ML20079L980

Proposed Tech Spec Change 10 Re Standby Gas Sys & Auxiliary Electric Sys,Core & Containment Cooling Sys,Torus Temp Monitoring Sys & Miscellaneous Revs

ML20079L980
Person / Time
Site:	Cooper
Issue date:	01/18/1984
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20079L977	List: ML20079L975 ML20079L980
References
NUDOCS 8401270085
Download: ML20079L980 (49)

Text

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. . Attachm:nt 1 ,

Revised Technical Specifications for the Standby Gas Treatment System and the Auxiliary Electrical System Revised Pages: 165 194 197 182 195 193 196 The current Technical Specifications for Cooper Nuclear Station require that both diesel generators be operable as active portions of the Standby Gas ,

Treatment System and are silent on a requirement for diesel generator operability during refueling operations.

Nebraska Public Power District requests a revision to the Technical Specifications to delete the requirement for diesel generator. operability in the Standby Gas Treatment section of the Technical Specifications and to add the requirement that at least one diesel generator be operable during refueling operations as shown on the attached pages. The proposed clarifications conform to NUREG-0123, Revision 3, Standard Technical Specification 3.6.5.3 which does not specifically require diesel generator operability for the Standby Gas Treatment System, and standard Technical Specification 3.8.1.2, which requires the operability of at least one diesel generator during refueling operations.

The need for clarification of the CNS Technical Specifications with regard to

. diesel generator operability during refueling operations was noted by the NRC Resident Inspector during the past outage as reported in NRC Inspection Report 83-12 and LER 50-298-83-07.

In addition, two typographical errors are corrected on page 165. Changes in Technical Specifications 3.9.A and 3.9.B improve the format of paragraph numbering.

Evaluation of this Revision with Respect to 10CFR50.92 A. The enclosed Technical Specification change is judged to involve no significant hazards based on the following:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation:

Because the proposed change brings this section of the Technical Specifications in agreement with the GE Standard Technical Specifications and clarifies the need for diesel generator operability during refueling operations and with regard to the Standby Gas Treatment System it does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

8401270085 840118 PDR ADOCK 05000298 P PDR

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NW Evaluation:

The operability requirements for the diesel generators are described in Technical Specification 3.9. The inclusion of a diesel generator requirement in the section of tha Technical Specifications on the Standby Gas Treatment System is unnecessary since operability of the system can be demonstrated with alternate sources of power.

Therefore, this change and the addition of the requirement to have at least one diesel gererator available during refueling operations do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Dces the proposed amendment involve a significant reduction in a margin of safety?

Evaluation:

Because these changes clarify the operability requirements of the Standby Gas Treatment System and the Auxiliary Electrical System and are in agreement with the GE Standard Technical Specifications they do not involve a significant reduction in a margin of safety.

B. Additional basis for proposed no significant hazards consideration determination:

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48FR14870). The examples include:

"(i) A purely administrative change to Technical Specifications . . . ",

"(11) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications . . .",

and "(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of tl change are cl.early within all acceptable criteria with respect to the system or component specified in the Standatd Review Plan . . ."

The proposed typographical and format corrections are clearly administrative changes encompassed by example (1). The proposed change adding the requirement for diesel generator availability during refueling operations is clearly encompassed by example (ii). The change bringing this Technical Specification in agreement with the Standard Technical Specifications with regard to the Standby Gas Treatment System is encompassed by example (vi) above.

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I LIMITING CONDITIONS FOR OPERATION fSURVEILLANCEREQUIREMENTS 3.7. (cont'd.) 4.7 (cont'd.)

B. Standby Gas Treatment Svstem B. Standby Gas Treatment System

1. Except as specified in 3.7.B.3 below, 1. At least once per operating cycle both standby gas treatment systems the following conditions shall be shall be operable at all times when demonstrated, secondary containment integrity is required, a. Pressure drop across the combined HEPA filters and charcoal adsorber 2.a. The results of the in-place cold DOP banks is less than 6 inches of and halogenated hydrocarbon tests at water at the system design flow design flows on HEPA filters and rate.

charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated b. Inlet heater input is capable of hydrocarbon removal. reducing R.H. from 100 to 70% R.H.

b. The results of laboratory carbon 2.a. The tests and sample analysis of sample analysis shall show >99%

Specification 3.7.B.2 shall be radioactive methyl iodide removal performed at least once per year at a velocity within 20 percent gf for standby service or after every actual system design, >1.75 mg/m 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and inlet methyl iodide concentration, following significant painting, l >70% R.H. and <30 C. fire or chemical release in any ventilation zone communicating

c. Fans shall be shown to operate within with the system.

+10% design flow.

b. Cold *D0P testing shall be performed

3. From and after the date that one after each complete or partial standby gas treatraent system is made replacement of the HEPA filter or found to be inoperable for any bank or after any structural reason, reactor operation or fuel maintenance on the system housing.

handling is permissible only during the succeeding seven days unless c. Halogenated hydrocarbon testing l such system is sooner made operable, shall be performed after each provided that during such seven days complete or partial replacement all active components of the other of the charcoal adsorber bank j standby gas treatment system shall or after any structural main-be operable. tenance on the system housing,

d. Each system shall be operated l with the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month,

e. Test sealing of gaskets for housing doors downstream of the HEPA filters and charcoal adsorbers shall be performed at, and in conformance with, each test performed for conpliance with Specification 4.7.B.2.a and Specification 3.7.B.2.a.

3. System drains where present shall be inspected quarterly for adequate water level in loop-seals.

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o 3.7.B & 3.7.C BASES (cont'd)

High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodido removal efficiency of at least 99 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents

. analyzed. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

Only one of the two standby gas treatment systems is needed to cleanup the reactor building atmosphere upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made. If neither system is operable.

l the plant is brought to a condition where the standby gas treatment system is not required.

4.7.B & 4.7.C BASES Standby Gas Treatment System and Secondary Containment Initiating reactor building isolation and operation of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary containment provides an adequate test of the operation of the reactor building isolation valvess leak tightness of the reactor building and performance of the standby gas treatment system. Functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing these tests prior to re-fueling will demonstrate secondary containment capability prior to the time the primary containment is opened for refueling. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treat-ment system performance capability.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A 7.8 kw heater is capable of maintaining relative humidity below 70%. Heater capacity and pressure drop should be determined at least once per operating cycle to show system performance capability.

The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed it. accordance with ANSI N510-1980. The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency tent. Each sample should be at least two inches in diameter and a lengLn equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced

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• . LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.9 AUXILIARY > ELECTRICAL SYSTEM 4.9 AUXILIARY ELECTRICAL SYSTEM Applicability: Applicability:

Applies to the auxiliary electrical Applies to the periodic testing power system, requirements of the auxiliary electrical systems.

Objective: Objective:

To assure an adequate supply of elec- Verify the operability of the auxiliary trical power for operation of.those electrical system.

systems required for safety. .

Specification: Specification:

A. Auxiliary Electrical Equipment A. Auxiliary Electrical Equipment

1. Emergency Buses Undervoltage l 1. The reactor shall not be made criti-Relays cal from a Cold Shutdown Condition unless all of the following condi-tions-are satisfied: a. Loss of voltage relays (a. Both off-site sources (345 KV and once every 18 months, loss 69 KV) and the startup transformer of voltage on emergency and emergency transformer are avail- buses is simulated to

, able and capable of automatically demonstrate the load shed-supplying power to the 4160 Volt -

ding from emergency buses emergency buses IF and 1G. and the automatic start lb. Both diesel generators shall be operable and there shall be a mini- b. Undervoltage relays mum of 45,000 gal. of diesel fuel in the fuel oil storage tanks. Once every 18 months, low voltage on emergency buses The 4160V critical buses IF and 1G is simulated to demonstrate

lc. and the 480V critical buses IF and 1G disconnection of the emer-are energized. gency buses from the offsite power source. The under-l 1. The loss of voltage relays and voltage relays shall be their auxiliary relays are calibrated once every 18 operable, months.

j. 2. The undervoltage relays and 2. Diesel Generators their auxiliary relays are operable. a. Each diesel-generator shall be startc manually and loaded to not less than

! d. The four unit 125V/250V batteries and 35% of rated load for no less than 2 their chargers shall be operable. hours once each month to demonstrate operational readiness.

{e. The power monitoriug system for the inservice RPS MG set or alternate source shall be operable.

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.: LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.9.A 4.9.A.2 (cont'd)

2. At least one diesel generator During the monthly generator test the shall be operable during diesel generator starting air compressor refueling operations, shall be checked for operation and its ability to recharge air receivers.

The operation of the diesel fuel oil transfer pumps and fuel oil day tank level switches shall be demonstrated, and the diesel starting time to reach rated voltage and frequency shall be logged.

b. Once every 18 months the condition under which the diesel generator is required will be simulated and a test conducted to demonstrate that it will start and accept the emergency load within the specified time sequence. The results shall be logged.

c. Specification 4.9.A.2.c deleted.

d. Once a month the quantity of diesel fuel available shall be logged.

e. -Every three man,ths and upon delivery a sample of diesel fuel shall be checked for quality. The quality shall be within the acceptable limits specified in Table 1 of ASTM D975-68 for Nos. 1D or 2D and logged.

f. Each diesel generator shall be given an annual inspection in accordance with instructions based on the manufacturer's recommendations.

3. Unit Batteries

a. Every week the specific gravity, the voltage and temperature of the pilot

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• LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMEhTS 3.9;A 4.9.A (cont'd.)

cell and overall battery voltage shall be measured and logged.

b. Every three months the measurements shall be made of the voltage of each cell to nearest 0.1 Volt, specific gravity of each cell, and temperature of every sixth cell. These measure-ments shall be logged.

c. Once each operating cycle, the stated batteries shall be subjected to a rated load discharge test. The specific gravity and voltage of each cell shall be determined after the discharge and logged.

