ML20116M911

Proposed Tech Specs,Incorporating Changes Re Standby Gas Treatment & Control Room Ventilation Sys,Clarification of lo-lo Set Design Change & Refueling Interlocks & Environ Qualification Program

ML20116M911
Person / Time
Site:	Cooper
Issue date:	04/26/1985
From:	NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20116M908	List: ML20116M904 ML20116M911
References
NUDOCS 8505060385
Download: ML20116M911 (28)

Text

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7. (cont'd.)

4.7 (cont'd.)

B.

Standby Gas Treatment System B.

Standby Gas Treatment System 1.

Except as specified in 3.7.B.3 below, 1.

At least once por 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 leak tests water ct the system design flow at < design flow (1780 CFM) and at a rate.

reictor Luilding pressure <.25" Wg on HEPA filters and charcoal adsorber b.

Inlet heater input is capable of banks respectiv.ly shall show >99%

reducing R.H. from 100 to 70% R.H.

DOP removal and >99% halogenated hydrocarbon removal.

2.a. The tests and sample analysis of Specification 3.7.B.2 shall be

b. The results of laboratory carbon performed at least once per year sample analysis shall show >99%

for standby service or after every radioactive methyl iodide removal 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and with inlet conditiogs of:

velocity following significant painting,

>42 FPM, >1.75 mg/m inlet methyl fire or chemical release in any iodide concentration >70% R.H.

ventilation zone communicating and <30 C.

with the system.

c. Each fan shall be shown to provide

b. Cold DOP testing shall be performed 1780 CFM !10%.

after each complete or partial replacement of the HEPA filter 3.

From and after the date that one bank or af ter any structural standby gas treatment system is made maintenance en the system housing, or found to be inoperable for any reason, reactor operation or fuel

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

standby gas treatment system, and its associated diesel generator,

d. Each system shall be operated shhl1 be operable, 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 compliance with Specification 4.7.B.2.a and Specification 3.7.B.2.a.

3.

System drains where present shall 8505060385 850426 be inspected quarterly for adequate PDR ADOCK 05000298 water level in loop-seals.

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3.7.B & 3.7.C BASES (cont'd) 1 1

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 HEPA filters. The laboratory carbon sample test results should indicate a radio-active methyl iodide removal efficiency of at least 99 percent for expected accident conditions.

If the performance of the HEPA filters and charcoal l

adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed.

i 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 i

system performance and reactor operation or refueling operation may continue while repairs are being made.

If neither system is operable, 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 valves, 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.

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Performing these tests prior to refueling 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 treatment system performance. capability.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less l

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 3

cycle to show system performance capability.

The' frequency of tests and sample analysis are necessary to show that the HEPA i

filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant-shall be performed in accor-dance with ANSI N510-1980.

The test cannisters that are installed with the adsorber trays should be used for the charcoal adsorber efficiency test.

Each sample should be at least two inches in diameter and a length equal to the thickness of the bed. If test results are unacceptable, all adsorbent in the system shall be replaced W

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[.7.B&4.7.C BASES with an adsorbent qualified according to Table 1 of Regulatory Guide l.52, Revision 2, March, 1978. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA fil-l ters with DOP aerosol shall be performed in accordance to ANSI N510-1980.

Any filters found defective shall be replaced with filters qualified pursuant 1

to Regulatory Position C.3.d. of Regulatory Guide 1.52, Revision 2, March, 1978.

l All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber system.

With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.

.s If system drains are present in the filter /adsorber banks, loop-seals

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must be used with adequate water level to prevent by-pass leakage from.

the banks..

i If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals j

or foreign material, the same tests and sample analysis _shall be performed as required for operational use. The determination -of significance shall be 3

made by the operator on duty at the time of the incident.- Knowledgeable I

staff members should be consulted prior to making this determination.

l Demonstration of the automatic' initiation _ capability and operability of filter cooling is necessary to assure system performance capability.

If one standby gas treatment system is inoperable, the other system must be-tested daily.

This substantiates the availability of the operable system and thus reacter operation or refueling operation can continue for a limited period of time.

i 3.7.D & 4.7.D BASES Primary Containment Isolation Valves

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Double isolation valves are provided on lines penetrat1ng the primary con-tainment and open to the free space of the containment. Closure of one-of l

the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation-is required to minimiz,e the potential leakage. paths from the containment in the event of a loss of- -

coolant accident.

The maximum closure times for the automatic isolation valves of-the primary-containment and reactor vessel. isolation control system have been selected in consideration of the design intent to prevent core uncovering following 2

pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.

These valves are highly-reliable, have a low service requirement, and most are.normally closed. The initiating sensors.and associated trip channels i

are also checked to demonstrate the capability for automatic isolation.

