ML20086N180

Amends 80,72,86 & 81 to Licenses DPR-19,DPR-25,DPR-29 & DPR-30,respectively,approving Tech Spec Changes Providing Addl Limiting Conditions for Class IE Power Monitoring Sys

ML20086N180
Person / Time
Site:	Dresden, Quad Cities, 05000000 (DPR-19-A-080, DPR-19-A-80, DPR-25-A-072, DPR-25-A-72, DPR-29-A-086, DPR-29-A-86, DPR-30-A-081, DPR-30-A-81)
Issue date:	02/14/1984
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20086N184	List: ML20086N180
References
NUDOCS 8402170276
Download: ML20086N180 (28)

Text

_. _.

I f* "% 4'e 4

UNITED STATES

[ %,g[:[gg NUCLEAR REGULATORY COMMISSION 5

s w ;9 s J V!ASHINGTON, D. C. 20555

% k M,5 COMMONWEALTH EDIS0N COMPANY DOCKET N0. 50-237 DRF3DE1 NUCLEAR P01!ER STATION, UNIT NO. 2

1. EENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 80 License No. DPR-19 1.

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

A.

The application for amendment by Comnonwealth Edison Company (the licensee) dated August 31, 1983, complies with the standards and reovirements of the Atonic 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 provisiens 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 activitics will be conducted in compliance with the Commission's regulations; D.

The-issuance of this emendment will not be inimical to the connon defense and security or to the health and safety of the public; ar.d E.

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

requirements have been satisfied.

9402170276 840214-POR ADOCK 05000237 4

P PDR

V 2.

Accordingly,ithe license is amended by changes to the Technical Specifications as indicated in the_ attachment to this license amendment and paragraph 3.B of Provisional Operating License No. DPR-19 is hereby amended to-read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment flo. 80, are hereby incorporated i

in the license. The. licensee shall operate the facility

~

in accordance with the Technical Specifications.

3.

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

FOR:THE NUCLEAR REGULATORY COMMISSION aiu

?

enni,s M. Crutchfie dj/ hief Operating Reactors Branch #5 Division of Licensing Attachrent:

Changes to the Technical

+

Specifications Date of Issuance:

February 14, 1984 I

e f

s n

=

gy

-9 e-rr7Y e-+F "r-

+-e'-*

ATTACHiiENT TO LICEf;SE A!'El:Df'ENT f!0S. 80 At'D 72 OPERATING LICENSE NOS. DPP-19 AND DPR-25 DOCKET N05. 50-237/249 Replace the following pages of the Technical Specifications with the encicsed pages. The revised pages are identified by the captioned amendment numbers and contain vertical lines indicating the area of change.

PAGES 22 22a 28 28a*

31 4

'This page is included for pagination purposes only; there are no chances to the: provisions contained thereon.

t-

.....-L.-.n.-..-.--~.~.--

DPR-19 t

3.I L1111T11!G CONDITION FOR OPERATI0tl 4.I SURVEILLANCE REQUIREMENT 3.I REACTOR PROTECTION SYSTEtt 4.1 REACTOR PROTECTION SYSTEM Applicability:

Applicability:

Applica to the instrumentation and associated devices which initiate a reactor scram.

Applies to the surveillance of the instrumentation and associated devices which initiate reactor scram.

Objective:

Objective:

To acaure the operability of the reactor protection ayaten.

To specify the type and frequency of surveillance to be applied to the protection instrumentation.

Specifications:

Specifications:

lA. AEACTOR PROTECTION SYSTEM lA. REACTOR PROTECTION SYSTEM

,l.

The setpoints, minimum number of trip systems, l1. Instrumentation systema shall be func-and minimum number of instrument channels that must be operable for each position of the tionally tested and calibrated as indi-reactor mode awitch shall be as given in cated in Tables 4.1.1 and 4.1.2, Table 3.1.1.

The system response times from respectively.

the opening of the sensor contact up to and including the opening of the trip actuator l2. Daily during reactor power operation 8'

contacte shall not exceed 50 milliseconds.

above 25% rated thermal power, the core power distribution shall be checked for:

l2.

If during operation, the maximum fractica of limiting power density for fuel fabricated by Maximum fraction of limiting power a.

GE exceeds the fraction of rated power when density for fuel fabricated by CE operating above 25% rated thermal power, (HFLPD) and compared with the frac-tion of rated power (FRP).

j either:

l I

l a.

The APRH scram and rod block settings shall lb. For compliance with assumptions of be reduced to the values given by the equa-the Fuel Design Analysis of over-tions in Specifications 2.1.A.I and 2.1.B.

power conditions for fuel fabricated This may be accomplished by increasing APRH by ENC.

gains as described therein.

l l

Amendment No. 75, 80 22

...\\..- -

m __ _ - _

.-.~...-.w.-

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

DPR-19 3.I LI!!ITI!!G CONDI710N FOR OPERATIO!!

4.I SURVEILLANCE REQUIREHENT lb. The power distribution shall be changed auch 3.

that the maximum fraction of limiting power The RPS power monitoring system instrumenta-l 4

density no longer exceeds the fraction of tion shall be determined OPERABLE:

rated power.

At least once per 6 months by performing j

a.

For fuel fabricated by ENC, operation of the a CilANNEL FUNGTIONAL TEST, and core shall be limited to ensure the power b.

At distribution is consistent with that assumed least once per operating cycle by in the Fuel Desi n Analysis for overpower demonstrating the OPERABILITY of over-8 conditions, voltage, undervoltage, and underfrequency protective instrumentation by performance 3.

Two RPS electric power monitoring channels for of a CilANNEL CALIBRATION including simu-cach inservice RPS HG set or alternate source lated automatic actuatica of the shall be OPERABLE at all times.

protective relays, tripping logic, and output circuit breakers, and verifying the 4.

With one RPS electric power monitoring channel following setpoints:

for an inservice RPS HG set or alternate power supply inoperable, restore the inoperable channel Surveillance Requirements:

?

Reactor Protection Buses to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the associated RPS HG set or alternate power supply (1) Overvoltage 126.5V 1 2.5%

from service.

Hin. 123.3V 5.

