ML20117F950

Proposed Tech Specs Section 3/4.3.2 Re Isolation Actuation Instrumentation

ML20117F950
Person / Time
Site:	Nine Mile Point
Issue date:	05/15/1996
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML20117F918	List: ML20117F913 ML20117F937 ML20117F950
References
NUDOCS 9605200318
Download: ML20117F950 (18)

Text

_ . _ . _ - .

4 ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION i

LICENSE NO. NPF-69 DOCKET NO. 50-410 i

Pronomad Chanaan to Technical Snacifications Replace existing pages v, 3/4 3-17,3/4 3-19A,3/4 3-25, and 3/4 3 28 with the attached revised pages. These pages have been retyped in their entirety with marginal markings to indicate changes to the text. Also, add page 3/4 3-19B, which contains Figure 3.3.2-1, 4

" Allowable and Trip Setpoint Values for the Main Steam Line Tunnel Lead Enclosure."

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9605200318 960515 PDR ADOCK 05000410 P PDR

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE REACTIVITY CONTROL SYSTEMS (Continued) 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Tank Volume vs. Concentration Requirements .............................................. 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS i

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ............ 3/4 2-1 3/4.2.2 AVERAGE POWER RANGE MONITOR SETPOINTS . . . . . . . . . . . . . . . . 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B. . . . . . . . . . . . . 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION R ATE . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION . . . . . . . . . . . . . ' 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation . . . . . . . . . . . . . . . . 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times ............... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance R e q ui rem e nt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-7 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . 3/4310 Table 3.3.2-1 Isolation Actuation Instrume a t i on . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-12 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints .................. 3/4 3-17 Figure 3.3.2-1 Allowable and Trip Setpoint Values for the Main Steam Line Tunnel Lead Enclosure .................................. 3/4 3-198 Table 3.3.2-3 Isolation System Instrumentation Response Time . . . . . . . . . . . . . . . . 3/4 3-20 Table 3.3.2-4 Valve Groups and Associated Isolation Signals . . . . . . . . . . . . . . . . . . 3/4 3-22 Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements .... 3/4 3-25 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION . . 3/4 3-29 Table 3.3.3-1 Emergency Core Cooling System Actuation instrumentation . . . . . . . . 3/4 3-30 Table 3.3.3-2 Emergency Core Cooling System Actuation Instrumentation S e t p oi n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-35 Table 3.3.3-3 Emergency Core Cooling System Response Times . . . . . . . . . . . . . . . 3/4 3-39 NINE MILE POINT - UNIT 2 v Amendment No. //

TABLE 3.3.2-2 .

E ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M

E TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE r-

1. Primary Containment Isolation Sianals o (Continued) h a. Reactor Vessel Water Level

• c 1) Low, Low, Low, Level 1 217.8 in. 210.8 in.

3 2) Low, Low, Level 2 2108.8 in. 2101.8 in.

[ 3) Low, Level 3 2159.3 in. 2157.8 in.

b. Drywell Pressure - High s1.68 psig s1.88 psig

c. Main Steam Line W 1) Radiation - High** s3x Full Power Background s3.6x Full Power Background

• 2) Pressure - Low 2766 psig 2746 psig y 3) Flow - High s121.5 psid s122.8 psid

d. Main Steam Line Tunnel

1) Temperature - High s 167.2* F s 170.6

• F

2) ATemperature - High s70.0'F s 71.7'F

3) Temperature - High MSL Lead Enclosure * *

• s 148.2'F s 151.6

• F l i

e. Condenser Vacuum Low 28.5 in Hg vacuum 27.6 in. Hg vacuum

f. RHR Equipment Area Temperature - High s135 F s 144.5'F g (HXs/A&B Pump Rooms)

E g. Reactor Vessel Pressure - High (RHR Cut-in s128 psig s148 psig y Permissive)

$ h. SGTS Exhaust - High Radiation ' s 5.7x10-3 pCi/cc s 1.Ox10-2pCi/cc z

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- -- - - - - ,-i--

ii - - - - - -

.. t Table 3.3.2-2 (Continued) z ISOLATION ACTUATION INSTRUMENTATION SETPOINTS E

m

$ ** Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of the hydrogen injection test and with the reactor power at greater than m 20% rated power, the normal full-power radiation background level and associated trip and alarm setpoints may be o changed based on a calculated value of the radiation level expected during the test. The background radiation level 9 and associated trip and alarm setpoints may be adjusted during the test program based on either calculations or

$ measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be

. determined and associated trip and alarm setpoints shall be reset within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after completion of the hydrogen  ;

c injection test. At reactor power levels below 20% rated power hydrogen injection shall be terminated, and control h rod withdrawal is prohibited until the Main Steam Line Radiation Monitor trip setpoint is restored to its pre-test value.

