Text

Proposed Tech Specs,Revising Axial Shape Index Allowable Ranges,Moderator Temp Coefficient Allowable Ranges,Insertion Limits for Part Length of Control Element Assemblies & Adding Reactor Vessel Level Monitoring Sys
	ML20203A976
Person / Time
Site:	Waterford
Issue date:	07/15/1986
From:	; LOUISIANA POWER & LIGHT CO.
To:	;
Shared Package
ML20203A921	List: ML20203A976; W3P86-2313, Application for Amend to License NPF-38,consisting of Tech Spec Change Requests NPF-38-24,-25,-26,-27 & -28 Re Revs to Axial Shape Index Allowable Ranges,Moderator Temp Coefficient Allowable Ranges & Insertion Limits.Fee Paid;
References
	NUDOCS 8607170372
	Download: ML20203A976 (45)
	v • d • e

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I POWER DISTRIBUTION LIMITS :

3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a. COLSS OPERABLE

-0.23 1 ASI 1 + 0.50

b. COLSS OUT OF SERVICE (CPC)

-0.15 $ ASI 1 + 0.50 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 less than 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.

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See Special Test Exception 3.10.2.

T 8607170372 860715 PDR ADOCK 05000382 P PDR .

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POWER DISTRIBUTION LIMITS 3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a. COLSS OPERABLE

0. 50- + C,2Eli

-0.23 1 ASI 1

b. COLSS OUT OF SERVICE (CPC)

-o,si-9-Fe 5 ASI $ 0.50 + 0 ,2 2.

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 less than 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.

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See Special Test Exception 3.10.2.

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-25 This is a request to revise Technical Specification 3.1.1.3, Moderator Temperature Coefficient.

Existing Specification See Attachment A.

Proposed Specification See Attachment B.

Description The proposed change revises the moderator temperature coefficient (MTC) limits of Technical Specification 3.1.1.3. This Technical Specification imposes limitations on the MTC to ensure that the assumptions used in the accident and transient analyses remain valid through the fuel cycle.

The proposed change is necessary to accommodate the change in core characteristics from Cycle 1 to Cycle 2.

Technical Specification 3.1.1.3 currently states that the MTC shall be less negative than -2.5x10-4 Ak/k/*F at rated thermal power, and less positive than +0.2x10- Ak/k/*F at or below 70% thermal power, and less positive than 0.0x10-4 Ak/k/*F above 70% thermal power. The proposed change will extend the most negative limit to -3.3x10-4 Ak/k/*F at all levels of thermal power and will extend the most positive limit to 0.5x10-4 Ak/

at or below 70% thermal power. Thecurrentpositivelimitof0.0x10g/*F Ak/k/*F above 70% thermal power is not modified by this change.

The above changes are necessary for Cycle 2 to address the MTC associated with higher fuel burnup at the end of the fuel cycle and higher boron concentration at the beginning of the fuel cycle. However, the proposed negative limit is more negative than is required for Cycle 2 to accommodate anticipated future cycles.

As part of the Cycle 2 reload analysis, the FSAR Chapter 15 accident analyses are reviewed for potential impact caused by this change. An example of a potential impact is for the main steam line break event which results in a rapid cooldown of the reactor coolant system. E.en a more negative MTC is used in the analysis, a more positive reactivity addition occurs due to the cooldown. For the steam line break event, n. greater return to power (although not necessarily worse consequences) could result. The accident analyses will use the applicable range of MTC values (based on the proposed change),

depending upon the power level assumed for the accident or transient. The results of the Cycle 2 analyses will comply with the acceptance criteria of the Standard Review Plan.

Approval of the proposed change is requested by December 15, 1986 in order to support Cycle 2 operation.

Safety Analysis The proposed changes 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 These changes are proposed to accommodate the wider range in MTC values necessary to support an increased cycle length for Cycle 2. As such, they are normal and expected' changes.

FSAR Chapter 15 events that are limiting with respect to MTC will be repeated using the new MTC values for Cycle 2. The results of the revised analyses will not show a significant increase in the probability or consequences of any accident previously evaluated since the standard acceptance criteria for each event will be met.

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 This revision addresses the change in core characteristics from Cycle 1 to Cycle 2. As such, no new failure or accident path is created. Therefore, this change does not create the possibility of any new or different kind of accident.

