Text

Proposed Tech Specs Changing Axial Shape Index to Implement Power Dependent Limits & Providing Administrative Changes to Shutdown Margin Requirements to Remove Mode 2 Applicability of Tech Spec 3.1.1.2
	ML20237H256
Person / Time
Site:	Waterford
Issue date:	08/28/1987
From:	; LOUISIANA POWER & LIGHT CO.
To:	;
Shared Package
ML20237H232	List: ML20237H256; W3P87-2049, Application for Amend to License NPF-38,changing Axial Shape Index to Implement Power Dependent Limits & Providing Administrative Changes to Shutdown Margin Requirements to Remove Mode 2 Applicability of Tech Spec 3.1.1.2.Fee Paid;
References
	NUDOCS 8709030185
	Download: ML20237H256 (49)
	v • d • e

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POWER DISTRIBUTION LIMITS 3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION

3.2.7 limits

The AXIAL SHAPE INDEX (ASI) shall be maintained within the following

a. COLS$ OPERA 8LE

-0.23 $ ASI $ + 0.28 b.

COLS$ OUT OF SERVICE (CPC)

-0.17 5 ASI $ + 0.22 APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.*

ACTION:

With the AXIAL SHAPE INDEX outside its above limits, restore the AXIAL SHAPE INDEX to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to le 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.7 The AXIAL SHAPE INDEX shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using the COLSS or any OPERABLE Core Protection Calculator channel.

See Special Test Exception 3.10.2.

!D D B2 P PDR WATERFORD - UN!T 3 3/4 2-12 AMENDHENT NO. 13 1

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POWER DISTRIBUTION LIMITS _

3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION

3.2.7 limits

The AXIAL SHAPE INDEX (ASI) shall be maintained within the following

-0. *

+f0 b ..

COLSS0TOFJERVICE

-0.17 $ ASI $ + 0.

PC)[

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.*

ACTION:

i With the AXIAL SHAPE INDEX outside its above limits, restore the AXIAL SHAPE INDEX to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.7 Theper lea:,t once channel.

AXIAL SHAPE INDEX shall be determined to be within its limit at 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using the COLSS or any OPERABLE Core Protection Cr,lculator

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a. C5LSS of>&CRt3LE

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b, 6CLSS our OF SERWcE (CPb

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-0.12$AST $ +0.p2 De 7HE&lAl f0NECS c 70 % el g/frfb 7NEAWAl l'OMER (

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See Special Test Exception 3.10.2.

WATERFORD - UNIT 3 3/4 2-12 AMENDMENT NO. 13 !

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DESCRIPTION AND SAFETY ANALYSIS-0F PROPOSED CHANGE NPF-38-64

'This is a request to~ revise Technical Specification 3/4.5.2, ECCS Subsystems -

Tayg Greater Than 350 F and Technical Specification 3/4.5.3, ECCS Subsystems -

Tavg Less Than 350 F.

Existing Specification

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See Attachment A.

Proposed Specification See Attachment B.

Description The proposed change would revise Technical Specification 3.5.2, ECCS_ Subsystems - i Tavg Greater Than 350 F and Technical Specification 3.5.3, ECCS Subsystems - Tavg ,

less Than 350 F by adding a note to the Applicability section of.both Tech Specs !

to indicate that two ECCS subsystems are required to be operable when the RCS average temperature is equal to or greater than 500 F.

Technical Specification 3.5.2 currently requires two independent emergency core-cooling system (ECCS) subsystems to be operable when the reactor is in Modes 1, 2 and 3; however, the requirements of this Tech Spec in Mode.3 are applicable only if the pressurizer pressure is equal to or greater than 1750 psia. - The proposed change will add a note to the Mode 3 applicability statement that will require both ECCS subsystems to be operable any time the RCS average temperature .

is equal to or greater tnan 500 F, regardless of the pressurizer pressure.

Technical Specification 3.5.3 currently requires one ECCS subsystem to be operable if the reactor is in Modes 3 or 4 with a Mode 3 requirement that the pressurizer pressure is less than 1750 psia. The proposed change to this Tech Spec is similar to the proposed change to' Tech Spec 3.5.2 in that a note will be added to the Mode 3 applicability statement that requires the RCS average temperature to be less than 500 F before it is acceptable to have only one ECCS subsystem in-service.

The reason for the proposed change to these Tech Specs is to ensure that at least one train of high pressure safety injection (HPSI) is available (even if a single failure is assumed) to mitigate the consequences of a postulated steam line break (SLB) accident initiated from an RCS average temperature of 500 F or greater. The Cycle 2 safety analysis has shown that borated water from HPSI is required to prevent the core from becoming critical during the uncontrolled RCS i cooldown (associated with a SLB) from greater than 500 F.

NS41294

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation.of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed change will require that two ECCS subsystems are operable whenever the average temperature of the RCS is equal to or greater than 530 F. This will ensure that, even if one ECCS subsystem is assumed to f ail, one train of HPSI will be available to inject borated water into '

the RCS during an SLB. As described in the safety analysis for Cycle 2, borated water (from HPSI) is required to mitigate the reactivity transient associated with the RCS cooldown and prevent the core from returning to a critical condition. Below 500 F the RCS cooldown (and associated re-activity transient) during the SLB is less severe and HPSI flow is not ;

Therefore, since the proposed required to maintain the core subcritical.

change reduces the consequences of a SLB it will not involve a significant increase in the probability or consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve any physi;al changes to plant systems, structures or components nor will there be any significant changes to plant operating procedures. The proposed change will simply 1 clarify the RCS conditions which must exist prior to taking one of the ECCS subsystems out of service. Thus, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?