The test interval of Lonce per operating cycle for automatic initiation 1

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.12 Additional Safety Related Plant 4.12 Additional Safety Related Plant Capabilities

' Capabilities Applicability:

Applicability:

Applies to the operating status of the Applies to the surveillance require-main control room ventilation system, ments for the main control room venti-the reactor building closed cooling lation system, the reactor building water system and the service water closed cooling water system and the 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 operability 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 conditions for which the capability is an essential response to station abnormalities.

A.

Main Control Room Ventilation A..

Main Control Room Ventilation 1.

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 HEPA treatment system, the diesel filters and charcoal absorber banks generators required for operation of shall be demonstrated to be less than this system and the main control room 6 inches of water at system design flow air radiation monitor shall be oper-

rate, able at all times when containment integrity is required.

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 per-at I design flow (341 CFM) and at formed at least once per year for control room pressure on HEPA fil-standby service or after every 720 ters and charcoal adsorber banks hours of system operation and fol-respectively shall show 3 99% DOP lowing significant painting, fire removal and 1 99% halogenated or chemical release in any ventila-hydrocarbon removal.

tion zone communicating with the system.

b.

The results of laboratory carbon b.

Cold D0P testing shall be performed sample analysis shall show 3 99%

after each complete or partial replace-radioactive methyl iodide removal ment of the HEPA filter bank or after withinletconditiogsof: velocity any-structural maintenance on the 322 FPM,11.75 mg/m inlet iodide system housing.

concentration, 1 95% R.H. and $30*C.

c.

Each fan shall be shown to' provide c.

Halogenated hydrocarbon testing shall 341 CFM 110%.

be performed after each complete or partial replacement of the charcoal absorber bank or after any structural maintenance on the system housing.

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1 12 BASES A.

Main Control Room Ventilation System The control room ventilation system is designed to filter the control room atmosphere for intake air and/or for recirculation during control room 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 absolute (HEPA) filters are installed before the.

charcoal adsorbers to prevent clogging of the iodine adsorbers.

The charcoal adsorbers are installed to reduce the potential intake of radioiodine to 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 HEPA filters. 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 performance 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 Power Plants, Appendix A to 10 CFR Part 50.

If the system is found to be inoperable, there'is no immediate threat to the control room and reactor operation or refueling ope' ration 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.

B.

Reactor Building Closed Cooling Water System The reactor building closed cooling water system has two pumps and one heat exchanger in each of two loops.

Each loop is capable of supplying the cooling requirements of the essential services following design accident conditions with only.one pump in either loop.

The system has additiona1' flexibility provided by the capability of inter-connection of the two loops and the backup water supply to the critical loop by the service water system.

This flexibility and the need for only one. pump in one loop to meet the design accident requirements justifies the 30 day repair time during normal operation and the reduced requirements during head-off operations requiring the 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 and instrumentation.

The pumps discharge to a common header from which independent piping supplies two Seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoff all supply 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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Attachmsnt 2 Revised Technical Specifications for

-Low-Low Set Design Change Revised Pages:

50 52b 83

Reference:

1) NPPD Letter, J. M. Pilant to D. B. Vassallo, " Safety Relief Valve (S/RV) Low-Low Set -(LLS) System and Lower Main Steam Isolation Valve (MSIV) Water Level Trip", Dated December 17, 1982 on March 4, 1983, Amendment 83 was issued to the Cooper Nuclear Station Facility Operating License to incorporate into the Technical Specifications the Low-Low Set System.

This system was implemented to mitigate excessive loads on Cooper's Mark I Containment brought about by subsequent safety relief valve actuations.

The setpoint for Main Steam Isolation Valve Water Level Trip was lowered by the change to provide additional safety margin and to reduce safety relief valve challenges as recommended in NUREG-0737, Item II.K.3.16.

A description and safety analysis of this change is contained in NEDE-22223, " Low-Low Set Logic and Lower-MSIV Trip for BWR's with Mark I Containment", an enclosure to Reference 1.

The modification was properly implemented and tested with all related drawings and procedures updated.

However, the initial Technical Specification change submitted in Reference 1 overlooked several points which should be corrected.

These points in no-way affect the operation, surveillance, or safety function of~the modification. Rather, they are editorial changes to correct errors and to achieve consistency of nomenclature.

Accordingly, Nebraska Public Power District proposes the following revisions be made to the Technical Specifications:

1.

Delete from page 50 the entry for reactor low-low water level instrument as it no longer initiates primary containment isolation.

2.

Change Group 7 isolation signal to reactor low-low-low water level

(>-145.5 in.) on page 52b and reflect this in the Bases for Section 3.2.A

_on page 83.

3.

Delete reference to HPCI and RCIC initiation under the " Primary Containment Isolation Functions" bases section on page 83 because'it is unrelated to primary system isolation.

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 pro ~ posed license amendment involve a significant increase in the probability or consequences. of an accident previously evaluated?