B. Operation with Inoperable Equipment 4. Power Monitoring System for RPS System

1. Whenever the reactor is in Run Mode or The above specified RPS power monitor-Startup Mode with the reactor not in a ing system instrumentation shall be Cold Condition,'the availability of determined operable:

electric power shall be as specified in 3.9.A.1, except as specified in a. At least once per operating cycle I 3.9.B.l. by demonstrating the operability

. . of over-voltage, under-voltage

a. Incoming Power and under-frequency protective instrumentation by performance of

1. From and after the date incoming power a channel calibration including

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is not available from a startup or emer- simulated automatic actuation of gency transformer, continued reactor the protective relays, tripping operation is permissible under this logic and output circuit breakers condition for seven days. At the end and verifying the following set-of this period, provided the second points.

source of incoming power has not been made immediately available, the NRC 1. Over-voltage < 132 VAC, with must be notified of the event and the time delay <,2 sec.

plan to restore this second source.

During this period, the two diesel gener- 2. Under-voltage > 108 VAC, with ators.and associated critical buses must time delay ;[ 2 sec.

be demonstrated to be operable.

3. Under-frequency > 57 Hz. with

2. From and after the date that incoming time delay ;[ 2 sec.

power is not available from both start-up l and emergency transformers (i.e., both failed), continued operation is per-missible, provided the two diesel generators and associated critical buses are demonstrated to be

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- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.B (cont'd.) 4.9.B operable, all core and containment cooling systems are operable, reactor power level is1 reduced to 25% of-the rated and NRC is notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the situation, the precau-tions to be taken during this period and the plans for prompt restora-

. tion of incoming power.

fb . Diesel' Generators

'1. From and after the date that one of

-the diesel generators or an_ associated critical bus is made or found to be inoperable for any reason, continued reactor operation is permissible in accordance with -

Specification 3.5.F.1 if-Specifica-tion 3.9.A.1 is satisfied.

l2. - From and .af ter the date that both diesel generators are made or found to be inoperable for any reason, continued reactor operation is

_ permissible only during the succeeding 2's hours in accordance with' Specification 3.5.F.2 if Specifi-cation-3.9.A.1 is satisified.

3. From and'after the date that one of-the diesel generators or associated critical buses and either the emer-gency or startup transformer power

source are made or.found.to be in-operable for any reason, continued reactor operation is permissible in

! accordance with Specification 3.5.F.1, provided the other off-site source, i startup transformer or emergency i transformer is available and l capable of automatically supplying

, power to the 4160V critical buses and the NRC is notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the occurrence and the plans for restoration of the inoperable compo-nents.

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.. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

~3.9.B.5 (cont'd.) 4.9.B

c. DC Power,-

1. From and after the date that one of the 125 or 250 volt battery systems is made or found to be inoperable for any reason,' continued reactor operation is permissible during the succeeding ten days within electrical safety considera-tions, provided repair work is initiated in the most expeditious manner to return the failed component to an operable state, ,

and Specifications 3.5.A.5 and 3.5.F are satisfied. The NRC shall be notified within 24. hours of the situation, the precautions to be taken during this period and the plans to return the failed components to an operable state.

d. RPS/MG Sets

1. - . With one RPS electric power monitoring

. channel for~an inservice RPS MG set or alternate power supply inoperable, restore the inoperable channel 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 remove the associated RPS MG set or alternate pcwer supply from service.

2. With both RPS electric power monitoring channels for an inservice RPS MG set or alternate power supply inoperable, restore at_least one to operable status within 30 minutes or remove the associated RPS MG set or alternate power supply from service.

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Attachmtnt 2-Revised Technical Specifications for-Core and Containment. Cooling Systems -

Minimum Low Pressure Cooling and Diesel Generator Availability Revised Pages:

121 122 Nebraska '. Public Power District requests a revision to the Technical Specifications to clarify. requirements for suppression chamber operability

~ during: refueling operations. The proposed revision is shown on the attached pages.

The current Technical Specifications are silent with respect to;the ability to conduct refueling operations with the suppression chamber drained. The proposed change permits - refueling . operations with the suppression chamber drained provided an operable core spray or~LPCI system is aligned to take a suction on the condensate storage ' tanks. This is in conformance with

.NUREG-0123, Revision 3, Standard Technical Specification 3.5.3.

Evaluation of this Revision with Respect to 10CFR50.92 A. ' The enclosed Technical Specification change is judged to involve no

. significant. hazards based on the following:

1. Does. the proposed-license amendment involve a significant increase ~

in the1 probability or consequences of an accident previously -

evaluated?

Evaluation:

CNS Technical Specification 3.5.F.6 requires that either the core spray or LPCI system be operable during refueling operations. This

! proposed change clarifies the demonstration of this operability by l specifying an-alternate source of water requirement while conducting o refueling.with the suppression chamber drained. Since this. change

[ does not alter , the availability of adequate emergency cooling

capability whenever irradiated fuel is in the reactor vessel it does j not involve a significant increase in the probability or t

consequences of an accident previously evaluated.

. .. Does the proposed license amendment create the possibility for a new j' or . different kind of accident from any accident previously evaluated?

i-Evaluation:

L Because this change clarifies the requirement for the availability L of emergency cooling capability during refueling operations and is

! in conformance with the GE Standard Technical Specifications it does not create _the possibility for a new or different kind of accident from any accident previously evaluated.

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a.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

-Since this change specifies that an ilternate source of water be available for emergency cooling when conducting refueling operations with the suppression pool drained it ensures that safety margins relat.ive to emergency cooling are not reduced.

B. A'dditional basis for proposed no significant hazards consideration determination:

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48FR14870). The examples include:

"(11) A change that constitutes an additional limitation, restriction, or e-control not presently included in the Technical Specifications . . ."

This Proposed Technical Specification -adds the requirement that the operable LPCI or core spray system be aligned to take a suction on the condensate storage tank to provide an alternate source of water during refueling operations with the suppression chamber drained. This requirement was implied but not specifically stated in the current Technical . Specifications. This addition is thus considered to be encompassed-by the above NRC example.

-^ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.F (cont'd.)' 4.5.F (cont'd.)

3. .Any combination of inoperable ccmpo-nents.in the core and containment cooling systems shall not defeat the capability of_the remaining operable components to fulfill the cooling functions.

4. When irradiated fuel is in the reactor vessel and the' reactor is in the Cold Shutdown Condition, both_ core spray systems,.the LPCI and containment cooling subsystems may be inoperable, provided no work is being done which has the potential for draining the-reactor vessel. Refueling require-ments are as specified in Specifi-cation 3.5.F.6.

5. With irradiated fuel in the reactor vessel, one control rod drive housing may be open.while the suppression chamber is completely drained'provided that:

la . The r(setor vessel head is removed.

b. The spent fuel pool gates are open and the fuel pool water level is maintained at a level > 33 feet.

.c. The condensate transfer system is operable and a minimum of 230,000 gallons of water is in the conden-sate storage tank,

d. The automatic mode of the drywell sump pump is disabled,

e. No maintainance is being conducted which-will prevent filling the suppression chamber to a level above the core spray and LPCI suctions.

f. With the exception of the suppres-sion chamber water supply, both core spray systems and the LPCI system are operable.

g. The control rod is withdrawn to the backseat.

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.F (cont'd) 4.5.F (cont'd)

h. A special flange, capable of sealing a leaking control rod housing, is available for immediate use.

1. The control rod housing is covered with the special flange following the removal of the control rod drive. ,

J. No work is being performed in the vessel while the housing is open.

6. During a refueling cutage, refueling operation may continue with one core spray system or the LPCI system in-operable for a period of thirty days.

Refueling is permitted with the suppression chamber drained provided an operable core spray or LPCI system is aligned to take a suction on the condensate storage tank containing at least 150,000 gallons

(>14 ft. indicated level).

7. The LPCI System is required to be operable while performing training startups at atmospheric pressure at power levels less than 17. of rated thermal power with the exception that the RHR system may be aligned in the shutdown cooling mode rather than the LPCI mode.

G. Maintenance of Filled Discharge Pipe G. Maintenance of Filled Discharge Pipe Whenever core spray subsystems, LPCI The following surveillance requirements subsystems, HPCI, or RCIC are required shall be adhered to, to assure that the to be operable, the discharge piping discharge piping of the core spray from the pump discharge of these sys- subsystems, LPCI subsystems, HPCI and tems to the last block valve shall RCIC are filled:

be filled.

1. Whenever the Core Spray, LPCI, HPCI or RCIC systems are made operable, the discharge piping shall be vented from the high point of the system and water flow observed initially and on a mcnthly basis.

2. The pressure switches which monitor the LPCI, core spray, HPCI and RCIC lines to ensure they are full shall be functionally tested and calibrated every three months.

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Attachmsnt 3 Revised Technical Specifications for Torus Temperature Monitoring System Revised Pages: 65 80 Nebraska Public Power District requests a revision to the Technical Specifications to include primary containment surveillance instrumentation installed as part of a new Torus Water Temperature Monitoring System as shown on the attached pages.

Evaluation of this Revision with Respect to 10CFR50.92 A. The enclosed Technical- Specification change is judged to involve no significant hazards based on the following:

1. Does the proposed licence amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation:

The replacement temperature monitoring system provides upgraded post-accident monitoring capability and does not degrade the function of any safety-related system. It therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Evaluatic,n:

This equipment provides indication only as an aid to operations for mitigation of a LOCA. Previous accident analyses remain bounding with no creation of a new or different kind of accident.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

The margin for 36fety is enhanced because this improved instrumentation provides additional aids for the operator in mitigating the effects of a LOCA.