Response: No.

The intent of this Specification is to ensure there will be sufficient emergency core cooling capability available in the event of a LOCA and a coincident single failure that results in the complete loss of one ECCS subsystem. The proposed change will not affect the LOCA analysis since NS41294

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it merely adds a restriction that requires both ECCS subsystems to be ,

operable whenever the RCS temperature is equal to or greater than 500 F. j This additional restriction ensures that sufficient borated water can be j added to the RCS to mitigate tne reactivity transient associated with the !

uncontrolled RCS cooldown that occurs during a steam line break. Since i the proposed change adds a restriction that was not already a part of the l Tech Specs and since this restriction ensures that the consequences of a {

broader range of steam line breaks can be initigated, the proposed change 1 will result in an increase in the margin of safety.

The Commission has provided guidance concerning the application of standards for l determining whether significant hazards consideration exists by providing certain I examples (48 FR 14870) of amendments that are considered not likely to involve j significant hazards consideration. Example (ii) relates to a change that con-stitutes an additional limitation, restriction, or control not presently included in the Technical Specifications, (e.g., a more stringent surveillance require-ment).

In this case, the proposed change is similar to Example (ii) in that it consti-tutes an additional restriction (i.e. , RCS temperature) that must be satisfied before it is acceptable to have only one ECCS subsystem in service.

Safety and Significant Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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i NPF-38-64 ATTACHMENT A NS41294

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE high pressure safety injection pump,

b. One OPERABLE low pressure safety injection pump, and

c. An independent OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal !

and automatically transferring suction to the safety injection system' sump on a recirculation actuation signal.

APPLICABILITY: MODES 1, 2, and 3*. I ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted .

the Commission pursuant to Specification 6.9.2 within 90 days des-cribing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

With pressurizer pressure greater than or equal to 1750 psia.

WATERFORD - UNIT 3 3/4 5-3

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EMERGENCY COAL COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN 350 F LIMITING CONDITION FOR OPERATION 3.5.3- As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE high pressure safety injection pump, and I

b. An OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal and auto-matically transferring suction to the safety injection system sump on a recirculation actuation signal. , 1 APPLICABILITY: MODES 3* and 4.

ACTION:

a. With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

l SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

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With pressurizer pressure less than 1750 psia.

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BASES l ECCS SUBSYSTEMS (Continued)

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. (

I The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment basement is provided to minimize the possibility of fI corrosion cracking of certain metal components during operation of the ECCS I following a LOCA. The TSP provides this protection by dissolving in the sump water and causing its final pH to be raised to greater than or equal to 7.0.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing pro-vide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The requirement to dissolve a representative sample of TSP in a sample of water borated within RWSP boron concentration limits provides assurance that the stored TSP will dissolve in borated water at the postulated post-LOCA temperatures.

The requirement to verify the minimum pump discharge pressure on recircula-tion flow ensures that the pump performance curve has not degraded below that used to show that the pump exceeds the design flow condition assumed in the safety analysis and is consistent with the requirements of ASME Section XI.

3/4.5.4 REFUELING WATER STORAGE POOL (RWSP)

The OPERABILITY of the refueling water storage pool (RWSP) as part of the ECCS also ensure's that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWSP minimum volume and boron concentration ensure that (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWSP and the RCS water volumes with all CEAs instrted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

WATERFORD - UNIT 3 8 3/4 5-2

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NPF-38-64 ATTACHMENT B NS41294

EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T avg GREATER THAN OR EQUAL TO 350 F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent emergency core cooling system (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE high pressure safety injection pump,

b. One OPERABLE low pressure safety injection pump, and

c. An independent OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal and automatically transferring suction to the safety injection system' sump on a recirculation actuation signal.

APPLICABILITY: MODES 1, 2, and 3*F ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 nours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days des-cribing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be previdad in this Special Report whenever its value exceeds 0.70.

With pressurizer pressure greater than or equal to 1750 psia.

$ kidh F?CS avenge twpemiure greeler kan or epal to SDo*E WATERFORD - UNIT 3 3/4 5-3 i

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T LESS THAN 350 F avo LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE high pressure safety injection pump, and

b. An OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal and auto-matically transferring suction to the safety injection system sump on a recirculation actuation signal. .

APPLICABILITY: MODES 3* and 4.

A_GTION:

a. With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTOOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable !

Surveillance Requirements of 4.5.2.

With pressurizer pressure less than 1750 psia c

od Me RCS "#57" "Y" "

less finn inno C l

WATERFORD - UNIT 3 3/4 5-8 l

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EMERGENCY CORE COOLING SYSTEMS BASES '

500 F and (4e 4t5 premre beloc /7ho b f i%

ECCS SUBSYSTEMS (Continued)

With the RCS temperature below , one OPERABLE ECCS subsystem is l acceptable without single failure consideration on the basis of the stable reactivity condition of the reacter and the limited core cooling requirements.

