ML17263A319

Proposed Tech Specs,Removing Containment Isolation Valve Table 3.6-1 from TS

ML17263A319
Person / Time
Site:	Ginna
Issue date:	07/15/1993
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A317	List: ML17263A316 ML17263A318 ML17263A319
References
NUDOCS 9307220182
Download: ML17263A319 (155)

Text

ATTACHMENT A Proposed Technical Specification Changes (9307220182, 930715 PDR ADOCK 05000244', P PDR I0 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 4.4-6 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 3.6-4 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 0 Q'cF r$k, s 4I Containment S stem A licabilit Applies to the integrity of reactor containment.

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

S ecification:

3.6.1 Containment

Inte rit a~Except as allowed by 3.6.3, containment integrity-shall not be violated unless the reactor is in the cold shutdown condition.

Closed valves may be opened on an intermittent basis under administrative control.b.The containment integrity shall not be violated when the reactor vessel head is removed unless the boron concentration is greater than 2000 ppm.o 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.CS 3.6-1 Proposed 1 I i py~@+4 gt'hi N'V y,+~~l'Ai l lt la I y l,III A~l'I 3.6.3 0 3.6.3.1 Containment Isolation Boundaries With a containment isolation boundary inoperable for one or more containment penetrations', either: a.Restore each inoperable boundary to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or b.Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, one closed manual valve, or a blind flange, or c.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 1 t t a II\If 4 II'0 1 tf t, t 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 a procedure under the control of Technical Specification 6.8.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 notaffected 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.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.CS 3.6-3 Proposed l P<A 7

References:

(1)Westinghouse Analysis,"Report II (2)UFSAR-Section 3.8.1.2.2 (3)UFSAR-Section 6.2.4 for the BAST Concentration Reduction for R.E.Gonna , 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 3.6-4 Proposed

'~I fiJ TIlk' REFUELING A licabilit Applies to operating limitations during refueling operations.

Ob ective 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.

Qoa~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.2,Ã8 Proposed

'1'I l'J V~l I flange.If this condition is not met, all 3.8.2 3.8.3 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 I ,~l'(I'$g>i~"~I>>l I~0 t~1 f ill 1I N provided on the lifting hoist to prevent movement of more than one fuel assembly at'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 the potential consequences of 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, 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 I~)J i r',~(~.P, Re f erences (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 I~~x'i Acce tance Criteria a 0 b.The leakage rate Ltm shall be<0.75 Lt at Pt.Pt is defined as the containment vessel reduced test pressure which is greater than or equal to 35 psig.Ltm is defined as the total measured containment leakage rate at pressure Pt.Lt is defined as the maximum allowable leakage rate at pressure Pt.I PC i~I~Lt shall be determined as Lt=LalzaJ which equals.1528 percent weight per day at 35 psig.Pa is defined as the calculated peak containment internal pressure related to design basis accidents which is greater than or equal to 60 psig.La is defined as the maximum allowable leakage rate at Pa which equals.2 percent weight per day.c~The leakage rate at Pa (Lam)shall be<0.75 La.Lam is defined as the total measured containment leakage rate at pressure Pa.Test Fre uenc a~A set of three integrated leak rate tests shall be performed at approximately equal intervals during each 10-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: 1~iii 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 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

'l I-i>.y$P"4p)4 n.>'C 5'3 4b.The local leakage rate shall be measured for each of the-following components:

1~~~lie 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.ills Fuel transfer tube.iv Ve 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.4.4.2e3 Corrective Action'a~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 I+4 k, 1e F&i b.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.

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

4.4.2.4 Test, Fre uenc a.Except as specified in b.and c.below, individual penetrations and containment isolation valves.shall be tested during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years.b..The containment equipment hatch, fuel transfer tube, steam generator inspection/maintenance penetration, and shutdown purge system flanges shall be tested at each refueling shutdown or after each use, if that be sooner.Amendment No.18 4.4-7 Proposed 0~S 4 C, 4 I 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.l'8 4.4-8 Proposed

'I l'I<1 1 0 J l S~e)l'q,e, r iQ!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 5%of the total wires are broken,-the.

reactor shall be shut:down~and:depressurized.

Pre-Stress Confirmation Test a 0 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,LL 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 temp'erature 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 s 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 I I 1~k a.~1.,~'i~0'I I" I g 4 n'f 4 t' The basis for specification of a total leakage of 0.60 La from'pen'etrations 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 I I'I J~i ci<w~Q i I T he 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.

o

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 I4>

ATTACHMENT B Safety Evaluation I vj C,'4'I Attachment B Pago 1 of 4The 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.The bases for Technical Specification 3.6 will include a statement that the listing of containment isolation valves and boundaries will be maintained in a procedure under the controls of Technical Specification 6.8.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 shutdown 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 necessary flanges and 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 various plant documents.

However, this amendment will remove Technical Specification Table 3.6-1.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 station procedures will maintain the listing of the containment isolation boundaries within a licensee controlled document.This listing is currently maintained in Procedure A-3.3 which is subject to the change control provisions of Technical Specification 6.8 as required by Generic Letter 91-08.A copy of Procedure A-3.3 is provided in Attachment D.

1 V+t 44 V 0 w>g.h 4~I I g l),ag Attachment B Page 2 of 4 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.A note related to"closed valves" only was added to Technical Specification

3.6.1 since

many test connections that are'required to be open dur'ing power operation for testing purposes are not locked closed at Ginna Station.These valves are maintained closed by system lineup procedures and"containment isolation boundary" control tags and verified closed by operator walkdowns.

This provides equivalent protection to locking devices since all plant personnel are trained with respect to the use of equipment control tags.A discussion of the necessary administrative controls required for opening these valves was also added to the bases for Technical Specification

3.6 consistent

with GL 91-08.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.

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 does 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.

l t l~4 ij s au<r', gg Q4~Ci 4 All 4*<<~<ac"~~Qt.w'J.1 T t j V<"'l U Attachment B Page 3 of 4 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, RGRE, 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 the shutdown purge flanges and mini-purge 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.