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

Response: No This change imposes limits on the MTC value to ensure that the assumptions used in the FSAR Chapter 15 accident analyses remain valid throughout the cycle. The accident snalyses using the proposed MTC limits will demonstrate continued compliance to the applicable acceptance criteria. Therefore, this change will not involve a significant reduction in a margin of safety.

Safety and Significant llazards Determination Based upon the above Safety Aanlysis, it is concluded that (1) the proposed change does not constitute a significant hazards consideration as defined by 10CFR50.92; (2) there is a 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.

4 NPF-38-25 ATTACHMENT A 1

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

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a. Less positive than 0.2 x 10 delta k/k/*F whenever THERMAL POWER is 5,70% RATED THERMAL POWER, and

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b. Less positive than 0.0 x 10 delta k/k/ F whenever THERMAL POWER is > 70% RATED THERMAL POWER, and

c. Less negative than -2.5 x 10 ~4 delta k/k/ F at RATED THERMAL POWER.

APPLICABILITY: MODES 1 and 2*#

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1. 3. 2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a. . Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. At any THERMAL POWER, within 7 EFPD of reaching 40 EFPD core burnup.

c. At any THERMAL POWER, within 7 EFPD of reaching two-thirds of expected core burnup.

With K,ff greater than or equal to 1.0.

See Special Test Exception 3.10.2.

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NPF-38-25 ATTACHMENT B 4

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

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a. Less positive than 4-t x 10 delta k/k/*F whenever THERMAL POWER is < 70% RATED THERMAL POWER, and

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b. Less positive than 0.0 x 10 delta k/k/*F whenever THERMAL POWER is > 70% RATED THERMAL POWER, and

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c. Less negative than -+:t x 10 delta k/k/*F at A f THERMAL POWER.

APPLICABILITY: MODES 1 and 2*# M MM#

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1.3.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.

b. At any THERMAL POWER, within 7 EFPD of reaching 40 EFPD core burnup.

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c. At any THERMAL POWER, within 7 EFPD of reaching two-thirds of expected core burnup.

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With K,ff greater than or equal to 1.0. .

See Special Test Exception 3.10.2.

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-26 This is a request to revise Technical Specification 3.1.3.7, Part Length CEA Insertion Limits, and 4.1.3.7, the associated Surveillance Requirements for this Technical Specification.

Existing Specification See Attachment A.

Proposed Specification See Attachment B.

Description Technical Specification 3.1.3.7 imposes limits on the allowable position of the part length CEA (PLCEA) groups and on the allowable burnup span during which the PLCEA may remain within a given position range during Modes 1 and

2. Adherence to this technical specification will:

1. Eliminate the potential for unexpected reactivity addition which otherwise might occur should a PLCEA drop or move from a less to a more reactive axial position,

2. Prevent undesirable perturbations on the axial distribution of core burnup due to PLCEA insertion, and

3. Prevent unacceptably high axial peaking upon subsequent movement of the PLCEA groups.

Technical Specification 3.1.3.7 currently states that the PLCEA groups shall be restricted in position between 0" - 17" withdrawn (i.e. between fully inserted and 11% inserted) for a maximum period of 7 effective full power days (EFPD) out of any 30 EFPD period. The proposed change would replace the entire current technical specification and would add Figure 3.1-3 which:

1. Allows a maximum PLCEA insertion to 75% withdrawal (112.5 inches) during long term steady state operation above 20% thermal power,

2. Allows any PLCEA insertion below 20% thermal power (i.e. PLCEA insertion below 20% power has negligible effect on items 1-3, above), and l 3. Allows a maximum transient PLCEA insertion to 15% withdrawal (22.5 inches) between 50% and'20% thermal power for a specified limit burnup duration. The more restrictive PLCEA insertion limits provided by the proposed changes to the technical specification, including Figure 3.1-3, will be used in the Cycle 2 safety analyses.