The trisodium phosphate dodecahydrate (TSP) stored in dissolving baskets located in the containment baseraent is provided to minimize the possibility of corrcsion cracking of certain metal components during operation of the ECCS following a LOCA. The TSP provides this protection by dissolving in the sump water and causing its final pH to be raised to greater than or equal to 7.0.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assurnptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing pro-vide assurance that proper ECCS flows will.be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total cump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA enalyses. The requirement to dissolve a representative sample of TSP in a sample of water borated within RWSP boron concentration limits provides assurance that the stored TSP will dissolve in borated water at the postulated post-LOCA temperatures.

The requirement to verify the minimum pump discharge pressure on recircula-tion flow ensures that the pump performance curve has not degraded below that {

used to show that the pump exceeds the design flow condition assumed in the safety analysis and is censistent with the requirements of ASME Section XI.

3/4.5.4 REFUELING WATER STORAGE POOL (RWSP)

The OPERABILITY of the refueling water storage pool (RWSP) as part of the ,

ECCS also ensures that a sufficient supply of borated water is available for I injection by the ECCS in the event of a LOCA. The limits on RWSP minimum volume and boron concentration ensure that (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWSP and the RCS water volumes with all CEAs inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

WATERFORD - UNIT 3 8 3/4 5-2 L

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-65 This is a request to change Technical Specification 3/4.1.1.1, Shutdown Margin -

Any CEA Withdrawn and Technical Specification 3/4.1.1.2, Shutdown Margin - All CEAs Inserted.

Existing Specification See Attachment A.

Proposed Specification See Attachment B.

Description The proposed change would revise Technical Specification Limiting Condition for Operation (LCO) 3.1.1.1, Shutdown Margin - Any CEA Withdrawn and Technical Specification LC0 3.1.1.2, Shutdown Margin - All CEAs Inserted. The reason for this change is to make the actual value of shutdown margin that is required when all CEAs are inserted (LC0 3.1.1.2) consistent with the value that was used in the Cycle 2 Safety Analysis. In addition, several administrative changes are being proposed to clarify LCOs 3.1.1.1 and 3.1.1.2 and make them consistent with normal plant operating procedures.

The proposed change consists of the following:

1. Eliminate the Mode 2 Applicability and associated notes from LC0 3.1.1.2.

Since Mode 2 requires a k-eff of equal to or greater than 0.99 and the Shutdown Margin LC0 when all CEAs are inserted requires a shutdown margin of at least 1.0% (higher if T-cold is greater than 400 F), the two conditions are mutually exclusive. That is, it is impossible to satisfy the Shutdown Margin requirements of LC0 3.1.1.2 and, at the same time, achieve Mode 2 operation.

2. Delete surveillance requirements 4.1.1.2.la and 4.1.1.2.2 because both of these requirements were applicable only during Mode 2 operation. Since LC0 3.1.1.2 will no longer be applicable in Mode 2, these surveillance require-ments are superfluous.

3. Delete the reference to Modes 3, 4 and 5 from surveillance requirement 4.1.1.2.lb and incorporate the remaining requirements into surveillance requirement 4.1.1.2. This is simply an editorial change to reflect the fact that there will now be only one surveillance requirement associated with this LC0 and it will be applicable whenever the LCO is applicable.

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4. Change the index page and the headings of both LC0 3.1.1.1 and LC0 3.1.1.2 to refer to " full length CEAs" rather than simply "CEAs". This is simply a clarification to reflect the fact that Part-Length CEAs (PLCEAs) were not credited in the safety analysis that was performed to justify this change nor are they credited in the actual calculation of Shutdown Margin as discussed in the Definitions (Section 1.0) of the Technical Specifications.

5. Change the required value of Shutdown Margin in Figure 3.1-0 (LCO 3.1.1.2) from 4.15% to 4.10% when T-cold is greater than 500 F. This change simply reflects the actual value of Shutdown Margin that was assumed in the Cycle 2 Safety Analysis.

Although the proposed change is being submitted to correct some minor inconsis-tencies between the current Shutdown Margin LCOs and the Cycle 2 safety analysis, it is anticipated that the proposed change will remain bounding for Cycle 3 as well.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No.

The proposed change is primarily administrative and will therefore have no impact on the Cycle 2 safety analysis. If the reactor has a neutron multi-plication factor (k-eff) of equal to or greater than 0.99 (Mode 2) it would be impossible to maintain a shutdown margin equal to or greater than 1.0%

with all full-length CEAs inserted. The proposed change simply removes this impossible condition from the Tech Specs.

Changing the required Shutdown Margin when all full-length CEAs are inserted and the RCS inlet temperature (T-cold) is greater than 500 F from 4.15% to 4.10% is being done to make the Tech Spec Shutdown Margin requirement consistent with the Cycle 2 safety analysis. That is, for those safety analysis events that are affected by Shutdown Margin with all CEAs inserted, a Shutdown Margin value of 4.10% was assumed. Since the proposed change is consistent with the safety analysis and since the results of the safety analysis have been shown to be acceptable for all events, the proposed change will not result in a significant increase in the probability or consequences of any accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

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I The proposed change does not involve any new equipment or procedures nor does it result in any physical change to plant systems, structures or components. Although approval of the proposed change would allow minor revisions to the Shutdown Margin surveillance procedure, these revisions will be reviewed and approved by appropriate plant personnel prior to implementation as required by the Administrative Controls in the Tech Specs.

