Text

Amends 180 to Licenses DPR-32 & DPR-37,respectively, Establishing Operating Requirements for Recirculation Mode Transfer Function
	ML20045H951
Person / Time
Site:	Surry
Issue date:	07/08/1993
From:	Berkow H; Office of Nuclear Reactor Regulation
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ML20045H952	List: ML20045H951; ML20045H957; ML20045H963;
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"[

'o UNITED STATES

- [ " 3,, ^,j NUCLEAR REGULATORY COMMISSION

.g WASHINGTON, D. C. 20555 s., u j VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.180 License No. DPR-32 1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated August 7, 1992, 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 authori7ed 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;

-l 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; 1

and E.

The issuance of this amendment is in accordance with 10 CFR Part I

51 of the Commission's regulations and all applicable requirements have been satisfied.

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, 2.

Accordingly, the license is amended by changes to the Technical Specifications as. indicated in the attachment to this license amendment, and paragraph 3.B of facility Operating License No. DPR-32 is hereby amended to read as follows:

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.180, 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 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION J

/

erbert N. Ber ow, Director Project Directorate Il-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 8, 1993

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UNITED STATES 8"

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NUCLEAR REGULATORY COMMISSION y,

,E WASHINGTON, D. C. 20555

'%.....,P' VIRGINIA ELECTRIC AND POWER COMPANY DOCKET N0. 50-281 SURRY PuWER STATION. UNIT NO. 2 AMENDMENT TO FACIllTY OPERATING LICENSE Amendment No.180 License No. DPR-37 1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated August 7, 1992, 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; i

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 i

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.

l

i s

e 2.

Accordingly, the license is amended by changes to the Technical i

Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-37 is hereby j

amended to read as follows:

i (B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.180, 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 30 days, i

FOR THE NUCLEAR REGULATORY COMMISSION

/ lJ

/

He ert N. Ber ow, Director Project Directorate II-2 i

Division of Reactor Projects - I/II j

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 8, 1993 I

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a ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.-

180 TOFhCILITYOPERATINGLICENSENO.DPR-32

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AMENDMENT NO. 180 TO FACILITY OPERATING LICENSE NO. DPR-37 QQfKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:

Remove Paaes Insert Paoes 1.0-1 thru 1.0-10 1.0-1 thru 1.0-7 3.3-1 thru 3.3-9 3.3-1 thru 3.3-7 3.4-1 thru 3.4-6 3.4-1 thru 3,4-4 3.7-1 thru 3.7-22 3.7-1 thru 3.7-29 4.1-7 4.1-7 4.1-9d 4.1-9d u

I

?

.J

V i

TS 1.0 i l

1.0 DEFINITIONS a

The following frequently used terms are defined fcr the uniform interpretation of the specifications.

A.

RATED POVER i

I A steady state reactor core heat output of 2441 MWt.

ii B.

THERMAL POWER

]

The total core heat transferred from the fuel to the coolant.

i C.

RE ACTOR OPERATION i

1.

REFUELING SHUTDOWN 1

i When the reactor is suberitical by.at least 5% Ak/k and Tavg s 5140*F and fuelis scheduled to be moved to or from the reactor e

l core.

a 2.

COLD SHUTDOWN When the reactor is subcritical by at least 1% Ak/k and Tavg s i

s200*F.

a 3.

INTERMEDIATE SHUTDOWN i

When the reactor is subcritical by at least 1.77% Ak/k and 200 F

< Tavg < 547*F.

j 4.

HOT SHUTDOWN When the reactor is suberitical by at least 1.77% Ak/k and Tavg is 2 547 F.

Amendment Nos.180 and 180

]

TS 1.0-2 5.

REACTOR CRITCAL When the neutron chain reaction is self sustaining and keff = 1.0.

6.

POWER OPERATION When the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

7.

REFUELING OPERATION Any operation involving movement of core components when the vessel head is unbolted or removed.

D.

OPERABLE A system, subsystem, train, component, or device shall be operable or have operability when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication.or other auxiliary -

equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

The system or component shall be considered to have'f 5is capability when: (1) it satisfies the limiting conditions for operation defined in Section 3, and (2) it has been tested periodically in accordance with Section 4 and meets its performance requirements.

E.

PROTECTIVE INSTRUMENTATION LOGIC 1.

ANALOG CHANNEL An arrangement of components and modules as required to generate a single protective action digital signal when required by a unit condition. An analog channel loses its identity when single action signals are combined.

Amendment Nos.180 and 180

TS 1.0-3 2.

AUTOMATIC ACTUATION LOGIC A group of matrixed relay contacts which operate in response to the digital output signals from the analog channels to generate a protective action signal.

F.

INSTRUMENTATION SURVEILLANCE 1.

CHANNEL CHECK The qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation on channels measuring the same parameter.

2.

CHANNEL FUNCTIONAL TEST Injection of a simulated signalinto an analog channel as close to j

the sensor as practicable or makeup of the logic combinations in a logic channel to verify that it is operable, including alarm and/or trip initiating action.

3.

CHANNEL CAllBHAliON Adjustment of channel output such that it. responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment action, alarm, or trip, and shall be deemed to include the CHANNEL FUNCTIONAL TEST.

G.

CONTAINMENT INTEGRITY Containment integrity shall exist when:

a.

The penetrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or Amendment Nos.180 and 180

I TS 1.0-4 2)

Closed by at least one closed manual valve, blind flange, or deactivated automatic valve secured in its closed position except as provided in Specification 3.8.C. Non automatic or deactivated automatic containment isolation valves may be opened intermittently for operational activities provided that the valves are under administrative control and are capable of beitig closed immediately,if required.

b.

The equipment access hatch is closed and sealed.

c.

Each airlock is OPERABLE except as provided in Specification 3.8.B.

d.

The containment leakage rates are within the limits of Specification 4.4.

e.

The sealing mechanism associated with each penetration (e.g.,

l welds, bellows, or O-rings) is OPERABLE.

H.

REPORTABLE EVENT i

i A reportable event shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

l j

t.

QUADRANT POWER TILT The quadrant power tilt is defined as the ratio of the maximum upper excore detector current to the Everage of the upper excore detector currents or the ratio of the maximum lower excore detector current to the average of the lower excore detector currents whichever is greater, if one excore detector is out of service, the three in-service units are used in computing the average.

J.

LOW POWER PHYSICS TESTS Low power physics tests conducted below 5% of rated power which measure fundamental characteristics of the core and 'related instrumentation.

Amendment Nos.180 and 180

TS 1.0-5 K.

FIRE SUPPRESSION WATER SYSTEM A fire suppression water system shall consist of: a water source (s),.

gravity tank (s) or pump (s), and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe, or spray system riser.

L.