There are two reasons for changing the technical specification.- First, the proposed change, by imposing more restrictive insertion limits, will provide i

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a i Surveillance Requirement 4.1.3.7 currently requires determination of the I

PLCEA group positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The proposed change would replace the current surveillance requirement with an equivalent require-

} ment to determine that the PLCEA groups are within the transient insertion

} range once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

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1. Will operation of the facility in accordance with this proposed change i

involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No i

) This proposed change provides additional assurance that adverse axial l shapes and rapid local power changes which affect radial power peaking i

factors and DNB considerations do not occur as a result of the part length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits along with surveillance i I requirements to ensure adherence with the insertion limits, this proposed j change does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

j 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 l accident previously evaluated?

i l Response: No j This proposed change provides additional assurance that adverse axial shapes and rapid local power changes which affect radial power peaking

factors and DNB considerations do not occur as a result of the part length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed.

change will impose more restrictive limits with respect to previously analyzed events, along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not create the possibility of a new or different kind of accident from any accident . j previously evaluated. 1 i l i

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

Response: No This proposed change provides additional assurance that adverse axial shapes and rapid local power changes which affect radial power peaking factors and DNB considerations do not occur as a result of the part length CEA group being positioned in the same axial segment of fuel assemblies for an extended period of time during operation. Because the proposed change will impose more restrictive limits along with surveillance requirements to ensure adherence with the insertion limits, this proposed change does not involve a significant reduction in 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 hazards considerations. Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Sepcifications; for example, a more stringent surveillance requirement.

In this case, the proposed change is similar to Example (ii) in that the change to Technical Specification 3.1.3.7 adds additional limitations on PLCEA insertion limits not presently included in 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 a 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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INDEX LIST OF FIGURES FIGURE PAGE 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND l TEMPERATURE AS A FUNCTION OF STORED BORIC ACID CONCENTRATION................................. 3/4 1-13 3.1-2 CEA INSERTION LIMITS VS THERMAL P0WER.............. 3/4 1-27 1

3.2-1 ALLOWA8LE PEAK LINEAR HEAT RATE VS BURNUP.......... 3/4 2-2 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON COLSS......... 3/4 2-8 3.2-3 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE). . . . . . 3/4 2-9 3.4-1 00SE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 pCi/ GRAM DOSE EQUIVALENT I-131....... 3/4 4-27 3.4-2 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (HEATUP)........................................... 3/4 4-30 3.4-3 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (C00LDOWN)......................................... 3/4 4-31 3.6-1 CONTAINMENT PRESSURE VS TEMPERATURE ............... 3/4 6-12 4.7-1 SAMPLING PLAN FOR SNUBBER FU;NCTIONAl.TEST.......... 3/4 7-26 5.1-1 EXCLUSION AREA........ ............................ 5-2 5.1-2 LOW POPULATION Z0NE........,........................ 5-3 5.1-3 SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS................................... 5-4 6.2-1 0FFSITE ORGANIZATION FOR MANAGEMENT AND TECHNICAL SUPP0RT................................. 6-3 l

6.2-2 PLANT OPERATIONS ORGANIZATION..............,........ 6-4 1

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REACTIVITY CONTROL SYSTEMS frs i: .i PART-LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.7 Part-length CEA groups positioned between 0" - 17" withdrawn shall be restricted to prevent the neutron absorber section of the part-length CEA group from covering the same axial segment of the fuel assemblies for a period in excess of 7 EFPD out of any 30 EFPD period.

APPLICABILITY: MODES 1 and 2. l ACTION:

With the neutron absorber section of the part-length CEA group covering any same axial segment of the fuel assemblies for a period exceeding 7 EFPD out of any 30 EFPD period, either:

a. Reposition the part-length CEA group to ensure no neutron absorber i section of the part-length CEA group is covering the same axial segment of the fuel assemblies within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

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 . \

SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of the part-length CEA group shall be determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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f N TH ex MA L Po weg IkX LIST OF FIGURES FIGURE PAGE 3.1-1 MINIMUM 80RIC ACID STORAGE TANK VOLUME AND TEMPERATURE AS A FUNCTION OF STORED 80RIC ACID CONCENTRATION......... ....................... 3/4 1-13 3.1-2 CEA INSERTION LIMITS VS THERMAL P0WER.............. 3/4 1-27 3.251 ALLOWA8LE PEAK LINEAR HEAT RATE VS 8URNUP.......... 3/4 2-2 3.2-2 DN8R MARGIN OPERATING LIMIT 8ASED ON COLSS......... 3/4 2-8 3.2-3 DN8R MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE).... .3/4 . 2-9 3.4-1 ~

DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 pCi/ GRAM DOSE EQUIVALENT I-131....... 3/4 4-27 3.4-2 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (HEATUP)........................................... 3/4 4-30 3.4-3 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITATIONS FOR 0-8 EFFECTIVE FULL POWER YEARS (C00LDOWN)......................................... 3/4 4-31 3.6-1 CONTAINMENT PRESSURE VS TEMPERATURE ............... 3/4 6-12 4.7-1 SAMPLING PLAN FOR SNUB 8ER FUNCTIONALTEST.......... 3/4 7-26 5.1-1 EXCLUSION AREA..................................... 5-2 S.1-2 LOW POPULATION Z0NE................................ 5-3 5.1-3 SITE 8OUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS...... . .......................... 5-4 6.2-1 0FFSITE ORGANIZATION FOR MANAGEMENT AND TECHNICAL SUPPORT.... . .........................

6-3 6.2-2 PLANT OPERATIONS ORGANIZATION...................... 6-4 WATERFORD - UNIT 3 XIX

REACTIVITY CONTROL SYSTEMS I

PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION f

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3.1.3.7 The part length CEA groups shall be limited to the insertion limits shown on Figure 3.1-3 with PLCEA insertion between the Long Term Steady State

Insertion Limit and the Transient Insertion Limit restricted to:

a. < 7 EFPD per 30 EFPD interval, and

b. < 14 EFPD per calender year.

APPLICABILITY: MODE 1 above 20% THERMAL POWER.

ACTION:

a. With the part length CEA groups inserted beyond the Transient 4 Insertion Limit, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours, either:

1. Restore the part length CEA group to within the limits, or

2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is

! allowed by the FLCEA group position using i

! Figure 3.1-3.

, b. With the part length CEA groups inserted between the Long Term Steady State Insertion Limit for intervals > 7 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, either:

1. Restore the part length group within the Long Term Steady State Insertion Limits within two hours, or

! 2. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i' SURVEILLANCE REQUIREMENT l

l 4.1.3.7 The position of the part length CEA group shall be determined to be within the Transient Insertion Limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

See Special Test Exception 3.10.2.

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-27 This is a request to modify Technical Specification 3.10.2, Moderator Temperature Coefficient, Group Height, Insertion, and Power Distribution Limits along with 4.10.2, the associated surveillance requirements for this technical specification, to allow suspension of the limitations specified in the proposed change to Technical Specification 3.1.3.7, Part Length CEA Insertion Limits (reference Technical Specification Change Request NPF-38-26) .

Existing Specification See Attachment A.

Proposed Specification See Attachment B.

Description This proposed change would revise Technical Specification 3.10.2 and the associated Surveillance Requirement 4.10.2 to allow suspension of the limits specified in the proposed change to Technical Specification 3.1.3.7.

The associated Bases Section is also revised to reflect technical terminology utilized at Waterford 3. This change is necessary to allow the performance of physics testing involving use of the part length CEA (PLCEA) groups. The physics testing will, in part, verify proper operation of the Core Protection Calculators following a refueling of the reactor core.

For Cycle 1, Technical Specification 3.1.3.7 did not affect the use of PLCEAs during physics testing. However, the proposed change to Technical Specifica-tion 3.1.3.7 for Cycle 2 will impose an insertion limit sLailar to that applied to the full length CEAs in Technical Specification 3.1.3.6.

In order to perform certain startup tests such as the verification of radial peaking factors at high power levels, it is necessary to insert the PLCEAs and CEAs beyond the limits specified in the proposed change to Technical Specification 3.1.3.7 and in Technical Specification 3.1.3.6. Currently, Technical Specification 3.10.2 allows suspension of the insertion limits for full length CEAs specified in Technical Specification 3.1.3.6. Because the proposed change to Technical Specification 3.1.3.7 will impose similar limits on the insertion of PLCEAs, it will be necessary to suspend these limits to perform these tests. Adherence to the restrictions in the limiting condition for operation and to the surveillance requirements assures that the operating safety limits are maintained.

Safety Analysis i

The proposed changes described above shall be deemed to involve a significant !

hazards consideration if there is a positive finding in any of the following ,

areas: l

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 i

The limits imposed on PLCEA position, which are more restrictive than those currently allowed, are used as input to the Cycle 2 safety analyses. Because the tests which take advantage of this special test exception are short in duration, core parameters related to the safety analyses are not adversely affected. There-fore, this change does not significantly increase the probability or consequences of any accident previously evaluated.