With both RPS electric power monitoring channels Hax. 129.6V for an inservice RPS 10 set or alternate power (2) Undervoltage 108V 1 2.5%

supply inoperable, restore at least one to OPERABLE status within 30 minutes or remove the Hin. 105.3V associated RPS HG set or alternate power supply Max. Il0.7V f rom service.

(3) Underfrequency 56.0 llz + 1% of 60 Itz i

Min. 55.4 IIz Max. 56.6 112 i

)

il I

l AmendmentNo.J5',80 22a

........-.....-..----c..

a.-.

=2

.---.--------r-Danent 3.1 The reactor protection ayotem autonatically intti-has a reliability greater than that of a 2 out of $~

ntes a reactor scram to syatem and somewhat Icas than that of a 1 out of 2 system.

l. Preserve the integrity of the fuel cladding.

2. Preacrve the integrity of the primary ayatem.

With the exception of the Average Power Range Honitor(APRH) and Intermediate Range Monitor Minimize the ener8y which nuat be abaarbed, (IRH) channels, each subchannel has one instru-3.

ment channel. When the minimum condition for and prevent criticality following a lona of coolant accident, operation on the number of operable instrument channela per untripped protection trip systen la This apecification provides the liniting conditions met or if it cannot be mot and the effected protec for operation necessary to preserve the ability of tion trip syntes la placed in a tripped condition the ayatem to tolerate single failures and still the effectiveness of the protection system la perform its intended function even during periods preserved; i.e., the ayatem can tolerate a single when instrument channela may be out of service failure and still perform its intended function of scramming the reactor. Three APRH instrument because of maintenance. When necessary, one channel ray be made, inoperable for brief intervals to conduct channels are provided for each protection trip system.

gw required functional tests and calibrations.

g.

4.., The reactor protection ayatem is of the dual channel APRH'a #1 and #3 operate contacts in a one sub-channel and APRH's #2 and #3 operate contacts in

!"i type.

Re f. Section 7.7.1.2 SAR. The ayates is the other subchannel. APRH's #4, #5, and #6 are

", nade up of two independent trip systems, each arranged similarly in the other protection trip having two subchannela of tripping devices. Each syntes.

~ subchannel has an input from at least one instru-Each protection trip system has one more APRM than la necessary to meet the minimum gent channel which monitors a critical parameter.

number required per channal. This allows the by-The outputa of the subchannels are combined in a '

passing of one APRH per protection trip system l'out qf 2 logic; i.e., an input algnal on either for maintenance, testing or calibreition. Additional IRH channels have also been provided to allow for one or both of the subchannels will cause a trip bypanning of one auch channel. The bases for the ayates trip. The outputa of the trip ayatema are arranged so that a trip on both ayatens la required scram settings for the IRM, APRH, high reactor to produce a reactor scram.

pressure, reactor.Iow water level, generator load rejection, and turbine stop valve closure are discussed in Specification 2.3.

]

Specifications are provided to ensure the operability

)

of the RPS Bus Electrical Protector Asocablies-(EPA's).

l f

Each channel from either overvoltage, undervoltage, Instrumentation (pressure switches) in the dryweli or under frequency will trip the associated HG act or are provided to detect a loss of coolant accident 1

alternate power source.

and. initiate the emergency core cooling equipment.

l This instrumentation la a backup to the water level Instrumentation which la discussed in Specification This synten rcets the requirenents of the proposed 2.2 A scram la provided at the same setting as IEEE Standard for Nuclear Power Plao* Protection the amargency core cooling system (ECCS) initia-Systema issued Septentar 13, 1966 The system tion to minimise the anargy which must be accon-rodated during a loss of coolant accident and to Amendment flo. 80

I e

e as et t

9 4

'f 1

.)

4 1

O e

.f 1

6 6

g i

I

.I ht C

l2o M

M

?

O y

M

,E C

f u

i u

l W

5*

U e

3 g

.C O

CC 1

1 s

j

$4 0

u Y

U IC 43b

• O CO M

O **

.:: 5 U "3, 6a O

l

=

U tJ Z

C u o eg 4

0 C

C C S

O s

ka 4

o- *e.

E O. a y

C w-8-9 cy

}.,,, ' e 2

E as

~ ~~

s

.a.

'*O

,r n

e,.

r

_.h

.ii

+, _ _.

_us,.+

m_

_.4 4

+

I I

nanen:

4.1

.Q 3.

Devicca which only serve a useful function 1

A.

The minimum functional testing frequency used during some restricted mode of operation, l

such as startup or shutdown, or for which in this specification la based on a reliability

'the only practical test is one that can be i

analysis using the concepts developed in refer-performed at shutdown.

ence (6). This concept was specifically adapted I

to the one out of two taken twice logic of the The sensors that make up group (A) are speci-reactor protection ayatem for Dresden 3.

The fically selected from among the whole family analyala shows that the sensors are primarily of industrial on-off sensors that have earned responsible for the reliability of the reactor protection ayeten. This analysis makes uso an excellent reputation for reliable operation.

ot "unanin failure" rate experience at conven-Actual history on this class of sensors operat-ing in nuclear power plants shows 4 failures in tional and nuclear power plants in a reliability 472 sensor years, or a failure rate of 0.97 X i

riodel for the system. An " unsafe failure" is 10-6/hr.

During design, a goal of 0.99999 defined an one which negates channel opera-probability of. success (at the 50% confidence bility and which, due to its nature, la revealed level) was adopted to assure that a balanced 4

only when the channel la functionally tested or and adequate design is achieved. The probability attempts to 'espond to a real signal.

Failures

.}

s.

r

!, * - such as blown fuses, ruptured bourdon tubes, of success is primarily a function of the sensor C., faulted amplifiers, faulted cables, etc., which failure rate and the test interval. A three-

}

.I result in "upacale" or "downscale" readings month test interval was planned for group (A)

".) on the reactor instrumentation are " safe" and This la in keeping with good operating i'

sensors.

will be easily recognized by the operators practice, and satisfies the design goal for the logic configuration utilized in the Reactor g

"during operation because they are revealed Protection System.

,by an alarm or a scram.

