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

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

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

S ecification:

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.c~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 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 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.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.

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

a~b.C~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.Radiation levels in the containment shall be monitored continuously.

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 3.8.2 3.8.3 flange.If this condition is not met, all operations involving movement of fuel or control rods in the reactor vessel shall be suspended.

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.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 e r g ate 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~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.b.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 lvo vo Fuel transfer tube.Isolation valves on the testable fluid systems lines penetrating the containment.

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.c~If repairs are not completed and conformance to the acceptance criterion of 4.4.2.2 is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion.

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~b.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.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 4.4.4.2 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.

Pre-Stress Confirmation Test a~Lift-off tests 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, if necessary, to a stress of 144,000 psi.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 4.4.5.1 Containment Isolation Valves 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 4.4.6.1 4.4.6.2 Containment Isolation Res onse 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.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, Reliability Handbook, February 1963.(8)FSAR Page 5.1-28 Autonetics 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: 2.3.involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident previously evaluated; or 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 II (2)(3)(4)AOV 959 for penetration 111 provides a containment isolation 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).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.

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(~)(6)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 it when containment integrity is required.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)(9)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.

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 would cost approximately

$15,000 to install the necessary vents.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 valve i s and should (not J be used to describe valves.If there is any clarification to be noted because valves are listed which are not cl assi fi ed as containment i sol ation val ves, it 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 contai nment i 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 it was consistent with the Appendix J testing program.Most 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 penetration with i ts associated val ves or boundary as a separate entry.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'I'I~-4A 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 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.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'i ed 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 P-109 Added 869A and 2856 as CIV's (pressure indicator root valves are now closed).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 P-209b No changes were made.See response to Question 23.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 P-308 No changes were made.See response to Question 26.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 CIB.Water 12500H (drain line)as a CIV and PI-2136 as a No root valves were added since the Service system is a CLIC.P-323Added-valves a CLIC FIA-2036 and TIA-2013 as CIBs.No root were added since the Service Water system is 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 li sti 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 6.2-22 6.2-30 6.2-56 6.2-73 864A, 2825, 2829 864B, 2826, 2830 497, 498, 567, 576 8049 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 P-109 P-121a Added 864A and 2829 as CIVs.Valve 2825 is a test connection used for Appendix J testing.Added 864B and 2830 as CIVs.Valve 2826 is a test connection used for Appendix J testing.'I 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 P-332a-d No changes were made.See response to Question 26.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 it ensures that the Residual Heat Removal (RHR)system is a closed system on a CIS and should be listed in Table 6.2-13.Also, the valves to the safety in j 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.If an exception is taken to this position, it 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 20.Footnote 14 states that valve 745 for enet (return)is to be manuall closed un manua y closed until it is modified to u orna ic closure si nal.h h d h h w y as it not been implemented?

RESPONSE: AOV 745 was t outage as stated in a l tt f to be modified by the end of the Johnson, NRC, dated July 9, 1990.Howe benefit analysis, and the fact that n this penetration is required based on t modification was canceled.0 erat'p mons zs stz.ll instruct d to o owing a CIS for additional redundancy.

See 91 letter from R.Mecred RGGE y as mo i ied to reflect this change.21.Please rovi de p'our 50.59 eval uati on and the change in the classification for v f 42, valves 1813A and 1813B should be exception i t k to thi n o is position, it should be justified.

RESPONSE: The 50.5 9 evaluation to remove valves 8 bl'tl d d'th th o'l Am d equest dated October 15, delete.on is contained in th A).However, the basis fo Max.erg RGB(EI to DE Crutchfield, NRC in in e August 30" 1982 SER SEP T'I 4 S'h f l NUREG-08'c-(see NRC letter posi.txon expressed in our A t 30, 21 both references and reflects the that the NRC had agreed that 851A and 8 our ugust 30, 1982 lette r, RGGE assumes 1813A d 1 to include the necessary"CIV" d an 1813B, the UFSAR fi ure ha IV designation.

22.Valve 1722 should be listed in Table 6.2-13 an be a 1ocked closed valve.If not provided.va ve.If not, a justification should be RESPONSE: Valve 1722 is a locked closed valve.been updated to reflect this.The UFSAR figure has l,

Attachment D Page ll of 1223.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 list each of the three penetrations individually.

A separate UFSAR 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 6.2-13 through 6.2-78 r

GINNA/UFSAR SI APERTURE CARD Also Availab1e Oo Aperture Card Table 6.2-15 CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING e~Sstem Penetration No valve/a~aaada Isolation Position'alve

~Te valve Operator 1~e Position Indication Zn Control Room Position Relative to Containment Normal asereadaa Position At Cold Shutdown Immediate Postaccident~

Power Pailure Trip on CZS Maximum Isolation Time~eea'P SAR~Pl re Class~Notes see end of table'team generator inspection/

maintenance Puel transfer tube charging line to B loop safety injection pump 1B dischazge 29 100 101 370B CLOC 870B 889B CLOC 12407 PZ-923A'PT-923'85B al a2 Blind flange Blind flange al a2 al al a2 bl bl bl b2 check NA Check check HA Globe NA NA Globe al, a2 Blind flange NA HA NA NA NA NA HA Manual NA HA Manual NA HA NA HA NA NA HA NA No NA NA No Inside Outside Inside Inside outside outside Outside Outside outside outside Outside Outside C C C 0 C C C C C NA NA a o/c o/c o/c C C C C C C NA NA 0 C C C C 0 0 C C NA NA 0 HA NA NA NA HA NA HA NA NA NA NA HA HA NA HA NA HA NA HA NA NA NA NA NA NA NA ,NA NA NA HA 6.2-13 6.2-13 6 2-13 6.2-14 6.2-14 6 2-15 6 2-15 6.2-15 6.2-15 6.2-15 6.2-15 6.2-15 3B 3B 3B 3B 3B 3B 3B 3B 3B 1, 2 1, 2 2I 3 Alternate charging to A cold leg construction fire service water containment spray pump 1A 102 103 105 383B CLOC NA 5129 862A CLOC 2829 Cap 869A 2856 2825 2825A 864A 859A 859B 859C al a2 al a2 al a2 bl b2 cl c2 dl d2 el e2 e2 e2 Check NA welded cap Gate check NA Globe NA Globe Globe Globe Ball Globe Globe Globe Globe NA NA HA Manual NA NA Manual NA Manual Manual Manual Manual Manual Manual Manual Manual HA NA NA No HA HA NA NA Ho No No No No No No No Inside Outside Inside Outside outside outside outside Outside outside Outside outside outside outside outside Outside outside C C C LC C C LC C C C LC C C 0/C o/c C C LC C C LC LC C C C 0 C LC C C C LC C C LC LC C NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA HA NA NA HA NA NA NA HA NA NA NA NA NA NA HA NA NA HA NA NA HA HA NA NA NA NA NA NA'A NA NA NA 6.2-16 6 2-16 6.2-17 6.2-17 6.2-18 6.2-18 6.2-18 6.2-18 6.2-18 6.2-18 6.2-18 6 2-18 6.2-18 6.2-18, 6 2-18 6.2-18 3B 3B 3Br 3B 3B 3B 3B 3B 3B 3B 3B 3B 3B 3B 9 10 10 9, 10 Reactor coolant pump A seal water inlet 106 304A CLOC al a2 check NA NA HA HA NA Inside outside 0 C 0 C C C HA NA HA NA NA NA r 6.2-19 3B 6.2-19 3B sump A discharge to waste holdup tank 107 1723 1728 al a2 Diaphragm Air Diaphzagm Air status Status Outside Outside 0 0 o/c o/c C C PC PC Yes Yes 60 60 6.2-20 2 6.2-20 2 Le end AI Aov BLC BLO Both C CIB ss CZS, T CIV CLIC Pails as is Air-operated valve Breaker locked closed Breaker locked open R/G and Status closed containment isolation boundary/bazrier containment

'solation signal containment isolation valve closed loop'nside containment CLOC CV D PC Po I YB J LC rov MV Closed loop outside containment check valve Drain Pails closed Pails open Inside missile barrier Appendix J connection Locked closed Motor-opezated valve Manual valve 0 0/C OMB R/G S Sov Status TC V Open Open or closed outside missile barrier Red/green light on main control board safety injection signal Solenoid-operated valve white status light Test connection Vent 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)

