ML19322B309

Amend 17 to License DPR-55 Incorporating Tech Spec Changes. Proposed Tech Specs Encl

ML19322B309
Person / Time
Site:	Oconee
Issue date:	03/25/1976
From:	Purple R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19322B304	List: ML19322B303 ML19322B305 ML19322B307 ML19322B309 ML19322B316
References
NUDOCS 7912020157
Download: ML19322B309 (34)

Text

1.

UNITE 3 STATES g'

t NUCLEAR REEULATCRY COMMISSION j

WASHINGTON, D. C. 20506

%*****/

G DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.17 License No. DPR-55 1.

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

A.

The application for amendment by Duke Power Company (the licensee) dated December 1, 1975, as supplemented Febntary 24, and 27, 1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of th6 Commission; C.

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

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

An environmental statement or negative declaration need not be prepared in connection with the issuance of this amendment.

i 7912026/

i i

2.

Accordingly, the license is amended by a change to the Techni %1 Specifications as indicated in the attachment to this license amendment.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert A. Purple, Chi f Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 25, 1976 I

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s ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 20 TO FACILITY LICENSE NO. DPR-38 AMENDMENT NO. 20 TO FACILITY LICENSE NO. DPR-47 AMENDMENT NO.17 TO FACILITY LJCENSE NO. DPR-55 DOCKET N05. 50-269, 50-270 AND 50-287 Revisc Appendix A as follows:

Remove Pages Insert Pages 2.1 2.1-1 2.1-2 2.1-2 2.1-3 2.1-3 2.1-4 2.1-4 2.1-7 2.1-7 2.1-10 2.1-10 2.3-1 2.3-1 2.3-2 2.3-2 2.3-3 2.3-3 2.3-4 2.3-4 2.3-5 2.3-5 2.3-8 2.3-8 2.3-11 2.3-11 'A #

3.1-1

'3.1-1

~

3.1-2 3.1-2 3.1-19 3.1-19 3.1-20 3.1-20 3.3-2 3.3-2 3.5-7 3.5-7 3.5-8 3.5-8 3.5-9 3.5-9 3.5-10 3.5-10 3.5-11 3.5-11 3.5-12 3.5-12 3.5-13 3.5-13 3.5-18 3.5-18 i

3.5-18a 3.5-21 3.5-21 3.5-21a 3.5-24 3.5-24 4.1-9 4.1-9 I

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2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETT LIMITS, REACTOR CORE

?

Applicability _

i Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and' coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification The com'bination of the reactor system pressure and coolant temperature shall-not exceed the safety limit as defined by the locus of points established in Figure 2.1-1A-Unit 1.

If the actual pressure / temperature point is below 2.1-1B-Unit 2 2.1-1C-Unit 3 and to the right of the line, the safety limit is exceeded.

The combination of reactor thernal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2A-Unit 1.

If the actual reactor-thermal-power / power 2.1-2B-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases - Unic 1 The safety limits presented for Oconee Unit 1 have been generated usinr,BA'W-2 critical heat flux (CHF) correlatienll)and the actual =casured flow raue at Oconce Unit 1 (2). This development is discussed 'n the Oconee 1. Cycle 3 Reload Report, reference (2). The flow rate utilize. is 107.6 percent of the 6 lbs/hr) based on four-pump operation. (2) design flow (131.32 x 10 To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under nor=al operating conditions. This is accomplished'by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coef ficient is large enovah so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure. Although DNB is nog an observable parameter during reactor operation, the observable parameters of noutron power, reactor coolant flow, temperature, and pressure 9

.a s

Amendmen No. 20, 20 & 17 2.1-1 March 25 1976

?*

f

t t

2 corrolatica (1). Tha BAW-2 can bo related to DNB thrcugh th2 uss of the BAW-d the lecction of DNB fo correlation has been developed to predict DNB an The loen1 DNB ibutions.

axially uniform and non-uniform heat flux distr flux that would cause DNB at a ratio (DNBR), defined as the ratio of the heatflux, is indicative of the margin ion, normal particular core location to the actual heatThe minimum value o A

i operational transients, and anticipated trans ents 95 percent confidence i

to DNB.

DNBR of 1.30 corresponds to a 95 percent probab tive margin to t m pressure has been DNB for all operating conditions. outlet pressure and the indicate The difference f ty limits.

considered in determining the core protection sa ein t l

30 psi drop was assumed in reducing the pressure trip setpontslocation ditions at which a l

The curve presented in Figure 2.1-1A represents the conpossible therm i

minimum DNBR of 1.30 is predicted for the ing (minimum reactor This curve is based on lbs/hr.).

flow is 107.6 percent of 131.3 x 10 the combination of nuclear power peaking nservative DNBR than any I

coolant densification and rod bowing, which result in a more co ion.

other shape that exists during normal operat i tive of two thermal and rod bowing: l The curves of Figure 2.1-2A are based on the more restr c ification limits and include the effects of potential fuel dens f the radial p I

The 1.30 DNBR limit produced by the combination o less than a 1.30 DN3R.

peak and position of the axial peak that yields no 1.

tral fuel melting The combination of radial and axial peak that causes cen

~

The limit is 20.15 ku/ft for Unit 1.

2 at the hot spot,.

d therefore limits have Power peaking is,not a directly observable quantity animbalance produ j

been established on the bases of the reactor power f

power peaking.

