ML20077N207

Proposed Tech Specs Re Deletion of Certain Instruments Not Classified as Category 1

ML20077N207
Person / Time
Site:	Nine Mile Point
Issue date:	01/06/1995
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML20077N194	List: ML20077N189 ML20077N207
References
NUDOCS 9501170102
Download: ML20077N207 (27)

Text

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ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 DOCKET NO. 50-410 Proposed Channes to the Technical Snecifications Replace existing pages 3/4 3-81,82,83,84,85,86,87,3/4 4-10 and 11 with the attached revised pages. These pages have been retyped in their entirety with marginal ;

markings to indicate changes to the text.

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9501170102 950106 PDR ADOCK 05000410 P PDR ,

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,, INSTRUMENTATION MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE.

1 APPLICABILITY: Operational Conditions 1 and 2. l ,

ACTION:

a. With one or more accident monitoring instrumentation channels inoperable, take the l ACTION required by Table 3.3.7.5-1.

b. The provisions of Specification 3.0.4 are not applicable. l SURVEILLANCE REQUIREMENTS - 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.51.  ;

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i NINE MILE POINT - UNIT 2 3/4 3-81 Amendment No. ,

TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION

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REQUIRED

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o NUMBER OF z INSTRUMENT QiANNELS ACTION h 1. Reactor Vessel Pressure 2 80 z

q 2. Reactor Vessel Water Level

a. Fuel Zone 2 80

b. Wide Range 2 80

3. Suppression Pool Water Level w

A a. Narrow Range 2 80 y b. Wide Range 2 80

4. Suppression Pool Water Temperature 2/Ouadrant 80

5. Suppression Chamber Pressure 2 80

6. Suppression Chamber Air Temperature 2 80

7. Drywell Pressure

a. Narrow Range 2 80

b. Wide Range 2 80 y 8. Drywell Air Temperature 2 80 3

g 9. Drywell Oxygen Concentration 2 80

$ 10. Drywell Hydrogen Concentration Analyzer and Monitor 2 80

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TABLE 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION sr-REQUIRED

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o NUMBER OF z INSTRUMENT CHANNELS ACTION

$ 11. Drywell High Range Radiation Monitors 2 81 z

q 12. Penetration Flow Path Primary Containment isolation Valve Position 2** 80**

9 Indication

• ta A

(J oo ta 3

E a

B G

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z p

• Not required for isolation valves whose associated flow path is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

• • Only one position indication channelis required for penetration flow paths with only one control room indication channel. If this one channel becomes inoperable, entry into ACTION 80b is required.

. .. . _- - . . -= - - - - - - - . _ - .

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. Table 3.3.7.5-1 (Centinued) e ACCIDENT MONITORING INSTRUMENTATION  ;

I ACTION ACTION 80 - a. With the number of OPERABLE accident monitoring instrumentation channels ,

for one or more functions one less then the Required Number of Channels -

shown in Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE  !

status within 30 days or prepare and submit a Special Report to the ,

Commission pursuant to Specification 6.9.2 within the following 14 days i outlining the preplanned alternate method of monitoring, the cause of the .:

inoperability, and the plans and schedule for restoring the instrumentation -l channel (s) of the function (s) to OPERABLE status.

b. With the number of OPERABLE accident monitoring instrumentation channels for one or more functions two less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12  ;

hours.

l ACTION 81 - a. With the number of OPERABLE Drywell High Range Radiation Monitors one l less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel to OPERABLE status within 30 days or ,

prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans l and schedule for restoring the instrumentation channel of the function to OPERABLE status, i b.' With the number of OPERABLE Drywell High Range Radiation Monitors two less than the Required Number of Channels shown in Table 3.3.7.5-1, f restore the inoperable channel (s) to OPERABLE status within 7 days or prepara and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans i and schedule for restoring the instrumentation channel (s) of the function to >

OPERABLE status.

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NINE MILE POINT - UNIT 2 3/4 3-84 Amendment No. ,

i This Page Not Used 1

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l NINE MILE POINT - UNIT 2 3/4 3-85 Amendment No.

TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

=

i o CHANNEL CHANNEL

{ INSTRUMENT CHECK CAUBRATION h 1. Reactor Vessel Pressure M R z

q 2. Reactor Vessel Water Level m a. Fuel Zone M R

b. Wide Range M R

3. Suppression Pool Water Level ,

w a. Narrow Range M R 3: b. Wide Range M R w

g 4. Suppression Pool Water Temperature M R*

5. Suppression Chamber Pressure M R

6. Suppression Chamber Air Temperature M R*

7. Drywell Pressure

a. Narrow Range M R

b. Wide Range M R

8. Drywell Air Temperature M R*

y 9. Drywell Oxygen Concentration M R

10. Drywell Hydrogen Concentration Analyzer and Monitor M O**

$ 11. Drywell High Range Radiation Monitors M Rt k 12. Penetration Flow Path Primary Containment isolation Valve Mtt R***

y Position Indication l

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., . TABLE 4.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUtREMENTS TABLE NOTATIONS

• Excludes sensors; sensor comparison shall be done in lieu of sensor calibration.

• • Using sample gas containing:

a. One volume percent hydrogen, balance nitrogen.

b. Four volume percent hydrogen, balance nitrogen.

• • • The CHANNEL CAllBRATION shall consist of position indication verification using ASME Section XI IWV-3300 test criteria.

t The CHANNEL CAllBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

tt Red, Green or other indication shall be verified as indicating valve position.

NINE MILE POINT - UNIT 2 3/4 3-87 Amendment No.

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., REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES

• LIMITING CONDITIONS FOR OPERATION 3.4.2 The safety valve function of at least 16 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings *: l 2 safety / relief valves @ 1148 psig i1%

4 safety / relief valves @ 1175 psig i1%

4 safety / relief valves @ 1185 psig 11%

4 safety / relief valves @ 1195 psig i1% '

4 safety / relief valves @ 1205 psig i1%  !

APPLICABILITY: OPERATIONAL CONDITIONS 1,2, and 3.

ACTION:

a. .With the safety valve function of one or more of the above required 16 safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. i

b. With one or more safety / relief valves stuck open, provided that the average water '

temperature in the suppression pool is less than 110'F, close the stuck open safety / relief valve (s); if unable to close the open valve (s) within 5 minutes or if the average water temperature in the suppression pool is 110*F or more, place the reactor mode switch in the Shutdown position.

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• The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

I NINE MILE POINT - UNIT 2 3/4 4-10 Amendment No.

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.. [ REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.2 No requirements other than Specification 4.0.5.

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NINE MILE POINT - UNIT 2 3/4 4-11 Amendment No. //

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4 ATTACHMENT B -

NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 DOCKET NO. 50-410 Sunoorting information and No Significant Hazards Consideration Analysis BACKGROUND .