Evaluation:

Because the. proposed change is editorial in nature and does not change existing equipment, surveillances, or procedures, it does not affect the probability or consequence of an accident previously evaluated.

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2.

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

Evaluation:

Because the proposed change does not introduce any new mode of operation, the possibility of an accident of a different type than

-analyzed.in the Final Safety Analysis Report.would not result from the change; therefore, the proposed license-amendment 'does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

)ves the' proposed amendment involve a significant ~ reduction in a

.argin of safety?

Evaluation:

Because the proposed change does not change existing facility equipment, surveillance, or procedure and is ~. intended to correct errors and achieve. consistency in nomenclature, it does not involve

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a significant reduction in a margin of safety.

B.

Additional basis for proposed no significant hazards censideration determination:

The Commission has provided guidance concerning the applicaticn of the standards for making a no significant hazards' consideration determinction by providing certain. examples (48FR14870).

The examples include "(1) A

_ purely administrative change to Technical Specifications:

for example, a change to achieve consistency throughout the. Technical Specifications, correction of an error, or a change-in nomenclature."

It is. the District's belief the proposed change is encompassed by the above example.

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COOPER NUCLEAR STATION TABLE 3.2.A (Page 1)

PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION Minimum Number

. Action Required of Operable When Component Instrument Components Operability is Instrument I.D. No.

Setting Limit Per Trip System (1) Not Assured (2)

Main Steam Line High RMP-RM-251, A,B,C,6D f 3 Times Full Power 2

A or B Rad.

Reactor; Low Water Level NBI-LIS-101, A,B,C,&D 3+12.5" Indicated Level.

2(4)

. A or B l

Reactor Low Low Low Water NBI-LIS-57 A & B #1 3-145.5" Irdicated Level 2

A or B L2 vel NBI-LIS-58'A & B #1 Main Steam Line Leak MS-TS-121, A,B,C,&D f 200*F 2(6)

B

'D2tection

'122, 123, 124, 143,.144, 145, 146, 147, 148,149, 150 di Main Steam Line'High MS-dPIS-Il6 A,B,C,&D f 140% of Rated Steam 2(3)

.B-

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Flow 117, 118, 119 Flow Main Steam Line Low HS-PS-134, A,B,C,6D 3 825 psig 2(5)

B Pressure High Drywell Pressure PC-PS-12, A,B,C,&D

$ 2 psig 2(4)

A or B liigh Reactor Pressure RR-PS-128 A & B f 75 psig I

D

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Main Condenser Low HS-PS-103, A,B,C,&D 1 7" Hg (7) 2 A or B Vacuum

'Rsactor. Water Cleanup RWCU-dPIS-170,A & B f 200% of System Flow 1

C System High Flow

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

-Isolations 1.

Secondary Containment Isolation '

2.

Start' Standby Gas Treatment System Group'7-Isolation Signals:

1.

Reactor Low Low Low Water Level'(>-145.5 in) 2.

Main Steam Line High Radiation (p times full power background)

Isolations:

1.

Reactor Water Sample Valves i

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i 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-

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l sequences. This set of specifications provides the limiting conditions of I

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 protective instrumentation when required even during periods when portions i

of such systems'are out of service for maintenance, and (2)~to prescribe i.

.the trip settings required.to. assure adequate performance. When necessary, 4

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.fwhere 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 actuation of the safety system involved and exposure to abnormal situations.

A.

Primary Containment Isolatio'n Functions q.

Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2.A which senses.the conditions for which isola-j tion is required.

Such instrumentation must-be available whenever primary j

containment integrity is required.

l 2

The instrumentation which initiates' primary system isolation is connected j,

in a dual bus arrangement.

l The low water level instrumentation, set to trip at 176.5" (+12.5") above the top of the active fuel, closes all isolation valves except those in Groups ll, 4, 5,

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and 7._ Details of valve grouping and required closing times are given 1Ln Specification 3.7._ For valves which isolate at this level this trip setting i-is adequate to prevent core uncovery in the case of.a1 break in the largest L

line assuming a 60 second' valve closing time.

Required. closing times are less than this..

The. low low low reactor. water level instrumentation is set-to trip when the

. ater level-is 19" (-145.5") above the top of'the active fuel.- This trip closes w

4 i

Groups-I and 7-Isolation Valves-(Reference 1), activates the remainder'of the

.l CSCS subsystems, and starts the. emergency diesel generators..These trip level; settings were chosen.co 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..

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Attachm:nt 3 Revised Technical Specifications for Refueling Interlocks Clarification Revised Page:

205 The current Technical Specifications for Cooper Nuclear Station require all refueling interlocks, except the one-rod-out interlock, to be operable during multiple control rod removal-regardless of whether fuel is present in the core or not.