~satan Penetration No.Valve/Isolation Baunda~Position" Valve 2ree Valve Operator~T8 Position Zndication In Control Room Position Relative to Containment Position At Immediate Postaccident~

s Hormal.Cold~ezarXon snctdoen Power Pailure Trip on cts Maximum Isolation Time~eec'PSAR~Pi ure Notes Classd See end of table'eactor coolant pump seal water return line and excess letdown to VCT Containment spray pump 1B 108 109 313 CLOC 862B CLOC 2830 Cap 869B 2858 2826 2826A 864B 859A 859B 859C al a2 al a2 bl b2 cl c2 dl d2 el e2 e2 e2 Gate HA Check HA Globe NA Globe Globe Globe Ball Globe Globe Globe Globe Motor NA NA NA Manual Na Manual Manual Manual Manual Manual Manual Manual Manual Both NA NA NA No NA No No No No No No No No outside outside outside outside outside outside outside outside outside outside outside Outside outside outside I 0 C C C LC C C C LC C C LC LC C o/c C C C o/c o/c C C LC C C LC LC C C C 0 C LC C C C LC C C LC LC C AI NA NA NA NA NA NA NA NA NA NA NA NA NA Yes NA NA NA NA NA NA NA HA NA NA NA NA NA 60 HA NA NA NA;HA NA NA'A'A'A , NA t NA NA 6.2-21 6 2-21 6.2-22 6.2-22 6.2-22 6.2-22 6 2-22 6.2-22 6 2-22 6.2-22 6.2-22 6.2-22 6.2-22 6.2-22 3B 3B 3B 3B 3B 3B 3B 3B 3B 3B 3B 3B 9 10 10 9, 10 Reactor coolant, pump B seal water inlet Safety injection test line 110a 110b 304B CLOC 879 al, a2 Check NA Globe NA NA Manual NA NA No Inside outside outside 0 C 0 C C C NA HA NA NA NA ,'A NA NA 6.2-23 3B 6.2-23 3B 6.2-15 1 12 Residual heat removal to B cold leg 720 959 CLOC al a2 a2 Gate Globe NA Motor Air NA R/G Status NA Inside outside HA C C C 0 o/c C C C C AI PC NA No Yes NA NA NA NA 6.2-24 3B 6 2-24 3B 6 2-24 3B 13, 14 15 Letdown to nonregenerative heat exchanger Safety injection pump~M discharge Standby auxil-iary feedwater line to steam generator 1A 112 113 119 200A 200B 202 371 427 870A 889A CLOC 12406 PI-922A PT-922 cap 885A 9704A 9723 CLIC al al al a2 NA al al a2 bl bl bl bl b2 al al a2 Globe Globe Globe Globe Globe Check check HA Globe HA HA NA Globe Stop-check Globe NA Air Air Air Air NA NA NA Manual NA HA NA Manual Motor Manual NA R/G R/G R/G Both R/G NA NA NA No NA NA NA No R/G No NA Inside Inside Znside outside Inside outside Outside Outside outside Outside Outside outside Outside Outside Outside Inside o/c o/c C 0 0 c C C C NA NA C 0 0 LC C C C C 0 o/c C C C C HA NA C 0 0 LC C C C C C C 0 0 C C NA NA C 0 0 LC C Ec Ec Pc PC Po NA NA, NA NA NA NA NA NA AI NA NA Yes Yes Yes Yes Yes NA NA NA NA NA NA NA NA No NA NA 60 60 60 60 NA NA NA NA NA NA NA NA NA NA HA HA 6.2-25 6.2-25 6.2-25 6.2-25 6.2-25 6.2-15 6.2-15 6.2-15 6.2-15 6.2-15 6e2-15 6.2-15 6 2-15 6.2-26 6.2-26 6.2-26 1 1 1 1 1 3B 3B 3B 3B 3B 3B 3B 3B 16 16 16 17 18 19 Nitrogen to accumulators Pressurizer relief tank to gas analyzer 120a 120b 846 8623 539 546 al a2 al a2 Globe check Globe Globe Air HA Air Manual Both NA Status No Outside Inside outside Outside C o/c C 0 o/c o/c o/c 0 C C C 0 Fc NA PC NA Yes NA Yes HA 60 HA 1 60 NA 6'-27 3A 6.2-27 3A 6 2-28 2 6.2-28 2 6.2-97 REV 8 7/92~

3'-)V~i~F I s~)

GINNA/UFSAR

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

~Sstem Penetration No.Valve/Isolation~Baunda poededon'alve

~Te Valve Operator e Position Zndication zn control Room Position Relative to Containment Normal ooeeandon Position At cold zmmediate Shutdovn Postaccident~

Pover Failure Trip on odd Maximum Isolation Time~deo'FSAR~Fi re Class~Notes see end of table'akeup water to pressurizer relief tank 121a 508 al Diaphragm 529 a2 check Air NA Both NA outside Znside C o/c o/c o/c FC NA Yes NA 60 6 2-29 3A 6.2-29 3A Nitrogen to pressurizer relief tank 121b 528 547 al a2 check Globe NA Manual NA No Inside , c Outside~LC 0/C o/c NA NA NA 6.2-30 6.2-30 3A 3A 20 Containment pressure transmitters PT945 and PT946 Reactor coolant drain tank to gas analyzer line Standby auxil-iary feedvater line to steam generator 1B Excess letdown heat exchanger cooling water supply Postaccident air sample to common return Excess letdown heat exchanger cooling water return Postaccident air sanple to c fan Component cooling water from reactor coolant pump 1B 121c 123a 123b 124a 124b 124c 124d 125 PT945 1819A PT946 1819B 1600A 1655 1789 9704B 9725 9724 CLIC 743 CLIC 1572 1573 1574 745 CLZC 1569 1570 1571 759B CLOC al a2 bl b2 NA al a2 al al al a2 al a2 al a2 a2 al a2 al a2 a2 al a2 HA Globe NA Globe Globe Globe Diaphragm Stop-check Globe Globe NA Check NA Diaphragm Globe Diaphragm Globe NA Diaphragm Globe Diaphragm Gate NA NA Manual NA Hanual solenoid Manual Air Motor Manual Manual NA Manual Hanual Manual Hanual Manual Manual Motor NA NA No NA No No No Status R/G No No NA NA NA No No No R/G NA No No No R/G NA Outside outside Outsi.de Outside Outside Outside outside Outside Outside Outside Znside Inside Inside Outside Outside Outside Outside Inside outside Outside outside outside Outside NA 0 NA 0 0 0 0 0 C C C C C C LC NA 0 HA 0 o/c 0 o/c 0 C C C C C NA 0 NA 0 C 0 C 0 C C C C C NA HA NA NA PC NA Fc AZ NA NA NA NA NA NA Fc NA NA NA NA AI NA NA HA NA NA Yes HA Yes No NA NA NA NA NA NA NA NA No NA No HA NA NA NA NA NA 60 NA NA NA NA NA NA NA NA NA NA'A NA , HA 6.2-31 6.2-31 6.2-31 6.2-31 6.2-32 6.2-32 6.2-32 6.2-26 6.2-26 6.2-26 6.2-26 6.2-33 6.2-33 6.2-34 6.2-34 6.2-34 6.2-33 6'-33 6.2-34 6'-34 6.2-34 6.2-35 6.2-35 5 5 5 17 18 19 22 21 22 23 Component, cooling water from reactor coolant pump 1A 126 759A CLOC al a2 Gate Motor NA'A R/G HA Outside Outside AI NA Ho NA NA NA 6.2-36 2 6'-36 2 23 Component cooling vater to reactor coolant pump 1A 127 749A 750A al a2 Gate Check Motor NA R/G NA Outside Inside AI NA No NA NA NA 6.2-37 3B 6.2-37 3B Component cooling vater to reactor coolant pump 1B 128 749B 750B al a2 Gate Check Hotor NA R/G HA Outside Inside AZ NA No NA NA 6.2-38 3B NA , 6.2-38 3B 6.2-99 REV 8 7/92 f ye p h i GINNA/UFSAR a"r~Sstem Penetration Valve/Isolation No.~Bounda roelrlon'alve T9(pe Valve operator~pe Position Indication Zn Control Room Position Relative to Normal contadntant

~erat1an SI APERTURE CARD Position At Cold shutdown Immediate Postaccident~

Power rallnre Trip on crc A1SO AVai1abIe On Aperture Card Maximum Isolation Time~ce Table 6.2-15 UPSAR Ei cCure class'otes See end oi table'ONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