OFFSITE DOSE CALCULATION MANUAL (ODCMJ The Offsite Dose Calculation Manual (ODCM) shall contain the i

methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section -

6.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi annual Radioactive Effluent Release Reports required by Specifications 6.6.B.2 l

and 6.6.B.3.

M.

DOSE EQUIVALENT l-131 The dose equivalent 1-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table 111 of TID 14844, " Calculation of j

Distance Factors for Power and Test Reactor Sites" or in NRC Regulatory Guide 1.109, Revision 1, October 1977.

i N.

GASEOUS RADWASTE TREATMENT SYSTEM A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to 3

release to the environment.

Amendment Nos.180 and 180 1

s

TS 1.0-6 O.

PROCESS CONTROL PROGRAM (PCP)

The process control program shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the [

processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, and other requirements governing the l disposal of the waste.

P.

PURGE - PURGING Purge or purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that l replacement air or gas is required to purify the confinement.

O.

VENTILATION EXHAUST TREATMENT SYSTEM A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents. Treatment includes passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.

l l

R VENTING Venting is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during venting. Vent, used in system I

names, does not imply a venting process.

Amendment Nos. 180 and 180

1 TS 1.0-7 i

S.

SITE BOUNDARY The site boundary shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

l T.

UNRESTRICTED AREA i

An unrestricted area shall be any area at or beyond the site boundary where access is not controlled by the licensee for purpose of protection of individuals from ex.nasure to radiation and radioactive materials or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, or recreational purposes.

l U.

MEMBER (S) OF THE PUBLIC Member (s) of the public shallinclude allindividuals who by virtue of their occupational status have no formal association with the plant. This category shallinclude non-employees of the licensee who are permitted l to use portions of the site for recreational, occupational, or other j

purposes not associated with plant functions. This category shall agi include non-employees such as vending machine servicemen or postmen who, as part of their formal job function, occasionally enter an

)

area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

.i Amendment Nos.180 and 180 1

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~TS 3.3-1 j

J 3.3 SAFETY INJECTION SYSTEM Aco!icability Applies to the operating status of the Safety injection System.

j Obiective j

.1 To define those limiting conditions for operation that are necessary to provide

l sufficient borated water to remove decay heat from the core in emergency j.

f situations.

Soecifications A.

A reactor shall not be made critical unless the'following conditions are met.

i 1.

The refueling water storage tank contains at least 387,100 gallons

]

of borated water at a maximum temperature of 45*F. The boron j

concentration shall be at least 2300 ppm but not greater than 2500 ppm.

2.

Each accumulator system is pressurized to at least 600 psia and contains a minimum of 975 ft3 and a maximum of 1025 ft3 of_

-i borated water with a boron concentration of at least 2250 ppm.

3.

Two channels of heat tracing shall be OPERABLE for the flow

'l paths.

4.

Two charging pumps are OPERABLE.

5.

Two low head safety injection pumps are OPERABLE.

-j 6.

All valves, piping, and interlocks associated with the above j

components which are required to operate under accident

?

conditions are OPERABLE.

Amendment Nos.180 and 180 1

TS 3.3-2 7.

The Charging Pump Cooling Water Subsystem shall be operP'.ing as follows:

a.

Make-up water from the Component Cooling Water Subsystem shall be available.

b.

Two charging pump component cooling water pumps and two charging pump service water pumps shall be OPERABLE.

C.

Two charging pump intermediate seal coolers shall be OPERABLE.

8.

During POWER OPERATION, the AC power shall be removed from the following motor-operated valves with the valves _in the open position:

Unit No.1 Unit No. 2 MOV 1890C

~ MOV 2890C 9.

During POWER OPERATION, the AC power shall be removed from the following motor-operated valves with the valves in the closed position:

Unit No.1 Unit No. 2 MOV 1869A MOV 2869A MOV 1869B MOV 2869B MOV 1890A MOV 2890A MOV 1890B MOV 2890B 10.

The accumulator discharge valves listed below shall be blocked open by de-energizing the valves motor operators when the reactor coolant system pressure is greater than 1000 psig.

Amendment Nos.180 and 180

TS 3.3-3

~

Unit No.1 Unit No. 2 MOV 1865A MOV 2865A MOV 1865B MOV 2865B MOV 1865C MOV 2865C 11.

POWER OPERATION with less than three loops in service is prohibited. The following loop isolation valves shall have AC power removed and be locked in open position during POWER OPERATION.

Unit No.1 Unit No. 2 MOV 1590 MOV 2590 MOV 1591 MOV 2591 MOV 1592 MOV 2592 MOV 1593 MOV 2593 MOV 1594 MOV 2594 MOV 1595 MOV 2595 12.

The total system uncollected leakage from valves, flanges, and pumps located outside containment shall not exceed the limit j

specified by Technical Specification 4.11.A.4.d.

B.

The requirements of Specification 3.3.A may be modified to allow one of l the following components to be inoperable at any one time. If the system is not restored to meet the requirements of Specification 3.3.A within the time period specified, the reactor shall be placed in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the requirements of Specification 3.3.A are not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the reactor shall be placed in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1.

One accumulator may be isolated for a period not to exceed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Two charging pumps per unit may be inoperable, provided immediate attention is directed to making repairs and one of the inoperable pumps is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendment Nos.180 and 180

a TS 3.3-4 3.

One low head. safety injection subsystem per unit may be inoperable provided immediate attention is directed to making i

repairs and the subsystem is restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

One channel of heat tracing may be inoperable for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , provided immediate attention is directed to making repairs.

5.

One charging pump component cooling water pump or one charging pump service water pump may be inoperable provided the pump is restor 9d to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

6.

One charging pur.1p intermediate seal cooler or other passive component may oe inoperable provided the system may still operate at 100 percent capacity and repairs are completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

7.

Power may be restored to any valve referenced in Specifications 3.3.A.8 and 3.3.A.9 for the purpose of valve testing or maintenance, provided that no more than one valve has power restored and the testing and maintenance is completed and power removed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

8.

Power may be restored to any valve referenced in Specification 3.3.A.10 for the purpose of valve testing or maintenance, provided that no more than one valve has power restored and the testing and maintenance is completed and power removed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

9.

The total uncollected system leakage for valves, flanges, and pumps located outside containment can exceed the limit stated in Specification 4.11.A.4.d provided immediate attention is directed

)

to making repairs and uncollected system leakage is returned to within limits within 7 days.

I Amendment Nos.180 and 180

TS 3.3-5 10.

Refueling water storage tank volume, temperature, and boron concentration may be outside the limits of Specification 3.3.A.1 provided they are restored to within their respective limits within one hour.

Basis The normal procedure for starting the reactor is, first, to heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing control rods and/or diluting boron in the coolant. With this mode of startup the Safety injection System is required to be OPERABLE as specified. During LOW POWER PHYSICS TESTS there is a negligible amount of energy stored in the system. Therefore, an accident comparable in severity to the Design Basis Accident is not possible, and the full capacity of the Safety injection System would not be necessary.

