ML17286A663

Proposed Tech Specs Re Isolation Actuation Instrumentation

ML17286A663
Person / Time
Site:	Columbia
Issue date:	03/11/1991
From:	Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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ML17286A661	List: ML17286A660 ML17286A663
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NUDOCS 9103210144
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ATTACHMENT2 91032i0140 910311 PDR ADOCK 05000397 PDR

INSTRUMENTATION 3/4.3. 2 ISOLATION ACTUATION INSTRUHEHTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1.

shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEH RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2"1.

ACTIONS

a. With an isolation actuation insirumentation channel trip setpoint less conservative than the value show~n the Allowable Values column of Table 3.3.2"2, declare the channe nhperable until the channel is restored to OPERABLE status with trip setpoint adjusted consistent with the Trip Setpoint e.

b. VR~e-numM~f OPGNBLE-c e4s~ss-tha~q &red-b~4e-

+QHeum-OPERABLE-Gh annus-p -em-requ+rement-fina-4Hy-j;<i'~4 A A

~~~pped-eondQ4o 4jy-one-houv~~e-peovi.~n'~

Spea4NzaMo ppQ-cab4e.

c. With the number of 0 E channels less than required by the Minimum OPERABLE Channels ip Sy requi.rement for both, trip systems, place at least on p syste " in the tripped condition within one hour .and take the IOH require by Table 3.3.2-1.

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%BI mqu4-ed-by-TaNe-3~~or that-TAy-FunetAon-sh~e-4akeo; I

-t4 m~opemlAe-charm&~~-the-4mpped-cond+tAo except mhe~~ou44-.cause-4heM4p-Funa4Aon-to-occ~.

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WASHINGTON NUCLEAR - UNIT 2 3/4 3-10

Insert A to Section 3.3.2b

'elete Section 3.3.2b and replace with the following:

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system:

l. If placing the inoperable channel(s) in the tripped condition would cause an isolation, the inoperable channel(s) shall be 'restored to OPERABLE status within a) 12 hours for trip functions common to RPS Instrumentation; "

and b) 24 hours for trip functions not common to RPS Instrumentation.

or the ACTION required by Table 3.3.'2-1 for the affected trip function shall be taken.

OR

2. If placing the inoperable channel(s) in the tripped conditions would not cause an isolation, the inoperable channel(s) and/or that trip system shall be placed in the tripped condition within a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common to RPS Instrumentation; and b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common to RPS Instrumentation.

The provisions of Specification 3.0.4 are not applicable.

Insert B to Note Referenced in Section 3.3.2 Action c Delete Note and replace with the following:

Place one trip system (with the most inoperable channels) in the tripped condition. The trip system need not be placed in the tripped condition when this would cause the isolation to occur.

I A

,ur CQMTRQLLE0 CQPV TABLE 3. 3. 2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 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 />.

ACTION 21 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or 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 />.

ACTION 22 Close the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and declare the affected system inoperable.

ACTION 23 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 24 Restore the manual initiation function to OPERABLE status..

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or'lose the affected system isqlation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOMN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOMN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 25 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 26 Lock close or close, as applicable, the affected system isolation valves within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.and declare the affected system inoperable.

TABLE NOTATIONS "May be bypassed with reactor steam pressure < 1037 psig and all turbine stop valves closed.

""When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

¹During CORE ALTERATIONS and operations with a potential for draining the reactor vessel. .C (a)A channel may be placed in an inoperable status for up to hours for required surveil.lance without placing the trip system in t e tripped con-dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b)Also actuates the standby gas treatment system.

(c)Also trips and isolates the mechanical vacuum pumps.

(d)A channel is OPERABLE if 2 of 4 detectors in that channel are OPERABLE.

(e)Also actuates secondary containment ventilation isolation dampers per Table 3. 6. 5. 2-1.

(f)Closes only RWCU system outboard isolation valve RWCU-V-4.

(g)Only valves RHR-V-123A and RHR-V-123B in Valve Group 5 are required for primary isolation.

