetration at Ginna may contain several process lines. As such, both the "supply" and "return" lines for a given system, or multiple lines performing the same function may go through a single penetration. However, to prevent. any misinterpretations, UFSAR Table 6.2-15 has been revised to show a separate entry for each process line. The penetration names have also been revised as necessary for consistency (i.e., eliminated use of "top" and "bottom"). See Attachment E for further details.

rta

~ 4A

'I 'I

Attachment D Page 3 of 12 Where footnote 9 is used in Table 6.2-13, the purpose for the automatic closure of the associated valve should be clarified.

For valve 427 on the letdown line, i ts closure on a containment isolation signal (CIS) 'is important if one of the orifice valves fails to close since it precludes the loss of reactor coolant to the pressurizer relief tank when reactor pressure is greater than the relief setting of relief valve 203 that is isolated by the closure of valve 371 on a CIS.

For penetrations 123, 205, 206a, 207a, and 210, the closure of a second valve on a CIS provides a degree of redundancy for containment isolation.

RESPONSE

Note 9 to UFSAR Table 6.2-15 is used to identify those valves which receive a containment isolation signal (CIS), but are not containment isolation valves based on missile barrier or class break criteria. These valves are only shown on the table to prevent any future questions relating to components which receive a CIS, but are not included on the table. With respect to AOV 427, this valve fails open on loss of instrument air which will occur shortly after receipt of a CIS since instrument air to containment is also isolated.

Consequently, AOV 427 will always fail open until instrument air is restored to containment. The importance of AOV 427 with respect to the failure of an orifice valve is an operational concern. It is not the intent of the UFSAR table to identify the potential significance of every valve, or to distinguish every scenario where the valve may be used (e.g.,

recovery from an accident). Instead, these issues are addressed by procedures and training. Consequently, Note 9 has not been revised.

For penetration 143 in the Table 6.2-13, valve 1722 should be added since it has been marked as a "CIV" (contai nment isolation valve) on Figure 6.2-43.

RESPONSE

UFSAR Table 6.2-15 has been updated to include valve 1722 as a CIV.

Note 17 in the table should be updated to reference correspondence whi ch granted reli ef from Appendix J leak testing, not just correspondence requesting such.

RESPONSE

The NRC agreed that testing to Appendix J requirements was not required for these valves during the SEP (see NUREG-0821, Section 4.22.3). Note 17 to UFSAR Table 6.2-15 has been changed to reference this NUREG.

<L Attachment D Pago 4 of 12 Figure 6.2-14 includes a note that "CIV" is used to designate containment isolation valves on this and subsequent figures.

This notation was also used on some other figures to designate an isolation boundary, but has been subsequently modified by deleting the letter "V." It" is recommended that you identify "CI, or preferably "CIB, as notation for a containment isolation boundary or barrier by the use of a note on this figure. Also on Figure 6.2-14, Figure 6.2-16, and Figure 6.2-18, an arrow is shown for the check valves to designate flow direction. On other figures it appears that marked changes for check valves (Fi gures 6.2-37 and 6.2-38) were for the purpose to clarify flow direction (it is presumed that the intent is that the flow direction is from the upper side marking) yet no convention for such is provided. It would appear to be clearer to use the arrow symbol for consistency, since the presumed intent of marking does not work for a check valve shown in a vertical line such as in Figure 6'.2-15.

RESPONSE

All UFSAR figures have been updated to identify containment isolation boundaries as "CIB" and valves as "CIV". All items identified with a CIB or CIV have also been added to the UFSAR table. The arrows on check valves have been deleted and new arrows have been placed on process lines to indicate direction of flow as necessary (i.e., incoming and outgoing).

On Figure 6.2-19, valve 304B was added and on Figure 6.2-23, Val ve 304A was added. One of these figures could now be deleted since they are redundant. Also both figures identify the penetration as P-110 rather than by its full designation "P-110a (top)" as identified in Table 6.2-13.

RESPONSE

All UFSAR figures have been replaced; consequently, there is typically a separate drawing for each penetration. The figure titles have also been updated to be consistent with UFSAR Table 6.2-15.

On Figure 6. 2-33, the containment penetrati ons should be labeled as "P-124a (Supply)" and "P-124a (Return)" to identify each.

RESPONSE

All UFSAR figures have been replaced; consequently, there is typically a separate drawing for each penetration. The figure titles have also been updated to be consistent with UFSAR Table 6.2-15.

Attachment D Page 5 of 12 On Figure 6. 2-34, the containment penetrati ons should be labeled as "P-124b (Top) " and "P-124b (Bottom) " or other appropriate means to distinguish between the two penetrations that are currently designated as "P-124b."

RESPONSE

All UFSAR figures have been replaced; consequently, there is typically a separate drawing for each penetration. The figure titles have also been updated to be consistent with UFSAR Table 6.2-15.

On Figures 6. 2-40 and 6. 2-44, the locati on of containment relative to P-131 and P-209 (Top) is the reverse of what is shown (Figures 6.2-33 and 6.2-44 have the proper configuration shown) .

RESPONSE

The new UFSAR figures correctly show the location of containment for these two penetrations.

The "CIV" designation is improperly used for the reactor compartment cooler 1B on Figure 6.2-44.

RESPONSE

The CIV designation associated with the compartment cooler has been removed on the new UFSAR figure.

On Figure 6.2-46, the two penetrations should be identified to distinguish them as separate penetrations and with the same P-202 designation used in the title block and the table.

RESPONSE

All UFSAR figures have been replaced; consequently, there is typically a separate drawing for each penetration. The figure titles have also been updated to be consistent with UFSAR Table 6.2-15.

On Figure 6 ' 72I the pressure transmitters should not be designated as "CIVs" but rather as an isolation boundary.

RESPONSE

The new UFSAR figure for penetration 332a correctly identifies the transmitters as CIBs.

Attachment D Page 6 of 12 All check valves on Figure 6.2-75 (P-403 6 P-404) should be designated and shown as "CIVs." Likewise, Footnote 11 should be deleted for these valves as shown in Table 6.2-13.

RESPONSE

The steam generator tubes form a containment isolation boundary for the main steam, feedwater and auxiliary feedwater penetrations. The first isolation valve(s) outside containment for these penetrations have been added to the UFSAR table as requested. In addition, Footnote ll revised to show that these valves are CIVs, but they do not has been require Appendix J leak testing consistent with previous conversations between RG&E and the staff. RGGE has agreed to this change even though the current Technical Specification Table 3.6-1 explicitly states that these are not containment isolation valves with the understanding that the NRC approves that no Appendix J testing is required.

Where instruments are connected to a line upstream of the containment isolation valve, the instrument and its root valve should be listed in Table 6.2-13, similar to the other listing of instruments and root valves. This includes valves 885A, 885B and the associated PTs (whi ch should be numbered) as shown on figure 6.2-15. This is true also of valve 2856 and PI-933A and an unidentified instrument on Figure 6.2-18, valve 2859 and PI-933B on Figure 6.2-22, valve 4588 and PI-2141 on Figure 6.2-44, valve 4590 and PI 2232 on Figure 6.2-45, PI-(uni denti f'ied number) on Figure 6. 2-49, valve 8052 and PI-(unidentified number) on Figure 6.2-56, valves and PIs and FIs shown on Figure 6.2-6'3.

RESPONSE

All UFSAR figures have been replaced with the CIV/CIB testing procedure drawings. The instrumentation lines were reviewed and added as necessary; however, the changes with respect, to the penetrations identified in this question are provided below.

Added PT-923 and PI-923A as CIB's and 885B and 12407 as CIVs.

P-105 Added 869A and 2856 as CIV's (pressure indicator root valves are now closed).

P-109 Added 869B and 2858 as CIV's (pressure indicator root valves are now closed).

Added PT-922 and PI-922A as CIBs, and 885A and 12406 as CIV's.

Attachment D Page 7 of 12 P-201b Added PI-2141 as a CIB. Root valve 4588 is not required to be a CIV since the Service Water system is a CLIC.

P-204 No changes were made. See response to Question 23.

P-209b Added PI-2232 as a CIB. Root valve 4590 is not required to be a CIV since the Service Water system is a CLIC.

P-300 No changes were made. See response to Question 26.

P-308 Added FIA-2033 and TIA-2010 as CIBs. No root valves were added since the Service Water system is a CLIC.

P-311 Added FIA-2034 and TIA-2011 as CIBs. No root valves were added since the Service Water system is a CLIC.

P-312 Added 12500K (drain line) as a CIV and PI-2144 as a CIB. No root valves were added since the Service Water system is a CLIC.

P-315 Added FIA-2035 and TIA-2012 as CIBs. No root valves were added since the Service Water system is a CLIC.

P-316 Added PI-2138 as a CIB. No root valves were added since the Service Water system is a CLIC.

P-319 Added PI-2142 as a CIB. No root valves were added since the Service Water system is a CLIC.

P-320 Added 12500H (drain line) as a CIV and PI-2136 as a CIB. No root valves were added since the Service Water system is a CLIC.

P-323 Added FIA-2036 and TIA-2013 as CIBs. No root valves were added since the Service Water system is a CLIC

Attachment D Page 8 of 12

16. Test, vent, and drain valves that are used for Appendi x J local leak rate testing need not be listed Tabl e 6. 2-13.

However, valves provided for other purposes including testing should be listed as locked closed valves and identified as containment isolation valves. Therefore, it is suggested to identify those valves which are test, vent, or drain valves used for local leak rate testing with some notation on the figures, or by listi ng them in Table 6. 2-13 with an appropriate footnote. By providing this identification it will be clear as to which of the remaining valves are "CIVs" and subject to the Appendix J requirements. Clarification of the function of the following valves on the following figures should be noted:

Ficiure Valve 6.2-18 864A, 2825, 2829 6.2-22 864B, 2826, 2830 6.2-30 497, 498, 567, 576 6.2-56 8049

6. 2-73 7448, 7452, 7456, 8437, 8438, 8439

RESPONSE

All UFSAR figures have been replaced with the CIV/CIB testing procedure drawings. Connections used for Appendix J testing are now specifically identified on the figures while other connections have been added to the UFSAR table as CIVs. The changes made with respect to the penetrations identified in the question are provided below.

P-105 Added 864A and 2829 as CIVs. Valve 2825 is a test connection used for Appendix J testing.

P-109 Added 864B and 2830 as CIVs. Valve 2826 is a test connection used for Appendix J testing.

'I P-121a Added 497 and 498 as CIVs. Valve 567 is a test connection used for Appendix J testing. Valve 576 is a test connection downstream of two CIVs (567 and 508) and is not required to be a CIV.

P-300 No changes were made. See response to Question 26.

P-332a-d Added 7452, 7456, and 7448 as CIVs. Valves 8437, 8438, and 8439 are used for Appendix J testing.

17. No Question

'T I v*

L

Attachment D Page 9 of 12 On Figure 6.2-24, valve 959 should be noted as a CIV since ensures that the Residual Heat Removal (RHR) system is a it closed system on a CIS and should be listed in Table 6.2-13.

j Also, the valves to the safety in ecti on system inside containment should be shown and listed in the table as CIVs as well. as the check valve and the parallel val ve shown connecting to the letdown line.

this position, it If an exception is taken to should be justified.

RESPONSE

AOV 959 was already listed on Table 6.2-15 as a CIV; however, the "CIV" designator was'missing for the valve on Figure 6.2-

24. The figure has been revised accordingly. With respect to the "valves to the Safety Injection System", the wording on the UFSAR figure was incorrect. These two MOVs (852A and 852B) are used for low pressure, injection to the reactor vessel. Consequently, both lines are completely inside containment and have no affect upon the integrity of RHR as a closed system (i.e., the failure of 720 to isolate would not create a release path from containment through the subject two lines). In addition, both MOVs open on a SI signal to provide a RHR injection path; consequently, they cannot be closed due to their function and were therefore not added as CIVs. This issue is addressed in a letter from the D. Crutchfield, NRC(

to J. Maier, RGGE dated September 29, 1981.

The flowpath to the letdown line connects to the CVCS between the two sets of containment isolation valves for Penetration 112. Consequently, the isolation valve outside containment for Penetration 112 (i.e., 371) must fail in addition to 720 to create a release path from containment. However, no credible single failure exists between 371 and 720 (e.g., AOV versus MOV, separate ESFAS trains and control power sources).

Therefore, the check valve to the letdown line was not added.

Valves 9704A and 9704B on Figure 6.2-26 should be shown as CIVs and listed in Table 6.2-13.

RESPONSE

The steam generator tubes form a containment isolation boundary for the main steam, feedwater and auxiliary feedwater penetrations. The first isolation valve(s) outside containment for these penetrations have been added to the UFSAR table as requested. In addition, Footnote 11 has been revised to show that these valves are CIVs, but they do not require Appendix J leak testing consistent with previous conversations between RGGE and the staff. RGGE has agreed to this change even though the current Technical Specification Table 3.6-1 explicitly states that these are not containment isolation valves with the understanding that the NRC approves that no Appendix J testing is required.

Attachment D Page 10 of 12 states that valve 745 for

20. Footnote (return) 14 is to manuall y closed un be manua ic closure si nal.

until itenet is modified to h h u orna d wh y h as it not been implemented?

RESPONSE

AOV 745 was t to outage as stated in a be modified by the end of the l tt f Johnson, NRC, dated July 9, 1990. Howe benefit analysis, and the fact that n this penetration is required based on tzs stz.ll instruct to modification was canceled. 0 erat'p mons d o owing a CIS for additional redundancy. See 91 letter from R. Mecred y RGGE as mo i ied to reflect this change.

21. Please p rovi de change in the

'our 50. 59 eval uati classification for v on and the 42, valves 1813A and 1813B should be f

i it should

't exception t k n too thiis position, be justified.

RESPONSE

The 50.5 9 evaluation to remove valves 8 bl ld d 'th th o ' l Am d equest dated October 15,

'I ). However, the basis delete.on is contained in th e August Max.erg RGB(EI to DE Crutchfield, NRC in S'h SER SEP T 'c 4 (see NRC letter A 30 " 1982 f l NUREG-08 21 both references and reflects the posi.txon expressed in our Augustt 30, 1982 lette r, RGGE assumes fo that the NRC had agreed that 851A and 8 1813A an d 1813B,1 the UFSAR fi ure ha to include the necessary "CIV" IV designation.

d

22. Valve 1722 should be listed in Table 6.2-13 an be a 1ocked closed va provided.

valve. ve. If not, not a justification should be

RESPONSE

Valve 1722 is a locked closed valve. The UFSAR figure has been updated to reflect this.

l, Attachment Page ll of D 12

23. Valve 8074 on Figure 6'. 2-49 should be shown as a CIV and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

Manual valves 8074, 8074A, and PI-2 are not CIV's or CIB's since the hinged flange provides two containment isolation boundaries (i.e., two 0-rings). AOV 5869 is only listed on the UFSAR table since it can be used in place of the hinged flange during refueling to provide the necessary barrier.

24. Valve 5749 on Figure 6'. 2-52 should be shown as a CIV and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

This penetration was modified during the 1992 Refueling Outage so that valve 5749 is only used for Appendix J testing. The new figure correctly shows the modified penetration.

25. Valve 5754 on Figure 6.2-54 should be shown as a CIV and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

This penetration was modified during the 1992 Refueling Outage so that valve 5754 is only used for Appendix J testing. The new figure correctly shows the modified penetration.