The OPERABLE status of the various systems and components is to be demonstrated by periodic tests, detailed in TS Section 4.11. A large fraction of these tests are performed while the reactor is operating in the power range. If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. A single component being inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures. In some cases, additional components (i.e., charging pumps) are installed to allow a component to be inoperable without affecting system redundancy.

If the inoperable component is not repaired within the specified allowable time period, or a second component in the same or related system is found to be inoperable, the reactor willinitially be placed in HOT SHUTDOWN to provide for reduction of the decay heat from the fuel, and consequent reduction of cooling requirements after a postulated loss-of-coolant accident. After 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in HOT SHUTDOWN, if the malfunction (s) is not corrected the reactor will be placed in COLD SHUTDOWN following normal shutdown and cooldown procedures.

Amendment Nos.180 and 180

TS 3.3-6 The Specification requires prompt action to effect repairs of an inoperable component or subsystem. Therefore, in most cases, repairs will be completed l in less than the specified allowable repair times. Furthermore, the specified repair times do not apply to regularly scheduled maintenance of the Safety injection System, which is normally to be performed during refueling shutdowns. The limiting times for repair are based on: estimates of the tims required to diagnose and correct various postulated malfunctions using safe and proper procedures, the availability of tools, materials and equipment, health physics requirements, and the extent to which other systems provide functional redundancy to the system under repair.

Assuming the reactor has been operating at full RATED POWER for at least 100 days, the magnitude of the decay heat production decreases as follows after a unit trip from full RATED POWER.

i Time After Shutdown Decav Heat. % of RATED POWER j

1 min.

3.7 30 min.

1.6

')

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1.3 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 0.75 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0.48 i

Thus, the requirement for core cooling in case of a postulated loss-of-coolant j

accident, while in HOT SHUTDOWN,is reduced by orders of magnitude below the requirements for handling a postulated loss-of-coolant accident occurring during POWER OPERATION. Placing and maintaining the reactor in HOT SHUTDOWN significantly reduces the potential consequences of a loss-of-coolant accident, allows access to some of the Safety injection System i

components in order to effect repairs, and minimizes the plant's exposure to j

thermal cycling.

Failure to complete repairs within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of going to HOT SHUTDOWN is considered indicative of unforeseen problems (i.e., possibly the need of major maintenance). In such a case, the reactor is placed in COLD SHUTDOWN.

i I

Amendment Nos.180 and 1eo

TS 3.3-7 The accumulators are able to accept leakage from the Reactor Coolant System without any effect on their operability. Allowable inleakage is based on the volume of water than can be added to the initial amount without exceeding the volume given in Specification 3.3.A.2. The maximum acceptable inleakage is 50 cubic feet per tank.

The accumulators (one for each loop) discharge into the cold leg of the reactor coolant piping when Reactor Coolant System pressure decreases below accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motor-operated valve to isolate the accumulator during reactor start-up and shutdown to preclude the discharge of the contents of the accumulator when not required.

These valves receive a signal to open when safety injection is initiated.

However, to assure that the accumulator valves satisfy the single failure criterion, they will be blocked open by de-energizing the valve motor operators l

when the reactor coolant pressure exceeds 1000 psig. The operating pressure of the Reactor Coolant System is 2235* psig and accumulator injection is initiated when this pressure drops to 600 psia.

De-energizing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal startup operation to perform the actions required to de energize the valve. This procedure will assure that there is an OPERABLE flow path from each accumulator to the Reactor Coolant System during POWER OPERATION and that safety injection can be accomplished, The removal of power from the valves listed in the specification will assure that the systems of which they are a part satisfy the single failure criterion, j

Total system uncollected leakage is controlled to limit offsite doses resulting from system leakage after a loss-of-coolant accident.

For Unit 2 Cycle 12, Reactor Coolant System nominal operating pressure may be reduced to 2135 psig.

Amendment Nos,180 and 180

TS 3.4-1l l

3.4 SPRAY SYSTEMS 3

i Acolicability

]

l Applies to the operational status of the Spray Systems.

-l Obiective i

To define those limiting conditions for operation of the Spray Systems i

necessary to assure safe unit operation.

j Soecification A.

A unit's Reactor Coolant System temperature or pressure shall not be made to exceed 350 F or 450 psig, respectively, unless the following Spray System conditions in the unit are met-1.

Two Containment Spray Subsystems, including containment j

spray pumps, piping, and valves shall be OPERABLE.

2.

Four Recirculation Spray Subsystems, including recirculation spray pumps, coolers, piping, and valves shall be OPERABLE.

3.

The refueling water storage tank shall contain at least 387,100 gallons of borated water at a maximum temperature of 45'F. The boron concentration shall be at least 2300 ppm but not greater than 2500 ppm.

4.

The refueling water chemical addition tank shall contain at least 4,200 gallons of solution with a sodium hydroxide concentration of at least 17 percent by weight but not greater than 18 percent by weight.

t 5.

All valves, piping, and interlocks associated with the above components which are required to operate under accident conditions shall be OPERABLE.

Amendment Nos.180 and 180

TS 3.4 2 6.

The total uncollected system leakage from valves, flanges, and pumps located outside containment shall not exceed the limit specified by Specification 4.5.B.4.

B.

During POWER OPERATION the requirements of Specification 3.4.A mhy i be modified to allow a subsystem or the following components to be inoperable. If the components are not restored to meet the requirements of Specification 3.4.A within the time period specified below, the reactor shall be placed in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the requirements of Specification 3.4.A are not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the reactor shall be placed in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

One Containment Spray Subsystem may be inoperable, provided 1.

immediate attention is directed to making repairs and the subsystem can be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

One outside Recirculation Spray Subsystem may be inoperable, provided immediate attention is directed to making repairs and the subsystem can be restored to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

One inside Recirculation Spray Subsystem may be inoperable, provided immediate attention is directed to making repairs and the subsystem can be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.

The total uncollected system leakage from valves, flanges, and pumps located outside containment can exceed the limit stated in Specification 4.5.B.4, provided immediate attention is directed to making repairs and uncollected system leakage is returned to within limits within 7 days.

5.

Refueling Water Storage Tank volume, temperature, and boron concentration may be outside the limits of Specification 3.4.A.3 provided they are restored to within their respective limits within one hour.

Amendment Nos.180 and 180

TS 3.4-3 Basis The spray systems in each reactor unit consist of two separate parallel Containment Spray Subsystems, each of 100 percent capacity, and four separate parallel Recirculation Spray Subsystems, each of 50 percent capacity.

