ML17263A390

Amend 54 to License DPR-18,removing Containment Isolation Valve Table 3.6-1 from TS

ML17263A390
Person / Time
Site:	Ginna
Issue date:	08/30/1993
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17263A391	List: ML17263A390 ML17263A392
References
NUDOCS 9309140045
Download: ML17263A390 (25)

Text

pe RK0I (4

~4 0

Cy P

0O C

/y gO

+w*w+

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 2055&0001 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.

E.

GINNA NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No.

DPR-18 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Rochester Gas and Electric Corporation (the licensee) dated October 15,

1990, as supplemented March 8,

1991, November 30,

1992, and July 13,

1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) 'of Facility Operating License No.

DPR-18 is hereby amended to read as follows:

9309140045 930830

(

.PDR ADOCK 05000244 P

PDR

I (2).

Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

54

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance August 30, 1993 Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE'MENDME T NO.

54 FACILITY OPERATING LICENSE NO.

DPR-18 DOCKET NO. 50-244 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

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 4.4-6 4.4-7 4.4-8 4.4-11 4.4-13 4.4-14 4.4-17

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.

Closed valves may be opened on an intermittent basis under administrative control.

b.

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

c ~

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

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

Amendment No.

45

~54 3.6-1

3.6.3 3.6.3.1 Containment Isolation Boundaries With a containment isolation boundary inoperable for one or more containment penetrations, either:

'a ~

b.

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 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

Basis:

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

The shutdown margins are selected based on the type of activities that are being carried out.

The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.

When the reactor head is not to be removed, a cold shutdown margin of 14,k/k precludes criticality in any occurrence.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig.t'>

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 not affected by a

single active failure.

Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.

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

Amendment No.

45

~

54 3.6-3

References:

Westinghouse Analysis,.

"Report for the BAST Concentration Reduction for R.

E.

Ginna",

August

1985, submitted via Application for Amendment to the Operating License in a

letter from R.W.

Kober, RG&E to H.A.

Denton, NRC, dated October 16, 1985 (2)

UFSAR Section 3.8.1.2.2 (3)

UFSAR Section 6.2.4 Amendment No.

54 3.6-4

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

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

a 0 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, 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. 2,1g

,S4

I 4t L-

~ U

'l

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.l.d are not

met, then in addition to the requirements of 3.8.2, isolate the shutdown purge and mini-purge penetrations within 4

hours.

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 Amendment NO.

54

3. 8-3

1u~

'E,q

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 willprevent 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.

2 s54 3.8-5

References (1)

UFSAR Sections 9.1.4.4 and 9.1.4.5 (2)

Reload Transient Safety Report, Cycle 14 (3)

UFSAR Section 15.7.3.3 Amendment NO.

54 3.8-6

4.4.1.4 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.

~ phyla Lt shall be determined as Lt = LalsaJ which equals

.1528 percent weight per day at 35 psig.

Pa is defined as the calculated peak containment internal pressure related to design basis accidents which is greater than or equal to 60 psig.

La is defined as the maximum allowable leakage rate at Pa which equals

.2 percent weight per day.

c ~

The leakage rate at Pa (Lam) shall be

<0.75 La.

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

4.4.1.5 Test Fre uenc a ~

A set of three integrated leak rate tests shall be performed at approximately equal intervals during each 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, ll~

l,iio 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.

Amendment No.

54

4. 4-4

b.

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

~

~lie iiie 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 Ve 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.

Amendment No.54 4.4-6

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 c ~

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 e Except as specified in b.

and c. below, individual b.

penetrations and containment isolation valves shall be tested during each reactor shutdown for refueling, or other convenient intervals, but, in no case at intervals greater than two years.

The containment equipment

hatch, fuel transfer

tube, steam generator inspection/maintenance penetration, and shutdown purge system flanges shall be tested at each refueling shutdown or after each use, if that be sooner.

Amendment No. f8 i54 4.4-7

'C

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 i 4.4-8

the tendon containing 6 broken wires) shall be inspected.

The accepted criterion then shall be no more than 4

broken wires in any of the additional 4 tendons. If this criterion is not satisfied, all of the tendons shall be inspected and if more than 5% of the total wires are broken, the reactor shall be shut down and depressurized.

4.4.4.2 Pre-Stress Confirmation Test a ~

Lift-offtests shall be performed on the 14 tendons identified in 4.4.4.1a

above, at the intervals specified in 4.4.4.1b. If the average stress in the 14 tendons checked is less than 144,000 psi (60% of ultimate stress), all tendons shall be checked for stress and retensioned, 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 frecjuencies 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,lÃ

~54 4.4-11

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

The duration and methods for the integrated leakage rate test established by ANSI N45.4-1972 provide a minimum level of accuracy and allow for daily cyclic variation in temperature and thermal radiation.

The frequency of the integrated leakage rate test is keyed to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.

Refueling shutdowns are scheduled at approximately one year intervals.

The specified frequency of integrated leakage rate tests is based on three major considerations.

First is the low probability of leaks in the liner, because of (a) the use of weld channels to test the leaktightness of the welds during erection, (b) conformance of the complete containment to a

0.1% per day leak rate at 60 psig during preoperational testing, and (c) absence of any significant stresses in the liner during reactor operation.

Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low value (0.60 La) of the total leakage that is specified as acceptable.

Third is the tendon stress surveillance program, which provides assurance than an important part of the structural integrity of the containment is maintained.

Amendment No.

54 4.4-13

The basis for specification of a total leakage of 0.60 La from penetrations and isolation boundaries is that only a portion of the allowable integrated leakage rate should be from those sources in order to provide assurance that the integrated leakage rate would remain within the specified limits during the intervals between integrated leakage rate tests.

Because most leakage during an integrated leak rate test occurs though penetrations and isolation

valves, and because for most penetrations and isolation valves a

smaller leakage rate would result from an integrated leak test than from a local test, adequate assurance of maintaining the integrated leakage rate within the specified limits is provided.

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

The test Amendment No.

54 4.4-14

'J tp

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

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

The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible.

Thus the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down the reactor.

The containment is provided with two readily removable tendons that might be useful to such a study.

In addition, there are 40 tendons, each containing a removable wire which will be used to monitor for possible corrosion effects.

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

Performance of cycling tests and verification of isolation times associated with automatic containment isolation valves is covered by the Pump and Valve Test Program.

Compliance with Appendix J to 10 CFR 50 is addressed under local leak testing requirements.

References:

(1)

UFSAR Section 3.1.2.2.7 (2)

UFSAR Section 6.2.6.1 (3)

UFSAR Section 15.6.4.3 (4)

UFSAR Section 6.3.3.8 (5)

UFSAR Table 15.6-9 (6)

FSAR Page 5.1.2-28 (7)

North-American-Rockwell Report 550-x-32, Reliability Handbook, February 1963.

(8)

FSAR Page 5.1-28 Autonetics Amendment No.

54 4.4-17