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: 1.involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.create the possibility of a new or different kind of accident previously evaluated; or 3.involve a significant reduction in a margin of safety.

~~~'~,l~~I' IJ I Attachment B Pago 4 of 4These 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 E.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.

I sf tv0 ATTACHMENT C Response To NRC Request For Additional Information Letter From-A.R.Johnson, NRC, to R.C.Mecredy,.RGRE,.dated March 11,.1993 IPj;]l'I Attachment C Page 1 of 17 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 Request for Additional Information (see letter from A.R.Johnson, NRC, to R.C.Mecredy, RG&E, dated March 11, 1993).The issues discussed in this RAI have already been addressed within the Amendment Request;however, a specific response to each of the six comments and questions is provided below.It should be noted that the responses to the 56 part Question P6 related to UFSAR Table 6.2-15 and the associated figures have not been incorporated to date.The necessary changes will implemented during the next UFSAR update currently scheduled for December of 1993.This is acceptable since the listing of containment isolation valves will be maintained in Ginna Station Procedure A-3.3.Consequently, the update of the UFSAR is not necessary with respect to the subject Technical Specification Amendment Request.RG&E will also perform a detailed review of UFSAR Table 6.2-15 and the associated figures at that time to ensure consistency and completeness as requested in your March ll, 1993 letter.The listing of CIVs contained in A-3.3 has been reviewed to ensure that it is complete.First paragraph of your Safety Evaluation, second sentence, refers to UFSAR Table 6.2-13, should this be referring to Table 6.2-15?The reference to UFSAR Table 6.2-13 was a typographical error.However, the necessary listing of containment isolation valves is now maintained in Ginna Station Procedure A-3.3.Consequently, all references to UFSAR Table 6.2-15 in previously submitted Amendment Requests have been replaced with Procedure A-3.3.2.According to Generic Let ter 91-08,"Removal.of Component Li sts from Technical Specifications (TS)," under the section entitled"Guidance on the Removal of Component Lists from TS," it states in part"...A list of those components must be included in a plant procedure that is subj ect to the change control provisions for plant procedures in the Administrative Controls Section of the TS Although some components may be listed in the Updated Final Safety Analysis Report (UFSAR), the FSAR should not be the sole means to identify these components.

Licensees are only required to update the FSAR annually, and they are only required to reflect changes made 6 months before the date of filing.Thus, the FSAR may be out of date by as much as 18 months...".Your Safety Evaluation does not address what TS controlled procedure covers this list of containment isolation valves.~Res ense The listing of containment isolation valves is now maintained in Ginna Station Procedure A-3.3.This procedure is subject to Technical Specification

6.8 which

requires review by the Ginna Station Plant Operations Review Committee (PORC)and approval by the Plant Manager for any changes.The safety evaluation contained in Attachment B has been updated to reflect this information.

ss a~$~g q a"~'a~al~a~sSdas s aar s~-4's ssl C a

)I I Attachment C Pago 2 of 17 3.'Proposed TS 3.6.3"Containment Isolation Boundaries," items b and~~~~c state: "b.Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of't least one deactivated automatic valve secured in the isolation position, one closed manual valve, or a blind flange, or C~Verify the operability of a closed system for the affected penetrations within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and either restore the inoperable boundary to OPERABLE status or i solate the penetration as provided in 3.6.3.1.b within 30 days, or" The basis for this change is given as"Specification now considers closed systems as an acceptable interim passive boundary and is more consistent with Standard Technical Specification." However, this does not reflect the Standard Technical Specifications (STS)requirement.

STS 3.6.3.C states: '"Isolate the affected penetration flow path by use of at least one closed and de-acti vated automatic valve, closed manual valve, or blind flange.(4-hour completion time)Verify the affected penetration flow path is isolated (once per 31 days)" Therefore, the proposed change to TS 3.6.3.C is not acceptable.

RGGE has"removed-the previously submitted TS 3.6.3.C with respect to the interim use of a closed system as an acceptable boundary for a failed containment isolation valve.TS 3.6.3 is now consistent with Standard Technical Specifications.

4.The term"Isolation Valve" is used in the proposed Bases Section of 4.4 (page 4.4-14), according to the SE, should have been replaced with the term"Isolation Boundary." Res onse: The term"Isolation Valve" is correct for this section of the Bases since most containment leakage observed during testing at Ginna Station and throughout the nuclear industry is through isolation valves and not through passive containment barriers such as blind flanges.Consequently, the bases section was not changed.Proposed TS 3.6.1.a states,"Closed valves may be opened on an intermittent basis under administrative control." Generic Letter 91-08 and your safety evaluation refer to"Locked or Seal Closed containment isolation valves" not j ust"closed valves." Should proposed TS 3.6.1.a be referring to locked or seal closed CIVs?

I%>>'tiV,'L'4 lsk1 I Attachment C Page 3 of 17 Res onse: .The'""locked or"=sealed" closed" terminology"was not" used-in TS 3.6;l.a since several test connections that.may.be-required,to be opened during power operation-for testing.purposes are,.not locked'"'closed at Ginna Station.These valves are-administratively

'maintained closed during power operation per system lineup procedures and have"containment isolation boundary" control tags installed.-This issue is also addressed.in the November 30, 1992 submittal, Attachment D, Item 428.The safety evaluation contained in Attachment B was revised to reflect this information.

6.Comments with regard to R.E.Ginna Updated Final Safety Analysis Report (UFSAR)Table 6.2-15 and Figures 6.2-13 through 6.2-78 are contained on the'ollowing pages.Identified discrepancies associated with proposed UFSAR Table 6.2-15.Valve/Penetration

~Bounder Discre anc 1.105 2829 Position indication in control room is marked"NA" for a manually operated valve.Should this be"No" for consistency7 Res onse: Yes.The position indication in control room column will be.updated to identify"No".for this valve.Valve/Penetration

~Boundar Discre anc 2.105 859A Valve does not.appear on the UFSAR Figure 6.2-18, as i ndi cated by proposed UFSAR Table 6.2-15.3.105 859B Valve does not appear on the UFSAR Figure 6.2-18, as indicated by proposed UFSAR Table 6.2-15.Res onse: UFSAR Figure 6.2-18 will be updated to include valves 859A and 859B.These valves are located on two branch lines between 864A and 859C.

I 1 I V~I Attachmont C Pago 4 of 17 Val ve/Penetration

~Boundar Discre anc 4.105 864A The normal operati ons.position of the valve is listed as"C" (closed)in proposed UFSAR Tabl e 6.2-15, however, it is indicated as"IC" (locked closed)on UFSAR Figure 6.2-18.Res onse: UFSAR Figure 6.2-18 is correct in showing that the valve is normally locked closed.Table 6.2-15 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 5.1 09" 859A Valve-does not appear on the UFSAR Figure 6.2-22, as i ndi cated by proposed UFSAR Table 6.2-15.6.109 859B Valve does not appear on the UFSAR Figure 6.2-22, as indicated by proposed UFSAR Table 6.2-15.Res onse: UFSAR Figure 6.2-22 will be updated to include valves 859A and 859B.These valves are located on two branch lines between 864B and 859C.Val ve/Penetration

~Boundar Discre anc 7.109 864B The normal operations position of the valve is listed as"C" (closed)in proposed UFSAR Tabl e 6.2-15, however, it is indicated as"LC" (locked closed)on UFSAR Figure 6.2-22.Res onse: UFSAR Figure 6.2-22 is correct in showing that the valve is normally locked closed.Table 6.2-15 will be revised to correct this discrepancy.

t~$+Yl: 1~5 Attachment C Page 5 of 17 Val.ve/Penetration

~Boundar Discre anc S.112 9.112 10.112 200A 200B 202 The valve type is li sted as a."Globe" valve in proposed-UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" valve on UFSAR Figure 6.2-25.Al so, proposed UFSAR Tabl e 6.2-15 indicates that this valve trips on CIS, however, this is not noted with a"T" on UFSAR Figure 6.2-25.The valve type is li sted as a"Globe" valve in proposed UFSAR Tabl,e 6.2-15, however, it is indicated as a"Gate" val ve on UFSAR Figure 6.2-25.Al so, proposed UFSAR Tabl e 6.2-15 indicates that this valve trips on CXS, however, this is not noted with a"T" on.UFSAR Figure 6.2-25.The valve type is listed as a"Gl obe" valve i n proposed UFSAR Tabl e 6.2-15, however, i t is indicated as a"Gate" valve on UFSAR Figure 6.2-25..Also, proposed UFSAR Table 6.2-15 indicates that this valve trips on CXS, however, this i s not noted with a"T" on UFSAR Figure 6.2-25.Res onse: Table 6.2-15 correctly identifies all three valves as globe valves which receive a containment isolation signal.Figure 6.2-25 will be revised to correct the discrepancies.

Valve/penetration

~Boundar Discre anc Il.112 371 The valve type is li sted as a"Globe" val.ve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" valve on UFSAR Figure 6.2-25.Res onse: Table 6.2-15 correctly identifies 871 as a globe valve.Figure 6.2-25 will be revised to correct this discrepancy.