Each Containment Spray Subsyste.n draws water independently from the refueling water storage tank (RWST). The water in the tank is cooled to 45 F or below by circulating the water through one of the two RWST coolers with one of the two recirculating pumps. The water temperature is maintained by two mechanical refrigerating units as required.

In each Containment Spray Subsytem, the water flows from the tank through an electric motor driven containment spray pump and is sprayed into the containment atmosphere through two separate sets o't spray nozzles. The capacity of the spray systems to depressurize the containment in the event of a Design Basis Accident is a function of the pressure and temperature of the containment atmosphere, the service water temperature, and the temperature in the refueling water storage tank as discussed in the Basis of Specification 3.8.

Each Recirculation Spray Subsystem draws water from the common containment sump. In each subsystem the water flows through a recirculation spray pump and recirculation spray cooler, and is sprayed into the containment atmosphere through a separate set of spray nozzles. Two of the recirculation spray pumps are located inside the containment and two outside the containment in the containment auxiliary structure.

With one Containment Spray Subsystem and two Recirculation Spray Subsystems operating together, the spray systems are capable of cooling and l depressurizing the containment to subatmospheric pressure in less than 60 minutes following the Design Basis Accident.

The Recirculation Spray Subsystems are capable of mp.intaining subatmospheric pressure in the containment indefinitely following the Design Basis Accident when used in conjunction with the Containment Vacuum System to remove any long term air in leakage.

Amendment Nos.180 and 180

l l

TS 3.4-4 In addition to supplying water to the Containment Spray System, the refueling l

water storage tank is also a source of water for safety injection following an accident. This water is borated to a concentration which assures reactor shutdown by approximately 5 percent Ak/k when all control rod assemblies are inserted and when the reactor is cooled down for refueling.

l i

i Total system uncollected leakage is controlled to limit offsite doses resulting from system leakage after a loss-of-coolant accident.

References UFSAR Section 4 Reactor Coolant System UFSAR Section 6.3.1 Containment Spray Subsystem UFSAR Section 6.3.1 Recircula'lon Spray Pumps and Coolers UFSAR Section 6.3.1 Refueling Water Chemical Addition Tank UFSAR Section 6.3.1 Refueling Water Storage Tank UFSAR Section 14.5.2 Design Basis Accident UFSAR Section 14.5.5 Containment Transient Analysis l

l 1

l l

Amendment Nos.180 and 180 j

l

-4

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l-TS 3.7-1 L

l 3.7 MSTRUMENTATION SYSTEMS 1

Ooerational Safety Instrumentation Acolicability Applies to reactor and safety features instrumentation systems.

Objectives To ensure the automatic initiation of the Reactor Protection System and the Engineered Safety Features'in the event that a principal process variable limit is exceeded, and to define the limiting conditions for operation of the plant instrumentation and safety circuits necessary to ensure reactor and plant safety.

)

l Soecification l

~l A.

During on-line testing or in the event of a subsystem instrumentation channel failure, plant operation at RATED POWER shall be permitted to continue in accordance with Tables 3.7-1 through 3.7 3.

B.

The Reactor Protection System instrumentation channels and interlocks shall be OPERABLE as specified in Table 3.7-1.

C.

The Engineered Safeguards Actions and Isolation Function Instrumentation channels and interlocks shall be OPERABLE as specified in Tables 3.7-2 and 3.7-3, respectively.

I D.

The Engineered Safety Features initiation instrumentation setting limits shall be as stated in Table 3.7-4.

l i-E.

The explosive gas monitoring instrumentation channel shown in Table l

3.7-5(a) shall be OPERABLE with its alarm setpoint set to ensure that the 1

limits of Specification 3.11.A.1 are not exceeded.

1.

With an explosive gas monitoring instrumentation channel alarm setpoint less conservative than required by the above specification, declare the channelinoperable and take the action shown in Table 3.7-5(a).

Amendment Nos.180 and 180

e 4

TS 3.7-2 2.

With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the action shown in Table 3.7-5(a). Exert best efforts to return the instruments to operable status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission (Region 11) to explain why this inoperability was not corrected in a timely manner.

F.

The accident monitoring instrumentation listed in Table 3.7-6 shall be OPERABLE in accordance with the following:

1.

With the number of OPERABLE accident monitoring instrumentation channe'. less than the Total Number of Channels shown in Table 3.7-6,

' ems 1 through 9, either restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum OPERABLE Channels requirement of Table 3.7-6, items 1 through 9, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

G.

The containment hydrogen analyzers and associated support equipment shall be OPERABLE in accordance with the following:

1.

Two independent containment hydrogen analyzers shall be OPERABLE during REACTOR CRITICAL or POWER OPERATION.

a.

With one hydrogen analyzer inoperable, restore the t

inoperable analyzer to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Amendment Nos.180 and 180 l

J

TS 3 7-3 b.

With both hydrogen analyzers inoperable, restore at least one analyzer to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

NOTE:

Operability of the hydrogen analyzers includes proper operation of the respective Heat Tracing System.

Basis Instrument Operating Conditions During plant operations, the complete instrumentation system will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with ce-tain instrumentation channels out of service since provisions were made for this in the plant design. This specification outlines the limiting conditions for operation necessary to preserve the effectiveness of the Reactor Protection Sys:em when any one or more of the channels is out of service.

1 Almost all Reactor Protection System channels are supplied with su. scient redundancy to provide the capability for channel calibration and test at power.

i Exceptions are back;:p channels such as reactor coolant pump breakers. The removal of one trip v.iannel on process control equipment is accomplished by placing that channel bistable in a tripped mode (e.g., a two-out-of-three circuit

.l becomes a one-out-of-two circuit). The Nuciear Instrumentation System (NIS) channels are not intentionally placed in a tripped mode since the test signal is superimposed on the normal detector signal to test at power. Testing of the NIS l

power range channel requires: (a) bypassing the dropped-rod protection from NIS, for the channel being tested, (b) placing the AT/Tavg rotection channel set p

that is being fed from the NIS channel in the trip mode, and (c) defeating the power mismatch section of Tavg control channels when the appropriate NIS I

channelis being tested. However, the Rod Position System and remaining NIS channels still provide the dropped rod protection. Testing does not trip the j

system unless a trip condition exists in a concurrent channel.

i Amendment Nos.180 and 180

TS 3.7-4 Instrumentation has been provided to sense accident conditioris and to initiate operation of the Engineered Safety Features.(1)

Safety injection System Actuation Protection against a loss-of-coolant or steam line break accident is provided by automatic actuation of the Safety injection System (SIS) which provides emergency cooling and reduction of reactivity.

The loss-of-coolant accident is characterized by depressurization of the Reactor Coolant System and rapid loss of reactor coolant to the containment. The engineered safeguards instrumentation has been designed to sense these effects of the loss-of-coolant accident by detecting low pressurizer pressure to generator signals actuating the SIS active phase. The SIS active phase is also actuated by a high containment pressure signal brought about by loss of high enthalpy coolant to the containment. This actuation signal acts as a backup to the low pressurizer pressure actuation of the SIS and also adds diversity to protection against loss of coolant.

