Text

Proposed Amend 72 Changing Tech Specs Sections 3.0,4.0 & 6.0 Re Emergency Power Supply Requirements,Valve Position Indicating Instrumentation for Inadequate Core Cooling, Containment Isolation & Auxiliary Feedwater Sys
	ML19347C388
Person / Time
Site:	Rancho Seco
Issue date:	10/08/1980
From:	; SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	;
Shared Package
ML19347C366	List: ML19347C365; ML19347C388;
References
	NUDOCS 8010170531
	Download: ML19347C388 (23)
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PAtlC110 SECO UtilT 1 TECHillCAL SPECIFICATIONS Limiting Conditions for Operation 3

LIMITillG C0ilDITl0fl5 FOR OPERATI0fl 3.1 REACTOR C00LAllT SYSTEM Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limi ting condi tions for opera tior of the reactor coolant sys tem which must be met to ensure safe reactor operations.

3.1.1 OPERATIOilAL COMPO!1EllTS Specification 3.1.1.1 Reactor Coolant Pumps A.

Pump combinations permissible for given power levels shall be as shown in specification table 2.3-1.

D.

The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

C.

Operation at ' power with two pumps shall be limited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 72 in any 30 day period.

3.1.1.2 Steam Generator A.

One steam generator shall be operable whenever the reactor coolant average temperature is above 280 F.

3.1.1.3 Pressurizer Safety Valves A.

The reactor shall not remain critical unless both pressurizer code safety valves are operable.

B.

When the reactor is subcritical, at least one pressurizer code safety valve shall be opercble if all reactor coolant system openings are closed, except for hydrostatic tes ts in accordance with ASME Boiler and Pressure Vessel Code, Section Ill.

3.1.1.4 Electronctic Poerated Valva The electrom.~ tic operated valve (EMOV) and its associated block valve shall be OPERABLE'.

A.

With the EMOV inoperable, within I hour either restore the 72 f

Proposed Amendment flo. 72

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Condi tions for Operation EMOV to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

B.

With the block valve inoperable, within I hour either restore the block valve to OPERABLE status or close the block valve and remove power from the block valve; otherwise, be in at Icast HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases A reactor coolant pump or decay heat removal pump is required to be in operation before tr.a boron concentration is reduced by dilution with makeup water.

Either pump will provide mixing which will prevent sudden posi tive reactivi ty changes caused by dilute coolant reaching the reactor.

One decay heat removal pump will circula e)the equivalent of the reactor coolant system volume in one half hour or less.

I The decay heat removal system suction piping is designed for 300'F and 300 psig; thus, the system ca7 eggye decay heat when the reactor coolant system is below this tempera ture.

One pressurizer-code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that re-quired by the sum of the available heat source which are pump energy, pressurizer heaters, and reactor decay heat. (4)

Both pressurizer code saf,ety valves are re-1 quired to be in service prior to criticality to conform to the system design relief capabi i The code safety valves prevent overpressure for rod withdrawal acci-dents. 5) ties.The pressurizer code safety valve lift set point shall be set at 2500 psig + 1 percent allowance for error and each valve shall be capabl'e of relieving 345,000 lb/h of saturated steam at a pressure not greater than 3 percent above the set pressure.

Two pump operation is limited until further ECCS analysis is performed.

The electromatic operated valve (Eh0V) onerates to relieve RCS pressure below the setting of the pressurizer code safety valves.

This relief valve 72 has a remotely operated block valve to provide a posi tive shutof f capability should the relicf valve become inoperabic., The electrical power for both the relief valve and the block valve is capable of being supplied from an caergency power source to ensure the ability to seal this possible RCS leakage path.

REFERENCES (1)

FSAR tables 9.5-2, 4.2-1, 4.2-2, 4.2-4, 4.2-5, 4.2-6 (2)

FSAR paragraph 9.5.2.2 and 10.2.2 (3)

FSAR' paragraph 4.2.5 (4)

FSAR paragraph.4.3.8 A and 4.2.4 (5)

FSAR paragraph 4.3.6 and 14.1.2.2.3 3-2 Proposed Amendment No. 72

RANCHO SECO UtilT 1 TECHillCAL SPECIFICATIO!!S Limiting Conditions for Operation 35 INSTRUMENTATI0ft SYSTEMS 3.5.1 OPERATIONAL SAFETY lilSTRUMENTATION Applicability Applies to unit instrumentation and control systems.