Nebraska Public Power District requests a revision to the Technical ~

Specifications to delete the above' requirement when fuel is not present in the reactor vessel.

The proposed change conforms to NUREG-0123, Revision 3, Standard Technical Specifications 3.9.10.2 which states the requirement is applicable when there is "f al in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed."

Evaluation of this Revision with Respect to 10CFR50.92 The enclosed Technical 1pecification 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 prop $ sed change conforms with the GE Standard Technical-Specifications and clarifies the need for refueling interlock operability during the time no fuel is present in the reactor vessel it does not involve a significant increase in the probabilitygor consequences of an accident previously evaluated.

2.

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

Evaluation:

Because the change does not affect the requirements for refueling interlock operability while' handling fuel-it does not create the possibility for 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:

Because this change clarifies the operability requirements of the refueling interlocks during the' time fuel is not present in the core and is in agreement with the GE Standard Technical Specifications it does not involve a significant reduction in a margin of safety.

1 W

LIMITING CONDITIONS FOR OPERATIOM SURVEILLANCE REOUIREMENTS 3.10.A (Cont'd) 4.10 (Cont'd) l 6.

Any number of control rods may be withdrawn or removed from the reactor core providing the following conditions are satisfied:

a.

The. reactor mode switch is locked in the " refuel" position. The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn con-trol rod af er the fuel assemblies in t

the cell containing (controlled by) that control rod have been removed from the reactor core.

When fuel is present in the reactor vessel, all other refueling interlocks shall be operable.

B.

Core Monitoring B.

Core Monitoring During core alterations two SRM's Prior to making any alterations to shall be operable, one in the core the core, the SRM's shall be quadrant where fuel or control rods functionally tested and checked for are being moved and one in an ad-neutron response. Thereafter, while jacent quadrant. For an-SRM to be required to be operable, the SRM's considered operable, the following will be checked daily for response conditions shall be satisfied:

(or every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until-3 cps is attained if the spiral reload 1.

The SRM shall be inserted to the normal technique is being used).

operating level.

(Use of special move-able, dunking type detectors during initial fuel. loading and major core alterations in picce of normal detec-tors is permissible as long as the detector is connected to the normal SRM circuit.)

2.

OperableRSRM's shall have a minimum of 3 cps except as specified in 3 and 4 below.

3.

Prior to spiral unloading, the SRM's shall have an initial count rate of 3 cps. During spiral unloading, the count rate on the SRM's may drop below 3 cps.-

-205-

Revised Techn1 cal Specifications for Environmental Qualification Program i

Revised Pages:

226 226a (deleted) 227 i

The. current Technical Specifications for Cooper Nuclear Station specify a deadline of June 30, 1982, for environmental qualification of all 1

safety-related electrical equipment.

On November 19, 1984, the U.S. Nuclear Regulatory Commission issued its final rule (49FR45571) eliminating the June 30, 1982, deadline for environmental qualification.

Nebraska Public Power District requests a revision to the Technical Specifications to delete the above deadline from the Administrative. Controls Section..The deletion of the deadline will have no effect on establishing an Environmental Qualification Program at Cooper Nuclear Station since. the District has already committed to implement the program in accordance with NRC guidelines.

Evaluation of this Revision with Respect to 10CFR50.92 The proposed amendment involves no significant hazards consideration since it will not 1) involve a significant increase in the possibility or consequences l

of an accident previously evaluated,' 2) create the possibility of~a new or different kind of accident from any accident previously evaluated, or

3) involve a significant reduction in a margin of safety.

The Commission has provided guidance concerning the application of the standards for making a no significant hazards consideration determination by providing certain examples (48FR14870). -The examples include "(1) A purely administrative change to Technical Specifications:

for example, a change to achieve consistency throughout the Technical Specifications, a correction of an error, or a change in nomenclature." Another example given is "(vii) A change-to make a license conform to changes in regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations."

It is the District's belief'the proposed change is encompassed.by the above examples.

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6 3. (cont'd)

.A._

High Radiation Areas 1

In-lieu of the " control device" or " alarm signal" required by Paragraph 20.203 (c) (2) of 10 CFR 20 each High Radiation Area (100 mrem /hr or greater)'shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by requiring notification and permission of the shift supervisor..Any. individual or group of indi-

. iduals permitted to enter.such' areas shall be provided with a radiation v

monitoring device which-continuously indicates the radiation' dose rate in the area.

6.3.5 Temporary Changes Temporary changes to procedures which do not change the intent of the original procedure may be made, provided such changes are approved by two members of the operating staff holding'SR0~ licenses.' Such changes-shall be documented and subsequently reviewed by the Division Manager of Nuclear Operations within one month.

6.3.6-Exercise of Procedures 1

Drills of the Emergency Plan procedures shall be conducted annually, including a check of communications with offsite support groups.