~I I y+4"I"43 pt-~k X I'4.,4 I Attachment C Page 6 of 17 Valve/Penetration

~Boundar Discre anc 12.112 13.112 820 204A This valve is indicated on UFSAR Figure 6.2-25, and in the current Technical Specifications as a CIV, however, it is not indicated in proposed UFSAR Table 6.2-15.This valve is indicated on UFSAR Figure 6.2-25, and in the current Technical Specifications as a CIV, however, it is not indicated in proposed UFSAR Table 6.2-15.Res onse: Manual valves 820 and 204A are no longer identified as containment isolation valves in the Ginna Station Technical Specifications

-(see-letter.from A.R.Johnson, NRC, to R.C Mecredy, RGGE,

Subject:

Issuance of Amendment No.52 to Facility Operating Li cense No.DPR-18, dated April 20, 1993).The CIV designations for these valves on UFSAR Figure 6.2-25 will be removed to reflect this change.Valve/Penetration

~Boundar Discre anc 14.123b 9 725 The normal operations position of the valve is listed as"C" (closed)in proposed UFSAR Tabl e 6.2-15, however, it is indicated as"LC" (locked closed)on UFSAR Figure 6.2-26.Res onse: UFSAR Figure 6.2-26 correctly shows 9725 as being normally locked closed.Table 6.2-15 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 15.127 749A The maximumi sol ation time as listed in proposed UFSAR Table 6.2-15 is"NA", however, it is listed in the current Technical Specifications as havi ng a maximum isolation time of 60 seconds.16.128 749B The maximum i sol ation time as listed in proposed UFSAR Table 6.2-15 is"NA", however, it is listed in the current Technical Specifications as having a maximum isolation time of 60 seconds.

I~J Q Wl\~I'l i'.L'b lh Attachment C Page 7 of 17 Res onse: The Technical Specifications contain a typographical error since-'these two.valves do-not-.=receive'..nor, require a.containment

--'"isolation-signal.-Consequently.,-a-..60>>second..maximumisolation time is not applicable.

This issue was addressed in a letter from R.C.Mecredy, RGGE, to A.R.Johnson, NRC,

Subject:

Containment Isolation Valves 745, 749A and 749B, dated July 9, 1990.Valve/Penetration

~Boundar Discre anc 17.143 1 721 Proposed UFSAR Table 6.2-15 indicates that this valve trips on CIS, however, this is not noted with a"T" on UFSAR Figure 6.2-45.~Res onse"Table.6.2-15 correctly identifies 1721 as receiving a containment isolation signal.Figure 6.2-45 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 18.201a NA The system is li sted in proposed UFSAR Table 6.2-15 as"Reactor compartment cooling unit A" and should be li sted as"Reactor compartment cooling unit A supply" for consistency.

Res onse: The system identification for Penetration 201a will be revised to include the word"supply".Valve/Penetration

~Boundar Discre anc 19.201b PI-2141 This instrument is sti ll not indicated in UFSAR Figure 6.2-46 (4 7 J as a CIB, even though you stated in your response to the September 26, 1991, RAI that this item was corrected.

24.209a PI-2140 This instrument i s i ndi cated on UFSAR Figure 6.2-46 (47]as a CIB, however, it is not indicated in proposed UFSAR Table 6.2-15.

I I I 0~'J e Attachment C Page 8 of 17 Res onse: The CIB designation was added to the wrong-pressure indicator on Figure 6.2-47.Consequently, a CIB designation.

will..be.added to-PI--2141 and removed from PI-2140.-Pressure-indicator.,PI-2140 is not a containment isolation valve since it located'pstream of valve 4635 (i.e., not between 4635 and containment).

Val.ve/Penetration

~Boundar Discre anc 20.206b 5 733 This valve is indicated in UFSAR Figure 6.2-54, and in the current Technical Specifications as a CIV, however, it is not indicated in proposed UFSAR Table 6.2-15.21.207b 3B.321 39.322 5734 5 701 5702 This valve is indicated in UFSAR Figure 6.2-56, and in the current Technical Specifications as a CIV, however, it is not indicated in proposed UFSAR Table 6.2-15.This valve is indicated on UFSAR Figure 6.2-71, and in the current Technical Specifications as a CIV, however, it is not indicated in proposed UFSAR Table 6.2-15.Thi s val ve is i ndi cated on UFSAR Figure 6.2-72, and in the current Technical Specifications as a CI V, however, it is not indicated in proposed UFSAR Table 6.2-15.Res onse: Manual valves 5733, 5734, 5701 and 5702 are no longer identified as containment isolation valves in the Ginna Station Technical Specifications (see letter from A.R.Johnson, NRC, to R.C Mecredyg RGEE,

Subject:

Issuance of Amendment No.52 to Facility Operating License No.DPR-IB, dated April 20, 1993).The CIV designations for these valves on UFSAR Figures 6.2-54, 6.2-56, 6.2-71 and.6.2-72 will be removed to reflect this change.Val ve/Penetration

~Boundar Discre anc 22.207b 5 736 The valve type is li sted as a"Globe" valve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" valve in UFSAR Figure 6.2-56.

(I'J y g'~,A~.*p>>}I Eg e*'A II' Attachment C Page 9 of 17 Res onse: Figure 6.2-56 is correct in showing that'736 is a-gate valve.-Table 6.2-15 will-be-revised-to--correct-.this-discrepancy..

-.,--Val.ve/Penetration

~Boundar Discre anc 23.209a NA The system is li sted as"Reactor compartment cooling unit B return" and according to UFSAR Figure 6.2-47 it should be listed as"Reactor compartment cooling unit B supply".Res onse: The system identification for Penetration 209a will be revised to replace"return" with"supply".Valve/penetration

~Bounder Discre anc 25.2095 NA The system is listed as"Reactor compartment cooling unit A supply" and according to VFSAR Figure 6.2-46 it should be listed as"Reactor compartment cooling unit B return".Res onse: The system identification for Penetration 209b will be revised to replace"A supply" with"A return" (not"B return" as suggested).

Valve/Penetration

~Boundar Discre anc 26.210 1 0214S Note 15 is listed in the proposed VFSAR Tabl e 6.2-15 as applicable.

However, note 17 appears to be more appropriate.

In addition, note 17 would make it consistent with valve 10215S.Res onse: Table 6.2-15 will be revised to correct the typographical error and replace note 15 with note 17.

I I I 4, I%l Attachment C Page 10 of 17 Valve/Penetration

~Boundar Discre anc 27.300 5879 This val ve is listed in proposed UFSAR Tabl e 6.2-15, and in the current Technical Specifications a as CIV, however, it is not indicated as a CIV on UFSAR Figure 6.2-58.Res onse: AOV 5879 is not a containment isolation valve.It is only used below cold shutdown conditions to provide containment integrity when the blind flange is removed.See UFSAR Table 6.2-15, Note 29 and Technical Specification Table 3.6-1, Note 22.Valve/Penetrati on~Bounder 28.305a 1556 Discre anc The maximum i sol ation time as listed in proposed UFSAR Table 6.2-15 is"NA", however, it is listed in the current Technical Specifications as having a maximum isolation time of 60 seconds.Res onse: The Technical Specifications contain a typographical error since manual valve 1556 does not receive nor require a containment isolation signal.Consequently, a 60 second maximum isolation time is not applicable.

This is a normally locked closed valve.Val.ve/Penetration

~Boundar Discre anc 29.307 9227 The maximum i sol ation time as listed in proposed UFSAR Table 6.2-15 is 6'0 seconds, however, the current Techni cal Specifications has the maximum isolation time listed as"note 18ne Res onse: A containment isolation signal was installed to AOV 9227 in 1981 under Engineering Work Request No.1833.Subsequent to this modification, the NRC accepted that no containment isolation signal was required for this valve (see letter from D.M.Crutchfield, NRC, to J.E.Maier, RG&E,

Subject:

Containment Isolation, dated May 22, 1982).RG&E has not removed the subject isolation signal.Since AOV 9227 is a containment isolation valve, a 60 second maximum isolation time was added in order to be consistent with other automatic containment isolation valves.

I I g i, I

~~I" I Attachment C Pago 11 of 17 Valve/'30.'308'IA-2010.'his.'nstrument

':is still--not ,.indicated in UFSAR Figure 6.2-65 as a CIB, even though you stated in your response to the September 26I 1991 RAI that this i tem was corrected.

32.311 TIA-2011 This i nstrument is sti l l not indicated in UFSAR Figure 6.2-65 as a CIB, even though you stated in your response to the September 26, 1991 RAI that thi s item was corrected.

34.315 TIA-2012 This.instrument is sti ll not indicated in UFSAR Figure 6.2-65 as a CIB, even though you stated in your response to the September 26, 1991 RAI that this item was corrected.

40.323 TI'A-2013 This instrument is sti ll not indicated in UFSAR Figure 6.2-65 as a CIB, even though you stated in your response to the September 26, 1991 RAI that this item was corrected.

Res onse: The necessary CIB designations will be added to UFSAR Figure 6.2-65 for TIA-2010, TIA-2011, TIA-2012, and TIA-2013.Valve/Penetration

~Boundar Discre anc 31.308 36e 319 NA NA Thi s penetration was indicated as penetration 319 on the current Technical Specifications.

This penetration was indicated as penetration 308 on the current Techni cal Specifications.

Res onse: The valves for penetrations 308 and 319 are reversed in Technical Specification Table 3.6-1.

~~1$1~fpt Attachment C Page 12 of 17 Val ve/Penetration

~Boundary Discre anc 33.313-Blind Flange The Blind Flange-is indicated in UFSAR Figure 6.2-69 as"CIV", should this be"CIB"?Res onse: Figure 6.2-69 will be revised CIB.to replace the CIV designation with Valve/Penetration

~Boundar Discre anc 35.31 7 Blind Flange The Blind Flange is indicated in UFSAR Figure 6.2-70 as"CIV", should this be"CIB"P Res onse: Figure 6.2-70 will be revised CIB.to replace the CIV designation with Valve/Penetration

