ML19322B417

Amend 10 to License DPR-55 Incorporating Tech Spec Change 10.Proposed Tech Specs Encl

ML19322B417
Person / Time
Site:	Oconee
Issue date:	10/31/1975
From:	Purple R Office of Nuclear Reactor Regulation
To:
Shared Package
ML19322B414	List: ML19322B413 ML19322B415 ML19322B416 ML19322B417 ML19322B418 ML19322B419
References
NUDOCS 7912020264
Download: ML19322B417 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION g

WASHINGTON. D. C. 20555 DUKE POWER COMPANY a

DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment' No.10 License No. DPR-55 1.

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

A.

The application for amendment by Duke Power Gompany (the licensee) dated August 26, 1975, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

2.

Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-55 is hereby amended to read as follows:

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"B.

Technical Spscifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change 'No.1 0."

3.

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

FOR TIIE NUCLEAR REGULATORY C05NISSION Originas signed by R.A.I'urple Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing

Attachment:

Change No.10 to the Technical Specifications Date of Issuhnce:

OCT 31 1975 I

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4 ATTACIMENT TO LICENSE A!DIDMENTS i

EE@E.YNO.13 TO 'FACILIW LICENSE NO. DPR-38 O!ANGE TO. g g TO TEONICAL SPECIFICATIONS; O

AMT.ND T G 0.13 TO FACILITY LICENSE NO. DPR-47 CIIANGL*NO.1 g TO TECl"iICAL SPECIFICATIONS; MCGGT NO.10 TO FACILIU LICENSE NO. DPR-55 CIANGE NO.10 TO TEQt1ICAL SPECIFICATIONS DOCEET HOS. 50-269, 50-270 AND 50-287

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Revise Appendix A as follows:

, i-f Recove pages 2.1-3b, 2.1-8, 2.1-9, 2.3-1, 2.3-2, 2.5-3, 2.3-4

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2.3-0, 2.3-10, 2.3-12, and 2.5-13; and insert identicit13y i

nu:nbered pages.

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Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.

The specified flow rates for Curves 1, 2, 3, and 4 of Figure 2.1-2B correspond 2.1-2C to the expected minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively.

The curve of Figure 2.1-1B is the most restrictive cf all possible reactor 2.1-1C coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.

2.1-3C The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR 2.1-3C of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 15%,(3) whichever condition is more restrictive.

Usin'g a local quality limit of 15 percent at the point of minimum DNBR as a basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even 2.1-3C though the quality of the exit is higher than the quality at the point of minimum DNBR.

The DNBR as calculated by the W-3 correlation continuall'y increases from point of minimum DNBR, so that the exit DNBR is 1.7 or higher, depending on the pressure.

Extrapolation of the W-3 correlation beyond its published quality range of +15 percent is justified on the basis of experimental data.(4)

The maximum thermal power for three pump operation.is 86.4% - Unit 2 23 86.4% - Unit 3

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I8 due to a power level trip produced by the flux-flow ratio 75% flow x 1.07.= 80%

IO 1.07 = 80%

power plus the maximum calibration and instrument error.

The maximum thermal power

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for other coolant pump conditions are produced in a similar manner. A flux-floci 23 ratio of 0.961 is used for single lonp conditions.

'l'18 For each curve of Figure 2.1-3B, a pressure-tempcrature point above and to the l'O

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2.1-3C lef t of the curve would result in a DNBR greater than 1.3' ar a local. quality at the point of minimum DNBR less than 15 percent for that particular reactor coolant pump situation.

The 1.3 DNBR curve for four-pump operation is more restrictive than any other reactor coolant pump situation because any pressure /

temperature point above and to the lef t of the four-pump curve will be above and to the lef t of the other curves.

REFERENCES 1

(1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FSAR, Section 3.2.3.1.1.k 2.1-3b gg] S 1 YOb

i Thermal Power Level, %

120

(-24.0,112)

(+20.0,112) kw/ft Limit kw/ft Limit 100

(-40.0, 97)

(+20.0, 86.4)

(-40.0, 86.4)

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(+40.0,80)

.'(+40.0, 73) 0MBR Limi

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(+26.0, 58.9)

.. 60 4

(-40,58.9)

(+40.'0,48)

ORBR Limit /.