Reactor coola~nt~99 1713 drain tank and 1799 pressurizer 17Q6 relief tank to 1787 containment vent header al a2 bl b2 check Diaphragm Diaphragm Diaphragm NA Manual Air Air NA No status Status outside outside Outside outside C LC 0 0 o/c o/c C C C LC C C NA HA Ec PC HA NA Yes Yes NA NA 60 60 6.2-39 6'-39 6.2-39 6.2-39 3A 3A 3A 3A 20 component cooling water from reactor support cooling component cooling water to reactor support cooling 130 131 814 CLIO 813 CLIC al a2 al a2 Gate NA Gate NA Motor NA Motor NA Both NA Both NA outside Inside Outside Inside 0 0 C C AI NA AZ NA Yes NA Yes NA 60 NA 60 NA 6.2-40 6.2-40 6.2-40 6.2-40 22 22 containment mini-purge exhaust Residual heat removal pump suction from A hot leg Residual heat;removal pump A suction from sump B 132 140 141 7970 7971 Cap 701 2763 2786 850A 1813A al a2 a2 al al al a2 al a2 bl, b2 Butterfly Butterfly NA Gate Globe Globe HA Gate NA Gate Air Air NA Motor Manual Manual NA Motor NA Motor Both Both NA R/G No Ho NA R/G NA R/G Inside outside Outside Inside Inside Inside outside Outside outside Outside o/c C C C C C C C C C o/c o/c C 0 C C C C C o/c C C C Pc Ec NA AZ NA NA NA Yes Yes NA Ho NA NA NA No NA Ho 3 3 NA NA NA NA NA NA NA NA 6.2-41 6.2-41 6.2-41 6.2-42 6.2-42 6.2-42 6.2-42 6.2-43 6.2-43 6.2-43 13, 14 15 24 15 14, 25 Residual heat, removal pump B suction from sump B 142 850B al CLOC a2 1813B bl, b2 Gate NA Gate Motor NA Motor'/G'NA R/G outside outside Outside C C o/c 0 C C AI NA AZ No NA'Ho NA NA NA 6.2-44 6.2-44 6.2-44 24 15 14, 25 Reactor coolant drain tank discharge line Reactor compartment cooling unit A Reactor compartment cooling unit B return B hydrogen recombiner (pilot)B hydrogen recombiner (main)Containment pressure transmitter PT947 and PT948 143 201a 201b 202a 202b 203a 1003A 1003B 1709G 1722 1721 4757 4775 CLIC 4636 4776 PI-2141 CLIC 1076B 10211%1.1084b 1021381 PT947 1819C PT948 1819D al al al al a2 al al a2 al al al a2 al a2 al a2 al a2 bl b2 Diaphragm Diaphragm Gate Diaphragm Diaphragm-Butterfly Gate HA Butterfly Gate NA HA Diaphragm Globe Diaphragm Globe NA Globe NA Globe Air Air Manual Manual Air Manual Manual NA Manual Manual NA NA Manual solenoid Manual solenoid'NA Manual NA Manual status status No No Status No No NA No No NA HA No Status No Status NA No NA No outside outside Outside Outside Outside Outside Outside Inside Outside outside outside Inside Outside Outside outside Outside outside Outside Ou side outside 0 0 C LC 0 0 C C 0 C HA C NA 0 NA 0 o/c o/c C 0 0 C C 0 C NA C NA 0 NA 0 C C C LC C 0 C C 0 C NA C NA 0 NA 0 PC Ec NA NA Fc NA NA NA NA NA NA NA HA Ec NA PC NA NA NA NA Yes Yes NA NA Yes HA NA NA NA NA NA NA NA Yes NA Yes NA NA NA NA 60 60 NA HA 60 NA NA NA NA NA NA NA NA 3 NA 3 NA NA NA NA 6 2-45 6.2-45 6.2-45 6.2-45 6.2-45 6.2-46 6.2-46 6.2-46 6.2-47 6.2-47 6.2-47 6.2-47 6.2-48 6'-48 6.2-48 5 6.2-48 5 6.2-49 2 6.2-49 2 6.2-49 2 6.2-49 2 26 27 30 27 28 28 6.2-101 REV 8 7/92 92j.mj40i59 I C t P l' GINNA/UFSAR SI ApERTURE CARD Table 6.2-15 CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

~sstem Postaccident air sample from D fan Postaccident air sample from common header Penetration No 203b 203c 1563 1564 1565 1566 1567 1568 al a2 a2 al a2 a2 Valve/Isolation Boundary poattton Valve~e Diaphragm Globe Diaphragm Diaphragm Globe Diaphragm Valve operator T~e Manual Manual Manual Manual Manual Manual Position Indication In Control Room No Ho No No No No outside outside outside outside outside outside LC C LC C Position Relative to Hormal tontatnnnnt

~eratioa Trip on CZS Immediate postaccident cold shutdown Power Pailure Position At HA NA NA HA NA Na Also Avadable On Aperture Card Maxim Isolation Time~(sec'A NA NA UPSAR~Fi re 6'-50 6.2-50 6.2-50 6.2-50 6.2-50 6.2-50 class'otes see end of table'urge supply duct Loop B hot leg sample Pressuriser liquid space sample 204 205 206a 953 956B 966B NA al a2 Globe Needle Globe HA al, a2 Blind flange 5869 NA Butterfly 955 , NA Globe 956D al.Needle 966C a2 Globe NA Air air Manual Air air Manual Air NA Both Status No Status status No Status Inside outside Inside Outside outside Inside outside outside C C C 0 C C 0 C 0 o/c NA PC Fc NA Fc FC Na Ec NA Yes Yes NA Yes Yes NA Yes NA NA NA NA 60 NA NA 60 6.2-51 6.2-51 6.2-52 6.2-52 6.2-52 6.2-53 6.2-53 6'-53 2, 29 29 17 17 Steam generator A sample Pressuriser steam space sample Steam generator B sample Reactor compartment.

cooling unit B return Reactor compartment cooling Unit A supply Oxygen makeup to A s B recombiners 206b 207a 207b 209a 209b 210 CLIC 5735 951 956F 966A CLIC 5736 4635 4637 CLIC 4638 4758 PZ-2232 CLIC 1080A 10214Sl 102148 10215Sl 102158 al a2 NA al a2 al a2 al al a2 al al al a2 al a2 NA a2 NA NA Gate Globe Needle Globe NA Globe Butterfly Gate NA Gate Butterfly NA NA Globe Globe Globe Globe Globe air Manual Air HA Air Manual Manual NA Mannual Manual NA NA Manual solenoid solenoid solenoid solenoid NA Status Status No Status HA Status No No HA No No NA NA No Status Status Status Status Inside Outside Inside outside outside Inside outside outside outside Inside outside outside outside Inside outside Outside outside outside outside C 0 C C 0 0 C NA C LC C C C C C 0 C 0 C NA C LC C C C C C 0 C 0 C NA C LC C C C C Ec HA Ec NA Fc NA NA HA NA NA NA NA NA PC Fc FC PC NA Yes Yes HA Yes NA Yes HA NA HA HA NA NA NA HA Yes Yes Yes Yes NA 60 NA HA 60 NA 60 NA NA NA NA NA Na NA NA 3 3 3 3 6.2-54 6.2-54 6.2-55 6.2-55 6.2-55 6.2-56 6.2-56 6.2-47 6.2-47 6.2-47 6.2-46 6 2-46 6.2-46 6 2-46 6.2-57 6.2-57 6.2-57 6.2-57 6.2-57 19 17 19 26 30 28 15, 28 28 17, 28 Purge exhaust duct auxiliary steam sunplv to containment 300 301 NA 5879 6151 6165 al, a2 NA al a2 Gate Gate Manual Manual Blind flange NA Butterfly Air NA Both No No Znside Outside Outside Outside 0 o/c NA PC NA NA NA Yes NA NA NA NA NA NA 6 2-58 5 6 2-58 5 6.2-59 4 6 2-59 4 2d 29 29 Auxiliary steam condensate return 303 6152 6175 al a2 Diaphragm Diaphragm Manual Manual No No outside outside HA NA NA HA NA NA 6.2-59 4 6.2-59 4 6.2-103 REV 8 7/92 S'I 41 l GINNA/UFSAR SI A,PERTURE CARD Table 6.2-15 CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

~sszem A hydrogen recombiner (pilot)Penetration No.304a 1076A 1020581 al a2 Valve/Isolation Boundaro posdndon'alve

~Te Diaphragm Globe Valve Operator TQQB Manual Solenoid Position Indication In Control Room No Status Position Relative to Containment outside Outside Normal~ei:asian C Cold Shutdown Position At Immediate Postaccident~

