Text

Amend 193 to License DPR-49,revising TS by Incorporating Extended Allowable out-of-svc Times & Surveillance Test Intervals for Rps,Isolation Actuation Sys,Eccs & Control Rod Block Function Instrumentation
	ML20056H404
Person / Time
Site:	Duane Arnold
Issue date:	04/14/1993
From:	Pulsifer R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20056H405	List: ML20056H404; ML20056H409;
References
	NUDOCS 9309090274
	Download: ML20056H404 (73)
	v • d • e

-

p@ ** cug

[. i l'%

UNITED STATES j

NUCLEAR REGULATORY COMMISSION p

g W ASHINGToN. D.C. 2055Fr0001 g

% ~...+ j IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER

{

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 193 License No. DPR-49 1.

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

A.

The application for amendment by Iowa Electric Light and Power Company, et al., dated December 19, 1991 and supplemented by addi-tional information on August 25, October 8, and November 24, 1992, and March 8, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Commission'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 3

51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:

9309090274 930814 "

PDR ADOCK 05000331 i

P PDR;

i i

(2) Technical Specifications i

The Technical Specifications contained in Appendix A, as revised l

through Amendment No.193, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of the date of issuance and shall be implemented within 180 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

,.bertM.Pulsifr,ProjectManager o

Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: April 14,1993 P

6 9

l f

v f

i

U 1

r 1

ATTACHMENT TO LICENSE AMENDMENT NO.

j I

FACILITY OPERATING LICENSE NO. DRP-49 DOCKET NO. 50-331 i

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages The revised areas are indicated by marginal lines.

Remove Insert i

i v

v 4

1.0-10*

3.1-1

- 3.1-4 3.1 3.1-4 3.1-4a 3.1 3.1-7 3.1 3.1-7 3.1-7a i

3.1 3.1-17 3.1 3.1-17 t

3.1 3.1-30 i

3.2-1

- 3.2-2 3.2-1

- 3.2-2 7

3.2-2a 3.2-3

- 3.2-4 3.2-3

- 3.2-4 3.2-4a 3.2-5 3.2-5 3.2-Sa 3.2 3.2-14 3.2 3.2-14 3.2-14a 3.2 3.2-23 3.2 3.2-23 3.2-23a - 3.2-23d 3.2 3.2-34 3.2 3.2-34 3.2-34a 3.2 3.2-36 3.2 3.2-36 I

3.2-36a i

3.2 3.2-45 3.2 3.2-45 3.2-45a - 3.2-45b 3.2 3.2-50 3.2 3.2-50 i

Denotes new page l

4 a m r,--

,s m.,,-

e-

DAEC-1 TECHNICAL SPECIFICATIONS 3

TABLE OF CONTENTS i

PAGE NO.

j 2.0 Definitions 1.0-1 FAFETY LIMITS LIMITING SAFETY SYSTEM SETTING 1.1 Fuel Cladding Integrity 2.1 1.1-1 1.2 Reactor Coolant System Integrity 2.2 1.2-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE j

REOUIREMENTS l 3.1 Reactor Protection System Instrumentation 4.1 3.1-1 3.2 Protective Instrumentation 4.2 3.2-1 A.

Isolation Actuation Instrumentation A

3.2-1 B.

Core and Containment Cooling Systems Initiation / Control Instrumentation B

3.2-11 C.

Control Rod Block Instrumentation C

3.2-20 D.

Radiation Monitoring Instrumentation D

3.2-25 l

E.

Drywell Leak Detection Instrumentation E

3.2-28 F.

Surveillance Instrumentation F

3.2-31 L

G.

Recirculation Pump Trip (RPT) and Alternate Rod Insertion (ARI) Instrumentation G

3.2-34 H.

Accident Monitoring Instrumentation H

3.2-37 I.

Explosive Gas Monitoring Instrumentation I

3.2-42 4

3.3 Reactivity control 4.3 3 3-1 i

A.

Reactivity Limitations A

3.3-1 B.

deram Discharge Volume B

3.3-3 C.

Reactivity Control Systems C

3.3-4 l

r D.

Scram Insertion Times D

3.3-5 3

E.

Reactivity Anomalies E

3.3-6 e

F.

Recirculation Pumps F

3.3-6 3.4 Standby Liquid control System 4.4 3.4-1 A.

Normal System Availability A

3.4-1 B.

Operation with Inoperable Components B

3.4-2 I

C.

Sodium Pentaborate Solution C

3.4-2 f

[

3.5 Core and Containment Cooling Systems 4.5 3.5-1 A.

Core Spray and LPCI Sabsystems A

3.5-1 B.

Containment Spray Cooling Capability B

3.5-4 Amendment No.

17A,7E0/7E#,193 i

DAEC-1 TECHNICAL SPECIFICATIONS LIST OF TABLES Table Number Title PJLq1 1.0-1 Operating Modes 1.0-10 3.1-1 Reactor Protection System Instrumentation 3.1-3 3.1-2 Protective Instrumentation Response Times 3.1-7 4.1-1 Reactor Protection System Instrumentation Surveillance Requirements 3.1-8 4.1-2 Deleted 3.2-A Isolation Actuation Instrumentation 3.2-3 4.2-A Isolation Actuation Instrumentation Surveillance Requirements 3.2-8 3.2-B Core and Containment Cooling Systems Initiation / control Instrumentation 3.2-12 4.2-B Core and Containment Cooling Systems Initiation / Control Surveillance Requirements 3.2-17 t

3.2-C Control Rod Block Instrumentation 3.2-21 4.2-C Control Rod Block Instrumentation Surveillance i

Requirements 3.2-23 3.2-D Radiation Monitoring Instrumentation 3.2-26 4.2-D Radiation Monitoring Instrumentation Surveillance

~i Requirements 3.2-27 3.2-I Drywell Leak Detection Instrumentation 3.2-29 4.2-E Drywell Leak Detection Instrumentation Surveillance Requirements 3.2-30 3.2-F Surveillance Instrumentation 3.2-32 4.2-F Surveillance Instrumentation Surveillance Requirements,3.2-33 3.2-G (ATWS) RPT/ARI and EOC-RPT Instrumentation 3.2-35 4.2-G (ATWS) RPT/ARI and EOC-RPT Instrumentation Surveillance Requirements 3.2-36 j

3.2-H Accident Monitoring Instrumentation 3.2-38 I

4.2-H Accident Monitoring Instrumentation Surveillance Requirements 3.2-41 i

i i

Amendment No. 17#,757,170,193 v

i

-, s.

~

i LAEC-1 l-!

(

TABLE 1.0-1 i

i OPERATING MODES l

i

)

REACTOR MODE SWITCH AVERAGE REACTOR COOLANT I

OPERATING MODE POSITION TEMPERATURE j

f 1.

RUN/ POWER OPERATION Run NA i

2.

STARTUP Startup/ Hot Standby or NA Refuel i

3.

HOT SHUTDOWN #

S hu t d own*"d'

> 212*F l

i 4.

COLD SHUTDOWN

Shutdown"*d""

5 212*F 5.

REFUELING

Shutdown or Ref uel""

NA f

i i

l (a) Fuel in the reactor vessel with the reactor vessel head closure bolts

[

fully tensioned.

+

(b) Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

t 5

(c) The reactor mode switch may be placed in the Run, Startup/ Hot Standby or Refuel position to test the switch interlock functions and related i

instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator.

l i

(d) The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.

t (e) The reactor mode switch may be placed in the Refuel position while a f

single control rod drive is being removed from the reactor pressure i

vessel per Specification 3.9.A.

(f) The reactor mode switch may be placed.in the Startup position for demonstration of shutdown margin per Specification 4.3.A.1.

j 1

l 1

I Amendment No.

193 1.o-lo t

i 4

l T

f y

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 PEACTOR PROTECTION SYSTEM INSTRUMENTATION INSTRUMENTATION A.

As a minimum, the reactor A.1 Each reactor protection system protection system instrumentation instrumentation channel shall be channels shown in Table 3.1-1 demonstrated OPERABLE by the shall be OPERABLE with the performance of the CHANNEL CHECK, PROTECTIVE INSTRUMENTATION CHANNEL FUNCTIONAL TEST and CHANNEL RESPONSE TIME as shown in Table CALIBRATION operations for the t

3.1-2.

OPERATING MODES and at the I

frequencies shown in Table 4.1-1.

l The designed system response times from the opening of the 2.

Response time measurements (from i

sensor contact up to and actuation of sensor contacts or trip including the opening of the trip point to de-energization of scram actuator contacts shall not solenoid relay) are not part of the exceed 50 milliseconds.

normal instrument calibration. The reactor trip system response time of l

Arolicability:

each reactor trip function shall be demonstrated to be within its limit r

As shown in Table 3.1-1.

once per operating cycle. Each test shall include at least one logic Action:

train such that both logic trains are tested at least once per 36 1.

With one channel required by months and one channel per function

'g Table 3.1-1 inoperable in one or such that all channels are tested at more Trip Functions, place the least once every N times 18 months inoperable channel (s) and/or that where N is the total number of trip system in the tripped redundant channels in a specific condition

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

reactor trip function.

2.

With two or more channels required by Table 3.1-1

[

inoperable in one or more Trip Functions:

i a.

Within one hour, verify sufficient channels remain OPERABLE or tripped

to maintain f

trip capability in the Trip Functions, and b.

Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the l

inoperable channel (s) in one trip system and/or that trip system **

in the tripped condition *, and c.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , restore the i

inoperable channels in the other s

trip system to an OPERABLE status or tripped *.

t Otherwise, take the ACTION required by Table 3.1-1 f or the Trip Function.

1 An inoperable channel or trip system need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, if the inoperable channel is not restored to OPERABLE status l

within the required time, the ACTION required by Table 3.1-1 for that Trip Function shall be taken.

This ACTION applies to that trip system with the most inoperable channels; if both systems have the same number of inoperable channels, r

the ACTION can be applied to either trip system.

I t

b Amendment No.

I20*743*I93 3.1-1 l

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE PEOUIREMENTS B.

Two RPS electric power monitoring B.

The RPS power monitoring system modules (or Electric Protective (EPA's) instrumentation shall be Assemblies - EPA's) for each in-determined OPERABLE:

service RPS MG set or alternate source shall be OPERABLE or 1.

By performance of a CHANNEL FUNCTIONAL TEST each time the plant is in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unless performed within the previous 6 months.

1.

With one RPS electric power 2.

At least once per OPERATING CYCLE by-monitoring module (or EPA) for an demonstrating the OPERABILITY of in-service RPS MG set or over-voltage, under-voltage and alternate power supply under-frequency protective inoperable, restore the instrumentation by performance of a inoperable module (EPA) to CHANNEL CALIBRATION including OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months simulated automatic actuation of the or remove the associated RPS MG protection relays, tripping logic set or alternate power supply and output circuit breakers and from service.

verifying the following limits:

2.

With both RPS electric power a.

Over voltage s 132 VAC monitoring modules (EPA's) for an in-service MG set or alternate b.

Under voltage 2 108 VAC power supply inoperable, restore at least one to OPERABLE status c.

Under frequency 2 57 Hz within 30 minutes or remove the l

associated RPS MG set or alternate power supply from service.

l l

I i

i i

l l

?

I i

i i

Amendment No. 19.743.785.797.193 3.1-2 I

~

TABLE 3.1-1

~

REACTOR PROTECTION SYSTEM INSTRUMENTATION N

MINIMUM OPERABLE o

APPLICABLE CHANNELS OPERATING PER TRIP TRIP FUNCTION TRIP LEVEL SETTING MODES SYSTEM (a)

ACTION N

F*

1.

Intermediate Range nonRo.; ib):

a.

Neutron Flux - High 5 120/125 of Full Scale 2

2 1

3,4 2

2 5

2 3

b.

Inoperative NA 2

2 1

3,4 2

2 5

2 3

'2.

Average Power Range Monitor (c):

a.

Neutron Flux - Upscale, in s 15% Power 2

2 1

w STARTUP 3,4 2

2 5

2 3

b b.

Neutron Flux - Upscale For Two Loop Operation:

1 2

4 s (.58W + 62%)*

For Single Loop Operation s (.58W + 58.5%)*

c.

Inoperative NA 1,2 2

1 3,4 2

2 5

2 3

3.

Reactor vessel Steam Dome s 1055 psig 1,2(d) 2 1

Pressure - High 4.

Reactor water Level - Low 2 170 Inches 1,2 2

1 5.

Main. Steam Line Isolation valve -

s 10% valve closure 1(e) 4 4

Closure

TABLE 3.1-1 (continued) g REACTOR PROTECTION SYSTEM INSTRUMENTATION MINIMUM OPERABLE

<D APPLICABLE CHANNELS OPERATING PER TRIP

_ TRIP FUNCTION TRIP LEVEL SETTING MODES SYSTEM (a)

ACTION 6.

Drywell' Pressure - High 5 2.0 psig 1,2(f) 2 1

yw

,tSa 7.

Scram Discharge Volume Water s 60 Gallons 1,2 2

1 Level - High 5(g) 2 3

n.

8.

Turbine Stop Valve - Closure s 10% Valve Closure 1(h) 4(1) 6

,w to 9.

Turbine Control Valve Fast Closure, Within 30 milliseconds 1(h) 2(1) 6 N

Valve Trip System Oil Pressure - Low of the Start of Control oy Valve Fast Closure 10.

Turbine First Stage Pressure s 165 psig 1

2 6

Permissive 11.

Reactor Mode Switch Shutdown NA 1,2 1

1 00 Position 3,4 1

7

-o 5

1 3

12.