Signals are also provided to actuate the SIS upor, sensing the effects of a steam line break accident. Therefore, SIS actuation following a steam line break is designed to occur upon senna high differential steam pressure between the steam header and stez sn.. ator line or upon sensing high steam line flow in coincidence with low i'actnr nulant average temperature or low steam line pressure.

The increase in the extraction of RCS heat following a steam line break results in reactor coolant temperature and pressure reduction.

For this reason, protection against a steam line break accident is also provided by low pressurizer pressure actuating safety injection.

Protection is also provided for a steam line break in the containment by actuation of SIS upon sensing high containment pressure, Amendment Nos.180 and 180

a TS 3.7-5 SIS actuation injects highly bor'ated fluid into the Reactor Coolant System in order to counter the reactivity insertion brought about by cooldown of the reactor coolant which occurs during a steam line break accident.

Containment Spray The Engineered Safety Features also initiate containment spray upon sensing a high-high containment pressure signal. The containment spray acts to reduce containment pressure in the event of a loss-of-coolant or steam line break accident inside the containment. The containment spray cools the containment directly and limits the release of fission products by absorbing iodine should it be released to the containment.

i i

Containment spray is designed to be actuated at a higher containment pressure than the SIS. Since spurious actuation of containment spray is to be avoided,it l is initiated only on coincidence of high-high containment pressure sensed by 3 out of the 4 containment pressure signals.

l Steam Line Isolation Steam line isolation signals are initiated by the Engineered Safety Features closing the steam line trip valves. In the event of a steam line break, this action prevents continuous, uncontrolled steam release from more than one steam generator by isolating the steam lines on high-high containment pressure or high steam line flow with coincident low steam line pressure or low reactor coolant average temperature. Protection is afforded for breaks inside or outside the containme-t even when it is assumed that there is a single failure in the steam line isolation system.

Feedwater Line Isolation The feedwater lines are isolated upon actuation of the SIS in order to prevent excessive cooldown of the Reactor Coolant System. This mitigates the effects of an accident such as a steam line break which in itself causes excessive coolant temperature cooldown. Feedwater line isolation also Amendment Nos.180 and 180

TS 3.7-6 reduces the consequences of a steam line break inside the containment by j

stopping the entry of feedwater.

Auxiliary Feedwater System Actuation The automatic initiation of auxiliary feedwater flow to the steam generators by instruments identified in Table 3.7 2 ensures that the Reactor Coolant System decay heat can be removed following loss of main feedwater flow. This is consistent with the requirements of the "TMI-2 Lessons Learned Task Force i

Status Report," NUREG-0578, item 2.1.7.b.

j Setting Limits 1.

The high containment pressure limit is set at about 10% of design containment pressure. Initiation of safety injection protects against loss J

of coolant (2) or steam line break (3) accidents as discussed in the safety analysis.

2.

The high-high containment pressure limit is set at about 23% of design containment pressure. Initiation of containment spray and steam line isolation protects against large loss-of-coolant (2) or steam line break accidents (3) as discussed in the safety analysis.

3.

The pressurizer low pressure setpoint for safety injection actuation is set 1

substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss-of-coolant accident as shown in the safety analysis.(2) 4.

The steam line high differential pressure limit is cet well below the j

differential pressure expected in the event of a large steam line break

)

accident as shown in the safety analysis.(3) 5.

The high steam line flow differential pressure setpoint is constant at 40%

full flow between no load and 20% load and increasing linearly to 110%

.l of full flow at full load in order to protect against large steam line break accidents. The coincident low Tavg setting limit for SIS and steam line isolation initiation is set below its HOT SHUTDOWN value.

The coincident Amendment Nos.180 and 1eo 4

1 TS 3.7-7 i

steam line pressure setting limit is set below the full load operating pressure.

The safety analysis shows that these settings provide protection in the event of a large steam line break.(3)

Accident Monitoring Instrumentation The operability of the accident monitoring instrumentation in Table 3.7-6 ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. On the pressurizer PORVs, the pertinent channels consist of redundant limit switch indication. The pressurizer safety valves utilize an acoustic monitor channe!

and a downstream high temperature indication channel. This capability is consistent with the recommendations of Regulatory Guide 1.97,

" Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975, and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations." Potential accident effluent release paths are equipped with radiation monitors to de*ect and measure concentrations of noble gas fission products in plant gaseous effluents during and following an accident.

The effluent release paths monitored are the process vent stack, ventilation vent stack, main steam safety valve and atmospheric dump valve discharge and the AFW pump turbine exhaust. These monitors meet the requirements of NUREG 0737.

Instrumentation is provided for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the Waste Gas Holdup System. The l operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60,63 and 64 of Appendix A to 10 CFR Part 50.

Containment Hydrogen Analyzers indication of hydrogen concentration in the containment atmosphere is provided in the control room over the range of zero to ten percent hydrogen concentration.

1 Amendment Nos.180 and 180 1

TS 3.7-8 These redundant, qualified hydrogen analyzers are sh'ared by Units 1 and 2 with instrumentation to indicate hnd record the hydrogen concentration.

l A transfer switch is provided for Unit 1 to use both analyzers or for Unit 2 to use both analyzers. In addition, each unit's hydrogen analyzer has a transferable emergency power supply from Unit 1 and Unit 2. This will ensure redundancy for each unit.

Indication of Unit 1 and Unit 2 hydrogen concentration is provided on the Unit 1 Post Accident Monitoring panel and the Unit 2 Post Accident Monitoring panel, respectively. Hydrogen concentration is also recorded on qualified recorders.

In addition, each hydrogen analyzer is provided with an alarm for trouble /high hydrogen content. These alarms are located in the control room.

The supply lines instal'ad from the containment penetrations to the hydrogen analyzers have Category l Class IE heat tracing applied. The heat tracing system receives the same transferable emergency power as is provided to the containment hydrogen analyzers. The heat trace system is de-energized during normal system operation. Upon receipt of a SIS, after a preset time delay, heat tracing is energized to bring the piping process temperature to 250 i 10 F within 20 minutes. Each heat trace circuit is equipped with an RTD to provide individual circuit readout, over-temperature alarm, and control the circuit to maintain the process temperatures.

The hydrogen analyzer heat trace system is equipped with high temperature, loss of D.C. power, loss of A.C. power, loss of control power, and failure of l automatic initiation alarms.

Non-Essential Service Water Isolation System The operability of this functional system ensures that adequate intake canal inventory can be maintained by the Emergency Service Water Pumps. l Adequate intake canal inventory provides design service water flow to the recirculation spray heat exchangers and other essential loads (e.g., control room area chillers, charging pump lube oil coolers) following a design basis loss of coolant accident with a coincident loss of offsite power. This system is common to both units in that each of the two trains will actuate equipment on each unit.