' 4 of Figure 2.1-2A correspond The specified flow rates for Curves 1, 2, 3 andh four pumps, three pum to the expected minimum flow rates witeach loop and tw i l f all possible reactor The curve of Figure 2.1-1A is the cost restrictive in Figure 2.1-3A.

h is 85.3 percent due to a The maximum thermal power for three-pump operation 74.7 percent f i

power level trip produced by the flux-flow rat o The d instrument error.

78.8 percent power plus the maximum calibration anditions are produc maximum thermal power for other coolant pump con similar manner.

um-i~.

2.1-2 Amendment No.

20, 20 6 17

• k March 25,1976

-~

b

n 4

a For Figure 2.1-3A, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30. The 1.30DNBR curve for four-l pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the four pump curve will be above and to the left of the oth'er curves.

References (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.

(2) Oconee 1, Cycle 3 - Reload Report - BAW-1427, December '1975.

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2.1-3 Amendment No. 20, 20 6 17 March 25, 197,6

.-g

t i

2400 2300 ACCEPTABLE OPERATION UNACCEPTABLE i

OPERATION

~

=

0 2100 S

a.

E 3

2000 a

1900 1800 I

I I

I 560 580 600 620 640 660 Reactor Outlet Temperaturs.*F

.1 LIIIT 1 CORE PETECTIm SAFEIY LIMITS OCONEE NUCLEAR STATION FIGmE 2.1.!A 2*I~'

Amend ent No. 20, 20 6 17 m

March 25, 1976 1..

t.

Inercal Po:er Level. 5

~

UNACCEPTABLE OPERATI0tt 120 (l12)

(+2s.I12)

(-29.I12) 110 ACCEPTABLE 4 PUNP 100 OPERATION 1

(+40. 90 )

90

( -37. 9 0 )

(es.3)

(+2s.ss.3)

( -29. s s. 3 )

- 80 ACCEPTABLE 314 70 PUMP

(*"o 63 3) 2 OPERATION

( -37. 6 3. 3 )

. 60 (ss.21'

(+2s.sa.2)

( 29.sa.2)

/

ACCEPTABLE

. 50 2.3 & 4 PUMP OPERATION 3&4

( -37,3 6. 2 )

30 20 10 I

I-f 60

-40

-20 0

+20

+40

+60 I

I Reactor Power Imbalance 5 CURVE REACTOR COOLANT FLOW (ID/hr) 6 1

141.3 x 10 6

2 105.6 x 10 6

3 69.3 x 10 6

4 64.7 x 10

'THE FLUX / FLOW SETPOINT FOR 2/0 PUNP OPERATION NUST BE SET AT 0.949 WIT' l CORE PROTECT 10ii SAFETY f

OCONEE NUCt. EAR STATION g

FicmE 2.1-2A

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Amendment No. 20, 20 4 17 2.1-7 March 25, 1976

n 2400

/

i 2300 ACCEPTABLE OPERATION 2200 E

N h

2l00 t

O E

2000 E

1900 1800 560 580 600 620 640 660 Reactor Outlet Ta.Tmerature,*F CURVE REACTOR COOLANT Fl.01 (lo.nr) 6 1

141.3 x 10 ONBR Limit 6

2 105.6 x 10 DNBR 1.imit 6

3 69.3 x 10 Quality Limit (o(& UillT 1 ninete OCONEE NUCLEAR STATION FIGURE 2.1-3A Amendment No. 20. 20.5 17 2.1-10 l

March 25, 1976

(

2.3 LlMlTING SAFETY SYSTDi SETTINGS, PROTECTIVE INSTRLMATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protective action to prevent any combination of process variables from exceeding a safety limit.

Specification The reactor protective syst'em trip setting limits and the permissible bypasses for the instrument channels shall be as_ stated in Table 2.3-1A - Unit 1 and 2.3-1B - Unit 2 2.3-lC - Unit 3 Tigure 2.3-2A - Unit 1 2.3 Unit 2 2.3-2C - Unit 3 The pump monitors shall produce a reactor trip for the following conditions:

Loss of two pumps and reactor power level is greater than 55% of rated a.

power.

b.

Loss of two pumps in one reactor coolant loop and reactor power level is greater than 0.0?. of rated power. (Power /RC pump trip setpoint is reset to SS: for all modes of 2 pump operation.)

l Loss of one or two pumps during two-pump operatien.

c.

Bases The reactor protective system consists of four instrument channels to monitor each of several selected plant conditions which vill cause a reactor trip if any one of these conditions deviates frem a pre-selected operating range to the degree that a safety limit c:ry be reached.

The trip setting limits for protective system instrumentation are listed in Table 2.3-1A - Unit 1.

The safety analysis has been based upon these protective 2.3-1B - Unit 2 2.3-lC - Unit 3 system instrumentation trip set points plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

Amendment No. 20, 20 $ 17 2.3-1 March 25, 1976 c

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x

During corno* 7,1cnt op2rctj vith cl1 rece:or coolant pu:py 7perating, level rasc.