Manual Scram NA 1,2 1

1 3,4 1

8 5

1 9

A e

l l

l

{

DAEC-1 TABLE 3.1-1 (Continued) i PEACTOP PPOTECTION SYSTEM INSTRUMENTATION i

t ACTION' I

ACTION 1 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 -

Verify all insertable control rods to be inserted in the core and lock the reactor-mode switch in the Shutdown position within one hour.

ACTION 3 -

Suspend all operations involving CORE i

ALTERATIONS' and insert all insertable control rods within one hour.

ACTION 4 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 -

Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 6 -

Initiate a reduction in THERMAL POWER I

within 15 minutes and reduce turbine first stage pressure to 5 165 psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

i ACTION 7 -

Verify all insertable control rods to be f

inserted within one hour.

i ACTION 8 -

Lock the reactor mode switch in the Shutdown position within one hour.

l ACTION 9 -

Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within one hour.

t i

(

.t I

i L

.I t

i t

'Except movement of IRM, SRM, or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Spec.

i 3.9.B.

' Amendment'No. 193 3.1-5 1

-1

- _ -~.

i DAEC-1 TABLE 3.1-1 (Continued) f PEACTOR PPOTECTION SYSTEM INSTRUMENTATION ACTION TABLE NOTATIONS I

See Section 2.1.A.1 (a)

A channel may be placed in an inoperable status for up to six hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

t (b)

This function shall be automatically bypassed when the reactor I

mode switch is in the Run position.

(c)

An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than the required" LPRM inputs to an APRM l

channel.

i (d)

This function is not required to be OPERABLE when the reactor pressure i

vessel head is unbolted or removed.

(e)

This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

l (f)

This function is not required to be OPERABLE when PRIMARY

{

CONTAINMENT INTEGRITY is not required.

(g)

With any control rod withdrawn.

Not required for control rods removed or withdrawn per Specification 3.9.A.

(h)

This function shall be tutomatically bypassed when turbine first stage pressure is s 1f5 psig, equivalent to THERMAL POWER less than 30% of RATED THEPMAL POWER.

The value of first stage pressure assumes that 'he second stage reheaters are not in service below 30% of rated core power.

l (i)

Also actuates the EOC-RPT system, i

t L

l c

I

't i

I l

I'

+

?

~APRMs A, B,

C, and D require 9 or more inputs. APRM's E and F require 13 or more inputs.

1 Amendment No. J75,193 3.1-6

g TABLE 3.1-2 ro:)

PROTECTIVE INSTRUMENTATION RESPONSE TIMES c1 S

Sensor Response Reactor Trip System Functional Unit Time Response Time z

O 1.

Reactor Vessel Steam Dome Pressure - High

<.5 seconds

<.55 seconds m

m Twa 2.

Reactor Water Level - Low

< 1.0 seconds

~ 1.05 seconds

?

a M

t.d (d

h eN

TABLE 4.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS OPERATING MODES CilANNEL CHANNEL FOR WHICII g

CHANNEL FUNCTION %L CALIBRATION SURVEILLANCE to FUNCTIONAL UNIT CilECK TEST (a)

ItEQUIRED

s 1.

Intermediate Range Monitors:

8 a.

Neutron Flux - fligh S/U(b),

S/U(c),W(1)

Controlled 2

Once/ Shift Shutdown mD Once/ Shift W(1)

Controlled 3,4,5 Shutdown to (a

b.

Inoperative NA S/U(c),W NA 2,3,4,5 2.

Average Power Range Monitor (f):

a.

Neutron Flux - Upscale in S/U(b),

S/U(c),W(1)

SA 2

STARTUP once/ Shift y

once/ Shift W(1)

SA 3,4,5 oo b.

Neutron Flux - Upscale once/ Shift Q(1)

D(d),R(e)

I c.

Inoperatio NA Q

NA 1,2,3,4,5 3.

Reactor Vessel Steam Dome once/ Shift Q

Q 1,2(h)

Pressure - High 4.

Reactor Water Level - Low Once/ Shift Q

Q 1,2 5.

Main Steam Line Isolation valve -

NA Q

R(g) 1 Closure 6.

Drywell Pressure - liigh NA Q

Q 1,2 7.

Scram Discharge volume Water NA Q

R(j) 1,2,5(1)

Level - High 8.

Turbine Stop Valve - Closure NA Q

R(g) 1

--m

--yrae

+,,

.-...,.e

..ms, e

c.

-m.,

. - - +,

-m,,,---

,.,,r,-

.-e.w..~e.,--.

n

-nr.,,--

..,m.,-.n-n,~

--n...,,,,

TABLE 4.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS OPERATING MODES CilANNEL CilANNEL FOR WHICII 3a CIIANNEL FUNCTIONAL CALIBRATION SURVEILLANCE g

FUNCTIONAL UNIT CHECK TEST (a)

REQUIRED o

r' 9.

Turbine control valve Fast m

Closure, Valve Trip System Oil

,0 Pressure - Low NA Q

R(k) 1

10. Turbine First Stage Pressure NA Q

SA 1

[*

Permissive u)

11. Reactor Mode Switch Shutdown NA R

NA 1,2,3,4,5 Position i

12. Manual Scram NA W

NA 1,2,3,4,5 e

M) 0

, ~ -. ~

nw

,.,-,--.,---r-~

, <.., ~, -,, - ~,

e

-n..

m ~

e,

~

- ~ - -

i DAEC-1 TABLE 4.1-1 (Continued)

REACTOR PPOTECTION SYSTEM INSTRUMENTATION SURVEILiANCE REOUIREMENTS TABiE NOTATIONS (a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

I (b)

The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering the STARTUP MODE and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

4 (d)

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during the_RUN MODE when THERMAL POWER >25% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e)

This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f)

The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFFH) using the TIP system.

t (g)

This calibration shall consist of the physical inspection and actuation of these position switches.

(h)

This function a not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed.

(i)

With any control rod withdrawn.

Not applicable to control rods removed or withdrawn per specification 3.9.A.

(j)

Calibrate trip unit at least once per 92 days.

(k)

Measure time interval baseline data for each operating cycle as follows:

From energization of fast acting solenoid, measure time interval to response of oil pressure switch, HFA relay (RPS) and position response j

of control valves.

(1)

This channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

1 f

a J

1

)

i Amendment No. 59.193 3.1-10 l

DAEC-1 i

3.1 BASES The reactor protection system automatically initiates a reactor scram to:

1.

Preserve the integrity of the fuel cladding.

2.

Preserve the integrity of the reactor coolant system.

3.

Minimize the energy which must be absorbed following a loss-of-coolant accident, and prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Allowed outage times have been incorporated consistent with General Electric i

topical report NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System," dated March, 1988.

The reactor protection system is of the dual channel type (Reference Subsection 7.2 of the Updated FSAR).

The system is made up of two independent trip systems, each having three subchannels of tripping devices. One of the three subchannels has inputs from the manual scram push buttons and the reactor mode switch.

Each remaining subchannel has an input from at least one independent instrument channel which monitors a critical parameter.

The outputs of the subchannels are combined in a 1 out of 2 logic; i.e.,

an input signal on either one or both of the subchannels will cause a trip system i

trip.

The outputs of the trip systems are arranged so that a trip on both trip systems is required to produce a reactor scram.

This system meets the intent of IEEE - 279 for Nuclear Power Plant Protection Systems.

The system has a reliability greater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.

The measurement of response time at the specified frequencies provides assurance that the protective, isolation and emergency core cooling functicns associated with each channel is completed within the time limit assumed in the accident analysis.

Pesponse time may be demonstrated by any series of sequential, overlapping or' total channel test measurements, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either:

1) inplace on-site or off-site test measurements, or i

2) utilizing replacement sensors with certified response times.

i With the exception of the Average Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valve closure and the Turbine Stop valve closure, each subchannel has one instrument channel. When the minimum condition for operation on the number of operable t

instrument channels per untripped protection trip system is met or if it cannot be met and the affected protection trip system is placed in a tripped condition, the effectiveness of the protectica rystem is preserved.

Three APRM instrument channels are provided for each protection trip system.

APRM's A and E operate contacts in one subchannel and APRM's C and E operate contacts in the other subchannel.

APRM's B, D and F are arranged similarly in i

the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.

This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels have also been provided to allow for bypassing of one such channel. The bases for the scram setting for i

the IRM, APRM, high reactor pressure, reactor low water level, MSIV closure, generator load rejection and turbine stop valve closure are discussed in.

Specifications 2.1 and 2.2.

Amendment No. 193 3.1-11

~

DAEC-1 Instrumentation (pressure switches) for the drywell are provided to detect a loss-of-coolant accident and initiate the emergency core cooling equipment. A high drywell pressure scram is provided at the same setting as the emergency core cooling systems (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent return to criticality.

This instrumentation is a backup to the reactor vessel water level instrumentation.

The reactor water level trip settings are referenced to the " top of the active fuel" which has been defined to be 344.5 inches above vessel zero.

These trip settings represent the indicated water level.

The MSIV closure scram is set to scram when the isolation valves are 10%

closed in 3 out of 4 lines. This scram anticipates the pressure and flux transient which would occur when the valves close.

By scramming at this setting, the resultant transient is less severe than either the pressure or i

flux transient which would otherwise result.

A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.

The manual scram function is active in all modes, thus providing for a manual l

means of rapidly inserting control rods during all modes of reactor operation.

The APRM (High flux in Startup or Refuel) system provides protection against excessive power levels and short reactor periods in the startup and intermediate power ranges.

The IRM system provides protection against short reactor periods in these ranges.

A source range monitor (SRM) system is also provided to supply additional neutron level information during startup but has no scram functions (reference paragraph 7.6.1.4 of the Updated FSAR).

Thus, the IRM and APRM are required in the " Refuel" and "Startup/ Hot Standby" modes.

In the power range the APRM system provides required protection (reference paragraph 7.6.1.7 of the Updated FSAR).

Thus the IRM System is not required in the "Run" mode.

The l APRM's cover only the power range. The IRM's and APRM's provide adequate coverage in the startup and intermediate range.

The control rod drive scram system is designed so that all of the water which l

is discharged from the reactor by a scram can be accommodated in the discharge piping. The scram discharge volume accommodates in excess of 60 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.

During normal operation the discharge volume is empty; however, should it fill I

with water, the water discharged to the piping from the reactor could not ce accommodated which would result in slow scram times or partial control rod insertion. To preclude this occurrence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water reaches 60 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to i

perform its function adequately.

The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume high level scrams are required for Startup and Run modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation.

i l

Turbine stop valve closure trip occurs at approximately 10% of valve closure.

l Below 165 psig turbine first stage pressure (corresponding to 30% of rated core power), the scram signal due to turbine stop valve closure is by-passed l

Amendment No. 23,743sIE2*l93 3.1-12

[

9

DAEC-1 because the flux and pressure scrams are adequate to protect the reactor below 30% of rated core power.

l Turbine control valve f ast closure scram trip shall initiate within 30 milliseconds of the start of control valve fast closure.

The trip level setting is verified by measuring the time interval from energizing the fast acting solenoid (from valve test switch) to pressure switch response; the measured result is compared to base line data taken during each refueling outage.

Turbine control valve fast closure is sensed by measuring disc dump electro-hydraulic oil line pressure (Relay Emergency Trip Supply) which decreases rapidly upon generator load rejection. This scram is only effective 4-when turbine first stage pressure is above 165 psig (corresponding to 30% of j

rated core power).

The APRM downscale trip signal in the Run mode reactivated the IRM upscale trip to the RPS logic.

It was determined that the operations addressed by this trip (startup and power descent) are adequately covered by the APRM upscale trip and the APRM downscale rod block. Consequently, the APRM j

downscale trip was deleted.

I P

l 1

t i

)

t Amendment No. 2).193 3.1-13 i

r F

~

f

l DAEC-1 4

4.1 BASES i

1.

The minimum functional testing frequency used in'this specification is based on a reliability analysis using the concepts developed in l

Reference 1.

This concept was specifically adapted to the one out of two taken twice logic of the reactor protection system.

The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system. This analysis makes use of " unsafe failure" rate experience at conventional and nuclear power plants in a 4

reliability model for the system. An " unsafe failure" is defined as one which negates channel operability and which, due to its nature, is i

revealed only when the channel is functionally tested or attempts to respond to a real signal.

Failures such as blown fuses, ruptured bourdon tubes, faulted am and faulted cables, which result in

" upscale" or "downscale" plifiers, readings on the reactor instrumentation are i

" safe" and will be easily recognized by the operators during operation t

because they are revealed by an alarm or a scram.

l The channels listed in Tables 4.1-1 are divided into three groups for functional testing.

These are:

A.

On-Off sensors that provide a scram trip function.

B.

Analog devices coupled with bi-stable trips that provide a scram function.

C.

Devices which only serve a useful function during some restricted mode of operation, such as startup or shutdown, or for which the only practical test is one that can be performed at shutdown.

The sensors that make up group 1 are specifically selected from among the whole family of industrial on-off sensors that have earned an excellent reputation for reliable operation.

During design, a goal of 0.99999 probability of success (at the 50% confidence level) was adopted to assure that a balanced and adequate design is achieved.

The probability of success is primarily a function of the sensor failure rate and the test interval. A three-month test interval is planned for Group 1 sensors. This is in keeping with good operating practices, and i

satisfiee the design goal for the logic configuration utilized in the l

Reactor Protection System.