B. Additional basis for proposed no significant hazards consideration determination:

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (48FR14870). The examples include:

"(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications . . ."

The Proposed Technical Specification adds replacement torus water temperature indication which was installed to comply with the intent of NUREG-0661 and Regulatory Guide-1.97 regarding. monitoring requirements of.

^

suppression pool temperature. Since it reflects the installation of an improved monitoring capability this Technical Specification change is an 4

exemple of the type discussed above.

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COOPER NUCLEAR STATION '

TABLE 3.2.F PRIMARY CONTAINMENT SURVEILLANCE INSTRUMENTATION Minimum Number Action Required When Instrument of Operable Minimum Condition Instrument I.D. No. Range Instrument Channels Not Satisfled (1)

Reactor Water Level NBT-LI-85A -150" to +60" 2 A,B,C NBI-LI-85B -150" to +60" Reactor Pressure RFC-PI-90A 0 - 1200 psig 2 A,B,C RFC-PI-90B 0 - 1200 psig Dryvell Pressure PC-PI-512A 0 - 80 psia 2 A,B,C PC-PR-512B 0 - 80 psia Drywell Temperature PC-TR-503 50 - 170*F 2 A,B,C PC-TI-505 50 - 350*F Suppression Chamber PC-TR-21A 0 - 300 F 2 A,B,C Air Temperature PC-TR-23, Ch I & 2 0 - 400*F hi Suppression Chamber PC-TR-24, Ch I to 16 0 - 250*F 4 A,B,C Water Temperature Suppression Chamber Water 1.evel PC-LI-10 (-4' to +6') 2 A,B,C PC-LR-11 (-4' to +6')

PC-LI-12 -10" to +10" 2 A,B,C.E PC-LI-13 -10" to +10" Suppression Chamber PC-PR-20 0 - 2 psig 1 B,C Pressure Control Rod Position N.A. Indicating Lights 1 A,B,C,D Neutron Monitoring N.A. S.R.M., I.R.M., 1 A,B,C,D LPRM 0 - 100% power Torus to Drywell PC-dPR-20 0 - 2 psid 1 A,B,C,E Differential Pressure Suppression Chamber / PC-PR-20/513 (2) 0 - 2 psig I Drywell Pressure (AP)

COOPER NUCLEAR. STATION TABLE 4.2.F PRIMARY CONTAINMENT SURVEILLANCE INSTRUMENTATION TEST AND CALIBRATION FREQUL,CIES Instrumeat Instrument I.D. No. Calibration Frequency Instrument Check Reactor Water Level NBI-LI-85A -

Once/6 Months Each Shift NBI-LI-85B Once/6 Months Each Shift RFC-PI-90A Once/6 Months Each Shift Reactor Pressure .

RFC-PI-908 Once/6 Months Each Shift PC-PR-512A Once/6 Months Each Shift Drywell Pressure PC-PI-512B Once/6 Months Each Sh'ft Drywell Temperature PC-TR-503 Once/6 Months Each Shift PC-TI-505 Once/6 Months Each Shift PC-TR-21A Once/6 Months Each Shift Suppression Chamber PC-TR-23, Ch. 1&2 Once/6 Months Each Shift Air Temperature g .

PC-TR-24, Ch. I to 16 Once/6 Months Each Shift

? Suppression Chamber Water Temperature l PC-LI-10 Once/6 Months Each Shift Suppression Chamber PC-LR-Il once/6 Months Each Shift Water Level PC-LI-12 Once/6 Months Each Shift  ;

PC-LI-13 Once/6 Months Each Shift Suppression Chamber PC-PR-20 Once/6 Months Each Shift f

Pressure l

l Coatrol Rod Position N.A. N.A. Each Shift Neutron Monitoring (APRM) N.A. Once/ Week Each Chift Torus to Drywell PC-dPR-20 Once/6 Months Each Shift Differential Pressure PC-PR-20/513 (2) Once/6 Months Each Shift Suppression Chamber /

Drywell Pressure (AP)

Attachmsnt 4 Miscellaneous Revised Technical Specifications Nebraska Public Power . District -requests revisions to. the Technical Specifications to make miscellaneous = administrative changes as indicated in the attached pages. These change are judged to involve no significant hazards based on the following:

1. Do the proposed license amendments involve a significant increase in the probability or consequences of an accident previously evaluated?

Evaluation:

Because these changes are of an administrative nature to correct errors or add components to lists they do not involve a significant increese in the probability or consequences of an accident previously evaluated.

2. Do the proposed license amendments create the possibility of a new or

-different kind of accident'from any accident previously evaluated?

Evaluation:

Because these changes are administrative or increase surveillance requirements they do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant. reduction in a margin of safety?

Evaluation:

Since - these changes involve the correction of errors in the current Technical Specifications or constitute an additional limitation they do not involve a significant reduction in a margin of safety.

Additional basis for proposed no significant hazards consideration determination:

The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain e::amples (48FR14870). The examples include: "(1) A purely administrative change to Technical Specifications . . ." and "(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications . . ." The changes listed below are encompassed by these examples. The applicable Federal Register example is indicated for each proposed Technical Specification change.

Federal Register Page Proposed Change Example 111 Corrects columnar typo error in first line. i 4' Deletes reference to reactor pressure less i than 1,000 psig which should have been corrected in LOW LOW set relief. Amendment 83 and makes minor editorial changes.

6 Changes ">" to ">" in Technical Specification ,

i 1.1.A and "<" to "<" in Specification 1.1.B.

7 Changes " values" to "value", i 28- Changes'">" to ">" and adds "2" to seventh column. 1 29 Adds underlining in column titles and deletes. i reference to Note 3 which should have been corrected in LOW LOW set relief Amendmen: 83.

30 Removes Note 3 which was erroneously left in i the LOW LOW set relief Amendment 83.

31 Capitalizes RUN. i 40 Clarifies the bases for the control rod drive i scram system to reflect two vice one scram discharge volume and changes FSAR to USAR.

50 Adds instrumentation inadvertently left out i and 11 of previous LO-LO set relief Amendment 83.

52 and 52b Makes corrections in Table 3.2.A as the i result of the LOW LOW set relief modifications which were inadvertently omitted from Amendment 83.

61 Changes "<" to "<". ,

i 68 Adds instrumentation inadvertently left out i and 11 of previous LO-LO set relief Amendment 83.

83 Corrects typo "n" and corrects description i of RV water level trip functions.

84 Adds " clad" and changes FSAR to USAR. i 133a Changes "followed" to "following". i 134 Changes " operating" to " operation" and adds i "at" in Technical Specification 4.6.B.3.a.

4

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Federal '

Register

-Page Proposed Change y Examp le --

s 1371 Add snubbers as. indicated. 11 Adds caps and underlining.

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159 1 (

161' P, and P should be expressed as psig vice i psia.

"F" 162a Changes to "f". i 'N s , i 166 " Adds " mph" in Technical Specification 4'.7.C. ,

1 167a Corrects spelling of " mode". s i

173 Changes "RPCI" to."HPCI". i >

204 -Adds " spiral" to Technical Specification , i 4.10.A.3.  !

208

~

Wording is added to recognize the use of the , i spiral refueling technique. t 215 Corrects typos in Technical Specification i .

3.12.A.2.b. \ , ,

215d -Changes "not" to no" . 3 i ,g 216b Corrects spelling of " instrumentation". 1 1 h.

1 I

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1

TABLE OF CONTENTS (Cont'd.)

4' Page No.

SURVEILLANCE LIMITING CONDITIONS'FOR OPERATICN REQUIREMENTS

'3.12 ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 4.12 215 - 215f A. Main Control Room Ventilation A 215 B. Reactor Building Closed Cooling Water System B 215b C. Service Water System C 215c D. . Battery Room Vent. D 215e 3.13 RIVER LEVEL . 4.13 216 3.14 FIRE DETECTION SYSTEl! 4.14 216b 3.15 FIRE SUP?hESSION WATER SYSTEM 4.15 216b 3.16 SPRAY AND/OR SPRINKLER SYSTEM (FIRE PROTECTION) 4.16 216e 3.17 CARBON DIOXIDE SYSTEM 4.17 216f 3.18 FIRE HOSE-STATIONS 4.18 216g 3.19 FIRE-BARRIER PENETRATION FIRE SEALS 4.19 216h 3.20 YARD FIRE HYDRANT AND HYDRANT HOSE HOUSE 4.20 2161 5.0 MAJOR DESIGN FEATURES 5.1 Site Features 217 5.2 Reactor- 217 5.3 Reactor Vessel 217 5.4 Containment 217 5.5 . Fuel Storage 218 5.6 Seismic Design 218 5.7 . Barge Traffic 218 s.

'6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 219 6.1.1 Responsibility 219 6.1.2 Offsite 219 6.1.3 Plant Staff - Shift Complement 219 6.1.4 Plant Staff - Qualifications 219a 6.2 , Review and Audit 220 6.2.1.A Station Operations Review Committee (SORC) 220 I A.1 Membership 220 A.2 Meeting Frequency 220 t  : A.3 Quorus 220 A.4 Responsibilities 220

, ;s.

4 A.5 Authority 221

i' A.6 Records 221 4 A.7 Procedures 222 r, 4 L\h_, -111-

K. Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent a margin with normal operation lying below these settings. The margin has been established so that with proper operation of *the instrumentation the safety limits will never be exceeded.