(h)Manual initiation isolates RCIC-V-8 only and only with a coincident reactor vessel level-low, level 3.

(i)Not required for RHR-V-8 when control is transferred to the alternate remote shutdown panel during operational conditions 1, 2 8c 3 and the isolation interlocks are bypassed. Mhen RHR-V-8 control is transferr'ed to the remote shutdown panel under operational modes 1, 2, and 3 the associated key lock switch will be locked with the valve in the closed position. Except RHR"V"8 can be returned to, and operated from, the control room, with the interlocks and automatic isolation capability reestablished in operational conditions 2 and 3 when reactor pressure is less than 135 psig.

MASHINGTON NUCLEAR " UNIT 2 3/4 3-15 Amendment No. 58

0 TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREHENTS CIIAHNEL OPERATIONAL CIIANHEL FUNCTIONAL CHANNEL CONDITIONS FOR MHICll TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE RE UIRED n

I 3. REACTOR MATER CLEANUP SYSTEH ISOLATION m

a. h, -F1ow - lligh 1, 2, 3

b. Ileat Exchanger Area Temperature - lligh 1, 2, 3 C. Heat Exchanger Area Ventilation h Temperature- o I I i gh Pump Area Temperature - Iligh 1 2 3 0 Pump Room A N.A SA -1 2 3 Pump Room 0 H.A. SA 1, 2, 3

e. Pump Area Ventilation h Temp. - Iligh Pump Room A H. A. SA R 1, 2, 3 Pump Room 0 N.A... SA R 1, 2, 3

f. SLCS Initiation N.A. R H.A. 1, 2, 3

g. Reactor Vessel Mater Level - Low Low, Level 2 H.A. 1, 2, 3

h. RWCU/RCIC Line Routing Area Temperature - lligh H.A. 1, 2, 3 1 RWCU Line Routing Area .

Temperature -. lligh H. A. SA 1 2 3 Hanual Initiation N.A. R 1 2 3

4. REACTOR CORE ISOLATION COOLING SYSTEtl ISOLATION

a. RCI C S team Line F 1 ow - Iligh S 1 2 3

b. RCIC/RllR Steam Line Flow - lligh S 1, 2, 3 C. RCIC Steam Supply Pressure-Low N.A. 1 2 3

d. RCIC Turbine Exhaust Diaphragm O Pressure - lligh H.A. 1, 2, 3

e. RCIC Equipment Room Temperature - lligh N.A. 1, 2, 3

f. RCIC Equipment Room 6 Temperature - lligh 1 2 3

TABLE 4.3. 2. 1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE UIRENEHTS CHANNEL OPERATIONAL CfIANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICfl.

TRIP FUNCTION CIIECK TEST CALIB RAT ION SURVEILLANCE RE UIREO

1. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level-M 1) Low, Level 3 S 1, 2, 3

2) Low Low, Level 2 N.A. 1 2 3

b. Drywell Pressure - High H.A. J4- Q 1,2,3

c. IIain Steam Line A

1) Radiation - High Pressure - Low S 1, 2, 3 0K 2)

3) Flow - Iligfi N.A. p @.

p-1 S 1 2 3

d. Hain Steam Line Tunnel e.

Temperature - fligh IIain Steam Line Tunnel H.A. SA 1,2,3 0 h, Temperature - High N. A. SA R 1, 2, 3

f. Condenser Vacuum - Low H.A. R 2* IVI Hanual Initiation 3A',

g. H.A. R N.A.= 2, 3 U

2. SECONDARY CONTAINMENT ISOLATION A

Reactor Building Vent Exhaust Plenum Q

'6 b.

Radiation - High Drywell Pressure - fligh S

.N.n.