To satisfy the long-term objective of maintaining an adequate level of f

safety throughout the plant lifetime, a minimum goal of 0.9999 at the 95% confidence level is proposed. With the (1 out of 2) X (2) logic, this requires that each sensor have an availability of 0.993 at the 95%

confidence level. This level of availability may be maintained by adjusting the test interval as a function of the observed failure history (Reference 1).

To facilitate the implementation of this technique, Figure 4.1-1 is provided to indicate an appropriat+ trend in test interval.

The procedure is as follows:

+

1.

Like sensors are pooled into one group for the purpose of data f

acquisition.

I 2.

The factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T (M = nT).

3.

The accumulated number of unsafe failures is plotted au an ordinate I

against H as an abscissa on Figure 4.1-1.

t 4.

After a trend is established, the appropriate monthly test interval to satisfy the goal will be the test interval to the left of the plotted points.

l S.

A test interval of 1 month will be used initially until a trend is established, which is based on system availability analysis and good j

engineering judgement plus operating experience.

Group 2 devices utilize an analog sensor followed by an amplifier and a bi-stable trip circuit. The sensor and amplifier are active components and a failure is almost always accompanied by an alarm and an indication of the source of trouble.

In the event of failure, repair or i

Amendment No.

3.1-14 l'

t

f I

DAEC-1 r

substitution can start immediately. An "as-is" failure is one that

" sticks" mid-scale and is not capable of going either up or down in i

response to an out-of-limits input.

This type of failure for analog devices is a rare occurrence and is detectable by an operator who j

observes that one signal does not track the other three.

For purposes of analysis, it is assumed that this rare failure will be detected within two hours.

The bi-stable trip circuit which is a part of the Group 2 devices can sustain unsafe failures which are revealed only on test.

Therefore, it is necessary to test them periodically.

6 A study was conducted of the instrumentation channels included in the Group 2 devices to calculate their " unsafe" failure rates.

The analog devices (sensors and amplifiers) are predicted to have an unsafe failure rate of less than 20 X 10+ failure / hour.

The bi-stable trip circuits are predicted to have unsafe failure rate of less than 2 X 10*

failures / hour.

Considering the two hour monitoring interval for the analog devices as assumed above, and a weekly test interval for the bi-l stable trip circuits, the design reliability goal of 0.99999 is attained l

with ample margin.

The bi-stable devices are monitored during plant operation to record i

their failure history and establish a test interval using the curve of Figure 4.1-1.

There are numerous identical bi-stable devices used throughout the plant's instrumentation system.

Therefore, significant data on the failure rates for the bi-stable devices should be accumulated rapidly.

The frequency of calibration of the APRM Flow Blasing Network is once f

per operating cycle. The flow biasing network is functionally tested at l

least once per quarter and in addition, cross calibration checks of the flow input to the flow biasing network can be made during the functional j

test by direct meter reading. There are several instruments which must be calibrated and it will take several days to perform the calibration i

of the entire network. While the calibration is being performed, a zero flow signal will be sent to half of the APRM's resulting in a half scram and rod block condition.

Thus, if the calibration were performed during operation, flux shaping would not be possible.

Based on experience at i

other generating stations, drift of instruments, such as those in the Flow Biasing Network, is not significant and therefore, to avoid spurious scrams, a calibration frequency of once per operating cycle is established.

Group 3 devices are active only during a given portion of the t

operational cycle.

For example, the IRM is active during startup and inactive during full-power operation.

Thus, the only test that is meaningful is the one performed just prior to shutdown or startup; i.e.,

the tests that are performed just prior to use of the instrument.

i Calibration frequency of the instrument channel is divided into two groups.

These are as follows:

1 1.

Passive type indicating devices that can be compared with like units on a continuous basis.

2.

Vacuum tube or semi-conductor devices and detectors that drift or I

lose sensitivity.

t Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.

For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4%/ month; i.e.,

in the period of a month a maximum drift of 0.4% could occur, thus providing for adequate margin.

l 1

For the APRM system, drift of electronic apparatus is not the only f

consideration in determining a calibration freguency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days.

Calibration on this frequency assures plant operation at or i

below thermal limits.

l i

)

Amendment No. 114,193 3.1-15

t i

DAEC-1 i

The mode switch in shutdown and manual scram trip functions are simple on-off switches and, hence, calibration during operation is not applicable.

2.

The peak heat flux is checked once per day to determine if the APRM scram re@ ires adjustment. This will normally be done by checking the LPRM readings. Only a small number of centrol rods are moved daily and i

thus the power distribution is not expected to change significantly and thus a daily check of the peak heat flux is adequate.

The sensitivity of LPRM detectors decreases with exposure to neutron f

flux at a slow and approximately constant rate.

This is compensated for t

in the APRM system by calibrating twice a week using heat balance data and by calibrating individual LPRM's every 1000 effective full power j

hours, using TIP traverse data.

i several channel functional test frequencies were extended from monthly to quarterly in accordance with NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System," dated March, 1988.

t t

i l

]

i 9

i j

i I

i

)

1 1

i i

Amendment No. 74,193 3.1-16 i

DAEC-1

4.1 REFERENCES

1.

Reliability of Engineered Safety Features as a Function of Testing Frequency, 1.

M. Jacobs, " Nuclear Safety", Volume 9, No.

4, July-August 1968, pp. 310-312.

l t

L L

l r

I i

L

-i l

I l

l i

I l

l 3

1

.t i

'I

\\

i i

j i

]

Amendment No.

133 3.1-17 j

r i' i

b

=.

DAEC-1 i

L!MITING CONDITIONS FOR OPERATION SURVEILLANCE PEOUIREMENTS l

i 3.2 PROTECTIVE INSTRUMENTATION 4.2 PROTECTIVE INSTRUMENTATION I

A.

ISOLATION ACTUATION INSTRUMENTATION A.

ISOLATION ACTUATION INSTRUMENTATION 1.

The isolation actuation 1.

Each isolation actuation l

instrumentation channels shall be instrumentation channel shall be OPERABLE as shown in Table 3.2-A.

demonstrated OPERABLE by the performance of the CHANNEL CHECK, 1

CHANNEL FUNCTIONAL TEST and CHANNEL l

CALIBRATION operations for the Applicability:

OPERATING MODES and at the frequencies shown in Table 4.2-A.

l As shown in Table 3.2-A l

Action:

2.

LOGI. SYSTEM FUNCTIONAL TESTS shall

(

be performed at least once per j

a.

With the number of OPERABLE operating cycle for the following:

channels less than required by the Minimum OPERABLE Channels per a.

Main Steam Line Isolation Valves Trip System requirement for one Main Steam Line Drain Valves trip system:

Reactor Water Sample Valves I

1)

If placing the inoperable b.

RHR-Iso!7. tion Valve Control J

channel (s) in the tripped Shutdes-Cooling Valves condition would cause an isolation, the inoperable c.

Reactor Water Cleanup Isolation channel (s) shall be restored i

to OPERABLE status within 6 d.

Drywell Isolation Valves hours TIP Withdrawal i

Atmospheric Control Valves or the ACTION required by Sump Drain Valves t

i Table 3.2-A for the affected trip function shall be taken.

e.

Standby Gas Treatment System

+

Reactor Building Isolation OR 4

j f.

HPCI Subsystem Auto Isolation

]

2)

If placing the inoperable l

1 channel (s) in the tripped g.

RCIC Subsystem Auto Isolation condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the j

tripped condition within 1

3 a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip t

functions common to RPS Instrumentation; and b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip l

functions not common to RPS Instrumentation.

I i

8 Amendment No. 193 3.2-1

DAEC-1 LIMITING CONDITIONS FOR OPERATION SUPVEILLANCE PFOUIREMENTS b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system

in the tripped condition within one hour and take the ACTION required by Table 3.2-A.

i l

t f

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 I

cause the isolation to occur.

i Amendment No.

7/d,780,193 3.2-2

Table 3.2-A ISOLATION ACTUATION INSTRUMENTATION MINIMUM VALVE El OPERABLE C20UF8 M

APPLICABLE CilANNELS ISOLATED 3

OPERATING PER TRIP BY TRIP FUNCTION TRIP LEVEL SETTING MODE SYSTEM

SIGNAL ACTION Common Isolation Sionals

'd 1,2,3 2

2 20 E

Reactor Water Level-Low 2 170 Inches 1,2,3 and

2 3

26 0,

1,2,3 2

4,5 23 N

Reactor Water Level - Low-Low-Low 2 18.5 Inches 1,2,3 2

1 21

-o 1,2,3 2

7 20 Drywell Pressure - High 5 2.0 psig 1,2,3 2

2 20 1,2,3 and

2 3 *'

26 1,2,3 2

4, 9.n 23 F

Main Steam Line Isolation N

Main Steam Line Pressure - Low 2 850 psig 1

2 1

22 Main Steam Line Flow ~ liigh 5 140% of Rated 1,2,3 2/line 1

20 Steam Flow Condenser Backpressure - High s 20 In. Hg 1,2**,3**

2 1

21 Main Steam Line Tunnei s 200*F 1,2,3 2/line 1

21 Temperature - High Turbine Building Temperature -

5 200*F 1,2,3 4

1 21 High Main Steam Line Radiation - Iligh s 3 x Normal Rated 1,2,3 2

1 *'

21 Power Backg rou nd9' i

~,...

o Tabic 3.2-A (Continued)

{

ISOLATION ACTUATION ITISTRUMENTATION S

MINIMUM VALVE OPERABLE GROUPS h

APPLICABLE CHANNELS ISOLATED OPERATING PER TRIP IW w

TRIP FUNCTION TRIP LEVEL SETTING MODE SYSTEM * '

SIGH %

-ACTION I

E Secondary containment tsa

.N Refuel Floor Exhaust Duct - High 3 9 mr/hr 1,2,3 and

1 3(cl 26 Radiation to i3 ci 26 Reactor Building Exhaust Shaft -

$ 11 mr/hr 1,2,3 and

1 High Radiation offgas vent Stack - High Radiation 5 1.5x10 cps 1,2,3 and

1 3(C) 26 8

RHR System Shutdown Cooling Reactor Vessel Pressure - High 5 135 psig 1,2,3 1

4 23 3

7 Reactor Water Cleanup e

RWCU Dif ferential Flow - Ifigh 5 40 gpmd 1,2,3 1

5 23 RWCU Area Temperature - High 5 130'F 1,2,3 1

5 23 RWCU Area Ventilation Differential A14*Fidl 1,2,3 1

5 23 Temperature - High Standby Liquid Control System NA Note i 1

5(*)

23 Initiation RWCU Area Near TIP Room 5 111.5'F 1,2,3 1

5 23

-Ambient Temperature - High l

l

9 Table 3.2-A (Continued)

_ ISOLATION ACTUATION INSTRUMENTATION

_,y MINIMUM-VALVE u)H W$

OPERABLE GROUPS APPLICABLE CHANNELS ISOLATED I

g OPERATING PER TRIP BY cc TRIP FUNCTION TRIP LEVEL SETTING MODE SYSTEM'"

SIGNAL ACTION N

Peactor Core Isolation Coolino RCIC Steam Line Differential s 155 Inches H;O 1,2,3 1

6A 23 Qj Pressure (Flow) - High

~

RCIC Steam Supply Pressure - Low 100 > P > 50 psig 1,2,3 2

6A 23 RCIC Turbine Exhaust Diaphragm s 10 psig 1,2,3 2

6A 23 to Pressure - High RCIC Equipment Room Temperature -

5 175'F 1,2,3 1

6A 23

.s High RCIC Room Ventilation Differential 5 A50*F 1,2,3 1

6A 23 w

Temperature - High RCIC Leak Detection Time Deiay s 30 Minutes 1,2,3 1

6A 23 (n

Suppression Pool Area Temperature 5 150'F 1,2,3 1

6A 23

- High Suppression Pool Area Ventilation s A50*F 1,2,3 1

6A 23 Differential Temperature - High Manual Initiation NA 1,2,3 1/RCIC 6 A5' 25 System RCIC System Initiation MO-2404 Not Full 1,2,3 1/RCIC 8

23 Closed System

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

Table 3. 2-A (Cont inue_dj ISOLATION ACTUATION INSTRUMENTATION e

d MINIMUM VALVE 3

OPERABLE GROUPS APPLICABLE CHANNELS ISOLATED r+

OPERATING PER TRIP BY TRIP FUNCTION TRIP LEVEL SETTING MODE S Y STEM *'

SIGNAL ACTION g

Hiah Pressure Coolant Iniection et HPCI Steam Line Differential 5 103 Inches H;O 1,2,3 1

6B 23 Pressure (Flow) - High (Outboard) gg 5 386 Inches H0 (Inboard)

U3 HPCI Steam Supply Pressure - Low 100 > P > 50 psig 1,2,3 2

6B 23 HPCI Turbine Exhaust Diaphragm s 10 psig 1,2,3 2

6B 23 Pressure - High u3 HPCI Equipment Room Temperature -

s 175'F 1,2,3 1

6B 23

(,

High Es HPCI Room Ventilation Differential s A50*F 1,2,3 1

6B 23 Temperature - High HPCI Leak Detection Time Delay s 15 Hinutes 1,2,3 1

6B 23 Suppression Pool Area Temperature s 150'F 1,2,3 1

6B 23

- High Suppression Pool Area Ventilation s A50'F 1,2,3 1

6B 23 Differential Temperature - High HPCI System Initiation MO-2202 Not Full 1,2,3 1/HPCI 8

23 Closed System Hanual Initiation NA 1,2,3 1/HPCI 6 B +'

25 System

I 4

DAEC-1 h

Table 3.2-A (Continued)

ISOLATION ACTUATION INSTPUMENTATION l

ACTION

\\

ACTION 20 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 -

Be in at least STARTUP with the associated isolation valves i

closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within i

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 23 -

Close the affected system isolation valves within one hour l

and declare the affected system inoperable.