26. Valve 8050 on Figure 6.2-56 should be shown as a CIV and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

Manual valves 8050, 8052, and PI-35 are not CIV's or CIB's since the hinged flange provides two containment isolation boundaries (i.e., two 0-rings). AOV 5879 is only listed on the UFSAR table since it can be used in place of the hinged flange during refueling to provide the necessary barrier.

27. The containment penetration should be shown on Figure 6.2-6'1.

If three penetrati ons, Top, Middle, and Bottom, exist, they should be identified and listed separately Table 6.2-13.

RESPONSE

UFSAR Table 6.2-15 was updated to listUFSAR each of the three penetrations individually. A separate figure is also provided for each penetration.

~ II 44 4

-" 1 s J

Attachment D Page 12 of 12

28. The drain val ves shown on Figure 6. 2-63 should be shown as CIVs and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

The drain valves were added to UFSAR Table 6.2-15 as CIVs; however, these valves are not locked closed. The drain valves are maintained normally closed during power operation per system lineup procedures and have "containment isolation boundary" control tags which are controlled by the CIV/CIB test procedures. This form of administrative control is considered acceptable since all plant personnel are instructed in the use of equipment tags. In addition, the Service Water system for these penetrations is a CLIC, thereby requiring a passive failure coincident with a LOCA before challenging the integrity of the drain valves.

29. Val ve 5 752 on Figure 6 269 shoul d be shown as a CIV and listed in the Table 6.2-13 as a locked closed valve.

RESPONSE

This penetration was modified during the 1992 Refueling Outage so that valve 5752 is only used for Appendix J testing. The new figure correctly shows the modified penetration.

30. Valves 3504A, 3505A, 3516, 3517, 3521, and 3506, 3507 or their associated atmospheric relief valves should be shown as CIVs on Figure 6.2-74, for penetrations 403 and 404, and listed in Table 6.2-13 as such.

RESPONSE

The steam generator tubes form a containment isolation boundary for the main steam, feedwater and auxiliary feedwater penetrations. The first isolation valve(s) outside containment for these penetrations have been added to the UFSAR table as requested. In addition, Footnote 11 has been revised to show that these valves are CIVs, but they do not require Appendix J leak testing consistent with previous conversations between RG&E and the staff. RG&E has agreed to this change even though the current Technical Specification Table 3.6-1 explicitly states that these are not containment isolation valves.

ATTACHMENT E UFSAR Table 6.2-15 and Figures r

6.2-13 through 6.2-78

GINNA/UFSAR e

SI APERTURE CARD Table 6.2-15 Also Availab1e Oo CONTAINMENT PIPING PENETRATIONS Aperture Card AND ISOLATION VALVING

~Sstem generator Penetration No valve/

a~aaada Isolation Position'alve al

~Te Blind flange valve Operator 1~e NA Zn Position Indication Control Room NA Position Relative to Containment Inside Normal asereadaa C

Cold Shutdown o/c Position At Immediate Postaccident~

C Power Pailure Trip CZS NA on Maximum Time

~eea 'P Isolation

~Pl SAR 6.2-13 re Class~

Notes see end of table 1, 2

'team inspection/ a2 Blind flange HA HA Outside C o/c C NA 6.2-13 1, 2 maintenance Puel transfer 29 al, a2 Blind flange NA Inside C o/c HA 6 2-13 2I 3 tube charging line to 100 370B al check NA HA Inside 0 C C HA HA NA 6.2-14 3B B loop CLOC a2 NA NA NA outside C C C NA NA NA 6.2-14 3B safety injection 101 870B al Check NA NA outside C C 0 NA HA NA 6 2-15 3B pump 1B dischazge 889B al check NA HA Outside C C 0 NA NA NA 6 2-15 3B CLOC a2 HA HA NA Outside C C C HA HA NA 6.2-15 3B 12407 bl Globe Manual No outside C C C NA NA ,NA 6.2-15 3B PZ-923A bl NA NA NA outside NA NA NA HA HA NA 6.2-15 3B

'PT-923'85B bl NA HA NA Outside NA NA NA NA NA NA 6.2-15 3B b2 Globe Manual No Outside a 0 0 NA NA HA 6.2-15 3B Alternate 102 383B al Check NA HA Inside C C NA HA NA 6.2-16 3B charging to A CLOC a2 NA NA NA Outside C C NA NA HA 6 2-16 3B cold leg construction fire 103 NA al welded cap HA NA Inside C NA NA NA 6.2-17 service water 5129 a2 Gate Manual No Outside LC NA HA NA 6.2-17 containment spray 105 862A al check NA HA outside C C 0 NA NA HA 6.2-18 3Br pump 1A CLOC a2 NA NA HA outside C C C NA NA HA 6.2-18 3B 2829 bl Globe Manual NA outside LC 0/C LC NA NA NA 6.2-18 3B Cap b2 NA NA NA Outside C o/c C NA HA NA 6.2-18 3B 869A cl Globe Manual Ho outside C C C NA NA NA 6.2-18 3B 2856 c2 Globe Manual No Outside C C C NA NA NA 6.2-18 3B 2825 dl Globe Manual No outside LC LC LC NA NA NA 6.2-18 3B Ball outside 2-18 2825A 864A d2 el Globe Manual Manual No No outside outside C

C C

C NA NA NA NA 'A NA 6 6.2-18 3B 3B 9 859A e2 Globe Manual No LC LC NA NA NA 6.2-18, 3B 10 859B e2 Globe Manual No Outside LC LC NA HA NA 6 2-18 3B 10 859C e2 Globe Manual No outside C C NA NA NA 6.2-18 3B 9, 10 r

Reactor coolant 106 304A al check NA HA Inside 0 0 C HA HA NA 6.2-19 3B pump A seal water CLOC a2 NA HA NA outside C C C NA NA NA 6.2-19 3B inlet sump A discharge 107 1723 al Diaphragm Air status Outside 0 o/c C PC Yes 60 6.2-20 2 to waste holdup 1728 a2 Diaphzagm Air Status Outside 0 o/c C PC Yes 60 6.2-20 2 tank Le end AI Pails as is CLOC Closed loop outside containment 0 Open Aov Air-operated valve CV check valve 0/C Open or closed BLC Breaker locked closed D Drain OMB outside missile barrier BLO Breaker locked open PC Pails closed R/G Red/green light on main control board Both R/G and Status Po Pails open S safety injection signal C closed IYB Inside missile barrier Sov Solenoid-operated valve CIB ss containment isolation boundary/bazrier J Appendix J connection Status white status light CZS, T containment 'solation signal LC Locked closed TC Test connection CIV containment isolation valve rov Motor-opezated valve V Vent CLIC closed loop 'nside containment MV Manual valve 6.2-95 REV 8 7/92 a

0 I

'B

'E C"'D la) 5j 1l'p J~

GINNA/UFSAR SI APERTURE CARD

. ~

Also Available O~

Aperture Card.

Table 6.2-15 CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

Position Position At Maximum Isolation Penetration Valve/ Isolation Valve Valve Operator Zndication In Control Position Relative to Hormal . Cold Immediate Power Trip on

~eec Time 'PSARure Notes of table

~satan No. Baunda~ Position" 2ree ~T8 Room Containment s

~ezarXon snctdoen Postaccident~ Pailure cts ~Pi Classd See end I

al Gate Motor Both outside 0 o/c C AI Yes 60 6.2-21 coolant 108 313 outside C C C NA NA HA 6 2-21

'eactor pump seal water CLOC a2 HA NA NA return line and excess letdown to VCT 862B al Check NA NA outside C C 0 NA NA NA 6.2-22 3B Containment spray 109 outside C C C NA NA NA 6.2-22 3B pump 1B CLOC a2 HA NA NA 2830 bl Globe Manual No outside LC o/c LC NA NA NA 6.2-22 3B b2 NA Na NA outside C o/c C NA NA ;HA 6.2-22 3B Cap 2-22 3B 869B cl Globe Manual No outside C C C NA NA NA 6 outside NA NA 6.2-22 3B 2858 2826 c2 dl Globe Globe Manual Manual No No outside C

LC C

LC C

LC NA NA HA 'A

'A 6 2-22 3B Ball outside NA 6.2-22 3B 2826A 864B d2 el Globe Manual Manual No No outside C

C C

C C

C NA NA NA 'A 6.2-22 3B 9 859A e2 Globe Manual No Outside LC LC LC NA NA , NA 6.2-22 3B 10 859B e2 Globe Manual No outside LC LC LC NA NA t NA 6.2-22 3B 10 859C e2 Globe Manual No outside C C C NA NA NA 6.2-22 3B 9, 10 Reactor coolant, 110a 304B Check NA NA Inside 0 0 C NA NA ,'A 6.2-23 3B NA NA outside C C C HA NA NA 6.2-23 3B pump B seal water CLOC NA inlet al, Globe Manual No outside NA NA 6.2-15 1 12 Safety injection 110b 879 a2 test line 720 al Gate Motor R/G Inside C 0 C AI No NA 6.2-24 3B 13, 14 Residual heat o/c 2-24 3B removal to B cold 959 a2 Globe Air Status outside C C C

PC NA Yes NA NA NA 6

6 2-24 3B 15 a2 NA NA NA HA C C leg CLOC al Globe Air R/G Inside o/c C C Ec Yes 60 6.2-25 1 16 Letdown to 112 200A o/c 6.2-25 16 nonregenerative 200B al Globe Air R/G Inside C C Ec Yes Yes 60 6.2-25 1

16 heat exchanger 202 al Globe Air R/G Znside C C C Pc Yes 60 60 6.2-25 1

371 a2 Globe Both outside 0 0 C PC 1 427 NA Globe Air R/G Inside 0 o/c C Po Yes NA 6.2-25 1 17 al outside c C 0 NA NA NA 6.2-15 3B Safety injection 113 870A Check NA NA Outside 0 NA NA 6.2-15 3B pump ~M discharge 889A al check NA NA C C NA, NA NA 6.2-15 3B CLOC a2 HA NA NA Outside C C C NA 12406 bl Globe Manual No outside C C C NA NA NA 6.2-15 3B PI-922A bl HA NA NA Outside NA HA NA NA NA NA 6.2-15 3B PT-922 bl HA HA NA Outside NA NA NA NA NA NA 6e2-15 3B bl NA NA NA outside C C C NA NA NA 6.2-15 3B cap 0 0 NA NA 6 2-15 3B 885A b2 Globe Manual No Outside 0 NA al Stop-check Motor Outside 0 0 0 AI No NA 6.2-26 18 Standby auxil- 119 9704A R/G Outside NA NA HA 6.2-26 iary feedwater 9723 al Globe Manual No LC LC LC 6.2-26 19 line to steam CLIC a2 NA NA NA Inside C C C NA NA HA generator 1A Nitrogen to 120a 846 al Globe Air Both Outside C o/c C Fc Yes 60 6 '-27 3A 8623 a2 check HA NA Inside o/c o/c C NA NA HA 6.2-27 3A accumulators al Globe Air Status outside C o/c C PC Yes 1 60 6 2-28 2 Pressurizer 120b 539 Outside 0 0 0 NA HA NA 6.2-28 2 relief tank to 546 a2 Globe Manual No gas analyzer 6.2-97 REV 8 7/92~

3

'-)V

~i ~

F I

s~

)

GINNA/UFSAR

.~

Sl--

APEQ,JURE CARD Table 6.2-15 CONTAINMENT PIPING PENETRATIONS A~ gvajta'ble OA AND ISOLATION VALVING (Continued)

Aperture Card Position Position At Maximum Valve Zndication Position Isolation

~Sstem Penetration No.

Valve/

~Baunda Isolation poededon'alve

~Te Operator e

zn control Room Relative to Containment Normal ooeeandon cold Shutdovn zmmediate Postaccident~

Pover Failure Trip odd on Time

~deo 'FSARre

~Fi Class~

Notes see end of table water to 121a 508 al Diaphragm Air Both outside C o/c FC Yes 60 6 2-29 3A 'akeup pressurizer 529 a2 check NA NA Znside o/c o/c NA NA 6.2-29 3A relief tank Nitrogen to 121b 528 al check NA NA Inside , c 0/C NA NA 6.2-30 3A pressurizer 547 a2 Globe Manual No Outside ~

LC o/c NA 6.2-30 3A 20 relief tank Containment 121c PT945 al HA NA NA Outside NA NA NA NA NA NA 6.2-31 pressure 1819A a2 Globe Manual No outside 0 0 0 HA HA NA 6.2-31 transmitters PT946 bl NA NA NA Outsi.de NA HA NA NA NA NA 6.2-31 PT945 and PT946 1819B b2 Globe Hanual No Outside 0 0 0 NA NA NA 6.2-31 Reactor coolant 123a 1600A NA Globe solenoid No Outside 0 o/c C PC Yes NA 6.2-32 17 drain tank to gas 1655 al Globe Manual No Outside 0 0 0 NA HA 6.2-32 analyzer line 1789 a2 Diaphragm Air Status outside 0 o/c C Fc Yes 60 6.2-32 Standby auxil- 123b 9704B al Stop-check Motor R/G Outside 0 0 0 AZ No NA 6.2-26 18 iary feedvater 9725 al Globe Manual No Outside C C C NA NA NA 6.2-26 line to steam 9724 al Globe Manual No Outside C C C NA NA NA 6.2-26 generator 1B CLIC a2 NA NA NA Znside C C C NA NA NA 6.2-26 19 Excess letdown 124a 743 al Check NA Inside C C C NA 6.2-33 heat exchanger CLIC a2 NA NA Inside C C C NA 6.2-33 22 cooling water supply Postaccident air 124b 1572 al Diaphragm Manual No Outside NA NA NA 6.2-34 sample to common 1573 a2 Globe Hanual No Outside NA NA NA 6.2-34 return 1574 a2 Diaphragm Manual No Outside NA NA NA 6.2-34 Excess letdown 124c 745 al Globe R/G Outside Fc No NA 6.2-33 21 heat exchanger CLZC a2 NA NA Inside NA NA NA 6 '-33 22 cooling water return Postaccident air 124d 1569 al Diaphragm Hanual No outside NA 6.2-34 5 sanple to c fan 1570 a2 Globe Manual No Outside NA '-34 1571 a2 Diaphragm Manual No outside C

LC NA 'A NA 6 6.2-34 5

5 Component cooling 125 759B al Gate Motor R/G outside AI No NA 6.2-35 water from CLOC a2 NA NA NA Outside NA HA , HA 6.2-35 23 reactor coolant

'A pump 1B Component, cooling 126 759A al Gate Motor R/G Outside AI Ho NA 6.2-36 2 water from CLOC a2 NA HA Outside NA NA NA 6 '-36 2 23 reactor coolant pump 1A Component cooling 127 749A al Gate Motor R/G Outside AI No NA 6.2-37 3B vater to reactor 750A a2 Check NA NA Inside NA NA NA 6.2-37 3B coolant pump 1A Component cooling 128 749B al Gate Hotor R/G Outside AZ No NA 6.2-38 3B vater to reactor 750B a2 Check NA HA Inside NA NA NA , 6.2-38 3B coolant pump 1B 6.2-99 REV 8 7/92

f ye p h i

GINNA/UFSAR a "r SI APERTURE Table 6.2-15 CARD PIPING PENETRATIONS (Continued)

'ONTAINMENT AND ISOLATION VALVING A1SO AVai1abIe On Position Position At Aperture Card Maximum Valve Indication Position Isolation

~Sstem Penetration No.