(+16.5,58.9)

-- 40 (440.0, 39) 20 e

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-60

-40

-20 0

+20

+40 460 Reactor Power Imbalance, *'

CURVE REACTOR COOLANT FLOW (LB/ljR) 6 1

131.3 x 106 2

98.1 x 10 6

3 64.4 x 10 6

4 60.1 x 10 CORE PROTECTION SAFETY LIlilTS UNIT 2

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o' ti t\\ OCONEE NUC EAR STATION k-

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Figure 2.1-sB 2.1-8 OCT 311975 t

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TiiEPJ4AL PO'!ER LEVEL, %

. 120

(-27.0,112)

(+22,112) kw/ft Limit kw/ft Limit

(-4.98.5)

(+20.0,86.4) l

(-40.0,86.4)

(+40.0,84)

DNBR Limit >~

O (+40.0, 73)

'(+26.0,58.9)

Limit

. 60

(-40, 58.9) h DNBR

(+40.0,48)

Limit

~-- 40 (440.0,39)

(+16.5, 58.9) 20

-60

-40

-20 0

+2'0

+40

+60 Reactor Power Imbalance, %

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CURVE REACTOR COOLANT FLOW (LB/IIR) 6 l

~ 131.3 x 10 2

. 98.1 x 106 3

64.4 x 106 4

60.1 x 10 CORE. PROTECTION SAFETY i_1MITS k

UNIT 3 OCONEE NUCLNAR STATION o

Figure 2.1-2C

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2.1-9 OCT 311975

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2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies'to instruments monitor'ing reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, nuaber of pumps in operation, and high reactor building pressure.

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

Specification The reactor protective system trap 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 Figure 2.3-2Al } Unit 1

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2.3-2A2 2.3-2B - 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% (0.0% for

.a.

Unit 1) of rated power.

b.

Loss of two pumps in one reactor coolant loop and reactor power level is greater t'han 0.0% of rated power.

(Power /RC pump trip setpoint is reset t 55% of rated power for single loop operation and for Units 2 and 3, the 23 flux-flow setpoint must be set at 0.961 prior to single loop operation.

yp Power /RC pump trip setroint is reset to 55% for all modes of 2 pump i

gp operation for Unit 1.)

Loss.of one or two pumps during two-pump operation.

c.

Bases The reactor protective system consists of four instrument channels to monitor each of seversi selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.

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

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

Nuclear Overpower l

n A reactor trip at high power level (neutron flux) is provided to prevent l

damage to the fuel cladding from reactivity excursions too rapid to be detected l

by pressure and temperature measurements.

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2.3-1 N

i During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 105.5% of rated power.

j Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actu-ated could.be 112%, which is more conservative than the value used in the safety analysis.(4)

Overpower Trip Based on Flow 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 accommodate the most severe thermal transient considered in the design, the loss-of-coolant.

Analysis has demonstrated that the specified flow accident from high power.

power-to-flow' ratio is adequate to prevent a DNBR of less than 1.3 should

,a low flow condition exist due to any electrical malfunction.

i ides

.The power level trip set point produced by the power-to-flow rat o prov

,l both high power level and low flow protection in the event the. reactor power I

level increases or the reactor coolant flow rate decreases.

The power level trip set point produced bythe power-to-flow ratio provides overpower DNB pro-tection for all modes of pump operation." For every flow rate there is a maxi-

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. mum permissible power level, and for every power level there is a minimum permissible low flow rate.

Typical power level and low flow rate combinations for the pump situations of Table 2.3-1A are as follows:

Trip would occur when four reactor-coolant pumps are operating if power 1.

is 108% and reactor flow rate is 100%, or flow rate is 93% and power level is 100%.

Trip would occur when three reactor coolant pumps are operating if power 2.

is 81.0% and reactor flow rate is 74.7% or flow rate is 69% and power level is 75%.

Trip would occur when two reactor coolant pumps are operating in a single 3.

loop if power is 59% and the operating loop flow rate is 54.5% or flow 33 rate is 43% and power level is 46%.

(For Tables 2.3-1B and 2.3-1C the valu>a g are 52% power if the operating loop flow rate is 54.5% or flow rate is 48%

[ '. I and power level is 46%.)

10 Trip would occur when one reactor coolant pump is operating in each l'oop 4.

(total of Fwo pumps operating) if the power is 53% and reactor. flow rate is 49.0% or flow rate is 45% and the power level' is 49%.

For safety calculations the maximum calibration and instrumentation errors for l

the power level trip were used.

The power-imbalance boundaries are established in order to pre' rent reactor l

thermal limits from being exceeded. These thermal limits are either power peaking-kw/ft limits or DNBR' limits.