L. Mode - The reactor mode is established by the mode selector switch. The modes include refuel, run, shutdown and startup/ hot standby which are defined as follows:

1. Refuel Mode - The reactor is in the refuel mode when the mode switch is in the REFUEL position. When the mode switch is in the REFUEL position, the refueling interlocks are in service.

2. Run Mode - In this mode the reactor sjstem pressure is at or above 825 psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.

3. Shutdown Mode - The reactor is in the shutdown mode when the mode switch is in the SHUTDOWN position.

l

. 4. Startup/ Hot Standby Mode - In this mode the reactor protection scram trips initiated by the main steam line isolation valve closure are l bypassed, the low pressure main steam line isolation valve closure ,

trip is bypassed, the reactor protection system is energized with .

APRM (15% SCRAM) and IRM neutron monitoring system trips and control rod withdrawal interlocks in service.

M. Operable - Operable means a system or component is capable of performing .

its intended function in its required manner.

N. Operating - Operating means a system or component is performing its l intended functions in its required manner.

O. Operating Cycle - Interval between the end of one refueling outage and l the end of the next subsequent refueling outage.

l P. Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of I the following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the reactor coolant system or. containment, and which are not required to be open during accident conditions, are closed.

. 2.- At least one door in each airlock is closed and sealed.

l l

l l- :

4-L

\

I SAFETY LIMITS (LIMITING SAFETY SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY j Applicability Applicability The Safety Limite established to The Limiting Safety System Settings preserve the fuel cladding integrity appi;' to trip settings of the instru-apply to those variables which cents and devices which are provided monitor the fuel thermal behavior, to prevent the fuel cladding integ-rity Safety Limits from being exceeded. ]

Cbjective Objective The objective of the Safety Limits is to establish limits below which The objective of the Limiting Safe-the integrity of the fuel cladding ty System Settings is to define the is preserved, level of the process variables at which automatic protective action Action is initiated to prevent the fuel cladding integrity Safety Limits l

'f a Safety Limit is exceeded, the from being exceeded, l reactor shall be in at least hot shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Specifications l Specifications A. Trip Settings A. Reactor Pressure >800 psia and The limiting safety system trip settings shall be as specified Core Flov ,)10% of Rated

_ below:

The existence of a minimum critical 1. Neutron Flux Trip Settings i power ratio (MCPR) less than 1.07 l shall constitute violation of the a. APRM Flux Scram Trip i fuel cladding integrity safety. Settine (Run Mode)

B. Core Thermal Power Limit (Reactor When the Mode Switch is l Pressure <800 psia and/or Core in the RUN position, the i Flow <10%) APRM flux scram trip setting shall be:

When the reactor pressure is <800 l

psia or core flow is less than 10% S,<0.66 W + 54%

of rated, the core thermal power shall not exceed 25% of rated where:

thermal power.

S = Setting in percent C. Power Transient of rated thermal l power (2381 MWt) l To ensure that the Safety Limit l established in Specification 1.1.A W = Loop recirculation I and 1.1.B is not exceeded, each flow rate l' rrcent required scram shall be initiated by of rated b . Loop recirculatiot. ; tow its expected scram signal. The 1 Safety Limit shall be assumed to be rate is that recirc exceeded when scram is acconplished ulation flow rate j which provides 100%

by a means other than the expected scram signal, coreflow at 100%

power)

• SAFETY LIMITS l LIMITING SAFETY SYSTEM SETTINGS 1.1 (Cont'd) 2.1.A (Cont'd)

D. Cold Shutdown a. In the event of operation with a maximum fraction of limiting power Whenever the reactor is in density (MFLPD) greater than the the cold shutdown condition fraction of rated power (FRP),

with irradiated fuel in the the setting shall be codified as reactor vessel, the water fcilows:

level shall not be less than 18 in, above the top of the S < (0.66 W + 54%) FRP normal active fuel zone (top MFLPD of active fuel is defined in Figure 2.1.1). where, FRP = fraction of rated thermal power (2381 MWt)

MFLPD = maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/ft for 8x8 fuel.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the l design value of 1.0, in which case the actual operating value wiil be used.

For no combination of loop recirculation flow rate and core thermal pcwer shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

b. APRM Flux Scram Trip Setting (Refuel or Start and Hot Standbv Mode)

When the reactor mode switch is in the REFUEL or STARTUP posi-tion, the APRM scram shall be l

set at less than or equal to I

15% of rated power.

c. IRM The IRM flux scram setting shall f be <120/125 of scale.

O A

COOPER NUCLEAR STATION TABLE 3.1.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION REyUIREMENTS Minimum Number Action Required ~

Applicability Conditions of Operable When Equipment Reactor Protection Mode Switch Position Trip Level Channels Per Opernbility is System Trip Function Shutdown Startup Refuel Run Setting Trip Systems (I) Not Assured (1)

Mode Switch in Shutdown X(7) X X X 1 A Manual Scram X(7) X X X 1 A IRM (17) X(7) X X (5) < 120/125 of in- 3 A High Flux dicated scale Inoperative X X (5) 3 A APR11 (17) X < (0.66W+54%) FRP 2 A or C liigh Flux (Flow biased) ~ (14) NFLPD h liigh Flux X(7) X(9) X(9) (16) <

_ 15% Rated Power 2 A or C l

Inoperative X(9) X(9) X (13) 2 A or C Downscale (11) X(12) > 2.5% of indi- 2 A or C cated scale liigh Reactor Pressure X(9) X(10) X < 1045 psig 2 A NHI-PS-55 A,B,C, 6 D liigh Drywell Pressure X(9)(8) X(8) X < 2 psig 2 A or D PC-PS-12 A It,C, & D l

Reactor Low Water Level X X X >+ 12.5 in. Indi- 2 A or D NBI-LIS-101 A,B,C, & D _ cated level Scram Discharge Instrument Volume X X(2) X < 92 inches .

3 (18) A liigh Unter Level CRD-LS-231 A & B CRD-LS-234 A & B CRD-LT-231 C & D l CRD-LT-234 C & D

COOPER NUCLEAR STATION

• TABLE 3.1.1 (Page 2)

REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS Minimum Number Action Required Applicability Conditions of Operable When Equipment Reactor Protection Mode Switch Position Trip Level Channels Per Operability is System Trip Function Shutdown Startup Refuel Run Setting Trip Systems (1) Not Assured (1)

Main Steam Line liigh Radiation X(9) X < 3 Times nor-al 2 A or D RMP-RM-251, A,B,C, & D full power back ground Main Steam Line Isolation Valve Closure X(6) (9) X(6) < 10% of valve 4 A or C l MS-LMS-86 A,B,C, & D closure 4 A or C MS-LMS-80 A,B,C, & D Turbine Control Valve X(4) > 1000 psig turbine 2 A or B l Fast Closure control fluid TCF-63/0PC-1,2,3,4 ra

, Y Turbine Stop Valve Closure X(4) <10% of valve 2 A or B SVOS-1(1), SVGS-1(2) Closure SVOS-2(1), SVOS-2(2)

Turbine First Stage Permissive MS-PS-14 X(9) X < 30% first 2 A or B A,B,C, & D stage press.

4

NOTES FOR TABLE 3.1.1

1. There shall be two operable *r tripped trip systems for each function. If the minimum number of operable instrument channels for a trip system cannot

be met, the affected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.

A. Initiate insertion of operable rods and complete insertion of all operable rods within four hours.

B. Reduce. power to less than 30% of rated.

C. Reduce power level to IRM range and place mode switch in the Startup position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and depressurize to less than 1000 psig. ,

i u D. Reduce turbine load and close main steam line isolation valves within ,

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

2. Permissible to bypass, with control rod block, for reactor protection

+

system reset in refuel and shutdown positions of the reactor mode switch.

3. This note deleted.

4. Permissible to bypass when turbine first stage pressure is less than 30%'of full load.

5. IRM's are bypassed when APRM's are onseale and the reactor mode switch is in the run position.

6. The design permits closure of any two lines without a full scram being initiated.

1

7. When the reactor is suberitical, fuel is in the vessel, and the reactor water I

' l temperature is less than 212*F, only the following trip functions need to'be operable:

a. Mode switch in shutdown. '

b.- Manual scram.

c. IRM high flux. 120/125 indicated scale. l

d. APRM (15%) high flux scram.

8. Not required to be operable when primary containment integrity is not required.

I 9. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).

10. Not. required to be operable when the reactor pressure vessel head is not bolted to the vessel.

4 L

. . _ _ . . -. . _ . , _ . . . . _ , . ._. _..__.. _ _ _ .. .. ._ ~ _ _ _ , .

11. Tha APRM downscale trip function is only active when the reactor modo switch is in RUN. l

12. The APRM dcwnscale trip is automatically bypassed ....- *he mode switch is not in RUN.

13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 11 operable LPRM detectors to an APRM.

14. W is the recirculation flow in percent of rated flow.

15. This note deleted.

16. The 15% APRM scram is bypassed in the RUN mode.

17. The APRM and IRM instrument channels function in both the Reactor Protection System and Reactor Manual Control System (Control Rod Withdraw Block, Section 3.2.C.). A failure of one channel will affect both of these systems.

18 .. The minimum number operable associated with the Scram Discharge Instrument Volume are three instruments per Scram Discharge Instrument Volume and three level devices per RPS channel.

• LYMfTfNC CORDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS J

3.1 - BASES'(cont'd.) 4.1 BASES (cont'd.)

against short reactor periods in revealed only on test. Therefore, these ranges. it is necessary to test them periodi-cally.