Objective To delineate the conditions of the unit ins trumentation and safety circui ts neces-sary to assure reactor safety.

Specifications 3.5.1.1 Startupand operation are not permi tted unless the requirements of table 3. 5.1-1, Columns A and B are me t.

3.5.1.2 in the event the number of protection channels operable falls below the limit given under table 3.5.1-1, Columns A and B, operation shall be limited as specified in Column C.

In the event the number of Process Instrumentation channels is less than the Total Number of Channel (s), restore the inoperable channels to operable status within 7 days, or be in at least hot shutdown wi thin the next 12 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

If the number of operable channels is less than the minimum chan-nels operable, either restore the inoperable channels to cperable within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.5.1.3 ror on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each reactor protection channel will. be used to lock the channel trip relay in the untripped state as indicated by a light.

Only one channel shall be locked in this untripped state at any one time.

3.5.1.4 The key operated shutdown bypass switch associated with each reactor protection channel shall not be used during reactor power operation.

3.5.1.5 During startup.vhen the intermediate range instrument comes on scale, the overlap between the intermediate range and the source range instru-mentation shall nut be less than one decade.

If the overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved.

3.5.1.6 In the event that one of the trip devices in either of the sources supplying power to the control rod drive mechuaisms fails in the un-tripped state, the power supplied to the rod drive mechanisms through the failed trip shall be manually removed within 30 minutes.

The con-dition will be corrected and the remaining trip devices shall be tested within eight hours.

if the condition is not corrected and the remaining i

trip devices are not tested wi thin the eight-hour period, the reactor l

shall be placed in the hot shutdown condition within an additional four l

hours.

3-25 Proposed Amendment No. 72

RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Condi tions for Operation Bases Every reasonabic effort will be made to maintain all safety instrumentation in operation.

A startup is not permitted unless three power range neutron instru-ment channels and two channels cach of the following are operable:

four reactor coolant tempera ture ins trument channels, four reactor coolant flow instrument cnannels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and fcur high reactor building pressure instrument channels.

The safety features actuaution system must have two analog channels functioning correctly prior to startup.

Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Calumn B (table 3.5.1-1).

This is in agreement wi th redundancy and single failure criteria of IEEE 279 as described in FSAR section 7 There are four reactor protection channels.

Normal trip logic is two out of four.

Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other ins trumen tation channels is one out of two.

The four reactor protection channels were provided with key operated bypass switches interlocked to allow on-line testing or maintenance on only one channel at a time during' power operation.

Each channel is provided alarm and lights to indicate when that channel is bypassed.

Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used.

There are four shutdown bypass keys in the control room under the administrative control of the shif t supervisor.

The keys will not be used during reactor power operation.

The source range and intermediate range nuclear flux instrumentation scales over-lap by one decade.

This decade overlap will be achieved at 10-10 amps on the intermediate range scale.

l Power is normally supplied to the control rod drive mechanisms from two s'cparate l

parallcl 480 volt sources.

Redundant trip devices are employed in each of thesc sources.

If an" n"-

of these trip devices falls in the untripped state on-line 1

repairs to the 1.

vice, when practical, will be made, and the remaining trip devices wi'

. > te d.

Eight hours is ample time to test the remaining trip devices and in many cases make on-line repairs.

j The OPERABillTY of the SFAS instrumentation systems and bypasses ensure that 1) the associated SFAS action will be initiated when the parameter monitored by each 72 channel or combination thereof reaches its setpoint, 2) the specified coincidence 3-26 Proposed Amendment No. 72 l

RANCl10 SECO UNIT 1 1

TECliNICAL SPECIFICATIONS Limiting Conditions for Operation Bases (Continued) logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for SFAS purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protec-tion and mitigation of accident and transient conditions.

The integrated opera-tion of each of these systems is consistent with the assumptions used in the accident analyses.

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these 4

variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-l Cooled fluclear Power Plants to Assess Plant Conditions During and Following an i

Accident", December 1975 and NUREG-0578, "Tril-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

REFERENCE FSAR, subsection 7.1 i

)

r 3-26a Proposed Admendment No. 72

.= -

TABLE 3 5.1-1 (Continued)

INSTRUMENTS OPERATING CONDITIONS (C)

(A)

(B)

Operator Action if Minimum Operable Minimum Degree Conditions of Colums A Functional Unit Channels of Redundancy and B cannot Be Met Safety Features

b.