~Boundar Discre anc 37.320 Res onse: 4641 This valve was indicated as 4647,in the current Technical Specifications.

Valve 4647 is a typographical error in the Technical Specifications.

This drain valve is in series with valve 12500H which is identified on UFSAR Table 6.2-15 as a CIV.The second containment boundary is a CLIC for this penetration.

Valve/Penetration

~Boundar Di sere anc 41.332a 922 The valve type is li sted as a"Gate" valve in proposed UFSAR Table 6.2-15, however, it is indicated as a"Globe" valve in UFSAR Figure 6.2-74.Also proposed UFSAR Table 6.2-15 indicates that'his valve's normal operating position is"C" (closed), however/it is indicated as open in UFSAR Figure 6.2-74.In addition, the maximum isolation time as listed in proposed UFSAR Tabl e 6.2-15 is 3 seconds, however, the current Technical Speci fi cati ons has the maximum isolation time listed as IINA II l P Attachment C Page 13 of 17 42.332a 924 The valve type is li sted as a"Gate" valve in proposed UFSAR Tabl e 6.2-15, however, it is indi cated as a"Globe" valve in UFSAR Figure 6.2-74.Also proposed UFSAR Table 6.2-15 indicates that this valve's normal operati ng position is"C" (closed), however, it is indicated as open in UFSAR Figure 6.2-74.In addition, the maximum isolation time as listed in proposed UFSAR Tabl e 6.2-15 is 3 seconds, however, the current Techni cal Specifications has the maximum isolation time li sted as"NA".43.332b 923 The val ve type is li sted as a"Gate" valve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Globe" valve in UFSAR Figure 6.2-74.Al so proposed UFSAR Table 6.2-15 indicates that this valve's normal operating position is"C" (closed), however, it is indicated as open in UFSAR Figure 6.2-74.In addition, the maximum isolation time as listed in proposed UFSAR Table 6.2-15 is 3 seconds, however, the current Technical Specifications has the maximum i solati on time listed as"NA".44.332d 921 The val ve type is listed as a"Gate" valve in proposed UFSAR Table 6.2-15, however, i t is indicated as a"Globe" valve in UFSAR Figure 6.2-74.Also proposed UFSAR Table 6.2-15 indicates that this val ve's normal operating position is"C" (closed), however, it is indicated as open in UFSAR Figure 6.2-74.In addition, the maximum isolation time as listed in proposed UFSAR Tabl e 6.2-15 is 3 seconds, however, the current Techni cal Specifications has the maximum i solation time listed as"NA".

I'cV4-~w t V 4 I 46 T'/P Attachment C Page 14 of 17 Res onse: Table 6.2-15 is correct in identifying 921, 922, 923, and 924 as gate valves and'in showing.that-'-these valves:are..normally closed.Figure 6.2-74 will-be.revised to-correct--these-discrepancies.

The three second isolation time for these solenoid valves is consistent with Standard Review Plan 6.2.4.II.6.n since these valves are open to containment atmosphere and receive a CIS.Valve/Penetration

~Boundar Discre anc 45.401 3521 The valve type is li sted as a"Gate" valve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"G1 obe" valve in UFSAR Figure 6.2-76.Res onse: Figure 6.2-76 is correct in showing 3521 as a globe valve.Table 6.2-15 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 46.401 PT-469A Instrument is indicated as Inside Containment in proposed UFSAR Table 6.2-15, however, it is indicated as outside containment in UFSAR Figure 6.2-76.Res onse: Figure 6.2-76 is correct in showing PT-469A is located outside containment.

Table 6.2-15 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 4 7.402 3520 The valve type is listed as a"Gate" valve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as.a"Globe" valve in UFSAR Figure 6.2-76.Res onse: Table 6.2-15 is correct in identifying 3520 as a gate valve.Figure 6.2-76 will be revised to correct this discrepancy.

I t C1 k t~1 Attachment C Pago 15 of 17 Valve/Penetration

~Boundar Discre anc 48.403 3995X The valve type is listed as a"Gl obe" val ve in proposed UFSAR Tabl e 6.2-15, however, i t is indicated as a"Gate" valve in UFSAR Figure 6.2-78.Res onse: Figure 6.2-78 is correct in showing 3995X as a gate valve.Table 6.2-15 will be revised to correct this discrepancy.

Valve/Penetration

~Boundar Discre anc 49.403 4011A The valve type is listed as a"Globe" valve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" valve in UFSAR Figure 6.2-78.Res onse: Table 6.2-15 is correct in identifying that 4011A is a globe valve.Figure 6.2-78 will be revised to correct this discrepancy.

Valve/Penetration

~Bounder Discre anc 50.404 3994E The valve type is li sted as a"Gl obe" val ve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" val ve in UFSAR Figure 6.2-78.Res onse: Figure 6.2-78 is correct in showing 3994E as a gate valve.Table 6.2-15 will be revised to correct this discrepancy.

Val,ve/Penetration

~Boundar Discre anc 51.404 4 012A The valve type is listed as a"Gl obe" val ve in proposed UFSAR Tabl e 6.2-15, however, it is indicated as a"Gate" val ve in UFSAR Figure 6.2-78.Res onse: Table 6.2-15 is correct in identifying that 4012A is a globe valve.Figure 6.2-78 will be revised to correct this discrepancy.

I 1 I w*%I'0'g I t Attachment C Page 16 of 17 Location Discre anc 52.Note 17 If this note describes valves that are not CIVs, then to avoid confusion, the note should state that these valves are not CIVs.Res onse: Table 6.2-15 note 17 will be revised to specifically state that the subject valves are not CIVs.Location Discre anc 53.Figure 6.2-13 There is no indication on the figure of where the"CIB" is for either penetration 2 or 29.Res onse: Figure 6.2-13 will be replaced with two separate figures for Penetration 2 and 29.These new figures will identify the location of the CIBs as necessary.

Location Discre anc 54.Fi gure 6.2-65 The"CIB" Cap downstream of 12500H/12500K doesn't show up on the proposed UFSAR Table 6.2-15 for either penetration 320 or 312.The figure does not indicate the association between penetrations and fan coolers.Res onse: The CIB designation is incorrect on Figure 6.2-65 since the CLIC and valves 12500H and 12500K provide the necessary two containment boundaries.

The figure will be revised to delete the CIB designation and provide a relationship between the fan coolers and associated penetrations.

Location Discre anc 55.Figure 6.2-76"CIV" appears on the figure (above CIV 11031 and to the left of valve 3409A)but does not appear to be associated with any particular val ve.Res onse: Figure 6.2-76 will be updated to remove the subject CIV designation.

I 1 I e~0~M 0 Attachment C Page 17 of 17 Location Discre anc 56.There is a lack of consistency for UFSAR Figures'6;2-'13':through

'6.2--78.with--respect to""the'ymbols-used-to--represent"the*directi on of flow through the check valves, and the symbols used to represent air operated valves.In addition, not all figures indicate"CLIC" or"Closed System" where it is applicable.

Res onse: All figures will be reviewed to ensure consistency with respect to air-operated valve designations, check valve flow directions, and the use of closed system indications.

1 f ATTACHMENT D Ginna Station Procedure A-3.3 f C ROCHESTER GAS AND ELECTRIC CORPORATION GINNA STATION CONTROLLED COPY NUMBER OC REV.NO, 1 NTAINMENT INTE RITY PR RAM TE HNI AL REVIEW PORC REVIEW DATE PLANT SUPERINTENDENT EFFECTIVE DATE CATEGORY 1.0 Fo~~<FORMATlOR Oev REVIEWED BY: THIS PROCEDURE CONTAINS~l PAGES

'0 A-3.3:1 NTAINMENT INTE RITV PR RAM 1.0~PPQ$E: To delineate the containment integrity program as required by Technical Specifications 3.6 and 3.8, and Generic Letter 88-17 for conditions above cold shutdown, refueling operations, and reduced inventory conditions, respectively.

2.0 2.1 2.2 2.3 Technical Specifications 3.6 and 3.8.Generic Letter 88-17, Loss of Decay Heat Removal.Updated Final Safety Analysis Report, Section 6.2.4.2.4 Design Analysis DA-NS-93402-21, EWR No.10084, Containment Isolation System Review.2.6 Letter from R.C.Mecredy, RG&E to A.R.Johnson, NRC-

Subject:

AOV-745, MOV-749A and MOV-749B, dated 7/9/90.Inter-Office Correspondence, John Cook and Mark Flaherty to PORC, Subject;Containment Integrity During Refueling, dated 2/20/92.2.7 0-1.1B-Establishing Containment Integrity.

2.&0-2.3.1A-Containment Closure Capability in 2 Hours During RCS Reduced Inventory Operation.

2.9 2.10 2.11 2.12 2.13 PTI'-23 Series.S-30.7, Containment Isolation Valve Verification.

PT-39, Primary System Leakage Evaluation Inservice Inspection.

0-15.2, Required Valve Lineup for Reactor Head Removal.0-15.7, Fuel Handling Instruction Pre-Loading and Periodic Valve Alignment Check.

I P A-3.3:2 3.0 The containment integrity program is designed to provide assurance that the necessary containment isolation boundaries are available for all required plant conditions.

This program is organized to address three plant conditions:

a.Containment Integrity during Refueling.

3.2 b.Containment Integrity during Reduced RCS Inventory.

c.Containment Integrity above Cold Shutdown.The requirements for each of these conditions is discussed below.Containment Integrity during Refueling.

3.2.1 During

plant conditions requiring containment integrity for refueling, each penetration must have a single barrier to the release of radioactive material.This single barrier may consist of any one of the following alternatives:

a.A closed system inside or outside containment such that a"direct access" release path to the outside of containment atmosphere is not provided.b.A closed isolation valve (including check valve with flow secured), blind flange or manual valve.c.An automatic isolation valve that closes on a Containment Ventilation Isolation (CVI)signal from high containment radioactivity.

3.2.2 In addition to the requirements above, Technical Specification

3.8 requires