Thus, operation of the facility in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in the margin of safety?

Response: No.

The intent of this Tech Spec is to ensure that the reactor remains sub-critical following any design basis event (DBE) or anticipated operation occurrence (A00). Since Mode 2 cannot be achieved under the conditions of LC0 3.1.1.2, removal of those portions of the LCO which involve Mode 2 operation is strictly an administrative change and will have no effect on the capability of the plant safety systems to maintain the reactor in a subcritical condition following any DBE or A00. In addition, since the proposed Shutdown Margin value of 4.10% for LC0 3.1.1.2 was used as a direct input to the Cycle 2 safety analysis and since the results of the Cycle 2 safety analysis are acceptable for all events, the proposed change will not involve a significant reduction in the margin of safety.

The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant harards consideration. Example (i) relates to a change which is purely administrative: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomen-clature.

In this case, the proposed change is similar to Example (i) in that removing an impessible condition from the Tech Specs (i.e., operation in Mode 2 with all full-length CEAs inserted) is purely an administrative change. Other changes proposed by this amendment are being submitted to provide more consistency between the Cycle 2 Safety Analysis and the Tech Specs.

Safety and Significant Hazards Determination Based on the abnve Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; (2) there is reasonable assurance that the health and safety of the public will not be endangored by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environ-ment as described in the NRC Final Environmental Statement.

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NS41304 s ;

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NPF-38-65 ATTACHMENT A i

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INDEX 1 l 3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

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l SECTION PAGE 3/4.0 APPLICABILITY.............................................. 3/4 0-1 3/4.1 ~ REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - ANY CEA WITHDRAWN.................. 3/4 1-1 SHUTDOWN MARGIN - ALL CEAS FULLY INSERTED............ 3/4 1-3 1 MODERATOR TEMPERATURE COEFFICIENT.................... 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY.................. 3/4 1-5 l 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUTD0WN................................ 3/4 1-6 FLOW PATHS - 0PERATING............................... 3/4 1-7 CHARGING PUMPS - SHUT 00WN............................ 3/4 1-8 ,

CHARGING PUMPS - 0PERATING........................... 3/4 1-9

.I BORIC ACID MAKEUP PUMPS - SHUT 00WN................... 3/4 1-10 BORIC ACID MAKEUP PUMPS - OPERATING.................. 3/4 1-11 BORATED WATER SOURCES - SHUTD0WN..................... 3/4 1-12 I i

BORATED WATER SOURCES - OPERATING.................... 3/4 1-14 j BORON DILUTION....................................... 3/4 1-15 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P0SITION......................................... 3/4 1-18 POSITION INDICATOR CHANNELS - OPERATING.............. 3/4 1-21 POSITION INDICATOR CHANNELS - SHUTDOWN............... 3/4 1-22 CEA DROP TIME........................................ 3/4 1-23 SHUT 00WN CEA INSERTION LIMIT......................... 3/4 1-24 REGULATING CEA INSERTION LIMITS...................... 3/4 1-25 PART-LENGTH CEA INSERTION LIMITS..................... 3/4 1-28 WATERFORD - UNIT 3 IV AMEN 0 MENT NO.11 l

3/4.1 REACTIVITY CONTROL SYSTEMS

, ,3/4.1.1 _B0 RATION CONTROL SHUTOOWN MARGIN - ANY CEA WITHORAWN LIMITING CONDITION FOR OPERATION 3.1.1.L The SHUTDOWN MARGIN shall be greater than or equal to 5.15%

delta k/k when T,yg is greater than 200*F or 2.0% delta k/k when T,yg is less than or equal to 200*F.

APPLICABILITY: MODE! 1, 2", 3, 4, and 5 with any full length CEA fully or partially withdrawn.

ACTION:

With the SHUTOOWN MARGIN less than that required above, immediately initiate and continue boratfor, at greater than or equal to 40 gpm of a solution con- l taining greater than or equal to 1720 ppm boron or equivalent until the required SHUTOOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 With any full length CEA fully or partially withdrawn, the SHUTOOWN MARGIN shall be determined to be greater than or equal to that required above:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTOOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable CEA(s),

b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Transient Insertion Limitt, of Specification 3.1.3.6.

c. When in MODE 2 with K,79 less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

I See Special Test Exception 3.10.1.

WATERFORD - UNIT 3 3/4 1-1 AMEN 0 MENT NO. 11

e REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e. When in MODE 3, 4, or 5, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration '

of at least the following factors:

1. Reactor Coolant System boron concentration, !

2. CEA position,

3. Reactor Coolant System average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le., above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPDs after each fuel loading.

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WATERFORD - UNIT 3 3/4 1-2 AMEN 0 MENT NO. 11

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I REACTIVITY CONTROL' SYSTEMS-SHUTDOWN MARGIN - ALL CEAS FULLY INSERTED LIMITING CONDITION FOR OPERATION l,- 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to that shown in i Figure 3.1-0.