Surveillance requirements are provided for the

'To satisfy the long-term objective of maintaic-ing an adequate level of safety throughout the RPS EPA's to demonstrate their operability. The plant lifetime, a minimum go.al of 0.9999 at setpoints for overvoltage, undervoltage and under frequency have been chosen based on analysia.

the 95% confidence level is proposed. With the

~

(1 out of 2) X (2) logic, this requires that each (Reference T. Raush letter to 11. Denton 02-04-83).

sensor have an availability of 0.993 at the 953 confidence level. This level of availability The 13 channels listed in Tables 4.1.1 and 4.1.2 are Jiivided into three groups respectin5 may be maintained by adjusting the test inter-val as a function of the observed failure q

functional testing. These ares his' tory (6). To facilitate the implementation

l]

1.

On-Off sensora that, provide a scram trip It func tio.n.

6.

Reliability of Engineered Safety Features as a

]p 2.

Analog devicca coupled with bi-stable Function of Testing Frequency, I.H. Jacobs,

]

trips that provide a sepam function.

Nuclear Safety, vol. 9. No. 4, July-Aug. 1968.

pp. 310-311 I

Amendment No. 80 l

31

[pn Rtc u

'Q g f(k UNITED STATES g

NUCLEAR REGULATORY COMMISSION j

5.

,"&'., E'/ly. E WASHINGTON, D. C. 20555 s ':;

E ms C0tE0NWEALTH EDIS0N COMPANY DOCKET N0. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Anendment No. 72 License No. DPR-25 1.

The Nuclear Reculatory Cornission (the Commission) has found that:

A.

The acplication for anendment by Comnonwealth Edison Company (the licensee) dated August 31, 1983, complies with the standards and recuirements of the Atonic 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 Comnission; C.

There is reasonable assurance (i) that the activities authorized by this anendment 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 anendment will not be inimical to the commor. defense and security or to the health and safety of the public; ard E.

The issuance of this amendment is in accordance with 10 CFR Part El of the Ccmission's regulations and all applicable recuirements have been satisfied.

. - ~.

I 2-i 2.

Accordingly, the license is anended by changes to the Technical i

Specifications as indicated in the attachment to this license amendnent and' paragraph.3.B of Facility Operating License l

No. DPR-25 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 72, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION M

Dennis k Crutchfield, hief

.s i

Operating Reactors Br nch #5 Division of Licensing

Attachment:

i Changes to_the Technical Specifications i

Date of Issuance:

February.14, 1984 i.

c w

a w

,,m v

m

=

i y-g T---

-e q

- 1st

-4 w

-g r-

--y'w

~

ATTACHMENT TO LICENSE f!'E!!DMENT NOS. 80 A!'D 72 OPERATIi;G LICENSE NOS. DPP-19 AND DPR-25 DOCKET NOS. 50-237/249 Replace 'he folicwing pages of the Technical Specifications with the enclosed pages.

The revised pages are identified by the captioned ctendment numbers and contain vertical lines indicating the area of change.

PAGES 22 22a 28 28a*

31

~

e

' is rcge is ire.iu ed 'er pagination pur,oses only; there are no chances to v

r,r c/c i; i( :

c.r-f re? therect.

l

i

...y DPR-25 3.I LilllTillG CO!!DITIO!I FOR OPERAT1011 4.1 SURVEIhLANCE REQUIREMENT 3.1 REACTOR PROTECT 1011 SYSTEtt 4.1 REACTOR PROTECTION SYSTEM

_ Applicability:

Applicability:

Applies to the. instrumentation and associatcd devices which initiate a reactor scram, Applies to the surveillance of the ina'trumentation and associated devices which initiate reactor scram.

Objective:

Objective:

5 To assure the operability of the reactor protection To specify the type and frequency of surveillance system.

to be applied to the protection instrumentation.

Specifications:

t Specifications:

lA. REACTOR PROTECTION SYSTEM lA. REACTOR PROTECTION SYSTEN l

I.

The. setpoints, minimum number of trip systems,.

l.

Instrumentation systems shall be func-and minimum number of instrument channels that tionally tested and calibrated as indi-1; must be operable for each position of the cated in Tables 4.l.1 and 4.1.2, reactor mode switch shall be as given in respectively.

Table 3.1.1.

The system response times from the opening of the sensor contact up to and 2.

including the opening of the trip actuator Daily during reactor power operation-ei above 25% rated thermal power, the core j.[

contacts shall not exceed 50 milliseconds.

pow'er distribution shall be checked fort f2.

If during operation, the maximum fraction of limiting power density for fuel fabricated by l *a. Maximum fraction of limiting power GE exceeds the fraction of rated power when density for fuel fabricated by GE operating above 25% rated thermal power, (HFhPD) and compared with the frac-either:

tion of rated power (FRP).

la. The APRM scram and rod block settings shall l b.

For compliance with assumptions of be reduced to the values given by ti.e equa-the Fuel Design Analysis of over-tions in Specifications 2.1.A.i and 2.1.B.

power conditions for fuel fabricated by EHC.

This may he accomplished by incretasing APHH o

gains as described therein.

i

.l

'i M

Amendment No.,42, 63, 72 M

22

..n.

a.

... i....

't - a -- ~ ~ ~ D -. A s'i.

"*s---

DPR-25 3.I lit!ITING CONDITION FOR OPERATION 4.I SURVEILLANCE REQUIRE!!ENT b.

The power distribution shall be changed such 3.

The RPS power monitoring system instrumenta-that the maxiruss fraction of limiting power tion shall be determined OPERABLE:

density no longer exceeds the fraction of rated power.

At least once per 6 months by perforning a.

l a CilANNEL FUNCTIONAL TEST, and

} [

For fuel fabricated by ENC, operation of the core shall be limited to ensure the power b.

At least once per operating cycle by distribution is consistent with that assumed demonstrating the OPEP. ABILITY of ovor-in the Fuct Design Analysis for overpower voltage, undervoltage, and underfrequency conditions, protective instrumentation by perfornance j l of a CilAllHEL CALIBRATION including simu-3.

Two RPS electric power moni.toring channels for lated automatic actuation of the 5

each inservice RPS HG set or alternate source protective relays, tripping logic, and shall be OPERABLE at all times, output. circuit breakers, and verifying the following setpoints:

4.