Valve/

~Bounda Isolation roelrlon'alveT9(pe operator

~pe Zn Control Room Relative to contadntant Normal

~erat1an Cold shutdown Immediate Postaccident~

Power rallnre Trip crc on Time

~ce UPSAR Ei cCure class'otesoi table See end Reactor coola~nt drain tank and

~99 1713 1799 al a2 check Diaphragm NA Manual NA No outside outside C

LC o/c o/c C

LC NA HA HA NA NA NA 6.2-39 6 '-39 3A 3A 20 pressurizer 17Q6 bl Diaphragm Air status Outside 0 C C Ec Yes 60 6.2-39 3A relief tank to 1787 b2 Diaphragm Air Status outside 0 C C PC Yes 60 6.2-39 3A containment vent header component cooling 130 814 al Gate Motor Both outside AI Yes 60 6.2-40 water from CLIO a2 NA NA NA Inside NA NA NA 6.2-40 22 reactor support cooling component cooling 131 813 al Gate Motor Both Outside 0 0 AZ Yes 60 6.2-40 water to reactor CLIC a2 NA NA NA Inside C C NA NA NA 6.2-40 22 support cooling containment mini- 132 7970 al Butterfly Air Both Inside o/c o/c C Pc Yes 3 6.2-41 purge exhaust 7971 a2 Butterfly Air Both outside C o/c C Ec Yes 3 6.2-41 Cap a2 NA NA NA Outside C C C NA NA NA 6.2-41 Residual heat 140 701 al Gate Motor R/G Inside C 0 AZ Ho NA 6.2-42 13, 14 removal pump 2763 al Globe Manual No Inside C C NA NA NA 6.2-42 15 suction from A 2786 al Globe Manual Ho Inside C C NA NA NA 6.2-42 hot leg a2 HA NA NA outside C C NA NA NA 6.2-42 Residual heat; 141 850A al Gate Motor R/G Outside C C No NA 6.2-43 24 removal pump A a2 NA NA NA outside C C NA NA 6.2-43 15 suction from 1813A bl, b2 Gate Motor R/G Outside C o/c Ho NA 6.2-43 14, 25 sump B Residual heat, 142 850B al Gate Motor outside C 0 AI No NA 6.2-44 24 removal pump B suction from CLOC 1813B bl, a2 b2 NA Gate Motor'/G NA 'NA R/G outside Outside C

o/c C

C NA AZ NA

'Ho NA NA 6.2-44 6.2-44 15 14, 25 sump B Reactor coolant 143 1003A al Diaphragm Air status outside 0 o/c C PC Yes 60 6 2-45 drain tank 1003B al Diaphragm Air status outside 0 o/c C Ec Yes 60 6.2-45 discharge line 1709G al Gate Manual No Outside C C C NA NA NA 6.2-45 1722 al Diaphragm Manual No Outside LC LC NA NA HA 6.2-45 1721 a2 Diaphragm Air Status Outside 0 0 C Fc Yes 60 6.2-45 Reactor 201a 4757 al -Butterfly Manual No Outside 0 0 0 NA HA NA 6.2-46 26 compartment 4775 al Gate Manual No Outside C C C NA NA NA 6.2-46 cooling unit A CLIC a2 HA NA NA Inside C C C NA NA NA 6.2-46 27 Reactor 201b 4636 al Butterfly Manual No Outside 0 0 0 NA NA NA 6.2-47 30 compartment 4776 al Gate Manual No outside C C C NA NA NA 6.2-47 cooling unit B PI-2141 al NA NA NA outside HA NA NA NA NA NA 6.2-47 return CLIC a2 HA NA HA Inside C C C NA NA NA 6.2-47 27 B hydrogen 202a 1076B al Diaphragm Manual No Outside HA NA NA 6.2-48 recombiner 10211%1. a2 Globe solenoid Status Outside Ec Yes 3 6 '-48 28 (pilot)

B hydrogen 202b 1084b al Diaphragm Manual No outside NA NA NA 6.2-48 5 recombiner (main) 1021381 a2 Globe solenoid Status Outside PC Yes 3 6.2-48 5 28 Containment 203a PT947 al NA 'NA NA outside NA NA NA NA NA NA 6.2-49 2 pressure 1819C a2 Globe Manual No Outside 0 0 0 NA NA NA 6.2-49 2 transmitter PT948 bl NA NA NA Ou side NA NA NA NA NA NA 6.2-49 2 PT947 and PT948 1819D b2 Globe Manual No outside 0 0 0 NA NA NA 6.2-49 2 6.2-101 REV 8 7/92 92j.mj40i59

I C

t P

l'

GINNA/UFSAR SI ApERTURE Table 6.2-15 CARD CONTAINMENT PIPING PENETRATIONS Also Avadable On AND ISOLATION VALVING (Continued)

Aperture Card Position Position At Maxim Valve Indication Position Isolation

~sstem Penetration No Valve/

Boundary Isolation poattton Valve

~e operator T~e In Control Room Relative to tontatnnnnt Hormal

~eratioa cold shutdown Immediate postaccident Power Pailure Trip CZS on Time

~(sec UPSAR

~Fi re class'otesof table see end Postaccident air 203b 1563 al Diaphragm Manual No outside LC HA 6 '-50 sample from D fan 1564 a2 Globe Manual Ho outside C NA 'A 6.2-50 1565 a2 Diaphragm Manual No outside NA 6.2-50 Postaccident air 203c 1566 al Diaphragm Manual No outside LC HA 6.2-50 sample from 1567 a2 Globe Manual No outside C NA NA 6.2-50 common header 1568 a2 Diaphragm Manual No outside Na NA 6.2-50

'urge al, Blind flange NA Inside C 0 NA NA NA 6.2-51 2, 29 supply duct 204 HA a2 NA 6.2-51 5869 NA Butterfly Air Both outside C o/c PC Yes NA 29 Loop B hot leg 205 955 NA Globe air Status Inside C Fc Yes NA 6.2-52 17 sample 956D al . Needle Manual No Outside 0 NA NA NA 6.2-52 966C a2 Globe Air Status outside C Fc Yes 60 6.2-52 Pressuriser 206a 953 NA Globe air status Inside C FC Yes NA 6.2-53 6.2-53 17 liquid space 956B al Needle Manual No outside 0 Na NA NA sample 966B a2 Globe Air Status outside C Ec Yes 60 6 '-53 Steam generator A 206b CLIC al NA NA Inside NA NA 6.2-54 19 sample 5735 a2 Gate Status Outside Yes 60 6.2-54 Pressuriser steam 207a 951 NA Globe air Status Inside C C C Ec Yes NA 6.2-55 6.2-55 17 space sample 956F al Needle Manual No outside 0 0 0 HA HA HA 966A a2 Globe Air Status outside C C C Ec Yes 60 6.2-55 Steam generator B 207b CLIC al NA HA HA Inside C NA NA NA 6.2-56 19 sample 5736 a2 Globe Air Status outside 0 Fc Yes 60 6.2-56 Reactor 209a 4635 al Butterfly Manual No outside NA HA NA 6.2-47 26 compartment. 4637 al Gate Manual No outside NA NA NA 6.2-47 cooling unit B CLIC a2 NA NA HA Inside HA HA NA 6.2-47 return Reactor 209b 4638 al Gate Mannual No outside 0 0 0 NA HA NA 6.2-46 30 compartment 4758 al Butterfly Manual No outside C C C NA NA NA 6 2-46 cooling Unit A PZ-2232 al NA NA NA outside NA NA NA NA NA Na 6.2-46 CLIC a2 NA NA NA Inside C C C NA NA NA 6 2-46 supply al outside LC LC NA HA NA 6.2-57 Oxygen makeup to 210 1080A Globe Manual No LC Yes 6.2-57 28 A s B recombiners 10214Sl a2 Globe solenoid Status Outside C C C PC 3 102148 NA Globe solenoid Status outside C C C Fc Yes 3 6.2-57 15, 28 10215Sl a2 Globe solenoid Status outside C C C FC Yes 3 6.2-57 28 102158 NA Globe solenoid Status outside C C C PC Yes 3 6.2-57 17, 28 al, Blind flange NA Znside 0 NA NA NA 6 2-58 5 2d 29 Purge exhaust 300 NA a2 NA 2-58 duct 5879 NA Butterfly Air Both Outside o/c PC Yes NA 6 5 29 auxiliary steam 301 6151 al Gate Manual No Outside NA NA NA 6.2-59 4 sunplv to 6165 a2 Gate Manual No Outside NA NA NA 6 2-59 4 containment Auxiliary steam 303 6152 al Diaphragm Manual No outside HA NA NA 6.2-59 4 condensate return 6175 a2 Diaphragm Manual No outside NA HA NA 6.2-59 4 6.2-103 REV 8 7/92

S

'I 4

1 l

GINNA/UFSAR SI A,PERTURE Table 6.2-15 CARD CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

Also Available On Position Position At Aperture Card Maximum Valve Indication Position Isolation Penetration Valve/ Isolation Operator In Control Relative to Trip

~sszem No. Boundaro posdndon'alve~Te TQQB Room Containment Normal

~ei:asian Cold Shutdown Immediate Postaccident~

Power Failure CIS on

~sea Time UESAR

~Pi re Class~

Notes See end of table

'A A hydrogen 304a 1076A al Diaphragm Manual No outside HA NA 6.2-60

'8 recombiner 1020581 a2 Globe Solenoid Status Outside C PC Yes 3 6.2-60 (pilot)

A hydrogen 304b 1084A al Diaphragm Manual No outside LC NA NA 6.2-60 recombiner (main) 1020981 &2 Globe solenoid status Outside C Fc Yes 6.2-60 28 Containment air 305a 1554 al Diaphragm Manual No outside LC NA NA HA 6.2-61 sample 1555 a2 Globe Manual Ho outside C NA HA NA q 6.2-61 postaccident 1556 a2 Diaphragm Manual No Outside LC NA NA NA 6.2-61 Containment air 305b 1598 al Diaphragm Air Both Outside 0 Fc Yes 60 6.2-62 3A sample inlet 1599 a2 Diaphragm Air Both outside 0 PC Yes 60 6.2-62 3A Containment air 305c 1557 al Diaphragm Manual No outside NA NA NA 6.2-61 sample 1558 a2 Globe Manual No outside NA NA NA . 6.2-61 postaccident 1559 a2 Diaphragm Manual No outside NA HA NA 6.2-61 Containment air 305d 1560 al Diaphragm Manual No outside NA 6.2-61 sample 1561 a2 Globe Manual No outside NA 6.2-61 postaccident 1562 a2 Diaphragm Manual No outside NA 6.2-61 Containment air 305e 1596 al Globe Manual No outside .0 NA HA HA 6.2-63 sample out 1597 a2 Diaphragm Air Both Outside '0 Ec Yes 60 6.2-63 Fire service 307 9227 al Gate Air Both outside ~ C Fc Yes 60 6.2-64 water 9229 a2 check NA NA Inside C NA NA NA 6.2-64 service water 308 4629 al Butterfly Manual No outside Lo o/c Lo NA NA NA 6.2-65 26 from A fan cooler 4633 al Gate Manual No outside ",c C C HA HA NA 6.2-65 PZA-2033 al HA HA NA outside 'NA NA NA HA NA NA 6.2-65 TZA-2010 al NA NA HA Outside ';NA NA NA NA NA NA 6.2-65 CLIC a2 NA NA HA Inside C C C NA NA NA 6.2-65 27 Mini-purge supply 309 7445 al Butterfly Air Both Outside 0/C Ec Yes 6.2-66 7478 a2 Butterfly Air Both Inside 0/C Ec Yes 6.2-66 Instrument air to 310a 5392 al Globe Air Both Outside i 0 FC Yes 60 6.2-67 3A containment 5393 a2 check NA NA Inside (

0 NA HA NA 6.2-67 3A Service air to 310b 7141 al Gate Manual Ho Outside o/c NA NA NA 6.2-68 3A containment 7226 a2 check NA NA Inside .C o/c NA NA NA 6.2-68 3A service 311 4630 al Butterfly Manual No outside 'Lo o/c Lo NA NA HA 6.2-65 26 fan cooler al water'rom B 4634 Gate Manual No Outside ,"'

C C NA NA NA 6.2-65 PZA-2034 al HA NA NA Outside )NA NA NA NA NA NA 6.2-65 TZA-2011 al NA NA NA Outside 'NA NA NA NA NA NA', 6.2-65 CLIC a2 HA HA NA Inside '

C C NA NA HA'A 6.2-65 27 Service water to 312 4642 al Butterfly Manual No outside 'LO o/c Lo NA HA 6.2-65 30 D fan cooler 4646 al Gate Manual No outside ' C C NA NA NA 6.2-65 12500K al Globe Manual No outside (C C C NA NA NA 6.2-65 PZ-2144 al HA HA NA Outside NA NA NA NA NA NA, 6.2-65 CLIC a2 NA HA NA Inside C C C NA NA NA 6.2-65 27

'A Leakage test 313 NA al Blind flange HA NA Inside ',

C HA NA 6.2-69 depressurization 7444 a2 Butterfly Motor status outside C AI Yes NA 6.2-69 Qa.i 2 i40159 6.2-105 REV 8 7/92

0 Pr' t

J 1

0 4

'0 p

C')

Wa~

GINNA/UFSAR SI

'PERTURE Table 6.2-15 CARD CONTAINMENT PIPING PENETRATIONS AND ISOIATION VALVING (Continued)

Also Availabte On Position Position At Aperture Card Maximum Valve Zndication Position Isolation

~sstem Penetration No.