that"...all automatic containment isolation valves shall be operable or at least one valve in each line shall be locked closed." Since the normal containment isolation signal is not available during the refueling mode of operation, for those penetrations with automatic isolation valves, those valves must be capable of being closed remotely.If those valves are not capable of being closed remotely (i.e.inoperable) thence affected penetration must be isolated by a locked closed manual valve or blind flange.If a manual valve or blind flange is not available, then a held closed auto valve (per A-1401)with motive power removed provides equivalent isolation.

3.2.3 It h not intended that the barriers provided for containment isolation during refueling be restricted to barriers tested to the requirements of Appendix I to 10CFR50.The basis for refueling integrity is to prevent the release of radioactivity resulting from a fuel handling event during refueling operations.

Since there is no potential for containment pressurization, any device which provides an atmospheric pressure boundary is sufficient.

3.2.4 Containment

integrity for refueling is verified through performance of 0-15,2 and 0-15.7.

A-3.3:3 Containment Integrity During Reduced RCS Inventory.

Containment integrity during reduced inventory conditions is provided by maintaining available one barrier for each penetration.

Since there is a potential for containment pressurization during loss of core cooling, this barrier should be one of the two barriers used for normal containment isolation with RCS greater than 200'F.All penetrations are required to be capable of being closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a loss of RHR.This 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame can be extended if the time to reach saturation and core uncovery is increased due to low decay heat levels.3.3.2 Containment integrity during reduced RCS inventory is verified through performance of 0-2.3.1A.3.4 Containment Integrity above Cold Shutdown including normal power operation.

3.4.1 Reference

2.4 provides the design basis for the containment isolation configuration and testing.Any change to this procedure, including Attachment A, must be reviewed by Nuclear Safety and Licensing.

3.4.2 Attachment

A provides a listing for each penetration of the valves and other boundaries required for containment integrity above cold shutdown.These boundaries are leak tested per Appendix J to 10CFR50 except where specific exemptions have been approved.This table is organized as follows: 3.4.2.1~~~5ggm-description of the system which penetrates containment.

3.4.2.2 3.4.2.3 3.4.2.4-unique identification number for the penetration.

-containment isolation valves or boundaries for the penetration.

d i'fh b are available for each penetration.

This is used since many process lines have multiple branch lines prior to entering or exiting containment.

The first character defines the branch line which the containment isolation valve or boundary isolates.The second character defines the isolation barrier which the valve provides (i.e., first or second).As an example, Penetration 107 lists the following containment boundaries:

1723 1728 al a2 AOV 1723 is one containment barrier while AOV 1728 is a second barrier.Above cold shutdown, both valves must be operable and capable of being closed.If AOV 1723 were inoperable, then AOV 1728 is the preferred valve to be closed in accordance with Technical Specification 3.6.3.Conversely, AOV 1723 is the preferred valve to be closed if AOV 172&were inoperable.

I C' A-3.3:4 As an example of penetrations with multiple branch lines, Penetration 124b lists the following containment boundaries:

1572 1573 1574 al a2 a2 Above cold shutdown, all three valves must be operable and capable of being closed.If manual valve 1572 were inoperable, then BOTH manual valves 1573 and 1574 must be closed in accordance with Technical Specification 3.6.3.However, if 1573 were inoperable, only 1572 must be closed (valve 1574 is not affected).

3.4.2.5 3.4.2.6 3.4.2.7~VLvV T~-type of containment isolation valve (e.g., MOV).~-Specific notes related to the containment isolation valve or boundary.-Maximum allowed.closure time in seconds for those valves which receive a containment isolation signal.3.4.3 Prior to heatup above cold shutdown, containment integrity is verified through performance of pr'ocedure 0-1.1B, PIT-23A, PT-39 and S-30.7, Closed systems inside and outside containment are verified through the required system lineups.3.53.5.1 Closed Systems: Closed systems inside and outside containment are used for several penetrations as a containment isolation barrier.The integrity of these closed systems as a barrier is typically confirmed by normal system operation or periodic test.Since these closed systems are exempt from testing per Appendix J to 10CFR50, except as noted below, the allowable leakage (e.g.packing leaks and heat exchanger tube leaks)has been based upon the guidance of ASME/ANSI OMa-1988, OM-10 for the size of isolation valve associated with the closed system.This guidance allows a leakage rate of.5 gpm per inch of nominal valve diameter.3.5.1.1 Service Water System (Penetrations 201a, 201b, 209a, 209b, 308, 311, 312, 315, 316, 319, 320 and 323)-All piping inside containment for these penetrations up to and including the first available isolation valve on all branch lines provide one containment barrier.'Ice integrity of this piping is verified by normal Service Water system operation and containment leakage detection systems.

A-3.3:5 Allowable leakage for the service water systems in containment are as follows: 201a/209 b 209 a/201b 319/308 316/311 320/315 312/323 SW to/from Rx Compartment Cooler A SW to/from Rx Compartment Cooler B SW to/from Fan Cooler A SW to/from Fan Cooler B SW to/from Fan Cooler C SW to/from Fan Cooler D 1.25 gpm 1.25 gpm 4.0 gpm 4.0 gpm 4.0 gpm 4.0 gpm Component Cooling Water System (Penetrations 124a, 124c, 125, 126, 127, 128, 130, and 131)-All piping inside containment for these penetrations up to and including the first available isolation valve on all branch lines provide one containment barrier.The integrity of this piping is verified by normal Component Cooling Water system operation and containment leakage detection systems.The only exception is for penetrations 124a and 124c (Excess Letdown Heat Exchanger cooling)which are normally isolated.Allowable leakage for the component cooling water systems inside containment are as follows:~L~R~124a/c CCW to/from Excess Ltd Hx 1.0 gpm 127/126 CCW to/from RCP A 2.0 gpm 128/125 CCW to/from RCP B 2.0 gpm 131/130 CCW to/from Rx Supt Cooling 3.0 gpm Steam Generator (Penetrations 119, 123b, 206b, 207b, 321, 322, 401, 402, 403, and 404)-The steam generator tubes, shell and all connected piping inside containment for these penetrations up to and including the first available isolation valve on all branch lines provide one containment barrier.The integrity of this piping is verified by normal power operation and containment leakage detection systems.Primary to secondary steam generator tube leakage is limited per Technical~Specification 3.1.5.2 to 0.1 gpm.The allowable leakage for the lines associated with the steam generator closed system are based on the nominal isolation valve size for that line.For main steam and main feedwater lines allowable leakage will be limited to that allowed for the Auxiliary and Standby Feedwater systems.5XKcHl Lu&hh 119 123b 401 402 403 404 206b 207b 321 322 SAFW to SG A SAFW to SG B MS from SG A MS from SGB MFW to SG A MFW to SGB SG A Sample SG B Sample SG A Blowdown SG B Blowdown 1.5 gpm 1.5 gpm 1.5 gpm 1.5 gpm 1.5 gpm 1.5 gpm.375 gpm.375 gpm 1.0 gpm 1.0 gpm I

A-3.3:6 Charging System (Penetrations 100, 102, 106, and 110a)-All piping outside containment from the penetration up to the discharge of the three positive displacement pumps, including the first available isolation valve on all branch lines, provide one containment barrier.The integrity of this piping is verified by normal Charging system operation and operator rounds.The allowable leakage for the lines associated with charging system outside containment is 1.0 gpm.~Pn 100 102 106 110a Charging to RCS Loop B Alt Charging to Loop A RCP A Seal Wtr Inlet RCP B Seal Wtr Inlet 1.0 gpm 1.0 gpm 1.0 gpm 1.0 gpm Safety Injection (Penetrations 101 and 113)-All piping outside containment from check valves 889A/B and 870A/B to the discharge of each Safety Injection pump, including the first available isolation valve on all branch lines, provide one containment barrier.The integrity of this piping is verified by system lineups and by the monthly and quarterly pump tests.The allowable leakage for the safety injection system is specified in PT-39.Containment Spray (Penetrations 105 and 109)-All piping outside containment from check valves 862A/B to MOVs 860A/B/C/D, including the first available isolation valve on all branch lines, provide one containment barrier.The integrity of this piping is verified by system lineup and by the monthly and quarterly pump tests.The allowable leakage for the containment spray system is specified in PT-39.Residual Heat Removal (Penetrations 111, 140, 141, and 142)-All piping outside containment including the first available isolation valve on all branch lines provide one containment barrier.The integrity of this piping is verified by monthly and quarterly pump tests and by normal system operation during shutdown.The allowable leakage for the residual heat removal system is specified in PT-39.Hydrogen Monitoring System (Penetrations 332a, 332b, and 332d)-All piping outside containment including the first available isolation valve on all branch lines provide one containment barrier.The integrity of this piping is verified by annual 10CFRSO Appendix J testing.Charging System-Seal Water Return (penetration 108)-All piping outside containment from MOV-313 to the VCT, including the first available isolation valve on all branch lines, provides one barrier.The integrity of this piping is verified by normal system operation and operator rounds.The allowable leakage for the seal water return lines outside containment is 1.5 gpm.MRCQIRK: None.