The operability requirements for post-accident monitoring (PAM) instrume' .! ion at Nine Mile Point Unit 2 (NMP2) are provided in Section 3/4.3.7.5 of the Techrucal Specifications,

" Accident Monitoring Instrumentation." A review of these requirements against current regulatory guidance indicates that the NMP2 Technical Specifications are more restrictive than necessary. Therefore, Niagara Mohawk proposes changes to the NMP2 Technical .l Specifications to eliminata unnecessary testing and prevent unwarranted plant shutdowns r when post-accident monitoring instrumentation channels are inoperable. .

The Nuclear Regulatory Commission Final Policy Statement on the improved Technical Specifications (ITS), issued in July of 1993, states that the purpose of the Technical j Specifications is to impose those conditions or limitations on reactor operations necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features which are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval. To achieve this purpose, the Final 6 Policy Statement provides four criteria that delineate those constraints on design and [

operation of a nuclear power plant that must be retained in the Technical Specifications:

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary (Criterion 1).

A process variable, design feature, or operating restriction that is an initial  ;

condition of a Design Basis Accident (DBA) or Transient analysis that either ,

assumes the failure of or presents a challenge to the integrity of a fission product barrier (Criterion 2).

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either j assumes the failure of or presents a challenge to the integrity of a fission  ;

product barrier (Criterion 3).

A structure, system or component which operating experience or probabil-  ;

istic safety assessment has shown to be significant to public health and l safety (Criterion 4).

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A review of Regulatory Guide (RG) 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,"

indicates that not all post-accident monitoring instrumentation in Technical Specification Table 3.3.7.5-1 meets the screening criteria specified in the Final Policy Statement. This conclusion is consistent with NUREG-1433, " Improved Standard Technical Specifications,"

which lists only RG 1.97 Category 1 variables (Type A and non-Type A).

Per RG 1.97, Type A variables provide primary information needed to permit the operators to take specified manually controlled actions for which no automatic control is provided, but which are required for safety systems to accomplish their safety functions for design basis events. Type A variables allow the operators to perform the diagnosis specified in the Emergency Operating Procedures for preplanned actions in the primary success path of the DBAs (e.g., LOCA) and enable operators to take the specified, preplanned, manually controlled actions required for safety systems to accomplish their safety function and for which no automatic action is provided. Therefore, as stated in the Bases of Specification 3.3.3.1 in the ITS, Type A variables meet Criterion 3 of the NRC policy and are included in the ITS.

In RG 1.97, Category 1 designates key variables which have the most stringent design requirements. Category 1 applies to all Type A variables and may apply to other variables as well. Category 2 generally applies to instrumentation which indicates system operating status and indicates less stringent design requirements. The ITS Bases state that non-Type A, Category 1 variables meet Criterion 4 of the NRC policy statement regarding risk in that they are the single variables that most directly indicate the accomplishment of a safety function. Consequently, the ITS also includes all Category 1 variables.

Type B variables provide information to indicate whether plant safety functions are being accomplished for reactivity control, core cooling, Reactor Coolant System integrity, and containment integrity. Type C variables provide information to indicate the potential for or the actual breaching of the barriers to fission product releases. Type D variables are those that provide information to indicate the operation of individual safety systems or other systems important to safety. Type D variables help the operators make appropriate decisions in using the individual systems important to safety in mitigating the consequences of an accident. Type E variables are required for use in determining the I magnitude of the release of radioactive materials and for continually assessing such releases. Thus, non-Category 1 Type B, C, D and E variables do not meet the screening criteria and are not included in the ITS.

RG 1.97 identifies neutron flux as a Type B, Category 1 variable that provides information to indicate whether plant safety functions are being accomplished. Because of the difficulty in meeting Category 1 design criteria, several licensees requested deviations from the Regulatory Guide position for Category 1 neutron flux monitoring instrumentation. In support of these requests, the BWR Owners' Group submitted NEDO-31558-A, " Position on NRC RG 1.97, Revision 3, Requirements for Post-Accident Neutron Monitoring System." The NEDO Report proposes criteria for neutron flux monitoring instrumentation in lieu of the Category 1 criteria in Regulatory Guide 1.97. By letter dated January 13, 1993, the Commission issued a Safety Evaluation of NEDO-31558-A. The Commission's i'

Safety Evaluation indicates that the BWR neutron flux monitoring instrumentation design criteria delineated in NEDO-31558-A are acceptable.

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• . l NMP2 Technical Specification Table 3.3.7.5-1, " Accident Monitoring Instrumentation,"

and Table 4.3.7.5-1, " Accident Monitoring instrumentation Surveillance Requirements,"

list all NMP2 Category 1 variables, two Type B, Category 3 variables (IRM, SRM), one l Type B variable that meets the requirements of NEDO-31558-A (APRM), and two Type D,  !

l Category 2 variables, (Suppression Chamber Air Temperature and Safety / Relief Valve (SRV) Position Indication). Consistent with the NRC Final Policy Statement, which  !

encourages licensees to use the screening criteria in plant-specific amendments, safety /

relief valve position indication, IRM, SRM, and APRM instrumentation are proposed for l removal. Suppression Chamber Air Temperature, although a Type D variable, is a process variable that is an initial condition of a Design Basis Accident (containment response analysis) and therefore does meet Criterion 2 of the screening criteria. in addition, the Residual Heat Removal (RHR) Heat Exchanger Service Water Radiation Monitors and the Refuel Platform Radiation Monitor are not classified as PAM instrumentation by RG 1.97  :

and are also proposed for removal. Niagara Mohawk also proposes to incorporate other  !

provisions of the ITS which provide operational flexibility by permitting continued operation  !

with one inoperable instrument channel, deleting the requirement that certain accident monitoring instrumentation be operable in Operational Condition 3 and providing an exception from the provisions of Specification 3.0.4. Finally, one of the SRV position indication systems, acoustic monitors, is also addressed in Specification 3/4.4.2,

" Safety / Relief Valves." The Limiting Condition for Operation (LCO), ACTIONS and Surveillance Requirements of Specification 3/4.4.2 are propossd for revision to remove requirements related to the safety / relief valve acoustic monitors, consistent with the changes proposed to Tables 3.3.7.5-1 and 4.3.7.5-1 and the ITS. i l

DESCRIPTION OF PROPOSED CHANGES The following changes are proposed:

1. Delete the Operability and Surveillance Requirements for the following functions I from Table 3.3.7.5-1, " Accident Monitoring instrumentation," and Table 4.3.7.5-1,

" Accident Monitoring Instrumentation Surveillance Requirements":

Safety / Relief Valve Position Indicators (item No.11)

RHR Heat Exchanger Service Water Radiation Monitor (item No.13)

Refuel Platform Area Radiation Monitor (Item No.14)

Neutron Flux (Item No.15)

2. Delete Operational Condition 3, Hot Shutdown, from the Reactor Vessel Water l Level, Suppression Pool Water Level, and Drywell High Range Radiation Monitor  ;

instrumentation located in the " Applicable Operational Conditions" column of Tables 3.3.7.5-1 and 4.3.7.5-1. Accordingly, revise LCO 3.3.7.5, Applicability, to indicate that accident monitoring instrumentation is required to be operable in Operational Conditions 1 and 2.