Valve/

Baundaxar Isolation Position'alve~Te Operator

~re Zn Control Room Relative to Containment Normal rrdaranddaann Cold shutdown Immediate Postaccident~

Power Failure Trip on dre Time

~aea 'PSARure

~F1 class4 Notes see end of table Service water 315 4643 al Butterfly Hanual No Outside ee Lo o/c Lo NA NA NA, 6.2-65 26 from C fan cooler 4647 al Gate Manual No outside C C C NA NA NA 6.2-65 PIA-2035 al NA HA NA outside NA NA NA NA NA NA 6.2-65 TZA-2012 al NA HA HA outside NA NA NA NA NA NA 6.2-65 CLIC a2 NA HA HA Inside C C C NA NA NA 6.2-65 27 Service water to 316 4628 al Butterfly Manual No outside Lo o/c Lo NA NA NA'A 6.2-65 30 B fan cooler 4632 al Gate Manual No outside C C C NA HA 6.2-65 PZ-2138 al NA NA NA outside NA NA HA NA NA NA, 6.2-65 CLZC a2 NA HA NA Znside C C C NA NA 6.2-65 27 NA'A'A Leakage test 317 NA al Blind flange HA HA Inside NA NA 6.2-70 supply 7443 a2 Butterfly Motor Status outside AI Yes 6.2-70 Deadweight tester 318 HA al, a2 NA NA NA NA NA NA NA NA NA HA NA 31 Service water to 319 4627 al Butterfly Manual No outside Lo o/c Lo NA 6.2-65 30 A fan cooler 4631 al Gate Manual No outside C C C NA 6 2-65 PI-2142 al NA HA NA outside NA HA NA NA 6.2-65 CLIC a2 NA HA NA Inside C C C NA 6.2-65 27 service water to 320 4641 al Butterfly Manual No outside Lo o/c NA NA HA 6.2-65 30 c fan cooler 4645 al Gate Manual No outside C C C NA NA NA 6.2-65 PZ-2136 al NA HA NA outside NA NA NA NA HA HA 6.2-65 12500B al Globe Manual No outside C C C HA HA HA 6.2-65 CLIC a2 HA NA NA Inside C C C HA NA NA 6.2-65 27 A steam generator 321 5738 al Globe Air status Outside o/c PC Yes 60 6.2-71 blowdown CLIC a2 NA NA NA Znside C NA NA 6.2-71 19

'a B steam generator 322 5737 al Globe Air status outside 0/C Pc Yes 60 6.2-72 blowdown CLIC a2 NA NA NA Inside C NA HA NA 6.2-72 19 service water 323 4644 al Butterfly Hanual No outside Lo o/c Lo NA NA NA 6.2-65 26 from D fan cooler 4648 al Gate Manual No Outside ,

c C C NA NA NA 6.2-65 PZA-2036 al NA NA NA Outside ~

NA NA HA NA NA NA 6.2-65 TZA-2013 al NA NA NA Outside NA NA HA NA NA NA 6 2-65 CLIC a2 NA NA NA Inside C C C HA NA HA 6.2-65 27 Demineralized 324 8418 al Globe Air Both outside C o/c Pc Yes 60 6.2-73 water to 8419 a2 check HA NA Znside C o/c NA NA NA 6.2-73 containment Hydrogen monitor 332a 922 al Gate solenoid Both Outside C C PC Yes 3 6.2-74 instrumentation 924 al Gate solenoid Both Outside C C FC Yes 3 6.2-74 line CLOC a2 NA NA NA Outside C C NA HA NA 6.2-74 32 7452 bl Globe Manual No Outside C C HA NA NA 6.2-74 cap b2 NA HA NA Outside C C NA NA NA 6.2-74 Bydrogen monitor 332b 923 al Gate Solenoid Both outside Pc Yes 3e 6.2-74 instrumentation a2 NA NA NA Outside NA NA NA 6.2-74 line 7456 bl Globe Manual No outside NA HA 6.2-74 Cap b2 HA NA NA Outside NA NA HA 6.2-74 containment 332c PT944 al NA NA NA Outside HA NA NA NA NA NA 6.2-75 pressure 1819G a2 Globe Hanual No outside 0 0 0 NA NA NA 6.2-75 transmitters PT949 bl NA NA NA outside NA NA HA NA NA NA 6.2-75 PT944, PT949, and 1819E b2 Globe Hanual No outside 0 0 0 NA HA NA 6.2-75 PT950 PT950 cl NA NA NA olltside NA NA HA NA NA NA 6.2-75 1819P c2 Globe Hanual No outside 0 0 0 NA NA HA 6.2-75

.OSSA >40 i 59'- 6.2-107 REV 8 7/92

~ ~ C

~ + e

~ 'a~~

J Q

\

C

(

.o n -w ~ GINNA/UFSAR SI APERTURE 6.2-15 CAR9 'able CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVIHG (Continued)

Also Available On Position Position At Aperture Card Maximum Valve Zndication Position Isolation penetration Valve/ Isolation Position'alve operator In Control Relative to Normal cold Immediate -

Pove Trip on Time UFSAR Notes

~sstem No. ~nonndo ~e ~pe Room Containment td<<erntdoonn don<<down Postaccident~ Failure CIS ~tool Ficiure Class~ see end of table Hydrogen monitor 332d 921 al Gate solenoid Both outside FC Yes 3 6.2-74 instrumentation CLOC a2 NA NA HA outside HA NA NA 6.2-74 32 line 7448 bl Globe Manual No Outside NA NA NA 6.2-74 Cap b2 HA NA NA Outside NA NA NA 6.2-74 Main steam from A 401 3413A al Globe Manual No outside 0 C o/c NA NA NA 6.2-76 4 18 steam generator 3455 al Globe Manual No outside C C C HA NA NA 6.2-76 4 3505A al Gate Motor R/G outside C C o/c AZ No

'A 6.2-76 4 3505C al Gate Manual No outside C C C NA NA NA 6.2-76 4 3507 al Gate Manual No outside 0 C o/c HA NA NA 6.2-76 4 3507A al Gate Manual No outside C C C NA NA NA 6.2-76 4 3517 al Sving check Air R/G outside 0 C C AI No NA 6.2-76 4 18 3521 al Gate Manual No outside 0 0 o/c NA NA NA 6.2-76 4 18 3615 al Gate Manual No Outside C C C HA HA NA 6.2-76 4 3669 al Gate Manual Ho outside 0 C o/c NA NA NA, 6.2-76 4 18 11027 al Gate Manual No Outside C C C NA HA NA 6.2-76 11029 al Gate Manual No outside C C C NA NA

'A 6.2-76 4 11031 al Gate Manual No outside C C C NA NA NA 6.2-76 4 PS-2092 al NA NA outside NA HA NA HA NA NA 6.2-76 4 6 PT-468 al NA NA NA Outside NA NA NA HA NA 6.2-76 4 6 PT-469 al NA NA NA Outside NA NA NA NA HA NA 6.2-76 4 6 PT-469A al NA NA NA Inside NA NA NA NA NA NA 6.2-76 4 6 PT-482 al NA HA NA outside NA NA NA HA NA NA 6.2-76 4 6 End caps al NA NA NA Outside C C C NA HA NA 6.2-76 4 33 CLZC a2 NA NA NA Inside C C C NA NA NA 6.2-76 4 19 Main steam from B 402 3412A al Gate Manual No outside 0 C o/c NA NA =

HA 18 steam generator 3456 al Globe Manual No outside C C C NA HA NA 6.2-77 3504A 3504C 3506 al al al Gate Gate Gate Motor Manual, Manual R/G No No outside Outside outside C

C 0

C C

C o/c C

o/c AI NA NA Ho NA NA NA NA NA

'.2-77 6.2-77 6.2-77 6.2-77 18 3506A al Gate Manual No Outside 'C C C HA NA NA 6.2-77 3516 al sving check Air R/G Outside 0 C C AI Ho

'A 6.2-77 18 3520 al Gate Manual No Outside 0 0 0 HA HA NA 6.2-77 18 3614 al Gate Manual No outside C C C NA HA HA 6.2-77 3668 al Globe Manual Ho outside 0 C o/c NA NA HA 6.2-77 18 11021 al Gate Manual No Outside C C C NA HA HA 6.2-77 11023 al Gate Manual No Outside C C C NA NA NA, 6.2-77 11025 al Gate Manual No Outside C C C NA NA HA 6.2-77 PS-2093 al NA NA NA Outside HA NA NA NA NA HA 6.2-77 6 PT-478 al NA NA NA Outside NA HA NA NA HA

'A 6.2-77 6 PT-479 al NA HA NA outside NA NA NA NA NA NA 6.2-77 6 PT-483 al Outside 6.2-77 End caps al NA NA ~

NA NA HA HA Outside NA C

NA C

NA C

NA HA 'AHA 'A 6.2-77 6

33 CLIO a2 HA HA HA Inside I C C C HA HA

'A 6.2-77 19 Feedvater line to 403 3993 al Check NA NA outside 0 C HA NA NA 6.2-78 34 A steam generator 3995X al Globe Manual No outside C C HA NA HA 6.2-78 4000C al Check NA NA outside C o/c NA NA NA 6.2-78 34 4003 al check NA NA Outside C o/c HA NA HA 6.2-78 34 4011A al Globe Manual No outside C C NA HA HA 6.2-78 4003A al Gate Manual No Outside C C HA NA NA 6.2-78 4099E al Gate Manual No outside C C NA NA HA'A 6.2-78 8651 al Gate Manual No Outside C C HA NA 6.2-78 CLZC a2 HA NA NA Inside C C NA HA NA 6.2-78 19 6.2-109 REV 8 7/92 9215140159-

a Q g~4

~A

GINNA/UFSAR SI APERTURE Table 6.2-15 CARD PIPING PENETRATIONS AND ISOLATION VALUING (Continued)

'ONTAiNMENT Aiso Availab1e On Position Position At Apert(ire Cat'~ .

Valve Indication Position Isolation Penetration Valve/ Isolation Position'alve operator Zn Contzol Relative to Normal cold Immediate Pover Trip on Time UPSAR Notes

~sstem Ho ~Bunda ~e ~T Room Containment cSs rand'n Shutdown Postaccident~ FFS.Sara CSF ~Pi re Class4 See end of table Peedvater line to 404 3992 al Check NA NA outside 0 C C NA NA NA 6.2-78 34 al B steam generator 3994E Globe Manual Na outside C C C NA NA NA "

6.2-78 4000D al check NA NA Outside C C o/c NA NA NA F 6 ' 78 34 4004 al Check NA NA outside C C o/c NA NA NA 6.2-78 34 4012A al Globe Hanual No outside C C C NA NA NA 6.2-78 3994X al Gate Manual Na outside C C C NA NA NA 6 2-78 4004A al Gate Manual No Outside C C C NA NA NA 6.2-78 8650 al Gate Hanual No outside C C C NA NA NA 6.2-78 CLZC a2 NA NA NA Inside C C C NA NA NA t, 6.2-78 Personnel hatch 1000 NA al NA NA NA Inside o/c NA HA NA 3. 8-31 NA NA a2 NA NA NA Outside o/c NA NA NA 3.8-31 NA Equipment hatch 2000 NA al NA NA Inside 0/C NA NA 3. 8-30 NA NA a2 NA NA outside 0/C NA NA 3.8-30 NA

'This tvo-chazacter designator identifies the branch line which contains the valve (a, b, c, d, or e) and the isolation boundary (1 or 2 since each line contains tvo bazriezs).

Refezs to position immediately folloving receipt of containment isolation signal and containment ventilation isolation signal.

'The maximum isolation time does not include diesel start time-nor instrument delay time.

'Refers to classes defined in Section 6.2.4.4.

Notes only used to supplement Section 6.2.4.4.

Notes (1), Penetration number ends.

2 vas added as a result of EWR 4998 to The innermost gasket for each flange (i.e., gasket facilitate steam generator maintenance activities during reduced inventory operation. This penetration is closed by a double-gasketed blind flange on both closest to containment vali) provides a containment barrier. Therefore, both flanges are necessary for containment integrity.

(2) This penetration is pzovided vith redundant seals and is closed during normal operation.

(3) The end of the fuel transfer tube inside containment is closed by a double-gasketed blind flange, to prevent leakage of spent fuel pool vater into the containment during plant operation. This flange also sezves as protection against leakage from the containment folloving a loss-of-coolant accident.

(4) The charging system is a closed system outside containment (CLOC). Verification of this closed system as a contain ent isolation boundary is accomplished via normal system operation (2235 psig).

(5) The safety injection system is a closed system outside containment (CZsOC). Verification of this closed system as a containment isolation boundary is accomplished via insezvice and/or shutdown leakage checks.

(6) The pressure transmitter assembly, by its design, provides a containment pressure boundary. The integrity af this boundary is verified by annual leakage tests.

(7) This penetration vas only utilized during initial plant construction and is maintained inactive.

(8) The containment spzay system is a closed system outside containment (CLoc). verification of this closed system as a containment isolation boundary is accomplished via inservice and/or shutdovn leakage checks.

(9) This valve may be opened during containment spray pump testing since there vill alvays be at least one isolation boundary betveen the valve and containment for the duration of the test.

(10) Manual valves 859A, 859B, and 859C are CZVs for both penetrations 105 and 109.

(ll) A second isolation barrier is provided by the volume control tank and connecting piping per letter fzom DE D Dilanni, NRC, to R. w. Rober, RGaE, dated January 30, 1987. This barrier is not required to be tested.

a 9218 140 I 59 6.2-111 REV 8 7/92

~

'~

I

'N t

l ~

I

't

G INNA/UFSAR SI APERTURE CARD Table 6.2-15 Also Available On CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

Aperture Card

( ) o y one isolation b~ier is pzovided since there are two Event V check valves in 'the safety injection cold legs, and two check valves and a normally closed motor-operated valve in the safety injectio hot legs.

This configuzation was accepted by the HRc during the sEp (MUREG 0821, section 4.22.2).

(13) 10 CFR 50, Appendix J containment leakage testing is not required per D.M. Crutchfield, MRC, letter to J. E. Maier, RGSE, dated May 6, 1981.

(14) MOVs 18)3A, 1813B, 720, and 701 are maintained closed at power with their breakers locked off.

(15) The residual heat removal system is a closed system outside containment (cLoc). verification of this closed system as a containment isolation boundary is accomplished via 'inservice and/or shutdown leakage checks.

(16) Containment isolation signals were added to AOVs 200A 200B, and 202 since AOV 427 fails open on loss of power. The isolation signal for these three valves is relayed from AOV 427.

(17) This valve receives a containment i,solation signal; however, credit is not taken for this function since the valve is inside the missile barrier or outside the necessary class break boundary. Thezefore, this valve is not subject to 10 CFR 50, Appendix J leakage testing, noz does it require a maximum isolation time. The, containment isolation signal only enhances isolation capability.

(18) This valve is normally open at power since required.

it is required during power operation or increases the reliability of a standby system. However, this valve can either be closed from the contzol room or locally when (19) The main steam, main feedwater, and standby auxiliary feedwater penetzations take credit for the steam generator tubes as a closed system inside containment. Verification of this closed system as a contain ent isolation boundary is accomplished via normal power operation. The isolation valves outside containment for these penetzations are not required to be Appendix J tested.

(20) Manual valves 547 and 1793 are locked closed and leak tested to provide equivalent pzotection for GDC 56 and 57 (see UFSAR Section 6.2.4.4.4.1, Class 3A).

(21) Operations is instructed to manually close AOV 745 following a containment isolation signal to provide additional redundancy.

(22) The component cooling water system piping inside containment for this penetration is a closed system (CLIC). Verification of this closed system as a containment isolation boundazy is accomplished via insezvice and/or shutdown leakage checks.

(23) The component cooling water system piping outside containment for this penetration is a closed system (CLOC). Verification of this closed system as a containment isolation boundary is accomplished via insezvice and/or shutdown leakage checks.

(24) sump lines are in operation and filled with fluid following an accident; therefore, 10 cFR 50, Appendix J leakage testing, is not required for this valve. See D. M. crutchfield, MRC, letter to J. E. Maier, RGCE, dated May 6, 1981.

1 (25) There is no second containment barrier for this branch line. However, Movs 1813A and 1813B are maintained closed at power and tested to Appendix J. These lines are also filled with water post LOCA, thus providing a barrier to the zelease of containment atmosphere.

(26) This manual valve is subject to an annual hydrostatic leakage test and is not subject to 10 CFR 50, Appendix J leakage testing.

(27) The service water system piping inside containment for this penetration is a closed system (CLIC). Verification of this closed system as a containment isolation boundary is accomplished via inservice and/or I

shutdown leakage checks.

(28) This solenoid valve is maintained inactive in the closed position by removal of its dc control power.

(29) The flanges and associated double seals pzovide containment isolation and ensure that containment integrity is maintained for all modes of operation above cold shutdown. During cold shutdown when the flanges are removed these valves provide isolation for containment shutdown purge and exhaust. These valves do not reouire 10 cFR 50, Appendix J leakage testing, nor a maximum isolation time.

(30) The service water system operates at, a higher pressure than the contain ent accident pressure and is missile protected inside containment. Therefore, this manual valve islused for flow control only and is not subject to 10 CFR 50, Appendix J leakage testing. See letter from J. E. Maier, RGCE, to D. M. Crutchfield, MRC, dated August 30, 1982.

(31) This penetration is decommissioned and welded shut.

(32) Acceptable isolation capability is provided for these instrument lines by two isolation boundaries outside containment. One of the boundaries outside containment is a Seismic Categozy I closed system which is subjected to Type C leakage testing under 10 CFR 50, Appendix J.

(33) These end caps include those found on the sensing lines for pS-2092, pT-468, FT-469, PT-469A, and PT-482 (penetration 401) and Ps-2093, pT-479, and pT-483 (Penetration 402).