Drills

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on the procedures specified in 6.3.2.A, B, and C above shall be con-ducted as part of the retraining program.

i 6.3.7 Programs The following programs shall be established:

-A.

Systems Integrity Monitoring Program 1

A program shall be established to reduce leakage to as low as practical levels from systems-outside the' primary containment during a serious accident that would or could_contain highlyLradioactive fluids.. This program shall include provisions establishing preventive maintenance and.

periodic visual inspection requirements, and leak testing _ requirements for each system at a frequency not to exceed refueling cycle intervals.

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~ B.

Iodine Monitoring' Program

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A program shall be established to-ensure'the capability to accurately determine the airborne iodine concentration in vital areas under accident' l

conditions. 'This program shall include training of personnel, procedures i

for monitoring-and provisions for maintenance of sampling and analysis equipment.

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C.

Environmental Qualification Program A.

By no later than December 1, 1980, complete and auditible records-i must be available and maintained-at a central location which l

describe the environmentalLqualification method used for all--

safety-related_ electrical equipment in sufficient detail to j

document the degree of compliance with.the DOR Guidelines or NUREG-0588.- Thereafter, such records should be updated and maintained current as equipment is replaced,-further. tested, or_otherwise further qualified.

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6.3 (cont'd)

B.

Post-Accident Sampling System (PASS)

A. program shall be established to ensure the capability to obtain and

' analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

This program shall include training of personnel, procedures for' sampling'and analysis and provisions for operability of sampling and analysis equipment.

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Attachm2nt 5 Revised Technical Specifications for Standby Liquid Control System RCIC Test and Calibration Frequencies Table Typographical Errors Revised Pages:

111 75

- To correct typographical errors discovered on two pages of the CNS Technical Specifications, Nebraska Public ' Power District proposes the following revisions be made to sames 1.

Change Subsection III.8.5 to read Subsection III.9.5 in paragraph 2 of 4.4 Bases, Standby Liquid Control System, page 111.

2.

Change Item 1.

Logic Buss Power Monitor, under item category Logic Systems (4)(6) on Table 4.2.B,. RCIC Test & Calibration Frequencies (page 6), page 75, to read 1. Logic Bus Power Monitor.

Evaluation of this Revision with Respect to 10CFR50.92 The ' proposed amendment incorporates changes that are of an administrative nature to correct errors and involves no significant hazards considerations since it will not 1) involve a significant increase in the possibility or i

consequences of an accident previously evaluated, 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or

3) involve a significant reduction in a margin of safety.

The Commission has provided guidance concerning the application of the standards.for making a no significant hazards consideration determination by providing certain examples (48FR14870).

The examples include "(i) A purely administrative change to Technical Specifications:

for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature."

It is the District's belief that the proposed change is,

encompassed by the above example.

i l

4 t

l

~$ 4 BASES'(cont'd.)

'The volume versus concentration' requirement of the solution is such that, should evaporation occur from any point within the curve, a low level alarm will annunciate before the temperature versus concentration requirements are exceeded.

l The quantity of stored boron includes an additional margin (25 percent) r beyond the amount needed to shutdown the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water.

A minimum quantity of 2650 gallons of solution having a 16 percent sodium pentaborate concentration, or the equivalent as shown in Figure 3.4.1, is required to meet-this shutdown requirement.

For the minimum required pumping rate of 38.2 gpm, the maximum net storage volume of the boron solution is established as 4780 gallons.

4.4 BASES STANDBY LIQUID CONTROL SYSTEM-Experience with pump operability indicates that the monthly test, in combination with the tests during each operating cycle, is sufficient to maintain pump

. performance. The only practical time to fully test-the liquid control system is during a refueling outage.

Various components of the system are individually tested. periodically, thus making unnecessary more frequent testing of the entire system.

The bases for the surveillance requirements.are given in subsection III.9.6 of the Final Safety Analysis Report, and the details of the various tests a're discussed in subsection III.9.5.

The solution temperature and volume are l

4 checked at a frequency to assure a high reliability of operation of the system should it ever be required.

1 J

-111-

COOPER NUCLEAR STATION TABLE 4.2.B (Page 6)

RCIC TEST &, CALIBRATION FREQUENCIES Instrument Item Item I.D. No.

Functional Test Freq.

Calibration Freq.

Check' Instrument Channels 1.

Reactor High Water Level NBI-LIS-101 A & C,.#2 Cnce/ Month (1)

Once/3 Months Once/ Day 2.

Reactor Low Water Level 10A - K79 A & B 10A-

'Once/ Month (1)

Once/3 Months Once/ Day K80 A & B 3.

RCIC High Turbine Exhaust RCIC-PS-72, A & B Once/ Month (1)

Once/3 Months None

-Press.

4.

RCIC Low Pump Suction. Press.

RCIC-PS-67-1 Once/ Month (1)

Once/3 Months None 5.