1, 2, 3, and 1 2 3

"" ~

C. Reactor Vessel Water Level - Low Low, Level 2 H~ A. 1,2,3, Hanual Initiation N.A. 2, 3, and andk',

C)

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUHEHTATIOH SURVEILLANCE RE UIREHEHTS C) CkkAHNEL OPERATIONAL Ck{AHHEL FUNCTIONAL CHANNEL CONDITIONS FOR WIIICkl TRIP FUNCTION CkkECK TEST CALIBRATION SURVEILLANCE RE UI RED n

4. REACTOR CORE ISOLATION COOLING SYSTEH ISOLATION (Continued)

Kl

g. RWCU/RCIC Steam Line Routing I

Area Temperature - High H.A.

~

SA 1 2 3 I

h. Drywell Pressure - kligh N.A. 1 2 3 lkanual Initiation N.A. R H.A. 1 2 3

5. RHR SYSTEM SkkUTDOMH COOLING NODE ISOLATION

a. Reactor Vessel Water Level-Low, Level 3 1, 2, 3

b. Reactor Vessel (RHR Cut-in Permissive) Pressure - High N.A. 1 2 3 c ~ Equipment Area Temperature-lkigh Equipment Area Ventilation SA 1 2 3 I

U 4 Temp. - kkigh N.A. SA 1 2 3 A Shutdown Cooling Return Q Flow Rate - High N.A. 1, 2, 3 RHR kkeat Exchanger Area Temperature - Higt> N.A. 1, 2, 3

g. Hanual Initiation H.A: .R 1, 2, 3 TABLE NOTATIONS When reactor steam pressure > 1037 psig and/or any turbine stop valve is open.

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

During CORE ALTERATIOH and operations with a potential for draining the reactor vessel.

ATTACHMENT3

o GE Nuclear Energy General Electric Company 175 Curtner Avenue, San Jose. CA 95125 Harch 5, 1991 Hr. P. L. Powell Washington Public Power Supply System North Power Plant Loop, Building 56 Hail Drop 956B Richland, Washington 99352

Subject:

Assessment of Hanford-2 Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses

References:

1) Sullivan, W. P., et. al., "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," NEDC-31677P-A, General Electric Company, July 1990.

2) Frederick, L. G., et. al., "Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," NEDC-30851P-A Supplement 2, General Electric Company, Harch 1989.

3) WPPSS letter G02-89-038 dated Harch 8, 1989, G. C. Sorensen to Nuclear Regulatory Commission,

Subject:

Nuclear Plant No.

2, Operating License NPF-21, Request for Amendment to Technical Specification Table 4.3.2. 1-1, Isolation Actuation Instrumentation Surveillance Requirements.

The Reference 1 document is an extension to the BWR Owners'roup Technical Specification Improvement analyses conducted for the BWR Reactor Protection System (RPS) and Emergency Core Cooling System (ECCS) actuation instrumentation. It provides an analysis of the isolation actuation instrumentation which is not common to the RPS and ECCS instrumentation.

The analysis for isolation actuation instrumentation common to the RPS and ECCS instrumentation is presented in Reference 2. Bases are provided for extending the isolation actuation instrumentation surveillance test intervals from 1 to 3 months and allowed out-of-service times for tests and repairs from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, respectively.

These test interval extensions reduce the potential for unnecessary plant scrams, excessive test cycles on equipment, and the diversion of plant personnel and resources to perform unnecessary testing. The allowed out-of-service time extensions enhance tests and repairs by reducing the potential for operator errors. The overall effect of the changes to the test intervals and allowed out-of-service times provides a net improvement to plant safety and operations.

J The USNRC Safety Evaluation Reports (SERs) for References I and 2 conclude that they provide an acceptable basis for extending surveillance test intervals and allowed outage times for isolation actuation instrumentation.

The SERs stipulate, that for plant-specific application of the technical specification changes for isolation instrumentation that are proposed, the licensee must confirm the applicability of the generic analyses to the plant. This letter responds to that stipulation.

References I and 2 provide bounding analyses of the impact of proposed technical specification changes for isolation actuation instrumentation.