Trip on CIS Power Failure HA PC NA Yes Maximum Isolation Time~sea'A 3 Also Available On Aperture Card UESAR~Pi re 6.2-60 6.2-60 Notes Class~See end of table'8 A hydrogen recombiner (main)Containment air sample postaccident Containment air sample inlet Containment air sample postaccident Containment air sample postaccident Containment air sample out Fire service water service water from A fan cooler Mini-purge supply Instrument air to containment Service air to containment service water'rom B fan cooler Service water to D fan cooler Leakage test depressurization 304b 305a 305b 305c 305d 305e 307 308 309 310a 310b 311 312 313 1084A 1020981 1554 1555 1556 1598 1599 1557 1558 1559 1560 1561 1562 1596 1597 9227 9229 4629 4633 PZA-2033 TZA-2010 CLIC 7445 7478 5392 5393 7141 7226 4630 4634 PZA-2034 TZA-2011 CLIC 4642 4646 12500K PZ-2144 CLIC NA 7444 al&2 al a2 a2 al a2 al a2 a2 al a2 a2 al a2 al a2 al al al al a2 al a2 al a2 al a2 al al al al a2 al al al al a2 al a2 Diaphragm Globe Diaphragm Globe Diaphragm Diaphragm Diaphragm Diaphragm Globe Diaphragm Diaphragm Globe Diaphragm Globe Diaphragm Gate check Butterfly Gate HA NA NA Butterfly Butterfly Globe check Gate check Butterfly Gate HA NA HA Butterfly Gate Globe HA NA Blind flange Butterfly Manual solenoid Manual Manual Manual Air Air Manual Manual Manual Manual Manual Manual Manual Air Air NA Manual Manual HA NA NA Air Air Air NA Manual NA Manual Manual NA NA HA Manual Manual Manual HA HA HA Motor No status No Ho No Both Both No No No No No No No Both Both NA No No NA HA HA Both Both Both NA Ho NA No No NA NA NA No No No NA NA NA status outside Outside outside outside Outside Outside outside outside outside outside outside outside outside outside Outside outside Inside outside outside outside Outside Inside Outside Inside Outside Inside Outside Inside outside Outside Outside Outside Inside outside outside outside Outside Inside Inside outside LC C LC C LC 0 0.0'0~C C Lo",c'NA';NA C 0/C 0/C i 0 (0.C'Lo ,"')NA'NALO'(C NA C', C C o/c C NA NA C o/c o/c o/c C NA NA C o/c C C NA C Lo C NA NA C Lo C NA NA C Lo C C NA C NA Fc NA NA NA Fc PC NA NA NA NA NA NA NA Ec Fc NA NA HA HA NA NA Ec Ec FC NA NA NA NA NA NA NA NA NA NA NA NA NA HA AI NA Yes NA HA NA Yes Yes NA NA HA HA Yes Yes NA NA HA NA NA NA Yes Yes Yes HA NA NA NA NA NA NA NA HA NA NA NA NA NA Yes HA NA NA 60 60 NA NA.NA HA 60 60 NA NA NA NA NA NA 60 NA NA NA HA NA NA NA', HA'A NA NA NA, NA'A NA 6.2-60 6.2-60 6.2-61 q 6.2-61 6.2-61 6.2-62 6.2-62 6.2-61 6.2-61 6.2-61 6.2-61 6.2-61 6.2-61 6.2-63 6.2-63 6.2-64 6.2-64 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-66 6.2-66 6.2-67 6.2-67 6.2-68 6.2-68 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-69 6.2-69 3A 3A 3A 3A 3A 3A 28 26 27 26 27 30 27 Qa.i 2 i40159-6.2-105 REV 8 7/92 0Pr't J 1 0 4'0 p C')Wa~

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

~sstem Penetration No.Valve/Baundaxar Isolation Position'alve

~Te Valve Operator~re Position Zndication Zn Control Room Position Relative to Containment Normal rrdaranddaann Position At Cold shutdown Immediate Postaccident~

Power Failure Trip on dre Also Availabte On Aperture Card Maximum Isolation Time~aea'PSAR~F1 ure class4 Notes see end of table'a Service water from C fan cooler Service water to B fan cooler Leakage test supply Deadweight tester Service water to A fan cooler service water to c fan cooler A steam generator blowdown B steam generator blowdown service water from D fan cooler Demineralized water to containment Hydrogen monitor instrumentation line Bydrogen monitor instrumentation line containment pressure transmitters PT944, PT949, and PT950 315 316 317 318 319 320 321 322 323 324 332a 332b 332c 4643 4647 PIA-2035 TZA-2012 CLIC 4628 4632 PZ-2138 CLZC NA 7443 HA 4627 4631 PI-2142 CLIC 4641 4645 PZ-2136 12500B CLIC 5738 CLIC 5737 CLIC 4644 4648 PZA-2036 TZA-2013 CLIC 8418 8419 922 924 CLOC 7452 cap 923 7456 Cap PT944 1819G PT949 1819E PT950 1819P al al al al a2 al al al a2 al a2 al, a2 al al al a2 al al al al a2 al a2 al a2 al al al al a2 al a2 al al a2 bl b2 al a2 bl b2 al a2 bl b2 cl c2 Butterfly Gate NA NA NA Butterfly Gate NA NA Blind flange Butterfly Butterfly Gate NA NA Butterfly Gate NA Globe HA Globe NA Globe NA Butterfly Gate NA NA NA Globe check Gate Gate NA Globe NA Gate NA Globe HA NA Globe NA Globe NA Globe Hanual Manual HA HA HA Manual Manual NA HA HA Motor NA Manual Manual HA HA Manual Manual HA Manual NA Air NA Air NA Hanual Manual NA NA NA Air HA solenoid solenoid NA Manual HA Solenoid NA Manual NA NA Hanual NA Hanual NA Hanual No No NA HA HA No No NA NA HA Status NA No No NA NA No No NA No NA status NA status NA No No NA NA NA Both NA Both Both NA No NA Both NA No NA NA No NA No NA No Outside outside outside outside Inside outside outside outside Znside Inside outside NA outside outside outside Inside outside outside outside outside Inside Outside Znside outside Inside outside Outside Outside Outside Inside outside Znside Outside Outside Outside Outside Outside outside Outside outside Outside Outside outside outside outside olltside outside ee Lo C NA NA C Lo C NA C NA Lo C NA C Lo C NA C C , Lo c~NA NA C C C C C C C C HA 0 NA 0 NA 0 o/c C NA NA C o/c C NA C NA o/c C HA C o/c C NA C C o/c C 0/C C o/c C NA NA C o/c o/c NA 0 NA 0 NA 0 Lo C NA NA C Lo C HA C NA Lo C NA C C NA C C Lo C HA HA C C C C C C NA 0 HA 0 HA 0 NA NA NA NA NA NA NA NA NA NA AI NA NA NA NA HA HA PC NA Pc NA NA NA NA NA HA Pc NA PC FC NA HA NA Pc NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA HA NA NA NA Yes NA NA NA NA NA NA NA HA HA NA Yes NA Yes HA NA NA NA NA NA Yes NA Yes Yes HA NA NA Yes NA HA NA NA NA NA HA NA NA NA, NA NA NA NA NA'A NA, NA'A'A NA HA NA HA HA NA 60 60 NA NA NA NA NA HA 60 NA 3 3 NA NA NA 3e NA HA NA NA NA NA NA HA 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-70 6.2-70 HA 6.2-65 6 2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-65 6.2-71 6.2-71 6.2-72 6.2-72 6.2-65 6.2-65 6.2-65 6 2-65 6.2-65 6.2-73 6.2-73 6.2-74 6.2-74 6.2-74 6.2-74 6.2-74 6.2-74 6.2-74 6.2-74 6.2-74 6.2-75 6.2-75 6.2-75 6.2-75 6.2-75 6.2-75 NA 26 27 30 27 31 30 27 30 27 19 19 26 27 32.OSSA>40 i 59'-6.2-107 REV 8 7/92

~~C~+e~'a~~J Q\C (

.o n-w~GINNA/UFSAR

~sstem penetration No.Valve/~nonndo Isolation Position'alve

~e Valve operator~pe Position Zndication In Control Room Position Relative to Containment Normal td<<erntdoonn SI APERTURE CAR9 Position At Also Available On Aperture Card cold don<<down Trip on CIS Immediate-Pove Postaccident~

Failure Maximum Isolation Time~tool UFSAR Ficiure Class~Notes see end of table'able 6.2-15 CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVIHG (Continued)