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

Response: No The limits imposed on PLCEA position, which are more restrictive than those currently allowed, are used as input to the Cycle 2 safety analyses. Because the tests which take advantage of this special test exception are short in duration, core parameters related to the safety analyses are not adversely affected.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Vill operation of this facility in accordance with this proposed change involve a significant reduction in margia of safety?

Response: No The limits imposed on PLCEA position, which are more restrictive than those currently allowed, will be used as input to the Cycle 2 safety analyses. Because the tests which take advantage of this special test exception are short in duration, core parameters related to the safety analyses are not adversely affected. There-fore, this change does not involve a significant reduction in margin of safety.

Safety and Significant liazards 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 a 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 signi-ficantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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SPECIAL TEST EXCEPTIONS 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, power distribution 3.2.2, 3.2.3, 3.2.7,limits of Specificationsand the Minimum Channels OPERABLE require Functional Unit 15 of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, requirements 3.2.2, 3.2.3, of Specificationsand the Minimum Channels OPERABLE requirement of 3.2.7, Functional Unit 15 of Table 3.3-1 are suspended, either:

a. Reduce THERMAL POWER suf ficiently to satisfy the requirements of Specification 3.2.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table S.3-1 are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, requirements 3.2.3, 3.2.7, of Specificationsor the Minimum Channels OPERABLE requirement of Fun Unit 15 of Table 3.3-1 are suspended.

WATERFORD - UNIT 3 3/4 10-2

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' hI. 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA worth I is immediately available for reactivity control when tests are performed for CEAs worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 MTC, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual CEAs to be positioned outside of their normal group heights and insertion limits during the perfor-mance of such PHYSICS TESTS as those required to (1) measure CEA worth and (2) determine the reactor stability index and damping factor under xenon oscillation conditions.

/ 3/4.10.3 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 CENTER CEA MISALIGNMENT This special test exception permits the center CEA to be misaligned during PHYSICS TESTS required to determine the isothermal temperature coefficient and power coefficient.

3/4.10.5 NATURAL CIRCULATION TESTING This special test exception permits all reactor coolant pumps to be secured during natural circulation testing and operator training for periods in excess of the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowed by Specification 3.4.1.2.

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SPECIAL TEST EXCEPTIONS 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS l

LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and power dist-ibution limits of Specifications 3.1.1. 3, 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 3, l,3. 7, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of i PHYSICS TESTS provided:

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1. 3, 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, '2, /, 3,7j l 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended, either:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or

b. Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1. 3, 3.1. 3.1, 3.1.3.5, 3.1.3.6,y3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE l 3,l.%7" requirement of Functional Unit 15 of Table 3.3-1 are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of l Specification 3.2.1 by monitoring it continuously with the Incore Detector l Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, -

3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended.

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(%.- 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA worth is immediately available for reactivity control when tests are performed for CEAs worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 MTC, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LINITS This special test exception permits individual CEAs to be positioned outside of their normal group heights and insertion limits during the perfor-mance m

of such PHYSICS TESTS as those required to (1) measure CEA worth en+

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/ 3/4.10.3 REACTOR COOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.4 CENTER CEA MISALIGNMENT This special test exception permits the center CEA to be misaligned during PHYSICS TESTS required to determine the isothermal temperature coefficient and power coefficient.

3/4.10.5 NATURAL CIRCULATION TESTING This special test exception permits all reactor coolant pumps to be secured during natural circulation testing and operator training for periods in excess of the I hour allowed by Specification 3.4.1.2.

a WATERFORD - UNIT 3 B 3/4 10-1

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-28 This is a request to revise Technical Specification 3.3.3.6, Accident Monitoring Instrumentation, to include the Reactor Vessel Level Monitoring System.

Existing Technical Specifications See Attachment A.

Proposed Technical Specifications See Attachment B.

Description I

The proposed change revises Technical Specification 3.3.3.6, Accident Monitoring Instrumentation, to add the Heated Junction Thermocouple System (HJTCS) - Reactor Vessel Level Monitoring System (RVLMS) to Tables 3.3-10 and 4.3-7, and the Bases. This implements item II.F.2, " Instrumentation for Detection of Inadequate Core Cooling", as requested by NRC Generic i Letter No. 83-37, "NUREG-0737 Technical Specification", dated November 1, 1983.