ACTION 24 Not Used ACTION 25 -

Restore the manual initiation function to OPERABLE status within B hours or close the affected system isolation valves within the next hour and declare the affected system i

3 inoperable.

ACTION 26 -

Establish SECONDARY CONTAINMENT INTEGRITY with the Standby i

Gas Treatment System operating within one hour.

i 1

i NOTES When handling irradiated fuel in the secondary containment and during k

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

When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the NORM position.

(a) When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed as follows: (1) for up to 3

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for RWCU Differential Flow-High, RCIC Manual Initiation, HPCI Manual Initiation; and (2) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function maintains isolation capability.

(b) Operates Group 1 valves except Main Steam Isolation valves.

Also trips Mechanical Vacuum Pump which results in a subsequent isolation of the

+

Mechanical Vacuum Pump suction valves.

(c) Also starts the Standby Gas Treatment System.

(d) Actual setpoint shall be 14'F above the 100% operation ambient j

temperature conditions as determined by DAEC plant test procedure.

t (e)

Closes MO-2701 and MO-2740 only.

(f)

Requires system steam supply pressure-low coincident with drywell pressure-high to close HPCI/RCIC exhaust vacuum breaker valves.

t (g)

Hanual isolation closes HO-2401 only, if RCIC initiation signal present.

(h) Manual isolation closes MO-2239 only, if HPCI initiation signql present.

1 i

(i) When the Standby Liquid Control System is required to be OPERABLE per l

Specification 3.4.A.

7 (j) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the planned start of the hydrogen injection test with the reactor power at greater than 20% rated power, the normal full-power radiation background level and associated trip setpoints may i

be changed based on a calculated value of the radiation level expected during the test.

The background radiation level and associated trip setpoints may be adjusted during the test program based on either calculations or measurements of actual radiation levels resulting from i

hydrogen injection. The background radiation level shall be determined

l and associated trip setpoints shall be set within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of I

reestablishing normal radiation levels after completion of the hydrogen injection test or within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of establishing reactor power levels i

below 20% rated power, while these functions are required to be operable.

i I

Amendment No.29.731,193 3.2-7

Table 4.2-A ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS

sr+

OPERATING m

CIIANNEL MODES FOR WHICH

.O CHANNEL FUNCTIONAL CllANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED y

w?

Common Isolation Sionals Reactor Water Level-Low (###)

Once/ Shift Q

Q 1,2,3 and

U Reactor Water Once/ Shift Q

Q 1,2,3 y

Level-Low-Low-Low (##)

h Drywell Pressure - liigh

(###)

NA Q

Q 1,2,3 and

Main Steam Line__ Isolation Main Steam Line Pressure - Low NA Q

Q 1

Main Steam Line Flow - High Once/ Shift Q

Q 1,2,3 C'

Condenser Backpressure - Ifigh NA Q

A 1,2**,3**

Main Steam Line Tunnel Temperature - Illgh D

Q A

1,2,3 Turbine Building Temperature - liigh D

Q A

1,2,3 Main Steam Line Radiation - Illgh once/ Shift Q

R 1,2,3 Secondary Containment Refuel Floor Exhaust Duct - High Radiation D

Q R

1,2,3 and

Reactor Building Exhaubt Shaft - High Radiation D

Q R

1,2,3 and

Offgas Vent Stack - High Radiation D

Q R

1,2,3 and

4 w

Table 4.2-A (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS El OPERATING j

CHANNEL MODES FOR WHICH 3

CilANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED m

?

RHR System Shutdown Coolina El Reactor Vessel Pressure - High NA Q

Q 1,2,3 N

}a Reactor Water Cleanup h

RWCU Differential Flow - High D

Q Q

1,2,3 RWCU Area Temperature - High NA Q'"

A 1,2,3 RWCU Area Ventilation Differential Temperature NA Q

A 1,2,3

- High Standby Liquid Control System Initiation NA R

NA Note b ta h)

Reactor Core Isolation Coolina RCIC Steam Line Differential Pressure (Flow) -

NA Q

Q 1,2,3 High RCIC Steam Supply Pressure - Low NA Q

Q 1,2,3 RCIC Turbine Exhaust Diaphragm Pressure - High NA Q

R 1,2,3 RCIC Equipment Room Temperature - High D

Q A

1,2,3 RCIC Room Ventilation Differential Temperature D

Q A

1,2,3

- High RCIC Leak Detection Time Delay NA NA A

1,2,3 Suppression Pool Area Temperature - High D

Q A

1,2,3 Suppression Pool Area Ventilation Differential D

Q A

1,2,3 Temperature - High Manual Initiation NA R

NA 1,2,3 RCIC System Initiation (HO-2404 Not Full NA R

NA 1,2,3 Closed)

5 o.

,M Table 4.2-A-fcontinued) o" ISOLATION ACTURTION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

T

,7 OPERATING CHANNEL MODES FOR WHICH CHANNEL FUNCTIONAL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED

~$

Hiah Pressure Coolant Iniecti?D w

HPCI Steam Line Differential Fressure (Flow)-

NA Q

Q 1,2,3

- High HPC7.'Sicam Supply Pressure - Low NA Q

Q 1,2,3 HPCI Turbine Exhaust Diaphragm Pressure -

NA Q

R 1,2,3 Higli HPCI Equipment Room Temperature - High D

Q A

1,2,3 HPCI Room Ventilation Differential D

Q A

1,2,3 w

Temperature - High fy HPCI Leak Detection Time Delay NA Q

A 1,2,3 Suppression Pool Area Temperature - High D

Q A

1,2,3 Suppression Pool Area Ventilation D

Q A

1,2,3 Differential Temperature - High HPCI System Initiation (MO-2202 Mot Full NA R

NA 1,2,3 Closed)

Manual Initiation MA R

NA 1,2,3

When handling irradiated fue1~in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

- When any turbine stop valve is greater than 90% open and/or when the key-locked bypass switch is in the Norm position.

1. These trip functions are common to the ECCS activation trip function.

1. 1. These trip functions are common to the RPS and ECCS activation trip functione.

(a)The functional test will consist of comparing the analog signal of the active thermocouple element feeding the isolation logic to a redundant thermocouple element.

(b)When the Standby Liquid Control System is required to be OPERABLE per Specification 3.4.A.

l (c)This channel functional tes?. will consist of injecting a simulated electrical signal into the measurement channels.

4 f

1 I

. ~

t DAEC-1 i

j LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS I

B.

CORE AND CONTAINMENT COOLING B.

CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS INITIATION / CONTROL INITIATION / CONTROL INSTRUMENTATION INSTRUMENTATION 1.

The core and containment cooling 1.

Each core and containment cooling i

system actuation instrumentation system actuation instrumentatacn channels shall be OPERABLE as channel shall be demonstrated shown in Table 3.2-B.

OPERABLE by the performance of the r

CHANNEL CHECK, CHANNEL FUNCTIONAL APPLICABILITY:

TEST and CHANNEL CALIBRATION operations for the OPERATING MODES i

As shown in Table 3.2-B.

and at the frequencies shown in Table 4.2-B.

ACTION:

With one or more core and 2.

LOGIC SYSTEM FUNCTIONAL TESTS shall i

containment cooling systems be performed at least once per actuation instrumentation operating cycle for the following:

channels inoperable, take the l

ACTION required by Table a.

Core Spray System 3.2-B.

I b.

Low Pressure Coolant Injection Mode f

of RHR System t

c.

Containment Spray Interlocks d.

HPCI System e.

ADS System f.

Area Cooling for Safeguard Systems g.

Low-Low Set Function I

i 1

j l

l i

Amendment No. ))P.777,193 1.2-11 1

i

l TABLE 3.2-8 fl CORE AND CONTAINMENT COOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION a$

e3e Minimum Operable Chanrels Applicab per Trip operating Trip Function Trip Level Setting Fu nct ion

Mode Action to "8

CORE SPRAY SYSTEM Reactor Water Level-Low-Low-Low 2 +18.5 Inches 4(b) 1,2,3,4*,5*

30 Drywell Pressure-High s 2.0 psig 4(b) 1,2,3 30 Reactor Pressure-Low (Permissive) 2 450 psig 4

1,2,3 31 4*,5*

32 Core Spray Pump Start Time Delay 5 see 1/ pump 1,2,3,4*,5*

31 P

LOW PRESSURE COOLANT h)

INJECTION MODE OF RHR h3 Reactor Water Level-Low-Low-Low 2 +18.5 Inches 4

1,2,3,4*,5*

30 Drywell Pressure-High s 2.0 psig 4

1,2,3 30 Reactor Precaure-Low (Permissive) 2 450 psig 4

1,2,3 31 4*,5*

32 LPCI Pump Start Time Delay 10 sec (A&B) 1/ pump 1,2,3,4*,5*

31 15 sec (C&D)

LPCI Loop Select Reactor Water Level-Low-Low 2 +119.5 Inches 4

1,2,3,4*,5*

31 Reactor Pressure-I.ow 2 900 peig 4

1,2,3,4*,5*

31 Recirculation Pump Differential s

2 psid 4/ pump 1,2,3,4*,5*

31 Pressure Recirculation Riser Differential O.5 < p < 1.5 psid 4

1,2,3,4*,5*

31 Pressure

TABLE 3.2-0 (Continued)

EU CORE AND CONTAINMENT COOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION a

R

)

co y

Minimum o

operable Channels Applicable to per Trip operating Trip Function Trip Level Setting Fu n c t ion

Mode Action HIGH PRESSURE COOLANT INJECTION SYSTEM (#)

C$

Reactor ifater Level-Low-Low 2 +119.5 Inches 4

1,2,3 34 t,a Drywell Pressure-High s 2.0 psig 4

1,2,3 34 Reactor Water Level-High s +211 Inches 2 'd' 1,2,3 31 Condensate Storage Tank Level-Low 2 12 Inches 2"

1,2,3 35 above tank bottom po (10,000 gallons) to 8,

Suppression Pool Water Level-High 5 5 Inches 2"

1,2,3 35 above nornal level REACTOR. CORE ISOLATION COOLING SYSTEM (F)

Reactor Water Levn1-Low-Low a +119.5 Inches 4

1,2,3 30 Beactor Water Level-High 5 +211 Inches 2*

1,2,3 31 Condensate Storage Tank Level-Low 2 12 Inches 2"

1,2,3 35 above tank bottom (10,000 gallons) s.

= - - -

--.w

w..

nm -

TABLE 3.2-D (Continued) gl CORE AND CONTAINMENT COOLING SYJTEMS INITIATION / CONTROL INSTRUMENTATION

. 's a

5 36 Minimum

=

Operable

?

Channels Applicable per Trip Operating 3.

go Trip Function Trip Level Setting Fu n c t ion

Mode Action f

AUTOMATIC DEPRESSURIZATION SYSTEM (##)

Reactor Water Level-Low-Low-Low 2 +18.5 Inches 4

1,2,3 30 Reactor Water Level-Low (Confirmatory) 2 +170 Inches 2

1,2,3 31 ADS Timer 120 see 2

1,2,3 31 Core Spray Pump Discharge Pressure-High 145 psig 2/ pump 1,2,3 31 (Permissive)

I" RHR(LPCI) Pump Discharge Pressure-High 125 psig 1/ pump 1,2,3 31 Q3 (Permissive)

-c, CONTAINMENT COOLING Reactor Water Level-Low 2 - 39 Inches 4

1,2,3 30 (Inside Shroud)

(2/3 core height)

' Containment Pressure-High 2 2.0 psig 4

1,2,3 30 t

..m

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

l TABLE 3.2-B (Continued)

E l

CORE AND CONTAINMENT JOOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION S

E E

Minimum en Operable

}n Channels Applicable to per Trip operating y*

Trip Function Trip Level Setting Fu n c t i o n

Mode Action w

{"

LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Loss 20 s V s 28 volts 2

1,2,3,4**,5**

33 us of Voltage) 4.16 kv Emergency Bus Degraded voltage a.

108 s V s 111 volts 8

1,2,3,4**,5**

36

b. 8.0 s t s 8.5 see time delay 4.16 kv Emergency Transformer Supply -

65% of Rated Voltage 4

1,2,3,4**,5**

36

.u, Undervoltage 4.16 kv Emergency Bus Sequential Loading 65% of Rate altage 2

1,2,3,4**,5**

36 d*

Relay l

NOTES i

(a)

When a channel is placed in an inoperable status solely for performance of required surveillances, entry into associated Limiting Conditions for Operation and required Actions may be delayed as follows: (1) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for HPCI Reactor Water Level-High, HPCI Condensate Storage Tank Level-Low, HPCI Suppression Pool Water Level-High, RCIC Reactor Water Level-High, RCIC Condensate Storage Tank Level-Low; and (2) for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the remaining Trip Functions provided the associated Trip Function maintains initiation / trip capability.

(b)

Also actuates the associated emergency diesel generators.

(c)

One trip system.

Provides signal to the pump suction valves only.

(d)

Provides signal to trip pump turbine only.

When the system is required to be OPERABLE per Specification 3.5.A.

l

- Required OPERABLE when ESF equipment is required to be OPERABLE.

l

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.

l

1. Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

1

- - - -..- - - - -. ~

m.