APPLICABILITY: MODE 2*#, 3, 4 and 5 with all CEAs fully inserted.

l ACTION:

With the SHUTOOWN MARGIN less than that shown in Figure 3.1-0, immediately ,

initiate and continue boration at greater than or equal to 40 gpm of a solu- j tion containing greater than or equal to 1720 ppm boron or equivalent until l the required SHUTDOWN MARGIN is restored. I SURVEILLANCE REQUIREMENTS 4.1.1.2.1 With all full length CEAs fully inserted, the SHUTOOWN MARGIN shall be determined to be greater than or equal to that shown in Figure 3.1-0.

a. When in MODE 2 with k,ff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6. l

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b. When in MODES 3 4, or 5, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by considera--

tion of the following factors: .J 1

1. Reactor Coolant System boron concentration,

2. CEA position,

3. Reactor Coolant System average temperature, '

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration,and

6. Samarium concentration. j 4.1.1.2.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1.0% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.2.lb, above. The predicted reac-tivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

With k,ff less than 1.0

See Special Test Exception 3.10.1 ll WATERFORD - ' UNIT 3 3/4 1-3 AMEN 0 MENT NO. 11

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l i I I I 0' 100' 2001 300 400! 500' 600' COLD LEG TEMPERATURE (*F) i Figure 3.1-0 SHUTDOWN MARGIN AS A FUNCTION OF COLD LEG TEMPERATURE WATERFORD - UNIT 3 3/4 1-3a AMENDMENT NO. 11

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l NPF-38-65 1

ATTACHMENT B j

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l NS41304 j

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INDEX 1 1

1 3 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION PAGE l

3/4.0 APPLICABILITY................................... .......... 3/4 0-1 i 3/4.1 -1tEACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL FML)EA

% SHUTDOWN MARGIN - ANY CEA WITHDRAWN.................. 3/4 1-1 Fm L.ENGTH SHUTDOWNMARGIN-ALDCEASFULLYINSERTED............ 3/4 1-3 MODERATOR TEMPERATURE COEFFICIENT.................... 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY.................. 3/4 1-5 3/4.1.2 BORATION SYSTEMS 1 FLOW PATHS - SHUTD0WN................................ 3/4 1-6 FLOW PATHS - 0PERATING............................... 3/4 1-7 CHARGING PUMPS - SHUTD0WN............................ 3/4 1-8 CHARGING PUMPS - 0PERATING........................... 3/4 1-9 BORIC ACID MAKEUP PUMPS - SHUTD0WN................... 3/4 1-10 B' 4CID MAKEUP PUMPS - OPERATING.................. 3/4 1-11 Bt WATER SOURCES - SHUTD0WN..................... 3/4 1-12 BORATtD WATER SOURCES - OPERATING.. ................. 3/4 1-14 GORON 0ILUTION....................................... 3/4 1-15 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P0SITION......................................... 3/4 1-18 POSITION INDICATOR CHANNELS - OPERATING.............. 3/4 1-21 POSITION INDICATOR CHANNELS - SHUTDOWN............... 3/4 1-22 CEA DROP TIME........................................ 3/4 1-23 SHUTDOWN CEA INSERTION LIMIT......................... 3/4 1-24 REGULATING CEA INSERTION LIMITS...................... 3/4 1-25 ,

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PART-LENGTH CEA INSERTION LIMITS..................... 3/4 1-28 WATERFORD - UNIT 3 IV AMENDMENT NO.11 i

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3/4.1 REACTIVITY CONTROL SYSTEMS

' 'e3/4 1..1 BORATION CONTROL h

Fult LMard SHUTDOWN MARGIN - ANY CEA WITHORAWN LIMITING CONDITION FOR OPERATION 3.1.1.L. The SHUTDOWN MARGIN shall be grester than or equal to 5.15%

delta k/k when T,yg is greater than 200'F or 2.0% delta k/k when T,yg is less than or equal to 200*F.

APPLICABILITY: MODES 1, 2*, 3, 4, and 5 with any full length CEA fully or partially withdrawn.

ACTION:

With the SHUTDOWN MARGIN less than that required above, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution con- l taining greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 With any full length CEA fully or partially withdrawn, the SHUTOOWN MARGIN shall be determined to be greater than or equal to that required abovo:

t

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperabic CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN chall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable CEA(s),

b.

When in MODE 1 or MODE 2 with K,7f greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Transient Ins,ertion Limits of Specification 3.1.3.6.

c. When in MODE 2 with K,ff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

WATERFORD - UNIT 3 3/4 1-1 AMENDMENT NO. 11

REACTIVITY CONTROL SYSTEMS - ~

FULL LEN6 TH SHUTDOWN MARGIN - ALL CEAS FULLY INSERTED LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal to that shown in Figure 3.1-0.

gg APPLICABILITY: H00E@ 3, 4 and 5 with all CEAs fully inserted, j

ACTION:

With the SHUTDOWN MARGIN less than that shown in Figure 3.1-0, immediate'1y initiate and continue boration at greater than or equal to 40 gpm of a solu-tion containing greater than or equal to 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 d With all full length CEAs fully insertedt the SHUTOOWN MARGIN shall be determined to be greater than or equal to that shown in Figure 3.1-O r g n n MODE"2 wi k le t n1 , wit 4h s to a iev or ti b erif ' ig th the g . dict rit#< al jogCre mits pos' ion ; thi the Specif4uti 3.1.3.6.

i MODE 3. 4 or .