(34) This check valve can be open when containment isolation is required in order to provide necessary feedwatez or auxiliazy feedwater to the steam generators. The check valve will close once feedwater is isolated to the afiected steam generatoz.

9212 3.40 l.S:9 6.2-113 REV 8 7/92

.0 J

tJ[

I BI TEST TEST CONNECTION CONNECTION t

DOUBLE-GASKETED BLIND FLANGE NO. 2

,. SPENT FUEL PIT DOUBLE~KETED BLIND FLANGE NO. 29 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-13 S/G Inspection/Maintenance, Penetration o. 2 Vi Fuel Transfer Tube, PenetrationiVo. 29 REV 6 12/90

RCS CHARGING LINE PENEHM.TION 100 I! I III CIy P100 K'OB ORS O

I Y/J NOTE DESCRIPTION ontanment Basement (Regenerative Hx Area)

Auxiliary Bldg Basement (RWST Area)

LRM should be located in Containment Basement (Regenerative Hx Area)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-14 Reactor Coolant System Charging Line Penetration 100 PTT-23.8 Revision 1 azv 8 7/92

SITO LOOP A SITO LOOP 8 and SITEST LINE PENETRATIONS 101 110b and 113 022 CLLL gsI g

'll 0210 12404 TC/J mal I

P113 a72A 0/J I

I LJ 21 A BLc a1 a1 I II ll LI arlrr TC/J oI 4< I al al oa 2010 Sl PVLLP C TO 201 TO P101 CV 200 I

aaaa LC 12401 Ca}

cw a21L' TO w 1 n" a2Q pr 023 h 02a LC 028 UNE P110b

( era 12ala 2$ LO I

I 1 0/J 0/J L

NOTE DESCRIPTION Containment near B Stairway Auxiliary Bldg Basement Sl Pump Area LRM should be located in Aux Bldg Basement ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT.

UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-15

~

Safety Injection System Penetrations 101, 110b, and 113 PTT'-23.19 Revision 1 tv 8 7/92

ALTERNATE CHARGING LINE PENETRATION 102 D

221 22)8 II CIY P102 Rg 383B 525 f.)

2227 I

TC/J V/J NOTE DESCRIPTION Containment Basement (Regenerative Hx Area)

Auxiliary Bldg Basement (Outside RWST Area)

LRM should be located in Containment Basement (Regenerative Hx Area),

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR F'OWER PLANl UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-16 Alternate Charging Line Penetration 102 PTl -23.10 Revision 1 REV 8 7/92

CONSTRUCTION FIRE SERVICE WATER PENETRATION 103 8!Z WELOED CAP LC (CAP)

Y/J CIB 5380 5129 CONSTRUCTION ClV FlRE SERVICE 5$ 29A WATER CONNECTIONS TC/J NOTE DESCRIPTION Containment Basement (Regenerative Hx Area)

Auxiliary Bldg (RWST Area)

(1) LRM should be located in Auxiliary Bldg Basement near RWST ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANAI YSIS REPORT Figure 6.2-17 Construction Fire Service Water Penetration 103 Pal=23.49 Revision 1 REV 8 7/92

"A" CONTAINMENTSPRAY HEADER PENETRATION 105 I

FROM PENElRATION 109 TEST IjNE Y/J TO RWST IQ 3[5 8&DA

! BLC eros! M S

~NQ 869A 8&DB CIV BLC R

PI

! 7 285e

! CIV I

11621 NOTE DESCRIPTION Containment Basement near RHX Auxiliary Bldg Basement (Behind RWS1)

LRM should be located in Auxiliary Bldg Basement near CS Pumps ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-18 Containment Spray Header A Penetration 105 PTT-23.18A Revision 1 REV 8 7/92

"A" RCP SEAL WATER LINE PENETRATION 106 Fl 1

298B 298A CIV P108 O 9303 300A 303A I 80ll 9304 303C cg TC/J V/J 277A

! LC

! 300B 275 301B 301A Fl I 2224 1 NOTE DESCRIPTlON Containment Basement (Regenerative Hx Area)

Auxiliary Bldg Basement (RWST Area)

LRM should be located in Containment Basement near Regenerative Hx (2) Operated with Reach Rods by the CS Pumps Located 20 ft above the floor behind RWST Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-19 Reactor Coolant Pump A Seal Water Line Penetration 106 PTT-23.9A Revision 1 aZV 8 7/92

SUMP "A" DISCHARGE PENETRATION 107 I

I..

g)B V/J TC/J 82 m) Pc 8 10006 I FC FC P1071 CIV 1002 1725 1723 10035 I

1072 0 I

10012 D/J 10036 10106 1759 1757 S 10023 1760 1758 SUMP A PUMPS SAMPIE PUMP NOTE DESCRIPTION CNMT Basement B Stairway Area Auxiliary Bldg Basement Fan Cooler Area LRM should be located in Containment Basement (8 Stairway Area)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-20 Sump A Discharge Penetration 107 Pal=23.23 Revision 1 azv 8 7/92

0 RCP SEAL WATER RETURN & EXCESS LETDOWN PENETRATION 108 I

TC/J g(3 117 FROM 314 o!Pz V

REACTOR 2213 tMB Pl COOLANT PUMP OMB 118 SEALS 318A FILTER P1O8) ctv 315C OMB 313 IMB I SEAL 322A RETURN I FtLTER 311e 2212 Lo I 2214 I V/J FROM EXCESS I LETOOVe HEAT EXCHANGER I,

3858 3828 385A FROM FROM

~ (It ~ Ao REACTOR REACTOR cootANT CoolANT PUMP PUMP SEALS SEALS NOTE DESCRIPTION Containment RHX Area Auxiliary Bldg Basement (RWST Area)

(1) LRM should be located in Containment Basement near RHX (2) On stairwell next to AOV-386 (3) Located in Regenerative Hx Area (4) Operated from outside the Seal Return Filter Room with reach rods ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-21 Reactor Coolant Pump Seal Hater Return and Excess Letdown Penetration 108 PTT-23.11 Revision 1 REV 8 7/92

'I Wl ~

I

"B" CONTAINMENT SPRAY HEADER PENETRATION 109 Pl 55 FROM PENETRATION 105 11620 2825A TEST LINE TO RWST 85QC 2825 CIV 2830 8648 LC LC CIV CIV M S

860 ~~5 BLC ~h P109 xo i

M ~W

~lO

~

gD

(

8698 2826 Qu 8600 Cl CIV CIV BLC CP I

o I 2826A Pl CIV PI I 77 I TC/J TC/J 2859 NOTE DESCRIPTION Containment Basement (RHX Area)

Auxiliary Bldg (RWST Area)

LRM should be located in Auxiliary Bldg near CS Pumps ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-22 Containment Spray Header B Penetration 109 Pal=23.18B Revision 1 aVr 8 7/92

"B" RCP SEAL WATER LINE PENETRATION 110a 303C 5

277A CO LC O 9301 P110a I 304B 275 3018 301A 9302 I TC/J V/J NOTE DESCRIPTlON Containment Basement (Regenerative Hx Area)

Auxiliary Bldg Basement (RWST Area)

LRM should be located in Containment Basement near Regenerative Hx (2) Operated with Reach Rods by the CS Pumps Located 20 ft above the Iloor behind RWST Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-23 Reactor Coolant Pump B Seal Water Line Penetration 110a PTT-23.98 Revision 1 REV 8 7/92

RESIDUAL HEAT REMOVAL TO B COLD LEG PENETRATION 111 (PENETRATION <4D)

RESIDVAL HEAT REMOVAL LOOP OUTlET VALVE I

INSTALLED 7198 BLANK (PENETRATION 112)

UNE 718A SAMPlZ SYSTEM FC T

958 702 ~I 0

721 P111 ~5 BLQ ES 627 S S 88 M Ll 2TEO BU'LC 852A 8528 TO REACTOR VESSEL 2780A 2748 51" ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure l).2-24 Residual Heat Removal to Loop B Cold Leg Penetration 111 Isolation from 33013-1247, Revision 19 REV 8 7/92

LETDOWN LINE FROM REACTOR COOLANT SYSTEM PENETRATION 112 TO PREESUIRZER REUEF TAHX HRHX RM E64~

I 1~)

2' I I CN I 2 I-5 LOOP AREA RHR FENCE V/J I IgR r R v

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FROM RHR STETEM I

I NOTE DESCRIPTION Containment Basement (Regenerative Hx Area)

Auxiliary Bldg Basement (RWST Area)

LRM should be located in Containment Basement Regenerative Hx Area Located on Mid Floor near Square Comer Located in B Loop area at base of ladder to RCP Platform (4) Located Auxiliary Bldg Basement nesr B CS Pump ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-25 Letdown Line from Reactor Coolant System Penetration 112 PTT-23.6 Revision 1 REV 8 7/92

STANDBY AUXILlARYFEEDWATER TO STEAM GENERATOR A PENETRATION I'I9 AND 123b I

ls us TC IL O

9719A s726 Sl 0 CIV 4J Q FROM

'X STANDBY 4J AUXIUARY 9706A g705A P 1 19 9702A 9704A FEEDWATER LO BLC PUMP C I I lA O 9727 I 9723 CIV IO 9724 O CIV CIV M FROll STANDBY AUXLJARY 9706B 0705 B P123b 9702B 97~B FEEDWATER LO LO PUMP D I

BLC lA O 9719B I

LC 0725 9722B CIV TC 'LJ D

ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAI. SAFETY ANALYSIS REPORT Figure 6.2-26 Standby Auxiliary Feedwater to Steam Generator A Penetrations 119 and 123b Isolation from 33013-1238, Revision 8 REV 8 7/92

NITROGEN TO ACCUMULATORS PENETRATION 120a Pl 627 TC/J I

~" 8621 VENT

~i 9 S627 8625 8626 o

5 O 624 CIV P120a CtV EP 8&23 846 862S 944 S628 g O

I I

2831 I

Ul TC/J V/J

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Vl &

4J g 0 ~

CC w>

Og e9

~s o:O QI0 NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg Mid Roor (SFP Hx)

LRM should be located in Containment Mid Floor (Square Comer)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-27 Nitrogen to Accumulators Penetration 120a Pl%-23.46 Revision 1 azV 8 7/92

PRT GAS ANALYZER PENETRATION 120b TO CONTAINMENT VENT HEADER hC 4 527 Ld jc LU hl CL lL a 4J t4 IK CIV 492 N 538 P)20b CIV M 546 539 Vl 4l FC O CL I

O 493, TC/J -..

NOTE DESCRIPTION Containment Mid Floor (Square Corner)

Auxiliary Bldg Mid Floor (SFP Hx)

LRM should be located in the Auxiliary Bldg Mid Floor near SFP HX (2) Downstream vent point ROCHESTER GAS AND ELECTRIC CORPORATION H. h. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-28 Pressurizer Relief Tank Gas Analyzer Penetration 120b PTT-23.1 Revision 1 Szv 8 7/92

PRT MAKEUP WATER PENETRATION 121a Ig h!m 497 O le O'J

~V O. CI 4~ P121o cg~ 9307 CIV CIV EL ~ 529 508 o~

I FC x5 568 567  ! 49S LJ TC/J  !

,0/J TC/J 576 NOTE DESCRIPTION Containment (Square Corner)

Auxiliary Bldg Mid Floor (SFP Hx)

LFIM should be located in the Containment Square Comer Area 3/4'ipe connection (3) Located 10 ft above the floor, adjacent to missile barrier (4) Located 15 ft above the floor ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-29 Pressurizer Relief Tank Makeup Nater Penetration 121a PTT-23.3 Revision 1 REV 8 7/92

PRT N PENETRATlON 121b bC 4 CL lal IL CF P121b! gy olC 545 441 1662 I 52B 547 'K 0

I 496 495  ! 494

.ci.  !

V/J I

NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg. Mid Floor (SFP Hx)

LRM should be located in Containment in the Square Corner ROCHESTER GAS AND ELECTRIC CORPORATION R. F- GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-30 Pressurizer Relief Tank N2 Penetration 121b PTT-23.2 Revision 1 REV 8 7/92

CONTAINMENT PRESSURE TRANSMITTERS PT-946 AND PT-946 PENETRATION 121 c GIB PT 946 V/J v/s CIV 1819A CIV 1819B 1818A 18188 OPEN PIPE TC/a P121 c NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg Mid Floor (SFP Hx)

(1) LRM should be located in Containment Mid'Floor in the Square Comer ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-31 Containment Pressure Transmitters PT-945 and PT-946 Penetration 12lc P 1T-23.17A Revision 1 tv 8 7/92

l gK RCDT TO GAS ANALYZER PENETRATlON 123a 1004 1717A Cl FC FC s T CIV 1020 1600A 1655 1789 5 O

I 1709F X

TC/J NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg Mid Floor (SFP Hx)

(1) LRM should be located in the Aux Bldg Mid Floor near SFP Hx.

(2) Downstream vent point ROCHESTER GAS AND ELECTRIC CORPORATION E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-32 Reactor Coolant Drain Tank to Gas Analyzer Penetration 123a PTT-23.21 Revision 1 azV 8 7/92

CCW TO FROM EXCESS LETDOWN HX PENETRATlON 124a and 124c

=l:

I

I I

I-

/ I I

'TC/J 2776 I

742C I I

743A I K REACTOR cooLANT I I

o o

I I

'XCESS I

I K

W I 'X I o LETDOWN HEAT iol PI24 I

X oo EXCHANCER ( 3I IH ]

I I X I 2725 o I 4 o a.au Aov hgh O s-vl

/

I I 744 I

I l~v/J 0 o uS> FC I n o3<

~ clv P124c I

ov) 742B 745 i I

I gg ~M

>oS Qo RHR FENCE I

o 2727 cg l

v/J NOTE DESCRIPTION Containment Mid Floor ("B" Stairway)

Auxiliary Bldg Mid Floor (RWST)

LRM should be located in containment basement near "B'tairway CV internals have been permanently removed ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-33 Component Cooling Water to and from Excess Letdown Heat Exchanger Penetrations 124a and 124c PTT-23.30 Revision 1 REV 8 7/92

CONTAINMENT POST ACCIDENT AIR SAMPLE C FAN PENETFIATION 124b and 124d I

I..

LC LC TC/J Y/J I~

124b CIV x

CIV O 1569 1571 1570 CIY I M

I I

X ID O

I gw z CP LC LC I Z+ TC/J

>I I X 124d CIY CIV Y/J

<o 1572 1574 NOTE DESCRIPTION Containment Mid Floor (B Stairway)

Auxiliary Bldg Mid Floor (RWST Area)

(1) LRM should be located in Containment Mid Floor near B Stairway ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-34 Containment Postaccident Air Sample (C Fan)

Penetrations 124b and 124d PTT-23.50C Revision 1 azv 8 7/92

CCW FROM B RCP PENETRATlON 126 oim 7548 I 765A 12305B 1 7568 Fl CCW 1 FROM IKACTOR Ll 765B COOLANT PULIP 18 P 125 CIV 762B cp~

7508 I 0 gA au@

2731 O

8 7578 O I

758B NOTE DESCRIPTION Containment Mid Floor (B Stairway)

Auxiliary Bldg. Mid Floor (RWST)

(1) LRM should be located in the Containment Mid Floor near 'B'tairway ROCHESTER GAS AND ELECTRIC CORPORATION R. K GINNA NUCLEAR" POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-35 Component Cooling Water from Reactor Coolant Pump 1B Penetration 125 PTT-23.29 Revision 1 REV 8 7/92

&tbsp 1" CCW FROM A RCP PENETRATION 126 FO I

765D I

'X 12308B 7$ 6A CCW Fl I FROM REACTOR 61 TC/J M COOLANT PUilP i 1A IC P126 CIV 702A 1 759A I 0

)

ug CP 2729 X O

O V

757A O 75BA NOTE DESCRIPTION Containment Mid Floor (B Stairway)

Auxiliary Bldg. Mid Floor (RWST)

(1) LRM should be located in the Containment Basement near the Rx Compt Coolers ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-36 Component Cooling Water from Reactor Coolant Pump 1A Penetration 126 PTT-23.28 Revision 1 REV 8 7/92

COMPONENT COOLING WATER TO REACTOR COOLANT PUMP 1A PENETRATION 127 N

$ 13 2723 MO OX I

N 0.