The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power i

' ' level trip produced by the power-to-flow ratio such that the boundaries of Figure 2.3-2Al') Unit 1 are Produced.

The power-to-flow ratio reduces'the power 2.3-2A2

~2.3-2B ~ - Unit 2 -

N 2.3-2C -Unit 3 2.3-2 OCT 311975

l' vel trip and associated rebetor power / reactor power-imbalance boundaries by e

1.08% - Unit 1 for a 1%. flow reduction.

l.07% - Unit 2 1.07% - Unit 3 23 For Units 2 and 3, the power-to-flow reduction factor is 0.961 during single 1g loop operation.

p p

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

The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio.

The pump monitors also restrict the power level for the number of pumps in operation.

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Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure set point is reached before the nuclear overpower 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 (1985) psig and variable low pressure (13.77 Tout-618D trip (1800) psig (16.25 T

-7756)

(1800) psig (16.25 Tout-7756) setpoints shom in. Figure 2.3-1A have been established to maintain 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 variabic low reactor coolant system pressure trip value of (13.77 Tout - 6221)

(16.25 T

-7796)

(16.25 T,

- 96) out Coolant Outlet Temperature 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 and instrumentation errors, the safety analysis used a trip set point of 620 F.

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 occident, even in the absence of a low reactor coolant system pressure trip.

2.3-3 OCT 3 A 1975

T Shatdown Bypass In order to provide for control rod drive tests, zero power physics testing, and star' tup procedures, thereis provision for bypassing certain segments of The reactor protection system segments which the reactor protection system.

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

2.3-1B 2.3-1C the bypass is used:

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

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

A high reactor coolant system pressure trip setpoint of 1720 psig is 2.

automatically imposed.

The purpos'e 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 normal low pressure trip set point The so'that the reactor must be tripped before the bypass is initiated.

over power trip set point of < 5.0% prevents any significant reactor power Sufficient natural from being produced uben performing the physics tests.

circulation (5) woulo be available to. remove 5.0% of rated power if none of the reactor coolant pumps were operating.

Two Pump Operation A.

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

After shutdown has occurred, the following actions will permit operation with one pump in cach loop:

1.

Reset the pump contact monitor power level trip setpoint to 55.0%.

2.

(Unit 1) Reset the protective system maximum allowable setpoint.as shown in Figure 2.3-2A2.

B.

Single Loop Operatior

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Singic loop operation is permitted only after the reactor has been tripped.

After the pump contact monitor trip has occurred, the following actions will permit single loop operation:

1.

Reset the pump contact monitor power level trip setpoint to 55.-0%.

2.

Trip one of the two protective channels receiving' outlet temperature information from sensors in the Idle Loop.

3.

(Unit 1) Reset the protective system maximum allowable setpoints as shown in Figure 2.3-2A2.

Tripping one of the two protective channels receiving outlet temperature information from the idle loop assures a protective system trip logic of one out of two.

f3 4.

(Units 2 and 3) Reset flux-flow setpoint to 0.961.

gg yp REFERENCES (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 (4) FSAR, Section 14.1.2.3 2.3-4

-0CT 311975

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Power Level, %

-- 120

(-14,107)

(+10,107)

FOUR PUMP SETP0ll!TS 100 TilREE PUMP SETI'0]IlTS

(-33,85) 80 (10,79.9

(-14, 79.9) 1110 PUMP

(.

(-33,57. 9)

SETP0ll:TS

" 60 i

(+25,57.9) i

(+10,52.4)

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(-14,52.4)

. 40 i

(-33,30.4)

(+25,30.4) 20

[

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-60

-40

-20 0

+20

+40

+60 I

1 1

Power Imbalance, %

• For two pumps in one loop, tbc flux-flow setpoint must be 0.961.

11 ES hT UNIT 2

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2.3-9 9ure 2.3-2B 00131 B75

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Power Level, %.

a

- 120 FOUR PUMP SETP0lllTS

(-17,107)

(+11.107) 100

(+25,89)

THREE PUMP

(-33* 86)

SETP01tlTS

(+11,79.9)

R0

(-17,79.9)

(+25,61.9)

(-33,58.9)

TWO PUMP

- 60 SETP0lllTS

(+11,52.4)

.(-17,52.4)

_ 40 s

- (- 33,' 31.~4 )

(+25,34.4)

-- 20

-60

-40

-20 0

20 40 60

^

Power Imbalance, %

• For two pumps in one loop, the flux-flow setpoint must be 0.961 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 3 fu ab. OCONEE NUCLEAR STATION Figure' 2,3-2C

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1 2.3-10 OCT 311975

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Table 2.3-1B s

Unit 2 Resetor Protective System Trip Settinz 1.imits One Reactor Bro Reactor Coolant Pumps Coolant Pump Three Reac*or Four Reactor Operating in A Operating in Coolant Pumps Coolant Pumps Single Loop Each Loop Shutdown Operating Operating (Operating Power (Operating Power (Operating Power (Operating Power

-75% R.*ted)

-46% Rated)

-49 Rated)

Bypass __

-100% Rated)

I3)

RPS Serment 105.5 5.0 105.5 105.5 105.5 1.

