Text

Amends 54 & 46 to Licenses DPR-77 & DPR-79,respectively, Changing Tech Specs to Add Requirement for Demonstrating Operability of Shunt Trip Attachment to Reactor Trip Breaker & Incorporating Testing Requirements
	ML20204K066
Person / Time
Site:	Sequoyah
Issue date:	03/16/1987
From:	Youngblood B; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20204K068	List: ML20204K066; ML20204K098;
References
	NUDOCS 8703270358
	Download: ML20204K066 (23)
	v • d • e

._

4

'o*t, UNITED STATES NUCLEAR REGULATORY COMMISSION g

.,i WASHINGTON, D. C. 20555

...../

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SE0VOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No. DPR-77 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment to the Sequoyah Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-77 filed by the 4

Tennessee Valley Authority (licensee), dated January 20, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the license, as amended,.

the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Appendix A

+

Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

8703270358 870316 DR p

ADOCK 050003g7 PDR

.,g C-r.-y h.

x s.

.3

)

2' '

, (2)1 Technical Specifications-The, Technical. Specifications ~ sootained in Appendix A, as revised through J

Amendment No.'.54 are hereby ' incorporated into the license. The licensee shall operate the facility in accordance with the Technical ' Specifications.

~3.

This license amendment is effec $1ve:as of its date of issuance, S,

FbR THE NUC1'Ef R' REGULATORY COMMISSION E

- \\

B..J. Youngblogd, Director PWR Project Directorate #4

's Division of PWR Licensi,q-A Attachment" Appendix A Technical Specification Changes f,

Date of Issuance: March 16, 1987 E

q-a l'

m A

/

x, 1

l-

\\j PWR#4/ WR-A PWR#/6PWR-A OG -B

',. PW MD R-A Mbudt'An@7 rad JHo,loriich BJ u'ng ood

/87 03/g/07 03

/87 03/4/87 03/

0 puuu>~.

Of 4

5&3 s

W-T

~

}--

, ;J a ^

y, S

OM.i

+

p

,y....

ATTACHMENT TO LICENSE AMENDMENT NO. 54 ~

1" FACILITY OPERATING LICENSE NO. DPR-77

['

DOCKET N0. 50-327 i.

Replace the followin the enclosad pages. g pages of the Appendix "A" Technical Specifications with The revised pages are identitied b contain vertical lines indicating the areas of change. y Amendment number and 4

Awnded Pages 4

.I' 3/4 3-4 5

u.

^

3/4 3-7 3/4 3-8

- i '

3/4 3-11 3/4 3-12 3/4 3-13 B 3/4 3-2 s

I 1

l s

l

~.

,t, I

t

\\

"we

'M c

4

i :-ef

,~

TABLE 3.3-1 (Continued) t REACTOR TRIP SYSTEM INSTRUMENTATION e,

a E

' MINIMUM e

TOTAL NO.

CHANNELS CHANNELS APPLICABLE-g

- FUNCTIONAL UNIT OF ' CliANNELS TO TRIP OPERABLE MODES

_ ACTION

[

19.- ' Safety Injection Input

'from ESF 2

1 2

1, 2 12 20.

Reactor Trip Breakers 2

1 2

1, 2, and *.12, 15 A.

Startup and Power Operation B.

Shutdown 2

0 1

3*, 4* and 5*

16 21.

Automatic Trip Logic 2

1 2

1, 2, and

12 w

A.

Startup and Power Operation i

B.

Shutdown 2

0 1

3*, 4* and 5*

16 22.

Reactor Trip Syr - Interlocks I

A.

Intermedi.a Ge +>

I Neutron Flux P-6 2

1 2

2, and*

8a

'l B.

Power Range Neutron Flux - P-7 4

2 3

1 8b C.

Power Range Neutron

_g Flux - P-8 4

2 3

1 8c

~

D.

Power Range Neutron Flux - P-10 4

2 3

1, 2 8d

{

E.

Turbine Impulse Chamber f'

i a

Pressure - P-13 2

1.: -

2 1

8b

{

F.

Power Range Neutron a

Flux - P-9 4

2 3

1 8e j

[

G.

Reactor Trip - P-4 2

1 2

1, 2, and* 14 y

'~'

TABLE 3.3-1 (Continued)

ACTION 8 -

With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-priate ACTION statement (s) for these functions.