The control rod drive scram system-

.is designed so that all of the water A study was conducted of the instru-which is discharged from the reactor mentation channels included in the by a scram can be accommodated in the Group (B) devices to calculate their discharge piping. The scram discharge " unsafe" failure rates. The analog volumes accommodate in excess of 18 devices (sensors and amplifiers) gallons of water in each volune and are predicted to have an unsafe -6 are the low points in the piping, failure rate of less than 20 X 10 No credit was taken for this volume failures / hour. The bi-stable trip in the design of the discharge piping circuits are predicted to have an as concerns the amount of water which unsafe _jailurerateoflessthan must be accommodated during a scram. 2 X 10 failures / hour. Consider-ing the two hour monitoring interval During normal operation the dis- for the analog devices as assumed charge volume is empty; however, above, and a weekly test interval should it fill with water, the water for the bi-stable trip circuits, discharged to the piping from the the design reliability goal of reactor could not be accommodated which 0.99999 is attained with ample margin.

would result in sicw scram times or partial control rod insertion. To pre- The bi-stable devices are monitored

-clude this occurrence, diverse indi- during plant operation to record their cation.(two level switches and two failure history and establish a test level transmitters for each discharge interval using the curve of Figure volume) has been pro'vided in the 4.1.1. There are numerous identical instrument volumes which alarm and bi-stable devices used throughout scram the reactor when the volume of the plant's instrumentation system; water reaches 92 inches in either therefore, significant data on the volume. -As indicated above, there failure rates for the bi-stable devices is sufficient volume in the piping should be accumulated rapidly.

to accommodate the scram without impairment of the scram times or The frequency of calibration of the

! amount of insertion of the control APRM Flow Biasing Network has been rods. This function shuts the reactor established as each refueling out-l down while sufficient volume remains age. The flow biasing network is to accommodate the discharged water functionally tested at least once l

, and precludes the situation in which per month and, in addition, cross ~

n' scram would be required but not be calibration checks of the flow able to_ perform its function adequately. input to the flow biasing network can be made during the functional j A source range monitor (SRM) system is test by direct meter reading. There

! also provided to supply additional are several instruments which must l neutron level information during start- be calibrated and it will take sev-up but has no scram functions (refer- eral days to perform the calibration L l ence paragraph VII.5.4 USAR). Thus, of the entire network. While the r

the IRMs and APRMs are required in the calibration is being performed, a l l. " Refuel" and "Startup/ Hot Standby" modes, i In the power range, the APRM system l provides required protection (refer-l l

1

COOPER NUCLEAR STATION TABLE 3.2.A (Page 1)

PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION Minimum Number Action Required When Instrument of Operable Components Component Operability Instrument I.D. No. Setting Limit Per Trip System (1) is Not Assured (2)

Main Steam Line liigh RMP-RM-251, A,B,C,&D < 3 Times Full Power 2 A or B Rad.

Reactor Low Water Level NBI-LIS-101, A,B,C,6D >+12.5" Indicated Level 2(4) A or B Reactor low Low Water NEI-LIS-57 A & B #2 >-37" Indicated Level 2 A or B Level NBI-LIS-58 A & B #2 Reactor Low Low Low Water NBI-LIS-57 A & B #1 >-145.5" Indicated Level 2 A or B Level NBI-LIS-58 A & B #1 Main Steam Line Leak MS-TS-121, A,B,C,6D < 200*F 2(6) B Detection 122, 123, 124, 143, 144, 145, 146, 147, 148, 149, 150 hi Main Steam Line Illgh MS-dPIS-Il6 A,B,C,&D < 140% of Rated Steam 2(3) B Flow 117, 118, 119 Flow Main Steam Line Low MS-PS-134, A,B,C,6D > 825.psig 2(5) B Pressure liigh Drywell Pressure PC-PS-12, A,B,C,6D < 2 psig 2(4) A or B liigh Reactor Pressure RR-PS-128 A & B < 75 psig 1 D l

Main Condenser Low MS-PS-103, A,B,C,6D > 7" Hg (7) 2 A or B Vacuum Reactor Water Cleanup RWCU-dPIS-170 A & B < 200% of System Flow I C System liigh Flow

. NOTES FOR' TABLE 3.2.A

1. Whenever Primary Containment integrity is required there shall be two operable or tripped trip systems tor each function.

2. If the minimum number of operable instrument channels per trip system requirement cm.not be met by a trip system, that trip system shall be tripped. If the requirements cannot be met by both trip systems, the appropriate action listed below shall be taken.

A. Initiate an orderly shutdown and have the reactor in a cold shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Initiate an orderly load reduction and have the Main Steam Isolation Valves shut within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. Isolate the Reactor Water Cleanup System.

D. Isolate the Shutdown Cooling System.

3. Two required for each steam line.

4. These signals also start the Standby Gas Treatment System and initiate Secondary Containment isolation.

5. Not required in the refuel, shutdown, and startup/ hot standby modes (interlocked with the mode switch).

6. Requires one channel from each physical location for each trip system.

7. Low vacuum isolation is bypassed when the turbine stop is not full open, manual bypass switches are in bypass and mode switch is not in RUN.

3. The instruments on this table produce primary containment and system isolations. The following listing groups the system signals and the system isolated.

Group 1 Isolation Signals:

1. Reactor Low Low Low Water Level (>-145.5 in.) l

2. Main Steam Line High Radiation (3 times full power background)

3. Main Steam Line Low Pressure (>825 psig in the RUN mode)

4. Main Steam Line Leak Detection (<200*F)  !

5. Condenser Low Vacuum (>7" Hg vacuum)

6. Main Steam Line High Flow (<140% of rated flow) l Isolations:

1. MSIV's

2. Main Steam Line Drains

r

. NOTES FOR TABLE 3.2.A (cont'd.)

Isolations

1. Secondary Containment Isolation

2. Start Standby Cas Treatment System Group 7 Isolation Signals:

1. Reactor Low Low Water Level Q-37 in)

2. Main Steam Line High Radiation (<3 times full power background)

Isolatiens:

1. Reactor Water Sample Valves

-52t-

TABLE 3.2.C -

CONTROL ROD WITilDRAWAL BLOCK INSTRUMENTATION 4

4 Minimum Number Of Function Trip Level Setting Operable Instrument Channels / Trip System (5)

APRM Upscale (Flow Bias) j[ (0.66W + 4'2%) FRP (2) 2(1)

APRM Upscale.(Startup) ;c 12% MFLPD 2(1) 1 APRM Downscale (9)

> 2.5% 2(1) -l APRM Inoperative (10b) 2(1)

RBM Upscale (Flow Bias) j[ (0.66W + 40%) (2) I RBM Downscale (9) j; 2.5% 1 RBM Inoperative (10c) 1.

IRM Upscale (8) ji 108/125 of Full Scale 3(1)

IRM Downscale (3)(8) 2;:2.5% 3(1) l T IRM Detector Not Full In (8) 3(1) i 1

IRM Inoperative (8) (10a) 3(1) 5 SRM Upscale (8) j[ 1 x 10 Counts /Second 1(1)'(6)

, SRM Detector Not Full In (4)(8) (> 100 cps) 1(1)(6)

SRM Inoperative (8) (10a) 1(1)(6)

Flow Bias Comparator ji 10% Difference In Recire. Flows 1

Flow Bias Upscale /Inop. ji 110% Recire. Flow 1 1

SRM Downscale (8)(7) j; 3 Counts /Second (11) .

1(1)(6) 4 SDV Water Level liigh j[46 inches 1(12) i CRD-231E, 234E t

l

~

• COOPER NUCLEAR STATION TABLE 4.2.A (Page 1)

PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION SYSTEM TEST AND CALIBRATION FREQUENCIES Instrument j Item . Item I.D. No. Function Test Freq. Calibration Freq. ' Check Instrument Channels I Reactor, Low Water Level NBI-LIS-101, A,B,C,&D. Once/ttonth (1) Once/3 Months once/ Day i Reactor Low Low Water Level NBI-LIS-57, A & B #2 Once/ Month (1) Once/3 Months Once/ Day NBI-LIS-58, A & B #2 Reactor Low Low Low Water Level NBI-LIS-57, A & B #1 Once/ Month (1) Once/3 Months once/ Day-

NBI-LIS-58, A & B #1 Main Steam Line Leak MS-TE-121, A,B,C,6D Once/ Month (1) Once/ Operating None i Detection 122, 123, 124, 143, 144, . Cycle

$' 145, 146, 147, 148, 149, 150 l Main Steam Line Iligh Flow MS-dPIS-Il6, A,B,C,6D Once/ Month (1) Once/3 Months None i_ 117 , once/ Month (1) Once/3 Months None 118 Once/ Month (1) Once/3 Months None l 119 once/ Month (1) once/3 Months None f Main Steam Line Low Press. MS-PS-134, A,B,C,6D Once/ Month (1) Once/3 Months None liigh Reactor Pressure RR-PS-128, A & B Once/ Month-(l) Once/3 Months None Condenser Low Vacuum MS-PS-103, A,B,C,&D Once/ Month (1) Once/3 Months None ,

Reactor Water C.U. liigh Flow RWCU-dPIS-170, A & B Once/ Month (1) Once/3 Months None Reactor Water C.U. liigh Space .KWCU-TS-150 A-D, 151, 152, once/ Month (1) Once/ Operating- None

, Temp. 153,.154, 155, 156,,157, Cycle

! 158, 159, RWCU-TS-81, A,B,E,F i RWCU-TS-81 C,D,G,Il 1.

. 3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious con-sequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the l protective instrumentation when required even during periods when portions of such systems are out of service for maintenance, and (2) to prescribe l

the trip settings required to assure adequata performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.

The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuatior. of the safety system involved and exposure to abnormal situations.

