ML18033B331

Proposed Tech Specs Re Reactor Pressure Instrument Channel

ML18033B331
Person / Time
Site:	Browns Ferry
Issue date:	05/24/1990
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18033B330	List: ML18033B328 ML18033B331
References
NUDOCS 9005310173
Download: ML18033B331 (21)

Text

ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION BROGANS FERRY NUCLEAR PLAN UNIT 2 (TVA BFN TS 287) 900524 9OOS3>Oa7S QSOooesO o

>0< "PDOCK pDC P

UNIT 2 EFFECTIVE PAGE LIST REMOVE INSERT 3.2/4.2-16 3.2/4.2-16 3.2/4.2-17 3.2/4.2-17*

3.2/4.2-40 3.2/4.2-40 3.2/4.2-41 3.2/4.2-41*

3.2/4.2-44 3.2/4.2-44*

3.2/4.2-45 '3.2/4.2-45 3.2/4.2-61 3.2/4.2-61 3.2/4.2-6la 3.2/4.2-61a

• Denotes overleaf or spillover page.

h TABLE 3.2.8 (Contipued)

Hinimum No.

Operable Per

~Tri ~l Fun ion Tri Level in ~Ati n R mark Instrument Channel 450 psig + 15 1. Below trip setting permissive Reactor Low Pressure for opening CSS and LPCI (PIS-3-74 A & 8) admission valves.

(PIS-68-95, 96)

Instrument Channel 230 psig + 15 1. Recirculation discharge valve Reactor Low Pressure actuat'ion.

(PS-3-74 A & 8)

(PS-68-95, 96)

Core Spray Auto Sequencing 6< t <8 sec. l. With diesel power Timers (5) 2. One per motor LPCI Auto Sequencing 0< t <1 sec. 1. With diesel power Timers (5) 2. One per motor I

RHRSW Al, B3, Cl, and D3 13< t <15 sec. 1. With diesel power Timers 2. One per pump I

Core Spray and LPCI Auto 0< t <1 sec. 1. With normal power Sequencing Timers (6) 6< t <8 sec. 2, One per CSS motor 12< t <16 sec. 3. Two per RHR motor 18< t <24 sec.

RHRSW Al, B3, Cl, and D3 27< t < 29 sec. 'A 1. With normal power Timers 2. One per pump

TABLE 3.2.B (Continued)

C tu Minimum No.

Operable Per

~Tri ~l Fn 'n Tr Lv in ~Ai n Rmrk 1(16) ADS Timer 105 sec + 7 l. Above trip setting in conjunction with low reactor water level permissive, low reactor water level, high drywell pressure or high drywell pressure bypass timer timed out, and RHR or CSS pumps running, initiates ADS.

1(16) ADS Timer (12 1/2 min.)

(High Drywell Pressure 12 1/2 min. g2 l. Above trip setting, in conjunction with low Bypass Timer) reactor water level permissive, low reactor water level, 105 sec.

delay timer, and RHR or CSS pumps running, 4J initiates ADS.

M Instrument Channel- 100 +10 psig 1. Below trip setting defers ADS RHR Discharge Pressure actuation.

lV Instrument Channel I 185 gl0 psig 1. Below trip setting defers ADS CSS Pump Discharge Pressure actuation.

1(3) Core Spray Sparger to Reactor Pressure Vessel d/p 2 psid +0.4 l. Alarm to detect core sparger pipe break.

RHR (LPCI) Trip System bus N/A 1. Monitors availability'. of power monitor power to logic systems.

Core Spray Trip System bus N/A 1. Monitors availability of power monitor power to logic systems.

ADS Trip System bus power N/A 1. Monitors ava>lability of monitor power to logic systems and valves.

TABLE 4.2.A SURVEILLANCE RE()UIREHENTS FOR PRIHARY CONTAINHENT AND REACTOR BUILDING ISOLATION INSTRUHENTATION

~Fn~iqn Fun i nl Tst libr i n Fr unc Instrum n h Instrument Channel '- (1) (27) Once/18 Honths (28) Once/day Reactor Low Mater Level (LIS-3-203A-D)

Instrument Channel- (31) Once/18 months None Reactor High Pressure (PS-68-93 8 94)

Instrument Channel (1) (27) Once/18 months (28) Once/day Reactor Low Water Level (LIS-3-56A-D)

Instrument Channel (1) (27) Once/18 Honths (28) N/A High Drywell Pressure (PIS-64-56A-D)

Instrument Channel 29 (5) Once/day High Radiation Hain Steam Line Tunnel Instrument Channel- (29) (27) Once/18 Honths (28) None Low Pressure Hain Steam .