With one RPS electric power monitoring channel for an inservice RPS HG set or alternate power Surveillance Requirements:

supply inoperable, restore the inoperable channel Reactor Protection Buses to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove the j

associated RPS HG set or alternate power supply (1) Overvoltage 126.5V 1 2.5%

from service.

Min. 123.3V 5.

With both RPS electric power monitoring channels for an inservice RPS HG set or alternate power (2) Undervoltage 108V 1 2.51 supply inoperable, restore at least one to Hin. 105.3V OPERABLE status within 30 minutes or remove the Max. Il0.7V associated RPS HG set or alternate power supply frors service.

(3) Underfrequency 56.0 tiz + II of 60 Ils Hin. 55.T IIz Hax. 56.6 Ils j

i i

' 72 Amenilment 11 0.

j

,63, 22a

I.

~. -. n

s...... s u n.

w.~...

s...s..?

..~.-n.~-a---.x-~-w..--s---. ~*~-~'..s*v

--~'.s Dawda :

I

,3.1 The reactor protection ayatem autonatically initi-nten a reactor scram to has a reliability greater than that ofa2outofh system and somewhat leas' than that of a 1 out of 2 ayatem.

1. Preserve the integrity of the fuel cladding.

t

2. Preservo the integrity of the primary ayatem.

With the exception of the Average Power Range Honitor(APRH) and Intermodlate Range Honitor Minimizu the energy which soust be absorbed, (IRH) channels, each subchannel has one instru-3.

ment channel. When the minimum condition for and prevent criticality following a loss of coolant accident, operation on the number of operable instrument channels per untripped protection trip system is This specification providea the lietiting conditions met or if it; cannot be mot and the ef fected protec for operation necessary to preserve the ability of tion trip system is placed in a tripped coadftion the system to tolerate single failures and still the effectiveness of the protection system is preservedi 1.e., the system can tolerate a single perform its intended function even during periods when instrument channels may be out of servico failure and still perform its intended function of scramming the reactor. Three APRH instrument because of maintenance. When necessary, one channel rsay be made inoperable for brief intervals to conduct channels are provided for each protection trip system.

.}

required functional tests and calibrations.

'f.

The reactor protection system is of the dual channel APRH'a #1 and #3 operate contacts in a one sub-i..,

channel and APRH's #2 and #3 operate contacts.in It type. Ref. 'Section 7.7.1.2 SAR.

The system is the other subchannel. APRH's #4, #5, and #6 are

'- nado up of two independent 'rlp' systems, each arranged similarly in the other protection trip h.aving two subchannels of tripping devices. Each

~ aubchannel has an input from at least one instru-Each prutection trip system has one more

. system.

,veent channel which monitors a critical parameter, APRH than is necessary to meet the minimum number required per channel. This allows the by-The outputa of the subchannels are combined in a

• passin'g of one APRH per protection trip system for maintenance, testing or calibrdtion. Additional l'out of 2 logic; i.e., an input signal on either IRH channels have also been provided to allow for one or both of the oubchannels will cause a trip bypassing of one such channel. The bases for the oystem trip. The outpute of the trip systems are scram settings for the IRM, APRM, high reactor arranged no that a trip on both systems is required to produce a reactor scram.

pressure, reactor, low water level, generator load rejection, and turbine stop valve closure are Specificatione are provided to ensure the operability of the RPS Ilua Electrical Protector Assemblies (EPA's).

Each ciannel from either overvoltage, undervoltage, Instrumentation (pressure switchen) in the drywell or under frequency will trip the associated HG set or are provided to detect a loss of coolant accident alternate power source.

and; initiate the emergency core cooling equipment.

{

This instrumentation is a backup to the water level This systen poeta tha requirements of the proposed instrumentation which is discussed in Specification 2.2 IEEE Standard for Nuclear Power Piso* Protection A scram is provided at the same setting as the emergency core cooling system (ECCS) initia-Systens issued Septemter 13, 1966 The systen tion to minimise the energy which must be accon-modated during a loss of coolant accident and to J'

1 Amendment No; 72,

w

w s

4-,

4 m.

4 w

a m

=

4 e

en N

e g

l 3

e s.

t g

1 8

1 e

(:

a g

+

9 4

4 e

6

)

O e

t e

e j

4

.O a

)

a h

i a

e t

gg s

w D

o M-H O

~

f M

e Of.

e4 e

b O

I g

s W

h'

• 7..

g C

I e e e.

4 g

-e u w

e e

Es.

p as g

g O

{

k,

N u

N et - g

=

, ~

. Ae r

e g ag 4.*"

o r

g Z:

MO

- /s*t =

)

Cg p

E-

. r- ~

- ).

t e.

o e.

y l.

.w c e-E t

C. 4J C

e 0

. t.

=

E-

... e-l.

4 f. a ~ ~

a..*.

a

.i be e.*

e

• eq*"

e g

, y a

g-e

~,

e,.

-W l

~. _.....

. _ _..u. _

a.

_.... ~...,

..m._,,__,._m_c..,_.,_,,____,,_

y. :.

.(

4.1 3.

Devices which only serve a useful function i"

A.

The nininun functionni tanting frequency uoed during some restricted rodo of operation, in thin specification la booed on a reliability

auch as startup or shutdown, or for which analyain using the coucopta developed in refor-the culy practical test la one that can be performed at shutdown.

ence (6). Thia concept was specifically adapted to the one out of tuo taken twice logic of the The nennora that nake up scoup (A) are spect-reactor protection nyaten for Dresden 3.

The fically ablected from among the whole family analysia nhows that the acusara are primarily of industrial on-off. sensors that have carned reaponsible for the reliability of the reactor protection ayaten. Thia cualyala makes uso an excellent reputation for reliable operation.

ot " unsafe failura" rata expertonce at conven-Actual history on this class of sensors operat-ing in nuclear power plants showa 4 failures in tional and nuclear power planta in a reliability 472 sensor years, or a failure rate of 0.97 X rodel.for the ayaten. An " unsafe failure" in 10 6/hr.