Section 5.5 of Reference I provides verification that the results of the generic analyses of the various product lines are applicable to the individual plant technical specification requirements. This evaluation included a comparison of isolation actuation instrumentation STIs and calibration intervals given in the current plant-specific technical specifications to those evaluated for the four product lines. Identified differences were then evaluated to verify that the product line analyses envelope these differences.

Enclosure I provides a matrix listing of STIs and calibration intervals given in current technical specifications of individual BWR5/6 plants included in this study with Hanford-2 added in the far right column for comparison (in place of Clinton, a BWR6 solid-state logic plant). The first column lists the isolation trips for Grand Gulf, the plant that was used in the analyses. The succeeding columns list the isolation trips for the remaining plants in the product line. As can be seen upon examination of this table,,isolation trips for Hanford-2 are essentially equivalent, by isolation function, to those of the two La Salle plants.

It should also be pointed out that semi-annual surveillance test intervals are specified in the Hanford-2 Technical Specification for a number of temperature and delta-temperature trip circuits. Justification for these extended test intervals is provided in Reference 3. It is based upon replacement of equipment using Riley Hodel 86 temperature switches in the leak detection system with General Electric NUHAC instrumentation.

According to Reference 3, all functions of the leak detection system remained unchanged, while overall system reliability was improved with channel functional test intervals extended to six months. These extensions are independent of the current assessment since their basis was improved hardware reliability.

The majority of the plants have the same logic and number of sensor variables that initiate system isolation, given a pipe break/leak or high radiation level. There are, however, specific cases where the number of sensor variables is less than the number for the majority of plants. The effect of these differences in the number of sensor variables on the overall analysis results was concluded to be small. This conclusion was based on the results from case studies performed for different types of instrumentation logic and number of sensor variables. Table 5.5 of Reference I (Enclosure 2) provides a summary of these case studies and results. As can be seen from this table there are, at most, negligible increases in isolation function failure frequencies when the STIs and AOTs are changed to the proposed values.

Hanford-2 isolation actuation instrumentation and logic were compared against the case studies summarized in Table 5.5 and were determined to be bounded by the envelope which they define. Therefore, the generic analyses of NEDC-30851P-A and NEDC-31677P-A are applicable to the Hanford-2 plant and provide an adequate basis for Technical Specification changes to extend the STIs and AOTs for Hanford-2 isolation actuation instrumentation.

R. P. Raftery, Principa Engineer A. E. Rogers, Hanager Reliability Engineering Services Reliability Engineering Services W/Enclosures

E II

ENCLOSURE 1 APPENDIX C-2 OF NEDC-31677P-A WITH HANFORD-2 ADDED FOR COMPARISON BWR-5/6 (RELAY) AND BWR-6 (SOLID STATE)

ISOLATION ACTUATION INSTRUMENTATION TECHNICAL SPECIFICATION REQUIREMENTS

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 1 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BWR-6 BWR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 PRIMARY CONTAINMENT

a. RPV Low Water Level 2 H M H H R(16) R(16) R(16) R(16)

b. RPV Low Water Level 2 M H (ECCS Div 3) R(16) R(16)

c. RPV Low Level 1 (19)

(ECCS Div 1 4 2) R(16) R(16)