Amendment Nos.180 and 180

I l

i TS 3.7-9 i

l References (1)

UFSAR - Section 7.5 (2)

UFSAR - Section 14.5 i

(3)

UFSAR - Section 14.3.2 i

l 1

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j Amendment Nos.180 and'1so

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TABLE 3.7-1 REACTOR TRIP INSTRUMENT OPERATING CONDITIONS Minimum Total Number OPERABLE Channels Permissible Functional Unit Of Channels Channels To Trio Dvoass Conditions Operator Action

17. Low steam 0enerator water 2/ loop-level and 1/!Oop-level 1/ loop-level 7

levelwith stearrV!eedwater 2/ loop-flow and 2/ loop-coincident flow rnismatch mismatch flow mismatch with 1/ loop-or 2/ loop-level flow mismatch and 1/ loop-flow in same loop mismatch 18.

a.

ReadorTrip Breakers 2

2 1

8

b. Reactor Trp 2

1 1

Bypass Breakers -Note C

19. Automatic Trip Logic 2

2 1

11 20.

Reactor Trip System interiocks - Note D a.

Intermediate range neutron 2

2 1

13 flux, P-6

b. Low power reactor trips block, P-7 Power range neutron flux, P-10 4

3 2

13 and k

Turbine impulse pressure 2

2 1

13 o

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c. Power range neutron flux, P-8 4

3 2

13 o"

d. Power range neutron flux, P-10 4

3 2

13 E

P

e. Turbine impulse pressure 2

2 1

13.

g Note C - With the Reactor Trip Breaker open for surveillance testing in accordance with Specification Table 4.1-1 (Item 30)

Note D - Reactor Trip System IrWeriocks are described in Table 4.1-A 4

d 3

O

TS 3.7-13 TABLE 3.7-1 (Continued)

TABLE NOTATION ACTION STATEMENTS ACTION 1.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 2.A.

With the number of OPERABLE channels equal to the Minimum OPERABLE Channels, POWER OPERATION may proceed provided the following conditions are satisfied:

1.

The inoperable channelis placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I 2.

The Minimum OPERABLE Channels requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of the redundant channel (s) per Specification 4.1.

3.

Either, THERMAL POWER is restricted to s 75% of RATED POWER and the Power Range, Neutron Flux trip setpoint is reduced to s 85% of RATED POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the OUADRANT POWER TILT is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l i

I I

Amendment Nos.180 and 180

l TS 3.7-14 TABLE 3.7-1 (Continued) 1 4.

The OUADRANT POWER TILT shall be determined to be within the limit when above 75 percent of RATED POWER with one Power Range Channelinoperable by using the moveable incore detectors to confirm that the normalized symmetric power distribution, obtained from 2 sets of 4 symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.B.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION 3.

With the number of OPERABLE channels one less than l required by the Minimum OPERABLE Channels requirement and with the THERMAL POWER level:

a.

Below the P-6 (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 10% of RATED POWER, restore the.

Inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED POWER.

c.

Above 10% of RATED POWER, POWER OPERATION may continue.

Amendment Nos.180 and 180

TS 3.7-15 TABLE 3.7-1 (Continued)

ACTION 4.

With the number of channels OPERABLE one less than required by the Minimum OPERABLE Channels requirement and with the THERMAL POWER level:

i a.

Below P-6, (Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

Two Source Range channels must be i

OPERABLE prior to increasing THERMAL POWER above I

the P-6 setpoint.

b.

Above P-6, operation may continue.

ACTION S.

With the number of OPERABLE channels one less than ]

required by the Minimum OPERABLE Channels requirement, verify compliance with the Shutdown Margin requirements within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

ACTION 6.A.

With the number of OPERABLE channels equal to the Minimum OPERABLE Channels requirement, REACTOR CRITICAL and POWER OPERATION may proceed provided the following conditions are satisfied:

1.

The inoperable channelis placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

The Minimum OPERABLE Channels requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.1.

6.B.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Amendment Nos.180 and 180

i TS 3.7-16 1

TABLE 3.7-1 (Continued)

ACTION 7.

With the number of OPERABLE channels equal to the Minimum OPERABLE Channels, REACTOR CRITICAL and POWER OPERATION may proceed provided the following conditions are satisfied:

1.

The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

The Minimum OPERABLE Channels requirement is met; i

however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification-3 4.1.

t ACTION 8.A.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In conditions of operation other than REACTOR CRITICAL or POWER OPERATIONS, with the number of OPERABLE channels one less' than the-Minimum OPERABLE Channels requirement, restore the i

inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour. However, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for i

surveillance testing per Specification 4.1 provided the other channelis OPERABLE.

8.B.

With one of the diverse trip features (undervoltage or shunt trip device) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply Action 8.A.'

The breaker shall not be bypassed while one of the diverse trip features is inoperable' ~except for the time required for performing maintenance to restore the breaker to OPERABLE status.

I 1

~

Amendment Nos.180 and 180

TS 3.7-17 TABLE 3.7-1 (Continued)

ACTION 9.

With one channel inoperable, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce THERMAL POWER to below the P 8 (Block of Low Reactor Coolant Pump Flow and Reactor Coolant Pump Breaker Position) setpoint-within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Operation below P-8 may continue pursuant to ACTION 10.

ACTION 10.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, operation may continue provided the inoperable channel is placed in the j

tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I ACTION 11.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, be in at least HOT SHUTDOWN within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . In conditions of operation other than REACTOR CRITICAL or POWER OPERATIONS, with the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour. However, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.1 provided the other channelis OPERABLE.

ACTION 12.

With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 13.

With the number of OPERABLE channels less than the Minimum OPERABLE Channels requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or be in at least HOT.

SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l Amendment Nos.180 and 180

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w TABLE 3.7-2 (Continued)

ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS Minimum Total Number OPERABLE Channels Permissible Operator Functional Unt Of Channels Channels To Trio Bypass Conddions Actions

2. CONTAINMENTSPRAY

a. Manual 1 set 1 set 1 set' 15 t

b. 'High containment pressure 4

3 3

17 i

H-HI)

c. Automaticactuationlogic 2

2 1

14-3.

AUXILIARY FEEDWATER '

{

a.

Steam generatorwaterlevel low-low

1) Start motor driven pumps 3/ steam 2/ steam 2/ steam 20 generator generator generator any I generator-i

2) Starts turbine delven pump 3/ steam -

2/ steam 2/ steam 20 generator-generator generator any 2 generators g

b. RCP undervoltage starts 3

2 2

20 g

turbine driven pung

&u!

-- c. Safety injection - start.