ATl'ACHMENT A A-3.3:7~astern Steam Generator Inspection/

Maintenance Fuel Transfer Tube Charging Line to Loop B Safety Injection Pump B Discharge Alternate Charging to Cold Leg A 29 100 101 102 Valval~BNI dI NA NA SAC05 8152 8153 370B CLOG 870B 889B CLOC 12407 PI-923A PT-923 885B 383B CLOG isolation Position al a2 al, a2 a2 a2 al a2 al al a2 bl bl bl b2 al a2 Valve~2e Blind Flange Blind Flange Blind Flange Manual Manual Check NA Check Check NA Manual NA NA Manual Check NA Notes Maximum isolation Yimo~sacs.Construction Fire Service Water 103 NA 5129 al Welded Cap a2 Manual 9 Containment Spray Pump A 105 862A CLOC 2829 869A 2856 2825 2825A 864A 859A 859B 859C al a2 NA bl b2 cl C2 dl d2 d2 d2 Check NA Manual Manual Manual Manual Manual Manual Manual Manual Manual 10 2 6, 13 6, 13 6 12 12 12 Reactor Coolant Pump A Seal Water Inlet Sump A Discharge to Waste Holdup Tank Reactor Coolant Pump Seal Water Return Line and Excess Letdown to VCT 106 107 108 304A CLOG 1723 1728 313 CLOG al a2 al a2 al a2 Check NA AOV AOV MOV NA 14 60 60 60 A ITACHMENT A A-3.3:8~Ss~te Containment Spray Pump B Reactor Coolant Pump B Seal Water Inlet Safety Injection Test Line Residual Heat Removal to Cold Leg B Letdown to Nonregenerative Heat Exchanger Safety Injection Pump A Discharge Standby Auxil-iary Feedwater Line to Steam Generator A Nitrogen to Accumulators Pressurizer Relief Tank to Gas Analyzer Penetration No.109 110a 110b 112 113 119 120a 120b Valvai~Blv all 862B CLOG 2830 869B 2858 2826 2826A 864B 859A 859B 859C 304B CLOG 879 720 2840 2847 2848 2853 959 CLOC 371 200A 200B 202 203 CLOG 371 427'70A 889A CLOG 12406 PI-922A PT-922 Cap(PT-922) 885A 9704A 9723 CLIC 846 8623 539 546 iaolation~aition al a2 NA bl b2 cl c2 dl d2 d2 d2 al a2 al,a2 al al al al al a2 a2 a2 al al al al al a2 NA al al a2 bl bl bl bl b2 al al a2 al a2 al a2 Valve~T Check NA Manual Manual Manual Manual Manual Manual Manual Manual Manual Check NA Manual MOV Manual Manual Manual Manual AOV NA AOV AOV AOV AOV Relief NA AOV AOV Check Check NA Manual NA NA NA Manual MOV Manual NA AOV Check AOV Manual~ates 10 2 6, 13 6, 13 6 12 12 12 15 17 6 6 6 6 35 16 36 16 36 11 18 Maximum iaoiation Tima 60 60 60 60 60 60 60

ATI'ACHMENT A A-3.3:9~sstem Makeup water to Pressurizer Relief Tank Nitrogen to Pressurizer Relief Tank Containment Pressure Transmitter PT945 and PT946 Reactor Coolant Drai.n Tank to Gas Analyzer Line Standby Auxil-iary Feedwater Line to Steam Generator B Excess Letdown Heat Exchanger Cooling Water Supply Post Accident Ai.r Sample to Common Return Excess Letdown Heat Exchanger Cooling Water Return Post Accident Ai.r Sample to Fan C Component Cooling Water from Reactor Coolant Pump B Component Cooling Water from Reactor Coolant Pump A Component Cooling Water to Reactor Coolant Pump A Component Cooling Water to Reactor Coolant Pump B Pcncttation 12la 121b 121c 123a 123b 124a 124b 124c 124d 125 126 127 128 Valve/~B 508 529 528 547 PT945 1819A PT946 1819B 1600A 1655 1789 9704B 9725 9724 CLIC 743 CLIC 1572 1573 1574 745 CLIC 1569 1570 1571 759B CLIC 759A CLIC 749A 750A CLIC 749B 750B CLIC bohtion Posiuon al a2 al a2 al a2 bl b2 NA al a2 al al al a2 al a2 al a2 a2 al a2 al a2 a2 al a2 al a2 al a2 a2 al a2 a2 Valve~TQB AOV Check Check Manual NA Manual NA Manual SOV Manual AOV MOV Manual Manual NA Check NA Manual Manual Manual AOV NA Manual Manual Manual MOV NA MOV NA MOV Check NA MOV Check NA Notes 6 18 19 20,37 19 19 19 37 30 19 37 30 19 Maximum Isolation Time~ceca.60 60 A%I'ACHMENT A A-3.3:10~astern Reactor Coolant Drain Tank and Pressurizer Relief Tank to Containment Vent Header Component Cooling Water from Reactor Support Cooling Component Cooling Water to Reactor Support Cooling Containment Mini,-Purge Exhaust Residual Heat Removal Pump suction from Hot Leg A Residual Heat Removal Pump A Suction from Sump B Residual Heat Removal Pump B Suction from Sump B Reactor Coolant Drain Tank Discharge Line Reactor Compartment Cooling Unit A Supply Reactor Compartment Cooling Unit B Return Hydrogen Recombiner B (Pilot)Penettation 129 130 131 132 140 141 142 143 201a 201b 202a Valve/~80UNI 1713 1793 1786 1787 814 CLIC 813 CLIC 7970 7971 Cap 701 2763 2786 CLOG 850A CLOG 851A 1813A 850B CLOG 851B 1813B 1003A 1003B 1709G 1722 1721 4757 4775 CLIC 4636 4658 4776 PI-2141 coats 2lal)CLIC 1076B 1021181 isolation Position ai a2 bl b2 al a2 al a2 a1 a2 a2 al al al a2 al a2 a2 bl,b2 al a2 a2 bl,b2 al al al al a2 al al a2 al al al al al a2 al a2 Valve~Te Check Manual AOV AOV MOV NA MOV NA AOV AOV NA MOV Manual Manual NA MOV NA MOV MOV MOV NA MOV MOV AOV AOV Manual Manual AOV Manual Manual NA Manual NA Manual NA NA NA Manual SOV Notes 19 19 29 17 6 6 16 21 16 30 32 21 16 30 32 23 28 22 28 Maximum isolation Time~i 60 60 60 60 60 60 60 ATTACHMENT A A-3.3:11~Ss~te Hydrogen Recombiner B (Main)Containment Pressure Transmitter PT947 and PT948 Post Accident Air Sample from Fan D Post Accident Air Sample from Common Header Penetration No.202b 203a 203b 203c Valve/~Sound 1084B 1021381 PT947 1819C PT948 1819D 1563 1564 1565 1566 1567 1568 iaolation Poat>on al a2 al a2 bl b2 al a2 a2 al a2 a2 Valve~ates Manual SOV NA Manual NA Manual Manual Manual Manual Manual Manual Manual Maximum iaolation Time~scca.Purge Supply Duct Hot Leg Loop B Sample Pressurizer Liquid Space Sample 204 205 206a ACD93 5869 955 956D 966C 953 956E 966B al, a2 NA NA al a2 NA al a2 Blind Flange AOV AOV Manual AOV AOV Manual AOV 25 60 60 Steam Generator 206b A Sample CLIC 5735 5749 al a2 a2 NA AOV Manual 18 60 Pressurizer Steam Space Sample 207a 951 956F 966A NA al a2 AOV Manual AOV 60 Steam Generator 207b B Sample CLIO 5736 5754 al a2 a2 NA AOV Manual 18 60 Reactor Compartment Cooling Unit B Supply Reactor Compartment Cooling Unit A Return Oxygen Makeup to Recombiners A 6 B 209a 209b 210 4635 4637 CLIC 4638 4758 4759 PI-2232 CLIO 1080A 1021481 10214S 1021581 102158 al al a2 al al al al al a2 al a2 NA a2 NA Manual Manual NA Manual Manual Relief NA NA NA Manual SOV SOV SOV SOV 23 28 22 28 ll ll ATI'ACHMENT A A-3.3:12~Sstem Purge Exhaust Duct Auxiliary Steam Supply to Containment Auxiliary Steam Condensate Return Hydrogen Recombiner A (Pilot)Hydrogen Recombiner A (Main)Containment Air Sample Post Accident Containment Air Sample Inlet Contai.nment Air Sample Post Accident Containment Air Sample Post Accident Containment Air Sample Out Fire Service Water Servi.ce Water from Fan Cooler A Mini-Purge SuPPlY Instrument Air to Containment Service Air to Contai.nment Penetration

<<o.300 301 303 304a 304b 305a 305b 305C 305D 305E 307 308 309 310a 310b Valve/~~eeaauU~ACD92 5879 6151 6165 6152 6175 1076A 1020581 1084A 1020981 1554 1555 1556 1598 1599 1557 1558 1559 1560 1561 1562 1596 1597 9227 9229 4629 4633 4655 FIA-2033 CeaeQXFIA.%33)

TIA-2010 CLIC 7445 7478 5392 5393 7141 7226 boiation~Posit on al, a2 NA al a2 al a2 al a2 al a2 al a2 a2 al a2 al a2 a2 al a2 a2 al a2 al a2 al al al al al al a2 al a2 al a2 al a2 Valve~Tp~Blind Flange AOV Manual Manual Manual Manual Manual SOV Manual SOV Manual Manual Manual AOV AOV Manual Manual Manual Manual Manual Manual Manual AOV AOV Check Manual Manual Relief NA NA NA NA AOV AOV AOV Check Manual Check otes 25 22 28 Maximum isolation Time~s.60 60 60 60 60

ATl ACHMENT A A-3.3:13~astern Service Water from Fan Cooler B Service Water to Fan Cooler D Leakage Test Depressuriza-tion Penetration N.311 312 313 Valve/~BNlee 4630 4634 4656 FIA-2034 TZA-2011 CLZC 4642 4646 12500K PI-2144 CLZC NA Cap 7444 bobtion Position al al al al al al a2 al al al al a2 al a2 a2 Valve~pe Manual Manual Relief NA NA NA NA Manual Manual Manual NA NA Blind Flange NA MOV Notes 22 28 23 28 26 Maximum boiation Time Service Water From Fan Cooler C Service Water to Fan Cooler B Leakage Test Supply Deadweight Tester Service Water To Fan Cooler A Service Water to Fan Cooler C 315 316 317 318 319 320 4643 4647 4659 FZA-2035 CstmCXFlh.xtLt)

TIA-2012 CLIC 4628 4632 PI-2138 CLIC SAT01 Cap 7443 NA 4627 4631 PI-2142 CLIC 4641 4645 12500H PZ-2136 CLZC al al al al al al a2 al al al a2 al a2 a2 al, a2 al al al a2 al al al a1 a2 Manual Manual Relief NA NA NA NA Manual Manual NA NA Blind Flange NA MOV NA Manual Manual NA NA Manual Manual Manual Nh NA 22 28 23 28 26 27 23 28 23 28 Steam Generator 321 A Blowdown Steam Generator 322 B Blowdown 5738 5752 CLIC 5737 5756 CLZC al al a2 al al a2 AOV Manual NA AOV Manual NA 18 18 60 60 I