Functions to

'be evaluated are:

a.

Source Range Reactor Trip b.

Reactor Trip Low Reactor Coolant Loop Flow (2 loops)

Undervoltage Underfrequency.

Pressurizer Low Pressure Pressurizer High Level c.

Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d.

Reactor Trip Intermediate Range Low Power Range Source Range e.

Reactor Trip Turbine Trip ACTION 9 -

Deleted ACTION 10 -

Deleted ACTION 11 -

Deleted l

ACTION 12 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for i

up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 l

provided the other channel is OPERABLE.

l l

SEQUOYAH - UNIT 1 3/4 3-7 Amendment No. 54 l

TABLE 3.3-1 (Continued)

ACTION 13 -

~With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above the P-7 (Block of Low Power Reactor Trips) setpoint, place the inoperable channel in the tripped condition within 6_ hours, operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTION 14 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 15 -

With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 12.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for performing maintenance to restore the breaker to OPERABLE status.

ACTION 16 -

With the number of OPERABLE channels one less than the minimum channels operable requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

i SEQUOYAH - UNIT 1 3/4 3-8 Amendment No. 54

_ ~, _ _ - -. _ _ _

TABLE 4.3-1 j

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8

5z

. CHANNEL MODES IN WHICH e

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED Z

1.

Manual Reactor Trip N.A.

N.A.

S/U(1) and R(9) 1, 2, and

g 2.

Power Range, Neutron Flux S

D(2), M(3)

Q 1, 2 and Q(6) 3.

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 High Positive Rate 4.

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 High Negative Rate 5.

Intermediate Range, S

R(6)

S/U(1) 1, 2, and

y Neutron Flux

[

6.

Source Range, Neutron Flux S(7)

R(6)

M and S/U(1) 2, 3, 4, 5, and

h 7.

Overtemperature Delta T S

R M

1, 2 l

8.

Overpower Delta T S

R M

1, 2 9.

Pressurizer Pressure--Low S

R Q

1, 2 10.

Pressurizer Pressure--High S

R Q

1, 2 2

1 11.

Pressurizer Water Level--High S

R Q

1, 2 12.

Loss of Flow - Single Loop S

R Q

1 1

13.

Loss of Flow - Two Loops S

R N.A.

1 I

s 14.

Main Steam Generator Water S

R Q

1, 2

'l

[

Level--Low-Low 5

Y'

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS S

CHANNEL MODES IN WHICH j

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E

15.

Steam /Feedwater Flow Mismatch and S

R Q

1, 2 Low Steam Generator Water Level Eq 16.

Undervoltage - Reactor Coolant N.A.

R M

1-Pumps g

17.

Underfrequency - Reactor Coolant N.A.

R M

1 4

Pumps 18.

Turbine Trip A.

Low Fluid Oil Pressure N.A.

N.A.

S/U(1) 1 B.

Turbine Stop Valve Closure N.A.

N.A.

S/U(1) 1 19.

Safety Injection Input from ESF N.A.

N.A.

M(4) 1, 2 20.

Reactor Trip Breaker N.A.

N.A.

M(5) and S/U(1) 1, 2, and

4

{

21.

Automatic Trip Logic N.A.

N.A.

M(5) 1, 2, and

l w

22.

Reactor Trip System Interlocks O

A.

Intermediate Range N.A.

R S/U(8) 2, and

Neutron Flux, P-6 B.

Power Range Neutron N.A.

R S/U(8) 1 Flux, P-7 i

C.

Power Range Neutron N.A.

R S/U(8) 1 Flux, P-8 i

D.

Power Range Neutron N.A.

R S/U(8) 1, 2 Flux, P-10 E.

Turbine Impulse Chamber N.A.

R S/U(8) 1 1

Pressure, P-13

{

F.

Power Range Neutron N.A.

R S/U(8) 1 Flux, P-9 s

[

G.

Reactor Trip, P-4 N.A.

R S/U(8) 1, 2, and

23.

Reactor Trip Bypass Breaker N.A.

N.A.

M(10)R(11) 1, 2, and

w b

1 e

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1)

If not performed in previous 7 days.

(2)

Heat balance only, above 15% of RATED THERMAL POWER.

Adjust channel if absolute difference greater than 2 percent.

(3) -

Compare incore to excore axial flux difference above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference greater than or equal to 3 percent.