" ~ The trip sotpoint and allowable value for a channel may be established based on Figure 3.3.2-1, if:

a. the actual ambient temperature reading for all operable channels in the Lead Enclosure area are equal to or  !

g greater than the ambient temperature used as the basis for the setpoint, and 3 the absence of steam leaks in the Main Steam Line Tunnel Lead Enclosure area is verified by visual inspection w b.

2. prior to increasing a channel setpoint, and e

1. a surveillance is implemented in accordance with Note "d" of Table 4.3.2.1-1.

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-- 1 90 100 110 120 130 140 Lead Enclosure Ambient Temperature (*F)

Used as Basis for Setpoint FIGURE 3.3.2-1 Allowable and Trip Setpoint Values for the Main Steam Line Tunnel Lead Enclosure NINE MILE POINT - UNIT 2 3/4 3-198 Amendment No.

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TABLE 4.3.2.1-1 ,

siE ISOLATION ACTUATION INSTRUMENTATION SURVElLLANCE REQUIREMENTS M

$ OPERATION E CHANNEL CONDITIONS FOR WHICH o CHANNEL FUNCTION CHANNEL SURVEILLANCE IS

9 TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED l z

1. Primary Containment isolation Sianals l

7

a. Reactor Vessel Water Level E._

• 1) Low, Low, Low, Level 1 S Q R(a) 1,2,3

" Low, Low, Level 2 S Q R(a) 1, 2, 3 and

• 2)

3) Low, Level 3 S Q R(a) 1,2,3

b. Drywell Pressure - High S Q R(a) 1,2,3

c. Main Steam Line w 1) Radiation - High S Q R 1,2,3 3 2) Pressure - Low S O R(a) 1 ,

w 3) Flow - High S Q R(a) 1,2,3

$ d. Main Steam Line Tunnel

1) Temperature - High S Q R(b) 1,2,3

2) ATemperature - High S Q R(b) 1,2,3 ,

3) Temperature - High MSL Lead S(d) Q R(b) 1,2,3 l Enclosure y e. Condenser Vacuum - Low S O R(a) 1,2**,3**

h f. RHR Equipment Area Temperature - S Q R(b) 1,2,3 z High (HXs/A&B Pump Rooms)

M g. Reactor Vessel Pressure High (RHR S Q R(a) 1,2,3 z Cut-in Permissive) z P

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

. TABLE 4.3.2.1-1 (Crntinued'

, ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

• During CORE ALTERATIONS and operations with a potential for draining the reactor vessel. This only applies to secondary containment isolation and automatic start of SGTS.

• • When any turbine stop valve is greater than 90% open and/or when the key-locked condenser low vacuum bypass switch is open (in Normal position).

t When handling irradiated fuel in the reactor building and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

(a) Perform the calibration procedure for the trip unit setpoint at least once per 92 days.

(b) Calibration excludes sensors; sensor response and comparison shall be done in lieu of.

(c) Manual isolation pushbuttons are tested at least once per operating cycle during shutdown. All other circuitry associated with manual isolation shall receive a CHANNEL FUNCTIONAL TEST at least once per 92 days as part of the circuitry required to be tested for the automatic system isolation.

(d) In addition to the normal shift channel check,if a channel setpoint has been established using Figure 3.3.2-1, then once per shift the actual ambient temperature reading for all operable channels in the Lead Enclosure area shall be verified to be equal to or greater than the ambient temperature used as the basis for the setpoint.

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NINE MILE POINT - UNIT 2 3/4 3-28 AMENDMENT NO. //, d/

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I ATTACHMENT 8 NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 DOCKET NO. 50-410 Marked Conv of Pronosed Chanaen to Current Technical Snecifications I

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INDEX  :

. i QMITING CONDITIC..S FOR OPERATION AND SURVEILLANCE.REOUIREMENTS

"- PAGE

{

REACTIVITY CONTROL SYSTEMS (Continued) 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................. 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Tank Volume vs. Concentration Requirements.............................................. 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..........'......