d. Drywell Pressure High M M R(16) R(16) R(16) R('16)

e. Drywell Pressure High M (ECCS Div 1 & 2) R(16) R(16)

f. Drywell High Pressure H (ECCS Div 3) R('16) R(16)

g. Containment 8 Drywell M(13)

Vent Exhaust Rad High A(13)

1. Cont. Bldg. NVAC

2. Cont. Honitoring 5 Process Sampling

h. Hanual Initiation H(17) R (2) H R NA NA (2) NA NA i'PV Low Water Level 3 R(16)

j. Standby Gas Treatment M Sys Exhaust-Rad High R TOP LETTER = SURVEILLANCE TEST INI'ERVAL BOI'TOH LETTER = CALIBRATION INI'ERVAL W = WEEKLY H = HONTNLY Q = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

ISOLATION ACTUATION INSTRUMENTATIOH SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 2 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BWR-6 BWR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 PRIMARY CONTAINHEHT (Continued)

k. Containment Purge Radiation High

l. Rx Bldg. Pipe Chase Temp-High (3 trips)

m. Rx Bldg. Temp. High

n. Cont. Bldg. Fuel Transf Pool Vent Plenum Rad

o. Main Steam Line Rad. HI

p. Cont/Drywell Purge Exhaust Plenum Rad High TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOM LETTER = CALIBRATIOH INTERVAL

'W = WEEKLY M = MONTHLY Q = QUARTERLY SA = SEMIANHUALLY A = ANNUALLY R = REFUELING OUTAGE

I ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREHENTS - BWR-5/6 RELAY PLANTS Page 3 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BWR-6 BWR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amnd. ¹70 SECONDARY CONTAINMENT

a. RPV Lou Water Level 2 H H R(16) R(16) R(16)

b. Drywell Pressure High H H R(16) R(16) R(16)

c. Fuel Handling Area/Vent H(12)

Exhaust Rad High R(12)

d. Fuel Handling Area Pool sweep Exhaust Radiation High

e. Hanual Initiation H('17) R (2) R NA NA (2) NA

f. Rx Bldg Annulus Vent Exhaust Rad. High

g. Containment Pres-High

h. Fuel Building Vent High Radiation i'x Bldg. Vent Exhaust Plenum Rad. - High

j. Fuel Pool Vent Exhaust Radiation - High

k. Rx Bldg. Below Refuel Floor - Rad. High TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOH LETTER = CALIBRATION INTERVAL W = WEEKLY H = MONTHLY Q = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

4 4

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREHENTS - BWR-5/6 RELAY PLANTS Page 4 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Nanford 2 BWR-6 BWR-6 BNR-6 BNR-5 BNR-5 BUR-5 BNR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 HAIN STEAH LINE ISOLATION

a. RPV Hater Level 1 H H H H H R(16) R(16) R(16) R R R(3)(16)

b. Hain Steam Line H H H(3) H(3) H(3)

High Radiation R R R(3) R(3) R(3)

c. Main Steam Line H H H M(3) H(3) H(3)

Pressure - Low R(16) R(16) R(16) Q(3) Q(3) R(3)(16)

d. Hain Steam Line H H M H(3) M(3) M(3)

Flow - High R(16) R(16) R(16) R(3) R(3) R(3)('16)

e. Condenser Vacuum-Low H H M H(3) H(3) H(3)

R(16) R(16) R(16) Q(3) Q(3) R(3)(16)

f. Hain Steam Line H(3) H(3) M(3) SA Tunnel Temp. High R(3) R(3) R(3) R

g. Hain Steam Tunnel H(3) H(3) H(3) SA Delta Temp High R(3) R(3) R(3) R

h. Hanual Initiation M(17), R (2)

NA NA (2) i'urbine Bldg. MSL H H Temperature High R R(16)

j. HSL Lead Enclosure H(3)

Temperature - High R(3)

TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOM LETTER = CALIBRATION INTERVAL ll = NEEKLY H = HONTNLY Q = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 5 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NMP 2 Nanford 2 BWR-6 BWR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹7 RWCU SYSTEH ISOLATION

a. RWCU Delta Flow-High H(3)

R(3)

b. RWCU Delta Flow Timer H(3)

R(3)

c. RWCU Equip. Area (9)

Temperature Nigh 1 ~ Pump Room A,B H(3) SA 8 C (for Clinton) R(3) R

2. MX Room Temp - High H(4) H(4) H(3) SA (2 rms. at Clinton) Q(4) Q(4) R(3) R

d. RWCU Equip. Area H(4) H(4) SA*

Delta Temp. High Q(4) Q(4) R

e. RPV Water Level 2 H H H R(16) R(16) R(16) f ~ HSL Tunnel Temp-High

g. HSL Tunnel Delta Temp M H H N I gh A R R