RCIC Steam Line Space Excess RCIC-TS-79, A,B,C, & D Once/ Month (1)

Once/Oper. Cycle None Temp.

RCIC-TS-80,'A,B,C, & D Once/ Month (1)

Once/Oper, Cycle None RCIC-TS-81, A,5,C, & D Once/ Month (1)

Once/Oper. Cycle None RCIC-TS-82, A,B,C, & D Once/ Month (1)

Once/Oper. Cycle None 6.

RCIC Steam Line High AP RCIC-dPIS-83 Once/ Month (1)

Once/3 Months None s

RCIC-dPIS-84' Once/ Month (1)

Once/3 Months None 3

7.

RCIC Steam Supply Press. Low RCIC-PS-87, A,B,C,.& D Once/ Month (1)

Once/3 Months None 8

8.

RCIC Low Pump Disch Flow RCIC-FIS-57 once/ Month (1)

Once/3 Months None 9.

Pump Disch. Line Low Pressure.

CM-PS-269 Once/3 Months once/3 Months None 10.

RCIC Turbine. Conditional RCIC-TDR - K9 Once/ Month (1)

Once/Oper. Cycle None Supv. Alarm Timer

11. -RCIC Steam Line High AP RCIC-TDR-K-12 Once/ Month once/Oper. Cycle None Actuation Timer RCIC-TDR-K-32 Once/ Month Once/Oper Cycle None Logic Systems (4)(6) l.

Logic Bus Power Monitor once/6 Months N.A.

l 2.

RCIC Initiation Once/6 Months N.A.

3.

Turbine Trip once/6 Months N.A.

4.

RCIC Automatic Isolation Once/6 Months N.A.

!i'

Attachmsnt 6 Revised Technical Specifications for Section 6 Administrative Controls Editorial Changes Revised Pages:

iv (Table of Contents) 221 222 223 224 225 225a (deleted) 231 235 Section 6, Administrative Controls, of the current CNS Technical Specifications has undergone various changes during the past several years which have introduced discontinuities between the pages of some subsections (i.e.; gaps).

To correct this, Nebraska Public Power District proposes a change in Technical Specifications (purely editorial in nature) which simply condenses related subsections which are spread out over several pages.

The

. content of this material remains completely _ unchanged - there are no l

deletions, wording modifications, syntax or sequence changes, etc.

Page 225a was deleted as a byproduct of compressing the contents of the various affected pages onto fewer pages.

This proposed editorial change is being submitted to provide for improved readability and understanding of the Technical Specifications.'

Evaluation of this Revision with Respect to 10CFR50.92 A.

The enclosed Technical. Specification change is judged to involve no significant hazards based upon the following:

1.

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

Evaluation:

No.

The proposed amendment does - not impact the probability or consequences of any accident previously-evaluated.-

2.

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

Evaluation:

No.

The proposed amendment does not impact upon any new or old accident analyses since it is purely editorial in nature and does not-change the content of the Technical. Specifications whatsoever.

7

3.

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

Evaluation:

No.

The proposed amendment is intended to clarify the Technical Specifications by introducing improved continuity and should improve safety, if anything.

B.

Additional basis for the proposed no significant hazards consideration determination:

The Commission has provided guidance concerning the application of the standards for making a no significant hazards consideration determination by providing certain examples (48CFR14870).

The examples include:

"(1) A purely administrative change to achieve consistency throughout the Technical' Specifications, correction of ~ an error, or a change in nomenclature... "

It is the District's belief the proposed change-is encompassed by the above example.

TABLE OF CONTENTS (Cont'd.)

Page No.

SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 1

6.2.1.B NPPD Safety Review and Audit Board (SRAB) 222 B.1 Membership 222 B.2 Meeting Frequency 222 B.3 Quorum 222 B.4 Review 222 B.5 Authority

-223 B.6 Records 223 B.7 Procedures 223 B.8 -Audits 223 6.3 Procedures and Programs 225 6.3.1 Introduction 225 6.3.2 Procedures 225 6.3.3 Maintenance and Test Procedures 225 6.3.4 Radiation Control Procedures 225

.A High Radiation Areas 226 6.3.5 Temporary Changes 226 6.3.6 Exercise of Procedures 226 6.3.7 Programs 226

.A Systems Integrity Monitoring Program 226

.B Iodine Monitoring Program 226

.C Environmental Qualification Program 226

.D Post-Accident Sampling System (PASS) 227 6.4 Record Retention 228 6'4.1 5 year retention 228 6.4.2 Life retention 228

-6.4.3 2 year retention 229 6.5 Station Reporting Requirements 230 6.5.1 Routine Reports 230

.A Introduction 230

.B. Startup Report 230

.C Annual Reports 230

.D Monthly Operating Report 231 6.5.2 Reportable Occurrences 231

.A Prompt Notification with Written Followup 232

.B Thirty-Day Written Reports 234 6.5.3 Unique Reporting-Requirements 235 l

-iv-

+

s

,. _

• k. 2 (cont'd) f.