Line (PIS-1-72, 76, 82, 86).

Instrument Channel- (29) (27) Once/18 Honths (28) Once/day High Flow Hain Steam Line (PdIS-1-13A-D, 25A-D, 36A-D, 50A-0)

TABLE 4.2.A (Cont'd)

SURVEILLANCE RE()UIREHENTS FOR PRIMARY CONTAINMENT AND REACTOR BUILDING ISOLATION INSTRUMENTATION

~Foe~in Fun ion 1 T 1 br in r un In mn Instrument Channel Once/3 months (27) Once/operating cycle None Steam Line Tunnel High

-'ain Temperature Instrument Chanriel- (1) (22) Once/3 months Once/day (8)

Reactor Building Ventilation High Radiation Reactor Zone Instrument Channel (1) (22) Once/3 Months Once/day (8)

Reactor Building Ventilation High Radiation Refueling Zone Instrument Channel- (4) N/A SGTS Train A Heaters Instrument Channel- (4) N/A SGTS Train 8 Heaters Instrument Channel- (4) N/A SGTS Train C Heaters Reactor Building Isolation (4) Once/operating cycle N/A Timer (refueling floor)

Reactor Building Isolation (4) Once/operating cycle N/A Timer (reactor zone)

TABLE 4.2.8 SURVEILLANCE RE(}UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS F nction Fun i nal T alibr i n In rumen Ch k Ins'trument Channel ('1) (27) Once/18 Honths (28) Once/day Reactor Low Mater Level (LIS-3-58A-D)

Instrument Channel (1} (27) Once/18 Honths (28) Once/day Reactor Low Water Level (LIS-3-184 8 185)

Instrument Channel (1) (27) Once/18 Honths (28} Once/day Reactor Low Water Level (LIS-3-52 5 62A)

Instrument Channel (1) (27) Once/18 Honths (28) none Drywell High Pressure (PIS-64-58E-H)

Instrument Channel (1) (27) Once/18 Honths (28) none Drywell High Pressure (PIS-64-58A-D)

Instrument Channel (1) (27) Once/18 Honths (28) none Drywell High Pressure (PIS-64-57A-D)

Instrument Channel (1) (27) Once/6 Honths (28) none Reactor Low Pressure (PIS-3-74A&B, PS-3-74ASB)

(PIS-68-95, PS-68-95)

(PIS-68-96, PS-68-96) rn C0

TABLE 4.2.B (Continued)

SURVEILLANCE RE()UIREHENTS FOR INSTRUHENTATION THAT INITIATE OR CONTROL THE CSCS Fn 'n Fun ion 1 T alibr ion n rum en h k Core Spray Auto Sequencing (4) Once/operating cycle none Timers (Normal Power)

Core Spray Auto'equencing (4) Once/operating cycle none Timers (Diesel Power)

LPCI Auto Sequencing Timers (4) Once/operating cycle none (Normal Power)

LPCI Auto Sequencing Timers (4) Once/operating cycle none (Diesel Power)

RHRSM Al, B3, Cl, D3 Timers (4) Once/operating cycle none (Normal Power)

RHRSW Al, B3, Cl, D3 Timers (4) Once/operating cycle none (Diesel Power)

ADS Timer (105 sec.) .(4) Once/operating cycle none ADS Timer (12 1/2 min.) (4) Once/operating cycle none (High Drywell Pressure Bypass Timer)

OTES OR ES 4 2 OUGH 4 2 e ce t 4.2 D ND 4 (Cont'd)

26. This instrument check consists of comparing the background signal levels for all valves for consistency and for nominal expected values (not required during refueling outages).

27. Functional test consists of the injection of a simulated signal into the electronic trip circuitry in place of the sensor signal to verify operability of the trip and alarm functions.

28. Calibration consists of the adjustment of the primary sensor and associated components so that they correspond within acceptable range and accuracy to known values of the parameter which the channel monitors, including adjustment of the electronic trip circuitry, so that its output relay changes state at or more conservatively than the analog equivalent of the trip level setting.

29. The functional test frequency decreased to once/3 months to reduce challenges to relief valves per NUREG-0737, Item II.K.3.16.