h. SLCS Initiation H(18) H(18) H(18) R R R(3) R NA NA NA NA NA NA(3) NA

i. Manual Initiation H(17) R (2) R NA NA (2) NA

j. Rx Bldg. Pipe Chase H(3) SA*~

Temp - Nigh (3 Trips) R(3) R Two Rooms

• • - Two Locations TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOM LETTER = CALIBRATION INTERVAL W = WEEKLY M = MONTHLY Q = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 6 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NMP 2 Nanford 2 BWR-6 BWR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 REACTOR CORE ISOLATION COOLING SYSTEH ISOLATION

a. RCIC Steam Line F low H M H H High R(16) R(16) R(16) R(16)

b. RCIC Steam Line Flow High - Time Delay

c. RCIC Turbine Exhaust H H H H Diaphram Pres. High R(16) R(16) R(16) R(16)

d. RCIC Steam Supply H Pressure - Low R(16) R(16) R(16) R(16)

e. RCIC Equip. Room Temp SA High R

f. RCIC Equip. Room H(11) H(11) SA Delta Temp - High Q R

g. MSL Tunnel Temp-High h ~ HSL Tunnel Delta Temp High

i. MSL Tunnel Temp Timer .M Q

j. RHR Equip. Room Temp.

High A

k. RHR Equip Room Delta Temp - High TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOM LETTER = CALIBRATION INTERVAL W = MEEKLY M = HONTHLY Q = QUARTERLY SA = SEMIANNUALLY A = ANHUALLY R = REFUELING OUTAGE

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 7 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BWR-6 8'WR-6 BWR-6 BWR-5 BWR-5 BWR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 REACTOR CORE ISOLATION COOLING SYSTEH ISOLATION (Continued)

l. Drywell Pressure-High H H R(16) R(16) R(16) R(16)

m. Hanual Initiation H R R H R NA NA NA NA NA

n. RHR/RCIC Steam Flow NI M H H R(16) R(16) R(16)

o. RCIC Tunnel Temp-High SA R

p. RCIC Tunnel Delta Temp- (15)

High TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOM LETTER = CALIBRATION INTERVAL W = WEEKLY H = HONTHLY Q = QUARTERLY SA = SEMIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE REQUIREHENTS - BMR-5/6 RELAY PLANTS Page 8 Grand Gulf Perry 1 Riverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BMR-6 BMR-6 BMR-6 BMR-5 BMR-5 BMR-5 BMR-5

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ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. <<70 ~

RHR SYSTEH ISOLATION

a. RHR Equip. Room Temp H(11) H(11) H(3) SA High Q(11) Q(11) R(3) R

b. RHR Equip. Room H(11) H(11) SA Delta Temp High Q(1'I ) Q(11) R

c. RPV Lou Mater Level 3 H H H(11) H(11)

R(16) R('16) R(16) R(11) R(11)

d. RPV Pressure High H H H(11) H(11) H(3)

R(16) R(16) R(16) Q(11) Q(11) R(3)

e. DryMeil Pressure High (9) H H R(16) R(16) R(16) 1 ~ RHR Test Line

2. Fuel Pool Cooling

f. Hanual Initiation H(17) R (2) R NA NA (2) NA 9 RPV Lou Mater Level 1 H R(16)

h. RHR Pump Suction Flou H(11) H(11)

High Q(11) Q(11)

i. RHR/RCIC Steam Line Floe - High

j. RHR Area Cooler Temp SA High R

k. RHR Heat Exch. Steam Supply Floe - High

l. RHR Equip Area Delta (15)

Temp - High TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOH LETTER = CALIBRATION INTERVAL M = MEEKLY H -" HONTHLY Q = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

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ISOLATION ACTUATIOH INSTRUHENTATIOH SURVEILLANCE REQUIREHENTS - BMR-5/6 RELAY PLANTS Page 9 Grand Gulf Perry 1 Rlverbend LaSalle 1 LaSalle 2 NHP 2 Hanford 2 BMR-6 BMR-6 BWR-6 BMR-5 BWR-5 BMR-5 BWR-5 ISOLATION SIGNALS Note No. Amend. ¹59 Amend. ¹16 Amend. ¹20 Amend. ¹50 Amend. ¹33 Amend. ¹5 Amend. ¹70 MANUAL INITIATION H(2) R(2) R(2)

NA(2) NA(2) NA(2)

TOP LETTER = SURVEILLANCE TEST INTERVAL BOTTOH LETTER = CALIBRATION INTERVAL W = WEEKLY H "- HONTNLY 0 = QUARTERLY SA = SEHIANNUALLY A = ANNUALLY R = REFUELING OUTAGE

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ISOLATION AC'IUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS - BWR-5/6 RELAY PLANTS Page 10 NOTES:

(1) Requirements stated under RCIC System Isolation. Noted in table under RHR since other plants note here.