During design, a goal of 0.99999 defined an one which negates channel opera-probability of success (at the 50% confidence bility and which, due to its nature, la revealed level) was adopted to assure that a balanced only when the channel la functionally tested or and adequate dealan la achieved. The probability ij S n'ttempts to raapond to a resi signal.

Failures if auch as blown fuses, ruptured bourdon tubes, of success la primarily a function of the acnaar failure rate and the test interval.

A threo-C, faulted amplifiars, faulted cables, etc., which month tent interval was planned for group (A)

I result in="upacale" or "downscale" readinga on the reactor instrumentation are " safe" and This in in keeping with good operating

.D sensors.

will be easily recognized by the operators practice, and antisfies the design goal for the r

durind op,eration because they are revcaled logic configuration utilized in the Reactor

.by an alern or a scran.

Protection System.

8' Surveillance requirements are provided for the To satisfy the long-term objective of maintain-ing an adequate level of safety throughout the RPS EPA's to demonstrate their operability. The plant lifetime, a minimum goal of 0.9999 at netpoints for overvoltage, undervoltage and under the 95% c'ontidence level is proposed.

With the frequency have been chosen based on analyala.

(L out of 2) X (2) logic, this requires that each (Raferenca 'f. Rauah letter to H. Denton 02-04-83).

acnaar have an availability of 0.993 at the 95%

4 confidence level. This level of availability The 13 channela listed in Tables 4.1.1 and may be maintained by adjusting the test inter-4.1.2 are dividad into three groups respecting val an a function of the observed failure functional testing. Theno are history (6). To facilitate the implementation 5

1.

f On-off acnsora that provide a scram trip function.

6.

Reliability of Engineered Safety Features na a l

j 2.

Analog devices coupled with bi-e able Tunct,lon of Testing Fraquency, I.H. Jacobo, trips that provida a sepam function.

Nuc*. ear Safety, Vol. 9 No. as, July-Aug. 1968.

pp. 310-312. -

Amendment No. 72

1. p e atauq

. fo UNITED STATES

'E g NUCLEAR REGULATORY COMMISSION 3 ghh~,.i),3 n

WASHINGTOrv, 0. C. 20555 pqWf 4.....,

COMMONWEALTH EDIS0N COMPAf4Y A!;D IOUA-ILLIN0IS GAS 775 ELECTRIC COMPANY DOCKET N0. 50-254 CUAD CITIES STATION, UNIT N0. 1 Af1ENDMENT TO FACILITY CPERATING LICENSE Amendment flo. 86 License No. DPR-29 1.

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

A.

The applicatinn for amendment by Ccnnonwealth Edison Company (the licer.see) dated August 31, 1983, complies with the standards crd requirenents of the Atom 1c Energy Act'of 1954, as amended (the Act), and the Ccmmission's rules and regulatiens set forth in 10 CFR Chapter I; P.. The facility will operate in ccnformity with the application, the previsions of the Act and the rules and regulationr of the Commission; C.

There is reasonable assurance (i) that the activ' ties authorized by this amendnent 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 comen defense and security or to the health and safety of the public; erd E.

The issuance of this anendment is in accordance with 10 CFR Part 51 o' the Commission's regulations and all applicable reauirements hrve been satisfied.

1 i_

2'-

2.

Accordingly, the. license is anended by changes to the Technical Specifications as indicated in the attachment to this license 1

amendment and paragraph 3.B of Facility Operating License No. DPR-29 is hereby amended to read as'follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B,

- as revised through Amendment No. 86, are hereby incorporated in~the license. The licensee shall operate the facility in accordance with the Technical Specifications.

i 3.

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

E CL COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 8

Division of Licensing Attachtent:

Changes to the Technical Scecifications Date of Issuance: February 14, 1984 I

f 1

f L

l

ATTACHl'ENT TO LICENSE Af!ENDI>ENT NOS. 86 AND al FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 DOCKET NOS. 50-254/265 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by the captioned amendment numbers and contain vertical lines indicating the area of change.

PAGES 3.9/4.9-4 3.9/4.9-5 3.9/4.9-6

=

^

QUAD-CITIES OPR-29 reactor shall be In the cold shut-Verifying de energi:stien of a.

down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

the emergency buses a d lead 2.

Specification 3.9.E.1 shall not shedcIng frem the emergency

buses, apply when a diesel gencrator has been made Inoperable for a period b.

Verifying the diesel starts'frem not to exceed 1-1/2 hours for the purpose of conducting preventative am: lent c:nditi:n en the aute-maintenance. Additionally, pre

• start signal, energizes the ventative maintenance shall not be errergency buses with ;:er anently undertaken unless two offsite lines c:nnected 1 cads, energizes the are available and the alternate ay o.c:nnecten emergency loads diesel generator has been demon" thrcugh the 1:ad sequencer, strated to be operable.

and c;erates fer greater than 5 minutes while its generator is 3.

When the reactor Is in the Cold

$hutdown or Refueling mode, a mini-nas of one diesel generator (either the Unit diesel generator or the Unit I/2 diesel generator) shall be operable' whenever any work is being done which has the potential for draining the vessel, secondary con-tainment Is required, or a core or containment cooling system is required.

F.

REACTOR PROTECTICN SUS POVER MONITCRING F.

REACTOR PROTECTION BU$ PCWER MONITORING system SYSTEM l.

Two RPS electric power eenitoring I.

The RPS Bus power monitoring system channels for each inservice RPS MG Instrwentattun shall be determined set or Inservice alternate power CPERAB.E:

source shall be OPERABLE except when the reactor is in the $HUTDOW f

L.

A ;least once per 6 noeths by a.

t rode.

performing a channel functional test, and 2.

a.

Vith one RPS electric power monl-tering channel for an Inservice b.

At RPS M3 set or Inservice alternate least once per operating cycle by demonstrating the operability power source Inoperable, restore the Incperable channel to CPERA8LE of overvoltage, undervoltage, anc underfrequency protect ive instru-status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove sentation by parformance of a the associated RPS MG set or al-channel calibration including ternate power source from service.

simulated automatic activation of b.