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

31. Functional Tests shall be performed once/3 months.

BFN 3.2/4.2-61 Unit 2

THIS PAGE INTENTIONALLYLEFT BLANK BFN 3.2/4.2-6la Unit 2

ENCLOSURE 2

SUMMARY

OF CHA GES

1. Revision to Table 3.2.B Delete "Instrument Channel Reactor Low Pressure" from the table.

2. Revision to Table 4.2.A (Instrument Channel Reactor High Pressure)

a. Add pressure switch numbers (PS-68-93 R 94).

b. Revise functional test note to (31).

c. Revise calibration frequency to once/18 months.

3 ~ Revision to Table 4.2.B Delete "Instrument Channel Reactor Low Pressure" from the table.

4~ Revision to notes for Table 4.2.A Add note 31 which reads as follows:

"31. Functional tests shall be performed once/3 months."

ENCLOSURE 3 Reasons and Justification For Changes Reasons Fo C es Installed pressure switches 2-PS-68-93 and 2-PS-68-94 are non-class.lE and have an adjustable setpoint range from 50 to 1200 psig. This 'wide pressure range and the excessive drift associated with the existing switches. causes unacceptable accuracy in the lower pressure ranges. Each switch contains two independently actuated microswitches which have separate, independent functions (SW Pfl, SW PP2). These microswitches open and close at a setpoint of 100 g 15 psig.

To resolve the problem of inadequate switch accuracy and excessive" drift, these pressure switches will be replaced with Class lE, Static-0-Ring (SOR) pressure switches with a setpoint range of 20 to 180 psig and an'accuracy of 1.0 percent of upper range limit. This gives better accuracy and; better stability near the 100 psig setpoint. The SOR pressure switches=-.contain only a single microswitch (there are no Class 1E pressure switches which contain two independently actuated microswitches).

To accommodate the use of the SOR pressure switch, one set of the presently used contacts must be deleted. This can be accomplished by deleting SW gl from the logic since it is redundant to the'limit switches of Residual Heat Removal (RHR) inboard/outboard isolation valves 2-FCV-'4-47 and 2-FCV-74-48.

SWSl contacts provide a low pressure permissive for isolation valves 2-FCV-74-53 and 2-FCV-74-67; which is not necessary since the isolation logic to these valves, is .also provided, whenever valves 2-FCV-74-47 and 2-FCV-74-48 are in the open position. Valves 2-FCV-74-47 and 2-FCV-74-48 cannot be open unless reactor pressure is below 100 psig.

The only safety function performed by switches 2-PS-68-93 and 2-PS-68-94 is the high pressure isolation signai to close the RHR inboard/outboard valves 2-FCV-74-47 and 2-FCV-74-48 at 100 psig. This function is accomplished by SW82.

This technical specification change revises Table 3.2.B and Table 4.2.B to take out references to the function of the deleted microswitch (SW81). Table 4.2.A is being revised to reflect the functional testing and calibration requirements for the new SOR pressure switches.

(Enclosure 3 cone.)

4 Page 2 of 2 JUSTIF CA ON OR THE CHANGES Pressure switches 2-PS-68-93 and 2-PS-68-94 currently contain two microswitches each (SW //1, SW /P2).

One microswitch (SW //2) closes on increasing reactor pressure at a setpoint of 100 + 15 psig. The function of this signal is to protect the shutdown cooling system piping and components from overpressurization. This function is accomplished by automatically closing inboard/outboard PHR isolation valves 2-FCV-74-47 and 2-FCV-74-48 when the reactor vessel pressure aoproaches the shutdown cooling piping design pressure. The maximum allowable pressure is 133 psig.

The other microswitch (SW /P1) closes, on decreasing reactor pressure at a setpoint of 100 + 15 psig. This microswitch provides a low pressure permissive signal to the isolation logic for RHR valves 2-FCV-74-53 and 2-FCV-74-67. A low pressure signal from either 2-PS-68-93 or 2-PS-68-94, and a signal indicating that both RHR shutdown cooling. suction isolation valves 2-FCV-74-47 and 2-FCV-74-48 are not fully closed, and a containment'isolation signal will cause closure of valves 2-FCV-74-53 and 2-FCV-74-67.