105.f of rce.ctor ::1p is initiated *(.

a :n3 reactor pewa:A4 ding to thin tha possiblo varintion in trip setpoints dus rcted p; war.

to calibra: ion and instru=ent er:crs, the maximum actual pcuer at which a trip would be actuated could be 1123, which is = ore conservative than :he value used in :he safety analysis. (4)

Overoover Trio Based on Flew and Imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to acco==odate the most severe thermal transient considered in the desi;n, th'e loss-of-coolant flow accident from high power.

Analysis has de=onstrated that the specified power-to-flow ratio is adequate to prevent a DN3R of less than 1.3 should a low flow c:ndition exist due to any electrical =alfunction.

The power level trip set point produced by the power-to-flow ratio provides both high power level and lov flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.

The power level trip set point produced by the power-to-flow ratio provides overpower D'~d pro-tection for all codes of pump operation.

For every flow rate there is a =axi-mum per=issible pcVer level, and for every power level there is a =ini um persissible low flow : ate.

Typical power le rel and low flow rate combina:1ons for the pu:p situcations of Table 2.3-1A are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if pouc is 105.5% and reactor flov ra:e is 100"., or flow rate is 94.S~ and power level is 100..

Trip vould occur when three reactor coolant pu=ps are operating if power 2.

is 78.S". and reactor flow rate is 74.7*. c: flow ra:e is 71.1" and power level is 75".

Trip would occur when two reactor coolant pu=ps a e operating in a singic 3.

loop if power is,51.7" and the operating loop flo:, rate is 54.5" or flev rate is 48.54 and power level is 46%.

4.

Trip would occ'u when one reactor coolant pu=p is operating in each loop (total of two pumps opera:ing) if the power is 51.7*. and reactor flov rate is 49.0". or flow rate is 46.4% and the power level is 49%.

The flux-co-flow ra:ios for Uni: 1 account for the =ax1=us varia: ion from the average value of the RC flow signal in such a =anner that the reactor pro:ective systes receives a conserva:1ve indication of :he RC flow.'

For safety calculations the maximum calibration and instru=entation-errors for the pcuer level trip were used.

reae:or The power-i= balance boundaries are es:ablished in order to preven:

thermal limits from being exceeded. These ther:al li=its are ei:her power lini:s or DN3R li=its.

The tactor power i= balance (;cuer in pt king kv/f:

the top half of core sinus power in :he bottos half of core) reduces the power the boundaries of level trip produced by the power-to-flov ratio such that l

Figure 2.3-2A - Unit 1 are produced. The power-to-flow ratio reduces the power [

l 2.3 Unit 2 p

2.3-2C - Unit 3 2.3-2 Y

mendment No. 20, 20 6 17 March 25,1976 1

t

'leval trip cnd ass cieted racetor pcvar/rc ctar p'swar-inbnlenca b:undcriss by 1. 0 5 5 %- Unit 1 for a 1% flow teduction.

l 1.07% - Unit 2 1.07% - Unit 3 For Unit 1, the power-to-flow reduction ratio is 0.949, and for Units 2 and 3, the power-to-flow reduction factor is 0.961 during single loop operation.

Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the locs of reactor coolant pump (s).

The circuitry monitoring pu=p operational status provides redur. dant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio. The pu=p monitors also restrict the power level for the number of pumps in operation.

Reactor Coolant System Pressure During a startup accident frco low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear over-power trip set point. The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)

The low pressure (1800) psig and variable low pressure (11.14 T

-4706) trip l

(1800) psig (16.25 T "*-7756)

(1800) psig (16.25 T "*-7756) setpoints shown in Figure 2.3-1A have been established to mainta2n the DNB 2.3-1B 2.3-1C ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2,3)

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T

-4746) l (16.25 T "" -7796)

(16.25 T "" -7796) out Coolant Outlet Temeerature The high reactor coolant outlet temperature trip setting limit (619 F).shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C temperatures in the operating range.

Due to calibration the safety analysis used a trip set point of 620,and instrumentation F.

errors, Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a loss-of-coolant accident, even in the absence of a low reactor ecolant system pressurc trip.

Amendment No.20,^20 6 17

~

i March 25, 1976 l

t Shutdtun Evonss_

In order to provide for control rod drive tests, xero power physics tese.ing, and startup procedures, there is provision for bypassing certain seg=ents of The reactor protection system seg=ents which the reactor protection system.

Two conditions are imposed when can be bypassed are shown in Table 2.3-1A.

2.3 'B 2.3-ic the bypass is used:

By administrative control the nuclear overpower trip set point must be 1.

reduced to a value 1 5.0% of rated power during reactor shutdown.

2.

A high reactor coolant system pressure trip setpoint of 1720 psig is auto =atically imposed.

The purpose of the 1720 psig high pressure trip set point is to prevent normal This high operation with part of the reactor protection system bypassed.

pressure trip set point is lower than the nor=al low pressure trip set point The over be tripped before the bypass is initiated.

so that the reactor =ust power trip set point of f 5.0% prevents any significant reactor power from Sufficient natural being produced when performing the physics tests.circulatien (5) would be a the reactor coolant pu=ps were operating.

Two Pucp Coerstion A.

Two Loop Operation Operation with one pucp in each loop will be allowed only following reactor shutdown.

Af ter shutdcun has occurred, reset the pucp contact monitor power level trip setpoint to 55.0%.

B.

Single Loop Operation Single loop operation is permitted only after the reactor has been Af ter the pump centact =onitor trip has occurred, the following tripped.

actions vill permit single loep operation:

conitor power level trip setpoint to 55.00.