Reactor Building Pressure Bring to hot shutdown within instrument channels 2

1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

Manual pushbutton 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

r, d.

Automatic Actuation Logic 2

1

' Bring to hot shutdown within 93 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 725 3; 99

2.,Lcwe pressure injection c-c un y,

v, 8

, es

[$

a.

Reactor coolant pressure Bring to hot shutdown within El o F1 ri instrument channels 2

1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Z2 g b.

Reactor Building pressure Bring to hot shutdown within R

-4 n _.

instrument channels 2

1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months E; g --

a =

c.

Manual pushbutton 2

1 Bring to hot shutdown-within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Jr

p. Reactor Building spray pump i'

.=

'c a.

Reactor Building pressure instrument channel 2

i Bring to hot shutdown within

[-

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

b.

Manual pushbutton 2

i Bring to hot shutdown within p

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

o":

(Con ti nued) 2.

o di f minimum condi tions are not met within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months af ter hot shutdown, the unit chall be placed in a cold shutdown condition within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Proposed Amendrent No. 72

TABLE 3 5.1-1 (Continued) 1NSTRUMENTS OPERATING CONDITIONS (C)

(A)

(B)

Operator Action i.

Minimum Operable Minimum Degree Condi ti'ons of Columns Functional Unit Channels of Rodundancy A and B Cannot Be Met 4

Safety Features *

4. Reactor Building spray valve a.

Reactor Building pressure Bring to shutdown within instrumen t channel 2

1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months b.

Manual pushbutton 2

1 Bring to hot shutdown within g

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months e,

Total Number Minimum Channels

[j a Process Instrumentation cf Channels Operabic gy l

r o 3

1. Pressurizer Water Level 3

1 See Section 3.5.1.2 o vs

, m 72 "

2. Auxi l ia ry Feedwa ter Flow **

I per loop 1 per loop u

=

$3. Reactor Coolant System Subcooling SE l

Margin Monitor 2

1 r-6 3' E

14. EPOV Posi tion Indicator (Primary

?!

~i Detec tor) power ind ica tor ***

1/ valve 1/ valve

5. EMOV Posi tion Indicator (Backup 2

Detector) acoustic or T/C***

1/ valve 0

L b

!6. EMOV Block Valve Position Indicator 1/ valve 1/ valve t

5 1

17. Safety Valve Position Indicator

{

(Primary Detector) T/C 1/ valve 1/ valve i.

o

8. Safety Valve Position Indicator 3

(Backup Detector) acoustic 1/ valve 0

2 1

o I

li

f ninicua condi tions are not met within.43 hours4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months after hot shutdown, the uni t shall be placed in a cold shutda.ca condition wi thin an addi tional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

- 0TSG 1evel may be used for flow.

- - Applies 72 when EMOV is OPERABLE.

Pro::osed Amendment No. 72

TABLE 3.5.1-1 (Continued)

INSTRUMENTS OPERATING CONDITIONS 4

(C)

(A)

(B)

Operator Action if Total Number of Minimum Channels Conditions of Colums A Functional Uni t Channels Operable and B Cannot Se Met Auxiliary Feedwater 1.

Low Main Feedwater Pressure See Section 3.5.1.2 Start Motor Driven Pump and 72 Turbine Driven Pump 2.

1 2.

Contact Monitor - P.CP Pump h

Start Motor Driven and Turbine gg Driven Pumps 2

1

-=

n n

> =.

t-

=

m v.

\\Ma i2o c

E n

A

c. s -

a =

-- O

-G n

9 E-

/,

)

o G

o'

.?

I

,c.

i

,'d'-

O I

Proposed Amendment No. 72

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.3 SAFETY FEATURES ACTUATION SYSTEM SETPOINTS Applicability This specification applies to the safety features actuation systen actuation setpoints.

Objective To provide for automatic initiation of the safety features actuation system in the event of a breach of reactor coolant system integrity.

Specification The safety features actuation setpoints and permissible bypasses shall be as follows:

Functional Unit Action Setpoint Reactor Building spray valves 430 psig High Reactor Building pressurc Reactor Building spray pumps 530 psic e

High pressure injection and start of Reactor Building cooling and Reactor Building isolation.