Nuclear Power Max.

(I Rated) 0.961 times flow 1.07 times flow sypassed 2

1.07 tintes flow i

2.

Nuclear Power Max. Based minus reduction minus reduction alous reduction 1.07 times flow on Flov (2) and Imbalance, due to imbalance due to imbalance due to imbalance 0

minus reduction due to imbalance Bypassed (I Rated) 55% (5)(6) 55%

NA 3.

Nuclear Power Max. Based NA 1720 'I

,on Pump Monitors. (%. Rated) 2355 2355 2355 2355 4.

High Reactor Coolant System Pressure, psig, Max.

,ha 1800' Bypas?ed 1800 La 5.

14v Reactor Coolant 1800 1800 (16.25 T,,g-7756)III (16.25 T,,g-7756)III (16.25 T,,g-7756)III (16.25 T,,g-7756)(II Bypassed System Pressure, psig, Min.

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6.

Variable Low Reactor Coolant Systeni Pressure 619 psig. Min.

619 (6) 619 619 619 7.

Reactor Coolar.t Temp.

4 F., Max.

4 4

4 4

8.

High Reactor Building Pressure, psig. Fm.---

(5) Reactor power level trip set point produced (1) T,,, is in degrees Fahrenheit ('F).

by pump 'centact monitor reset to 55.0%.

(2) Reactor Coolant Systen Flow, 1.

(6) Specification 3.1.8 applies. Trip one of the two protection channels' receiving outlet temper-(3) Administratively controlled reduction set ature information from sensors in the idle loop.

only during reactor shutdown.

'(4) Automatics 11y set when other segments of the U 3 are bypassed.

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Table 2.3-IC L

Unit 3 Reactor Protective Systen Trip Setting Limits Dro Reactor One Reactor Tour Reactor Three Reactor Coolant Pumps coolant Pump Coolant Pumps Ceolant Pumps Operating in A Operating is Operating Operating Single Loop Each Loop Shutdown (Operating Power (Operating Power (Operating Power (Operating Power

-100% Rated)

-75% Rated)

-46% Rated)

-49% Rated)

Bypass I3I RPS Segment _

105.5 5.0 105.5 105.5 105.5 1.

Nuclear Power Max.

eE[ g (I Rafed) 0.961 times flow 1.07 times flow syp.

1.07 times flow 1.07 times flow minus reduction minus reduction minus reduction minus reduction

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2.

Nuclear Poser Max. Based due to imbalance due to imbalance due to imbalance due to imbalance 5

on Flow (2) and labalance,

(% Rated) 55% (5)(6) 53%

Bypassed NA 3.

Nuclear Power Max. Based NA I

on Pump Monitors. (2, Rated) 1720 2355 2355 2355 2355 4.

High Reactor Coolant pay System Pressure, psig. Max.

1800 1800 Bypassed I

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'5. Low Reactor Coolant 1800 1800 System Pressure, psig Min.

1 (16.25 T

-7756)(1)

Bypassed

-7756)(1)

(16.25 T,,g-7756)(1I (16.25 T,,,-7756)III 6.

Variable Low Reactor 0 6.25 Tout Coolant System Pressure psig. Min.

619 619 (6) 619 -

619 619 7.

Reactor Coolant Temp.

i j

F., Max.

4 4

4 4

8.

High Reactor Build'ing 4

Pressure, psig. Max'.

(5) Reactor power level trip set point produced is in degrees Fahrenheit ('F).

by pump contact sionitor reset to 55.02.

(1) T,,g (2) Reactor Coolant Systen Flow, 2.

(6) Specification 3.1.8 applies. Trip one of the two protection channels receiving outlet temper-

,(3) Administratively controlled reduction set ature information from sensors in the idle loop.

only durir.g reactor shutdown.

(4) Autonatic111y set when other segments of H

the U 3 are bypassed.

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