Hydrogen monitor instrumentation line Main steam from A steam generator 332d 401 921 CLOC 7448 Cap 3413A 3455 3505A 3505C 3507 3507A 3517 3521 3615 3669 11027 11029 11031 PS-2092 PT-468 PT-469 PT-469A PT-482 End caps CLZC al a2 bl b2 al al al al al al al al al al al al al al al al al al al a2 Gate NA Globe HA Globe Globe Gate Gate Gate Gate Sving check Gate Gate Gate Gate Gate Gate NA NA NA NA NA NA solenoid NA Manual NA Manual Manual Motor Manual Manual Manual Air Manual Manual Manual Manual Manual Manual NA NA NA NA HA NA NA Both HA No NA No No R/G No No No R/G No No Ho No No No NA NA NA NA NA NA NA outside outside Outside Outside outside outside outside outside outside outside outside outside Outside outside Outside outside outside outside Outside Outside Inside outside Outside Inside 0 C C C 0 C 0 0 C 0 C C C NA NA NA NA NA C C C C C C C C C 0 C C C C C HA NA NA NA NA C C o/c C o/c C o/c C C o/c C o/c C C C NA NA NA NA NA C C FC HA NA NA NA HA AZ NA HA NA AI NA HA NA NA NA NA HA NA NA HA NA NA Yes NA NA NA NA NA No NA NA NA No NA HA NA HA NA NA NA HA HA NA NA HA NA 3 NA NA NA NA NA'A NA NA NA NA NA NA NA, NA'A NA NA NA NA NA NA NA NA 6.2-74 6.2-74 6.2-74 6.2-74 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 6.2-76 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 32 18 18 18 18 6 6 6 6 6 33 19 Main steam from B steam generator 402 3412A 3456 3504A 3504C 3506 3506A 3516 3520 3614 3668 11021 11023 11025 PS-2093 PT-478 PT-479 PT-483 End caps CLIO al al al al al al al al al al al al al al al al al al a2 Gate Globe Gate Gate Gate Gate sving check Gate Gate Globe Gate Gate Gate NA NA NA NA NA HA Manual Manual Motor Manual, Manual Manual Air Manual Manual Manual Manual Manual Manual NA NA HA NA~NA HA No No R/G No No No R/G No No Ho No No No NA NA NA HA HA HA outside outside outside Outside outside Outside Outside Outside outside outside Outside Outside Outside Outside Outside outside Outside Outside Inside 0 C C C 0'C 0 0 C 0 C C C HA NA NA NA C I C C C C C C C C 0 C C C C C NA HA NA NA C C o/c C o/c C o/c C C 0 C o/c C C C NA NA NA NA C C NA NA AI NA NA HA AI HA NA NA NA NA NA NA NA NA NA HA HA NA HA Ho NA NA NA Ho HA HA NA HA NA NA NA HA NA HA'A HA=HA NA NA NA NA NA'A NA HA HA HA NA, HA HA'A NA'A'A'.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 6.2-77 18 18 18 18 18 6 6 6 6 33 19 Feedvater line to A steam generator 403 3993 3995X 4000C 4003 4011A 4003A 4099E 8651 CLZC al al al al al al al al a2 Check Globe Check check Globe Gate Gate Gate HA NA Manual NA NA Manual Manual Manual Manual NA NA No NA NA No No No No NA outside outside outside Outside outside Outside outside Outside Inside 0 C C C C C C C C C C o/c o/c C C C C C HA HA NA HA NA HA NA HA NA NA NA NA NA HA NA NA NA HA NA HA NA HA HA NA HA'A NA 6.2-78 6.2-78 6.2-78 6.2-78 6.2-78 6.2-78 6.2-78 6.2-78 6.2-78 34 34 34 19 9215140159-6.2-109 REV 8 7/92

'$a Q g~4~A GINNA/UFSAR

~sstem Penetration Ho Valve/~Bunda Isolation Position'alve

~e Valve operator~T Position Indication Zn Contzol Room Position Relative to Containment Normal cSs rand'n Position At cold Shutdown Immediate Postaccident~

SI APERTURE CARD Aiso Availab1e On Apert(ire Cat'~.Isolation Pover Trip on Time FFS.Sara CSF UPSAR~Pi re Table 6.2-15 Class4 Notes See end of table'ONTAiNMENT PIPING PENETRATIONS AND ISOLATION VALUING (Continued)

Peedvater line to B steam generator 404 3992 3994E 4000D 4004 4012A 3994X 4004A 8650 CLZC al al al al al al al al a2 Check Globe check Check Globe Gate Gate Gate NA NA Manual NA NA Hanual Manual Manual Hanual NA NA Na NA NA No Na No No NA outside outside Outside outside outside outside Outside outside Inside 0 C C C C C C C C C C C C C C C C C C C o/c o/c C C C C C NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA , 6.2-78" 6.2-78 F 6'78 6.2-78 6.2-78 6 2-78 6.2-78 6.2-78 t, 6.2-78 34 34 34 Personnel hatch 1000 NA NA al a2 NA NA NA NA NA NA Inside Outside o/c o/c NA NA HA NA NA NA 3.8-31 NA 3.8-31 NA Equipment hatch 2000 NA NA al a2 NA NA NA NA Inside outside 0/C 0/C NA NA NA 3.8-30 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 2 vas added as a result of EWR 4998 to facilitate steam generator maintenance activities during reduced inventory operation.

This penetration is closed by a double-gasketed blind flange on both ends.The innermost gasket for each flange (i.e., gasket 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 I NNA/UFSAR SI APERTURE CARD Table 6.2-15 Also Available On Aperture Card CONTAINMENT PIPING PENETRATIONS AND ISOLATION VALVING (Continued)

()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)(19)(20)(21)(22)This valve is normally open at power since 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 required.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.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).Operations is instructed to manually close AOV 745 following a containment isolation signal to provide additional redundancy.

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)(25)(26)(27)(28)(29)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 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.

This manual valve is subject to an annual hydrostatic leakage test and is not subject to 10 CFR 50, Appendix J leakage testing.I 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 shutdown leakage checks.This solenoid valve is maintained inactive in the closed position by removal of its dc control power.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[

TEST CONNECTION I BI TEST 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 Vi o.2 Fuel Transfer Tube, PenetrationiVo.

29 REV 6 12/90

RCS CHARGING LINE PENEHM.TION 100 I O!!!I!I I I I!!CIy P100 K'OB ORS Y/J NOTE DESCRIPTION ontanment Basement (Regenerative Hx Area)Auxiliary Bldg Basement (RWST Area)LRM should be located in Containment Basement (Regenerative Hx Area)PTT-23.8 Revision 1 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 azv 8 7/92 SITO LOOP A SITO LOOP 8 and SITEST LINE PENETRATIONS 101 110b and 113 022 CLLL gs g I 0210'll mal I P113 12404 TC/J o 4<I a72A I I LJ 21 A BLc I LI arlrr I oa 0/J TC/J 2010 Sl PVLLP C a1 a1 II ll al al TO 201 P101 CV I aaaa 12401 Ca}cw TO 200 LC a21L'pr 023 LC P110b (12ala era 2$LO I I 1 0/J 0/J L TO w 1 n" a2Q h 02a 028 UNE 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 REPORTFigure 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 383B!I 2227 TC/J Rg 525 f.)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 CIB 5380 5$29A TC/J LC (CAP)Y/J 5129 CONSTRUCTION ClV FlRE SERVICE WATER CONNECTIONS 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" CONTAINMENT SPRAY HEADER PENETRATION 105!I IQ 3[5!eros!!!!!I FROM PENElRATION 109 869A CIV PI 7 285e CIV TEST IjNE TO RWST Y/J 8&DA BLC M S 8&DB BLC~NQ R 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 O I 9303 9304 CIV P108 80ll!TC/J!!!I 300A 300B 2224 V/J 301B Fl 1 301A 303A 303C cg 277A LC 275 NOTE (2)DESCRIPTlON Containment Basement (Regenerative Hx Area)Auxiliary Bldg Basement (RWST Area)LRM should be located in Containment Basement near Regenerative Hx Operated with Reach Rods by the CS Pumps Located 20 ft above the floor behind RWST Area PTT-23.9A Revision 1 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 aZV 8 7/92

SUMP"A" DISCHARGE PENETRATION 107 TC/J 1002 10035 10036 10106 I I..g)B 82 m)Pc 8 I FC P1071 I 1725 1072 FC CIV 1723 D/J V/J 10012 10006 0 I1759 1757 1760 1758 S 10023 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 FROM REACTOR COOLANT PUMP SEALS TC/J 2213 I g(3 314 o!Pz V tMB OMB 117 Pl 118 318A OMB IMB 322A 2214 311e Lo 2212 FROM LETOOVe EXCESS HEAT EXCHANGER P1O8)I I I I I I, ctv 313 FILTER 315C SEAL RETURN FtLTER V/J 3858 3828 385A FROM~(It REACTOR cootANT PUMP SEALS FROM~Ao REACTOR CoolANT PUMP 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 PTT-23.11 Revision 1 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 REV 8 7/92