1 4 psig i

Low pressure injection 1 4 psig Low reactor coolant system liigh pressure injection and start

~

pressure **

of Reactor Building cooling and

~

Reactor Building isolation.

1 1600 psig Low pressure injection 1 1600 psig Automatic Actuation Logic All above Not Applicable Manual All above Not Applicable 72 Loss of all RC Pumps Starts Auxiliary Feedwater Pumps Not Applicable Low Feedwater Pressure Starts Auxiliary Feedwater Pumps 1 750 psig AMay be bypassed during Reactor Building leak rate test.

- May be bypassed below 1850 psig and is automatically reinstated above 1850 psig.

- - Five-minute time delay.

3-34 Proposed Amendment No. 72

rat C110 SECO UNIT 1 TECli.SICAL SPECIFICATIONS Limiting Conditions for Operations SAFETY FEATURES CONTAIN'!ENT ISOLATION VALVES VALVE N!'MBER DESCRIPTION MAXIMUM CLOSURE TIME (SEC)

SFV 53612 RB Atm. & Purge Sample, AB Side.

.3 SFV 53613 RB Atm. & Rad Sample, AB Side.

3 SFV 60003 RC Sys. Drain Isol, AB Side.

14 SFV 66308 RB Normal Sump Drain, AB Side.

15 SFV 92520 Przr. Nitrogen Isol., AB Side.

5 SFV 53503 RB Purge Inlet, AB Side.

5 SFV 53604 RB Purge Outlet, AB Side.

3 SFV 53610 RB Press. Equalizer, AB Side.

15 SFV 60002 RC System Vent Isol., AB Side.

6 SFV 60004 RC System Drain Isol., AB Side.

14 SFV 66309 RB Normal Sump Drain, AB Side.

8 SFV 70002 Przr. Liquid Sample Isol., AB Side.

8 SFV 72502 Przr. Gas Sample Isol., AB Side.

5 llV 20611 OTSG's Blowdown Isol., AB Side.

18 IIV 20593 OTSG-A Sampic Isol., AB Side.

12 ItA 20594 OTSC-B Sample Isol., AB Side 5

7 SFV 35304 RB Purge Inlet, RB Side.

8 SFV 53603 RB Press. Equalizer, RB Side.

9 SFV 53605 RB Purge Outlet, RB Side.

8 SFV 6000.

RC Sys. Vent 1 sol, RB Side.

12 SFV 70001 Przr. Liquid Sample Isol., RB Side.

18 SFV 70003 Przr. Vapor Sample Isol., RB Side.

21 SFV 72501 Przr. Gas Sample Isol., RB Side.

9 SFV 46014 RB CCW Supply, AB Side.

15 SFV 46203 RB CCW Return, RB Side.

14 SFV 46204 RB CCU Return, AB Side.

18 SFV 46906 CRD Cooling Water Supply, AB Side.

9 SFV 46907 CRD Cooling Water Return, RD Side.

14 SFV 46908 CRD Cooling Water Return, AB Side.

8 IIV 20609 OTSC-A Blowdown Isol., RB Side.

15 llV 20610 OTSC-B Blowdown Isol., RB Side.

13 3-35 Proposed Amendment No. 72

S RANClio SECO UNIT 1 TECllNICAL SPECIr1 CATIONS Limiting Conditions for Operations Bases liigh Reactor Building Pressure The basis for the 30 poig and 4 psig setpoints for the high pressure signal is to establish a setting which would be reached in adequate time in the event of a DBA, cover a spectrum of break sizes and yet be far enough above normal opera-tion maximum internal pressure to prevent spurious initiation.

Low Reactor Coolant System Pressure l

l The basis for the 1600 psig low reactor coolant pressure setpoint for high and low pressure injection initiation is to establish a value which is high enough such that protection is provided for the entire spectrum of break sizes and is l

far enough below normal operating pressure to prevent spurious initiation. (1)

J l

CONTAI'O!ENT ISOLATION VALVES k

The OPERABILITY of the containment isolation valves ensures that the containment l

atmosphere will be isolated from the outside enviroment in the event of a release of radioactive material to the containment atmosphere by pressurization of the 72' contalment. Conta!.nment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for LOCA.