O O.

5>

cc z 749B o~

"80 2732 Nx LJ z Oa 742A I

7$ 1A VW N 740A N ~>g zg) e aL 0Ã CIY P127 i 617 750A BLO Q co 75DC 2761 2730 R~~

oI 761r TC/J tc/J 0/J NOTE DESCRIPTION Containment Mid Floor (B Stairway)

Auxiliary Bldg. Mid Floor (RWST)

LRM should be located in Containment Basement near Rx Compt Coolers and Auxiliary Bldg. Mid Floor near RWST ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-37 Component Cooling Mater to Reactor Coolant Pump lA Penetration 127 PTT-23.26 Revision 1 REV 8 7/92

CCW TO "8'CP PENETRATION 128 I,

lh/SZ y

I

$ 13 a5 I M P Og, g>

a8 07M Up o~

85 aH I

742A 7511 CIV g 740B $ 8 le OA 8

CN 7608 P1Zg ] gl 75QD a I u. Q8 I

761E TC/J P/J NOTE DESCRIPTION Containment Mid Floor (B Stairway)

Auxiliary Bldg Mid Floor (RWST)

LRM should be located in Containment near the 'B'tairway and Auxiliary Bldg Mid Floor near RWST ROCHESTER GAS AND ELECTRIC CORPORATION H. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6. 2-38 Component Cooling Water to Reactor Coolant Pump B Penetration 128 PTT-23.27 Revision 1 azV 8 7/92

RCDT GAS HEADER PENETRATION 129 FC FC CIY CIV 1675 1787 1786 1676B Y/J

! 1016 IL

! 40 0 O'ISA K W

TC/J ClY Y/J 1014 16Oa 2

NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg Mid Floor (Behind SFP Hx)

LRM should be located in Auxiliary Bldg Mid Floor (Behind SFP Hx)

Disconnect tubing for downstream vent ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWEH PLA'NT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-39 Reactor Coolant Drain Tank Gas Header Penetration 129 PTT-23.20 Revision 1 REV 8 7/92

CCW FROM 0 RX SUPPORT CLRS PENETRATION 130 and 131 SUPPLY fROQ COMPONENT COOUNQ WATER PUMPS

))5 eIlg 817 M BLO I O)

I I 3I "I"

o)~

5s-I

~(8 oE)~

I HICH I YENTS TC/J TC/J QwQ 1734, 1733

-s.E 815A CY P130 P131i 8th I REACTOR 815 8 SUPPORT

~

818 COOLERS I 271 g 1723 I I Y/J 0/J D/J Y/J i

I I NOTE DESCRIPTION nta1nment Ml Foor ("B" talrway)

Auxiliary Bldg Mid Floor (Behind RWST)

(1) LRM should be located in Containment Mid Floor near the 'B'tairway ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-40 Component Cooling Hater from and to Reactor Support Coolers Penetrations 130 and 131 PTT-23.24 Revision 1 REV 8 7/92

MINI PURGE EXHAUST PENETRATION 132 r-! 9 I

Lal FC FC K '

8 CIS T m~

CP P)52 CIV CIV 7970  ! 7971 C) Pg aK CC

! C I

79710 I TC/J NOTE DESCRIPTION Containment Mid Floor (Square Comer)

Auxiliary Bldg Mid Floor (Behind SFP Hx)

LRM should be located in Auxiliary Bldg Mid Floor near SFP Hx.

Open pipe with debris screen ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-41 Mini-Purge Exhaust Penetration 132 PTT-23.34 Revision 1 REV 8 7/92

RESIDUAL HEAT REMOVAL FROM A HOT LEG PENETRATION 140 REFUEUNO WATER STQRAGE TANK 252 5

M g~

X P140 701 27d5 nss CN 00 PENETRATION 111) $54 I!'-

TO RESIDUAL HEAT REMOVAL SYSTEM ROCHESTER GAS'ND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-42 Residual Heat Removal from Loop A Hot Leg Penetration 140 Isolation from 33013-1 247 Revision 19 REV 8 7/92

SUMP "B" TO "A" RCDT PUMP PENETRATloN 1a1 REFUEUN0 WATER STQRACE TANK TO RCDT PUMPS (CASED SmrZM)

~I!I Y/J 1813C CV

$ 8$ 8A !R BLD CO BLO M 1818D TC/J 705C

!i5 C 851A 850A BLC P

NOTE 'DESCRIPTION CNMT Sump B Auxiliary Bldg Basement LRM should be located in Auxiliary Bldg Basement (connected in RHR Sub Basement)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-43 Sump B to Reactor Coolant Drain Tank Pump A Penetration 141 P1T-23.5A Revision 1 aZV 8 7/92

SUMP "B" TO "B" RCDT PUMP PENETRATlON 't42 REFVEUNO WATER STORAGE TANK (CLOSED SmrZM)

M

~ 7048 BLO M 705D  !

~s

~ III 8518 850B 48 TC/J 8LC 1818F CIV 18158 i P BLD Y/J 1818K TO RCDT PUMPS NOTE DESCRIPTION CNMT Sump B Auxiliary Bldg Basement LRM should be located in Auxiliary Bldg Basement (connected in RHR Sub Basement)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-44 Sump B to Reactor Coolant Drain Tank Pump B Penetration 142 PTT-23.5B Revision 1 REV 8 7/92

I A" + Pe\-

0

RCDT DISCHARGE PENETRATION 143 Pl 01 179SG 172@A LC FC FROM FUEL LC TRANSFER CANAL 1726 DRAINS CIY CIY 1722 100$ A 1811A CONTAINMENT SUMP RCDT PUMPS I

I Pl I s- 01 I

I go I 8~

~C 1725A I

0 I P145) CIV FC I ceo 1721 I ~

o I I 1727 L CIV S43 I 10058 I

I TC/J I I 17090 18118 I

I I 1 NOTE DESCRIPTION Containment Sump B Auxiliary Bldg Sub Basement LRM should be located at entrance to Containment Sump B ROCHESTER GAS AND ELECTRIC CORPORATION R. E GlluNA'NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-45 Reactor Coolant Drain Tank Discharge Penetration 143 PUT-23.22 Revision 1 REV 8 7/92

REACTOR COMPARTMENT COOLING UNIT A SUPPLY AND RETURN PENETRATIONS 201a AND 209b I I ml g

I! I CIS tO 47940 4590 I

P201a CIV CIV 4757 4758 REACTOR S

, COMPARTLtENT COOmR 4BSA 5 O [

)A I

l ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-46 Reactor Compartment Cooling Unit A Supply and Return Penetrations 201a and 209b Isolation from 33013-1 250-3 Revision 7 REV 8 7/92

REACTOR COMPARTMENT COOLING UNIT B SUPPLY AND RETURN PENETRATIONS 201b AND 209a l

sl. .Ig I

nl5 Pl

)4O CIB N

CI CLIC P20$ b cv 4BJ5 REACTOR X 4837A COMPARTMENT CIY 4778 4778A 5 O

$ B C I

I ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-47 Reactor Compartment Cooling Unit B Supply and Return Penetrations 201b and 209a Isolation from 33013-1250-3 Revision 7 REV 8 7/92

B HYDROGEN RECOMBINER AIN AND PILO PENETRATION 202a and 202b I

Ri"

)

IL "I Lc S P202b I- 1076B cN 102'11$

c)v 10211$ 1 a'

1075B 10203$ 1 I a, V/J I m Tc/J )O 10203S Q

P

)a 10204S I o

5) 5 5 Lo 10204S1 P202a 102'l3$

O~ o3

>w a 1084B c)v CIV 1021351 8426 1063B

.c/. I V/J NOTE DESCRIPTION Containment Mid Floor (above "A Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Containment Mid Floor above "A" Accumulator (2) Located in Intermediate Bldg Basement outside Hot Shop ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNa NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6,2-48 Hydrogen Recombiner B (Main and Pilot)

Penetrations 202a and 202b PTT-23.51B Revision 1 REV 8 7/92

CONTAINMENT PRESSURE TRANSMITTERS PT-947 AND PT+48 PENETRATION 203a CIB CIB Y/J Y/J CIV CIV 1819C 1819D 1818C 1818D DPEN P)PE TC/J P205a NOTE DESCRIPTION Containment Mid Floor (Above "A Sl Accumulator)

Intermediate Bldg. (Sample Shed)

(1) LRM should be located in Containment Mid Floor above "A" Sl Accumulator ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT'PDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-49 Containment Pressure Transmitters PT-947 and PT-948 Penetration 203a PT1=23.17B Revision 1 REV 8 7/92

CONTAINMENT POST ACCIDENT AIR SAMPLE FAN PENETRATION 203b and 203c I

KQ V/J P2Mb CIV CIV 1505 1565 1584 CV 8

o>

I TC/V V/J P205o CIV OV 4o CC~

1560 ~1500 1587 CIV t

NOTE DESCRIPTION Containment Mid Floor (above A Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Containment Mid Floor above 'A" Accumulator ROCHESTER GAS AND'LECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-50 Containment Postaccident Air Samp1e (Fan D)

Penetrations 203b and 203c PTT-23.50B Revision 1 REV 8 7/92

PURGE SUPPLY PENETRATION 204 TC/J o

[g FC

~CO T

P204 CIV (BLIND FLANGE)

I I

8074 I

I 8074K Pl V/J NOTE DESCRIPTION Containment Mid Floor (behind "A'ccumulator)

Intermediate Bldg Basement (near Controlled Access Fans)

LRM should be located in Containment Mid Floor behind 'A'ccumulator ROCHESTER GAS AND ELECTRIC CORPORATION H. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-51 Purge Supply Penetration 204 PTT-23.35.1 Revision 1 azv 8 7/92

~ Q M I RCS LOOP B HOT LEG SAMPLE PENETRATlON 205 h!L'ml j 10M'27C DELAY FROll RCS LOOP P205 B-HOT IMn O

I TC/J Y/J FROM RCS LOOP A-HOT NOTE DESCRIPTION Containment Mid Floor (Above 'A" Accumulator)

Intermediate Bldg (Sample Shed)

LRM should be located in Intermediate Bldg near Sample Shed ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR PovlER PLANT '

UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-52 Reactor Coolant System Loop B Hot Leg Sample Penetration 205 PTT-23.12C Revision 1 RVr 8 7/92

PRESSURIZER LIQUID SAMPLE PENETRATION 206a 10001 FC 9g1 P206a CN 927 958K 9668 O

95SH SSQD TC/J V/J Y/J NOTE DESCRIPTION Containment Mid Floor (above "A" Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Containment Mid Floor above "A" Accumulator ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-53 Pressurizer Liquid Sample Penetration 206a PTT-23.12B Revision 1 aEv 8 7/92

"A" STEAM GENERATOR SAMPLE PENETRATION 206b y/J 5711 5781 I I I w 5769D FC I

Ld 0 3 N

aQ gx I I 0>

X P206b ~5 5748A su, 5785 y/J Oa TO A S/G 8 LOWDOWN 5749 5105 TC/J NOTE DESCRIPTION Containment Mid Floor (above "A" Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Intermediate Bldg (Sample Shed)

(2) Located on S/G Platform ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-54 Steam Generator A Sample Penetration 206b PTT-23.13A Revision 1 REV 8 7/92

PRESSURIZER STEAM SAMPLE PENETRATION 207a I

10000 g! g

~8 51 FC I

40 g

921A 958F 9510 958G 999 E TC/J I Y/J Y/J NOTE DESCRIPTION Containment Mid Floor (above "A" Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Containment Mid F!oor above 'A'ccumulator.

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINhlA NUCLEAR'POVVER PL'ANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-55 Pressurizer Steam Sample Penetration 207a PTI -23.12A Revision 1 REV 8 7/92

"8'TEAM GENERATOR SAMPLE PENETRATlON 207b I

II

I s ~soE 8! I I

5786 y/J TOSS G 5754 570e TC/J NOTE DESCRIPTION Containment Mid Floor (above "A" Accumulator)

Intermediate Bldg (Sample Shed)

LRM should be located in Intermediate Bldg (Sample Shed)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-56 Steam Generator B Sample Penetration 207b PTT-23.13B Revision 1 REV 8 7/92

"A" AND "B" HYDROGEN RECOMBINER OXYGEN MAKEUP PENETRATION 210 1079 m/~ ~La S T 31:

~O 1021 4S1 1021 4S CIY 55 1080A Clg CIV S T S T g&

1021 5S1 10215S CIY NOTE DESCRIPTION Containment Mid Floor (Above "A" Accumulator)

Intermediate Bldg (Sample Shed)

(1) LRM should be located in Containment Mid Floor Above "A'ccumulator (2) Spool pieces located in Intermediate Bldg Basement below Sample Shed ROCHESTER GAS AND ELECTRIC CORPORATlON R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-57 Hydrogen Recombiner A and B Oxygen Makeup Penetration 210 PTT-23.51C Revision 1 REV 8 7/92

PURGE EXHAUST PENETRATtON 300 I.

!m my~ ~~! y.