'I I Wl~

"B" CONTAINMENT SPRAY HEADER PENETRATION 109 Pl 55 2830 LC CIV FROM PENETRATION 105 85QC CIV 8648 LC CIV 11620 2825A TEST LINE TO RWST 2825 M S P109 i~(I I I I Pl 77 8698 CIV 2826 CIV 2826A CIV PI TC/J 860 BLC M 8600 BLC~~5~h xo~W~lO gD Qu Cl CP o 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 LC CO O I P110a 304B I 9301 275 3018 301A 9302 V/J TC/J NOTE DESCRIPTlON Containment Basement (Regenerative Hx Area)Auxiliary Bldg Basement (RWST Area)Operated with Reach Rods by the CS Pumps Located 20 ft above the Iloor behind RWST Area (2)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 REV 8 7/92 PTT-23.98 Revision 1 LRM should be located in Containment Basement near Regenerative Hx RESIDUAL HEAT REMOVAL TO B COLD LEG PENETRATION 111 (PENETRATION

<4D)RESIDVAL HEAT REMOVAL LOOP OUTlET VALVE I (PENETRATION 112)UNE INSTALLED BLANK 7198 718A SAMPlZ SYSTEM FC T 958 721 BLQ 2748 702 S S M Ll 852A 8528 BU'LC TO REACTOR VESSEL 627 P111!2TEO 2780A 51"!!~I 0~5 ES 88 Isolation from 33013-1247, Revision 19 ROCHESTER GAS AND ELECTRIC CORPORATION R.E.GINNA NUCLEAR POWER PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Figurel).2-24 Residual Heat Removal to Loop B Cold Leg Penetration 111 REV 8 7/92 LETDOWN LINE FROM REACTOR COOLANT SYSTEM PENETRATION 112 R LOOP r I I I mi I 545427 8 AREA 7I v////////////////HX/TTC C/2 I I I I I I I o/4 I I I FROM RHR STETEM ee I I I I0 g TO PREESUIRZER REUEF TAHX RHR FENCE I 0/i I 2000 I (CTV 2'V/J 669$STION I'I l awsr I r---1 I FC I~TT2I 21 I I I I LC y I I+I I 22ET I I I I 0/J I L J I I I I I o>>)L Jg CIV I I HRHX RM E64~I 1~)I I CN I 2 I-5 I IgR I I I I I I I I I I lg~~J I>>Q I lhn NOTE (4)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 Located Auxiliary Bldg Basement nesr B CS Pump PTT-23.6 Revision 1 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 REV 8 7/92 STANDBY AUXILlARY FEEDWATER TO STEAM GENERATOR A PENETRATION I'I9 AND 123b IL O 4J'X 4J I lA O I TC 9719A I ls us s726 Sl 0 9706A g705A P 1 19 LO I 9727 9723 CIV CIV Q 9702A 9704A BLC FROM STANDBY AUXIUARY FEEDWATER PUMP C IO O I lA O I TC 9706B LO 9719B 0705 B P123b 9724 CIV CIV M LC 0725 CIV'LJ D 9702B 97~B LO BLC FROll STANDBY AUXLJARY FEEDWATER PUMP D 9722B Isolation from 33013-1238, Revision 8 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 REV 8 7/92

NITROGEN TO ACCUMULATORS PENETRATION 120a o 5 8625 TC/J I~"~i 9 S627 8626 Pl 627 8621 VENT O EP O I Ul~CC~O Vl&4J g CC~0 w>Og e9~s o:O QI 0 624 CIV P120a 8&23 I I TC/J I CtV 846 2831 V/J 862S 944 S628 g 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 Ld LU CL lL 4J t4 IK M Vl 4l CL O 527 jc 538 P)20b CIV 546 493, CIV 539 FC 492 hl a N O I TC/J-..NOTE (2)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 Downstream vent point PTT-23.1 Revision 1 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 Szv 8 7/92 PRT MAKEUP WATER PENETRATION 121a O'J 4~cg~9307 EL~o~I 568 CIV 529 LJ TC/J 567 Ig h!m 497!P121o!49S!,0/J CIV 508 FC O le~V O.CI x5 TC/J 576 NOTE (3)(4)DESCRIPTION Containment (Square Corner)Auxiliary Bldg Mid Floor (SFP Hx)LFIM should be located in the Containment Square Comer Area 3/4'ipe connection Located 10 ft above the floor, adjacent to missile barrier Located 15 ft above the floor PTT-23.3 Revision 1 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 REV 8 7/92 PRT N PENETRATlON 121b bC 4 lal 0 I 545 496 CF 52B 495 494 P121b!gy!547!.ci.!I V/J 441 1662 CL IL o lC I'K 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 OPEN PIPE TC/a P121 c 1818A 18188 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 P 1T-23.17A Revision 1 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 tv 8 7/92 l gK RCDT TO GAS ANALYZER PENETRATlON 123a 1004 1717A Cl X 1655 FC CIV 1789 1709F 1020 FC s T 1600A 5 O I 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 PTT-23.21 Revision 1 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 azV 8 7/92 CCW TO FROM EXCESS LETDOWN HX PENETRATlON 124a and 124c o a.au hgh O uS>o s-vl o3<~ov)742B gg~M>oS Qo cg o FC clv 745 2727 P124c i I=l:I;I I/I-I'XCESS LETDOWN HEAT EXCHANCER (I Aov/I I'TC/J 2776 742C I I REACTOR cooLANT I I I I I iol 3I IH I I I I 744 I I I RHR FENCE I I 743A I I I I PI24]2725 l~v/J I I l K o o I K W'X o X o o X o 4 0 n 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 PTT-23.30 Revision 1 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 REV 8 7/92

CONTAINMENT POST ACCIDENT AIR SAMPLE C FAN PENETFIATION 124b and 124d I I..TC/J 124b 124d I gw z TC/J Z+>I I X<o LC CIV 1569 1570 CIY LC CIY 1572 LC CIV 1571 LC CIV 1574 Y/J I~x O I M I I X ID O CP I Y/J 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 PTT-23.50C Revision 1 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 azv 8 7/92 CCW FROM B RCP PENETRATlON 126 7548 7568 Fl CCW 1 FROM IKACTOR COOLANT PULIP 18 7578 765A-12305B 765B oim I!1 Ll P 125 CIV 7508 2731 762B cp~I 0 gA au@O 8 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 PTT-23.29 Revision 1 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 REV 8 7/92

&tbsp 1" CCW FROM A RCP PENETRATION 126 FO 765D I I 7$6A CCW Fl FROM REACTOR 61 COOLANT PUilP 1A 757A 12308B I TC/J i M P126 CIV 1)759A 2729 IC'X 702A I 0 ug CP X O V O 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 PTT-23.28 Revision 1 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 REV 8 7/92 COMPONENT COOLING WATER TO REACTOR COOLANT PUMP 1A PENETRATION 127 N$13 2723 MO OX I O O.5>cc z o~"8 0 N 0.749B 2732 Nx LJ z Oa I 742A VW aL 0Ão I 7$1A 761r 75DC CIY P127 i 750A TC/J N 740A 2761 2730 tc/J 0/J N~>g zg)e 617 BLO Q co R~~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 PTT-23.26 Revision 1 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 REV 8 7/92