Reset the pu=p centact 1.

temperature Trip one of the two protective channels receiving outlet 2.

information from sensors in the Idle Loop.

3.

Reset flux-flow setpoint to 0.949 (Unit 1).

0.961 (Units 2,3)

REFERENCES (4) FSAR, Section 14.1.2.3 (1) FSAR, Section 14.1.2.2 (5) FSAR, Section 14.1.2.6 (2) FSAR, Section 14.1.2.7 (3) FSAR..Section 14.1.2.8 2.3-4 Amendment No. 20, 20 6 17 i

Plarch 25, 1976 W

I

w 2400 P = 2355 psig T = 619*F 2300 2200

~

m C.

f ACCEPTABLE

=

OPERATION 2l00 g

Z "e."

O 2000 M

j UNACCEPTAa'E M

OPERATION E

8 e

1900 c'

4 4

1 P = 1600 psig 1800 t

i I

I 540 560 580 600 620 640 Reactor Out :: T !.e.c e r a tu r a,

• F UillT 1 PROTECTIVE SYSTEM ?AXIFUM ALLOWABLE SET POINTS na OCONEE NUCLEAR STATION FIGURE 2.3-1A Amendment No. 20, 20 G 17 2.3-5 March 25, 1976 I

I Po::er Level, t 120 UNACCEPTABLE OPERATION

. 110 (ios.5)'

ACCEPTABLE Q

l 0P ION l

'e,

?

l l

- - 90 l

l e

c I

I

+

( -2 s. so) l 80(7s.33 (si,so)

IACCEPTABLE l

3 8 4 PUNP l' OPERATION..

70 l

l l

l 60 l

( -2a, s3. 3 )

(si.7}-l (31 s3.3)

ACCEPTABLE 50 l

2.3 & 4 l OPERATION l

PUMP 40 I

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I 30 l

(-28.26.2)

(31,26.2) l

-_ 20 g

U lT 10 7g is

, is ti n

i y

!e I

I e

1 i

-60

-40

-20 0

+20

+40

+60 Reacter Power Imbalance. 5

• THE FLUX / FLOW SETPOINT lIIIT 1 FOR 2/0 PUMP OPERATION PRDIECTIGl SYSTEM i',tXI!i IAUST BE SET AT 0.949 ALL0lABLE SETP0 lifts Ceninus.

OCONEE NUCLEAR STATI FIGURE 2.3-2A Amendment No. 20, 20 G 17

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2.3-11 k ndment No. 20, 20 6 17 March 25, 1976

4 k.

3 LIMITING CONDITIONS FOR.'PERATION 3.1 REACTOR C001. ANT SYSTD1 Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operation of the reactor coolant system co=ponents which must be met to ensure saf reactor operation.

Specification 3.1.1 Operational Components a.

Reactor Coolant Pumns 1.

Whenever the reactor is critical, single pu=p operatio'n shall be pro-hibited, single-loop operation shall be restricted to testing, and i

~

other pump combinations permissible for F ven power levels shall be as shown in Table 2.3-1.

2.

Except for test purposes and limited by Specification 2.3, power

('

ope-rtion with one idle reactor coolant pu=p in cach loop shall bc restt:cted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the reactor is not returned to an j

acceptabic RC pump operating combination at the end of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the reactor shall be in a hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.

The boron concentration in the reactor coolant system shall not be reduced unless at 1 cast one reactor coolant pump or one low pressure injection pump is circulating reactor coolant.

b.

Steam Cencrator 1.

One steam generator shall be operable whencver the reactor coolant average temperature is above 250 F.

c.

Pressurizer Safety Valves 1.

All pressurizer code safety valves shall be operable whenever the reactor is critical.

2.

At least one pressurizer code safcty valve shall be operable whenever all reactor coolant system openings are closed, except for hydrostatic tests in accordance with the ASME Section Ill Boiler and Pressure

{'

Vessel Code.

)

1 Amendment No, 20, 20 6 17 3.1-1 March 25, 1976

9 o

(

.(

Bases The. limitation on power operation with one idle RC pu=p in each loop has been imposed since the ECCS cooling performance has not been calculated in ac-cordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation.

A time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allovad for operation with one idle RC pump in each loop to effect repairs of the idle pu=p(s) and to return the reactor to an acceptable combination of operatir.g RC pumps.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The for this mode of operation is acceptable since this mode is expected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is considered very remote.

A reactor coolant pu=p or low pressure injection pu=p is required to be in operation before the boron. concentration is reduced by dilution with makeup Either pump will provide mixing which will prevent sudden positive water.

reactivity changes caused by dilute coolant reaching the reactor.

One low pressure injection pump will circulate the equivalent of the reactor coolant system volume in one-half hour or less. (1)

The low pressure injection system suction piping is designed for 300 F and 370 psig; thus the system with its redundant components can remove decay heat when the reactor coolant system is below this temperature.

(2,3)

One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than

(~

that required by the sum of the available heat scurces which are pomp energy, N*

pressurizer heaters, and reactor decay heat.

valves are required to be in service prior to criticality to conform to the(4) Both press system design relief capabilities.

The code safety valves prevent overpressure for a rod withdrawal accident at hot shutdown.