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by c

[tionofthefollowingfactors:

[ 1. Reactor Coolant System boron concentration,

2. CEA position,

3. Reactor Coolant System average temperature,

4. Fuel burnup based on gross thermal energy generation, S. Xenon concentration,and I

6. Samarium concentration.

' 4.1. 2.2 The overall core reactivity balance shall be compared to predicted va s to demonstrate agreement within i 1.0% delta k/kjtdeast o qe per 31 E ectivefullPower, Days (EFPD). T) tis' co arison s..Tl consi at 13 aft hose fjetors stated'in Specification 4. 1.2.lb, ove. T predigtid re tivity values shal'1 be adjuste ((normp ized) to rrespo to thq 4ctua ore conftionspr)'oftoexceedi a fueVburnup of 6 EFPD after each fuel loading.

/ /

[*Withk[f less than .0

See (Decial Test Exception 3.10.1, WATERFORD - UNIT 3 3/4 1-3 AMENDMENT NO. 11

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Figure 3.1-0 SHUTDOWN MARGIN AS A FUNCTION OF COLD LEG TEMPERATURE WATERFORD - UNIT 3 3/4 1-3a AMENDMENT NO. 11

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NPF-38-66 i

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- ' " - - ~ - - - - - . _ _ _ . . _ _ _ _ _ _ _ , _ _ _ _ _ _

i DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-66

.This is a request to revise Technical Specification'3/4.7.10.2,. Fire Suppression Systems - Spray and/or Sprinkler Systems. !

l Existing Specification c See Attachment A. l l

Proposed Specification See Attachment B.

I Description The proposed change would revise Technical Specification Limiting Condition for Operation (LC0) 3.7.10.2 and its associated Surveillance Requirement 4.7.10.2. ,

The reason for this change is to implement operability and surveillance require- l ments for newly installed fire suppression equipment in Waterford 3's charcoal filtration system.

LC0 3.7.10.2 currently. specifies the number and location of sprinkler and ' spray systems that must be maintained in an operable condition. The proposed change would simply add the newly installed charcoal filter spray system to this LCO. ,

In addition, the proposed change adds surveillance requirement 4.7.10.2d which l requires an air flow test through each open. head spray or sprinkler system every :

three years to verify that the spray nozzles are unobstructed. This surveillance requirement is being proposed to make the Waterford 3 Tech Spec surveillance requirements more closely conform with those in the standard Technical Speci-

'fications. However, since the charcoal filter spray system nozzles can be accessed only when the charcoal has been removed, a note has been added to the 4 proposed surveillance requirement indicating that a visual inspection of the ;

charcoal filter system spray nozzles verifying that they are unobstructed may be performed in lieu of the air flow test. The visual inspection will be performed every time the charcoal is changed.

Safety Analysis i The proposed change described above shall be deemed to involve a significant i hazards consideration if there is a positive finding in any of the following areas:

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

( l Response: No.

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NS41307.

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.. 1 The charcoal filter units in the Shield Building Ventilation System (SBVS)- I are used to adsorb iodine.and other radionuclides that leak into the reactor annulus-building following a LOCA. Similarly, the charcoal filter i units in the Controlled Ventilation Area System (CVAS) are used to adsorb iodine and other. radionuclides that leak'from the safety-related equipment used to mitigate the consequences of a LOCA (primarily high pressure safety injection (HPSI) pumps during the recirculation phase). The oper- ,

ability of at least one SBVS train and one CVAS train is required to meet the offsite dose criteria established by 10CFR100. .The addition of fire suppression equipment specifically designed to extinguish fires.in these two filtration systems increases the probability that the SBVS and CVAS will be able to perform their design function following.a LOCA. There- t fore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated. l

}

2. Will operation of the facility in accordance with the proposed cha'nge create the possibility of a new or different kind of accident from any accident previously evaluated? l Response: No.

New fire suppression equipment has-been added to increase the overall level of fire protection at Waterford 3 and the proposed change simply implements the operability and surveillance requirements necessary to ensure the equipment remains functional. Since the new equipment-is designed to protect portions of the SBVS and CVAS and these systems are required only during post accident conditions, the proposed change will not create the possibility of a new or different kind of accident from )

any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change !

involve a significant reduction in the margin of safety?

Response: No.

The intent of this Specification is to ensure that adequate fire suppres-sion capability is available to confine and extinguish fires occurring in i any portion of the facility where safety-related equipment is located.

Since the addition of fire suppression equipment in the charcoal

' filtration systems will add to the capability of extinguishing fires in 1 the SBVS and CVAS, the proposed change will result in an increase in the margin of safety.

The Commission has provided guidance concerning the application of standards for determining whether significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to f involve significant hazards consideration. Example (ii) relates to a change i that constitutes an additional limitation, restriction or control not presently I included in the Technical Specifications (e.g., a more stringent surveillance requirement). l l

NS41307 I I

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In this case the proposed change is similar to Example (ii) in that it consti- J tutes additional operability and surveillance requirements for newly installed fire suppression equipment that is not now a part of the Technical Specifica-tions.