With both RPS electric power mont-the protective relays, tripping logic, and output circult break-toring channels for an Inservice RPS MG set or Inservice alternate ers, and verifying the following setpoIntst power source Inoperable, restore at least one channel to OPERA 8LE (1) Overvoltage 126.5 V - 2 5%

status within 30 minutes, or re-ecve the associated RPS MO set or Min. 123 3 y Max. 129 6 V alternete power source from service.

(2) Undervoltage 108 V = 2.5%

Min. 105 3 V Max. 1I0 7 V (3) Underfrequency 56.0 Hz k 1%

of 60 fz Min. 55.4 Hz

. Max. 56.6 Hz 3 9/4 9-4 4

Acen &ent Mo. 77, 86

QUAD-CITIES DPR-29 39 LIMITING CONDITIONS FOR OPERATION BASES r)

/

A.

The general objective of this specification is to assure an adequate source of electrical power to operate the auxiliaries during plant operation, to operate facilities to cool and lubricate the plant during shutdown, and to operate the engineered safety features following on accident. There are two sources of electrical energy available, namely, the 345-kV transmission system and the di-l esel generators.

B.

The d-c supply is required for control and motive power for switchgear and en-gineered. safety features.

The electrical power required provides for the maxi-mum availability of power, i.e., one active offsite source and or.e backup source of offsite power and the maximum numbers of onsite sources.

C.

Auxiliary power for the Unit is supplied from two sources, either the Unit auxill-ary transformer or the Unit reserve auxiliary transformer.

Both of these trans-forcers are sized to carry 100% of the auxiliary load.

If the reserve auxiliary transformer is lost, the unit can continue to run for 7 days, since the Unit aux 111.ary transformer is available and both diesel generators are operational.

A 7-day period is provided if one source cf offsite power is lost.

This period is based on having two diesels operable which are adequate,to handle an acclient assuming a sing 1'e failure.

In addition, auxiliary power from the other unit can be obtained through the 4160-volt bus tie.

If both offsite lines are lost, power is reduced to 40% of rated so that the turbine bypass system could accept the steam flow without reactor trip should the generator be separated from the systcm or a turbine trip occur.

In this condition, the turbine generator is capable of supplying house load and ECCS load if necessary through the unit auxillary tranc-

!, -l former.

If the unit were shut down on. loss of -both lines, fewer sourc~es of power would be available than for sustained operation at 40% power.

Attention will be given to restoring normal offsite power to minimize the length of time i

operation is allowed in a ccndition where both sources are available.

4

. In the normal mode of operation, the 345-kV system.is oper:ble and two diese' generators are operable. One diesel generator may be allowed out of service for a short period of time to conduct preventative maintenance provided that power is available from the 345-kV system through a 4160-volt bus tie to supply the emergency buses, and the alternate diesel generator is proven operable. Offsite power is quite reliable, and in the last :25 years there has been only one instance in which all offsite power was lost at a Commonwealth Edison Generating Station.

When the unit or shared diesel generator Is made or found inoperable for reasons other than preventative maintenance, the remaining diesel generator and its j

associated low pressure core cooling and containment cooling systems, which pro-vide sufficient engineered safety features equipment to cover all breaks, will be proven operable.

D.

The diesel fuel supply of 10,000 gallons will sup' ply each diesel generator with a minimum of 2 days of full load operation or about 4 days at 1/2 load. Additi-onal diesel fuel can be obtained and delivered to the site within an 8-hour 4

period; thus a 2-day supply provides-for adequate margin.

1 E.

Diesel generator operability is discussed in Paragraph 3.S.C above.

t tI,--

F.

Specifications are provided to ensure the operability of the RPS Bus. electrical I _?

protection assemblics (EPA's).

Each RPS MG set and the. alternate power source has 2 EPA channels wired in series.

A trip of either channel from either over-i voltage, undervoltage, or underfrequency will trip the* associated MG set or alternate power source.

1.

3 9/4.9-5 Amendment No. 86

. ~--+ --

.,.....,.n.,,.~,.,

,.,.-v-

+..-,...~.-.- n - -

s CUAD-CITIES OPR-29 49 $URVEILLANCE REqulREMENTS BA$ES m

A.

The monthly test of the diesel generator is conducted to check for equipment failures and deterioration.

conditions to demonstrate proper operation at thise conditions.,. The dieselTest will be manually started, synchronized to the bus and load picked up.

The diesel shall be loaded to at least half load to prevent fouling of the engine.

it is expected that the diese! generator will be run for I to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Olesel-generator experience at other Conronwealth Edison generating stations Indicates that the testing frequency is adequate and provides a high rellability of opere-tion should the system be required. In addition, during the test. the generator Is synchronized to the offsite power sources and thus not completely Independent of this source.

To maintain the maximum amount of Independence, a 30 day test-Ing Interval is also desirable.

Each diesel generator has two air compres, sors and four air tanks.

Two air tanks are piped together to form an air receiver. Each air compressor supplies an air receive r.

This arrangement provide $ redundancy in starting capability. It is expected that the ' air congressors will run only Infrequently.

During the monthly check of the diesel, the receivers will be drown down below the point at whIch the compressor automatically starts to check operation and the ability of the compressors to recharge the receivers. Pressure Indicators are provided on each of the receivers.

Following the monthly test of the diesels, the fuel oil day tank will be approxi-mately half full based on the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> test at full load and 205 sph at full load.

At the end of the monthly load test of the elesel generators, the fuel oil trans-fer pumps will be operated to refill the day tank and to check the operation of these pumps from the emergency source.

The test of the emergency diesel generator during the refueling outage will be mere comprehensive in that It will functionally test the system, i.e..

It will check diesel starting, closure of diesel breaker, and sequencing of loads on the diesel.

The diesel will be started by simulation of a less-of coolant accident.

In addition, an undervoltage condition will be imposed to simulate a loss of i

the time required.

The only load on the diesel is that due to friction and s.-

windage and a small amount of bypass flow on each pump.

y Perledic tests between refueling outages verify the ability of the diesel to run at full load and the core and containment cooling pumps to deliver full flow.

Periodic testing of the various components plus a functional test at the refuel-Ing Interval are sufficient to nelntain adequate sellability.

S.