Following the May 1979 accident at Three Mile Island Unit 2, many features were added to nuclear power plants to enhance the ability of the operator to manage accidents and transients. The RVLMS is one of these enhancements and serves to provide trending information to the plant operators relative to RCS inventory. The proposed change adds the RVLMS to the technical specifications, to reflect its incorporation into the plant.

The Combustion Engineering Owners Group (CE0G) has reviewed and endorsed a generic Technical Specification for the RVLMS as noted in their letter to Mr. Hugh Thompson dated February 19, 1985 (RWW-85-12). The proposed change accurately reflects the Waterford 3 implementation of the CE00 generic Technical Specification.

Approval of this change is requested by December 15, 1986 in order to support Cycle 2 operation.

Safety Analysis The proposed changes described above shall be deemed to involve a signifi-cant 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 an accident previously evaluated?

Responses No 1

The RVLMS is neither credited nor required in the mitigation of any previously evaluated accident. It is not relied upon for reactor trip or initiation of any plant safety systems. Therefore, the proposed change does not affect the probability or consequences of an 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 Although the RVLMS will be utilized in the emergency operating proce-dures for corroboration of selected indications no change to normal operating procedures is involved. Thus no new path or mode of operation is created which may lead to a new or different kind of accident. The proposed change is intended solely to enhance the ability of the operator with additional corroborative information.

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

Response: No The specific purpose of the proposed change is to enhance accident and transient monitoring capability and therefore to increase the margin of safety.

The Commission has provided guidance concerning the applicaton of standards for determining whether a significant hazards consideration exists by provid-ing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations. Example (ii) relates to a change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications; for example, a more stringent surveillance requirement.

In this case, the proposed change is similar to Example (ii) in that the change to Technical Specification 3.3.3.6 is an addition to the accident monitoring instrumentation required by the NRC post TMI-2 Action Plan.

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 a 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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INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10; either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

! c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHAhNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

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j y ACCIDENT MONITORING INSTRUMENTATION 3 o MINIMUM l 5 REQUIRED NUMBER OF CHANNELS i e i CHANNELS OPERA 8LE j c INSTRbMENT ;

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1. Containment Pressure 2

] 2 1 1 2. Reactor Coolant Outlet Temperature - THot (Wide Range)

Range) 2 1

3. Reactor Coolant Inlet Temperature - TCold (

Reactor Coolant Pressure - Wide Range 2 1 l 4. ,

2 1

! 5. Pressurizer Water Level Steam Generator Pressure 2/ steam generator 1/ steam generator

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Steam Generator Water Level - Narrow Range 2/ steam generator 1/ steam generator i

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Steam Generator Water Level - Wide Range 1/ steam generator ** 1/ steam generator **

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Refueling Water Storage Pool Water Level 2 1 Y 9.

a 1/ steam generator **

10. Emergency Feedwater Flow Rate 1/ steam generator **

Reactor Cooling System Saturation Margin Monitor 2 1 ;

11.

12. Safety Valve Position Indicator 1/ valve 1/ valve j

i 13. Containment Water Level (Narrow Range) 1*** 1*** i Containment Water Level (Wide Range) 2 1 14.

Core Exit Thermocouples 4/ core quadrant 2/ core quadrant

! 15. '

4 1/ valve # :

16. Containment Isolation Valve Position Indicators

1/ valve #

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Condensate Storage Pool Level 2 1 .

j 17.

If the containment isolation valve is declared inoperable and the provisions of Specification 3.6.3 are complied with, position indicators may be inoperable; otherwise, comply with the provisions of i

Specification 3.3.3.6. !

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Containment isolation valves listed in Table 3.6-2 (Category 1).

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- - 0peration may continue for up to 30 days with less than the Minimum Channels OPERA 8LE requirement.

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M ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

o x CHANNEL CHANNEL CHECK CALIBRATION INSTRUMENT E M R Z 1. Containment Pressure R

M

" 2. Reactor Coolant Outlet Temperature - THot (Wide Range)

M R

3. Reactor Coolant Inlet Temperature -TCold (Wide Range)

M R

4. Reactor Coolant Pressure - Wide Range M R

5. Pressurizer Water Level M R

6. Steam Generator Pressure M R

7. Steam Generator Water Level - Narrow Range M R y 8. Steam Generator Water Level - Wide Range M R w 9. Refueling Water Storage Pool Water Level M R

$ 10. Emergency Feedwater Flow Rate M R l

11. Reactor Coolant System Saturation Margin Monitor M R

12. Safety Valve Position Indicator M R

13. Containment Water Level (Narrow Range)

M R

14. Containment Water Level (Wide Range) l M R

15. Core Exit Thermocouples M R

16. Containment Isolation Valve Position M R

17. Condensate Storage Pool Level l

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! 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERA 8ILITY of the accident monitoring instrumentation ensures that ;

i sufficient information is available on selected plant parameters to monitor '

and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for ,

i Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and

} Following an Accident," December 1980 and NUREG-0578, "TMI-2 Lessons Learned .