A%I'ACHMENT A A-3.3:14~astern Service Water from Fan Cooler D Demineralized Water to Containment Hydrogen Monitor Instrumentation Line Hydrogen Monitor Instrumentation Line Containment Pressure Transmitters PT944, PT949, and PT950 Penetration 323 324 332a 332b 332c Valvci l~~4644 4648 4660 FIA-2036 Ceca Ot(FIA 3t3at TIA-2013 CLIC 8418 8419 922 924 CLOG 7452 Cap@452)923 CLOC 7456 Capp456)PT944 1819G PT949 1819E PT950 1819F al al al al al al a2 al a2 al al a2 bl b2 al a2 bl b2 al a2 bl b2 cl c2 Valve~22+Manual Manual Relief NA NA NA NA AOV Check SOV SOV NA Manual NA SOV NA Manual NA NA Manual NA Manual NA Manual Notes 22 28 31 31 Maximum bolation Time~a.Hydrogen Monitor Instrumentation Line Main Steam from Steam Generator A 332d 401 921 CLOC 7448 Cap(7448)3411 3413A 3455 3505A 3505C 3509 3511 3513 3515 3517 3521 3615 3669 11027 11029 11031 PS-2092 PT-468 PT-469 PT-469A PT-482 End Caps CLIC al a2 bl b2 al al al al al al al al al al al al al al al al al al al al al al a2 SOV NA Manual NA Relief Manual Manual MOV Manual Relief Relief Relief Relief AOV Manual Manual Manual Manual Manual Manual NA NA NA NA NA NA Nh 31 24 24 24 24 8 8 8 8 8 33 18 ATl'ACHMENT A A-3.3:15~Sstem Penetration Valve/~Bo nda hoiation Position Valve~Te Notes Maximum Solatioa Time~secs.Main Steam from 402 B Steam Generator 3410 3412A 3456 3504A 3504C 3508 3510 3512 3514 3516 3520 3614 3668 11021 11023 11025 PS-2093 PT-478 PT-479 PT-483 End caps CLZC al al al al'al al al al al al al al al al al al al al al al al a2 Relief Manual Manual MOV Manual Relief Relief Reli.ef Reli.ef AOV Manual Manual Manual Manual Manual Manual NA NA NA NA NA NA 24 24 24 24 8 8 8 8 33 18 Feedwater Line to Steam Generator A Feedwater Line to Steam Generator B 403 404 3993 3995X 4000C 4003 4003A 4011A 4099E 8651 CLIC 3992 3994E 3994X 4000D 4004 4012A 4004A 8650 CLZC al al al al al al al al a2 al al al al al al al al a2 Check Manual Check Check Manual Manual Manual Manual NA Check Manual Manual Check Check Manual Manual Manual NA 34 34 34 18 34 34 34 18 Personnel Hatch 1000 Equipment Hatch 2000 NA NA NA NA al a2 al a2 NA NA NA NA ATTACHMENT A A-3.3:16 (2)(3)~ates This penetration is closed by a double-gasketed blind flange on both ends.Both flanges are necessary for containment integrity purposes since the test connections between the two gaskets for each flange do not meet the requirements of ANSI-56.8.

Therefore, the innermost gasket for each flange (i.e., gasket closest to containment wall)provides a single containment barrier.This valve is not a containment isolation valve due to the installed downstream welded flange, but is normally maintained locked closed to provide additional assurance of containment integrity.

The end of the fuel transfer tube inside containment is closed by a double-gasketed blind flange to prevent leakage of spent fuel pit water into the containment during plant operation.

Each gasket provides a single containment isolation barrier.This flange also serves as protection against leakage from the containment following a loss-of-coolant accident.(4)(5)(6)(8)The charging system is a closed system outside containment (CLOG).Verification of this closed system as a containment isolation boundary is accomplished via normal system operation (>>2235 psig).The safety infection system is a closed system outside containment (CLOG).Verification of this closed system as a containment isolation boundary is accomplished via inservice and/or shutdown leakage checks.(Safety In)ection Pump discharge pressure is~1500 psig)This valve is not locked closed~however, the valve is maintained closed by testing and system lineup procedures and has a"Boundary Control Tag" per PTT-23A.This provides equivalent assurance of proper valve position.The pressure indicator only provides local indication; therefore, a second closed isolation device is required (i.e., indicator's root valve).However, the root valve (12406 or 12407)is listed with the indicator, not as a second barrier due to the design of the line.The pressure transmitter assembly, by its design, provides a containment pressure boundary.Since the transmitter provides direct indication to the control room, operators would be aware of its failure.Therefore, the transmitter's root valve(s)is normally maintained open.(9)This penetration was only utilized during initial plant construction and is maintained inactive.Since there is no test connection between 5129 and the threaded cap, all observed leakage during testing is applied to 5129.Therefore, the outside cap is not a CIB.(10)The containment spray 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.(Containment Spray pump discharge pressure is~285 psig)This valve receives a containment isolation signalg however, credit is not taken for this function since the valve is inside the missile barrier or outside the necessary class break boundary.Therefore, this valve is not a containment isolation valve and not subject to 10 CFR 50 ATI'ACHMENT A A-3.3:17 Appendix J testing nor Technical Specification 3.6.3.isolation signal only enhances containment isolation.

The containment (12)(13)Both containment spray test lines have a locked closed manual valve that leads to a common line with two normally closed manual valves.The valves in this common line may be opened during a pump test since necessary containment isolation is maintained (see Safety Evaluation NSL-OOOO-SE015).

The test line and root valves for the pressure indicators can be opened during testing of the CS pumps since manual valves 868 A/B are closed, thus providing the necessary containment boundary for the short duration of the test.(14)The second isolation barrier (CLOC)is.provided by the volume control tank and connecting piping per letter from D.D.DiIanni, NRC, to R.W.Kober, RG&E, dated January 30, 1987.This barrier is not required to be tested.(15)(16)(17)(18)(19)(2o)(21)(22)Only one isolation barrier is provided since there are two Event V check valves in the SI cold legs, and two check valves and a normally closed motor-operated valve in the SI hot legs.This configuration was accepted by the NRC during the SEP (NUREG-0821, Section 4.22.2).The residual heat removal lines for this penetration are a closed loop outside containment (CLOG).Verification of this closed system as a containment isolation boundary is accomplished via inservice and/or shutdown leakage checks.(Residual Heat Removal pump discharge pressure is~175 psig)Appendix J containment leakage testing is not required per letter from D.M.Crutchfield, NRC, to J.E.Maler, RGGE, dated May 6, 1981.The Main Steam, Main Feedwater, Standby Auxiliary Feedwater and S/G Blowdown penetrations take credit for the steam generator tubes and shell as a closed system inside containment (CLIC).Verification of this closed system as a containment isolation boundary is accomplished via normal power operation (750 psig).The isolation valves outside containment for these penetrations do not require Appendix J testing.The component cooling water lines inside containment for this penetration are a closed loop inside containment (CLIC).Verification of this closed system as a containment isolation boundary is accomplished via inservice and/or shutdown leakage checks.(Component Cooling Water pump discharge pressure is~85 psig)Operations is instructed to manually close AOV 745 following a containment isolation signal to provide additional redundancy.

Sump lines are in operation and filled with fluid following an accident;therefore, 10CFR50, Appendix J leakage testing is not required for this penetration.

See letter from D.M.Crutchfield, NRC, to J.E.Maier, RGM, dated May 6, 1981.This manual valve is sub)ected to an annual hydrostatic leakage test (>60 psig)and is not sub)ect to 10CFR50, Appendix J leakage testing.See NUREG-0821, Section 4.22.3.

ATI'ACHMENT A A-3.3:18 (23)(24)(25)The Service Water System operates at a higher pressure (80 psig)than the containment accident pressure (60 psig)and is missile protected inside containment.

Therefore, this manual valve is used for flow control only and is not subject to 10CFR50, Appendix J leakage testing.See NUREG-0821, Section 4.22.3.h This valve does not receive an automatic containment isolation signal but is normally open at power since it either improves the reliability of an essential standby system or is required for power operation.

However, this valve can either be closed from the control room or locally when required.The flanges and associated double seals provide containment isolation and ensure that containment integrity is maintained for all modes of operation above cold shutdown.When'the flanges are removed during cold shutdown conditions, containment integrity is provided by the valve.This valve is not required to bo operable above cold shutdown and does not require 10CFR50, Appendix J leakage testing, nor a maximum isolation time.(26)(29)(30)Motor<<Operated Valves 7443 and 7444 are powered from non-safety-related Bus 15.However, this is acceptable since the valves are maintained closed at power and are in series with a blind flange.In addition, operators would be aware of a loss of Bus 15 by a loss of control room indication for these two valves (Safety Evaluation NSL-OOOO-SE021).

This penetration is decommissioned and welded shut.The service water system piping inside containment for this penetration is a closed system inside containment (CLIC).Verification of this closed system as a containment isolation boundary is accomplished via inservice and/or shutdown leakage checks.(Service Water Pump discharge pressure is~80 psig)This end cap is used for flow balancing.

However, it cannot be opened above cold shutdown without first performing a safety evaluation.

This valve will no longer be classified as a CIV following NRC approval of the Amendment Request to remove the listing of CIVs from Technical Specifications since another boundary has been identified.

However, in the interim, the valve will continue to be identified and tested as a CIV consistent with Technical Specifications.

This note applies to valves 750A, 750B, 851A and 851B.(31)(32)(33)(34)Acceptable isolation capability is provided for these instrument lines by two isolation boundaries outside containment.

One of the boundaries is a Seismic Category I closed system which is subject to Type C leakage testing under 10 CFR 50 Appendix J.There is no second containment barrier for this branch line.This is addressed by Safety Evaluation NSL-OOOO-SE015.