A. Primary Containment Isolation Functions Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2.A which senses the conditions for which isola-tion is required. Such instrumentation must be available whenever primary containment integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation, set to trip at 176.5" (+12.5") above the top l

of the active fuel, closes all isolation valves except those in Groups 1, 4, and 5. Details of valve grouping and required closing times are given in Specification 3.7. For valves which fsolate at this level this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time. Required closing times are less than this.

The low low reactor water level instrumentation is set to trip when reactor water lesel is 127" (-37") above the top of the active fuel. This trip closas Recire Sample Valves (Group 7) and initiates the HPCI and RCIC. The low low low reactor water level instrumentation is set to trip when the water level is 19" (-145") above the top of the active fuel. This trip closes Main Steam Line Isolation Valves (Reference 1), Main Steam Drain Valves, and activates the remainder of the CSCS subsystems, and starts the emergency diesel generators. These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished,

c 3.2 Bg ESj (Cont'd) and the guidelines of 10CFR100 will not be exceeded. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are~ initiated in time to meet the above criteria. Reference Paragraph VI.S.3.1 USAR. l The high drywell pressure instrumentation is a diverse signal for mal-functions to the water level instrumentation and in addition to initiating CFCS, it causes isolation of Group 2 and 6 isolation valves. For the breaks discussed above, this instrumentation will generally initiate CSCS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also. The water level instrumen-tation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instru-

. mentation is to detect a break in'the main steam line. For the worst case of accident, main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel clad temperatures peak at approximately l 1000*F and release of radioactivity to the environs is below 10CFR100 l

guidelines. Reference Section XIV.6.5 USAR. {

Temperature monitoring instrumentation is provided in the main steam .

tunnel and along the steam line in the turbine building to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves. See Spec. 3.7 for Valve Group. The setting is 200*F for the main steam leak detection system.

I For large breaks, the high steam flow instrumentation is a backup t'o the temp. instrumentation.

High radiation monitors in the main steam tunnel have been provided to ,

l detect gross fuel failure as in the control rod drop accident. With the established setting cf 3 times normal background, and main steam line isolation valve cloaure, fission product release is limited so that 10CFR100 guidelines are not exceeded for this accident. Reference Sec-l tion XIV.6.2 USAR. h Pressure instrumentation is provided to close the main steam isolation val"es in RUN Mode when the main steam line pressure drops below Speci-

-fication 2.1.A.6. The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not in the RUN L Mode is less severe than the loss of feedwater analyzed in Section XIV.5 of the USAR, therefore, closure of the Main Steam Isolation valves for j thermal transient protection when not in RUN mode is not required.

The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar to that for the HPCI. The trip settings are such that core uncovery is prevented and fission product release is within limits, l

l

o LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6'(cont'd.) 4.6 (cont'd.)

B. Coolant Chemistry B. Coolant Chemistry

1. The reactor coolant radioactivity 1.a. A sample of reactor coolant shall be concentration shall be maintained collected and analyzed for gross within the following limits: gamma activity as follows:

a. Whenever the reactor is critical, 1. At least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> whenever the reactor coolant activity shall the reactor is critical, not exceed the equilibrium value of ~

3.1 pC1/gm of dose equivalent I-131. 2. Prior to reactor startup.

b. The limit of 3.6.B.1.a above may be 3. In the STARTUP mode, at 4-hour exceeded by a factor of 10 or less intervals following a power for a maximum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following change exceeding 5% of rated power transients. The reactor shall power in one hour or less, not be operated more than 5% of its annual power operation under this 4. In the RUN mode, at 4-hour inter-exception. vals following a power change

[

exceeding 20% of rated power

c. If the iodine concentration in the in one hour or less, coolant exceeds the equilibrium limit by a factor greater than 10, the 5. At 4-hour intervals following reactor shall be shutdown in an an off-gas activity increase of orderly manner and in the cold shut- 10,000 pC1/sec measured at the down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and SJAE.

the steam line isolation valves shall be closed. 6. At 4-hour intervals whenever measurements indicate the equilibrium iodine concentration limit of 3.6.B.1 is exceeded, until a stable value below the equilibrium limit is established.

The samples required in 4.6.B.1.a.3, 4, and 5 shall be collected for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> but nay be discontinued if the reactor coolant concentration is shown to be less than 1% of the equilibrium value specified in 3.6.3.1 or when a stable iodine con-centration below the limiting equilibrium value is established.

Whereas a single measurement may be used to show an activity level below 1%, at least 3 consecutive samples with the last 2 yielding activities below the equilibrium value are required to establish a stable concentration below the

- equilibrium limit.

-133a-

LIMITING CONDITIONS FOR OPERATION fSURVEILLANCEREQUIREMENTS 3.6.B. (cont'd) 4.6.B.1 (cont'd) l

2. Prior to startup and during the b. If the gross activity counts of a l operation of the reactor up to sample indicate an activity con-10% of rated power, and during centration above 3.1 uC1/gm of hot standby, the reactor coolant dose equivalent 1-131, an isotopic shall not exceed the followine, analysis shall be performed and limits: quantitative measurements made to determine the dose equivalent

a. Conductivity < 5 umho/cm at 25 C I-131 concentration,

b. Chloride 0.1 ppm c. An isotopic analysis of a reactor coolant sample shall be made at The reactor shall be shut devn least once per month, if pH is <5.6 or >8.6 for a 24-hour period. 2. Reactor coolant shall be continuously, monitored for conductivity.

3. During reactor operation in excess of 10% of rated power, the 3. Prior to startup, during the operation reactor coolatt shall not exceed of the reactor and during hot the following limits: standby, a sample of the reactor

a. Conductivity I umho/cm at 25 C

a. At least every 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> for

b. Chloride 0.2 ppm conductivity and chloride ion content when the continuous conductivity

4. During the reactor operation in monitor reading is <0.7 umho/cm at

~ i excess of 10% of rated power, 25 C. l the reactor coolant may exceed the limits of Paragraph 3.6.B.3 b. At least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for only for the time limits specified conductivity and chloride ion content here. If these time limits or the when the continuous conductivity following maximum limits are exceeded, monitor reading is $0.7 but the reactor shall be shutdown -<2.0 umho/cm at 25 C.

and placed in the Cold Shutdown Condition. c. At 1 cast every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for conductivity anc chloride ion

a. Conductivity~

Time above 1 umho/cm content when the continuous at 25 C, 2 weeks / year conductivity monitor reading is >2 Maximum limit-10 but -<3.5 umho/cm at 25 C.

umho/cm at 25 C

d. At least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for

b. Chloride Time above 0.2 ppm, conductivity, chloride ion content, 2 weeks / year and pH, when the continuous Maximum limit-0.5 ppm conductivity monitor reading is

>3.5 umho/cm at 25 C or when the The reactor shall be shut down if continuous conductivity monitor is pH is <5.6 or >8.6 for a 24-hour inoperable, period.

4. When the reactor is not pressurized,

5. When the reactor is not pressurized a sample of the reactor coolant shall (i.e. at or below 212 F), reactor be analyzed at least every 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> coolant shall be maintained below for conductivity and chloride ion the following limits: content.

s. Conductivity 10umho/cm at 25 C

b. Chloride 0.5 ppm

-134- , __ -_. .

, Table 3.6.3 INACCESSIBLE SAFETY PEpf.TED MECHA9ICAL SHOCK SUPP2ESSORS (SNUBBERS) (cont'd)

Snubber Location RF-SNUB-(RF-S17) DW-921 RF-SNUB-(RF-S18) DW-921 RF-SNUB-(RF-S19) DW-921 RF-SNUB-(RF-S8) DW-921 RF-SNUB-(RF-S9) DW-921 RHR-SNUB-(RH-SIO) DW-901 RHR-SNUB-(RH-S11) DW-901 RHR-SNUB-(RH-S13) DW-921 i RHR-SNUB-(RH-5 F4 DW-921 RHR-SNUB-(RH-14A) DW-901 l RHR-SNUB-(RH-S15) DW-921 RHR-SMUB -(RH-S 16) DW-901 RHR-SNUB-(RH-S17) DW-901 RHR-SNUB-(RH-S18) DW-901 RHR-SNUS-(RH-S19) DW-901 RHR-S NU B-(RH-53) DW-FLG AREA RHR-SNUB-(RH-S4) DW-FLG AREA RHR-SNUB-(RH-SS) DW-921 RHR-SNUB-(RH-S6) DW-921 RHR-SNUB-(RH-S67) DW-901 RHR-SNUB-(RH-S68) DW-901 RiiR-SNUB-(RH-S 69A) -

DW-901 RHR-SNUB-(RH-S69B) DR-901 RHR-S!UB-(RH-S7)

• DW-921 RHR'-SNUB-(RH-S70) DW-901 RHR-SSUB-(RH-S71) DW-901 RHR-SNUB-(RH-S72) DW-901 RHR-SNUB-(RH-S72A) DW-901 RHR-SNUB-(RH-S73) DW-901 RHR-SNUB-(RH-SBA) DW-901 RHR-SNUB-(RH-S8B) DW-901 RHR-SNUB-(RH-S8C) DW-901 l RHR-SNUB-(RH-S9) DW-901 RR-SNUB-(S S-1 A) DW-888 RR-SNUB-(SS-1B) DW-888 RR-SNUB-(SS-2A) DW-888 RR-SNUB-(SS-2B) DW-883 RR-SNUB-(SS-3A1) DW-901 RR-SNUB-(SS-3A2) DW-901 RR-SNUB-(SS-3B1) DW-901 RR-SNUB-(SS-332) DW-901 RR-SNUB-(S S-4 A) DW-901 RR-SNUB-(SS-4B) DW-901 RR-S NUB-(S S-5 A) DW-888 RR-SNUB-(SS-5B) DW-888 RR-SNUB-(SS-8A1) DW-901 RR-SNUB-(SS-8A2) DW-901 RWCU-SNUB-(CU-S3A) DW-921 RWCU-SNUB-(CU-S3B) DW-921

-1371-

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability: Applicability:

Applies to the operating status of Applies to the primary and secondary the primary and secondary contain- containment integrity, ment systems.