3. Revise the ACTIONS of Table 3.3.7.5-1 to allow 30 days to restore one inoperable channel and 7 days to restore two inoperable channels. With one inoperable channel, the ACTIONS will require the filing of a Special Report after 30 days and, with two inoperable channels, the ACTIONS will require a plant shutdown after 7 days (except for the Drywell High Range Radiation Monitors where no shutdown is required). Also, the ACTIONS for Suppression Pool Water Level will be made Page 3 of 17

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consistent with remaining accident monitoring instrumentation (except Drywell High Range Radiation Monitors) to reflect the deletion of the requirement to be operable 1 in Operational Condition 3.  !

4. Add an exception from Specification 3.0.4 to the required ACTIONS of LCO 3.3.7.5.

5. Revise item 16 in Tables 3.3.7.5-1 and 4.3.7.5-1 from " Primary Containment isolation Valve Position Indication" to " Penetration Flow Path Primary Containment isolation Valve Position Indication" and increase the " Required Number of Channels" for this function from one to two.

6. Delete the " Minimum Channels Operable" column from Table 3.3.7.5-1 and the

" Applicable Operational Conditions" column from Tables 3.3.7.5-1 and 4.3.7.5-1.

7. Remove reference to acoustic monitors from LCO 3.4.2 and delete associated required ACTION "c" and Surveillance Requirement 4.4.2.1 and associated notes.

8. Add Surveillance Requirement 4.4.2 to Specification 3/4.4.2 to reference the SRV testing performed under the Inservice inspection Program (Specification 4.0.5).

EVALUATION i Deletion of Non-Cateaorv 1 Parameters Table 7.5 2 of the NMP2 Updated Safety Analysis Report (USAR), "Conformance to Regulatory Guide 1.97," provides a list of NMP2's RG 1.97 instrumentation along with their category and type. To evaluate their possible removal from the Technical Specifications, Niagara Mohawk has evaluated those PAM instruments which are not classified as Category 1 against the screening criteria of the Final Policy Statement.

Instrumentation classified as Category 1 (Type A or non-Type A) will be retained in the Technical Specifications.

Safety / Relief Valve Position Indication The SRV Position Indicators instrumentation listed in the NMP2 Technical Specifications consists of Type D, Category 2 instrumentation (acoustic monitors) and non-RG 1.97 instrumentation (tailpipe temperature). The SRVs themselves are a part of the primary success path in the USAR accident analysis in that they can actuate to mitigate a DBA and therefore meet Criterion 3 of the Final Policy Statement. Accordingly, their operability is required by Technical Specification 3/4.4.2, " Safety / Relief Valves." However, SRV position indication does not detect or indicate a significant abnormal degradation of the reactor coolant pressure boundary (Criterion 1). This is consistent with the Commission's Final Policy Statement which indicates that :the first criterion was intended to assure that

- Technical Specifications controlled those instruments specifically installed to detect reactor coolant leakage but not to include instrumentation to identify the source of actual leakege (e.g., valve position indication). Nor is it a process variable, design feature or operoting restriction that is an initial condition of a DBA or transient analyses (Criterion 2). While the function of SRVs is part of the primary success Page 4 of 17

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i path and the SRVs actuate to mitigate a DBA or transient, position indication for  !

the SRVs does not form a part of the primary success path since the USAR l accident analysis assumes the SRVs function as designed. That is, the USAR l analysis assumes no operator action based on SRV valve position for the SRVs to perform their primary success path function (Criterion 3). Finally, failure of SRV position indication would not pose a significant challenge to the ability of the operating crew to respond to a DBA or transient, since the Emergency Operating Procedures provide symptom-based instruction to the crew in mitigating an upset -

condition of the plant, (i.e., level, pressure, and temperature provide EOP direction l

regardless of SRV status). The loss of this instrumentation has no effect on  !

probabilistic safety assessment and has not been shown to be significant to public j health and safety (Criterion 4). Consequently, SRV position indication does not meet any of the screening criteria of the Final Policy Statement. NUREG-1433, which does not list SRV Position Indication in Specification 3.3.3.1, " Post-Accident ,

Monitoring Instrumentat;on," supports this conclusion. l i

Residual Heat Removal Heat Exchanger Service Water Radiation Monitor As discussed in USAR Section 9.2.1, each RHR heat exchanger at NMP2 has a radiation monitor in the service water retum line. RG 1.97 does not discuss or >

require these radiation monitors. Rather, the monitors were installed in response to item III.3.d of Section 9.2.1 of the Standard Review Plan:

" Provisions are made in the system to detect and control leakage of radioactive contamination into and out of the system. It will be acceptable if the system P&lDs show radiation monitors located on the system dischargc and at components susceptible to leakage, and these components can be isolated by one automatic and one manual valve in series." l Section 7.6.1.1 of the NRC SER reflects this design:

"An offline liquid monitor is located in the service water effluent on each of the two RHR heat exchangers. These monitors detect and alarm contamination of the service water effluent resulting from leaks in the heat exchangers following a l LOCA or under normal operating conditions "

USAR Section 11.5.1 designites these monitors as " Radiation Monitors Required  :

for Safety" and states the objective of these radiation monitoring systems is to initiate appropriate manual or automatic protective action to limit the potential release of radioactive materials from the reactor vessel, primary and secondary containment if radioactivity exceeds predetermined levels in major process effluent streams. USAR Section 11.5.2.1.2 states, "These monitors function to detect and alarm on contamination of the service water effluent due to leaks in the heat exchangers following a LOCA or under normal operating conditions." That is, the monitors alert plant personnel to potential plant releases due to leakage in the heat exchangers and allow plant personnel to terminate such releases. Thus, the

, monitors are not relied upon to detect reactor coolant pressure boundary l

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t degradation (Criterion 1). The radiation level in service water effluent is not a  !

process variable, design forture or operating restriction that is an initial condition  !