In order to accommodate use of the available class lE pressure switches, one set of the presently used contacts must be deleted. This will be accomplished by deleting SW 81 from the logic since it is redundant to the limit switches of valves 2-FCV-74-47 and 2-FCV-74-48. The contacts provide a low reactor pressure permissive for isolation valves~2-FCV-74-53 and 2-FCV-74-67, which is not necessary since the permissive signal'o these valves is also provided whenever valves 2-FCV-74-47 and 2-FCV-74-48"are in the open position. These-valves, 2-FCV-74-47 'and 2-FCV-74-48, cannot be open unless reactor pressure is below 100 psig. The emoval of SW ill in effect provides a permanent low pressure permissive since the contacts are replaced with a jumper.

The only remaining safety function performed by switches 2-PS-68-93 and 2-PS-68-94 is the high pressure isolation signal to close the RHR-inboard/outboard isolation valves 2-FCV-74-47 and 2-FCV-74-48. This function is not arfected by .this technical- specification cnange. The calibration frequencies for the new pressure switches are based on the manufacturer's recommendation and are supported by calculations.

0 hl ENCLOSURE 4 Proposed Determination of No S igni ficant Hazards Consideration DESCR PT ON OF PROPOSED T CHNICAL SPECIFICAT ON AMENDMENT The Browns Ferry unit 2 technical specifications are being revised as follows: (1) Delete. references to the function "Instrument Channel-Reactor Low Pressure" from Tables 3.2.B and 4.2.B and (2) incorporate revised functional. testing and calibration frequencies for replacement pressure switches PS-68-93 and 94 in Table 4.2.A.

BASIS FOR PROPOS D 0 S GN FICA H ZARDS CONSIDERAT ON DETERMINATION NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating, license involves no significant hazards consideration if. operation of the facility in accordance with the proposed amendment would not 1) involve a significant increase in the probability or consequences of an accident previously evaluated, or 2) create the possibility of a new or different J'ind of accident from any accident previously evaluated, or 3) involve 'a significant reduction in a margin of safety.

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The existing non-class lE pressure switches (2-PS-68-93 and 94) are being replaced by class 1E pressure switches to resolve the problems of-inadequate pressure switch accuracy and excessive drift. The existing pressure switches contain two internal microswitches (SV/P1, SM/$ 2) whereas the replacement pressure switches contain one internal microswitch. As a result, the function oi SWlkl, which is to provide a low pressure permissive signal to the isolation logic for RHR valves 2-FCV-74-53 and 2-FCV-74-67, is being deleted from Tables 3.2.B and 4.2.B by this change.'his function is redundant to the limit switches on RHR valves 2-FCV-74-47 and 2-'FCV-74-48. As such, it is not required nor was it considered in the FSAR analysis. Changes are also being made to table 4.2.A to reflect the revised functional testing and calibration requirements for the new pressure switches.

No new failure modes have been identified for the proposed changes.

Misoperation of the replacement pressure switches could not cause the initiation of any accident previously evaluated in plant Safety Analysis Report (SAR). Further, the replacement pressure switches do not require relocation, do not adversely affect system function or operations, and do not adversely affect other systems or components. Therefore, this change will not significantly increase the probability of occurrence or consequences of any accident previously evaluated in the SAR.

P p ~

l~

-(Enclosure 4 cont.)

2 ~ The proposed change does not create the possibility of a new or different kind of accident from any accident previously analyzed. The function and operation of the affected systems are not changed by the amendment.

Seismic qualification of the affected components remain intact due to this modification and other systems will not be adversely affected.

Operation and failure modes of the replacement switches can cause no diffezent effects than the existing switches. Thus, the credible failure modes of the replacement pressure switches would be bounded by existing FSAR Section 14.6.3.3.2 accident analysis. Therefore, this modification will not create the possibility of an accident of a different type than any previously evaluated in the SAR.

3. The Proposed change does not involve a significant reduction in a margin of safety.

This change replaces the existing non-Class lE pressure switches with Class lE pressure switches which are more accurate.

In addition, one of the two contacts from each pressure switch will be removed from the cuzrent valve control logic . This contact was.

redundant to other logic which controls these valves and is not required for proper operation of any logic required for Technical Specification compliance.

The margin of safety defined by the bases for Technical Specifications 3.2.A/4.2.A (Primary Containment and Reactor Building Isolation Functions) and 3.2.3/4.2.B (Core and Containment Cooling Initiation Ec Control) is not reduced by this modification. This modification results in increased instrument accuracy and a reduction of failure modes caused, by the deletion of redundant contacts.

Jp V