(4)

Manual ESF functional input check every 18 months.

(5)

Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

The test shall independently verify the operability of the undervoltage and auto.3atic shunt trip circuits.

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8)

Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.

(9)

The channel function test shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function.

(10) -

Local manual shunt trip prior to placing breaker in service.

Each train shall be tested at least every 62 days on a staggered test basis.

s

!1 (11) -

Automatic and manual undervoltage trip.

l SEQUOYAH - UNIT 1 3/4 3-13 Amendment No.

54

)

INSTRUMENTATION l BASES 3

i-2 The measurement of response time at the specified frequencies provides I

assurance that the protective and ESF action function. associated with each' channel is completed within the time' limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response. times indicated as not applicable.

2

._ Response time may be demonstrated by any series of sequential, overlapping 7

or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2). utilizing replacement-sensors with certified response times.

Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing maintenance. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on a Westinghouse analysis performed.in WCAP-10271, Supplement 1, which determines bypass breaker availability.

t.

I.

3/4.3.3 MONITORING INSTRUMENTATION

. 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION I

-The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation ~ levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 MOVABLE INCORE DETECTORS i

The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of j

this system accurately represent the spatial neutron flux distributio.n of the reactor core.

The OPERABILITY of this system is demonstrated by irradiating each detect'or used and determining the acceptability of its voltage curve.

ForthepurposeofmeasuringF(Z)orF5Hafullincorefluxmapisused.

q Quarter core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

1 3/4.3.3.3 SEISMIC INSTRUMENTATION i

The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the i

SEQUOYAH - UNIT 1 8 3/4 3-2 Amendment No.

54 i

[

'o,,

UNITED STATES

.E NUCLEAR REGULATORY COMMISSION o

]

E wAsmNGTON, D. C. 20656

\\...../

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SE000YAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. OPR-79 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment to the Sequoyah Nuclear Plant, Unit 2 (the facility) Facility Operating License No. DPR-79 filed by the Tennessee Valley Authority (licensee), dated January 20, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety.of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by page changes to the Appendix A

(

Technical Specifications as indicated in the attachment to this license L

amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:

l-i 4-r-----,-e---

er-----* - - - - - - -

e-~- - - - - - + - + - - - - - - - " " ' - - - ' ' ~ ' " ' " " " " ~ * " " " ' ' ' " "

. _ =..

(2) Technical' Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 46 are hereby incorporated into the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 6\\

B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A Attachment Appendix A Technical Specification Changes Date of Issuance:

March 16, 1987 h

nl su, c/

1 M(Duncan/radPRygpFWR-A PWR#4/DPWR-A OGC-Beth, PW # ADE R-A JHplori'ich SETtATL BJ pungblood 03/2 /87 03/4/87 03/9 /87 03 % /87 bpo#

JW*

ATTACHMENT TO LICENSE AMENDMENT NO. 46 FACILITY OPERATING LICENSE N0. DPR-79 DOCKET NO. 50-328 Replace the followin the enclosed pages. g pages of the Appendix "A" Technical Specifications with The revised pages are identified b contain vertical lines indicating the areas of change. y Amendment number and Amended Page 3/4 3-4 3/4 3-7 3/4 3-8 3/4 3-11 3/4 3-12 3/4 3-13 B 3/4 3-2 B 3/4 3-3 B 3/4 3-4 B 3/4 3-5 J

lp: ', ;

e l

i.'

e I

l

- ~ ~

- -~ ~ ~ ~

~-

,~

TABLE 3.3-1 (Continued)

Mg REACTOR TRIP SYSTEM INSTRUMENTATION 5!g MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION c5 19.

Safety Injection Input N

from ESF 2

1 2

1, 2 12 20.

Reactor Trip Breakers 2

1 2

1, 2, and

12,15 A.

Startup and Power Operation B.

Shutdown 2

0 1

3* 4* and 5*

16

21. Automatic Trip Logic 2

1 2

1, 2, and

12 A.

Startup and Power Operation B.

Shutdown 2

0 1

3* 4* and 5*

16 22.

Reactor Trip System Interlocks A.

Intermediate Range A

Neutron Flux, P-6 2

1 2

2, and*

8a B.

Power Range Neutron Flux, P-7 4

2 3

1 8b C.

Power Range Neutron Flux, P-8 4

2 3

1 8c D.