3/4 2-1 3/4.2.2 AVERAGE POWER RANGE MONITOR SETP0INTS..................... 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO (00YN OPTION B)..............

3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION F. ATE............................... 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................. 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation .......... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times............ 3/4 3-6 h Table 4.3.1.1-1 Reactor Protection System Instrumentation 3/4 3-7 Surveillance Requirements.................................

3/4 3-10 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.......................

Table 3.3.2-1 Isolation Actuation Instrumentation................. 3/4 3-12 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints....... 3/3 3-17 Wy 3- 198 AttewdW M TW .5*&*f.[fCg],ffL".r f;me s. 3. z-j 3/4 3-20 Table 3.3.2-3 Isolation System Instrumentation Response Time ..... l Table 3.3.2-4 Valve Groups and Associated Isolation Signals....... 3/4 3-22 l

Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance 3/4 3-25 l Requirements..............................................

3/4 3-29 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION... l Table 3.3.3-1 Emergency Core Cooling System Actuation 3/4 3-30 Instrumentation...........................................

Table 3.3.3-2 Emergency Core Cooling System Actuation 3/4 3-35 Instrumentation Setpoints.................................

Table 3.3.3-3 Emergency Core Cooling System Response Times........ 3/4 3-39 he, v Amenduent No. 17 NINE MILE POINT - UNIT 2

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5 TABLE 3.3.2-2 g .

g ISOLATION ACTUADON INSDtUMENTA110N sea rOiNTS ' I E

m g ALLOWABLE g TRIP FUNCTION TRIP SETPOINT VAI'UE s

, I. Primary Containment Isolation Sinnals (Continued)

a. Reactor Vessel Water Level

• w I) IAw, Low, Imw, Level I k 17.8 in. k 10.8 in.

2) law, Iow, Level 2 h 108.8 in. k 101.8 in. i

3) law, Level 3 2159.3 in. k 137.8 in. -

b. Drywell Pressure - High s1.68 psig s1.88 psig

c. Main Steam Line I

' R

1) Radiation - High** s3x Full Power Background s3.6x Full Power Background

2) Pressure -low h766 psig 2746 psig

'f 3) Flow - High s121.5 psid s122.8 psid I i

d. Mala Steam Line Tunnel '

t

1) Temperature - High s 167.2*F s 170.6*F

2) .ATemperature - High s70.0*F $71.7'F  :

3) Temperature - High MSL Lead Enclosu s 148.2*F s 151.6*F  !

e. Condenser Vacuum Imw 28.5 in Hg vacuum 27.6 in. Hg vacuum

f. RHR Equipment Area Temperature - High s135'F s 144.5'F  ;

(HXs/A&B Pump Rooms) i g. Reactor Vessel Pressure - High s128 psig s148 psig  !

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(RHR Cut-in Permissive) it  :

g h. SGTS Exhaust - High Radiation 55.7x104 pCi/cc  !

r, s 1.0x10pCi/cc

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Table 3.3.2-2 (Continued) i x .

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 1

k ** Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of the hydrogen injection test and with o the reactor power at greater than 20% rated power, the normal full power radiation S background level and associated trip and alarm setpoints may be changed based on a 2i calculated value of the radiation level expected during the test. The background

, radiation level and associated trip and alarm setpoints may be adjusted during the c- test program based on either calculations or measurements of actual radiation levels 5 resulting from hydrogen injection. The background radiation level shall be determined H

and associated trip and alarm setpoints shall be reset within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after completion N

of the hydrogen injection test. At reactor power levels below 20% rated power hydrogen injection shall be terminated, and control rod withdrawal is prohibited until the Main Steam Line Radiation Monitor trip setpoint is restored to its pre-test value.

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80 90 100 110 120 130 140 Lead Enclosure Ambient Temperature (*F)

Used as Basis for Setpoint l

r j FIGURE 3.3.2-1 Allowable and Trip Setpoint Values for the Main Steam Line Tunnel Lead Enclosure l

l NINE MILE POINT - UNIT 2 3/4 3-19B Amendment No.