Objective: Objective:

To assure the integrity of the pri- To verify the integrity of the primary mary and secondary containment systems. and secondary containment.

Specification: Specification:

A. Primary Containment A. Primary Containment

1. Suppression Pool 1. Suppress ~on Pool At any time that the nuclear system a. The suppression pool water level is pressurized abcve atmospheric and temperature shall be checked pressure or work is being done once per day, which has the potential to drain the vessel, the suppression pool b. Whenever there is indication of water volume and temperature shall relief valve operation or testing be maintained within the following which adds heat to the suppression limits except as specified in pool, the pool temperature shall 3.7.A.2. and 3.5.F.5. be cont'nuelly monitored and also bserved and logged every 5

a. Minimum water volume - 87,650 ft 3 minutes until the heat addition 3

b. Maximum water volume - 91,000 ft

c. Whenever there is it.dication of

c. Maximum suppression pool temperature relief valve operatioa with the during normal power operation - 95 F, temperttura of the suppression pool reaching 160 F or more and l d. During testing which adds heat to the primary coolant system pres-the suppression pool, the water sure greater than 200 psig, an temperature shall not exceed 10 F external visual examination of l above the normal power operation the suppression chamber shall limit specified in c. above. In be conducted before resuming connection with such testing, the power operation.

pool temperature must be reduced to

, below the normal pcwer operation d. A visual inspection of the l limit specified in c. above within suppression chamber interior, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. including water line regions, shall be made at each major

e. The reactor shall be scrammed from refueling outage.

any operating condition if the pool temperature reaches 110 F. Power

( operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in c.

above.

l l

1

-159-

1 I

LIMITING CONDITIONS FOR OPERATION l SURVEILLANCE REOUIREMENTS 3.7.A-(cont'd.) 4.7.A.2.b. (cont'd.) I l

where i P = peak accident pressure, 58 psig l P = appropriately measured test pres-sures (psig)

{

b for em > 0.7 b

am

c. The ILRT's shall be performed at the following minimum frequency:

1. Prior to initial unit operatfor.

2. At approximately three and

, one-third year intervals so that any ten-year interval would include four ILRT's. These intervals may be extended up to eight months if necessary to coincide with refueling outage.

d.

{hemeasuredleakagerates.ghall am, t and be l 0.75 a for the reduced pressure tests

- and peak pressure test respectively.

e. Except for the initial ILRT, all ILRT's shall be performed without any pre-liminary leak detection surveys and leak repairs immediately prior to the test. If an ILRT has to be ter-minated due to excessive leakage through identified leakage paths, the leakage through such paths shall be determined by a local leakage test and recorded. After repairs are made another ILRT shall be conducted.

If en ILRT is completed but the acceptance criteria of Specification 4.7.A.2.d is not satisfied and repairs are necessary, the ILRT need not to I

-161-

LIMITING CONDITIONS FOR OPERATION I SURVEILLANCE REOUIREMENTS 3.7.A (Cont'd) 4.7.A.2.f (cont'd) l

4. Main steam line and feedwater line expansion bellows as specified in Table 3.7.3 shall be tested by pressurizing between the laminations of the bellows at a pressure of 5 psig.

This is an exemption to Appendix J of 10CFR50.

5. The personnel airlock shall be tested at 58 psig at intervals no longer than six months. This testing may be extended to the next refueling catage (not to exceed 24 months) provided that there have been no airlock openings since the last successful test at 58 psig. In the event the personnel airlock is not opened between refueling outages, it shall be leak checked at 3 psig at intervals no longer than six months.

Within three days of opening (or every three days during periods of frequent opening) when containment integrity is required, test the

. personnel airlock at 3 psig. This is an exemption to Appendix J of 10CFR50.

g. Continuous Leak Rate Monitor When the primary containment is inerted the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements. This monitoring system may be taken out of service for maintenance but shall be returned to service as soon as practicable.

h. Drywell Surfaces The interior surfaces of the drywell and torus shall be visually inspected each operating cycle for evidence of torus corrosion or leakage.

J

-162a-

LIMITING CORDITIONS FOR OPERATTON SURVEILLANCE REOUIREMENTS 3.7.C (cont'd.) 4. 7.C (cont 'd.)

n. The reactor is suberitical and Speci- a. A preoperational secondary containment fication/3.3.A is met, capability test shall be conducted after isolating the reactor building and

b. The reactor water temperature is belrw placing either standby gas treatment 212'F and the reactor coolant system system filter train in operation. Such is vented. tests shall demonstrate the capability to maintain 1/4 inch of sater vacuum

c. No activity is being performed which under calm wind (2<U<5 mph) conditions l can reduce the shutdown margin below with a filter train flow rate of n'ot that specified in Specification 3.3.A. more than 100% of building volume per day. (u= wind speed)

d. Icradiated fuel is not being handled in the secondary containment, b. Additional tests shall be performed during the first operating cycle under

e. If secondary containment integrity an adequate number of different envir-cannot be maintained, restore onmental wind conditions to enable secondary containment integrity valid extrapolation of the test results, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or;

c. Secondary containment capability to

a. Be in at least Hot Shutdown maintain 1/4 inch of water vacuum within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and under calm wind (2<U<5 mph) conditions in cold shutdown within the with a filter train flow rate of not following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. more than 100% of building volume per day, shall be demenstrated at

b. Suspend irradiated fuel handling each refueling outage prior to operations in the secondary con- refueling.

tainment and all core alterations and activities which could reduce d. After a secondary containment viola-the shutdown margin. The pro- tion is determined, the standby gas visions of Specification 1.0.J treatment system will be operated are not applicable, immediately after the affected zones are isolated from the romainder of the secondary containment to confirm its ability to maintain the remainder of the secondary containment at 1/4 i

inch of water negative pressure under calm wind conditions.

D. Primary Containment Isolation Valves D. Primary Containment Isolation Valves

1. During reactor power operating condi- 1. The primary containment isolation tions, all isolation valves listed in valves surveillance shall be performed Table 3.7.1 and all instrument line as follows:

flow check valves shall be operable except as specified in 3.7.D.2. a. At least once per operating cycle the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.

-166-

l

. LI.MITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7 (cont'd.) 4.7 (cont'd.) )

l E. Drvuell-Suppression Chamber E. Drywell-Suppression Chamber )

Differential Pressure Differential Pressure 1

1

1. Differential pressure between the 1. The pressure differential l drywell and suppression chamber between the drywell and l shall be maintained at equal to suppression chamber shall l or greater than 1.0 psid except be recorded at least once ,

as specified in a, b, and c below. each shift. I l

a. This differential shall be established within 26 haurs after placing the mode switch in run,

b. This differential may be de-creased to less than 1.0 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to placing I mode switch in refuel or shut-l down.

c. This differential may be decreased to less than 1.0 psid for a maximum of four (4) hours during required operability testing of the HPCI system pump, the RCIC system pump and the drywell- i pressure suppression chamber vacuum breakers.

2. If the differential pressure of specification 3.7.E.1 cannot be maintained, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initi-ated and the reactoc shall be in Hot Standby in six (6) hours and in a Cold Shutdown condition within l the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

3. The specifications of 3.7.E.1 and l l 3.7.E.2 are not applicable during a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> continuous period l between the dates of March 22, 1982 and March 25, 1982.

-167a-

t

• TABLE 3.7.4 PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO. VALVE NUMBERS MEDIA X-7A MS-AO-80A and MS-AO-86A, Main Steam Isolation Valves Air X-7B MS-AO-80B and MS-AO-86B, Main Steam Isolation Valves Air X-7C MS-AO-80C and MS-AO-86C, Main Steam Isolation Valves Air X-7D MS-A0-80D and MS-AO-86D, Main Steam Isolation valves Air X-8 MS-MO-74 and MS-MO-77, Main Steam Line Drain Air X-9A- RF-15CV and RF-16CV, Feedwater Check Valves Air X-9A RCIC-AO-22, RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater Air X-9B RF-13CV and RF-14CV, Feedwater Check Valves Air X-9B HPCI-AO-18 and HPCI-M0-57, HPCI Connection to Feedwater Air X-10 RCIC-MO-15 and RCIC-MO-16, RCIC Steam Line Air X-ll HPCI-M0-15 and HPCI-MO-16 HPCI Steam Line Air l X-12 RHR-M0-17 and RHR-MO-lS, RHR Suction Cooling Air X-13A RHR-MO-25A and RHR-M0-27A, RHR Supply to RPV Air X-13B RHR-MO-25B and RHR-MO-27B, RHR Supply to RPV Air X-14 RWCU-MO-15 and RWCU-MO-18, Inle: to RWCU System Air X-16A CS-MO-llA and CS-M0-12A, Core Spray to RPV Air X-16B CS-MO-llB and CS-MO-12B, Core Spray to RPV Air X-17 RHR-MO-32 and RHR-MO-33, FPV Head Spray Air X-18 RW-732AV and RW-733AV, Drywell Equipment Sump Discharge Air X-19 RW-765AV and RW-766AV, Drywell Floor Drain Sump Discharge Air X-25 PC-232MV and PC-238AV, Purge and Vent Supply to Drywell Air X-25 ACAD-1305MV and ACAD-1306MV, Supply to Drywell Air X-26 PC-231MV and PC-246AV, Purge and Vent Exhaust from Drywell Air X-26 ACAD-1310MV, Bleed from Dryvell Air

-173-

= __

• LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.10.A , Refueling Interlocks 4.10.A (Cont'd)

3. The fuel grapple hoist load switch 3. Whenever the reactor is in the refuel shall be set at -< 650 lbs. mode and rod block interlocks are being bypassed (for spiral core (

unloading), one licensed operator and one member of the reactor engineering staff will verify that all fuel has been removed before the corresponding control rod is withdrawn.