I DAEC-1 TABLE 3.2-B # Continued)

?

COPE AND CONTAINMENT COOLING SYSTEMS INITIATION / CONTROL INSTPUMENTATION ACTION ACTION 30 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated system inoperable.

b.

With more than one channel inoperable, declare the associated system ineperable

,i ACTION 31 -

With the number of OPERABLE channels less than required by i

the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, declare the associated ECCS inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With more than one channel inoperable, declare the associated ECCS inoperable within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With more than one channel inoperable, declare the associated ECCS inoperable within I hour.

ACTION 33 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated emergency diesel generator inoperable and take the action required by Specification 3.5.G.I.

ACTION 34 -

With the number of OPERABLE channels less than regaired by the Minimum OPERABLE Channels per Trip Function requirement:

a.

For one channel inoperable, place the inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCI system inoperable.

b.

With more than one channel inoperable, declare the HPCI system inoperable.

ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated system inoperable.

ACTION 36 -

With the number of OPERABLE channels one less than required-by the Minimum OPERABLE Channels per Trip Function, place the inoperable channel in the tripped condition within 1 j

hour; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST.

5 Amendment No. J20,193 3.2-16 1

E

F mo Q

TABLE 4.2-B e

CORE AND CONTAINMENT COOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION SURVEILLANCE REQUIREHENTS

.E Channel s,

Channel Functional Channel Operating Modes for Which Ty Trip Function Check Test Calibration Surveillance Required CORE SPRAY SYSTEM w

Reactor Water Level-Low-Low-Low D

Q Q

1, 2, 3, 4*,

5*

Drywell Pressure-High NA Q

Q 1, 2, 3 Reactor Pressure-Low (Permissive)

NA Q

Q 1,

2, 3, 4*,

5*

Core Spray Pump Start Time Delay NA NA A

1, 2,

3, 4*,

5*

.L_OW PRESSURE COOLANT INJECTION MODE OF RHR 3

Reactor Water Level-Low-Low-Low D

Q Q

1, 2, 3, 4*,

5*

'J Drywell Pressure-High NA Q

Q 1, 2, 3 Reactor Pressure-Low (Permi,alve)

NA Q

Q 1, 2, 3, 4 *,

5*

LPCI Pump Start Time Delay NA NA A

1, 2, 3,

4*,

5*

LCPI Loop Select Reactor Water Level-Low-Low D

M Q

1, 2, 3,

4*,

5*

Reactor Pressure-Low NA M

Q 1, 2, 3,

4*,

5*

Recirculation Pump Differential Pressure D

M Q

1, 2, 3, 4*,

5*

Recirculation Riser Differential Pressure D

M Q

1, 2,

3, 4*,

5*

TABLE 4.2-B (Continued) gl CORE AND CONTAINHENT COOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION SURVEILLANCE.REOUIREMENTS BaD Channel Channel Functional Channel Operating Modes For Which 6

Trip Function Check Test Ca l ib ra t i rgn Surveillance Required a

P HIGH PRESSURE COOLANT INJECTION SYSTEM

(#)

h Reactor Water Level-Low-Low D

Q Q

1, 2,

3 G$

Drywell Pressure-High NA Q

Q 1,

2, 3

w Reactor Vessel Water Level-High D

Q Q

1, 2,

3 Condensate Storage Tank Level-Low NA Q

Q 1,

2, 3

j Suppression Pool Water Level-Migh NA Q

Q 1,

2, 3

REPCTOR CORE ISOLATION COOLING SYSTEM

(#)

Reactor Water Level-Low-Low D

Q Q

1, 2,

3 00 Reactor Water Level-High D

Q Q

1, 2,

3 Condensate Storage Tank Level-Low NA Q

Q 1,

2, 3

AUTOMATIC DEPRESSURIZATION SYSTEM (##)

Reactor Water Level-Low-Low-Low D

Q Q

1, 2,

3 Reactor Vessel Water Level-Low D

Q Q

1, 2,

3 (Confirmatory)

ADS Timer NA Q

R 1,

2, 3

Core Spray Pump Discharge Pressure-High NA Q

Q 1,

2, 3

(Permissive)

RHR(LCPI) Pump Discharge Pressure-High NA Q

Q 1,

2, 3

(Permissive)

Low-Low Set Low-Low Set Function Setpoints D

M SA 1,

2, 3

TABLE 4.2-B (Continued)

?>l CORE AND CONTAINMENT COOLING SYSTEMS INITIATION / CONTROL INSTRUMENTATION SURVEILLANCE REOUIREMENTS 4au Channel r+

Channel Functional Channel Operating Modes for which a

Trip Function Check Test Calibration Surveillance Required P

my CONTAINMENT COOLING ff Reactor Water Level-Low (Inside Shroud)

D M

Q 1,

2, 3

containment Pressure-High NA M

Q 1,

2, 3

LOSS OF POWER 4.16 kv Emergency Bus NA A

A 1,

2, 3,

4**,

5**

Undervoltage (Loss of Voltage) 4.16 kV Emergency Bus Degraded Voltage NA M

A 1,

2, 3,

4**,

5**

u, 4.16 kv Emergency Transformer Supply-NA A

A 1,

2, 3,

4**,

5**

g Undervoltage 4.16 kv Emergency Bus Sequential Loading NA A

A 1,

2, 3,

4**,

5**

Relay When the system is required to.be OPERABLE per Specification 3.5.A.

Required OPERABLE when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 150 psig.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

~

OAEC-1 LIMITING COND!TIONS FOR OPERATION SURVEILLANCE PEOUIPEMENTS C.

CONTROL ROD BLOCK INSTPUMENTATION C.

CONTROL ROD BLOCK INSTRUMENTATION i

1.

The control rod block 1.

Each of the required control rod instrumentation channels shall be block instrumentation channels shall OPERABLE as shown in Table 3.2-C.

be demonstrated OPERABLE by the performance of the CHANNEL CHECK, Arrlicability:

CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the As shown in Table 3.2-C.

OPERATING MODES and at the frequencies shown in Table 4.2-C.

Action:

With the numoer of OPERABLE i

channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.2-C.

2.

Rod Block Monitor (RBMI a.

The RBM control rod block setpoints are given in Table 3.2-C.

The upscale High Power Trip Setpoint shall be applied w.*n the core thermal power is greater than or equal to 85% of rated (F 2 851).

The upscale Intermediate Power Trip setpoint shall be applied when the core thermal power is greater than or equal to 65% of rated and less than 85% of rated (65% 5 P < 85%).

The upscale Low Power Trip Setpoint shall be applied when the core thermal power is greater than or equal to 30% of rated and less than 5% of rated (301 sP<

65%).

The RBM can be bypassed when core thermal power is less than 30% of rated.

The RBM bypass time delay (td,) shall be less than er equal to 2.0 seconds.

b b

3 Amendment No. 769,193 3.2-20

Table 3.2-C d

CONTROL ROD DLOCK INSTRUMENTATION a

c1 MINIMUM M

OPERABLE gg CHANNELS APPLICABLE PER TRIP OPERATING

[

TRIP FUNCTION TRIP LEVEL SETTING FUNCTION MODES ACTION v

Rod Block Monitor *

%g Upscale (Power Refertnced) a)

Low Power Trip Setpoint 5 115/125 of full scale 2

1*

40 b)

Intermediate Power Trip Setpoint 5 109/125 of full scale ye c) High Power Trip Setpoint 5 105/125 of full scale Downscale 2 94/125 of full scale 2

1*

40 tR Inoperative NA 2

1*

40 APRM Flow-Biased Upscale Two loop operation:

4 1

41 s (0.58W + 50%)#

Single loop operation:

s (0.58W + 46.5%)#

PJ Upscale in Startup 5 12% of RATED THERMAI, POWER 4

2,5 41 na Inoperative NA 4

1,2,5 41 Downscale 2 5% of RATED THERMAL POWER 4

1 41 Intermediate Ranoe Monitors Detector not full in NA 4

2,5 41 Upscale 5 108/125 of full scale 4

2,5 41 Inoperative NA 4

2,5 41 Down sca le"'

2 5/125 of full scale 4

2,5 41 Source Ranae Monitors Detector not full [n*'

NA 3

2 41 2

5 41 Upsca l e*'

s 10' cps 3

2 41 2

5 41 I noperat ive*'

NA 3

2 41 2

5 41 Downscale*

2 3 cps 3

2 41 2

5 41 Scram.Discharoe Volume Water Level - High 5 24 gallons 1

1,2,5**

42 Recirculation Flow Upscale 5 110%

2 1

42 Inoperative NA 2

1 42 Comparator 5 10% flow deviation 2

1 42 Reactor Mode. Switch-Shutdown. Position NA 2

3,4 43 9

.m m

a e-m w

+.. ~, -

DAEC-1 Table 3.2-C (Continued)

CONTPOL POD BLOCK INSTPUMENTATION i

ACTION ACTION 40 -

Declare the RBM inoperable and take the ACTION required by Specification 3.3.C.3.

l ACTION 41 -

With the number of OPERABLE Channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

l b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one j

inoperable channel in the tripped condition within one hour.

ACTION 42 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

ACTION 43 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.

NOTES l

f With THERMAL POWER > 30% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed or withdrawn per Specification 3.9.A.

W is the recirculation loop flow in percent of design.

(a)

The RBM shall be automatically bypassed when a peripheral control rod is selected.

l t

(b)

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c)

This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

(d)

This function shall be automatically bypassed when the IRM channels are on range 3 or higher.

1 (e)

This function shall be automatically bypassed when the IRM channels are on range 1.

i Amendment No. 734.755.193 3.2-22

e Table 4.2-C COllTPOL POD BLOCV lilSTPUl1Ft3TATION SURVEILLANCE REOUIREMENTS OPERATIONAL g

CONDITIONS 5

CilANNEL

'FOR WHICil D

CilANNEL FUNCTIONAL Cil AtlN EL SURVEILLANCE TRIP FUNCTION CllECK TEST CAL I H RATIOtr**

REQUIRED 3

Rod Block Monitor 2

Upscale (Power Referenced)

D S / U**', Q"*

SA 1*

.U Downscale D

S / U**", Q "

SA 1*

g inoperative NA S / U**", Q" '

NA l'

w APPJi y

Flow Biased Upscale D

S/U*,Q Q

l m7 Upscale in Startup D

S/U*,Q Q

2,5 g

Inoperative NA S/U,Q NA 1,2,5 ta Downscale D

S / U **, Q Q

1 Intermediate Ranae Monitors Detector not full in NA S / U+', W R

7,5 Upscale D

S / U+', W Prior to Startup or Controlled Shutdown 2,5 Inoperative NA S / U+', W NA 2,5 N

Downscale D

S/U*,W Prior to Startup or Controlled shutdown 2,5 N"

Source Ranae Monitors Detector not ful1 i r, NA S / U*', W R

2,5 Upscale D

S / U+', W Prior to Start up or controlled shutdown 2,5 Inoperative NA S/U*',W NA 2,5 Downscale D

S/U*,W Prior to Startup or Controlled Shutdown 2,5 Scram Discharce Volume Water Level-High NA Q

R 1,2,5**

Recirculation Flow Upscale NA S/U*,Q SA 1

Inoperative NA S/U*,Q NA 1

Comparator NA S/U*,Q SA 1

Peace.or Mode Switch -

NA R

NA 3,4 Shutdown Position

- ~,,.,,.. -

v.

---,a

DAEC-1 1

I r

Table 4.2-C (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS H21El i

(a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

t (c)

Includes reactor manual control multiplexing system input.

With THERMAL POWER 2 30% of RATED THERMAL POWER..

With more than one control rod withdrawn.

Not applicable to control rods removed or withdrawn per Specification 3.9.A.

I 1

i i

i f

l i

Amendment No. E4.123,J/3,750,193 3.2-24 l

[

t I

I DAEC-I l EIMITING OONDITIONS FOR OPERATION SURVE7LLANCE REOUIREMEhTS D.

RADIATION MONITORING D.

RADIATION MONITORING INSTRUMENTATION INSTRUMENTATION r

1.

The radiation monitoring 1.

Each radiation monitoring l

instrumentation channels shall be instrumentation channel'shall be OPERABLE as shown in Table 3.2-D.

demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, CHANNEL CALIBRATION and SOURCE CHECK I

operations for the OPERATING MODES and at the frequencies shown in Table 4.2-D.

Applicability:

2.

LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all As shown in Table 3.2-D.

channels shall be performed at least

[

once per operating cycle for the Action:

following:

a.

With one or more radiation a.

Steam Jet Air Ejector Offgas Line monitoring channels inoperable, Isolation take the ACTION required by Table 3.2-D.

b.

Steam Jet Air Ejector Charcoal-Bed Bypass c.

Mechanical vacuum Pump Trip and Isolation 2.

In the event the noble gas flow in the air ejector offgas exceeds the equivalent of 1.0 Ci/see after 30 minutes delay in the offgas holdup line as indicated on the offgas pre-treatment radiation monitor, restore the rate to less than this limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Amendment No.

//,79).)pp.193 3.2-25

o Cshf Table 3.2-D co m BADIATION MONITORING INSTRUMENTATION a

MINIMUM APPLICABLE VALVE (s) m CHANNELS OPERATING ALARM / TRIP OPERATED INSTRUMENTATION OPERABLE MODES SETPOINT BY SIGNAL ACTION

.o Offgas Post-Treatment Radiation 1

(a)

(a) 50 QQ Monitors I*

Offgas Pre-Treatment Radiation 1

(b)

NA 51

.EI Monitors

$1 Main Steam Line Radiation Monitors 2

s 3X Normal (c)

,0*

Full Power gg

Background

.D'

}d

When the offgas system is cperating.

p

- Refer to Specification 3.7.F.