(5) The pressurizer code safety valve li f t setpoint shall be set at 2500 psig + 1* allowance for error and cach valve shall be capable of relieving 300,000 lb/hr of saturated steam at a pressure no greater than 3% above the set pressure.

REFERE"CES (1) FSAR Tables 9-11 and 4-3 through 4-7.

(2) FSAR Sections 4.2.5.1 and 9.5.2.3.

(3) FSAR Section 4.2.5.4.

(4) FSAR Sections 4.3.10.4 and 4.2.4.

(5) FSAR Sections 4.3.7 and 14.1.2.2.3.

f n))

I i

s Amendment No.

20. 20 6 17 March 2S, 1976 3.1-2

i 3.1.8 Single Loop Restrictions i

Specification The following special limitations are placed on single loop operation in addition to the limitations set forth in Specification 2.3.

Single loop operation is authorized for test purposes only and

~

3.1.8.1 requires prior Commission approval.

23 incore detectors meeting the requirements of Technical least 3.1.8.2 AtSpecification 3.5.4.1 and 3.5.4.2 shall be available throughout this test to check gross core power distribution.

I 3.1.8.3 The pump monitor trip setpoint shall be set at no gr' eater than 50 percent of rated power.

The outlet reactor coolant temperature trip setpoint shall be set 3.1.8.4 at no greater than 610 F.

15 percent of rated power and every 10 percent of rated power 3.1.8.5 At measurements shall be taken of each operable above 15 percent, incore neutron detector and each operable incore thermoco'uple, reactor coolant loop flow rates and vessel inlet and outlet temperature, and evaluation.of this data determined to be ac-ceptable before proceeding to higher power levels.

covering single loop operation. permitted by Specification 3.1.8.6 A report 3.1.8, shall be submitted within 90 days after completion of testing.

shall include the data obtained together with analyses This report and interpretations of these data which denonstrate:

(1) Coolant flows in the idle loop and operating loop are as predicted.

(2) Relative incore flur. and temperature profiles remain es-sentially the same as for four pump operation at cach power 26/21 level taking into account the reduced flow in single loop operation.

(3) Operating loop temperatures and flows are obtained which justify the revised safety system setting prescribed for the temperature and flow instruments located,in the operating loop (which must sense the combined core flow plus the cooler bypass flow of the idle loop).

l Bases The purpose of singic loop testing is to (1) supplement the 1/6 scale model test information, (2) verify predicted flow through the idle loop, (3) verify that changes in power level do not affect flow distribution or core power Amendment No. 20, 20 4 17 3.1-19 March 25, 1976 i

3.1.9 Low Power Physics Testing Restrictions Specification The following special limitations are placed on low power physics testing.

3.1.9.1 Reactor Protective System Requirements Below 1720 psig shutdown bypass trip setting limits shall apply in a.

accordance with Table 2.3-1A - Unit 1.

2.3-1B - Unit 2.

2.3-1C - Unit 3.

b.

Above 1800 psig nuclear overpower trip shall be set at less than 5.0 Other settings shall be in accordance with Table 2.3-1A - Unit 1.

Percent.

2.3-1B - Unit 2.

2.3-IC - Unit 3.

3.1.9.2 Startup rate rod withdrawal hold shall be in effect at all times. This applies to both the source and intermediate ranges.

3.1.9.3 Shutdown =argin cay not be reduced below 1.0% ak/k as required by Specification 3.5.2.1 vich the exception that the stuck red worth criterion does not apply during rod worth =easurements.

Bases Technical Specification 3.1.9.2 vill apply to both the source and intermediate ranges.

The above specification provides additional safety cargins during low power physics testing.

i 1

3 3.1-20 Amendment No. 20, 20 6 17 i

March 25, 1976 0

i r

3.3.2 In addition to 3.3.1 above, the following ECCS equipment shall be operable when the reactor coolant system is above 350 F and irradiated fuel is in the core:

(a) Two high pressure injection pumps shall be maintained operable to provide redundant and independent flow paths.

(b) Engineered Safety Feature valves and interlocks associated with 3.3.2a above shall be operable.

3.3.3 In addition to 3.3.1 and 3.3.2 above, the following ECCS equipment shall be operable when the reactor coolant system is above 800 psig:

(a) The two core flooding 3)anks shall each contain a minimum of 13 +

t

.44 ft. (1040 + 30 ft of borated water to 600 + 25 psig.

(b) Core flooding tank boron concentration shall not be less than 1,800 ppm boron.

(c) The electrically-operated discharge valves from the core flood tanks shall be open and breakers locked open and tagged.

the electrically-operated core flood tank vent valves CF-5 and CF-6 l

(d) shall be closed and the breakers locked open and tagged except when

(

adjusting core flood tank pressure.

(e) One pressure instrument channel and one icyc1 instrument channel per core flood tank shall be operable.

3.3.4 The reactor shall not be made~ critical unicss the following equipment in addition to 3.3.1, 3.3.2, and 3.3.3 is operabic.

(a) The other reactor building spray pump and its associated spray nozzle header.

(b) The remaining reactor building cooling fan and associated cooling unit.

(c) Engineered Safety Feature valves and interlocks associated with 3.3.4a and 3.3.4b shall be operable.

3.3.5 Except as noted in 3.3.6 belou, tests or maintenance shall' be allowed during power operation on any component (s) in the high pressure injection, low pressure injection, low pressure service water, reactor building j

spray, reactor building cooling or penetration roem ventilation systems which will not remove more than one train of each system from service.