I REFERENCES 1

(1)

FSAR, paragraph 14.2.2.5 i

1 I

3-35a e

Proposed Amendment No. 72

=

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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Condi tions for Operation B.

Both startup transformers shall be in service except that one will be sufficient if during the time one startup transformer is inoperable, a diesel generator is started and run continuously.

C.

Both diesel generators shall be operable except that from and after the date that one of the dicsci generators is made or found to be inoperable for any reason, reactor operation is permissible for the succeeding 15 days provided that during such 15 days the operable dicsci generator shall be load tested daily and both startup trans-formers are available.

If the diesel is not returned to service at the end of 15 days, the other diesel will be started and run with at least minimun load continuously for an additional 15 days.

If at the end of the second 15 days the diesel is not returned to service, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

D.

If the plant is separated from the system while carrying its own auxiliaries, or if all 220 kV lines are lost, continued reactor operation is permissible provided that one emergency diesel generator is started and run continuously until a transmission line is restored.

E.

The essential nuclear service electrical buses, switchgear, load shedding, and automatic diesel start systems shall be operable except as provided in C above and as required for surveillance testing.

F.

Nuclear service batteries are charged and in service except that one nuclear service battery may be removed from service for not more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G.

Both nuclear services busses are operable except that one nuclear service bus may be removed f rom service for not more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided that all equipment on the other nuclear service bus is operable.

3.7.3 If both dies,el generators become inoperable, the unit shall be placed in the cold shutdown condition.

3.7.4 The pressurizer shall be OPERASLE with at leest (126) kw of pressurizer heaters. With the pressurizer inoperable due to inoperable emergency power supplies to the pressurizer heater either restore the inoperable 72 emergency power supply wi thin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAUDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3-42 Proposed Amendment No. 72 l

l i

I P,ANCil0 SECO UtilT I J

TEtllNICAL SPECIFICATI0tJS Limiting Conditions for Operation Bases The auxiliary electrical power systems are arranged so that no single failure can inactivate enough safety features equipment to jeopardize plant safety.

The normal souce of power to the redundant nucicar service loads is the two Startup transformers connected to the 220-kV station switchyard.

All of the normal power supplied to plant auxiliary loads is provided through the two uni t auxiliary transformers connected to the generator bus.

Emergency power for the nuclear service loads is obtained f rom two on-site diesel generators.

The startup transformers are sized to carry full plant auxiliary loads.

When p la n t auxiliary power is not available f rom the uni t auxiliary trans-former,it will be obtained from the startup transformers.

4 I

4 i

l I

3-42a Proposed Amendment No. 72

)

I

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The five 220-kV transmission lines are not under the direct control of the Rancho Seco station. Therefore, all five cannot be assumed to be available at all times. However, extensive reliabili ty and protective features are utilized so that the probability of losing more than one source of 220-kV power from faults is low.

By requiring that two 220-kV lines are in service prior loss of the to startup, one circuit will be immediately available following a onsite alternating current diesel power supplies and the other offsite 220-kV If there is a loss of all 220-kV remote connections, power to the line.

safety features will be supplied by the diesel generators.

The 35,000 gallons of fuel stored in each storage tank permit operation of the two diesel generators for seven days.

It is considered unlikely not to be able to secure fuel oil from an outside source during this time under the worst of weather conditions.

loss of one The four 120-volt d-c control panelboards are arranged so that bus will not preclude safe shutdcan or operation of safety features systems.

During periods when one plant battery is de-energized for test or maintenance, the associated 125-volt d-c bus can be supplied f rom its battery charger.

Each redundant pair ("A" and "C", "B" and "D") of safety features actuation and reactor protection 125-volt d-c buses has a standby battery charger in addi tion to the two bus battery chargers.

Loss of power f rom one ba ttery charger per pair of redundant d-c buses has no signi ficant consequence since a standby battery charger is available.

In addition, each 125-volt d-c bus can continue to receive power from its respective battery without i n te r rup t i on.

Suf ficient redundancy is available wi th any three of the four 120-vol t a-c vi tal power buses in service such that reactor safety is assured.

Every reasonable ef fort will be made to maintain all safety instrumentation in operation.

During periods of station operation under the condition of electrical system degradation, as described above in Specification 3.7.2, the operating action required is to start and-run sufficient standby power supplies so as not to comprom?se the safety of the plant.