Safety and Significant' Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed ;

change does.not constitute a significant hazards consideration as defined by 10CFRb0.92; (2) there is reasonable assurance that the health and safety of the ]

4 public will not be endangered by the proposed change; and (3) this action will j not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement. l NS41307

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NPF-38-66 ATTACHMENT A f

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i NS41307

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PLANT SYSTEMS <l 1

SPRAY AND/OR SPRINKLER SYSTEMS

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LIMITING CONDITION FOR OPERATION 3.7.10.2 The following spray and/or sprinkler systems shall be OPERABLE:

Sprinkler No. Bida./Elev. Location FPM-1 RCB Reactor Coolant Pumps 1A, 1B FPM-2 RCB Reactor Coolant Pump 2A, 2B FPM-3A RAB +21, +46 Diesel Generator Area A, Feed Tank Room A !

FPM-48 RAB +21, +46 Diesel Generator Area B, Feed Tank Room B FPM-11A RAB -35 Emergency D/G Fuel Oil Tank A FPM-128 RAB -35 Emergency D/G Fuel Oil Tank B FPM-16 FWPH +15 Fire Water Pump House FPM-17 RAB +35 Cable Vault Area FPM-18 RAB +35 Electrical Penetration Area 1 FPM-19 RAB +35 Electrical Penetration Area 2 FPM-22 RAB -4 Corridor and Blowdown Tank Rooms FPM-23 RAB -35 Corridor, Shutdown Heat Exchanger Rooms, EFW Pump Room FPM-24 RAB +21 Corridors, CCW Area FPM-258 RAB +21 North High Voltage Switchgear Room FPM-26 RAB +46 Ventilation Equipment Rooms FPM-27 RAB +7 HVAC Rooms FPM-28 RAB -35 Auxiliary Component Cooling Water Pump Rooms FPM-29 RAB +35 Relay Room, Corridor FPM-30A RAB +21 South High Voltage Switchgear Room APPLICABILITY: Whenever equipment protected by the spray / sprinkler system is required to be OPERABLE.

A_ CTION:

a. With one or more of the above required spray and/or sprinkler systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged unless the spray and/or sprinkler system (s) is located inside the containment, then inspect that containment area at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor air temperature at.least once per hour at the locations listed in Specifi-cation 4.6.1.5; for other areas, establish an hourly fire watch patrol,

b. The provisions of 0 specifications 3.0.3 and 3.0.4 are not applicable.

)

WATERFORD - UNIT 3 3/4 7-32 h

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PLANT SYSTEMS i

SURVEILLANCE REQUIREMENTS i

4.7.10.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secure.d in position is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

c. At least once per 18 months:

1. By performing a system functional test which includes simulated automatic actuation of the system, and:

a) Verifying that the automatic valves in the flow path actuate to their correct positions on a thermal /preaction test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2. By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity, and

3. By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.

WATERFORD - UNIT 3 3/4 7-33

, , _ _ _ - - - -.7--- -, . - - - - . - - - - _ _ - - . . . - - - - _ - - , . _ - - - - - - - , . , _ , _ , - - - - - , - , - - - - - - - - , _ , , . - - - - . - - - - - - , , - - - - - - - - - - - - , - - - - - - . - - , , , , , - - - . - - - , _ _ . , - , - - - - , . _ - - - - - , - - - -

--,-------.--_-----,,,.,-,7---.---

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NPF-38-66 ATTACHMENT B i

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NS41307

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i 3 ANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS

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LIMITING CONDITION FOR OPERATION 3.7.10.2 The following spray and/or sprinkler systems shall be OPERABLE:

Sprinkler No. Bldo./Elev. Location FPM-1 RCB Reactor Coolant Pumps 1A, IB FPM-2 RCB Reactor Coolant Pump 2A, 28 FPM-3A RAB +21, +46 Diesel Generator Area A, Feed Tank Room A FPM-4B RAB +21, +46 Diesel Generator Area B, Feed Tank Room B FPM-11A RAB -35 Emergency D/G Fuel Oil Tank A FPM-12B RAB -35 Emergency D/G Fuel Oil Tank B FPM-16 FWPH +15 Fire Water Pump House FPM-17 RAB +35 Cable Vault Area FPM-18 RAB +35 Electrical Penetration Area 1 FPM-19 RAB +35 Electrical Penetration Area 2 FPM-22 RAB -4 Corridor and Blowdown Tank Rooms FPM-23 RAB -35 Corridor, Shutdown Heat Exchanger Rooms, I

EFW Pump Room FPM-24 RAB +21 Corridors, CCW Area '

FPM-258 RAB +21 North High Voltage Switchgear Room FPH-26 RAB +46 Ventilation Equipment Rooms FPM-27 RAB +7 HVAC Rooms FPM-28 RAB -35 Auxiliary Component Cooling Water Pump Rooms FPM-29 RAB +35 Relay Room, Corridor JNSERT FPM-30A RAB +21 South High Voltage Switchgear Room

_- -4 APPLICABILITY: Whenever equipment protected by the spray / sprinkler system is required to be OPERABLE.