Components shall not be removed from service so that the affected system train is inoperable for more than 24 consecutive hours.

If the system is not restored to meet the requirements of Specification 3.3.1, 3.3.1, 3.3.3, or 3.~.4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the requitecents of Specification 3.3.1, 3.3.2, 3.3.3, or 3.3.4 are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a condition below that reactor coolant j

c system condition required in Specification 3.3.1, 3.3.2, 3.3.3, or 3.3.4 for the component degraded.

Amendment No. 20, 20 6 17 3.3-2 March 25, 1976

=

If within one (1) hour of determination of an inoperable rod, g.

it is not determined that a 1%ik/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the het standby condition until this margin is established.

Following the determination of an inoperable rod, all rods shall h.

be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and ex'ercised weekly until "the rod problem is solved.

If a control rod in the regulating or safety rod groups is 1.

declared inoperable, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump com-bination.

If a control rod in the regulating or axial power shaping groups J.

is declared inoperable, operation above 60 percent of rated power may continue provided the rods in the group are positioned such that the red that was declared inoperable is caintained within allowable group average position limits of Specification 3.5.2.2.a and the withdrawal limits of Specification 3.5.2.5.c.

3.5.2.3 The worths of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the control rod casition limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Power Tilt a.

Except for physics tests, if the maximum positive quadrant power tilt exceeds +3.41% Unit 1, either the quadrant power tilt shall 4.92% Unit 2 4.92% Unit 3 be reduced to less than +3.41% Unit I within two hours or the 4.92% Unit 2 4.92% Unit 3 following actions shall be taken:

(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced below the power level cutoff (as identified in specification 3.5.2.5) and further reduced 1.

by 2% of full power for each 1% tilt in excess of 3.41% Unit 4.92% Unit 2 4.92% Unit 3 (2) If less than four reactor coolant pumps are in operation, the allowable thernal power for the reactor coolant pump combination shall be reduced by 2% of full power for ecch 1% tilt.

s Amendment No.' 20, 20 6 17 3..* -7 March 25, 1976

o t

(3) Except as provided in specification 3.5.2.4.b the reactcr shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 3.41% Unit I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.92% Unit 2 4.92% Unit 3 b.

If the quadrant tilt exceeds +3.41% Unit I and there is simultaneous 4.92% Unit 2 4.92% Unit 3 indication of a misaligned controi rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60%

of the thermal power allowable for the reactor coolant pump combination.

c.

Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, 11.07% Unit 2 11.07% Unit 3 a controlled shutdown shall be initiated immediately, and the reactor shall be brought to the hot shutdown condition within four hours.

d.

Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the ther:al power and the power range high flux serpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each I percent tilt for the maximum tilt observed prior to shutdown.

e.

Quadrant power tilt shall be conitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.

3.5.2.5 Control Rod Positions Technical Specification 3.1.3.5 does not prohibit the exercising a.

of individual safety rods as required by Table 4.1-2.or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

l b.

Operating rod group overlap shall be 25% + !% between two sequential groups, except for physics tests.

Except for physics tests or exercising control rods, the control c.

rod withdrawal li=its are specified on Figures 3.5.2-1A1 and 3.5.2-1A2, (Unit 1), 3.5.2-1B1, 3.5.2-132 and 3.5.2-133 (Unit 2), l and 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump s operation and on Figures 3.5.2-2A1,3.5.2-2A2 (Unit 1), 3.5.2-2B

\\

(Unit 2), and 3.5.2-2C (Unit 3) for three or t,wo pump. operation.

If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable I

3.5-8 Amendment No. 20, 20 6 17 March 25, 1976'

i control rod position. Acceptable control rod position shall then be attained within two hours. The minimum shutdown nargin all times.

required by specification 3.5.2.1 shall be amintained at

~

l for physics tests, power shall not be increased above the

1),

d.

Except power level cutof f as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Unit 3.5.2-1B1, 3.5.2-1B2, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.

(1) The xenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.

(2) The xenon reactivity shall be asymptotically approaching the value for operation at the power level cutoff.

i

~

to Reactor power imbalance shall be monitored on a frequency not 3.5.2.6 rated power.

exceed two hours during power operation above 40 percent for physics tests, imbalance shall be maintained within the Except i

envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B, and If the imbalance is not within the envelope defined by 3.5.2-3C.

Figure 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B. and 3.5.2-3C, corrective If an measures shall be taken to achieve an acceptable imbalance.

acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.

The control rod drive patch paucis shall bc locked at all times with 3.5.2.7 limited access to be authorized by the manager.

If 3.5-9

^**ndment No. 20, 20 6 17 March 25; 1976

v t

Bases The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2 3.5.2-3B, and 3 5.2-3C is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Final Acceptance Criteria.

Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary.

Operation in a eituation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distri-bution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.**

Conservatism is introduced by application of:

a.

Nuclear uncertainty factors b.

Thermal calibration c.

Fuel densification effects d.

Hot rod manufacturing tolerance factors The 25% + 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.

Control rods are arranged in groups or banks defined as follows:

Group Function l

1 Safety 2

Safety 3

Safety 4

Safety 5

Regulating 6'

Regulating 7

Xenon transient override 8

APSR (axial power shaping bank)

The rod position limits are based on the most limiting of the following three criteria:

ECCS power peaking, shutdown margin, and potential ejected rod worth.

Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The misi=um available rod worth, consis-tent with the rod position limits, prevides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod,that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.5" ak/k (Unit 1) or 0.65% ak/k (Units 2 and 3) at rated power.

These values have been shown to be safe by the safety analysis (2,3,4) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% ak/k is allowed by the rod positions limits at hot zero power.

A single inserted control red worth of 1.0; ik/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, there-fore, less severe environmental consequences than a 0.5% ak/k (Unit 1) or 0.65% ak/k (Units 2 and 3 ejected rod worth at rated power.

• • Actual operacing limits depend on whether or not incere or excere dete: tors are used and their respective instrument and calibratica errors.

The cethod used to define the operating limits is defined in plant operating procedures.

s 3.5-10 Amendment No. 20, 20 6 17 March 25, 1976

.s...

..._ m

n.

Control red groups are withdrawn in esquznce beginning with Group 1.

Groups 5, 6, and 7 are overir.pecd 25 percent.

The normal position at power is for Groups 6 and 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established with consideration of potential effects of rod bowing (Unit and fuel densification to prevent 1 only) associated with a positive quadrantthe linear heat rate peaking increase power tilt during normal power operation f rom exceeding 5.10% for Unit 1.

The limits shown in Specification 3.5.2.4 7.36% for Units 2 & 3 are measurement system independent.

The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.

The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally vill be perfor=ed in the process The two-hour frecuency for monitoring these quantities will computer.

provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours withour specification violation.

Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specification 3.5.2.5d to prevent excessive power peaking by transient xenon.

The xenen 1

reactivity must be beyond the "undershcot" region and asytptotically approaching its equilibrium value at the power level cutoff.

REFERENCES IFSAR, Section 3.2.2.1.2 2FSAR, Section 14.2.2.2 FSAR, SUPPLEMENT 9 B&W FUEL DENSIFICATION REPORT BAN-1409 (UNIT 1)

BAW-1396 (UNIT 2)

BAW-1400 (UNIT 3) i l

d v

T e

Amendment No. 20, 20 6 17 3.5-11 March 2S, 1976

(

RCO POSITICN LIM.LTS FCR 4 PLNP CPERATICN APP'._ICA2t_E TC THE PERICO FRCM 0 TO 230 2 5 EFPD (iss.ioz) (22s.io2) 100 RESTRICTED REGION POWER LEVEL CUTOFF 226.92) 90 -

(,,,,,,)

(soo.so)

(i72.so) 80 OPERATION IN THIS REGION RESTRICTED 15 NOT ALLOWED REGION 60 E

g so

. (iss.so) c.

2 MINIMUM SHUT 00WN 40 MARGIN LIMIT PERMISSIBl.E d

OPERATING REGION E

30 a.

20

~

. (88.15) 10

(" (0.6) 0 0

100 200 300 Rod index (5 withdrawn) 0 25 50 75 100 0

25 50 75 100 g

g g

f I

Group 7 Group 5 I

1 l

0 25 50 75 100 i

i

(

Grbup 6 i

Rod inaer is tne percentage sum of tne witnarawal of' Groups 5.6 and 7.

LIIIT 1 IMITS

% g RCD POSITICil L umin CCONEE NUCLEAR STATION

,U FIGURE 3.5.2-1A1 3.5-12 Amendnent No. 20, 20 6 17 March 25, 1976

,,~,

RCO,PCSITICN ' I.MITS FCR 4 PLW CPERATIC" APPLICABLE TO THE PERICL, AFTER 230 2 5 EFPO (2ss.1o2)(295. 802) 100 (25o.1o2)

• O POWER LEVEf CUTOFF (2ss. s2 )~ * ( 29*. 92,.

-a 90 80 (soo.m, 3 OPERATION IN THIS REGION RESTRICTEC

~

IS NOT ALLOWED IONS 60

=

=y 50

• ('88 50)

O 40 E

MINIMUM SHUTOOWN 30 o

MARGIN LIMIT PERMISSIBLE 20 OPERATING (los.is)

REGION e

10 (o.5) 0 i

0 100 200 300 Rod index. 5 Withdrawn 0

25 50 75 100 0

25 50 75 100 e

t t

t e

I e

t e

i Group 5 Group 7 0

25 50 75 100 i

t f

e

)

Group 6 Rod index is the percentage sum of the withdrawal of Grsuas 5.5 and 7 tilIY l OIM RCDFCSITICNLIMIT OCONEE NUCLEAR STATION FIGUPE 3.5.2.!A2 Amendment No. 20, 20 6 17 3.5-13 March 25, 1976

o i

ROD POSITICN LAMITS FCR 2 AND 3 PLMP CPERxs'ICN APPLICABLE TO THE PERIOD FRCM 0 TC 230 t 5 EFPC (145.1o2)'

(175.102) 100 OPERATION IN THIS REGION OPERATION IN THIS REGION 90 13 NOT Al.LOWE0 WITH 2 OR IS NOT All.01E0 WITH 3 3 PUMPS PUMPS E 80 E

E y

70 PERMISSISLE 5

OPERATING

[

60 REGION

=

0 50

. ('55 50) j 40 MINIMUM SHUT 00sN URGW 30 LIMIT

~

20 (72,15) g n.