As seen in Specification 3.7.2, a time is placed on operation during certain degraded conditions based on the limit reliability of the available power supply.

The requi rement that (126) kw of pressurizer heaters and their associated controls supplied with electrical power from an emergency bus 72 being capable of being provides assurance that these heaters can be energized during a loss of of fsite power condition to maintain natural circulation at HOT SHUTDOWN.

REFERENCE FSAR, section 8 I

3-43 Proposed Amendment No. 72 l

l

I TABLE 4.1-1 (Continued)

INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks 20.

High pressure injection, MA M

NA l72 Reactor Building isolation, ar.d Reactor Building emer-gency cooling Channel A manual trip.

21.

High pressure injection NA M

NA l72 Reactor Building isola-tion, and Reactor Build-ing emergency cooling M

Channel B manual trip.

ngg

-- =

22.

Low pressure injection NA R

NA 5; @

Channel A manual trip r o 7

m 23 Low pressure injection MA R

NA

?; 8 Cahnnel B manual trip

e

__ =

n -

24.

Reactor Building spray NA R

NA R*

pump Channel A manual 5 --

trip 5

25 Reactor Building spray NA R

NA g,

pump Channel B manual 5

trip 5

26.

Reactor Building spray NA R

NA h'

valves Channel A manual 5

trip o

oo E~

1 w

Proposed Amendment No. 72 l

1

TABLE 4.1-1 (Continued) lNSTRUMENT SURVEILLANCE REQUIREMENTS' Channel Description Check Test Calibrate Remarks 42.

Reactor Building drain accumulation tank icvel NA NA R

43 Incore neutron detectors M(1)

NA NA (1) Check functioning, including functioning of computer read out amd/or recorder readout.

44.

Process and area radiation troni toring systems V

M q

45.

Energency plant radiation instruments M(1)

NA R

(1) Battery check 46.

Environmental air monitors M(1)

NA R

(1) Check functioning 47.

Strong notion accelerometer q(1)

NA R

(1) Battery check

48. Auxiliary Feedwater p

Start Circuit E R

a. Phase imbalance /Under-yg power RCF 5

M R

>=

b. Low Mair, Feedwater

%{

p Pressure NA NA R

U 49 Pressurizer Wa ter Level

M R

50.

Auxiliary Feedwater Flow 9Q 72 y -

Rate M

NA R

51.

Reactor Coolant Sys tem sub-5

~

cooling Marnin MonI tor M

NA R

52.

EMOV Power Position m

E (e,nfra ryien tgetec tor) 7 r

r M

NA R

g

53. DIOV Posi tion Indicator JD cLun De tectp-M NA R

E T/C or Acousti)c R

9s.

EMOV Block valve Position o

Indicator M

NA R

g 55.

Safety Valve Positi.on In-NA

(

dicator(Primary Detector)f/C M

R-I

56. Safety Valve Position in-dicator (Backup Detector)

M NA R

.A.c.o.u. s. t ic S = Each shift M = Monthly P = Prior to each startup if not done previous week D = Daily Q = Quarterly R = Once during the refueling interval.

Vital Power Supply Requirerents p r-,,a.

e aw - s:n 7?

RANCil0 SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Specification (Continued)

B.

2.

The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal and all appropriate pump breakers shall have opened or closed, and all valves have completed their travel.

3 Decay heat pump casings shall be vented monthly and prior to any ECCS flew tests.

C.

Core Flooding System 1.

During each refueling interval, a core flooding system test shall be conducted to demonst rate proper operat ion of the system.

During depressurization of the reacter coolant system, verification shall be made that the check valves in the core flooding tank discharge lines operate.

2.

The test will be considered satisfactory if control board indication of core flood tank level verifies that all check valves have opened.

4.5 1.2 Components Tests A.

Pumps At Jeast quarterly, the high pressure, makeup and decay beat removal pumps shall be started and operated to verify operation.

i B.

Valves -- Power Operated 1.

At least quarterly each safety features valve in the emergency core cooling systems and each safety features valve associated with emergency core coolina in the decay heat removal syster.

1 shall be tested to verify operability.

2.

The EMOV shall be demonstrated OPERABLE each 31 days and the block valve demonstrated OPERABLE each 92 days.

3 The EMOV shall be CHANNEL CAllBRATED at icast once cach 18 months.

C.