S (a)

The monitors shall be set to initiate immediate closure of the charcoal bed bypass valve and the air ejector offgas isolation valve at a setting equivalent to or below the dose rate limits in ODAM Section 6.2.2.1.

(b)

The monitors shall be set to initiate an alarm if the monitor exceeds a trip setting equivalent to 1.0 ci/sec of noble gases after 30 minutes delay in the offgas holdup line.

(c)

Trips Mechanical vacuum Pump which results in a subsequent isolation of the Mechanical vacuum Pump suction valves.

ACTION RADIATION MONITORING INSTRUMENTATION ACTION 50 - With the number of OPERABLE channels less than required by the Minimum Channels operable requirement, gases from the steam air eiector offgas system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided (1) the charcoal bed of the offgas system is not bypassed, and (2) the offgas stack noble gas activity monitor is operable.

Otherwise, be in at least HOT STANDBY within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 51 - With the number of OPERABLE channels less than required by the Minimum Channels Operable requirement, gases from the steam air ejector offgas system may be released for up to 30 days provided (1) the charcoal bed of the offgas system is not bypassai, (2) Grab samples are collected and analyzed weekly, and (3) the offgas stack noble gas activity monitor is OPERABLE or at least 1 post-treatment monitor is OPERABLE.

l Otherwise, be in at least HOT STANDBY within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

F l

Table 4.2-D 4

R

- Sl RADIATION MONITORING INSTRUMENTATION SURVFILLANCE REOUIREMENTS

?

U OPERATING MODES y

CHANNEL FOR WHICH w

CHANNEL FUNCTIONAL CHANNEL SOURCE SURVEILLANCE g

INSTRUMENTATION CIIECK TEST CALIBRATION CHECK REQUIRED L

Offgas Post-Treatment Radiation Monitors D

Q**

R H

y Offges Pre-Treatment Radiation Monitors D

Q**

R M

to L

Main Steam Line Radiation Monitors once/ shift Q

R R

w i

. Yy

When the offgas system is operating

- The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists Instrument indicates measured levels above the alarm / trip setpoint a.

b.

Instrument indicates a downscale failure c.

Instrument controls not set in the operate mode.

This channel functional test will consist of injecting a simulated electrical signal into the measurement channels.

l Refer to Specification 3.7.F.

T 5

3

-. =

w m

- -m m

.-.a

.....-a-,,w..-.

maww..

...d+--.3-w n.e-w..w.

.<w.ry,..,.r_reww-,

..co4,

.,m.m,,,-.,.w,-

3

.u w. - m.,,,ww.co m a e, ca=+,e.-ee es--emw-..=ww-ns*-ree--*+,w

", e e

i f

DAEC-1 l

1 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE PEOUIREMENTS i

I i

E.

DPYb' ELL LEAK DETECTION E.

DPYWELL LEAK DETECTION i

INSTRUMENTATION INSTRUMENTATION

)

2.

The drywell leak detection 1.

Each drywell leak detection i

instrumentation channels shall be instrumentation channel shall be OPERABLE as shown in Table 3.2-E.

demonstrated OPERABLE by the performance of the CHANNEL CHECK,

[

CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATING MODES and at the i

frequencies shown in Table 4.2-E.

I Aeolicability 1

As shown in Table 3.2-E.

r betion With the number of OPERABLE

[

channels less than required by the Minimum OPERABLE Channels requirement, take the ACTION required by Specification 3.6.C.

i i

l t

i

?

r i

?

5 i

t i

[

t i

I l

f f

i Amendment No. 2E*73, (3,193 3.2-2s l

y, Table 3.2-E U

DRYWELL LEAK DETECTION INSTRUMENTATION fl MINIMUM APPLICABLE c0 OPERA BLE OPERATING g

INSTRUMENT CHANNELS MODES ACTION Ok, Sump System *'

1 1,2,3*

60 va

$7' Air Sampling System *'

1 1,2,3*

60 ss w?

%?

ww!"

When irradiated fuel is in the vessel.

(a) The Sump System is comprised of the Equipment Drain Sump and Floor Drain Sump Sub-systems.

P3 The Equipment Drain Sump Sub-system consists of one Equipment Drain Sump Flow Integrator and two QJ Equipment Drain Sump Flow Timers.

The Floor Drain Sump Sub-system likewise consists of one Floor Drain Sump Flow Integrator and two Floor Drain Sump Flow Timers.

The Sump System is OPERABLE when no u3 any one of these six devices is OPERABLE.

(b) The Air Sampling System provides a backup system to the Sump System.

The Air Sampling System is OPERABLE when any one of the six available channels is OPERABLE.

ACTION ACTION 60 - See specification 3.6.C.

Table 4.2-E y

DRYWELL LEAK DETECTION INSTRUMENTATION SURVEILLANCE REOUlREMENTS a

a r+

OPERATING m

CilANNEL MODES FOR WilICll f

CilANNEL FUNCTIONAL CHANNEL SURVEILLANCE INSTRUMENT CHECK TEST CALIBRATION REQUIRED yw Equipment Drain Sump Flow Integrator D

NA Q

1,2,3*

Floor Drain Sump Flow Integrator D

NA Q

1,2,3*

8 Equipment Drain Sump Flow Timer NA Q

A 1,2,3+

Floor Drain Sump Flow Timer NA Q

A 1,2,3+

Air Sampling System D

M Q

1,2,3*

F N

d, When irradlated fuel is in the vessel.

O y

n n,

.c r---,,

,w u

,r e

l DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l

F.

SURVEILLANCE INSTRUMENTATION F.

SURVEILLANCE INSTRUMFNTATION 1.

The surveillance instrumentation 1.

Each surveillance instrumentation channels shall be OPERABLE as channel shall be demonstrated shown in Table 3.2-F.

OPERABLE by the performance of the CHANNEL CHECK and CHANNEL l

CALIBRATION operations at the frequencies shown in Table 4.2-F.

Aeolicability i

As shown in Table 3.2-F.

Action With the number of OPERABLE channels less than required by the Minimum Operable Channels requirement, take the ACTION required by Table 3.2-F.

l f

r i

l i

t t

b 1

r F

4 Amendment No.19.f// ))),193 3.2-31

Table 3.2-F SURVEILLANCE INSTRU M TATION o

MINIMUM TOTAL a

OPEliABLE CHANNELS INSTRUMENT TYPE / RANGE CilANNELS PROVIDED ACTION eo 2

Reactor Water Level Recorder, indicator 2

3 90

.O 158 to 218 Inches to Reactor Pressure Recorder, indicator 2

3 90 W

0-1200 psig 1

Drywell Pressure Recorder 2

2 90

-10 to +90 psig Drywell Temperature.

Recorder 2

8 90 0-350'F Torus Water Temperature Recorder 2

2 90 w

20-220'F fu 8

Torus Water Level Recorder 2

2 90

-10 to +10 Inches H;O Source Range Monitoring 10 ' to 10' cps 3*'

4 90 IRM/APRM O to 125%

2/ Trip *'

3/ Trip System 90 System

(a) The Source Range Monitors and Intermediate Range Monitors are not required in Operating i

Mode 1.

(b) These instruments are considered to be redundant to each other.

ACTIONS ACTION 90 -

a.

From and after the date that one of these parameters is reduced to one indication, when required, continued operation is permissible during the succeeding thirty days unless such instrumentation is sooner made OPERABLE.

b.

From and after the date that one of these parameters'is not indicated in the Control Room, continued operation is permissible during the succeeding seven days unless such-instrumentation is sooner made OPERABLE.'

c.

If the requirements of (a) and (b) cannot be met, an orderly shutdown shall be 1

initiated and the reactor.shall be in a COLD SHUTDOWN condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

<_-.,..w,--m

-#n,.,

.~ -

.w-r+-,.

.~-,~.,----,r.

.....e--

.-e..

r-

---w

.--e.---.-

y

-.-r.

sw.......--.

.= --..-.

,-n4

- - - - - - +. ~

---,.2.

Table 4.2-F SURVEILLANCE INSTRUMENTATION SURVEILLANCE REOUIREMENTS 6

J CilANNEL INSTRUMENT CHANNEL CHECK CALIBRATION bs Reactor Water Level Once/ Shift SA

' Ma)"

Reactor Pressure once/ Shift SA w

Drywell Pressure once/ Shift SA ga Drywell Temperature once/ Shift SA g;

Torus Water Temperature Once/ Shift SA n+.

Torus Water Level Once/ Shift SA Control Rod Position Once/ Shift NA w

k)

Average Powar Range Monitoring Once/ Shift *

(a)

In STARTUP or RUN Mode only.

(a) Prior to reaching 20% power and once per day when in RUN Mode (APRM Gain Adjust when in RUN Mode).

me.

c.

4

--.,--,4.

v

-~

g DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS f

l G.

RECIRCULATION PUMP TRIP (RPT) AND G.

RECIRCULATION PUMP TRIP (RPT) AND j

ALTERNATE ROD INSERTION (ARI)

ALTERNATE ROD INSERTIUN fARI)

INSTRUMENTATION INSTRUMENTATION l.

(ATWS) - RPT/ARI 1.

Each RPT and ARI instrumentation channel shall be demonstrated i

The instrumentation that trips OPERABLE by the performance of the the recirculation pumps and CHANNEL CHECK, FUNCTIONAL TEST and initiates ARI as a means of CHANNEL CALIBRATION operations at limiting the consequences of a the frequencies shown in Table 4.2-failure to scram during an C.

anticipated transient shall be OPERABLE as shown in Table 2.

LOGIC SYSTEM FUNCTIONAL TESTS and 3.2-G.

simulated automatic operation of all ATWS-RPT/ARI instrumentation channels shall be performed at least (EOC)-RPT cnce per operating cycle.

The instrumentation that trips 3.

Time response testing of the RPT

[

the recirculation pumps during breakers shall be performed at least i

stop valve or control ve.ive fast once per operating cycle.

closure for transi.c.!.na rgin improvement (especially at end-of-cycle) shall be OPERABLE as

}

shown in Table 3.2-G with the RPT SYSTEM TIME RESPONSE as shown in Table 4.2-G.

I Arolicability:

l As shown in Table 3.2-G Action:

With one or more (ATWS)RPT/ARI or

,(EOC)-RPT instrument channels inoperable, take the ACTION required by Table 3.2-G.

}

i i

i t

f i

i i

i r

t I

i i

I i

Amendment No.$bef47,IIYel93 3.2-34 i

'I

a R

m Table 3.2-G (ATWS)RPT/ARI AND EOC-RPT INSTRUMENTATION o

MINIMUM OPERABLE APPLICABLE

.s 43 CHANNELS PER OPERATING TRIP FUNCTION TRIP LEVEL SETTING TRIP SYSTEM

MODE ACTION (ATWS) RPT/ARI 5 1140 poig 2 "**

1 80 Reactor High Pressure (ATWS) RPT/ARI 2 +119.5 inches 2 * *'

1 80 Reactor Water Level-Low-Low (EOC) RPT Logic NA IN 1

81 Q)

(a)

There shall be one OPERADLE trip system for each parameter.

If this cannot be met, the w

indicated ACTION shall be taken.

us (b)

There are 2 trip systems. The instruments are arranged in a two-out-of-two once logic.

(c)

If an instrument (s) is(are) inoperable, it may be considered to be OPERABLE if placed in a tripped condition.

(d)

Two (EOC)RPT systems exist, either of which will trip both recirculation pumps.

(e)

If the duration of required surveillance testing for a RPT system exceeds two consecutive hours, the RPT system shall be declared inoperable.

ACTION ACTION 80 - a.

With one instrument channel inoperable, restore the inoperable instrument channel to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With both instrument channels inoperable, restore at least one instrument channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 81 - If both RPT systems are inoperable or if one RPT system is inoperable for more than 72 consecutive hours, an orderly power reduction shall be initiated and reactor power shall be less than 85% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

=, -

Table 4.2-G fATWSIRPT/ARI AND EOC-RPT INSTRUMENTATION SURVEILLANCE REQUIREMENTS a

8 3d OPERATING CHANNEL MODES FOR W111Cil 2

CHANNEL FU'4CTIONAL CHANNEL SURVEILLANCE P

TRIP FUNCTION CHECK TEST CALIBRATION REQUIRED O

(ATWS) RPT/ARI NA A

A 1

Reactor High Pressure w

(ATWS) RPT/ARI NA A

A 1

Reactor Water Level-Low-Low (EOC) RPT Logic NA M

NA 1

RPT Breaker NA R

NA 1

Y 7

W END-OF-CYCLE (EOC) RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME TRIP FUNCTION RESPONSE TIME RPT System s 140 msec

This response tirne is from initiation of Turbine Control Valve Fast Closure or Turbine Stop Valvo closure to actuation of the breaker secondary (auxiliary) contact.

.=

3 DAEC-1 LIMITING CONDITIONS FOP OPERATION SURVEILLANCE PEOUIREMENTS H.

ACCIDENT MONITORING H.

ACCIDENT MONITORING INSTRUMENTATION INSTRUMENTATION I

1.

The accident monitoring 1.

Each accident monitoring instrumentation channels shown instrumentation channel shall be shall be OPERABLE as shown in demonstrated OPERABLE by the l

Table 3.2-H.

performance of the CHANNEL CHECK and l

CHANNEL CALIBRATION operations at the frequencies shown in Table 4.2-H.