Task Force Status Report and Short-Tern Recommendations." Table 3.3-10 includes :

Regulatory Guide 1.97 Category I key variables. The remaining Category I

! variables are included in their respective specifications. !

3/4.3.3.7 CHEMICAL DETECTION SYSTEMS 1 i l The OPERA 8ILITY of the chemical detection systems ensures that sufficient !'

i capability is available to promptly detect and initiate protective action in i the event of an accidental chemical release. This capability is required to protect control room personnel and is consistent with the recommendations of

Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room j Operators Against an Accidental Chlorine Release," February 1975 and the l l recommendations of Regulatory Guide 1.78, " Assumptions for Evaluating the ,

l Habitability of a Nuclear Power Plant Control Room During a Postulated j Hazardous Chemical Release," June 1974.

i 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION

OPERA 81LITY of the fire detection instrumentation ensures that adequate i warning capability is available for the prompt detection of fires. This ;

i capability is required in order to detect and locate fires in their early I stages. Prompt detection of fires will reduce the potential for damage to '

i safety-related equipment and is an integral element in the overall facility l fire protection program. l

{ In the event that a portion of the fire detection instrumentation is j inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the' inoperable j instrumentation is restored to OPERA 81LITY. !

j 3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION i 1

4 j The OPERASILITY of the loose part detection instrumentation ensures that 4

sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The

. allowable out-of-service times and Surveillance Requirements are consistent j with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection i Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

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i INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION I LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-10, :!ther

-reeter th: in;r:ti: :hr xi t: OPC"?.0 L ;tetus withia 7 d;ys, co-L ,,, ll0! :llUT00= ithi , th; .,ut 12 h;r;. FME THc /3CT/OA>

IDthparlFIc1b IN 'Th Est.c 3 3-10,

b. With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10; either

. e s te, s .... ,,, ;bi; ;hr=1( ) to OPE"f,0LE ;tatus withir, 40 hevre vi uw m ou lmost ll0! SCUIDOW withir, the r.emt 12 hes, v. T'A" NE THE

/Strlou IDSMi*WI E b JH TA 6LE 3* 5 ~/Q.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

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TABLE 3.3-11 ACCICENT MONITORING INSTRUMENTATION il o

E REQUIRED MINIMUM NUMBER of CHANNELS g INSTRUMENT CHANNELS OPERA 8tE

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1. Containment Pressure 2 1 7f 30 f

2. Reactor Coolant Outlet Temperature - THot (W Rang *) 2 1 29,50

3. Reactor Coolant Inlet Temperature - T Cold I

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4. Reactor Coolant Pressure - Wide Range 2 1 2.% YO gg

5. Pressurtzer Water Level 2 1

6. Steam Generator Pressure g

2/ steam generator 1/ steam generator 7 Steam Generater Water Level - Narrow Range gg 3o

, 2/ steam generator 1/ steam generator gc 3c) 1 8. Steam Generator Water Level - Wide Range 1/ steam generator ** 1/ steam generator **

9. Refueling Water Storage Pool Water Level y g yo

{

2 1 39 FO f

10. Emergency Feedn.ater Flow Rate 1/ steam generator ** 1/ steam generator **

Reactor Cooling Systee Saturation Margin Monitor 4 3g II. 2 1

12. Safety valve Pesition Indicator 1/ valve 1/ valve

13. Containment Water Level (Narrow Range) 1*** 1*** I

14. Containment water Lemi (Wide Range) 2 1 2 CQ 30

15. Core Eaft Thermocouples 4/ core quadrant 2/ core quadrant gg

16. Containment Isolation Valve Position Indicators

1/ valved 1/ valve # g 3a 17 Condensate Storage Pool Level 2 1 y g

If the containment isolation valve is declared inoperable and the provisions of Specification 3.6.3 are complied with, position indicators may be inoperable; otherwise, comply with the provisions of Spectfication 3.3.3.6.