Nuclear Service Cooling and Raw Water System At least quarterly, the nuclear service cooling and raw water system shall be operatea to verify performance.

4-27 Proposed Amendment No. 72 l

RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Specification (Continued)

B.

3 Acceptable performance for the nucicar service cooling water and raw water pumps shall be that the pumps start and operate for 15 minutes at the design flow rate with the required differential pressure.

4.

The acceptable performance of each power-operated valve in the emergency core cooling system will be that notion is indicated upon actuation by' appropriate signals.

~

The PORV shall be demonstrated OPERABLE:

a.

At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and b.

At least once per 18 months by performance of a CHANNEL CAllBRATION.

72 The block valve shall be demonstrated OPERA 5LE at least once per 92 days by opera ting the valve through one complete cycle of full travel.

Bases The emergency core cooling systems are the principal reactor safeguards in the event of a loss-of-coolant accident. The removal-of heat f rom the core provided by these systems is designed to limit core damage.

The decay heat removal pumps are testej singularly for operability by opening the borated water storage tank outlet valves and the test line valves to the borated water s torage tank. This allows water to be pumped from the borated water storage line.

tank through each of the injection lines and back to the tank through a test With the reactor shut down, the check valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the core flood tanks verifies the check valves have opened.

REFERENCES FSAR subsection 6.2.

.4-28 Proposed Amendment No. 72

~

RANCHO SECO UUlT 1 TECHNICAL SPECIFICATIONS Surveillance Standards Specification (Continued)

D. Acceptance 1.

Acceptable performance for the high pressure injection pumps shall be that the pump starts and operates for 15 minutes discharging through the miniflow and the discharge pressure indicates flow is within 10 percent of the initial level of performance and at least equal to the minimum design flow rate.

2.

Acceptable performance for the decay heat pumps shall be that the pump starts and operates for 15 minutes discharg'.3 through the test flow path and the dischargc pressure and flow are within 10 percent of the initial level of performanc2 and at least equal to the minimum design flow rate.

3 Acceptable performance for the nuclear service cooling water and raw water pumps shall be that the pumps start and operate for 15 minutes at the design flow rate with the required differential pressure.

4.

The acceptable performance of each power-operated valve in the emergency core cooling system will be that motion is indicated upon actuation by appropriate signals.

I 5

The acceptable performance of the EMOV shall include a s

CHANNEL FUNCTIONAL TEST, excluding valve operation.

6.

The acceptable performance of the block valve shall be by 72 operating the valve through one complete cycle of full travel.

Bases The emergency core cooling systems are the principal reactor safeguards in the event of a loss-of-coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The decay heat removal pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the test line valves *s the borated water storage tank.

This allows water to be pumped f rom the borat d water storage tank through each of the injection lines and back to the tank thrcJgh a test line.

With the reactor shut down, the check valves in each core flooding line are checked for operability by reducir,g the reactor coolant system pressure until the indicated level in the core flood tanks verifies the check valves have opened.

REFERENCES FSAR subsection 6.2 4-28 Proposed Amendment No. 72 t

RANCHO SE:3 UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.6.5 Diesel generator fuel oil supply shall be tested as follows:

A.

During the monthly diesel generator test, the diesel fuel oil transfer pumps shall be monitored for operation.

B.

Once a month, quantity of the diesel fuel oil shall be logged and checked against minimum speci fica tions.

The tests specified will be considered satisfactory if control room indicat ion and/or visual examinction demonstrates that all components have operated properly.

4.6.6 The pressurizer shall be tested as follows:

A.

The pressurizer water level shall be determined to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

72 B.

The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the heaters.

Bases The tests specified are designed to demonstrate that the diesel generators will provide power for operation of safety features equipment.

They also assure that the emergency generator control system and the control systems for the safety features equipment will f.nction automatically in the event of a loss of all normal a-c station service power, or upon receipt of a safety features actuation signal.

The testing frequency specified is intended to identify and ~ permit correction of any mechanical or cicctrical deficiency before it can result in a system failure.

The fuel oil supply, starting cir-cuits and controls are continuously monitored and any faults are alarmed and l

indicated. An abnormal condition in these systems would be signaled without having to place the diesel generators on test.

Precipitous failure of the plant battery is extremely unlikely.