Applicability:

t As shown in Table 3.2-H.

t Action:

With one or more accident L

monitoring instrumentation channels inoperable, take the ACTION required by Table 3.2-H.

I y

h i

)

P j

k i

l t

)

t I

i i

s 1

i 1

2 Amendment No.IJA,193 3.2-37 i

l k

. - - =

Table 3.2-H ACCIDENT MONITORING INSTRUMENTATION

>k

)$

MINIMUM APPLICABLE CHANNELS OPERATING INSTRUMENT OPERABLE MODES ACTIOF Safety / Relief Valve Position Indicator (Primary Detector) 1/va lve

1,2,3 90 m

0 Safety Valve Position Indicator (Primary Detector) 1/ Va lve

1,2,3 90 y

m7 Reactor Coolant, Containment Atmosphere, and Torus Water Post-1(each) 1,2,3 91 g

Accident sampling Extended Range Effluent Radiation Monitors b

a)

Reactor Building Exhaust Stack 1

1,2,3 92 b)

Turbine Building Exhaust Stack 1

1,2,3 92 c)

Offgas Stack 1

1,1,3 92 N

Drywell Radiation Monitor rv 1

1,2,3 92 w

Torus Radiation Monitor 1

1,2,3 92 Drywell Pressure Monitor (0-250 psig) 2 1,2,3 93 Drywell Pressure Monitor (-5 to +5 psig) 2 1,2,3 93 Torus Water Level Monitor (1.5 to 16 feet) 2 1,2,3 94 Containment Hydrogen / Oxygen in-line Monitor 2*

1,2,3 95 t

i (a)

Each channel is comprised of three instruments (pressure switches) which are arranged in a two cut of three" logic.

(b) Normal condition is with monitor in Standby Mode.

4

.m..

--,v---,4.-

e-~..

..----r.

-u.-=

~,w-e v-

,,w--.

,,-wr we v--w.v.

1--v-

-w-.

-1

.-,v--

--,~c

DAEC-1 Table 3.2-H (Continuedi ACCIDENT MONITORING INSTRUMENTATION ACTION ACTION 90 - From and after the date that a channel is inoperable, the torus temperature will be monitored at least once per shift to observe any unexplained temperature increase which might be indicative of an open SRV; continued reactor operation is permissible only during the succeeding 30 l

days, unless such channel is sooner made OPERABLE.

ACTION 91 - When the ability to obtain a sample has been loats a.

Within 7 days, confirm a sample can be obtained within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the time a decision is made to sample; and b.

Within 90 days, restore the sampling capability.

l Otherwise, be in at least HOT SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l When the ability to analyze a sample has been lost:

l a.

Within 7 days, confirm that alternative sample analytical support

services can be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the time a decision is made to sample; and b.

Within 90 days, restore sample analysis capability.

Otherwise, be in at least HOT SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 92 - With the number of OPERABLE channels (both indicator and recorder inoperable) less than the Minimum Channels Operable requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , ands l

a.

Either restore the inoperable channel (s) to OPERABLE status within 7 days following the event, or s

l b.

Prepare and submit a Special Report to the Commission within 14 days following the event describing the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 93 - If the number of OPERABLE channels (both inuicator and recorder inoperable) is reduced to one channel, follow either step (a) or (b) below.

j a.

Operation may continue for the next 30 days provided at least one (1) channel of instrumentation specified in Table 3.2-F for the j

identical parameter is OPERABLE ** or follow step (c) below.

b.

Restore the inoperable channel to OPERABLE status within 7 days, should neither channel of instrumentation specified in Table 3.2-F for the identical parameter be OPERABLE, or follow step (c) below.

c.

Within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the c. ext 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SHUTDOWN.

[

If the number of OPERABLE channels (both indicator and recorder inoperable) is reduced to zero (e.g., no channels available) restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SHUTDOWN.

2he requirement for alternative analytical support for containment hydrogen and oxygen may be satisfied by verifying that at least one of the HP2 in-line monitors is on-line and capable of being lined up to each of the two sampling points (drywell and torus).

The instruments in Table 3.2-F which measure the identical parameters are the

-10 to 90 psig drywell pressure monitors.

Amendment No. 717,131,181,18f.193, 3.2-39

l DAEC-1 Table 3.2-H Lpontinued)

ACCIDENT MONITOPING INSTRUMENTATION 8.GI19H ACTION 94 - If the number of OPERABLE channels (both indicator and recorder inoperable) is reduced to one channel, follow either step (a) or (b) below.

l a.

Operation may continue for the next 30 days provided at least one torus water level channel and one containment water level l

channel is available.a**

If these conditions cannot be met, follow step (b) below.

b.

Operation may continue for the next 7 days if one torus water level channel is available and there are no other containment water level channels availnble.

If these conditions cannot be met, follow step (c) below.

c.

Within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SKUTDOWN.

If the number of OPERABLE channels (both indicator and recorder inoperable) is reduced to zero (e.g., no channels available) restore at least one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SHUTDOWN.

ACTION 95 - If the number of OPERABLE channels (both indicator and recorder inoperable) is reduced to one channel, follow either step (a) or (b) below a.

Within 30 days, increase the number of OPERABLE channels to the Minimum Number Channels Required or follow step (c) below, b.

Within 30 days, and at least once every 7 days thereafter, i

demonstrate the ability to obtain and analyze containment samples for hydrogen and oxygen or follow step (c) below.

If this sampling is done, but the number of OPERABLE channels is not increased to the Minimum Number Channels Required within 60 days from the time of initial loss, follow step (c) below, c.

Within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SHUTDOWN.

If the number of OPTRABLE channels (both indicator and recorder inoperable) is reduced to zero (e.g., no channels available) follow either step (a) or step (b) below.

a.

Restore at least one channel to OPERABLE status within 7 days or follow step (c) below.

)

i b.

Within 7 days, and at least every other day thereafter, demonstrate the ability ', obtain and analyze containment samples for hydrogen ana oxygen or follow step (c) below.

If this sampling is done, but the number of OPERABLE channels is not increased to one channel within 14 days from the time of 1

initisl loss, follow step (c) below.

c.

Within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months be in at least HOT STANDBY and within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months be in COLD SKUTDOWN.

i i

l Ine containment water level monitors provide indication from 0 to +98 feet.

i Amendment No. f6.l8/,193 3.2-40

s Table 4.2-H ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REOUIREMENTS T

APPLICABLE CilANNEL CHANNEL OPERATING g

INSTRUMENT CilECK CALIBRATION MODE to S

Safety / Relief Valve Position Indicator (Primary Detector)"

M R

1,2,3 i

Safety / Relief Valve Position Indicator (Backup-Thermocouple)

M R

1,2,3 f

Safety Valve Position Indicator (Primary Detector)*

M R

1,2,3 Safety Valve Position Indicator (Backup-Thermocouple)

M R

1,2,3 Reactor Coolant, Containment Atmosphere, and Torus Water NA A*

1,2,3 Post-Accident Sampling Extended Range Effluent Radiation Monitors a)

Reactor Building Exhaust Stack W

A" 1,2,3 b)

Turbine Building Exhaust Stack W

AM 1,2,3 h) c)

Offgas Stack W

A 1,2,3 M

u Drywell Radiation Monitor M

R 1,2,3 N

Torus Radiation Monitor M

R*

le2e3 3

1,2,3 Drywell Pressure Monitor (O to 250 psig)

M Drywell Pressure Monitor (-5 to +5 psig)

M g

1,2,3 p

1,2,3 Torus Water Level Monitor (1.5 to 16 feet)

M Containment Water Level Monitor M

R I'2*3 Containment Hydrogen / Oxygen In-line Monitor M

sam 1,2,3 NOTES (a)

Functional test of the relay is done once/3 months.

(b)

CHANNEL CALIBRATION shall consist of an electronic calibration of the channel for ranges above 10 R/hr and a one point calibration check of the detector below 10 R/hr with a portable gamma source.

(c)

Accident range effluent monitors shall be calibrated by means of a built-in check source or a known radioactive source.

(d)

Not a calibration, but a demonstration of system operability.

(e)

Monitors shall be tested for operability using standard bottled H: and 0.

DAEC-1 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS I

I.

EXPLOSTVE GAS MONITORING I.

EXPLOSIVE GAS MONITORING TRFTEUMENTATION INSTRUMENTATION 1.

A minimum of 2 offgas Hydrogen 1.

Each explosive monitoring Monitoring Instrument channels instrumentation channel shall be (R2)** shall be OPERABLE with demonstrated OPERABLE by:

their alarm / trip setpoints set to ensure that the limits of f

Specification 3.2.I.2 are not exceeded.

r Applicability a.

Daily channel check

During Offgas System Operation Action:

b.

Quarterly channel calibration

which.

ahall include the use of at least With an explosive monitoring /

two standard gas samples, each a.

instrumentation channel alarm containing a known volume percent trip setpoint less conservative hydrogen in the range of the than required by the above instrument, balance nitrogen.

I specification, declare the channel inoperable.

c.

Menthly Channel Functional Testa b.

With one channel OPERABLE, t

operation of the Offgas System may continue provided the recombiner temperature sensor is OPERABLE. When only one of the preceding methods is ourable the Offgas System may be opera,ted provided gas samples are i

collected at least once per day and analyzed for hydrogen within the ensuing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c.

If the minimum required instrumentation is not returned to OPERABLE status within 30 days, prepare and submit to the Commission within 30 days, pursuant to specification 6.11.3, a Special Report in lieu of any other report, why the instrument was not made operable in a timely manner.

2.

The concentration of hydrogen in the e

2.

The concentration of hydrogen in Offgas System shall be determined by the Offgas System downstream of monitoring the offgases in the the recombiners shall be limited Offgas System downstream of the i

to 5 4% by volume.

recombiners with the hydrogen monitors.

Applicabilitv:

i During Offgas System operation.

i Action:

a.

With the concentration of hydrogen in the main condenser offgas treatment system downstream of the recombiners i

exceeding the limit restore the i

concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

l t

b.

In the event the hydrogen concentration is not reduced to s 4% within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , be in at least HOT SHUTDOWN or within the limit within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

During Offgas System Operation Refer to CDAM Figure 3-1 for location of effluent monitoring point R2.

Amendment No.120.193 3.2-42

~

c

+, -

-w

DAEC-1 3.2 BASES In addition to reactor protection instrumentation which initiates a teactor scram, protective instrumentation has been p.ovided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious j consequences. The objectives of the Specifications are:

1.

To ensure the effectiveness of the protective instrumentation when required including periods when portions of such systems are out of service for maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

2.

To prescribe the trip settings required to assure adequate performance.

i Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal Leerating range to prevent inadvertent actuation of the safety system l

involved and exposure to abnormal situations.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The trip level settings given for reactor water level represent the indicated water level.

The reactor water level trip settings are defined or described in " inches" above the top of active fuel.

The term top of active fuel, however, no longer has a precise physical meaning since the length of the fuel pellet columns has changed over time from that of the initial core load.

Since the basis of all safety analyses is the absolute level (inches above vessel zero) of the trip settings, the " top of the active fuel" has been arbitrarily defined to be 344.5 inches above vessel zero.

This definition is the same as that given by Figure 5.1-1 of the Updated FSAR for the initial core and maintains the consistency between the various level definitions given in the FSAR and the technical specifications.

The low water level instrumentation set to trip at 170" above the top of the active fuel closes all isolation valves except those in Groups 1, 6, 7 and 9.

For valves which isolate at this level this trip setting is adequate to prevent uncovering the core in the case of a break in the largest line assuming a 60 second valve closing time.

Required closing times are less than this.

The low-low reactor water level instrumentation is set to trip when reaeter water level is 119.5" above top of the active fuel. This trip initiates the HPCI and RCIC and trips the recirculation pumps. The low-low-low reactor water level instrumentation is set to trip when the water level is 18.5" above j the top of the active fuel. This trip activate the remainder of the ECCS subsystems, closes Group 7 valves, closes Main Steam Line Isolation valves, Main Steam Drain Valves, Recirc Sample Valves (Group 1) and starts the Amendment No. 109,728,193 3.2-43

P 0AEC-1 emergency diesel generators.

These trip level settings were chosen to be high l enough to prevent spurious actuation but low enough to initiate ECCS opers icn and primary system isolation so that post accident cooling can be accomplithed and the guidelines of 10 CFR 100 will not be exceeded.

For large breaks up to lthecompletecircumferentialbreakofa22-inchrecirculationlineandwiththe trip setting given above, initiated in time to meet the above criteria.

Reference Sections 6.3 and 7.3 t

of the Updated FSAR.

The high drywell pressure instrumentation is a diverse signal for malfunctions j to the water level instrumentation and in addition to initiating ECCS, it causes isolation of Group 2 and 3 isolation valves.

For the breaks discussed l above, this instrumentation will gener4Lly initiate ECOS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also.

The water level instrumentation initiates protection for the full spectrum o.' loss-of-coolant accidents and causes isolation of all isolation valves except Group 6.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line.

For the worst case accident, main sicam line break outside the drywell, a trip setting of 140% of rated steam f2ow in conjunction with the flow limiters and consequently main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel clad temperatures peak at approximately 1000*F and release of radioactivity to the environs is below 10 CFR 100 guidelines.

Reference Subsection 15.6.5 of the Updated FSAR.

Temperature monitoring instrumentation is provided in the main steam line tunnel and turbine building to detect leaks in this area.

Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves, j

The setting is 200*F for the main steam line tunnel detector.