' See #1 above (at Si initiating functions and requirements)

S motor dHven purnps

~ f

d. Stationbladeout-start 1/ bus -

1/ bus -

.2' 21 motor driven pufmps 2 transfer 2 transfer

+

lg-buses / unit buses / unit ~

.g.

(n

'o w

A.

.. Must actuale 2 switches simultaneously H..

4,a.

o 2:.

,-. - - -.. ~. - -.... -..

....,u-2---...

, =.. -.

. -.... - -.... ~.. -. -

"1 TABLE 3.7-2 (Continued)

ENGINEERED SAFEGUARDS ACTION INSTRUMENT OPERATING CONDITIONS Minimum Total Number OPERABLE Channels permissible Operator Functional Unit Of Channels Channels To Trio D) pass Conditions

. Actions AUXILIARY FEEDWATER (continued)

e. Trip of main feedwater 2/MFW pump 1/MFW pump 2-1 each 21 pumps - start motor MFW pump driven pumps f.

Automatic actuation logic 2

2 1

22 4.

LOSS OF POWER a 4.16 kv emergency bus 3/ bus 2/ bus 2/ bus 20 undervoltage (loss of voltage) b.

4.16 kv emergency bus 3/ bus 2/ bus 2/ bus 20 undervoltage (degraded voltage) 5.

NON-ESSENTIAL SERVICE WATER ISOLATION a low intake canallevel 4

3 3

20

b. Automatic actuation logic 2

2 1

14 6.

ENGINEERED SAFEGUARDS ACTUATION INTERLOCKS-Note A a Pressurizer pressure, P-11 3

2 2

23 k.

b. Low-low Tavg. P-12 3

2 2

23 2

c.

Reactor trip, P 2 2

1 24 x

7.

RECIRCULATION MODE TRANSFER Y

.a e,a 8

a RWST Leveli ow '

4 3

2

'25 y

L k

~ b.. Automatic Actuation' Logic _.

'2 2_

'1 14 t

and Actuation Relays

,g Note A - Engmeered Safeguards Actuation Interlocks are described in Table 4.1-A a

a

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+ '

4-me e

'TV s

m"-

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a.t.

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4 TABLE 3.7-3 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS Minimum Total Number OPERABLE Channels Permissible Operator Functional unit Of Channels Channels To Trio Bvoass Conditions Actions 1.

CONTAINMENTISOLATION a.

Phasei

1) Safety injection (SI)

See item #1, Table 3.7-2 (all Si initiating functions and requirements) l

2). Automaticinitiation logic 2

2 1

14 i

3) Manual 2

2 1

21

b. Phase 2

1) High containment pressure 4

3 3

17

2) Automatic actuation logic 2

2 1

14

3) Manual 2

2 1

15 c.

Phase 3

1) High containment pressure 4

3 3

17 (Hi-HI setpoint) g

2) Automatic actuation logic 2

2 1

14 Eg

3) Manual 1 set i set 1 set' 15 2.

STEAMLINEISOLATION k

a. High steem flowin 2f3 Enes See item #1.e Table 3.7-2 for operability requirements coincident wnh 2/3 low Tavg or g

g 2/3 low steam pressures en -

ta

+

Must actuate 2 swnches simutaneously i

b H

O 4

TABLE 3.7-3 (Continued)

INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS Minimum Total Number OPERABLE Channels Permissible Operator -

Funchonal unt Of Channels Channels To Trio Bvoass Conditions.

Actions STEAMLINE ISOLATION (continued)

b. High containmert pressure 4

3 3

17 (Hi-Hi setpoint)

c. Manual 1/steamline 1/steamline 1/steamline 21

d. Automaticactuationlogic 2

2 1

22 3.

TURBINETRIP AND FEEDWATER ISOLATION

a. ~ Steam generatorwater-level 3/ steam 2/ steam 2/in any one 20 high-high -

generator generator steam generator

b. Automatic actuation logic 2

2 1

22 and actuation relay.

c. Safety injection See item #1 of Table 3.7-2 (all Si initiating functions and requirements) y 3

B-

$t' d

g-IM y

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.w

TS 3.7-23 TABLES 3.7-2 AND 3.7 3 TAEsLE NOTATIONS ACTION 14.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . One channel may be bypassed for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing per Specification 4.1, provided the other channelis OPERABLE.

ACTION 15.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 17.

With the number of OPERABLE channels or,e less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the Minimum OPERABLE Channels requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.1.

ACTION 19.

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable ch3nnel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 20.

With the number of OPERABLE channels one less than the Total Number of Channels, REACTOR CRITICAL and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped a.

condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

The Minimum OPERABLE Channels requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.1.

l Amendment Nos.180 and 180

TS 3.7 l TABLES 3.7-2 ANDS 3.7-3 (Continued)

TABLE NOTATIONS ACTION 21.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirements, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 22.

With the number of OPERABLE channels one less than the Minimum OPERABLE Channels requirement, be in at least HOT SHUTDOWN within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months and reduce pressure and temperature to less than 450 psig and 350* within the next 8 [

hours; however, one channel may be bypassec' for up to 8 1

hours for surveillance testing per Specification 4.1 provided the other channelis OPERABLE.

]

ACTION 23.

With the number of OPERABLE channels less than the 4

Minimum OPERABLE Channels requirement, within one hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24.

With the number of OPERABLE channels less than the Total Number of Channels, restore the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or reduce pressure and temperature to less than 450 psig and 350 F within the next 12 l hours.

ACTION 25.

With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channelin the bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.1.

Amendment Nos.180 and 180

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING NQ.

Functional Unit Channel Action Settina Limit l

1 High Containment Pressure (High Containment a) Safety injection s 5 psig Pressure Signal) b) Containment Vacuum Pump Trip c) High Press. Containment Isolation d) Safety injection Containment isolation e) F.W. Line Isolation 2

High-High Containment Pressure (High-High a) Containment Spray 510.3 psig l

Containment Pressure Signals) b) Recirculation Spray c) Steam Line Isolation d) High-High Press. Containment isolation l

3 Pressurizer Low-Low Pressure a) Safety injection 21,700 psig b) Safety injection Containmt nt Isolation c) F.W. Line isolation 4

High Differential Pressure Between a) Safety Injection s 150 psig Steam Une and the Steam Lins Header b) Safety injection Containment isolation l

c) F.W. Line Isolation 5

High Steam Flow in 2/3 Steam Lines a) Safety injection s 40% (at zeroload) of full steam flow s 40%(at 20% load) of full steam flow s 110% (at fullload) of full steam flow k

b) Steam Line Isolation D

c) Safety injection Containment isolation l

[

d) F.W. Line Isolation v#

Coincident with Low Tavg or 2 541'F Tavg h

Low Steam Line Pressure 2500 psig steamline pressure Q

td L

8 4

a m

O

TABLE 3.7-4 ENGINEERED SAFETY FEATURE SYSTEM INITIATION LIMITS INSTRUMENT SETTING thL Fundional Unit Channel Action Settino Limit l

6 AUXILIARY FEEDWATER l

a.