. _ _ _ . - .___.___.._.___-_.___m._ . _ - - - . . . . . _ . - . . _ . _ _ _ _ . _ . _ _ _ . . _ _ . _ . _._.-_m. . . _ _ . . . .

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.. :: 1 TABLE 4.3.2.1-1

• ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS ,

N n g OPERATIONAL P.

H CONDITIONS FOR h CNANNEL CNANNEL CNANNEL WNICN SURVEIL-TRIP FUNCTION CHECK._ FUNCTION TEST CALIBRATION LANCE IS REOUIRED M 1. Primary Containment Isolation Sianals N a. Reactor Vessel Nater Level ..

I h

1) Low, Low, Low, Level 1 S Q R(a) 1, 2, 3 h 2) Low, Low, Level 2 S Q R(a) 1, 2, 3 and * ,

E , 3) Low, Level 3 S Q R(a) 1, 2, 3 N b. Drywell Pressure - High S Q -

R(a) 1, 2, 3  ;[

c. Main Steam Line [

1) Radiation - High S Q R 1, 2, 3 R* - 2) Pressure - Low S Q R(a) 1

3) Flow - High S Q R(a) 1, 2, 3 Y d. Main Steam Line Tunnel L w h v' 1) Temperature - High S Q R(b) 1, 2, 3 p

2) ATemperature - High S Q R(b) 1, 2, 3 [

3) Temperature - High MSL Lead S (d) Q R(b) 1, 2, 3 8-Enclosure i:

e. Condenser Vacuum - Low S Q R(a) 1, 2 * -2 , 3**  ;

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f. RHR Equipment Area Temperature - S Q R(b) 1, 2, 3 g High (HXs/AEB Pump Rooms) y g. Reactor Vessel Pressure High (RHR S Q R(a) 1, 2, 3 o

x Cut-in Permissive)

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- TABLE 4.3.2.1-1 (C:ntinuid)

ISOLATION ACTUATION INSTRUMENTATION SURVElli.ANCE REQUIREMENTS

(

TABLE NOTATIONS

• During CORE ALTERATIONS and operations with a potential for draining the reactor vessel. This only applies to secondary containment isolation and automatic start of SGTS.

• • When any turbine stop valve is greater than 90% open and/or when the key-locked condenser low vacuum bypass switch is open (in Normal position).

t When handling irradiated fuelin the reactor building and during CbiiE ALTERATIONS and operations with a potential for draining the reactor vessel.

(a) Perform the calibration procedure for the trip unit setpoint at least once per 92 days.

(b) Calibration excludes sensors; sensor response and comparison shall be done in lieu of.

.(c) Manual isolation pushbuttons are tested at least once per operating cycle during shutdown. All other circuitry associated with manual isolation shall receive a CHANNEL FUNCTIONAL TEST at least once per 92 days as part of the circuitry required to be tested -

for the automatic system isolation. {

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. 1 ATTACHMENT C 1

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! NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 l DOCKET NO. 50-410 s

hanortina information and No Sionificant Harards Consideration Analvnis I

l INTRODUCTION Item 1.d of Table 3.3.2-2, entitled " Isolation Actuation Instrumentation Setpoints," of the Nine Mile Point Unit 2 (NMP2) Technical Specifications (TS) identifies the isolation actuation I instrumentation temperature setpoints associated with the main steam line tunnel.

Operating data indicates that actual temperatures for the Main Steam Line Lead Enclosure (i.e., item 1.d.3 of Table 3.3.2-2), particularly in the summer, approach the isolation actuation setpoint. In fact, at times during hot summer conditions, operating margin between the actual temperature and the actuation setpoint is often negligible. Niagara Mohawk has confirmed by inspection of the main steam line tunnel that these high tunnel temperatures are not the result of steam leakage in the area. Under these conditions, a i minor disturbance in the turbine building ventilation system, while not otherwise l compromising safe operation, could result in an unwarranted isolation actuation at full ]

power with resulting Main Steam isolation Valve closure and reactor scram.

The main steam tunnel high temperature isolation actuation instrumentation is part of the Leak Detection System (LDS). A schematic diagram of the system is shown on Figure 7.6-1, sheet 2,in the Updated Safety Analysis Report (USAR). The LDS complies with General Design Criterion (GDC) 54, and is discussed in Section 5.2.5 of the USAR. The portion of the LDS in the main steam tunnel is used to detect leakage from the main steam line and initiate signals used for automatic closure of the Main Steam isolation Valves. The monitors located in the main steam tunnel and Lead Enclosure area have sensitivities suitable for detection of increases in ambient air temperature resulting from reactor coolant leakage into the area.