Investigate all violations of Technical Specifications,; including reporting evaluation and recommendations to prevent recurrence, i

to:the Assistant General Manager - Nuclear and to the Chairman of the NPPD Safety Review and Audit Board.

g.

Perform special reviews and investigations and render reports thereon as requested by the Chairman of the Safety Review and 4

Audit-Board.

h.

Review all reportable events specified in Section 50.73 to 10CFR Part 50.

i.-

Review drills on emergency procedures-(including plant evacuation) and adequacy of communication with off site groups.

j.

Periodically review procedures required by Specifications 6.3.1, 6.3.2, 6.3.3, and 6.3.4 as set forth in administrative procedures.

~

5.

Authority

}

a.

The Station Operations Review ~ Committee shall be advisory.

b.

The Station Operations Review Committee shall recommend to the Division Manager.of Nuclear Operations approval or disapproval i

of proposals under items 4, a through e and j above.

In case of disagreement between the recommendations of the Station l

Operations Review Committee and the Division Manager of Nuclear i

Operations, the' course determined by the Division Manager of l

Nuclear Operations to be the more conservative will be followed.

)

A written summary of the disagreement will be sent to the

.Assistant General Manager - Nuclear and to the NPPD Safety i

Review and Audit Board.

c.

The Station'0perations Review Committee shall report to the Chairman of the NPPD Safety Review and Audit Board on all re-views and investigations conducted under items 4.f, 4.g, 4.h,

'and 4.1.

d.

.The Station Operations Review Committee shall make determinations regarding whether or not proposals considered by the Committee involve unreviewed safety. questions.

This determination shall be subject to review by the.NPPD Safety Review and Audit Board.

6.

Records:

j Minutes shall be kept for all meetings of the Station Operations Review Committee and shall include identification of all documen-tary material reviewed; copies of the minutes shall be forwarded to the Chairman of the NFPD Safety Review and Audit Board and the Assistant General Manager - Nuclear within one month.

7.

Procedures:

~

Written administrative procedures for Committee operation shall l

be prepared and maintained describing the method for submission and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, dissemination of minutes,~and:such"other matters as may be appropriate..

i i-

-221-4-

.. ~.

,-,__m..-~..

-. - '.'.i-

.E..

, '* -6.2 :(Cont'd)

B.

NPPD Safety Review and Audit Board (SRAB)

Function:

The Board shall function to provide independent review and audit of designated activities.

1..

Membership:

.a.

Chairman

+

b.

Vice-Chairman c.

Five Members d.

Consultants (as required) i-The Board members shall collectively have the capability required to review problems in the following areas:

nuclear power plant operations, nuclear engineering.. chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering,' quality assurance l

practices, and other appropriate fields associated with the unique characteristics of.the nuclear power plant involved.

When the. nature of a particular problem dictates, special consultants will be utilized.

i Alternate members shall be appointed in writing by the Board Chairman to serve on a temporary basis; however, no more than two alternates shall serve on the Board at any one time, i

i 2.

Meeting frequency:

Semiannually, and as required on call of the

. Chairman.

3.

Quorum: Chairman or Vice Chairman, plus four members including i

alternates.

No more than a minority of the quorum shall be from groups holding line responsibility for the operation of the plant.

4.-

Review: The following subjects shall be reported to and reviewed by the NPPD Safety Review and Audit Board.

i The safety evaluations for 1) changes to procedures, equipment a.

or. systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such acticns did not constitute an unreviewed safety question.

b.

. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR. 4

-222-g

"6. 2 (cont'd) c.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d.

Proposed changes to Appendix A Technical Specifications or the CNS Operating License.

e.

Violations of applicable codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance, f.

Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

g.

All reportable events specified in Section 50.73 to 10CFR Part 50.

h.

Any indication of an unanticipated deficiency in some aspect of design er operation of safety related structures, systems, or components,

i. Minutes of meetings of the Station Operations Review Committee..

j. Disagreement between.the recommendations of the Station-Operations Review Committee and the Division Manager of Nuclear Operations.

k.

Review of events covered under e,f,g, and h above include reporting to appropriate members of management on the results of investiga-tions and recommendations to prevent or reduce the probability of recurrence.

5.

Authority: The NPPD Safety Review and Audit Board shall report to and be advisory to the Assistant General Manager - Nuclear on those areas of responsibility specified in Specifications 6.2.1.B.4 and 6.2.1.B.7.

4 6.

Records:

Minutes shall be recorded for all meetings of the NPPD Safety Review and Audit Board and shall identify all documentary material reviewed. Copies of the minutes shall be forwarded to the Assistant General Manager - Nuclear and the Division Manager of Nuclear Operations, and such others as the Chairman may designate within one month of the meeting.