The surveillance speci fied is that which has been demons trated over the years to provide an indi-cation of a cell becoming unserviceable long-before it fails.

REFERENCE (1)

IEEE 308 4-35 Proposed Amendnen t No. 72 3

RANCil0 SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feedwater pumps.

Objective To verify that the auxiliary fe2dwater pump and associated valves are operable.

Specification 4.8.1 At least every 92 days on a staggered test basis at a time when the average reactor coolant system temperature is 1 305 F, the turbine / motor driven and motor driven auxiliary feedwater pumps shall be operated on recircula-tion to the condenser to veri fy proper operation.

The 92-day test frequency requirement shall be brought current within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the average reactor coolant system temperature is > 305 F.

Acceptable performance will be indicated i f the pump starts and operates for fifteen minutes at the design flow of 780 gpm.

This flow will be verified using tank level decrease and pump di f ferential pressure.

4.8.2 At least once per 18 months during a shutdown:

1.

Verify that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation 72 tes t s i gna l.

i d.

2.

Verify that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal.

Bases The quarterly test frequency will be sufficient to verify that the turbine / motor driven and motor driven auxiliary feedwater pumps are operalle.

Verification of correct operation will be made both from the control roon insb umentation and direct visual observation of the pumps.

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant Systeia can be cooled down to less t:ian 305 F f rom normal operating conditions in the event of a total loss of of f-si te poacr.

Each electric driven auxiliary feedwater pump is capable of delivering a total 72 feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam gene ra tors. The steam driven auxiliary feedwater pump

'.s capable of de-livering a total feedwater flow of 780 gpm at a pressure of 1050 psig to the entrance of the steam gene.ators.

This capacity is sufficient to ensure tha t ade-quate teedwater flow is available to renove decay heat and reduce the P.cactor Coolant System temperature to less than 30fF when the Decay Heat Removal System r.ny be placed into ope;ation.

REFERENCE 4-39 Proposed Amendment No. 72

RANCil0 SECO UNIT 1 TECliNICAL SPECIFICAT10tlS TABLE 6.2-1 SillFT CREW PERSONNEL AtlD LICENSE REQUIREMENTS REACTOR liODE RANCHO SEC0 JOB TITLE COLD OTilER TilAN SHUTDOWN COLD SHUTDOWN Shi 4 Spervisor 1 - SL 1 - SL Sr. Control Room Operator or Con trol Room Operator 1-L 2 - L*

Auxiliary Operator or Equipment Attendant i

I Equipment Attendant or Power Plant Helper 1

Shift Technical Advisor 0

l Minimum Total Personnel 3

6 72

0nc licensed operator when the reactor is shut down greater than 1% Ak/k.

- In the event that any member of a minimum shif t crew is absent or incapacitated due to illness or injury, a qualified replacement shall be designated to report onsite within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SL -NRC Senior Licensed Operator

~-NRC Licensed Operator 72 6-2 Proposed Amendment No. 72

'AD?9tNISTRATIVE CONTROLS o

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the unit staf f shall meet or exceed the minimum quali fica-tions of ANSI N18.1-1971 for comparable posi tio.ns, except for (1) the (Chemical-Radiation Supervisor)who shall meet or exceed the qualifica-tions of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical 72 Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline.

The STA shall receive specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and rep'lacement training program for the operating staff shall be maintained under the direction of the Training Supervisor and shall meet.or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Apnendix "A" of 10 CFR Part 55 6.4.2 A training program for the Fire Brigade shall be maintained under the di rection of the Safety Technician and shall meet or exceed the requirements of Section 27 of the NFPA code - 1975, except refresher classroom training shall be on a quarterly schedule.

6.5 REVIEW AND AUDIT 6.5.1 PLANT REV!EU COMMITTEE (PRC)

FUNCTION 6.5.1.1 The Plant Review Comnittee shall function to advise the Manager of Nuclear Operations and the Plant Superintendent on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Plant Review Committee shall be composed of the:

Chairman:

Technical Assistant Member:

Supervisor Nuclear Operations Member:

Engineering & Quality Control Supervisor 72 Member:

Supervisor Nuclear Maintenance Member:

Chemical and Radiation Supervisor Other rembers as the Manager of Nuclear Operations may appoint from time to time.

l 6-3 Proposed Amendment No. 72 l

ML19347C388

Text

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