The design calculations for NMP2 estimated an average ambient temperature for the main steam tunnel of 85'F during the winter and 110*F during the summer. To provide the necessary sensitivity year round for a single setpoint, the transient analysis for a steam leak in the Main Steam Tunnel Lead Enclosure area utilized the winter temperature as an initial condition. The temperature setpoints for the main steam tunnelisolation actuation instrumentation were determined using an 80 F initial temperature and a 25 gallon per minute (gpm) steam leak. The isolation instrumentation setpoints assure that a main steam line leak in this area would be isolated before a pipe break occurred.

Niagara Mohawk Power Corporation (NMPC) has determined that the temperatures in the Main Steam Line Tunnel Lead Enclosure resulting from a postulated steam leak are dependent on the initial ambient temperature in the area. The acceptable trip setpoint and allowable temperatures, based on initial ambient temperatures, have been calculated and Page 1 of 4

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plotted on proposed Figure 3.3.2-1. The range of temperatures for the allowable value and l trip setpoint for main steam line isolation are the result of calculating the temperature j increase for a main steam line leak for a range of initial temperatures. Since these i calculated values are based on normal operating conditions, the absence of any steam leaks i in the Main Steam Line Tunnel Lead Enclosure area is required to be visually verified prior to

increasing the setpoint of any temperature instrument channel. Furthermore, a new >

surveillance requirement has been established to verify the continued validity of a setpoint established using Figure 3.3.2-1. The surveillance involves confirming, once per shift, that the actual temperature reading for all operable temperature channels in the Lead Enclosure area are equal to or greater than the ambient temperature used as the basis for the setpoint.

This charige does not affect the setpoint, function, or operation of the temperature

j. instruments in the main steam line tunnel which are not in the Lead Enclosure area.

! EVALUATION i

The effects of the increased main steam tunnel temperatures were evaluated. The results ,

concluded that all equipment and components in the main steam tunnel would remain 4

operable and would perform their intended safety function.

l The increase'in the main steam lead enclosure temperature setpoints has been reviewed

! against Equipment Qualification (EO) documentation. The increase has no adverse impact.  !

j The EO program at NMP2 has a process in place to monitor ambient temperatures in j different environmental zones, including the zones in the main steam lead enclosure.

i Qualified lives of EO equipment are calculated and adjusted by the NMPC EQ group, based j on actual ambient temperatures, i

i The peak temperature as a result of a main steam line break will not change, since the I

dominant e'fect is the energy released by the break.

I j The structural design was also evaluated and found to be acceptable for the increased 4

temperatwe. Structural impacts are bounded by the steam line break analysis.

I A small steam leak in the Lead Enclosure area could increase the ambient temperature in the area without causing a main steam line isolation. Raising the trip setpoint based on an l ambient temperature affected by a steam leak could compromise the ability to detect a leak

of 25 gpm. Therefore, a requirement to confirm the absence of steam leaks in the Lead

Enclosure area, before increasing the setpoint of any temperature instrument channel, has  ;

been incorporated into this proposed TS change. I Temperatures in the Main Steam Line Lead Enclosure area can vary based on environmental conditions, such as the temperature of the outside air or the temperature of the lake water, i which is the source of cooling water for the affected ventilation systems. Since the ambient temperature in the Lead Enclosure area is an input in establishing the trip setpoint l

4 and allowable values, this proposed TS change also includes a surveillance requirement to j verify, once per shift, that the actual temperature reading for all operable temperature

channels in the Lead Enclosure area are equal to or greater than the ambient temperature

used as the basis for the setpoint. This frequency is considered adequate to preclude j ~

operation outside the allowable range and to permit compensatory action should actual Lead Enclosure temperatures be trending down toward the setpoint basis temperature, j Experience has shown that actual Lead Enclosure temperatures respond relatively slowly to l Page 2 of 4

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even large and sudden environmental changes, such as the rapid reduction in lake water temperatures which have been experienced at Nine Mile Point. Available compensatory actions include reducing the trip setpoint consistent with the actual ambient temperature

and adjusting ventilation system parameters to maintain an elevated ambient temperature.