10 b (o.6) 0 e

i i

0 100 200 300 Rod index, " Witnarawn 0

25 50 75 100 0

25 50 75 100 0

1 t

f a

t t

t t

t Group 5 Groua 7 0

25 50 75 100 t

t t

Grcup 6 Roo inces is the p e rc en ta ge som of the withorawal of Groups 5.5 ans 7 IJIIT 1

'. g ROD POSITICf1 LIf1ITS l

cat 'entti OCONEE NUCLEAR STATION W

RGURE 3.5.2-2Al 3.5-18 i

endment No. 20, 20 6 17 l

March 25, 1976

(

R00 POSITICN LIMITS FOR 2 ANO 3 Pl>P CPERATION APPLICABLE TO THE PERICO AFTER 230 t 5 EFPD (205.102)

(236.102) 100 g

OPERATION IN THIS REGION IS NOT ALLOWED WITH 2 OR 3 PUMPS OPERATION IN

.=

THIS REGION IS 80 g

NOT ALLOWE0

'j WITH 3 PUMPS e

m.

60 n-E 3

(iss.so)

E j

E 40 5

MINIXUM SHUTOONN 3

MARGIN LIMIT PERMISSISLE g

n OPERATING 20 REGION g

j (106.15) b (o.6) i e'

9

+

i i

e 0

50 100 150 200 250 300 Rod index. 5 Withdrawn 0

25 50 75 100 0

25 50 75 100 t

t t

t a

t t

t a

Group 5 Group 7 0

25 50 75 100 Group 6 Rod indes is the percentage sum o f the wi thdrasal of 3roucs 5.5 and 7.

LUIT 1 9.FDDFOSITIGlL OCONEE NUCLEAR STATION Frans 3.5.2-2.a2 ;

3.5-18a b endment No. 20, 20 6 17 March 25. 1976

..o Power, ', of 2568 M t RESTRICTEo REGlon-(102.-15) 9 100 --

(s2 -is) 90 -

80 --

• (so.+20)

(so.-20) 70 -

PERMIS$181.5 OPERATt1G 60 REGian 50 40 - -

30 20 10 30

-20

-10 0

+10

+20

+30 Core imoalance. 5 lJilIT 1 OPEPATIG!AL FGIR ItTALEE B.6ELCFE FOR CFEFATICN FFCM 0 TO 230 t 5 E=D out rem OCONEE NUCt. EAR STATION FIGWE3.5.2-3Al g

Y

~

3.5-21 Amendment No.

20, 20 6 17 March 25, 1976

~

~

(

Power, 5 of 2568 Mt RESTAICTEo REGlon

(+ 5,102)-

(-lo,102) '

100

(+ 5,s2)

(-tz,st) 90 1

80

(+20,ao)

(-20.So) 70 PERM 13313L*

OPERATING 50 REGlon 40 30 20 10

+k0 +20

+30 k0

.0

-30

-20 Core Imaalancs.,3 tillT 1 CPEPATIGlAL FGER li"FJLA':CE BNELCFE FCR CFEPATICil ANtx

, A 230 t 5 EFFD OCONEE NUCLEAR STATION qut Fiems3.5.2-3A2 H

l 3.5-21a bendmentNo.20,20617 i

1976 March 257

t 20

~

r 18

)

M,-

x

/

/

/

~

h 16

/

$ 14 3

o' GE.NE.9IC I

2 g

UNIT 1. SATCH 4 ---

h 12 2

10 O

2 4

6 8

10 12 Axial Location of Peak Poner From Bottem o f Core. f t LOCA LIMITED i'AXIful ALLD!AEli LIIEE FEAT PATE 9 OCONEE NUCLEA FIGURE 3.5.2 !4 3.5-24 Amendment No. 20, 20 6 17 March 25, 1976 r _ _..

heme.

i Tchla 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Test Frequency' Item

' Movement of Each Rod Bi-Weekly 1.

Control Rod Movement 50% Annually 2.

Pressurizer Safety Valves Setpoint 25% Annually 3.

Main Steam Safety Valves Setpoint 4.

Refueling System Interlocks Functional Prior to Refueling 5.

Main Steam Stop Valves Movement of Each Step Monthly I1}

Valve 6.

Reactor Coolant System Evaluate Daily Leakage 7.

Condenser Cooling Water Funcr' nal Annually System Gravity Flow Test Functional Monthly 8.

High Pressure Service Water Pumps and Power Supplies Prior to Functional 9.

Spent Fuel Cooling System Refueling 10.

Hydraulic Snubbers on Visual Inspgetion Annually Safety-Related Systems I

High Pressure and Low ( )

Vent Pump Casings Monthly and Prior 11.

to Testing Pressure Injection System

12. ~ Reactor Coolant System Flow Validate F1'ow~to be Once Per Fuel Cycle j

at least:

]

Unit 1 141.30 x 106 lb/hr 6

Unit 2 131.32 x 10 lb/hr Unit.3 131.32 x 10 lb/hr (1) Applicabic only when the reactor is critical.

(2) Applicable only'when the reactor coolant is above 2000F and at a steady-state temperature and pressure.

(3) Operating pumps excluded.

Amendment No.~ 20, 20 4 17

'4.1-9 March 25, 1976 o

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