$ NA I! I

'!N P300 CIB (euwn ru~eE)

PI 55 v/v AIR SUPPLY NOTE DESCRIPTION Containment Top Floor (Mezzanine)

Intermediate Bldg. (floor above steam header)

(1) LFIM shoUld be located in Containment on Top Floor Mezzanine (2) Intermediate Bldg Top Floor ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINrIA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-58 Purge Exhaust Penetration 300 PTT-23.36.1 Revision 1 BEV 8 7/92

AUX STEAM SUPPLY AND AUX STEAM CNDST RETURN PENETRATIONS 301 AND 303 STEAM FROM HOUSE HEATINC BOILER

.Ig 7050 Q )cr 7941 I

LC PM1 CIY CIY 6151 6165 7040 I

7945 V/J I

T 4J I 7944 I 7946 mg D/J O~

'5O5~

W~

975 car CIV TC/J 6175 6152 STRAINER OX O

SPACE HEATERS Y/J NOTE DESCRIPTION Containment Mid Floor ('A" Fan Area)

Intermediate Bldg (TDAFWP Area)

(1) LRM should be located in Intermediate Bldg (TDAFWP Area)

(2) V-7941 is in overhead above 'A" Chiller Unit ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-59 Auxiliary Steam Supply and Condensate Return Penetrations 301 and 303 PTT-23.40 Revision 1 aZV 8 7/92

A HYDROGEN RECOMBINER AIN AND PILO PENETRATION 304a and 304b IB

~la K I cL'~f LC S

+lal Q QM ID 0I 107~ CIV 10205S 8$ CL CIV 10205S1 g

1075A 10207S1 TC/J v/J 10207S X oO IY O

I I 10202S S

10202S1 10%A CIV 10209S 10209S1 L TC/J v/J NOTE DESCRIPTION Containment Mid Floor ("A Recirc Fan Area)

Intermediate Bldg (TDAFW Pump Area)

(1) LRM should be located in Containment Mid Floor 'A'ecirc Fan Area (2) Located in Intermediate Bldg Basement outside Hot Shop ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-60 Hydrogen Recombiner A (Main and Pilot)

Penetrations 304a and 304b PTT-23.51A Revision 1 aVr 8 7/92

CONTAINMENTPOST ACCIDENT AIR SAMPLE PENETRATION 305a 305c and 305d FROM 8 FAN Y/J ClV 1555 CV 1555 FROM A FAN I Y/J CIV g 1557 TC/J P305d LC COMMON RElURN Y/J CIV 156O 1552 CIV 1551 NOTE DESCRIPTION Containment Mid Floor ( A" Recirc Fan Area)

Intermediate Bldg (TDAFW Pump Area)

(1) LRM should be located in Containment Mid Floor near 'A" Recirc Fan Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E GiNNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-61 Containment Postaccident Air Sample Penetrations 305a, 305c, and 305d Pl%-23.50A Revision 1 REV 8 7/92

CONTAINMENTAIR SAMPLE ETURN PENETRATION 305b 10010 FROM POST ACCIDENT SAMPUNO SYSTEM OPEN PIPE TC/J PM5b ~y FROM 1599 15S8 RADIATION MONITORS 1599A NOTE DESCRIPTION Containment Mid Floor ('A'ecirc Fan Area)

Intermediate Bldg (TDAFWP Area)

(1) LRM should be located in Containment Mid Floor near "A'ecirc Fan, (2) Main Steam Header Adjacent to Containment Wall ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR PGMiER F'LANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-62 Containment Air Sample (Return)

Penetration 305b PTT-23.14 Revision 1 REV 8 7/92

CONTAINMENTAIR SAMPLE OUTLET PENETRATION 306e Il 10009 P0Sr/ACCInm 05 SAMPUNQ SYSTEM O FC C)

OPEN PIPE tc/a cy CIV

! 1596 1597 Cl 15M v/a NOTE DESCRIPTION Containment Mid Floor ("A" Recirc Fan Area)

Intermediate Bldg (TDAFW Pump Area)

(1) LRM should be located in Containment Mid Floor near "A'ecirc Fan.

(2) Located on Main Steam Header Floor ROCHESTER GAS AND ELECTRIC CORPORATION R. F- GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-63 Containment Air Sample Outlet Penetration 305e PTT-23.15 Revision 1 tv 8 7/92

FIRE SERVICE WATER PENIS'RATION 307 Ig TC/J 8!g m gQ KCC P507! ~y 9nS XQ 9231 ClY 3~ 9229 TC/J NOTE DESCRIPTION Containment Mid Floor ("A" Fan Area)

Intermediate Bldg (TDAFWP Area)

(1) LRM should be located in Containment Mid Floor "A" Fan Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E. 'GINNA NUCLEAR POWER PLANT UPDATED P!NAL SAFETY ANALYSIS REPORT Figure 6.2-64 Fire Service Nater Penetration 307 PTT-23.52 Revision 1 REV 8 7/92

SERVICE WATER FOR CONTAINMENT FAN COOLERS PENETRATIONS 308 311 312 315 316 319 320 323 fj (av) 4627 I

0 4626 5 4641

( 4642 P319 (090 4514 4631 45 I 5 4513 451 d I (CIV) 12500 H I 0 125006 (312)

I (as) 2142 Pl 2'136

.I 2156 2144 I

5s (as) 0 I gO I

2034 Q

I I

4522A 45226 4524A 45246 4592A 45926 4594A 4655 465d I 4659 4524 4660 4592 4594 P306 (0N) 311 4633 4630 315 ) 4634 tI 323 I 4641 4644 125020 12502R 12502T

~ 20!0 2011 2012 0 12502U 2013 ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEA'O'Po'PER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-65 Service Water for Containment Fan Coolers, Penetrations 308, 311, 312, 315, 316, 319, 320, and 323 Isolation from 33013-1250-3, Revision 7 REV 8 7/92

MINl PURGE SUPPLY PENETRATlON 309 LC FROLI IIIN 748O SUPPLY FAN TC/J CIV PSOQ v~n (BVND FLANCE ON INTERMEDIATE BUILDING ROOF ILRT VENT)

STANDBY 7481 CONNECTION NOTE DESCRIPTION Containment Mid Floor ("A" Recirc Fan Area)

Intermediate Bldg (TDAFW Pump Area)

(1) LRM should be located in Intremediate Bldg near TDAFW Pump.

(2) Located in Intermediate Bldg. above Steam Header (3) Open pipe with debris screen ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-66 Mini-Purge Supply Penetration 309 P1T-23A4 Revision 1 tv 8 7/92

INSTRVMENT AIR PENETRATION 310a Ig 14100 7045 g

Zg FC

~g P$ 50a CY MQS tHQ2 96 MQ5 1 5450 TC/J 5450A 5450S NOTE DESCRIPTION Containment Mid Floor ("A" Fan Area)

Intermediate Bldg (TDAFWP Area)

(1) LRM should be located in Containment Mid Floor 'A'an Area (2) N, Bottle connection point ROCHESTER GAS AND ELECTRIC CORPORATION R. F GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-67 Instrument Air Penetration 310a PTT-23.33 Revision 1 REV 8 7/92

SERVICE AIR PENETRATION 310b ll

= IS lal fJ Q

~CP V/J 4l~

>i~

X Q 7227 ~y P310b

~CO CV 714' 7225 7141 O~

TC/J TC/J NOTE DESCRIPTION Containment Mid Floor ("A Fan Area)

Intermediate Bldg (TDAFW Pump Area)

LRM should be located in Containment Mid Floor near 'A'an, and Intermediate Bldg near TDAFW Pump (2) Located approximately 10 ft above floor ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GiNNA NUCLEAR PONER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-68 Service Air Penetration 310b PTI -23.32 Revision 1 aZV 8 7/92

0 LEAKAGE TEST DEPRESSURIZATION PENETRATION 313 BLIND FLANGE V/J ClY 7478 ClY TC/J 7 NOTE DESCRIPTION Containment Mid Floor ('A'ecirc Fan Area)

Intermediate Bldg. (TDAFWP Area)

LRM should be located in Intermediate Bldg. near TDAFWP Intermediate Bldg. Roof adjacent to CNMT Dome platform Door 54, Cap removal/replace should be done when valve is positioned.

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-69 Leakage Test Depressurization Penetration 313 PTl -23.42 Revision 1 aVr 8 7/92

LEAKAGE TEST SUPPLY PENETRATION 317 Ig

. I.Q i5 744$

I FLANcE 8 I

I 7473 I

Cy P3$ 7 1

7475 Y/J NOTE DESCRIPTlON Containment Mid Floor ( A Recirc Fan Area)

Intermediate Bldg. (TDAFWP Area)

(1) LRM should be located in Intermediate Bldg. near TDAFWP ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-70 Leakage Test Supply Penetration 317 PTT-23.43 Revision 1 azv 8 7/92

"A" STEAM GENERATOR BLOWDOWN PENETRATlON 321 I

lm

<<! e g

g lm I

CW 28 e 570$

4 5705A V/J NOTE DESCRIPTION Containment Mid Floor (above "A" Accumulator)

Intermediate Bldg (TDAFWP Area)

LRM should be located in Intermediate Bldg (TDAFWP Area)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-71 Steam Generator A Blowdown Penetration 321 PTT-23.16A Revision 1 REV 8 7l92

"B" STEAM GENERATOR BLOWDOWN PENETRATION 322 I

II

~! aa x! Ig Q Id p822 5702 9h 5VO5 5768A v/~

I NOTE DESCRIPTION Containment Mid Floor (above 'A'ccumulator)

Intermediate Bldg (TOAFWP Area)

(1) LRM should be located in Intermediate Bldg (TDAFWP Area)

ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-72 Steam Generator B Blowdown Penetration 322 P IT-23.168 Revision 1 aVr 8 7/92

DEMINERALIZEDWATER PENETRATION 324 yB

%!g gh 502$ ~ cr TC/J v/z NL'OTE t

DESCRIPTION Containment Mid Floor ( A Fan Area)

Intermediate Bldg (TDAFWP Area)

LRM should be located in Containment Mid F!oor 'A'an Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-73 Demineralized Hater Penetration 324 PTT-23.39 Revision 1 aZV 8 7/92

CONTAINMENTH MONITORS PENETRATION 332a 332b and 332d I

X S

T C

CIY 921 7448 CIY CN TC/J T S

CN 922 P5$ 2a T

7452 924 CIY Y/J 0208 S

P532b 5 020A g I

8451 7456 CIY TC/J NOTE DESCRIPTION Containment Mid Floor ( A" Recirc Fan Area)

Intermediate Bldg (TDAFW Pump Area)

(1) LRM should be located in Intermediate Bldg TDAFW Pump Area ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6. 2-74 Containment H2 Monitors Penetrations 332a, 332b, and 332d PTT-23.45 ReViSion 1 azv 8 7/92

CONTAINMENT PRESSURE TRANSMITTERS PT-944 PT449 AND PT-950 PENETRATION 332c CIB CIB CIB Y/J V/J Y/J CIV CIV CIV 181M 1818F 1819E Hl-1818G 1818F 1818E OPEN PIPE NOTE DESCRIPTION Containment Mid Floor ('A" Recirc Fan Area)

Intermediate Bldg. (TDAFWP Area)

(1) LRM should be located in Containment Mid Floor near 'A'ecirc Fan ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLIAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-75 Containment Pressure Transmitters PT-944, PT-949, and PT-950 Penetration 332c Pal=23.17C Revision 1 aZV 8 7/92

MAIN STEAM FROM STEAM GENERATOR A PENETRATION 401 CB CB CB CB II'10K 11025 3409 C CV ~ CV 11020 (4 lYP) 5500 5511 CV 54IIB . 5515 11051 AllQSPHERE TO lEMPKRATlJRE CONF ENSATEQ SUPPORTS i5 I

8517 5510 CV 5

3411 TO AUXIUAlA'EEOWATER 3415C $ 413B ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-76 Hain Steam from Steam Generator A Penetration 401 Isolation from 33013-1 231, Revision 19 aZV 8 7/92

MAIN STEAM FROM STEAM GENERATOR B PENETRATION 402 l

TD AUXIUARZ I TEEDWATER (4 TTP]

Cruasua) 5505

] 551 0 551 2 I

IV ATQCSPHERE 5504A 5504C I

I 5504 551 d 5515 g lg 5514 CIV 1120 PS B

I 541 2C d721 11021 I

CIV 54455 I

I 5520 541 0 I 11024A 11022 I 11025 CIV I

PT D PT 11025 47B 470 CIV ad Cld D

ROCHESTER GAS AND ELECTRIC CORPORATION R. E GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figure 6.2-77 Main Steam from Steam Generator B Penetration 402 Isolation from 33013-1 231, Revision 19 REV 8 7/92

MAIN AND AUXILIARYFEEDWATER TO STEAM GENERATORS A AND B PENETRATIONS 403 AND 404 KNNI IN!ON OXIVKM AOXRWtf AZDVAlKN WMt 1A CIV I

m IMOM MAIN Pco FIZOWA1KN I IIIMI IA 8415C 0001 CIV CIV I lc 10 I

I I.

i5 IXOM lMKNMK OXIVKN AUXILIAXV IKXOMAllà tWt jib I

I I

I

~M MAIN IZGWAIKN 400K tVW I ~

CIV o

FKKAI QKOO ONVKM AXOIAINK IKXOMAIKN l%NP 1~

NOIX l0IKA CIV IXV ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCI'EAR POWER PL'ANT UPDATED FINAL SAFETY ANALYSIS REPORT

,Figure 6.2-78 Main and Auxiliary Feedwater to Steam Generators A and B Penetrations 403 and 404 Isolation from 3301 3-1236-2, Revision 5 and 33013-1 237, Revision 25 avr 8 7/92

ATTACHMENT F Table of Technical Specification Changes

Attachment P Page 1 of 3 Technical Specification Changes I Changes Effect Removed reference to Table No technical change.

3.6-1 from Technical Specifications are now Specifications 3.6.3.1, consistent with Generic 4.4.5.1, and 4.4.6.2. Added Letter 91-08.

statement to Bases for Technical Specification 3.6 that containment isolation boundaries are listed in UFSAR Table 6.2-15.

2. Removed Table 3.6-1 from Valve listing remains in a Technical Specifications and licensee controlled document placed information in UFSAR under 10 CFR 50.59 program.

Table 6.2-15.

3. Revised action statement of Specification now considers Technical Specification closed systems as an 3.6.3.1. acceptable interim passive boundary and is more consistent with Standard Technical Specifications.

4 ~ Removed definition of Definition is found in leakage inoperability from Technical Specification Technical Specification 4.4.2.2. Eliminated 3.6.3.1. redundant discussion of leakage acceptance criteria.

5. Added statement related to No technical change.

intermittent operation of Specification now consistent boundaries to both Technical with Generic letter 91-08.

Specification 3.6.1 and the bases.

6. Removed note associated with Mini-purge valves have been Technical Specification installed so specification 3.6.5. is considered effective. No technical change.

7. Added definition of No technical change.

"isolation boundary" to Clarification of "isolation Bases for Technical boundary" provides Specification 3.6. consistency with UFSAR Table 6.2-15.

8. Updated reference list No technical change.

contained in Bases for Technical Specifications 3.6, 3.8, and 4.4.

Attachment P Pago 2 of 3 Technical Specification Changes Changes Effect Revised action statement of Clarification only.

Technical Specification Specification now consistent 3.8.1 section a. with Standard Technical Specifications.

Revised action statement of No technical change.

Technical Specification Specification now 3.8.3. specifically addresses affected containment penetrations.

Revised bases for Technical No technical change. Bases Specification 3.8. are now consistent with Standard Technical Specifications and support changes to 3.8.1 section a and 3.8.3.

Added "Pt" and necessary Addition of "Pt" definition definitions to Technical provides clarification of Specification 4.4.1.4 testing type consistent with section a. 10 CFR 50, Appendix J. All terms in 4.4.1.4, section a are now fully defined. No technical change.

Added to the definition of Addition of "Lt" definition "Lt" in Technical provides clarification Specification 4.4.1.4 consistent with 10 CFR 50, section b. Appendix J. All terms in 4.4.1.4, section b are now fully defined. No technical change.

Added definition of "Pa" and Addition of "Pa" and "Lam" "Lam" to Technical provides clarification Specification 4.4.1.4. consistent with 10 CFR 50, Appendix J. All terms in 4.4.1.4 now fully defined.

No technical change.

Added steam generator Addition of this penetration inspection/maintenance provides, testing criteria penetration to Technical similar to the equipment Specification 4.4.1.5 hatch and containment air section a (ii). locks.

Revised first line of Minor clarification only.

Technical Specification No technical change.

4.4.1.5, section a (ii).

Attachment F Page 3 of 3 Technical Specification Changes Changes Effect

17. Revised acceptance criteria Clarification only. No provided in Technical technical change.

Specification 4.4.2.2

18. Replaced "isolation valve" Minor clarification only.

with "isolation boundary" in Specification and bases are Technical Specification now consistent with the 4.4.2.3 and the Bases for revised Technical section 4.4. Specification 3.6.3.

19. Removed notes associated Mini-purge valves have been with Technical Specification installed so specification 4.4.2.4 section a. Also, is considered effective.

deleted reference to section Section d will be removed

d. from Technical Specifications with this amendment.