7.

Audits:

Audits of selected aspects of plant operation shall be performed under the cognizance of SRAB with a frequency commensurate with their safety significance. Audits performed by the Quality Assurance Department which meet this specification shall be considered to meet the SRAB audit requirements if the audit results are reviewed by SRAB. A representative portion of procedurcs and records of the activities performed during the audit period shall be audited and, in addition, observations of performance of operating and maintenance activities shall be included.

These audits shall encompass:

-223-l

l 6.2 (cont'd) i a.

Verification of. compliance with internal rules, procedures (for example:. normal, off-normal, emergency, operating, maintenance, surveillance, test, and radiation control procedures) and applicable license conditions at least once i

per 24 months.

b.

The training, qualification, and performance of the operating _

staff at least once per 24 months.

c.

The Emergency Plan and implementing procedures at least once per 12 months.

d.

The Security Plan and implementing procedures at least once per 12 months.

e.

The facility fire protection and its implementing procedures at least once per 24 months.

f.

A fire protection and loss prevention inspection will be performed utilizing either qualified off-site licensee personnel or an out-side fire protection consultant at least once per 12 months.

g.

An inspection and audit by an outside qualified fire-protection consultant shall be performed at least once per 36 months.

F

-224-I

9 "6. 3 PROCEDURES AND PROGRAMS 6.3.1

-Introduction Station personnel shall be provided detailed written procedures to be used for operation and maintenance of system components and systems

(

that could have an effect on nuclear safety.

6.3.2 Procedures Written procedures.and instructions including applicable cNeck off lists shall be provided and adhered to for the following:

^

r A.

Normal startup, operation, shutdown and fuel handling operations of the station including all systems and components involving nuclear safety.

Actionstebetakentocorrectspecificandforeseenhotential B.

or actual malfunctions of. safety related syst' ems or components including responses to alarms, primary system leaks and abnormal reactivity changes.

Emergencyconditionsinvolvingpossibleorabtualreleasesofradio-C.

active materials.

D.

Implementing procedures of the Security Plan an'd the Emergency Plan.

E.

Implemencing procedures for the fire protection program.

F.

Administrative procedures for shift overtime.

6.3.3 Maintenance and Test Procedures The following maintenance and test procedures will be provided to satisfy routine inspection, preventive maintenance programs, and operating license requirements.

A.

Routine testing of Engineered Safeguards and equipment as required by the facility License and the Techniedl Specifications.

B.

Routine testing of standby and redundant equipment.

s C.

Preventive or correctide' maintenance of plant equipment and systems that could have an.effect on nuclear safety.

D.

Calibration and preventive maintenance of-instrumentation that could affect the nuclear safety of the plant.

E.

-Special testing of equipment for proposed changes.co operational.

procedures or proposed system design changes.

6.3.4 Radiation Control Procedures Radiation control procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure, and shall be consistent with the requirements of 10 CFR 20.

-225-g

)

7-

.s' 1.

A. tabulation on an annual basis of tha number of station,

~',

utility and other personnel. (including contractors) re-ceiving exposures greater than 100 mrem /yr and their

~._

associated man rem exposure according to work and job functions, 1/

e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special main-tenance (describe maintenance), waste processing, and refueling.

The dose. assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20%

of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

2.

A summary description of facility changes, tests or experi-ments in accordance with the requirements of 10CFR50.59(b).

3.

Pursuant to 3.8. A, a report of radioactive source leak testing. This report is required only if the tests reveal the presence of 0.005 microcuries or more of removable contamination.

4.

Documentation of all challenges to relief valves or safety valves.

D.

Monthly Operating Report Routine reports of operating statistics, shutdown experience, and a narrative summary of operating experience relating to safe operation of the facility, shall be submitted on a monthly basis to the individual designated in the current revision of. Reg.

' Guide 10.1 no later than the tenth of each month following the calendar month covered by the report.

6.5.2 Eeportable Events

A Reportable Event shall be any of'those conditions specified.in Section 50.73 to 10CFR Part 50.

The NRC shall be notified and a report submitted pursuant to the requirements of Section 50.73.

Each Reportable Event shall be reviewed by SORC and the results of this review shall be submitted to SRAB and the Assistant General Manager - Nuclear.

6.5.3 Unique Reporting Requirements Reports shall be submitted to the Director, Nuclear Reactor Regulation, USNRC, Washington, D.C. 20555: as follows:

A.

Reports on the following areas.shall be submitted as noted:

None.

1/

This tabulation supplements the requirements of 520.407 of 10CFR Part 20.

-231-

t'

.W O

C6

/

4

" INTENTIONALLY LEFT BLANK e

/

-232, 233, 234, 235-l

'\\

...... -