CCW TO"8'CP PENETRATION 128 I,;lh y/SZ I$13 a5 I Og, g>a8 M P 07M o~85 aH I 742A UpOA a 7511 761E 75QD CN P1Zg]7608 I I CIV g 740B TC/J P/J$8 le 8 gl u.Q8 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 PTT-23.27 Revision 1 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 azV 8 7/92 RCDT GAS HEADER PENETRATION 129 FC FC CIY 1787 CIV 1786 1675 1676B Y/J!!!0 O'ISA 1016 IL 40 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 PTT-23.20 Revision 1 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 REV 8 7/92 CCW FROM 0 RX SUPPORT CLRS PENETRATION 130 and 131 QwQ-s.E 8 815A Y/J CY 8th I I 3I"I"~(8 P130 I~818 I 0/J I))5 eIlg O)I o)~5s-I oE)~I HICH I YENTS TC/J TC/J 1734, 1733 SUPPLY fROQ COMPONENT COOUNQ WATER PUMPS 817 M BLO REACTOR SUPPORT COOLERS P131i 815 I 271 g 1723 I i D/J Y/J 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 PTT-23.24 Revision 1 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 REV 8 7/92 MINI PURGE EXHAUST PENETRATION 132 I Lal K CP C)CC FC CIV 7970 r-!9 8'CIS!P)52!79710 I TC/J FC T m~CIV 7971 Pg aK C I 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 140 252 REFUEUNO WATER STQRAGE TANK 5 g~M X 27d5 P140 701 nss CN I!'-00 PENETRATION 111)$54 TO RESIDUAL HEAT REMOVAL SYSTEM Isolation from 33013-1 247 Revision 19 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 REV 8 7/92 SUMP"B" TO"A" RCDT PUMP PENETRATloN 1a1 TO RCDT PUMPS REFUEUN0 WATER STQRACE TANK (CASED SmrZM)851A CO BLO M C~I!I 1813C CV$8$8A BLD 1818D Y/J TC/J 705C 850A BLC!R!i5 P NOTE'DESCRIPTION CNMT Sump B Auxiliary Bldg Basement LRM should be located in Auxiliary Bldg Basement (connected in RHR Sub Basement)P1T-23.5A Revision 1 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 aZV 8 7/92 SUMP"B" TO"B" RCDT PUMP PENETRATlON

't42 REFVEUNO WATER STORAGE TANK (CLOSED SmrZM)BLO M 8518 1818F CIV 18158 BLD M 705D 850B TC/J 8LC~7048!~s~III 48 i P 1818K Y/J TO RCDT PUMPS NOTE DESCRIPTION CNMT Sump B Auxiliary Bldg Basement LRM should be located in Auxiliary Bldg Basement (connected in RHR Sub Basement)PTT-23.5B Revision 1 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 REV 8 7/92 I A"+Pe\-0 RCDT DISCHARGE PENETRATION 143 Pl 01 179SG LC FROM FUEL LC TRANSFER CANAL DRAINS CIY 1722 FC CIY 100$A 172@A 1726 CONTAINMENT SUMP I I I s-I g I o I 8~~C I 0 I ceo I~I o L S43 I I I TC/J I I I 1 I P145)I CIV 1721 I 17090 I FC CIV 10058 RCDT PUMPS 1725A Pl 01 1811A 1727 18118 NOTE DESCRIPTION Containment Sump B Auxiliary Bldg Sub Basement LRM should be located at entrance to Containment Sump B PUT-23.22 Revision 1 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 REV 8 7/92 REACTOR COMPARTMENT COOLING UNIT A SUPPLY AND RETURN PENETRATIONS 201a AND 209b tO I O CIV 4757 I ml g!!!!47940 P201a REACTOR , COMPARTLtENT

[COOmR)A!l I I!I!CIS 4590 CIV 4758 S 4BSA 5I Isolation from 33013-1 250-3 Revision 7 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 REV 8 7/92 REACTOR COMPARTMENT COOLING UNIT B SUPPLY AND RETURN PENETRATIONS 201b AND 209a l sl.I nl5.Ig)4O CIB Pl N X O cv 4BJ5 4837A CLIC P20$b REACTOR COMPARTMENT

$B I CIY 4778 CI 4778A 5 C I Isolation from 33013-1250-3 Revision 7 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 REV 8 7/92 B HYDROGEN RECOMBINER AIN AND PILO PENETRATION 202a and 202b)IL I-a'I Ri""I Lc P202b 1076B c)v 1075B Tc/J S cN 10211$1 V/J 102'11$10203$1 I a, I m)O 10203S Q P)a 10204S I o 5)5 O~o3>w a Lo P202a 8426 1084B c)v 1063B.c/.I 5 CIV 1021351 V/J 102'l3$10204S1 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 PTT-23.51B Revision 1 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 REV 8 7/92 CONTAINMENT PRESSURE TRANSMITTERS PT-947 AND PT+48 PENETRATION 203a CIB CIB Y/J Y/J CIV 1819C CIV 1819D TC/J DPEN P)PE P205a 1818C 1818D 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 PT1=23.17B Revision 1 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 REV 8 7/92

CONTAINMENT POST ACCIDENT AIR SAMPLE FAN PENETRATION 203b and 203c o>TC/V 4o CC~P205o!!!t!I KQ P2Mb CIV 1505 1584 CV CIV 1565 CIV OV 1560~1500 1587 CIV V/J 8 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 PTT-23.50B Revision 1 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 REV 8 7/92 PURGE SUPPLY PENETRATION 204 TC/J o[g P204 FC~CO T CIV (BLIND FLANGE)I I I I Pl 8074K 8074 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 FROll RCS LOOP B-HOT DELAY h!L'ml j P205 IMn 10M'27C 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 PTT-23.12C Revision 1 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 RVr 8 7/92 PRESSURIZER LIQUID SAMPLE PENETRATION 206a 10001 FC 9g1 P206a 958K CN 9668 927 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 0 N X TO A S/G 8 LOWDOWN 5748A 5781 I I y/J 5711 I I w 5769D 3 Ld I P206b FC su, 5785 y/J 5749 I aQ gx 0>~5 Oa 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 40 g FC 9510 I g!g 51~8 958F TC/J I Y/J 958G 999 E Y/J 10000 921A 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~O g&m/~~La 31: 1080A CIV 1021 4S1 CIY S T S T 1021 4S S T 55 Clg 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 PTT-23.51C Revision 1 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 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.Ig Q)cr I LC STEAM FROM HOUSE HEATINC BOILER 7050 7941 TC/J STRAINER SPACE HEATERS 975 PM1 CIY 6151 I 7040 I I I'5O5~car 6175 CIY 6165 7944 CIV 6152 D/J Y/J 7945 V/J T 4J 7946 mg O~W~OX O 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 PTT-23.40 Revision 1 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 aZV 8 7/92 A HYDROGEN RECOMBINER AIN AND PILO PENETRATION 304a and 304b K I cL'~f+lal Q QM ID I$0 8 g CL IB~la LC S 107~CIV CIV 10205S1 1075A 10205S 10207S1 TC/J v/J o O IY O I 10207S 10202S X I S 10202S1 10%A CIV 10209S1 L 10209S 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 PTT-23.51A Revision 1 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 aVr 8 7/92 CONTAINMENT POST ACCIDENT AIR SAMPLE PENETRATION 305a 305c and 305d FROM 8 FAN ClV 1555 CV 1555 Y/J FROM A FAN CIV 1557 I Y/J g TC/J P305d LC COMMON RElURN 156O CIV 1552 CIV 1551 Y/J 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 Pl%-23.50A Revision 1 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 REV 8 7/92 CONTAINMENT AIR SAMPLE ETURN PENETRATION 305b 10010 FROM POST ACCIDENT SAMPUNO SYSTEM OPEN PIPE TC/J PM5b~y 1599 15S8 FROM 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 CONTAINMENT AIR SAMPLE OUTLET PENETRATION 306e!Il OPEN PIPE tc/a!cy 1596 FC CIV 1597 10009 P0Sr/ACCInm SAMPUNQ SYSTEM 05 O C)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 KCC X Q 3~TC/J 9231 ClY 9229!!Ig 8!g m P507!~y gQ 9nS 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 5 ((av)I 4627 4626 0 4641 4642 5s gO 0 Q P319 I I I.I I I I I I I (090 4631 4514 45 I 5 4513 451 d (CIV)12500 H 125006 (as)2142 Pl 2'136 2156 2144 (as)2034 45226 45246 45926 (312)0 4522A 4524A 4592A 4594A 4655 465d 4659 4524 4660 4592 4594 P306 (0N)311)4633 315 4634 323 I 4641 4630 tI 4644 125020 12502R 12502T 0 12502U~20!0~2011 2012 2013 Isolation from 33013-1250-3, Revision 7 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 REV 8 7/92 MINl PURGE SUPPLY PENETRATlON 309 TC/J LC FROLI IIIN 748O SUPPLY FAN 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 g Zg~g 14100 7045 CY MQS 96 MQ5 TC/J!!Ig FC P$50a 1 tHQ2 5450A 5450 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 lal fJ Q~CP 4l~>i~X Q~CO O~7227 l l=IS~y P310b 7225 CV 7141 V/J 714'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 PTI-23.32 Revision 1 Figure 6.2-68 Service Air Penetration 310b aZV 8 7/92 0

LEAKAGE TEST DEPRESSURIZATION PENETRATION 313 BLIND FLANGE ClY TC/J 7 ClY 7478 V/J 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 PTl-23.42 Revision 1 Figure 6.2-69 Leakage Test Depressurization Penetration 313 aVr 8 7/92