20. Added steam generator Addition of this penetration inspection/maintenance provides testing criteria penetration to Technical similar to the equipment Specification 4.4.2.4 hatch and containment air section b. locks.

21. Removed Technical Blind flanges have been Specification 4.4.2.4 installed so specification section d and associated is considered effective. No note. technical change.

22. Revised statement for Specification now consistent Technical Specification with Standard Technical 4.4.5.1. Specifications.

23. Revised statement for Specification now consistent Technical Specification with Standard Technical 4.4.6.2. Specifications.

3.6 Containment S stem A licabilit Applies to the integrity of reactor containment.

To define the operating status of the reactor containment for plant operation.

S ecification:

3.6.1 Containment Inte rit a ~ Except as allowed by 3.6.3, containment integrity shall not be violated unless the reactor is in the cold shutdown condition.g>'j:;:'.;jCX'osedpp;;.',VO3VpSpjNSyjggi8 op" 'neddy";:,":);.,:.oxi,'::,,'.:,:;";,~";

o'cia;::n:x'8~57!AOLv8'$,",'con'cx!0~3'~

b. The containment integrity shall not be violated when the reactor vessel head is removed unless the boron concentration is greater than 2000 ppm.

c ~ Positive reactivity changes shall not be made by rod drive motion or boron dilution whenever the containment integrity is not intact unless the boron concentration is greater than 2000 ppm.

3.6.2 Internal Pressure If the internal pressure exceeds 1 psig or the internal vacuum exceeds 2.0 psig, the condition shall be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the reactor rendered subcritical.

Amendment No. 3.6-1 Proposed

3. 6. 3 Containment Isolation-VakveeF~SoQ da~kes 3.6.3.1 With epee-aao~aj'oontai::one'nc,',,:.i~'~olati~oP jbon'nclsgf::;:.,~iinaperab1l'eFf OsniytOraa'Or'e'i'!O'On:,,':a'; 'nJa'e'nr '"lip'a'n'etr'i" O'ne'j

': 'it er i ox>

a~ Restore tba'ilia'a'Ob inoperable ,';:.':.;b'ygjd~y: to eperahke~OPEHABG'8 status within 4 hours, ox

b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation~osition, ihO+j'iici'i8 laanxail~v8'ivy!,:0'iixii,a!.;:".blin'R!'i!gianqsg or c ~

4:-large ,"::::;Verif j,i: tb,pg opaeratigl4t y~;,:,:::! oi:::.::':; o'o'sel::,':;"syipem a

"':::"-":-"::::':-i"'::i''i:a::--.'-:::::::ll::::::,:-::-"li5'::::: '

-" '-'"-; "'-:'::i -Jt"::-- "e:-*":::::is

d. Be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.4 Combustible Gas Control 3.6.4.1 When the reactor is critical, at, least two independent containment hydrogen monitors shall be operable. One of the monitors may be the Post Accident Sampling System.

3. 6.4. 2 With only one hydrogen monitor operable, restore a second monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3. 6.4. 3 With no hydrogen monitors operable, restore at least one monitor to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.6s5 Containment Mini-Pur e Whenever the containment integrity is required, emphasis will be placed on limiting all purging and venting times to as low as achievable. The mini-purge isolation valves will remain closed to the maximum extent practicable but may be open for pressure control, for ALARA, for respirable air quality considerations for personnel entry, for surveillance tests that may require the valve to be open or other safety related reasons.

Amendment No. P,gP 3. 6-2 Proposed

Amendment No- 9 r P 3. 6-2 Proposed

,+, I

~ 1 0 lt'>o ~li r

Basis:

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.

The shutdown margins are selected based on the type of activities that are being carried out. The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances. When the reactor head is not to be removed, a cold shutdown margin of 14~k/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig. " The containment is designed to withstand an internal vacuum of 2.5 psig. The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.

'.4: W~iii(!,,47,:

Amendment No. 3. 6-3 Proposed

References:

(1) Westinghouse Analysis, "Report for the BAST Concentration Reduction for R. E. Ginna", August 1985p)gpQRi)i5!%pep'phyla:.

(2) UFSAR Section 6.2.1.4 L(:8';)::;;:p~GPSA'Ri:i:.Tibia,p6','::::2.'.:,'-:.:15 3.6-3a Proposed

3.8 REFUELING A licabilit Applies to operating limitations during refueling operations.

To ensure that no incident could occur during refueling operations that would affect public health and safety S ecification 3.8. 1 During refueling operations the following conditions shall be satisfied.

a~

~ ~

z ~~~+,.~~~+~~x+t~wyn g+~~vox+w~<< ..g~fl~gxccRwc~~~+Pg~j~>l~

st'at~ue::.:

jL~~g;,::,',<~iTh'eppgixiymei5~ha5ch::::;aliall~!bsefsi,'n~plaei:::;:::yah:;::i8,::::

,illgl APW';,,if:,:::,"-::,'--:.,:::, '--'-:, Bit,,-"::,i4':::.,4h ii y~@! i':- -::"::":

m'eg~y&RPoxc E~>&'h~+xoPxco~+. xc ~t~,. NhccQ~ >~~ 'ttt'too&5PY.,~Y+Ao?Po&x'.

y 0$ ~~~7dw+'~PPx

/ha', y!i qiita4'nmianKi'..;::::;::ytmo'+phagia-;:,it ii;"'j~eg~,:,ou>sjd~

~~~@o~%~hp~k~~sh~aL~~be ~e"~~K T"-" 'Wi'CLCi886~hj":::::~"'S~n"": X~5i'8tiSIY" "v8;3;vb"': ':. 811%8 "tl: ~i:-::::::::::::: ;:::::;,:. 1!",,-::;::,:,::;:::,

Amendment No. g, g.g 3.8-1 Proposed

-"IK,::::,:;--,:::::,::,::, -~.,:,: 1,,-",;:"-;,:::,;, F,-;:::,,,:,,: Qf Iigainagi~a: '-.:Ship3owri: :;!'.:W,,:iqe ':::ai~RiKi.;::;:::1%iqa

~i:lyg;.

b. Radiation levels in the containment shall be monitored continuously.

c ~ Core subcritical neutron flux shall be continuously monitored by at least two source range neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment and control room available whenever core geometry is being changed. When core geometry is not being changed at Amendment No. g,gg Proposed

flange. If this condition is not met, all operations involving movement of fuel or control rods in the reactor vessel shall be suspended.

3.8.2 If any of the specified limiting conditions for refueling is not met, refueling of the reactor shall cease; work shall be initiated to correct the violated conditions so that the specified limits are met; no operations which may increase the reactivity of the core shall be made.

3.8.3 If the conditions of 3.8.1.d are not met, then in addition to the requirements of 3.8.2, Q. 8'oil'ilt'Q thaF5hnt8~ova,';.3?uz cga':-:;ka pcti!MipiPPutqa:,:':,:pa~pa: i,'at f nna within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Basis:

The equipment and general procedures to be utilized during refueling are discussed in the gFSAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating'uilt-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard 3.8-3 Proposed

F 4 I ~

provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time. Xn addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over stored racks containing spent fuel.

The operability requirements for residual heat removal loops will ensure adequate heat removal while in the refueling mode. The requirement for 23 feet of water above the reactor vessel flange while handling fuel and fuel components in containment is consistent with the assumptions of the fuel handling accident analysis.

The analysis~~~~ for a fuel handling accident inside containment establishes acceptable offsite limiting doses following rupture of all rods of an assembly operated at peak power. No credit is taken for containment isolation or effluent filtration prior to release.

Requiring closure of penetrations

h sVS e"-ma!~i!i '- -"-i!'4. - -:i -e e--- "Eiii t --0 "',--i!lit '!4h" OVCSXde"",:",:.age'O'Sgh8de establishes additional margin for the fuel handling accident and establishes a seismic envelope to protect against seismic events during refueling. P~:,::j::;1ipgei8+~ai:;<@: <!~i:, ese gurge:::~~vilvelaiu&lx%pl~iiiger,:;::~%vcsoFataciÃ>valvi.: ti;:ijul,.; opHRKiiE sws'"":,:::::,::::.":::,::::::.::: '"e'eii! i !w,::::::.-,"t,::j(-,:)i!i:,'::,. -,:-(::::::::: i:::::-.'::::;*, pà autcma&icgll~y!:::::i's'alated~<ib jRLg':,o:.:..L;~818:,.,:~ %Pen'et:

px',ovine::: 4'i~ii,,it;:,~~jjePP li::::,:":ii,.::-,-,-",:-- -:::;*:::;::;, '-",,*,:y,,::,:,:"'l,e::;::.;:-;:-:~i'!,,:---;::-:-::::::,,-,-,,::,-,-,g:-,,::.':e pre s arses,::;:s~fgggpgh'mendment No. g 3.8-5 Proposed

~ ~

I 1

References (1)  :::,-l,'.l:0::,a.g,i!!I tli:?",'d!::::,:::::0,',::: t,-::!:::::::::::~4!%!"It:::::,:-4):

(2) Reload Transient Safety Report, Cycle 14 (3) . ~UPBEAR':;"..seehge$ g;9;:5~F7~!:9';:Pp 3.8-6 Proposed

b. The local leakage rate shall be measured for each of the following components:

Containment penetrations that employ resilient seals, gaskets, or sealant compounds, piping penetrations with expansion bellows and electrical penetrations with flexible metal seal assemblies.

lie Air lock and equipment door seals.

~ ~

3.i3. Fuel transfer tube.

3.V e Isolation valves on the testable fluid systems lines penetrating the containment.

v Other containment components, which require leak repair in order to meet the acceptance criterion for any integrated leakage rate test.

4.4.2.2 Acce tance Criterion

'c t' ',"""'1F-  %" 'd are Q!'.::'inoperable,::! if'r'osis.",:;,:::::;:,:,:a:&silage~>i";ihiiiidp:',ax'nt A'sheii<",

s,-B-,e',-:, "Wl:, lt,-:.;::"-%!e::, -:i~R::".:i;::::-

ga5a~X'j':,:Xeak8gePaf~3a1l~lhoun'da'x'::.,le'smm3':s'(q'red~'4 4.4.2s3 Corrective Action If at any time it is determined that. the total leakage from all penetrations and isolation vaBree hcnTTdarieg exceeds 0.60 La, repairs shall be initiated immediately.

4.4-6 Proposed

b. If repairs are not completed and conformance to the acceptance criterion of 4.4.2.2 is not demonstrated within 48 hours, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion.

c ~ If it is determined that the leakage through a mini-purge supply and exhaust line is greater than 0.05 La an engineering evaluation shall be performed and plans for corrective action developed.

4.4.2.4 Test Fre uenc a~ below, individual penetrations and containment isolation valves shall be tested during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years.~

b. The containment equipment hatch, fuel transfer tube~xgtseemi!iy,,:gene'xgfoi!:: :ji::':;,-:,in-'i,,p'eigkci'irma'i,:ii'tiananPe pen'eWrpFion, and shutdown purge system flanges shall be tested at each refueling shutdown or after each use, if that be sooner.

Amendment No. 4.4-7 Proposed

II ,~" ~ 4~

c ~ The containment air locks shall be tested at intervals of no more than six months by pressurizing the space between the air lock doors. In addition, following opening of the air lock door during the interval, a test shall be performed by pressurizing between the dual seals of each door opened, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the opening, unless the reactor was in the cold shutdown condition at the time of the opening or has been subsequently brought to the cold shutdown condition. A test shall also be performed by pressurizing between the dual seals of each door within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of leaving the cold shutdown condition, unless the doors have not been open since the last test performed either by pressurizing the space between the air lock doors or by pressurizing between the dual door seals.

Amendment, No. g,P 4.4-8 Proposed

Amendment No. gP 4.4-8 Proposed J M

the tendon containing 6 broken wires) shall be inspected.

The accepted criterion then shall be no more than 4 broken wires in any of the additional 4 tendons. If this criterion is not satisfied, all of the tendons shall be inspected and if more than 54 of the total wires are broken, the reactor shall be shut down and depressurized.

4.4.4.2 Pre-Stress Confirmation Test a Lift-off tests shall be performed on the 14 tendons

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identified in 4.4. 4. la above, at the n e r v a s specified in 4.4.4.1b. If the average stress in the i t l 14 tendons checked is less than 144,000 psi (604 of ultimate stress), all tendons shall be checked for stress and retensioned, of 144,000 psi.

if necessary, to a stress

b. Before reseating a tendon, additional stress (64) shall be imposed to verify the ability of tendon to sustain the added stress applied during t h e accident conditions.

4.4.5 Containment Isolation Valves 4.4.5.1 Each comma'a%ment')isolation valve

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with the Ginna Station Pump an8 Valve Test ',,t!mme'ccordance program submitted in accordance with 10 CFR 50.55a.

4.4.6 Containment Isolation Res onse 4.4.6. 1 Each containment isolation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.1-1.

4.4.6.2 The ILespons'e,::.'-.,:,time

' ' of "'achN-the containment isolation valve , , shall be demonstrated to be within 4heg4'Cs limit at least once per 18 months. The response time includes only the valve Amendment. No. P,gg Proposed

sar.

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The Specification also allows for possible deterioration of the leakage rate b'etween tests, by requiring that the total measured leakage rate be only 754 of the maximum allowable leakage rate.

The duration and methods for the integrated leakage rate test established by ANSI N45.4-1972 provide a minimum level of accuracy and allow for daily cyclic variation in temperature and thermal radiation. The frequency of the integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best, be performed during refueling shutdowns. Refueling shutdowns are scheduled at approximately one year intervals.

The specified frequency of integrated leakage rate tests is based on three major considerations. First is the low probability of leaks in the liner, because of (a) the use of weld channels to test the leaktightness of the welds during erection, (b) conformance of the complete containment to a 0.14 per day leak rate at 60 psig during preoperational testing, and (c) absence of any significant stresses in the liner during reactor operation. Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60 La) of the total leakage that is specified as acceptable Third is the tendon stress surveillance program, which provides assurance than an important part of the structural integrity of the containment is maintained.

4.4-13 Proposed

The basis for specification of a total leakage of 0.60 La from penetrations and isolation vakvee<5'ogA~iiieg is that only a portion of the allowable integrated leakage rate should be from those sources in order to provide assurance that the integrated leakage rate would remain within the specified limits during the intervals between integrated leakage rate tests. Because most leakage during an integrated leak rate test occurs though penetrations and isolation valves, and because for most penetrations and isolation valves a smaller leakage rate would result from an integrated leak test than from a local test, adequate assurance of maintaining the integrated leakage rate within the specified limits is provided.

The limiting leakage rates from the Recirculation Heat Removal Systems are judgement values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test 4.4-14 Proposed

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The pre-stress confirmation test. provides a direct measure of the load-carrying capability of the tendon.

If the surveillance program indicates by extensive wire breakage or tendon stress relation that the pre-stressing tendons are not behaving as expected, the situation will be evaluated immediately.

The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible. Thus the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down the reactor. The containment is provided with two readily removable tendons that might be useful to such a study. In addition, there are 40 tendons, each containing a removable wire which will be used to monitor for possible corrosion effects.

Operability of the containment isolation vakveeKbogq'4yx'if'nsures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

Performance of c cling tests and verification of isolation times Appendix J to 10 CFR 50 is addressed under local leak testing requirements.

References:

(2)

(3)

(4)

(5)

(6) FSAR Page 5.1.2-28 (7) North-American-Rockwell Report 550-x-32, Autonetics Reliability Handbook, February 1963.

(8) FSAR Page 5.1-28 4.4-17 Proposed