Power Range Neutron Flux, P-10 4

2 3

1, 2 8d E.

Turbine Impulse Chamber Pressure, P-13 2

1 2

1 8b y

B

[

F.

Reactor Trip, P-4 2

1 2

1, 2, and

14 5

?.

F

TABLE 3.3-1 (Continued)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the appro-

.priate ACTION statement (s) for those functions.

Functions to be evaluated are:

a.

Source Range Reactor Trip.

b.

Reactor Trip Low Reactor Coolant Loop Flow ('2 loops)

Undervoltage Underfrequency Turbine Trip Pressurizer Low Pressure Pressurizer High Level c.

Reactor Trip Low Reactor Coolant Loop Flow (1 loop) d.

Reactor Trip Intermediate Range Low Power Range Source Range ACTION 9 - Deleted ACTION 10 - Deleted ACTION 11 - Deleted ACTION 12 - With the nutber of OPERABLE channels one less than required by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1 provided the other channel is OPERABLE.

i l

SEQUOYAH - UNIT 2 3/4 3-7 Amendment No. 46 l

R

~

TABLE 3.3-1 (Continued)

ACTION 13 - With the number of'0PERABLE channels one less than the Total Number of Channels and with the THERMAL' POWER level above the P-7 (enable reactor trips) setpoint place-the inoperable channel in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , operation may continue until performance of the next required CHANNEL' FUNCTIONAL TEST.

ACTION 14 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in at-least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 15 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to operable-status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 12.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except_for up to 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s-for performing maintenance to restore the breaker to OPERABLE status.

ACTION 16 - With the number of OPERABLE channels one less than the minimum channels operable requirement, restore the inoperable-channel to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

I i

t i

i I

i SEQUOYAH - UNIT 2 3/4 3-8 Amendment No. 46 I

=.

n.

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS S

CHANNEL MODES FOR WHICH l

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS I

E FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED l

N 1.

Manual Reactor Trip N.A.

N.A.

S/U(1)and R(9)-

1, 2, and

2.

Power Range, Neutron Flux S

D(2),M(3)

Q 1, 2 and Q(6) i 3.

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 High Positive Rate

4..

Power Range, Neutron Flux, N.A.

R(6)

Q 1, 2 i

High Negative Rate

,s

[

5.

Intermediate Range, S

R(6)

S/U(1) 1, 2, and

4 Neutron Flux l

6.

Source Range, Neutron Flux S(7)

R(6)

M and S/U(1) 2, 3, 4, 5, and

7.

Overtemperature AT S

R M

1, 2 7

l 8.

Overpower AT S

R M

1, 2 9.

Pressurizer Pressure--Low S

R Q

1, 2 10.

Pressurizer Pressure--High S

R Q

1, 2

]

11.

Pressurizer Water Level--High 5

R Q

1, 2 12.

Loss of Flow - Single Loop S

R Q

1 i

13.

Loss of Flow - Two Loops S

R N.A.

1 E

14. Steam Generator Water Level--

S R

Q 1, 2 1

Low-Low i

e

]

m 4

1 TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E5 CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS 7

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRE 0 g

15. Steam /Feedwater Flow Mismatch and S

R Q

1, 2 Low Steam Generator Water Level m

16. Undervoltage - Reactor Coolant N.A.

R M

1 Pumps

17. Underfrequency - Reactor Coolant N.A.

R M

1 Pumps 18.

Turbine Trip A.

Low Fluid Oil Pressure N. A.

N.A.

S/U(1) 1 B.

Turbine Stop Valve Closure N.A.

N.A.

S/U(1) 1 g

19.

Safety Injection Input from ESF N.A.

N.A.

M(4) 1, 2

[

20.

Reactor Trip Breaker N.A.

N.A.

M(5) and S/U(1) 1, 2, and *

21. Automatic Trip Logic N.A.

N.A.

M(5) 1, 2, and

22.

Reactor Trip System Interlocks A.

Intermediate Range N.A R

S/U (8) 2, and

Neutron Flux, P-6 B.

Power Range Neutron N.A.

R S/U (8) 1 Flux, P-7 C.

Power Range Neutron N.A.

R S/U (8) 1 Flux, P-8 D.

Power Range Neutron N.A.

R S/U (8) 1, 2 Flux, P-10 E.