The methodology utilized to determine the allowable values and setpoints is in accordance with Regulatory Guide 1.105, " Instrument Setpoints for Safety-Related Systems," Revision 2, February 1986, and Standard ISA S67.04, "Setpoints for Nuclear Safety Related j Instrumentation used in Nuclear Power Plants," 1982. The instrumentation in the main j steam lead enclosure has been evaluated for the effect of increased temperatures on drift, accuracies, and allowances for environmental effects. The allowable values contain

, sufficient margin to account for instrument accuracy and calibration capability in the new environment. The differences between the trip setpoints and allowable values are adequate to account for expected drift between calibrations.

CONCLUSIONS a

i Based on the:;e evaluations, it is concluded that the revised main steam lead enclosure .

temperature setpoints will not adversely affect any design / operational consideration.

Specifically, the LDS will continue to provide a main steam line isolation for a leak prior to a q pipe break, and the performance of safety-related equipment and structures as a result of j increased steam line lead enclosure temperature has been determined to be acceptable, d

NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS i 10CFR50.91 requires that at the time a licensee requests an amendment, it must provide to j the Commission its analysis using the standards in 10CFR50.92 as to whether no significant i hazards consideration associated with the amendment exists. Therefore,in accordance with i 10CFR50.91, the following analysis has been performed:

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The anaration of NMP2 In accordance with the oronomed amendment will not involve a slanificant incran== In the orobahliity or conneouences of an accident nreviousiv evaluated.

I i The LDS instrumentation in the main steam line tunnel isolates the Main Steam isolation g Valves upon sensing a steam leak of 25 gpm. For an elevated ambient temperature in the l Lead Enclosure area, a setpoint established using the prcposed Figure 3.3.2-1 ensures that l the Main Steam isolation Valves continue to receive an isolation signal upon sensing a steam

! leak of 25 gpm. Verifying the absence of any steam leak in the area prior to raising any

, temperature instrument setpoint ensures that the ability ta sense a 25 gpm :en i. not i compromised by an increased ambient temperature resulting from a smalbr steam leak. The periodic surveillance to verify the actual ambient temperature ensures the continued validity

of the ambient temperature used for the setpoint basis, and provides sufficient advance  ;

) indication to take appropriate compensatory action. Accordingly, this change will not l j involve a significant increase in the consequences of any accident previously evaluated.

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Furthermore, the LDS function provides a mitigation action for a postulated main steam line pipe leak which could lead to a pipe break. This function does not affect any accident i precursors, and the proposed change does not affect the function of the LDS system.

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Accordingly, this change will not involve a significant increase in the probability of any l accident previously evaluated.

i Tha =-- :* . of NMP2. In meenrdance with the cronomad amendment, will not create the l

pasalbluty of a new or different kind of rdd M from any oraviouniv avaluated.  ;

The qualification of safety-related equipment in the main steam lead enclosure is evaluated using actual temperatures and component qualified life is adjusted accordingly. The ,

temperature elements are the only safety-related equipment affected by this change,  ;

therefore, the instrumentation response to previously evaluated accidents will not be adversely affected. This change will not affect the performance of safety related structures.  ;

Accordingly, the design capabilities of those structures, systems and components affected l by the proposed change are not challenged in a manner not previously evaluated so as to '

create the possibility of a new or different kind of accident from any previously evaluated.

The anaration of NMP2 In accordance with the pronomad amendment. will not involve a minniflemt radnetton in a marain of safetv. '

The proposed change provides a range of setpoints and allowable values for the Main Steam Line Tunnel Lead Enclosure temperatures. The calculation of the allowable values and trip setpoints was performed using the same methodologies as previously employed. For an elevated ambient temperature in the Lead Enclosure area, a setpoint established using the proposed Figure 3.3.2-1 ensures that the Main Steam isolation Valves receive an isolation signal upon ser. sing a steam leak of 25 gpm, resulting in a main steam line isolation prior to a pipe break. Therefore, the proposed change provides the same level of protection against a main steam line break as the existing setpoint values. The proposed setpoints will provide increased scram avoidance, and thereby reduce unnecessary challenges to the plant shutdown systems. Accordingly, the proposed change does not result in a significant reduction in a margin of safety.

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