used in any analysis (Critenon 2). The monitors are not required by RG 1.97 and do not function or actuate to mitigate DBAs or transients (Criterion 3). In addition to '

the RHR Heat Exchanger Eervice Water Radiation Monitors, the service water system return headers contain radiation monitors to monitor potential releases to the environment. The RHR Heat Exchanger Service Water Radiation Monitors are ,

an enhancement to the NMP2 effluent monitoring system but are not required to 1 assure public health and safety (Criterion 4). Therefore, these monitors do not l meet any of the screening criteria and qualify for removal from the Technical '

Specifications. I Refuel Platform Area Radiation Monitor l

The Refuel Platform Area Radiation Monitor is an area radiation monitor as discussed in USAR Section 12.3.4. The monitor is required to alert plant personnel working on the refuel platform area to increasing or abnormally high radiation levels and is declared operable prior to any fuel movement. This monitor has no function '

relative to the safe shutdown of the plant or to the quantitative monitoring of releases of radioactive material to the environment. The radiation monitor does not detect any significant abnormal degradation of the reactor coolant pressure boundary (Criterion 1) nor is it a process variable, design feature or operating restriction that is an initial condition of any analyses (Criterion 2). The monitor is not part of the primary success path associated with any DBA or transient (Criterion 3) and has not been shown to be significant to public health and safety (Criterion 4). The monitor is installed for personnel safety and is not required by RG 1.97. Therefore,it does not meet any of the four screening criteria.

Neutron Flux (SRM/lRM/APRM)

The APRM instrumentation listed in the NMP2 accident monitoring Technical Specifications are Type B variables and conform to the design and qualification requirements of NEDO-31558-A in lieu of the requirements of Regulatory Guide 1.97, Category 1. Consequently, APRMs are not considered Category 1 variables.

As indicated in the Commission's Safety Evaluation of NEDO-31558-A, the l NEDO-31558-A Report analyzed event scenarios to determine the consequences of neutron flux monitoring unavailability and concluded that the failure of this instrumentation would not prevent the operator from determining reactor power levels. Altemate parameter status will be available from which power can be  !

inferred and used to base operational decisions. Further, NEDO-31558-A contains criteria regarding the range, power supplies, and qualifications for neutron flux monitoring instrumentation that provide sufficient confidence that 150 neutron flux monitoring instrumentation will be available to confirm reactor shutdown for a wide range of events including ATWS. Also, for BWR design' basis events, recriticality is not a significant contributor to core melt risk for BWR accident scenarios that go beyond the design basis. Based on the BWR Owners' Group submittals, the Director of NRR determined that Category 1 neutron flux monitoring instrumentation is not needed for existing BWRs to cope with LOCA, ATWS, or other accidents that do not result in r,evere core damage conditions.

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. . . j RG 1.97 indicato neutron monitoring is a Type B, Category 1 variable. However, the IRM and SRM dstrumentation listed in the NMP2 accident monitoring Technical Specifications are Typs B, Category 3 variables as indicated in USAR Table 7.5-2. 3 This is based on the fact that at NMP2 neutron flux level below the APRM range is  !

not a key variable for accomplishing mitigative actions for any design basis accident or transient (including those a'nticipated operational occurrences required to be )'

considered in the implementation of the ATWS Rule (10CFR50.62). Also, required operator actions specified in the plant Emergency Operating Procedures for such events can be accomplished without reliance on reactor power information below ,

the APRM range. On this basis, the designation of Category 3 (in lieu of Category l 1 instrumentation as recommended by Regulatory Guide 1.97) is appropriate for instrumentation monitoring intermediate range and source range neutron flux. In  :

addition, the EOPs provide guidance when reactivity is unknown in the unlikely  ;

event that al: monitoring is lost.

Although neutron flux instruments can be used to determine core reactivity, they do not detect or indicate a significant abnormal degradation of the reactor coolant l pressure boundary (Criterion 1) nor are they process variables, design features, or i operational restrictions that are an initial condition of a DBA or transient analysis  !

(Criterion 2). Although certain IRM and APRM channels provide inputs to the l Reactor Protection System, the accident monitoring functions of these instruments i do not form part of a primary success path which functions or actuates to mitigate '

a DBA or transient. The IRM and APRM LCOs and surveillance requirements of the Reactor Protection System will romain in the Technical Specifications (Criterion 3). ,

As indicated above, the loss of neutron flux monitoring would not prevent the i operator from determining reactor power levels and making appropriate operational l decisions. Therefore, loss of this instrumentation has no effect on probabilistic

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safety assessment and has not been shown to be significant to public health and  !

safety (Criterion 4). Consequently, neutron flux does not meet any of the screening i criteria of the Final Policy Statement and removal of non-Category 1 i instrumentation from the accident monitoring Technical Specification is consistent with the ITS.

The operability and surveillance requirements for the above instruments will be removed from Table 3.3.7.5-1 and Table 4.3.7.51 of Specification 3/4.3.7.5 and from the LCO, 1 ACTIONS, and Surveillance Requirements of Specification 3/4.4.2. These proposed changes will provide consistency between NMP2's RG 1.97 instrumentation, the ITS, and the NMP2 Technical Specifications and are consistent with the purpose of the Technical Specifications as stated in the Commission's Final Policy Statement. That is, the instruments being proposed for removal are not of controlling importance to safety or necessary to obviate the possibility of an abnormal situaticn or event giving rise to an immediate threat to public health and safety. All Type A and Category 1 instruments are being retained in the Technical Specifications and, as demonstrated above, the instruments proposed for removal do not meet any of the screening criteria contained in the Final Policy Statement. The Commission's Final Policy Statement recommends that when a licensee 4 elects to apply the "four criteria" to remove a section of the Technical Specifications that the associated requirements be relocated to another licensee controlled document.

Conceming existing Technical Specification surveillance requirements, Niagara Mohawk plans to continue to perform testing of the Safety / Relief Valve Position Indication, Residual J Heat Removal Heat Exchanger Service Water Radiation Monitor, Refuel Platform Radiation Page 7 of 17

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monitor and the neutron flux instrumentation to assure a high level of component reliab:lity. Testing will be performed in accordance with plant procedures as described in the Updated Final Safety Analysis Report. Also, the APRM and IRM Reactor Protection System instrumentation surveillance requirements will remain in Technical Specifications.

Concerning the existing Technical Specification ACTION requirements, Niagara Mohawk concludes that relocating these requirements associated with the Safety / Relief Valve Position Indication, Residual Heat Removal Heat Exchanger Service Water Radiation Monitor, Refuel Platform Radiation Monitor and neutron flux instrumentation is unnecessary based on the following:

1. The Residual Heat Removal Heat Exchanger Service Water Radiation Monitor and Refuel Platform Radiation Monitor instrumentation are not RG 1.97 instruments and accordingly, should not have had ACTION or Surveillance Requirements previously located in the accident monitoring section of Technical Specifications.

2. The APRM and IRM Reactor Protection System instrumentation ACTION requirements will remain in Technical Specifications.