4 If the frame-mounted auxiliary hoist, 4. Following the withdrawal and bypassing the monorail-mounted auxiliary hoist, of a control rod, two licensed operator:.

or the service platform hoist is to will verify that the interlock bypassed be used for handling fuel with thw is on the correct control rod.

head off the reactor vessel, the load limit switch on the hoist to be used shall be set at < 400 lbs.

5. A maximum of two nonadjacent control 5. Prior to loading fuel in a control cell rods may be withdrawn from the core (using the spiral reload technique),

for the purpose of performing the control room operator and a license; control rod and/or control rod drive operator and a member of the reactor maintenance, provided the following engineering staff on the refueling floo:

conditions are satisfied; shall verify that the control rod is inserted in the c' ell to be loaded,

a. The reactor mode switch shall be -

locked in the " refuel" position.

The refueling interlock which pre-vents more than one control rod from being withdrawn may be by-passed for one of the control rods on which maintenance is being per-formed. All other refueling in-terlocks shall be operable.

b. A sufficient number of control rods shall be operable so that the core can be made suberitical with

, the strongest operable control rod fully withdrawn and all other operable control rods fully in-serted, or all directional control valves for remaining control rods shall be disarmed electrically and sufficient margin to criticality shall be demonstrated.

c. If maintenance is to be performed on two control rod drives, they must be separated by more than two control cells in any direction.

d. An appropriate number of SRM's are available as defined in Specification 3.10.B.

-W- ____. ._ __._ _ __ _

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

v

-,: 3.10 BASES (Cont'd)

During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time. The mainten-ance is performed with the mode switch in the " refuel" position to .

provide the refueling interlocks normally available during refueling operations. In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one control rod

, from being withdrawn at the same time. The requirement that an adequate

^

shutdown margin be demonstrated or that all remaining control rods have ,

their directional control valves electrically' disarmed ensures that inadvertent criticality cannot occur during this maintenance. The adequacy of the shutdown margin is verified by demonstrating that the core is shutdown by a margin of 0.38 percent ak.vich the strongest operable control rod fully withdrawn, or that at least 0.387; Ak shutdown margin is available if the remaining control rods have had their directional control valves disarmed. Disarming the directional control valves does not inhibit control rod scram capability.

Specification 3.10.A.6 allows unloading of a significant portion of the reactor core. This operation is performed with the mode switch in the

" refuel" position to provide the refueling interlocks normally available during refueling operations. In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdre.wn.

, at a time. .The requirement that the fuel assemblies in the cell controlled '

l by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality. Prior to removal of the last two

. diagonal fuel assemblies, a double blade guide shall be inserted to properly support the control rod and fuel assemblies. After removal of the last two fuel assemblies and withdrawal of the control rod, the double blade guide may be removed.

Each control rod provides primary reactivity control for the fuel assemblies in the cell associated with that control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core. The requirements for SRM operability during these core alterations assure sufficient core monitoring.

To minimize the possibility of loading fuel into a cell containing no control rod, when refueling interlock input signals are bypassed for the

, spiral unload / reload technique, it is required that the control' room operator and a licensed operator and a member of the reactor engineering staff on the refueling floor verify that the control rod is inserted in the cell to be loaded. Prior to insertion of the control rod, it shall be verified that a double blade guide was placed in the cell to be loaded to properly support the control rod and fuel assemblies.

-208-

r s LIMITING CONDITIONS FOR OPERATION l SURVEILLANCE REOUIREMENTSl  !

, , ,a 3.12 Additional Safety Related Plant 4.12 Additional Safety Related ?lant Capabilities ^

Capabilities ,

, s  ;

Applicability: Applicability: ,  ; 7

,1

'I Applies to the operating status of the Applies to the surveillance require- i main control room ventilation system, nents for the main control room venti-\ ,

t.he reactor building closed cooling lation system.,tle'rcactor building 3' *

. water system and the service water ' closed cooling! water s 9 tem and the *i system, service water system'which are required by _the corresponding"'

Limiting Conditions for Operation.

Objective: Objective:

To assure the availability of the main To verify that opdrability or availa-control room. ventilation system, the bility under conditions for which these  !

reactor building closed cooling water capabilities are an essential response system and the service water system to station abnormalities.

~

upon the_condftions for which the il capability is an essential response r

to station abnormalities. tj A. Main Control Room Ventilation A. Main Control Room Ventilation

l. Except as specified in Specification 1. At least once per operating- cycle, the 3.12.A.3 below, the control room' air pressure drop across the combined HIPA treatment system, the diesel filters and charcoal absorber banks' i generators required for operation of shall be demonstrated to be less tihatt.

this system and the nain control room 6 inches of water at : ystem design flow air radiation monitor shall be oper- rate. t yp q able at all times when containment integrity is required. ,

y 2.a. The_results of the in-place cold DOP 2.a. The tests and sample: analysis of and halogenated hydrocarbon tests Specification 3.12.A.2 shall be performed at design flows on HEPA filters at least once per year for standby serviccq

~'

and charcoal absorber banks shall or after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />spf system show > 99% DOP removal and > 99% operation and following significant paint-halogenated hydrocarbon removal. ing, fire or chemical telease in any ventilation zone communicating with the system. ,,

b. The results of laboratory carbon b. Cold DOP testing shal! be performed sample analysis shall show 1 99% after each complete or pyrtial' replace-radioactive methyl iodide removal ment of the HEPA filter bank or after atavelocitywithjn20%ofsystem any structural maintenance on the system l design, 11.75 mg/m inlet iodide housing.  ;

concentration, > 95% R.H. and l <30'C.

4 1

c. Fans shall be shown to operate with- c. Halogenated hydrocarbon testing shall [

in + 10% design-flow, be performed after each complete er partial replacement of the charcoal absorber bank or after any scructural maintenance on the system housing,i i

, z<

i ,

\j l

-215-

TJ .

s? N v h ( $ R '

\

r 3. lT GASES 4~ _e .A. thin Control Rocm Ventilation System g,

4 U The control room ventilation system is designed to filter the control room 5> atmosphere for intake air and/or for recirculation during control room k isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage should be out leakage.

High efficiency particulate absoluto (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine cdsorbers. The charcoal adsorbers are installed to reduce the potential intake of radiplodine tc the' control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of D0P particulates.

.The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for

. Nuclear Pdwer Plants, Appendix A to 10 CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.

t. If the system is found to be inoperabic, there is no immediate threat to l

^

the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or refueling operations are terminated.

5. Reactor Building Closed Cooling Water System The reactor building closed cooling water system has two pumps and one heat exchanger in ecch of two loops. Eech loop is capable of supplying a the cooling requirements of the essential services following design I '

accident conditions with only one pump in either loop.

p , 3 The system has additional flexibility provided by the capability of inter-m*b connection of the two loops and the backup water supply to the critical y loop by the ecrvice water system. This flexibility and the need for only one pump in one loop to meet the design accident requirements justifies 1 the 30 day repair time during normal operation and the reduced requirements

, duringshead-off operations requiring thc availability of LPCI or the core spray systems.

C. Service Water System The service water system consists of four vertical service water pumps located in the intake structure, and associated strainers, piping, valving g and instrumentation. The pumps discharge to a common header from which independent piping supplies two Seismic Class I cooling water loops and one 4 turbine building loop. Automatic valving is provided to shutoff all supply T' to the turbine building loop on drop in header pressure thus assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR service water booster pumps, one control room basement fan coil unit and one RBCCW

+

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1 e LIMITXNG CONDITIONS FOR OPERATION bORVEILLANCEREOUIREMENTS .

3,14 FIRE DETECTION SYSTEM 4.14 FIRE DETECTION SYSTEM APPLICABILITY APPLICABILITY Applies to'the operational status of the Applies to the operational status of the Fire Detection _ System. Fire Detection System.

OBJECTIVE To assure-continuous automatic surveillance throuEh out the Main Plant.

SPECIFICATIONS SPECIFICATIONS The Fire Detection System instrumen- A. Each detector on Table 3.14 shall be lA. ration for each fire detection zone demonstrated operable every 6 months shown in Table 3.14 shall be operable. by performance of a channel functional test.

B. With one or more of the fire detection B. The NFPA Code 72.D Class B supervised instrument (s) shown in Table 3.14 circuits supervision associated with inoperable: the detector alarms of each of the above required fire detection

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire instruments shall be demonstrated watch patrol to inspect the OPERABLE at least once per 6 months.

zone (s) with the it. operable instru-ment (s) at lesst once per hour, and

2. Restore the inoperable instrument (s) to OPERABLE status within 14 days 4 or prepare and submit a Special Report to the Commission pursuant to Specification 6.7.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for re-storing the instrument (s) to OPERABLE status.

3.15 FIRE SUPPRESSION WATER SYSTEM 4.15 FIRE SUPPRESSION WATER SYSTFM APPLICABILITY APPLICABILITY Applies to the availability of water for Applies to the availability of water fire fighting purposes. for fire fighting purposes.

OBJECTIVE To assure a centinuous operable water supply for fire fighting systems fren 2 fire pumps.

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