Although station batterias will deteriorate with time, utility experience Indl-cates there Is almost no possibility of precipitous failure. The type of sur-veillance described In this specification is that which has been demonstrated over the years to provide an Indication of a cell becoming irregular or unservice-able long before it becornes a failure.

In addition, the checks described also provide adequate Indication that the batteries have the specified ampere-hour capability.

C.

Eecause the availability of electricity to the system is a normal operating function a check of the status of these systems provides adequate surveillance.

D.

The diesel fuel oil cuality must be checked to ensure proper operation of the diesel generators.

Vater centent should be minimized, because water in the fuel would contribute to excessive corrosion of the system, causing decreased rell-ability. The growth of micro organisms results in slime formations, which are one of the chief causes of jellying in hydrocarbon fuels.

Minimizing of such silres is also essential to assuring high reliability.

E.

Olesel generater operability surveillance Is discussed in Paragrach 4.9.A above.

F.

Surveillance recuirerwnts are provided for the RP5 EPA's to demonstrate their operability. The setpoints for overvoltage, undervoltage and uncerfrequency have bean chosen based on analysis (ref. Febr6ary 4.1983 letter to M.

from T. Rausch).

Centon 39/4.9-q d

AT.encent No. 86

e rsauq

+f

. jo,}

UNITED STATES i

5%

• [

NUCLEAR REGULATORY COMMISSION

% % '

e wasair:GTON, D. C. 20555 A *...*p

~ ~; /,

g COMMONUEALTH EDISON COMPANY 4

AND IOWA-ILLIN0IS GAS 702T ELECTRIC COMPANY DOCKET NO. 50-265 OUAD CITIES STATION, UNIT NO. 2 1

AMENDMENT T0 FACILITY OPERATING LICENSE Amendment No. 81 Lice'nse No. DPR-30 1.

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

A.

Ti'e application for amendment by Commonwealth Edison Company (the licensee) dated August 31, 1983, complies with the standards-and requirements of the Atomic Energy Act'of 1954, as atended 4

(the Act), and the Commission's rules and regulations set forth-in 10 CFR Chapter I; B. - The facility will operate in conforaity with the application, the provisiens of the Act and the_ rules and regulationslof the Commission; C.

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

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

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

i

~.

2-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendnent and paragraph 3.B of-Facility Operating License No. DPR-30 is hereby amended to read as follows:

B, Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 81, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLE R REGULATORY C0lillISSION

'I 4

Donenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

i Chances to the Technical Specifications Date of Issuance: February 14, 1984 i

't i-4.g t

v.

-. - - -~

m- - - - -,.,-- +

,--yr f

--w e-,

~c.

--n,,-,

w-r---.

3, 3

y

ATTACH!'U:T TO LICEf!SE A!!Ef'CI'EfiT I:0S 86 Af;D al FACILITY CPERATIl!G LICEf:SE h'0S. DPP-29 AlO DPR-30 DOCKET f05. 50-254/265 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by the captioned arendment nutbers and contain vertical lines indicating the area of change.

PAGES 3.9/4.9-4 3.9/4.9-5 3.9/4.9-6 E

e t

DPR-30 risctor shall be in th:n cold shut-Verifying de energization of a.

down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

the emer.gency bus'es, and Icad 2.

Specification 3 9.E.1 shall not shedding from the emercency

~

buses.

apply when a diesel generator has been made inoperable for a period b.

Verifying the diesel starts from not to exceed 1-1/2 hours for the purpose of conducting preventative ambient ccndition on the auto-maintenance. Additionally, pre-start signal, energizes the ventative maintenance shall not be emergency bus'es with permanently undertaken unless two offsite l,nes connected loads, energizes the i

are available and the alternate auto-connected emergency bads diesel generator has been demon-through the load sequencer, strated to be operable.

and operates for greater than 5 minutes while its generator is 3

When the reactor is in the Cold loaded with the emergency loads.

Shutdown or Refueling mode, a mini-mum of one diesel generator (either the Unit diessi generator or the Unit 1/2 diesel generator) shall be operable whenever any work is being done which has the potential for draining the vessel *, secondary con-tainment is required, or a core or containment cooling system is required.

F.

REACTOR PROTECTION BUS POWER MONITORING F.

REACTOR PROTECTION BUS POWER MONITORING SYSTEM SYSTEM 1.

Two RPS electric power monitoring 1.

Th RPS Bus power monitoring system channels for each inservice RPS MG Instrumentation shall be determined set or Inservice alternate power OPERABLE:

source shall be OPERABLE except when the reactor is in the SHUTDOWN a.

At least once per 6 months by mode.

performing a channel functional test, and 2.

a.

With one RPS electric power moni-toring channel for an inservice b.

At RPS MG set or Inservice alternate least once per operating cycle by demonstrating the operability power source inoperable, restore the inoperable channel to OPERABLE of overvoltage, undervoltage and underfrequency protective itstru-status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or remove mentation by performance of a the associated RPS MG set or al-channel calibration including ternate power source from service.

simulated automatic activation of b.

With both RPS electric power moni-the protective relays, tripping logic, and output circuit break-toring channels for an inservice RPS MG set or inservice alternate ers, and verifying the following setpoints:

power source inoperable, restore at least one channel to OPERABLE (1) Overvoltage 126.5 V - 2 5%

status within 30 minutes, or re-Min. 123 3 y move the associated RPS MG set or Max. 129.6 y alternate power source from service.

(2) Undervoltage 108 V - 2.5e s

Min. 105 3 V Max. I10 7 V (3) ;:nderfrecuency 56.0 Hz :2.: l' of 60 H:

l Imndrent No. 71, 81 3 9/4.9-4 Min 55.t 9:

I eax. 5 6. 6 H..

1

UFR-30 3.9 LIMITING CONDITIONS FOR OPERATION BASES

-4, A.

The general objective of this specification is to assure an adequate source of electrical power to operate the auxiliaries during plant operation, to operate s-l facilities to cool and lubricate the plant during shutdown, and to operate the enginaered safety features following an accident.

There are two sources of electrical energy available, namely, the 345-kV transmission system and the di-esel generators.

B.

The d-c supply is required for control and motive power for switchgear and en-gineered safety features.