3. The instruments proposed for removal from the Technical Specifications were evaluated against the Final Policy screening criteria used to identify parameters that should be consiCared for retention in Technical Specifications. This evaluation indicated these instruments did not meet any of the screening criteria and that their removal from Technical Specifications posed no danger to the health and safety of the public. This same logic, in part, provides the basis for Niagara Mohawk's determination that the subject ACTION requirements need not be relocated to anothar licensee controlled document.

4. ACTION requirements by their nature are not readily relocated to other licensee controlled documents. Also, placing the ACTION requirements in a specific testing procedure would not capture those occasions when a component was found to be inoperable outside of the required testing process. The Technical Specifications are where ACTION requirements should be located if they affect safety. As discussed above, there is no safety basis for keeping these items in the Technical Specifications. Therefore, relocating the associated ACTION requirements is not required.

5. Testing and maintenance will continue to be performed in accordance with approved procedures to assure high reliability of the items removed from the Technical Specifications.

6. Niagara Mohawk has a Deviation / Event Reporting (DER) system that is used to identify, document, evaluate, correct and trend conditions, events, activities and concerns that have the potential for affecting plant safety or reliability. Although the DER provides no specific direction in the event of a specific component becoming inoperable, the DER system provides a formalized process to identify such concerns and bring these concerns to the attention of plant operations and management. This assures that each deviation is evaluated by the proper personnel and that corrective actions are taken in a timely manner. This evaluation, when requested, would include an Operability Supporting Analysis to determine the Page 8 of 17

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• capability of the affected structure, system or component to perform their safety function. If a situation were to occur affecting instrumentation proposed for -

removal from the Technical Specifications, the DER system would provide flexibility over the existing Technical Specification ACTION requirements to address the situation without having an adverse affect on plant safety or reliability. j l

Deletion of ADD liCabl2 ODefational Condition 3 j NMP2 Technical Specification Tables 3.3.7.5-1, " Accident Monitoring Instrumentation,"

and 4.3.7.5-1, " Accident Monitoring Instrumentation Surveillance Requirements," indicate that the Reactor Water Level, Suppression Pool Water Level, and Drywell High Range Radiation Monitor instrumentation are required to be operable in Operational Conditions 1, 2 and 3. Niagara Mohawk proposes to delete the requirement to have these instruments operable in Operational Condition 3, HOT SHUTDOWN. Post-Accident monitoring  ;

variables are related to the diagnosis and preplanned actions required to mitigate Design  ;

Basis Accidents (DBAs). The applicable DBAs are assumed to occur in Operational '

Conditions 1 and 2. In Operational Condition 3 (as well as 4 and 5), plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely l low. Therefore, PAM instrumentation is not required to be operable in Operational Condition 3. This position is consistent with NUREG-1433, " improved Standard Technical l Specifications," which requires PAM instrumentation to be operable in Operational Conditions 1 and 2.

The existing ACTION s. lements for Suppression Pool Water Level, with the number of l operable channels less than the Minimum or Required would require placing the plant in i COLD SHUTDOWN. As indicated in the ITS, Bases Section 3.3.3.1, if any required i ACTION and associated completion times could not be met, the plant must be brought to an operational condition in which the LCO does not apply. Currently, because suppression ,

pool water level monitoring is required in Operational Conditions 1, 2 and 3, placing the  !

plant in Operational Condition 4, COLD SHUTDOWN, was appropriate if the required i actions could not be met. However, as noted above, Operational Condition 3 will be '

deleted as a required condition for suppression pool water level. Accordingly, the ACTION requirement associated with Suppression Pool Water Level will be revised such that the plant will only need to be brought to Operational Condition 3, HOT SHUTDOWN. This change will make the ACTION for Suppression Pool Water Level consistent with ACTION requirements of the remaining accident monitoring instrumentation (except for Drywell High Range Radiation), as discussed in the next section.

Revision to Reavired ACTIONS of Table 3.3.7.5-1 1

The current NMP2 Technical Specifications require the plant be placed in HOT or COLD I SHUTDOWN after 7 days if one required PAM instrumentation channelis inoperable and after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if two required PAM instrumentation channels are inoperable. Exceptions to this requirement are the various radiation monitors, where continued operation is allowed provided a Special Report is filed, and isolation valve position indication which refers to Specification 3.6.3, Primary Containment Isolation Valves. With the existing Technical l Specification, when an isolation valve position indication channel becomes inoperable, the associated isolation valve is considered inoperable and the ACTIONS of Specification 3.6.3 apply. Considering an isolation valve inoperable based solely on the inoperability of position indication is conservative and inconsistent with other plant licenses and the ITS. I Page 9 of 17 l

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The provisions of the iTS allow continued operation with one required PAM instrument  ;

channelinoperable provided a Special Report is filed if the channel is not restored within  :

30 days. The Special Report must provide a root cause evaluation and identify corrective  !

actions and alternate methods of monitoring the parameter. With two required PAM instrument channels inoperable, the ITS generally requires shutdown within 7 days, except for the hydrogen monitors (shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and the radiation monitors (no shutdown required).

l The proposed changes would incorporate the ITS provisions for inoperable PAM instrumentation channels into the required ACTIONS of Specification 3.3.7.5, except for the required ACTIONS for two inoperable hydrogen monitor channels. i The proposed 30 day restoration time for one inoperable channelis acceptable based on  ;

the availability of the remaining operable PAM channel, the passive nature of the PAM instrumentation (i.e., no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation. ,

Continued operation with an inoperable channel is acceptable since the remaining RG 1.97 channel is still available and a preplanned alternate method of monitoring is identified in the associated Special Report. The 7 day completion time for two inoperable PAM instrumentation channels is acceptable based on the relatively low probability of an event requiring PAM instrumentation, the passive natute of the PAM instrumentation, and the availability of alternate means to obtain the required information. Continued operation with two inoperable channels is not acceptable since the alternate indications may not [

meet all the qualification requirements of RG 1.97. Limiting the restoration time to 7 days  !

limits the risk that a PAM function will be in a degraded condition should an accident occur. These 30 and 7 day restoration times are consistent with the ITS and have been previously approved at Sequoyah Nuclear Plant, Units 1 and 2 (Amendment Nos.159 and 149, respectively). I As stated above, the ITS allows a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restoration time with two inoperable hydrogen monitor channels. However, Generic Letter (GL) 83-36, "NUREG-0737 Technical Specifications," previously recommended a 7 day restoration time for two inoperable hydrogen monitor channels. The Staff approved a 7 day restoration time for Clinton Power Station in Amendment No. 31, dated February 8,1990, which was based on GL 83-36.