Steam Generator Water Level Low-Low Aux. Feedwater Initiation 2 5% narrow range S/G Blowdown isolation b.

RCP UndervoRage Aux. Feedwater Initiation 2 70% nominal

c. Safety injection Aux. Feedwater initiation All S.I. setpoints d.

Station Blackout Aux. Feedwater initiation 2 46.7% nominal

e. Main Feedwater Pump Trip Aux. Feedwater initiation N.A.

7 LOSS OF POWER l

a.

4.16 KV Emergency Bus UndervoRage Emergency Bus Separation and 75 (11.0)% volls with a (Loss of Voltage)

Diese! start 2 (+5, -0.1) second time delay

b. 4.16 KV Emergency Bus UndervoRage Emergency Bus Separation and 90 (11.0)% volts with a (Degraded Voltage)

Dieselstart 60 (13.0) second time delay (Non CLS, Non SI) 7 (i.35) second time delay l

(CLS or St Conditions)

ET 8

NON-ESSENTIAL SERVICE WATER ISOLATION a

g g-a.

Low Intake Canal Level isolation of Service Water flow to 23 feet-6 inches l

g non-essentialloads m:

9 RECIRCULATION MODE TRANSFER Initiation of Recirculation Mode 2 18.93%

8 a.

RWST Level-Low Transfer System s 19.43 %

i$o

.a ba e

8

.. m m

TABLE 3.7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM Automatic Function Monitoring Alarm Setpoint Monitor Channel At Alarm Conditions Recuirements uCI/cc l

1.

Corrponent cooling water radiation Shuts surge tank vent valve See Specification Twice Background monitors HCV-CC-100 3.13

2. Containment particxslate and ges Trips affected unit's purge supply See Specificatoin Particulate s 9 x 10-9 monitors (RM-RMS-159 &

fans, closes affected unit's purge 3.10 Gas s 1 x 10-5 RM-RMS-160, RM-RMS-259 &

air txstterfly valves (MOV-VS-100A, RM-RMS-260) 8 C & D or MOV-VS-200A, B. C & D) 3 Manipulator crane area monitors Trips affected unit's purgo supply See Speci!' cation s 50 mrern'br (RM-RMS-162 & RM-RMS-262) f ans, closes affected unit's purge 3.10 air butterfly valves (MOV-VS-100A.

B, C & D or MOV-VS-200A, B, C & D)

E B

Be h

?

w 9

4 i

q t

4 TABLE 3.7-5(a)

EXPLOS1VE GAS MONITORING INSTRUMENTATION

't Minimum l

Total No.

OPERABLE Instrument of Channels Channels Action

1. Waste Gas Holdup System Explosive Gas Monitoring System Oxygen Monitor 1

1 1

ACTION 1 - With the number of channels OPERABLE less than required by the minimum OPERABLE channels requirement, operation of this waste gas holdup system may continue provided grab samples are collected (1) at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months dunng degassing operations to the waste gas decay tank and (2) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations. Samples shall be analyzed wthin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after collection.-

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qeeo s

a S F 1

2 eRRC Y Y y

y i i i

l t

t il d

R R e

e n

n n

n c

t i

e' e

t t

x n

u a

O O a

a o

o o

n ia u

o o

f f

x t

t e

A n... P P S S C C C C P V M A N

N s

r abc I

w 0

1 2

3 4

.-m 1

2 3

4 5

6 7 8

9 1

1 1

-w

-v e

@'8* yy $ { -@

i

t l

TABLE 4.1-1 (Continued)

MINIMUM FREQUENCIES FOR CHECK _ CALIBRATIONS. AND TEST OF INSTRUMENT CHANNEL _S l

l Channel DescritAlon Check Calibrate les!

Remarks 10.

Rod Postion Bank Counters S(1,2)

N.A.

1) Each six inches of rod motion O(3) when dataloggeris out of service

2) Wdh analog rod postion

3) For the control banks, the bench-boardirdcators shall be checked against the output of the bank j.

overlap unit

11. Steam Generator Level S

R M

12. Charging Flow N.A.

R N.A.

13. ResidualHeat Removal Pump Flow N.A.

R N.A.

14.

Boric Acid Tank Level

D R

N.A.

15. Recirculation Mode transfer a.

Refueling Water Storage Tank Level-Low S

R M

b. Automatic Actuation Logic and N.A.

N.A.

M Actuation Relays

16. Volume ControlTank Level N.A.

R N.A.

h-

17. Reactor Containment Pressure-CLS

D R

M(1)

'1) Isolation valve signal and spray signal o

k

18. Boric Acid Control N.A.

R N.A.

o#

19. Containment Sump Level N.A.

R N.A.

5 A

P

20. Accumulator Leveland Pressure S

R N.A.

m b

ag

21. Containment Pressure-Vacuum S

R N.A.

'g Pump System i

E

22. Steam Line Pressure S

R M-O m

1

____m._;-..___._

E m 1.

w

=

...-r w,,

c e

TABLE 4.1-2A ICONTINUED)

MINIMUM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION DESCRIPTION TEST "10UENCY REFERENCE 18.

Primary Coolant System Functional 1.

Periodic leakage testing (a) on each valve listed in Specification 3.1.C.7a shall be accortplished prior to entering power operation condition after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in cold shutdown condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not been accomp-lished in the preceeding 9 months, and prior to retuming the valve to service after maintenance, rrpair or replacement work is performed.

19.

Containment Purge MOV Leakage Functional Semi-Annua! (Unit at power or shutdown) if purge valves are operated during interval (c)

20. Containment Hydrogen Analyzers a.

Channel Check Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

b. Channel Functional Test Once per 31 days

~

c.

ChannelCalibration using Once per 92 days on staggered basis sarrple gas containing:

1.

One volume percent (10.25%) hydrogen, balance nitrogen 2.

Four volume percent (t 0.25%) hydrogen, balance nitrogen

3. Channetcalbration test willinclude startup andoperationof the Heat Tracing System 9

21.

RCS Flow Flow 2 273,000 gpm Once per refueling cycle 14 9

22. RWSTparameters

a. Temperature s 45'F Once per shift

b. Volume 2 387,100 gallons Once per shift 5'u

-4 (a)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure 'ndicators) If isvirWed in accordanca m

8 with approved procedures and supported by computations showing that the method is capable of demonstrXing valve compliance with the leakage criteria.

(

(b)

Minimum differential test pressure shall not be below 150 psid.

k-(c)

Refer to Section 4.4 for acceptance criteria.

a 8

See Specification 4.1.D

ML20045H951

Text

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