These end caps include those found on the sensing lines for PS-2092, PT-468, PT-469, PT-469A, and PT-482 (Penetration 401)and PS-2093, PT-479, and PT-483 (Penetration 402).This check valve can be open when containment isolation is required in order to provide necessary feedwater or auxiliary feedwater to the steam ATI'ACHMENT A A-3.3:19 (35)generators.

The check valve will close once feedwater is isolated to the affected steam generator (NUREG-0821, Section 4.22.1).AOV 959 cannot be tested to 10 CFR 50 Appendix J requirements since there are no available test connections.

Therefore, the fuses for AOV 959 are removed with boundary control tags in place to maintain this valve closed.Manual valve 957 is also maintained closed to provide additional assurance of containment lntegrltyy however, valve 957 is not a containment isolation valve sub)ect to Technical Specification 3.6.3.(36)AOV 371 is a containment isolation valve for both penetrations ill and 112.(37)The Technical Specifications currently identify a 60 second maximum isolation signal for this valve (745, 749A and 749B).However, there is no automatic containment isolation signal to this valve and none required.

ATTACHMENT E Table of Technical Specification Changes Pg Attachment P.Page 1 of 3 Changes Technical Specification Changes Effect 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 Procedure A-3.3.Removed Table 3.6-1 from Technical Specifications and placed information in Procedure A-3.3.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.Revised action statement of Technical Specification

3.8.1 section

a.No technical change.Specifications are now consistent with Generic Letter 91-08.Valve listing remains in a licensee controlled document under Technical Specification change controls.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.Clarification only.Specification now consistent with Standard Technical Specifications.

I I l Mf 1~I v: II q"('~~ec II a~I~*

Attachment E Page 2 of 3'Changes Technical Specification Changes Effect 10.12.'13.14.15.'6.'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).Revised acceptance criteria provided in Technical Specification 4.4.2.2 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'ir locks.Minor clarification only.No technical change.Clarification only.No technical change.

t>>I'i~'I y g'a mL 4 4-Attachment E Page 3 of 3 Changes Technical Specification Changes Effect 17.18.19.20.21.22.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.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.

f I I~i4~V r I~%lt II I p~4

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.pg;"',pl'ossa)yi1je's,.';.':~~'imp'he

4'col'3.'.n4 svx8,'5x'v8:i<,::::const',03+!~

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 l

3.6.3 Containment

Isolation-Vakvee.:4'oGFdai:i~e'8 3.6.3.1 With epe~nd~!afjccint'ainus',:;i:,:@platinum'houndarg';::a ppe~~SIe,;::;..';for one..:.ex

'::,miieIj'.co%tegn'meie$

~j4rii:,:;-::,:,8

',::,-:::,:h'

-..6-::y e~~keFOPERABX:8 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, on,::-:,':ll:::c'las'el n'a'jn'u'a~j@Lue~",;!or'.:g.;,:;;:Jjgggg',::;'g'jl'ap~g'e." or de.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,hours,.

3.6.4 Combustible

Gas Control 3.6.4.1 When the reactor is critical, at least two independent containment hydrogen monitors shall be operable.One of the monitors may be the Post Accident Sampling System.3.6.4.2 With only one hydrogen monitor operable, restore a second monitor to operable status within 30 days or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.3.6.4.3 With no hydrogen monitors operable, restore at least one monitor to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be at least hot shutdown within'the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.3.6e5 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

'I I h'T C.1 ll.a4'I I+'L'+4 0 fl l'1$o i'.~'~'"~,~~r'lf.<II Y Ig~I l I%J 6~+44lllh F I h jp\+<~$

Basis: The reactor coolant system conditions of cold shutdown assure that'o steam will be formed and hence-there-would be no pressure buildup in the containment if the reactor coolant system ruptures.'he-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.Amendment No.3.6-3 Proposed Jp,>4'cacti r I.>Af>>r~II r)

References:

(1)Westinghouse Analysis,"Report for the BAST Concentration Reduction f or R.E.Ginna", August 198 5Pj~i~ip55'Xt;pppg~~N~.

8: I'i'i""':':fioiii:":

R: N"':>'Kobi""""'RGB'"'

(2)UFSAR-Section 6.2.l.4]f3'ggj~GPSA'Rq:.'::.-,, e8'actin'n;::~6,".!2~~4:

3.6-4 Proposed

'it~II, h~A

'Ie\gal b.c~a:::~ii-:::;:,-p Et'::,:~:~':-pic-:: '::;i:,:.:!8!'::.:::,:, KiihogaSi~cs!:,zguudovnj:::,ipux'ga'<ll,a:i'iiiiiimLii il:;,,pii~7ja yves'l~v8.:".i 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, g.g 3.8-1 Proposed I~~

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.l.d are not met, then in addition to the requirements of 3.8.2, pi~MMKCC44Xw.'w'i&+5 55e~sh~g...6own~pqx'cge';:;and::ljqi,:ni:::,:.,'.p'urge;;..:penetiat,,:io'ns 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 PFSAR.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 I r~I kya~/rL~*ad 4"~*a~I d~Q g g C C 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 reanalysis~~~~

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

."--h'1!hlly.

-,-,, 16"-,':is%res:::::":"::::::::,:4:,'::::ilia:".::,,::ni::::.::;

ilia::: t"-Ph-:: ilia%i!!ah: Out81'd'eimahtmCiajihere" establishes additional margin f or the fuel handling accident and establishes a seismic envelope to protect ,,:::;::s.,ii"i:::-':k:::"'\l!":-:i:,".I

~i!i"::-:,'-:i'!]i refueling Wax:.sole!toin~a I,jttTi'ie'j:,j efng aritioen's'Cmaliiibe:':

ii--i'---,::-s,,:::, e::.,:-:.::~;-::::::-g-;pd':--i'i~)::-:,""'g,",::::::ii.

-:"""-:-,-:,'i~

":.":-;:::.',":"K",::ii'!".'.

aFiiipherei...,,c;it

'ej'outsi'de,;atmc sybil', e jsyu'oit'.".",'

on knai~t,:::;::~

'it i r'iMi-: ':.',: ieihrCh::;:!!Oan::

Zi!p rcnu'B aiei';:,::a::;,:::;:::temp Origay::::::::::::y,~t'-: li,,:,--,-,e:;:--

.,",:-,:::ii!)i!,:.'!imari., ml:--:-, E)!4!.",:ii:::: T:-":.--'!!4 il::-,I ii---J,.pc,veLentgl Amendment No.g 3.8-5 Proposed I I QJ 4'I I, 4>A t I F References (1)'2)(3)Re~~,':,Ug@Wg4@ct',::j:ovals>gbYg~.':::.:4~and;",:;;9.'.g.~@-:.'8 load Transient Safety Report, Cycle 14-:!UF BAR::","!SPP&Tp fli'i'5!~gi'3!'i!~3:

3.8-6 Proposed 1 4, 4'I, l I 4*tf iI I f I\

Acce tance Criteria~"..s.-~~~"~!.-g.'i!j,;."'i',',,',";:P'S,,"%',8,'3l, S,",'ll!RS!ii,,!il':C:Oll ct'31lhlSll

.,:"!~!'VB'8'88' pgp/a b.Lt shall be determined as Lt=La~>>~VMeh(~egup~f~Q QU~81~8%2!..::p'8x::,c5'At,::::v8'xgÃit~!$.ex.'::>'Aay8~~eajI.A9'~.';::ra~e.ski'=:,-".:-,pi::~assur'e:

4a~~j~Test Fre uenc a~A set of three integrated leak rate tests shall be performed at approximately equal intervals during each 10-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.~~lie following-eaeP;ea'c8 in-service inspection, the containment airlockj>"..:,,'gath~ePj>:"':.>jpy5jj'ii) y::e.:n e:r,:a.:t;ale"':::5;-:;:gg;:,;,:,i'.,:",.n:s,.p,:;e~c':,':!i~a";,'n'>jjm'::a':;:-'i..-':n"t.,':i'nYaiiic.',e l~eak tested prior Wo returning the plant to operation, and 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 I I I S I I j 4 L 0 I

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

~~ll~ill.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.iv>>Isolation valves on the testable fluid systems v~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 p,CwA'Mt~!uxsaN>ss>>spg>>>>

wAl~gt&san,sas&~dANx>>!>>>>s>>esi a@a!p, i'noperab'li

',i,::".':if rlo!mj!!!a",;i!!ieaki'gal!i~>>>scan'dgoinC~>>iwhe'n,,:gha dem'oniti'."a'tk'd~fieaga'j~e",<or!ira!L:;::sanglijijb'oun,dayr:

oai~)ga'umui rCa'iy'e 4.4.2>>3 Corrective Action a~If at any time it is determined that the total leakage from all penetrations and isolation valves pcun'd'ariaS exceeds 0.60 La, repairs shall be initiated immediately.

4.4-6 Proposed I ,1 I a l FJ+

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~Except as specified in b.-, and.;)c., 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.-Xa b.The containment equipment hatch, fuel transfer Ipiiitdatx'oa, 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 1 I I)I 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.Zn 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.4.4-8 Proposed

\P, I~!

Amendment No.4.4-8 Proposed IO'A'h the tendon containing 6 broken wires)shall be inspected.

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

4.4.4.2 Pre-Stress Confirmation Test a~b.Lift-off tests shall be performed on the 14 tendons identif ied in 4.4.4.1a 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 (6 4)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 contiiame'ntg>:isolation valve b" I6::,i(i:i::1gj.i accordance with the Ginna Station Pump anda 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 peSp'Ofi's'@~time of pi'ehj~e containment isolation valve ,', shall be demonstrated to be within Cheggts limit at least once per 18 months.The response time includes only the valve va'1Vee'~+giehiithejaa'f aVy'::;";aaa,:

/Amendment No.4.4-11 Proposed

'i CI~k 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 th'e 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 O.l>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

.p a II'L 0$rt The basis for specification of a total leakage of 0.60 La from penetrations and isolation~ee~SFugŽdaige8 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 V'J t CV~p 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 vakvee~hnund'ix'fi'8 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 covere by e ump an Va ve Tes 5'rogram.Comp iance wi Appendix J to 10 CFR 50 is addressed under local leak testing requirements.

References:

(2)(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 gQ (ft I f C v 1 e.)t'I I~~e)~%gbqgg I Q