The proposed changes include a 7 day restoration time for two inoperable hydrogen i monitoring instrument channels. The 7 day restoration time is necessary due to the difficulty in repairing inoperable hydrogen monitors and is acceptable for several reasons. i First, the probability of a Loss-of-Coolant-Accident which generates hydrogen in amounts exceeding the flammability limit is low. Second, the Post-Accident Sampling System can independently determine hydrogen concentration in the drywell. Finally, as discussed in USAR Section 6.2.5.2.'2, plant personnel would have at least two days to place the recombiners in service to prevent hydrogen concentration from exceeding its flammability limit, which is more than enough time to obtain hydrogen concentration readings with the Post-Accident Sampling System.

Exbeation to Soecification 3.0.4 The proposed changes add an exception from the provisions of Specification 3.0.4 to LCO 3.3.7.5. This exception is consistent with the ITS and will permit entry into the applicable conditions of LCO 3.3.7.5 (Operational Conditions 1 and 2) with inoperable .

Page 10 of 17

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instrumentation channels. GL 87-09, " Sections 3.0 ar.d 4.0 of the Standard Technical Specifications on the Applicability of Limiting Conditions for Operations and Surveillance Requirements," proposed a generic exemption from Specification 3.0.4 and discussed the conditions under which it applies. The GL states that Specification 3.0.4 unduly rastricts facility operation when conformance to the required ACTIONS provides an acceptable level of safety for continued operation. Further, the GL states that LCOs with required ACTIONS which permit continued operation for an unlimited time should permit entry into their applicable operational conditions. In the proposed required ACTIONS of LCO 3.3.7.5, continued operation with one inoperable channel is permitted and an acceptable level of safety is assured since the remaining RG 1.97 channel is still available, the PAM instruments have no active safety function, and a preplanned alternate method of monitoring is identified in the required Special Report. Therefore, a 3.0.4 exception for single inoperable PAM channels is consistent with GL 87-09 and the ITS.

The proposed Specification 3.0.4 exception would also apply to two inoperable PAM instrument channels. As stated in GL 87-09, the intent of Specification 3.0.4 is to ensure that a higher mode of operation is not entered when equipment is inoperable or when parameters exceed their specified limits. This precludes a plant startup when actions are being taken to satisfy an LCO which, if not completed within the specifiec8 time, would result in a plant shutdown to comply with the required ACTIONS. However, for PAM instrumentation, the proposed 7 day restoration time provides adequate time to restore inoperable channels without placing undue pressure on plant personnel. Consequently, a plant shutdown due to inoperable PAM instrumentation would not be anticipated. While this proposed change goes beyond the scope of GL 87-09, the passive nature of the PAM instrumentation, the ability to diagnose an accident using alternative instruments and methods, and the low probability of an event requiring PAM instrumentation assures that an acceptable level of plant safety is maintained during the 7 day completion time, thus meeting the intent of GL 87-09. Several nuclear facilities, including V. C. Summer Nuclear Station, and Sequoyah Nuclear Plant, Units 1 and 2, have a Specification 3.0.4 exception for PAM instrumentation. The 3.0.4 exception for two inoperable PAM instrument channels is consistent with the ITS.

Prmarv Containment isolation Valve Position Indication in the iTS, the number of required channels for Primary Containment isolation Valve (PCIV)

Position Indication (Line item 8 in ITS Table 3.3.3.1-1) is "1 per valve." While this format accurately reflects the requirements of RG 1.97 and provides proper direction for loss of indication on one PCIV, i.e., restore within 30 days, it does not adequately address the actions necessary when both PCIVs on a penetration lose indication and is not consistent with the ITS ACTION requirements and Bases. The ITS, as written, would allow continued operation with two inoperable PCIV Position Indication channels on a penetration. The appropriate action requirement for this case should be to restore at least one channel within 7 days or, as a minimum, isolate the penetration.

To assure correct implementation of the required ACTIONS, the proposed changes revise

. the description of PCIV Position Indication from " Primary Containment isolation Valve Position Indication" to " Penetration Flow Path Primary Containment isolation Valve Position Indication," and revise the " Required Number of Channels" from 1 channel per PCIV to 2 channels por penetration. This accurately reflects the design of most penetrations, i.e., two PCIVs per penetration and one channel per PCIV. A footnote is Page 11 of 17

s o added to clarify that only one instrument channelis required for penetrations with only one PCIV and that the more restrictive ACTION requirements must be taken upon loss of that one channel. This is consistent with the ITS Bases which states that one position indicator is required for each active PCIV. The ITS also adds an additional footnote which states that indication is not required for PCIVs whose associated flow path is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. Since the PCIVs on a penetration isolated by a passive device would no longer be called upon to change state (i.e., they would not be considered active), this is also consistent with the ITS Bases.

Delete " Minimum Ooerable Channels" and "Acolicable Ooerational Conditions ~ Column l In the Final Policy Statement, the Commission stated its intent that, to the extent practicable, license amendment applications use the wording and Bases of the ITS. I Currently, in Table 3.3.7.5-1 the minimum operability tequirements for PAM instrumen- l tation are provided in two columns, " Required Number of Channels" and " Minimum j Channels Operable." The current required ACTIONS are based on whether the number of operable channels is less than the " Required Number of Channels" or less than the

" Minimum Channels Operable." The ITS consolidates those same operability requirements into one " Required Number of Channels" column. Consistent with the Final Policy Statement and the proposed ACTIONS for LCO 3.3.7.5, the changes proposed to Table 3.3.7.5-1 contain one column for the operability requirements. The proposed ACTIONS are based on whether the number of operable channels is one less or two less than the

" Required Number of Channels" specified in the table. Also, because of this change, the ACTION requirements cannot be properly applied to suppression pool water temperature instrumentation where the " Required Number of Channels" is 8,2/Ouadrant. To be consistent with the new ACTION requirements, the number "8" will be deleted. This is an administrative change which does not change any operability requirements.

Also, Tables 3.3.7.5-1 and 4.3.7.5-1 contain a column entitled " Applicable Operational Conditions." As indicated in these tables, certain accident monitoring instrumentation is required to be operable in Operational Conditions 1 and 2 and others in Operational Conditions 1,2 and 3. This is consistent with existing LCO 3.3.7.5 which indicates that the accident monitoring specification is applicable as shown in Table 3.3.7.5-1. As discussed above, the requirement to have PAM instrumentation operable in Operational Condition 3 has been deleted. Accordingly, all PAM instrumentation will be required only in Operational Conditions 1 and 2. In order to make LCO 3.3.7.5 consistent with other NMP2 specifications in which all relevant instrumentation is required to be operable in the same operational conditions, the Applicability section of LCO 3.3.7.5 will be revised to read " Operational Conditions 1 and 2." Subsequently, the " Applicable Operational Conditions" column will be deleted as it is no longer required.