" Containment isolation valves Ifsted in Table 3.6-2 (Category 1).

- These corresponding instruments may be substituted for each e*her.

- - 0peration may continue for up to 30 days with less than the Minimum Channels OPERA 8LE requirement, ff M A 6m uusz.In En n r Sewscas la a Paose . A en n'wa ls OPcamn E If f~ewa. On Notet Susogg Ont on JL4o4Lg_ k DE. OPPE a / o+se E.G ann S esae.08 kcs112 W f"y w ea fivs,A*cOPt=RABLl=.

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TABLE 3.3-10 ACTION STATEMENTS ACTION 29 - With the number of OPERABLE accident monitoring channelsiless than the Required Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . --

ACTION 30 - With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10; either restore the inoperable channel (s) to OPERABLE ~

status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 31 - With the number of OPERABLE accident monitoring channels, less than the Required Number of Channels, either restore the system to OPERABLE status within 7 days if repairs are feasible

without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause ,

of the inoperability and the plans and schedule for restoring -

the system to OPERABLE status. g ACTION 32 - With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE in Table 3.3-10, either .

restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1. Initiate an alternate method of monitoring the reactor vessel inventory;

2. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

3. Restore the system to OPERABLE status at the next scheduled refueling, s

WATERFORD - UNIT 3 3/4 3-45a ;

i

)

o

TABLE 4.3-7 2

y ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS B

=

CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION E

Z 1. Containment Pressure M R

2. Reactor Coolant Outlet Temperature - THot (Wide Range) ., M R

3. Reactor Coolant Inlet Temperature -TCold (Wide Range) M R

4. Reactor Coolant Pressure - Wide Range M R

5. Pressurizer Water Level M R

6. Steam Generator Pressure M R

7. Steam Generator Water Level - Narrow Range M R y 8. Steam Generator Water Level - Wide Range M R w 9. Refueling Water Storage Pool Water Level M R

$ 10. Emergency Feedwater Flow Rate M R

11. Reactor Coolant System Saturation Margin Monitor M R

12. Safety Valve Position Indicator M R

13. Containment Water Level (Narrow Range) M R

14. Containment Water Level (Wide Range) M R

15. Core Exit Thermocouples M R

16. Containment Isolation Valve Position M R

17. Condensate Storage Pool Level M R

18. BL AscoR- V5 SS5L I#V#' NomTb12 % Ssren M R l

)

v'

l INSTRUMENTATION t

BASES 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent :

with the recommendations of Regulatory Guide 1.97, " Instrumentation for !

p Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and C,

Following an Accident," December 1980 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations." Table 3.3-10 includes N Regulatory Guide 1.97 Category I key variables. The remaining Category I V1 variables are included in their respective specifications.

N 4 I 3/4.3.3.7 CHEMICAL DETECTION SYSTEMS The OPERABILITY of the chemical detection systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chemical release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," February 1975 and the ,

recommendaticns of Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," June 1974.

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection prcgram.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.9 LOOSE-PART DETECTION INSTRUMENTATION i

The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect 1cose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and Surveillance Requirements are consistent with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981.

WATERFORD - UNIT 3 B 3/4 3-3

INSERT The Subcooled Margin Monitor (SMM), the Heated Junction Thermocouple (HJTC),

and the Core Exit Thermocouples (CET) comprise the Inadequate Core Cooling (ICC) instrumentation required by Item II.F.2 NUREG-0737, the Post TMI-2 Action Plan. The function of the ICC instrumentation is to enhance the ability of the plant operator to diagnose the approach to existence of, and recovery from ICC. Additionally, they aid in tracking reactor coolant inventory. These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37. These are not required by the accident analysis. nor to bring the plant to Cold Shutdown.

In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outage.

This is because the sensors are accessible only after the missile shield and reactor vessel head are removed. It is not feasible to repair a channel except during a refueling outage when the missile shield and reactor vessel head are removed to refuel the core. If only one channel ic inoperable, it should be restored to OPERABLE status in a refueling outage as soon as reasonably possible. If both channels are inoperable, at least one channel shall be restored to OPERABLE status in the nearest refueling outage.

L_ _

ML20203A976

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