(2) Requirements called out under "Manual Initiation>> in Tech Spec. Other plants have specified under each isolation.

(3) Requirements stated under Primary Containment Isolation instead of location indicated in Table.

(4) Title for LaSalle 1 & 2 is RWCU Heat Exchanger Area versus RWCU Equipment Area.

Judged to be the same measurement with different titles'5)

Entry appears in two locations - one under >>RHR HighsSystem Steam Condensing Mode Isolation" and another under

>>RHR System Shutdown Cooling Mode Isolation." Judged to be the same signals (6) Data entered under "Primary & Secondary Isolation>> heading. No separation between Primary & Secondary Isolations.

(7) Same locations as >>RWCU Equipment Area Temp "

(8) Called out under both RCIC & RHR Isolations.

(9) Specific areas are not specified in all Tech Specs. Where general requirements are specified in Tech Specs, the calibration &

functional test intervals are entered under the general title. If specific areas are specified these are entried under the specific titles noted.

(10) Deleted.

(11) Lasalle 1 & 2 have two divisions for RHR System Isolations. One is <<RHR System Condensing Mode Isolation. The second is

>>RHR Shutdown Cooling Mode Isolations.>> These heading have been combined in the table under >>RHR System Isolations.>>

(12) Titled >>Rx Bldg. Above Refuel Floor Exhaust Radiation - High." Judged to be the same as "Fuel Handling Area Exhaust Radiation High" for other plants.

('13) Call out is general. Separation or additional detai l not provided in Tech Spec.

(14) Titled >>RHR Heat Exchanger A/B Ambient (Delta) Temperature - High." Judged to be simi liar to RHR Room Ambient (Delta) Temperature - High.>>

(15) Tech Spec. for Lasalle 2 is stated as >>W>> Temp. We have noted this as Delta Temp. since other plants were specified this way.

(16) Calibrate trip unit at least once per 31 days.

(17) Manual initiation a~itches tested once per 18 months during shutdown. All other circuitry associated with manual isolation shall receive channel functional test once per 31 days as part of circuitry required to be tested for automatic system actuation.

(18) Each train or logic channel shai.l be tested at least every other 31 days.

(19) Level 2 for Clinton technical specification.

0 ee ENCLOSURE 2 TABLE 5.5 OF NEDC-31677P-A

SUMMARY

OF CASE STUDIES PERFORMED

• FOR PLANTS IN DIFFERENT PRODUCT LINES Absolute Increase in Isolation Failure Freq. (Events/Yr)

Number of Logic Type Sensor Per Value Variables Per Variable Minimum Maximum 2 out of 2 -3.5E-08 -1.5E-08 1-out -of-2 Twice -1.1E-08 1 out of 1 -1.0E-08 -4.4E-09 3 1 out of 1 -4.2E-10 5.0E-11 1 out of 1 -4.0E-11 2.1E-10 1 out of 1 2.9E-09 3.2E-09 1 1outof 2 4.1E-ll 9.1E-11 1 2 out of 2 5.5E-11 2.2E-08

• Does not include BWR-6 solid-state logic plant.

• • High area and delta temperatures considered as one variable.

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