Inservice Testino of Safetv/ Relief Valves Surveillance Requirement 4.4.2.1, which specifies the testing requirements for the acoustic monitors, is being removed consistent with the removal of the acoustic monitor operability requirements. A new surveillance requirement will be added (Surveillance Requirement 4.4.2) which will clarify that all required surveillances for the safety function of the SRVs are accomplished under the Section XI Inservice Testing Program, as required by Surveillance Requirement 4.0.5. During each refueling outage, the Code required Page 12 of 17

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• l number of installed valves are removed for verification of set pressure and are stroke {

tested using the pneumatic power actuator. Code required testing is documented in the ,

inservice Testing Program Plan. Therefore, adequate testing of the safety function of the l SRVs is provided. I CONCLUSION  ;

The proposed changes would delete those instruments not classified as Category 1 or '

Type A, as defined by RG 1.97, and revise the ACTIONS of LCO 3.3.7.5 to permit continued operation with one inoperable channel and 7 days with two inoperable channels.

With one inoperable channel, the ACTIONS will require the filing of a Special Report after 30 days and, with two inoperable channels, the ACTIONS will require a plant shutdown after 7 days (except for the Drywell High Range Radiation Monitors where no shutdown is required). In addition, an exception to Specification 3.0.4 is being added to LCO 3.3.7.5 and changes are proposed to Tables 3.3.7.5-1 and 4.3.7.5-1 for consistency and clarity.

Removal of non-Type A and non-Category 1 PAM instruments from the Technical Specifications is consistent with the Final Policy Statement on the Improved Technical Specifications. The instruments proposed for removal have been evaluated against the four criteria contained in the Final Policy Statement and qualify for removal under the i guidelines of the Final Policy Statement. SRV position indication is not necessary 1 information for the completion of the primary success path for accident mitigation as  !

outlined in the USAR. The function of SRVs is part of the primary success path and they i actuate to mitigate a DBA or transient. However, position indication for the SRVs is not part of the primary success path since the USAR accident analysis assumes the SRVs function as designed. That is, no operator action based on SRV valve position is assumed in the USAR analysis. The RHR Radiation Monitors serve only to monitor a potential l effluent pathway by alarming upon contamination of the service water effluent resulting from leaks in the heat exchangers. The Refuel Platform Area Radiation Monitor serves to alert plant personnel working on the refuel platform to increasing or abnormally high radiation levels and is installed for personnel safety. NMP2 neutron flux monitoring instrumentation are not Category 1 instruments and, therefore, their removal is consistent  :

with the ITS.

Continued operation with an inoperable channel is acceptable based on the availability of the remaining operable PAM channel, the passive nature of the PAM instrumentation (i.e.,

no critical automatic action is assumed to occur from these instruments), the low probability of an event requiring PAM instrumentation, and the fact that a preplanned alternate method of monitoring is identified in the Special Report required after the first i 30 days of inoperability. The 7 day completion time for two inoperable PAM instrumentation channels is acceptable based on the relatively low probability of an event requiring PAM instrumentation, the passive nature of the PAM instrumentation, and the availability of alternate means to obtain the required information. The 30 day Special Report for one inoperable channel and the 7 day restoration time for two iv perable channels is consistent with the ITS (except for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restoration time with two inoperable hydrogen monitor channels) and GL 83-36 and has been previously approved by the Staff for Sequoyah Nuclear Plant, Units 1 and 2 (Amendment Nos.159 and 149, respectively) and Clinton Power Station (Amendment No. 31).

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u o The requirement to maintain the Reactor Water Level, Suppression Pool Water Level and Drywell High Range Radiation Monitor instrumentation operable in Operational Condition 3 will be deleted. Post-accident monitoring variables are related to the diagnosis and preplanned actions required to mitigate DBAs assumed to occur in Operational Conditions 1 and 2. Because Suppression Pool Water Level indication will no longer be required in Operational Condition 3,its ACTION requirement was revised to delete the requirement to place the plant in COLD SHUTDOWN, Operational Condition 4. This is consistent with the ITS, which requires that the plant be brought to an operational condition in which the LCO does not apply if a required action cannot be met.

The proposed changes would add a Specification 3.0.4 exception to LCO 3.3.7.5. The proposed required ACTIONS permit continued operation with one iroperable channel and assure an acceptable level of safety since the remaining RG 1.97 channel is still available, the PAM instruments have no active safety function, and a preplanned alternate method of monitoring is identified in the required Special Report. The proposed 7 day restoration time for two inoperable PAM channels provides adequate time to restore inoperable channels without placing undue pressure on plant personnel. Consequently, a plant shutdown due to inoperable PAM instrumentation would not be anticipated. In addition, the passive nature of the PAM instrumentation, the ability to diagnose an accident using alternative instruments and methods, and the low probability of an event requiring PAM instrumentation assure that an acceptable level of plant safety is maintained during the 7 day completion time.

To adequately address the compensatory actions necessary when both PCIVs on a penetration lose indication, the proposed changes revise the description of PCIV Position Indication from " Primary Containment isolation Valve Position Indication" to " Penetration Flow Path Primary Containment Isolation Valve Position Indication," and revise the

" Required Number of Chann31s" from 1 channel per PCIV to 2 channels per penetration.

Thus, when two PCIV indication channels on the same penetration are inoperable, the proposed changes appropriately require restorstion of at least one channel within 7 days or, as a minimum, require isolation of the penetration. Requiring only one instrument channel for penetrations with only one PCIV is acceptable since RG 1.97 and the ITS Bases require only one position indicator for each active PCIV. The PCIVs on an isolated  !

penetration are not required to change state during an accident and would not be I considered active. Therefore, not requiring indication for PCIVs associated with isolated flow paths is also acceptable.

The editorial changes to Tables 3.3.7.5-1 and 4.3.7.51 are administrative in nature and do not reduce any Technical Specification requirements. Deleting the " Minimum Number of Channels" column does not affect the implementation of the Technical Specifications since the proposed required ACTIONS of LCO 3.3.7.5 refer only to the " Required Number of Channels." Deleting the " Applicable Operational Conditions" column and revising the applicability section of LCO 3.3.7.5 will provide consistency between the accident monitoring specification and other NMP2 Technical Specifications sections and is consistent with the proposed changes. Referencing Specification 4.0.5 in Surveillance Requirement 4.4.2 does not reduce any operability requirements or alter any existing surveillance requirements for the safety relief valves.

On an aggregate basis, these proposed changes will provide consistency between NMP2's RG 1.97 instrumentation, the NMP2 Technical Specifications, and the ITS, and are Page 14 of 17

y<. e e-consistent with the purpose of the Technical Specifications as stated in the Commission's Final Policy Statomsnt. Therefore, there is reasonable est arence that operation of NMP2 in the proposed manner will not endanger the public health and safety and that issuance of the proposed amendment will not be inimical to the common defense and security.