Turbine Impulse Chamber N.A.

R S/U (8) 1 Pressure, P-13 E

F.

Power Range Neutron E

Flux, P-9 N.A.

R S/U (8) 1

[

G.

Reactor Trip, P-4 N.A.

R S/U (8) 1, 2, and

3 23.

Reactor Trip Bypass Breaker N.A.

N.A.

M(10)R(11) 1, 2, and

o

( _

TABLE 4.3-1 (Continued)

NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1)

If not performed in previous 7 days.

(2)

Heat balance only, above 15% of RATED THERMAL POWER.

Adjust channel if absolute difference greater than 2 percent.

(3)

Compare incore to excore axial flux difference above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference greater than or equal to 3 percent.

(4)

Manual ESF functional input check every 18 months.

(5)

Each train or logic channel shall be tested at least every 62 days on a ST/GGERED TEST BASIS.

The test shall independently verify the operability of the undervoltage and automatic shunt trip circuits.

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8)

Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.

(9)

The channel function test shall independently verify the operability of the undervoltage and shunt trip circuits for the manual reactor trip function.

(10) -

Local manual shunt trip prior to placing breaker in service.

Each train shall be tested at least every 62 days on a staggered test basis.

(11) -

Automatic and manual undervoltage trip.

SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46

INSTRUMENTATION BASES REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the protective and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows-the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing maintenance.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on a Westinghouse analysis performed in WCAP-10271, Supplement 1, which determines bypass breaker availability.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equi? ment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

N For the purpose of measuring F (Z) or F a full incore flux map is used.

q 3H Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100.

The instrumentation is consistent with the

.I SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment No. 46

m INSTRUMENTATION BASES 3/4 3.3.3 SEISMIC INTRUMENTATION (Continued) recommendations of Regulatory Guide 1.12. " Instrumentation for Earthquakes,"

April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data ia available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "0nsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Apinst an Accidental Chlorine Release," February 1975.

<l 3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that l

sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975.

Sequoyah has four separate methods of determining safety valve position (i.e.,

open or closed).

l a.

Acoustic flow monitors mounted on each safety valve line (one per valve).

A flow indicating module in the main control room is calibrated to detect failure of a valve to reclose.

An alarm in the main control room will l

actuate when any valve is not fully closed.

i SEQUOYAH - UNIT 2 B 3/4 3-3 Amendment No. 46 i

+,

.=

e

.c h

- INSTRUMENTATION E

BASES ~

3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION (continued) b.

. Temperature sensors downstream of each safety valve (one per valve).

Tem-

-perature indication.and alarm are provided in the main control room.

Pressurizer relief tank temperature, pressure and level indication, and c.

1

. alarm in main control room.

i

..d.

Pressurizer pressure indication and alarm in the main control room.

i Although all the above position indicators for the pressurizer safety valves

.and the.PORVs are acceptable as one of the channels, the acoustic monitors must be one of the two required operable chanels.

In addition to the four methods described above, the PORVs use an electromagnetic " reed"-switch to determine 2

valve position.

The stem mounted switches are no longer in use since the PORVs were changed.

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION 1

OPERABILITY of the fire detection instrumentation ensures that adequate i

warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the potential for damage to-safety related equipment and is an integral' element in the overall-facility

{

fire protection program.

i' In the event that a portion of the fire detection-instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas i

is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.9 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and controT, as applicable, the releases of radioactive materials in liquid i

effuents during actual or potential releases of liquid effluents.

The alarm /

trip setpoints for these instruments shall be calculated in accordance with the procedures in the 00CM to ensure that the alarm / trip will occur prior to i

exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this l

instrumentation is consistent with the requirements' of General Design l

Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

j 3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLUENT INSTRUMENTATION l

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip setpoints for these instruments shall be calculated in accordance l

with the procedures in the ODCM to ensure that the alarm / trip will occur prior 4

i SEQUOYAH - UNIT 2 B 3/4 3-4 Amendment No. 46 1

r INSTRUMENTATION BASES 3/4.3.3.10 RADI0 ACTIVE GASEOUS EFFLUENT INSTRUMENTATION (Continued) to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 cf Appendix A to 10 CFR Part 50.

l l

i l

SEQUOYAH - UNIT 2 B 3/4 3-5 Amendment No. 46 I

ML20204K066

Text

Navigation menu

Search