LEAKAGE TEST SUPPLY PENETRATION 317!Ig.I.i5 I Q FLANcE 8 I I I Cy P3$7!!!1 744$7473 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!I lm g<<!e g lm CW 570$28 e 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 g~!aa x!I Q Id p822 5702 9h 5VO5!!I 5768A v/~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 DEMINERALIZED WATER PENETRATION 324 yB%!g!!!!TC/J t v/z gh 502$~cr NL'OTE 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

CONTAINMENT H MONITORS PENETRATION 332a 332b and 332d I X C P5$2a P532b 7448 CIY CN TC/J 7452 CIY Y/J S T CIY 921 S T CN S 922 T 924 0208 5 020A g 8451 7456 I 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 181M CIV 1818F CIV 1819E Hl-!OPEN PIPE 1818G 1818F 1818E 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 FanPal=23.17C Revision 1 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 aZV 8 7/92 MAIN STEAM FROM STEAM GENERATOR A PENETRATION 401 II'10K CB CB CB CB 11025 3409 C CV 11051 54IIB CV CV~11020 (4 lYP)5500 5511.5515 AllQSPHERE TO lEMPKRATlJRE CONF ENSATEQ SUPPORTS 8517 CV i5 I 5510 5 TO AUXIUAlA'EEOWATER 3411 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 402l TD AUXIUARZ I TEEDWATER]Cruasua)I IV 5504A 5504C I I 5504 (4 TTP]5505 551 0 551 2 ATQCSPHERE lg I I 1120 11021 CIV PS B 541 2C d721 551 d 5515 g 5514 CIV 54455 I I I 11024A I I 11025 CIV 11022 541 0 5520 PT D 47B ad PT 11025 470 CIV 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 AUXILIARY FEEDWATER TO STEAM GENERATORS A AND B PENETRATIONS 403 AND 404 IXOM lMKNMK OXIVKN AUXILIAXV IKXOMAllÃtWt o I m Pco I 8415C I lc I I I.i5 jib I I I I 10 IMOM MAIN FIZOWA1KN IIIMI IA 0001 CIV CIV~M MAIN IZGWAIKN 400K tVW I~CIV KNNI IN!ON OXIVKM AOXRWtf AZDVAlKN WMt 1A CIV NOIX l0IKA IXV FKKAI QKOO ONVKM AXOIAINK IKXOMAIKN l%NP 1~CIV 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 I Changes Technical Specification Changes Effect 2.3.4~5.6.7.8.Removed reference to Table 3.6-1 from Technical Specifications 3.6.3.1, 4.4.5.1, and 4.4.6.2.Added statement to Bases for Technical Specification 3.6 that containment isolation boundaries are listed in UFSAR Table 6.2-15.Removed Table 3.6-1 from Technical Specifications and placed information in UFSAR Table 6.2-15.Revised action statement of Technical Specification 3.6.3.1.Removed definition of leakage inoperability from Technical Specification 3.6.3.1.Added statement related to intermittent operation of boundaries to both Technical Specification 3.6.1 and the bases.Removed note associated with Technical Specification 3.6.5.Added definition of"isolation boundary" to Bases for Technical Specification 3.6.Updated reference list contained in Bases for Technical Specifications 3.6, 3.8, and 4.4.No technical change.Specifications are now consistent with Generic Letter 91-08.Valve listing remains in a licensee controlled document under 10 CFR 50.59 program.Specification now considers closed systems as an acceptable interim passive boundary and is more consistent with Standard Technical Specifications.

Definition is found in Technical Specification 4.4.2.2.Eliminated redundant discussion of leakage acceptance criteria.No technical change.Specification now consistent with Generic letter 91-08.Mini-purge valves have been installed so specification is considered effective.

No technical change.No technical change.Clarification of"isolation boundary" provides consistency with UFSAR Table 6.2-15.No technical change.

Attachment P Pago 2 of 3 Changes Technical Specification Changes Effect Revised action statement of Technical Specification 3.8.1 section a.Revised action statement of Technical Specification 3.8.3.Revised bases for Technical Specification 3.8.Added"Pt" and necessary definitions to Technical Specification 4.4.1.4 section a.Added to the definition of"Lt" in Technical Specification 4.4.1.4 section b.Added definition of"Pa" and"Lam" to Technical Specification 4.4.1.4.Added steam generator inspection/maintenance penetration to Technical Specification 4.4.1.5 section a (ii).Revised first line of Technical Specification 4.4.1.5, section a (ii).Clarification only.Specification now consistent with Standard Technical Specifications.

No technical change.Specification now specifically addresses affected containment penetrations.

No technical change.Bases are now consistent with Standard Technical Specifications and support changes to 3.8.1 section a and 3.8.3.Addition of"Pt" definition provides clarification of testing type consistent with 10 CFR 50, Appendix J.All terms in 4.4.1.4, section a are now fully defined.No technical change.Addition of"Lt" definition provides clarification consistent with 10 CFR 50, Appendix J.All terms in 4.4.1.4, section b are now fully defined.No technical change.Addition of"Pa" and"Lam" provides clarification consistent with 10 CFR 50, Appendix J.All terms in 4.4.1.4 now fully defined.No technical change.Addition of this penetration provides, testing criteria similar to the equipment hatch and containment air locks.Minor clarification only.No technical change.

Attachment F Page 3 of 3 Changes Technical Specification Changes Effect 17.18.19.20.21.22.23.Revised acceptance criteria provided in Technical Specification 4.4.2.2 Replaced"isolation valve" with"isolation boundary" in Technical Specification 4.4.2.3 and the Bases for section 4.4.Removed notes associated with Technical Specification 4.4.2.4 section a.Also, deleted reference to section d.Added steam generator inspection/maintenance penetration to Technical Specification 4.4.2.4 section b.Removed Technical Specification 4.4.2.4 section d and associated note.Revised statement for Technical Specification 4.4.5.1.Revised statement for Technical Specification 4.4.6.2.Clarification only.No technical change.Minor clarification only.Specification and bases are now consistent with the revised Technical Specification 3.6.3.Mini-purge valves have been installed so specification is considered effective.

Section d will be removed from Technical Specifications with this amendment.

Addition of this penetration provides testing criteria similar to the equipment hatch and containment air locks.Blind flanges have been installed so specification is considered effective.

No technical change.Specification now consistent with Standard Technical Specifications.

Specification now consistent with Standard Technical 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,'::,,'.:,:;";,~";

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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 ox>OsniytOraa'Or'e'i'!O'On:,,':a';

'nJa'e'nr'"lip'a'n'etr'i"': O'ne'j'it er i a~b.Restore tba'ilia'a'Ob inoperable

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to eperahke~OPEHABG'8 status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, ox 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~d.4:-large-,"::::;Verif j,i: tb,pg opaeratigl4t y~;,:,:::!

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'-:-"'-'"-;"'-:'::i-Jt"::--"e:-*":::::is 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 3.6.4.2 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.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.

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

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~i:lyg;.Radiation levels in the containment shall be monitored continuously.

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 3.8.2 3.8.3 flange.If this condition is not met, all operations involving movement of fuel or control rods in the reactor vessel shall be suspended.

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.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

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'!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.

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

~~lie 3.i3.3.V e v Containment penetrations that employ resilient seals, gaskets, or sealant compounds, piping penetrations with expansion bellows and electrical penetrations with flexible metal seal assemblies.

Air lock and equipment door seals.Fuel transfer tube.Isolation valves on the testable fluid systems lines penetrating the containment.

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,::!

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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 c~within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be shutdown and depressurized until repairs are effected and the local leakage meets the acceptance criterion.

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!'

4.4.4.2 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.

Pre-Stress Confirmation Test a~b.Lift-off tests shall be performed on the 14 tendons identified in 4.4.4.la above, at the i n t e r v a l s specified in 4.4.4.1b.If the average stress in the 14 tendons checked is less than 144,000 psi (604 of ultimate stress), all tendons shall be checked for stress and retensioned, if necessary, to a stress of 144,000 psi.Before reseating a tendon, additional stress (64)shall be imposed to verify the ability of t h e tendon to sustain the added stress applied during accident conditions.

4.4.5 Containment

Isolation Valves 4.4.5.1 4.4.6 Each comma'a%ment')isolation valve~,::.:,,:::.,',,t!mme'ccordance with the Ginna Station Pump an8 Valve Test program submitted in accordance with 10 CFR 50.55a.Containment Isolation Res onse 4.4.6.1 4.4.6.2 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.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.I 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 a,P'r 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, Reliability Handbook, February 1963.Autonetics (8)FSAR Page 5.1-28 4.4-17 Proposed