10 CFR 50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis using the standards in 10 CFR 50.92 concerning the issue of no significant hazards consideration. Therefore,in accordance with 10 CFR 50.91, the following analysis has been performed:

The onoration of NMP2 In acceiderice with the nronosed amendment, will not involve a alSGificant increase in the orok.hmtv or coriseaumaeas of an .,+sant orevleudv eva!usted.

l PAM instruments are used to help p operator response to postulated accidents. Thus, I the status or operability of PAM irt montation does not affect the probability of L 'previously analyzed accidents. Th ,on-Category 1 PAM instruments being removed from I the Technical Specifications do not meet 6ny of the Commission's screening criteria and are not of controlling importance to safety or necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to public health and safety.

The operability of critical parameters necessary to assure proper response to previously analyzed accidents (i.e., Category 1 instruments) is still controlled by the Technical Specifications. Thus, deleting non-Category 1 instruments will not increase the consequences of any accident previously evaluated.

i PAM instruments are related to the diagnosis and preplanned actions required to mitigata l DBAs assumed to occur in Operational Conditions 1 and 2. A DBA during Operational l Condition 3 is extremely unlikely. The requirement to maintain the Reactor Water Level, l Suppressicn Pool Water Level and Drywell High Range Radiation Monitor instrumentation operable in Operational Condition 3 will be deleted. Because Suppression Pool Water Level indication will no longer be required in Operational Condition 3, its ACTION requirement was revised to delete the requirement to place the plant in COLD SHUTDOWN, Operational Condition 4. This is consistent with the ITS, which requires that the plant be brought to an operational condition in which the LCO does not apply if a required action cannot be '

met. Therefore, deleting the requirement that PAM instruments be operable during Operational Condition 3 and changing the ACTION requirement for Suppression Pool Water Level Monitoring does not affect the probability or consequences of an accident. i The passive nature of the Category 1 PAM instruments (i.e., those instruments that initiate no critical automatic action) and_ the alternate means available to obtain the required information assure an acceptable level of safety is maintained during operation with instrument channels out of service. Since an acceptable level of safety is maintained with inoperable channels, plant startup or operation with inoperable channels will not alter plant response to analyzed accidents. Thus, the proposed changes to the required ACTIONS and the proposed exemption to Specification 3.0.4 will not increase the conseqwces of previously analyzed events.

The proposed changes to the requirements for PCIV indication are consistent with the proposed required ACTIONS. Position indication will still be required for each operable PCIV and penetrations without adequate PCIV indication status will be isolated, thus 1

Page 15 of 17

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. 1 s 9 1 assuring containment integrity in the evant of an accident. Deletion of the " Minimum i Required Actions" column in Table 3.3.7.51 is consistent with the proposed ACTIONS for LCO 3.3.7.5, since compensatory actions are based on compliance with the " Required Number of Channels." Deleting the " Applicable Operating Conditions" column is consistent with the proposed changes and other NMP2 Technical Specifications sections.

Finally, referencing Specification 4.0.5 is an administrative change which does not alter any existing surveillance requirements for the safety relief valves. )

In aggregate, the proposed changes do not affect the plant in a way that could directly contribute to causing or mitigating the effects of an accident. Therefore, the operation of NMP2, in accordance with the proposed amendment, will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The operation of NMP2 In accordance with the oroonsed amendment, will not create the 1 D2ssibility of a new or different kind of accident from any accident previousiv evaluated.

The proposed changes do not represent a physical change to the plant as described in the NMP2 USAR. The proposed changes do not modify any plant equipment and the initial conditions used for the design basis accident analysis are still valid. Thus, no potential initiating evcats are created which would cause any new or different kinds of accidents.

PAM instrumentation is used to guide operator response during postulated accidents.

Those PAM instruments considered of controlling importance to safety are retained in the Technical Specifications. Thus, plant response to previously analyzed events is not altered so as to create any new or different kinds of accidents. Therefore, operation of Nine Mile Point Unit 2 in accordance with the proposed change will not create the possibility of a now or different kind of accident from any previously assessed.

The operation of NMP2, in accordance with the oronosed amendment, will not involve a sianificant reduction in a marain of safety.

The non-Category 1 PAM instruments being removed from the Technical Specifications do not meet any of the Commission's weening criteria. That is, the instruments being proposed for removal are not of contrc!!ing importance to safety or necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to public health and safety. Thus, they are not critical to any margin of safety.

PAM instruments are related to the diagnosis and preplanned actions required to mitigate DBAs assumed to occur in Operational Conditions 1 and 2. A DBA during Operational Condition 3 is extremely unlikely. The requirement to maintain the Reactor Water Level, Suppression Pool Water Level and Drywell High Range Radiation Monitor instrumentation operable in Operational Condition 3 will be deleted. Because Suppression Pool Water Level indication will no longer be required in Operational Condition 3, its ACTION requirement was revised to delete the requirement to place the plant in COLD SHUTDOWN, Operational Condition 4. This is consistent with the ITS, which requires that the plant be brought to an operational condition in which the LCO does not apply if a required action cannot be met. Therefore, deleting the rnquirement that PAM instruments be operable during Operational Condition 3 and changing the ACTION requirement for Suppression Pool Water Level Monitoring does not significantly reduce a margin of safety.

Page 16 of 17

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f Since the Category 1 PAM instruments are passive in nature (i.e., no critical automatic action is assumed to occur from these instruments) and alternate means exist to obtain the required information, an acceptable level of safety is assured when instrument channels are out of service. Also, the probability of an event requiring PAM instrumentation is low.

Continued operation with one channel out of service, and limited plant operation with two channels out of service, does not compromise plant safety margins. An acceptable level of safety is maintained during plant startups and operation with instrument channels out of service. Thus, the proposed changes to the required ACTIONS and the proposed exemption to Specification 3.0.4 will not significantly reduce a margin of safety.

The proposed changes to PCIV indication will assure correct implementation of the  !

ACTIONS discussed above. Isolating the flow path associated with one or two inoperable PCIV indicetion channels is conservative since the subject valve will be positioned as required to assure primary containment integrity. The remaining editorial changes are administrative in nature and by definition do not affect safety margins. Deleting the

" Minimum Operable Channels" and " Applicable Operating Conditions" columns is consistent with the proposed changes. Finally, referencing the requirements of Specification 4.0.5 is an administrative change and by definition does not reduce tN margin of safety. f Therefore, the operation of NMP2 in accordance with the proposed change will not involve a significant reduction in a margin of safety.

Page 17 of 17 l

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