For large i

breaks, the high steam flow instrumentation is a backup to the temporsture j

instrumentation.

j i

High radiation monitors in the main steam line tunnel have been provided to j

detect gross fuel failure as in the control rod drop accident. A trip setting of 3 times normal full-power background is established to close the main steam line drain valves, recirculation loop sample valves, and trip the Mechanical vacuum Pump.

For changes in the Hydrogen Water Chemistry hydrogen injection rate, the trip setpoint may be adjusted based on a calculated value of the radiation level expected. Hydrogen addition will result in an increase in the nitrogen (N-16) activity in the steam due to increased N-16 carryover in the main steam.

Reference Subsection 15.4.7 of the Updated FSAR.

Pressure instrumentation is provided to close the main steam isolation valves l in the RUN Mode when the main steam line pressure drops below 850 psig.

The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the l

turbine bypass valves when not in the RUN Mode is less severe than the loss of f

feedwater analyzed in Subsection 15.6,3 of ths Updated FSAR, therefore, closure of the Main Steam Isolation valves for thermal transient protection l when not in the RUN Mode is not required.

l The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in j

Amendment No. 707,128,J82,193 3.2-44 1

i t

DAEC-1

{

actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1 out of 2 logic.

Temperature is monitored at two (2) locations with four (4) temperature

~,

sensors at each location. Two (2) sensors at each location are powered by "A" direct current control bus and two (2) by "B" direct current control bus.

Each pair of sensors, e.g.,

"A" or "B",

at each location are physically separated and the tripping of either "A" or "B" bus sensor will actuate HPCI isolation valves.

L The trip settings of 103 inches Hp (outboard instrument) and 386 inches H:0 (inboard instrument) which correspond to 300% of design flow for high flow and

(

i 175*F and A50* for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as L

that for the HPCI.

The trip setting of 155 inches Hp for high flow and 175'F IandA50* for temperature are based on the same criteria as the HPCI.

1 l The instrumentation which initiates ECCS action is arranged in a dual bus system.

As for other vital instrumentation arranged in this fashion, the specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed. An exception to this is when l logic system functional testing is being performed.

[

The controi rod block functions are provided to prevent excessive control rod withdrawal so that the MCPR does not decrease below the Safety Limit.

The trip logic for this function is 1 out of n:

e.g.,

any trip on one of six APRM's, six IRM's, or four SRM's will result in a rod block.

l 4

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criterion is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or i

calibration. This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow.

The APRM provides gross core protection; i.e.,

limits the gross core power increase.

from withdrawal of control rods in the normal withdrawal sequence. The trips l are set so that MCPR is maintained greater than the safety limit.

The RBM rod block function provides local protection of the core; i.e.,

the prevention of boiling transition in a local region of the core, for a single rod withdrawal error from a Limiting Control Rod Pattern.

The IRM rod block function provides local as well as gross core protection.

The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented. The downscale trips are set at 5 indicated on scale for APRM's and 5/125 full scale for IRM's.

Amendment No. 157,769,193 3.2-45 i

i

=

e DAEC-1 l

Both of the scram discharge volume high level channels provide input to the "B"

logic.

l The refueling interlocks cpe.7te one 1sgic channel, and are required for safety only when the mode switch is in the refueling position.

For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time.

The Automatic Depressurization System (ADS) is provided as a backup to HPCI.

The arrangement of the ADS logic is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met.

The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadrertent operation; i.e.,

only one instrument channel out of service.

Two air ejector offgas post-treatment monitors are provided. They are designed so that an instrument failure gives a downscale trip or an inoperative trip.

When both instruments reach an upscale trip point, or when one reaches an upscale trip point and the other reaches a downscale trip point or an inoperative trip, a trip is actuated. The post-treatment monitors have three upscale trip setpoints, one (Hi) to initiate charcoal bed bypass valve closure (CV-4134A open and CV-4134B closing to route offgas through the 6

charcoal) and another (Hi-Hi-Hi) to initiate offgas system isolation valve (CV-4108) closure. The third trip point (Hi-Hi) is for alarm initiation, and will initiate prior to the offgas isolation trip.

l Two sets of two radiation monitors are provided which initiate the Reactor Building Isolation function and operation of the standby gas treatment system.

Two instrument channels monitor the radiation from the refueling area ventilation exhaust ducts and two instrument charnels monitor the building ventilation below the refueling floor.

Trip settings of < 9 mr/hr for the monitors in.the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation and r

standby gas treatment system operation so that none of the activity released during the refueling accident leaves the Reactor Building via the normal i

ventilation path but rather all the activity is processed by the standby gas treatment system.

High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure.

In the event of a gross fuel failure, the established setting of 3 times normal full power background radiation levels (accounting for the N-16 carryover due to Hydrogen Water Chemistry) will trip the Mechanical Vacuum Pump, which in turn isolates the suction of the Mechanical Varuum Pump from the high and low pressure condensers. This prevents the elease of untreated fission products to the environment via the Mechanical Vacuum Pump.

Flow integrators are used to record the integrated flow of liquid from the I

drywell sumps. The alarm unit in each timer is set to annunciate before the values specified in Specification 3.6.C are exceeded. An air sampling system is also provided, as a backup to the sump system, to detect leakage inside the primary containment.

Amendment No.193 3.2-46 t

[

f DAEC-1 i

t l For each parameter monitored, as listed in Table 3.2.F, there are at least two (2) channels of instrumentation. By comparing readings between the two (2) l channels, a near continuous surveillance of instrument performance is i

available. Any deviation in readings will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

On July 26, 1984 the NRC published their final rule on Anticipated Transients Without Scram (ATWS), (10 CFR 550.62).

This rule requires all BWR's to make l

certain plant modifications to mitigate the consequences of the unlikely i

occurrence of a failure to scram during an anticipated operational transient.

i The bases for these modifications are described in NEDE-31096-P-A,

Anticipated Transients Without Scram; Response to NRC ATWS Rule, 10 CFR 50.62," December, 1985.

The Standby Liquid Control System (SLCS) was modified i

for two-pump operation to provide the minimum required flowrate and boron concentration required by the ATWS rule (see section 3.4 Bases). The existing ATWS Recirculation Pump Trip (RPT) was modified from a one-out-of-two-once logic to trip each recire. pump to a two-out-of-two-once logic to trip both recire. pumps, ("Monticello" design). This logic will also initiate the Alternate Rod Insertion (ARI) system, which actuater solenoid valves that bleed the air off the scram air header, causing the control rods to insert.

The instrument setpoints are chosen such that the normal reactor protection i

system (RPS) scram setpoints for reactor high pressure or low water level will be exceeded before the ATWS RPT/ARI setpoints are reached.

Because ATWS is considered a very low probability event and is outside the normal design basis for the DAEC, the surveillance frequencies and LCO requirements are less i

stringent than for safety-related instrumentation.

The End-of-Cycle (EOC) recirculation pump trip was added to the plant to improve the operating margin to fuel thermal limits, in particular Minimum i

Critical Power Ratio (MCPR).

The EOC-RPT trips the recire. pumps to lessen the severity of the power increases caused by either a closure of turbine stop

'i valves or fast closure of the turbine control valves with reactor power greater than 30% and a simultaneous failure of the turbine bypass valves to t

open.

The operating limit MCPR of section 3.12.C is calculated assuming an operable EOC-RPT system.

If the requirements of Table 3.2-G are not met, then the reactor power level is reduced to a level (85% of rated) which will ensure i

that the full power MCPR limits of section 3.12.C will not be violated if such a transient were to occur.

l i

1 The accident monitoring instrumentation listed in Table 3.2-H were specifically added to comply with the requirements of NUREG-0737 and Generic Letter 83-36.

The instrumentation listed is designed to provide plant status for accidents that exceed the design basis accidents discussed in Chapter 15 of the DAEC UFSAR.

l Action 94 of Table 3.2-H deviates from the guidance of Generic Letter 83-36 as f

continued operation for 30 days (instead of 7 days as recommended in the generic letter) is allowed with one of two torus water level monitor (TWLM) l channels inoperable.

Redundancy is available in that at least one channel of i

a Amendment No. 133 3.2-47 i

)

l o

DAEC-1

)

l the containment water level monitor (CWLM) instrumentation must be available.

{

Since the CWLM envelopes the span measured by the TWLM, the torus water level can be monitored by the CWLM system.

Therefore, continued operation is justified.

Main Condenser Offgas i

Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently

['

discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

Explosive Gas Mixture j Specification 3/4.2.I is provided to ensure that the concentration of I

potentially explosive gas in the offgas Treatment System downstream of the recombiners is maintained below the flammability limit sf a hydrogen and oxygen mixture in the syrtem. Keeping the mixture below its flammability limit will provide assurance that offgas Treatment System integrity and operability is maintained and that the radioactive material concentration in the offgas will be controlled in conformance with 10 CTR Part 50, Appendix A, criterion 60.

Calibration gas concentrations will be within the range of interest for hydrogen concentration and will not include Ot or 100% hydrogen concentrations.

t l Explosive Gas Monitoring Instrumentation shall be OPERABLE and inservice except that channels out of service are permitted for the purpose of required tests, checks, calibrations, and preventive maintenance without declaring the channel to be inoperable.

t i

l 3

I 1

l s

Amendment No. l93 3.2-48 l

i

DAEC-1 4.2 BASES The instrumentation listed in Table 4.2-A through 4.2-F will be functionally tested and calibrated at regularly scheduled intervals. The same design i

reliability goal as the Reactor Protection System of 0.99999 is generally I

applied for all applications of (1 out of 2) X (2) logic. Therefore, on-off sensors are tested once/3 months, and bi-stable trips associated with analog insors and amplifiers are tested once/ week.

2 hose instruments which, when tripped, result in a rod block have their contacts arranged in a 1 out of n logic, and all are capable of being bypassed. For such a tripping arrangement with bypass capability provided, there is an optimum test interval that should be maintained in order to maximize the reliability of a given channel (7).

This takes account of the fact that testing degrades reliability and the optimum interval between tests is approximately given by:

I=

Where i = the optimum interval between testo.

t = the time the trip contacts are disabled from perform.ng their function while the test is in progress.

r = the expected failure rate of the relays.

To test the trip relays requires that the channel be bypassed, the test made, and the system returned to its initial state.

It is assumed this task j

requires an estimated 30 minutes to complete in a thorough and workmanlike manner and that relays have a failure rate of 104 failures per hour. Using this data and the above operation, the optimum test interval ist i =) 2 (0. 5) =1x10 bours 8

10'

= 40 days For additional margin a test interval of once per month will be used initially.

The sensors and electronic apparatus have not been included here as these are analog devices with readouts in the control room and the sensors and electronic apparatus can be checked by comparison with other like instruments.

The checks which are made on a daily basis are adequate to assure operability of the sensors and electronic apparatus, and the test interval given above provides for optimum testing cf the relay circuits.

The above calculated test interval optimizes each individual channel, considering it to be independent of all others. As an example, assume that there are two channels with an individual technician assigned to each. Each technician tests his channel at the optimum frequency, but the two technicians are not allowed to communicate so that one can advise the other that his channel is under test.

Under these conditions, it is possible for both i

channels to be under test simultaneously.

Now, assume that the technicians are required to communicate and that two channels are never tested at the same tLme.

Amendment No.193 3.2-49

DAEC-1 C

Forbidding simultaneous testing improves the availability of the system over that which would be achieved by testing each channel independently. These one out of n trip systems will be tested one at a time in order to take advantage of this inherent improvement in availability.

optimizing each channel independently mry not truly optimize the system considering the overall rules of system operation. However, true system optimization is a complex problem. The optimums are broad, not sharp, and optimizing the individual channels is g6nerally adequate for the system.

The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the availability is illustrated by Curve No. 1 of Figure 4.2-2 which assumes that a channel has a f ailure rate of 0.1 x 104/ hour and that 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months is required to test it.

The unavailability is a minimum at a test interval 1, of 3.16 x 103 hours0.00119 days 0.0286 hours 1.703042e-4 weeks 3.91915e-5 months .

If two similar channels are used in a 1 out of 2 configuration, the test interval for minimum unavailability changes as a function of the rules for testing. The simplest case is to test each one independent of the other.

In this case, there is assamed to be a finite probability that both may be bypassed at one time.

This case is shown by Curve No. 2.

Note that the unavailability is lower as expected for a redundant system and the minimum occurs at the same test interval.

Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability.

A more usual case is that the testing is not done independently.

If both channels are bypassed and tested at the same time, the result is shown in curve No. 3.

Note that the minimum occurs at about 40,000 hours0 days 0 hours 0 weeks 0 months , much longer than for cases 1 and 2.

Also, the minim 2m is not nearly as low as Case 2 which indicates that this method of testing does not take full advantage of the redundant channel.

Bypassing both channels for simultaneous testing should be avoided.

The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested and restored. This is shown in Curve No. 4.

Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.

The best test procedure of all those examined is to perfectly stagger the tests.

That is, if the test interval is four months, test one or the other channel every two months. This is shown in curve No. 5.

The difference between Cases 4 and 5 is negligible. There may be other arguments, however, that more strongly support the perfectly staggered tests, including reductions in human error.

The conclusions to be drawn are these:

1.

A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval.

2.

More than one channel should not be bypassed for testing at any one t ime.

Amendment No.193 3.2-50

ML20056H404

Contents

Text

Attachment:

4.1 REFERENCES

Background

Navigation menu

ML20056H404

Text

Attachment:

4.1 REFERENCES

Background

Navigation menu

Search