ACTION:

a. With one or more of the above required spray and/or sprinkler systems inoperable, within I hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged unless the spray and/or sprinkler system (s) is located inside the containment, then inspect that containment area at least once per B hours or monitor air temperature at.least once per hour at the locations listed in Specifi-cation 4.6.1.5; for other areas, establish an hourly fire watch patrol.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

WATERFORD - UNIT 3 3/4 7-32

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INSERT A I

'FPM-33 RAB +46 E-17(3A-SA) Shield Building Ventilation System Filter, Train A Charcoal Adsorber FPM-34 RAB +46 E-17(3B-SB) Shield Building Ventilation System Filter, Train B Charcoal Adsorber FPM-36 RAB +46' E-23(3A-SA) Controlled Ventilation Area System Filter, Train A Charcoal Adsorber FPM-37 RAB +46 E-23(3B-SB) Controlled Ventilation Area System Filter, Train B Charcoal Adsorber i

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.10.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

c. At least once per 18 months:

4

1. By performing e system functional test which includes simulated automatic actuation of the system, and:

a) Verifying that the automatic valves in the flow path actuate to their correct positions on a thermal /preaction test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2. By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity, and

3. By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.

d. M kst once pe< s years by pg,,,,io, an air ,q , tese ihrcugh each open head symy end synnkler header " Sydem lis}ed in % bon 3. 710.1 cnd verWying the spity riozzles are unobskucted.

To lieu of en gir flow best, %e charemt fIter sys)en, spmy ncerles l need only be visucIly inspected and veri 6ed to be unob.sfrucled ecch fime the chcrtocl is changed WATERFORD - UNIT 3 3/4 7-33 1

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L NPF-38-67 l1 2

I NS41320

DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-67 This is a request to revise Technical Specification Table 3.8-1, Containment Penetration Conductor Overcurrent Protective Devices as it applies to the Polar Crane's containment electrical penetration.

Existing Specification See Attachment A.

Proposed Specification See Attachment B. :

Description 4 The proposed change would add the as-built primary and backup overcurrent protective devices used to protect the polar crane's containment electrical penetration to Technical Specification Table 3.8-1. The reason for the proposed change is to document the fact that the polar crane's containment electrical ,

penetration nas adequate overcurrent protection so that it may be energized when the reactor is in Modes 1, 2, 3 or 4.

Technical Specification Limiting Condition for Operation (LCO) 3.8.4.1 currently requires all containment penetration conductor overcurrent protection devices shown on Table 3.8-1 to be operable whenever the reactor is in Modes 1, 2, 3 or i

4. With respect to the polar crane, Table 3.8-1 currently states that the primary breaker is locked out in the open position during operation in Modes 1, 2, 3 and 4; hence, there is no requirement to list the actual primary or backup protection. The proposed change will simply remove the sentence stating that the primary breaker is locked out in the open position during operation in Modes 1, 2, 3 and 4, and insert the as-built overcurrent protective devices that are actually installed.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas: 1 l

1. Will operation of the facility in accordance with the proposed change involve a significant increase in the probability or consequences of ,

any accident previously evaluated? I Response: No.

The primary safety concern to be addressed in the use of the polar crane is the potential for heavy loads that, if dropped, could impact irradiated fuel in the reactor vessel or equipment necessary for the safe shutdown of the reactor. As a result of this concern the NRC, in July 1980, published '

NS41320

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! NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants", followed by a j Generic Letter (December 22, 1980) which detailed the degree of compliance required for NUREG-0612.

LP&L has made a number of submittals to the NRC in response to the heavy l loads requirements and the subsequent generic letter. Requirements for safe load paths, crane operator qualifications, crane inspection and testing, etc., have been addressed to ensure that heavy load drops will either not occur or will occur with consequences no more severe than what has been previously evaluated.

In Section 9.1 of Supplements 6 and 8 to the Waterford 3 Safety Evaluation ;

Report (NUREG-0787), the NRC indicated that Waterford 3 has satisfied all i applicable requirements of NUREG-0612. Thus, the proposed change will l not involve a significant increase in the probability or consequences of 1 any accident previously evaluated. !

2. Will operation of the facility in accordance with the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve any physical modifications to plant systems, structures or components. The overcurrent protection devices described previously reflect the actual as-built components that are already an integral part of the polar crane's electrical system and meet all applicable requirements for overcurrent protection. In addition, since the polar crane is used very infrequently in Modes 1 through 4, LP&L plans to continue the current practice of locking out the primary breaker in the open position whenever the polar crane is not in use (as described in the existing Tech Spec). Thus, since there has been no change to the level of overcurrent protection for the polar crane's containment electrical penetration, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with the proposed change involve a significant reduction in the margin of safety?

Response: No.

The intent of this Specification is to ensure there is adequate over-current protection for all containment electrical penetrations. As described in Tech Spec Bases 3/4.8.4, " Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the operability of primary and backup overcurrent protection circuit breakers during periodic surveillance". Since the proposed change utilizes both the above mentioned methods of overcurrent protection, and since the heavy loads requirements of NUREG-0612 have been satisfied, the proposed change will not involve a significant reduction in the margin of safety.

NS41320

The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards consideration. Example (1) relates to a purely administrative change to Technical Specifications (i.e., a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature. Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specification (i.e., a more stringent surveillance requirement).

In this case, the proposed change is similar to Examples (i.) and (ii). It is similar to Example (i) because the proposed change simply updates the Tech Specs to reflect the actual equipment already installed in the plant. In addition, the proposed change makes the polar crane's overcurrent protection requirements l

consistent with all of the other items from Tech Spec Table 3.8-1 that require power from the 480 volt low voltage switchgear. It is similar to Example (ii) in that the proposed change adds operability and surveillance requirements for the primary and backup overcurrent protection devices that are not currently part of the Technical Specifications.

Safety and Significant Hazards Determination Based upon the above Safety Analysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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