The electrical power required provides for the maxi-i mum availability of power, i.e., one active offsite source and one backup source of offsite power and the maximum numbers of onsite sources.

C.

Auxiliary power for the Unit is supplied from two sources, either the Unit auxili-ary transformer or the Unit reserve auxiliary transformer.

Both of these t.ans-formers are sized to carry 100% of the auxiliary load.

If the reserve auxiliary trans former-is lost, the unit can continue to run for 7 days, since the Unit auxiliary.ansformer is available and teth diesel generators are operational.

A 7-day-period is provided if one source of offsite power is lost. 'This period is based on having two diesels operable which are adequate to handle an accident assuming a single: failure.

In addition, auxillary power from the other unit can be obtained through the 4160-volt bus tie.

If both offsite lines are lost, power is reduced to 40% of rated so that the turbine bypass system could accept the steam flow without reactor trip should the generator be separated from the systen or a turbine trip occur.

In this condition, the turbine generator is capable of supplying house load and ECCS foad if necessary through the unit auxiliary tranc-former.

If the unit were shut down on loss of _both lines, fewer sources of g,-

power would be available than for sustained operation at 40% power.

Attention s.

will be given to restoring normal offsite power to minimize the length of time operation is allowed in a condition where both sources are available.

In the normal mode of coeration, the 345-kV system is operable and two diesel generators are operable.

One diesel generator may be allowed out of service for

'a short period of time to conduct preventative maintenance provided that power is available from the 345-kV system through a 4160-volt bus tie to supply the emergency buses, and the alternate diesel generator is proven operable.

Offsite power is quite reliable, and in the last 25 years there has been only one. instance in which all offsite power was lost at a Commonwealth Edison Generating Station.

1 When the unit or shared diesel generator is made or found inoperable for reasons other than preventative maintenance, the remaining diesel generator and its associated low pressure core cooling and containment cooling systems, which pro-vide sufficient engineered safety features equipment to cover all breaks, will be proven operable.

D.

The diesel fuel supply of 10,000 gallons will supply each diesel generator with

-a minimum of 2 days of full load operation or about 4 days at 1/2 load.

Additi-onal diesel fuel can be obtained and delivered to the site within an 8-hour period; thus a 2-day supply provides for adequate' margin.

E.

Diesel generator operability is discussed in Paragraph 3 9.C above.

F.

Specifications are provided to ensure the operability of the RPS Bus electrical

(

protection assemblies (EPA's).

Each RPS MG set and the. alternate co'wer source i

has 2 EPA channels wired in series.

A trip of either channel from eitner over-voltage, undervoltage, or underfre,quency will trip the associated MG set or alternate power source.

3 9/4.9-5 Amendment No. 81

QUAD $0

• lTIES crR-k.9 $URVEILI.ANCE REqulREMENT5 BASES A.

The renthly test of the diesel generator is conducted to check for equipment failures and deterioration. Testing Is conducted up to ecullibrium operating conditions to demonstrate proper operation at these conditions. The diesel will be manually started, synchronized to the but, and load picked up.

The diesel shall te loaded to at least half load to prevent foullng of the engine.

It is expected that the diesel generator will be run for I to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Diesel-generator experience at other Comonwealth Edison generating stations indicates that the testing frequency is adeewate and provices a high reliability of opera-tion should the system be required. In addition, during the test. the generator Is synchronized to the of fsite power sources and thus not completely independent of this source. To maintain the maximum amount of independence, a 3D-day test-Ing Interval is also desirable.

Each diesel generator has two air compressors and four air tanks. Two air tanks are piped together to form an air receiver. Each air compressor supplies an air re ce i ve r.

This arrangement provides redundancy In starting capability. It is expected that the air compressors will run only Infrequently.

During the monthly c5eck of the diesel. the receive s will be drawn down below the point at which the coepressor automatically starts to check operation and the ability of the cor.oressors to recharge the receivers. Pressure indicators are provided on each of the receivers.

Following the conthly. test of the diesels, the fuel oil day tank will be approml-mately half full based og the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> test at full load and 205 sph at full load.

At the end of the renthly 'oad test of the diesel generators. the fuel oil trans-fer pumps will be operated to refill the day tank and to check the operation of these pumps from the emergency source.

The test of the emergency diesel generator duricg the refueling outage will be core co-prehensive in that.it will functionally test the system. I.e..

It will check diesel starting, closure of diesel breaker, and secuencing of load; on the

( ',

~

diesel. The diesel will be started by simulation of a loss of-coolant accident.

in addition, an undervoltage condition will be imposed to sim~ulate a loss of

' ~

the time recuired. The only load on the diesel is that due to friction and windage and a small amount of bypass flow on each puro.

Periodic tests between refueling outages verify the abl.llty of the diesel to run at full load and the core and containment coollag pumps to deliver full flow.

Period 3c testing of the various components plus a functional test at the refucl-eng Interval are sufficient to rubintain adequate reliability.

6.

Although station batteries will deteriorate with time, utility experience indl-cates there is almost no possibility of precipitous failure. The type of sur-t veillance described in this specification is that which has been demonstrated over the years to provice an indication of a cell becoming irregular or unservice-able long before it becomes a failure.

fn addition, the checks described also provide adequate Indication that the batteries have the specified ampere-hour capability.

C.

Because the availability of electricity to the system is a normal operating function. a check of the status of these systems provides adequate surveillance.

D.

The diesel fuel oil cuality must be checked to ensure proper operation of the diesel generators. Vater content should be riinimized. because water in the fuel would contribute to excessive corrosion-of the system. causing decreased cell-ability. The growth of micro-organisms results in slime formations, which are one of the chief causes of Jellying in hydrocarbon fuels. Minimizing of such slines is also essential to assuring high reliability.

E.

Diesel-generator operability surveillance Is discussed in Paragraph 4.9.A above.

F.

Surveillance recuirements are provided for t e EPA's to demonstrate their h

operability. The setpoints for overvoltage, vio'e voltage. and underf reawency have been chosen based on analysis (ref. Februp, 4.1983 letter to M. Denton from T. Rausch).

s 3 9/4.9-6 Amendment No. 81