ML19331E151

Amends 6,6 & 3 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Changing Tech Specs Requirements for Fuel & Core Design Re Second Fuel Cycle Operation

ML19331E151
Person / Time
Site:	Oconee
Issue date:	11/26/1974
From:	Goller K US ATOMIC ENERGY COMMISSION (AEC)
To:
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ML19331E150	List: ML19331E149 ML19331E151 ML19331E152
References
NUDOCS 8009050562
Download: ML19331E151 (97)

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-[4 UNITED STATES

-g urOMIC ENERGY COMMISSION p

WASHINGTON. D.C. 20545 g

7 tra DUKE POWER COMPANY DOCKET No. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 6 License No. DPR-38 1.

h Atomic Energy Commission (tha Commission) having found that:

A.

h application for amendment by Duke Power Company (the

~

licanuee) (ated September 20, 1974, as supplemented October 8 and 31, 1974, complias with the standards and requirements of the Atomic Energy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10 CFR Chapter I; 5.

h facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the C==4,= ion; C.

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

N issuance of this amendment will not be inimical to the common defense and security or to the tealth and safety of the public; i

and E.

Prior public notice of this amendment is not required since the amendmant does not involve a significant hazarda consideration.

2.

Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amen h nt and Paragraph 3.B of Facility License No. DPR-38 is hereby amended to read as follows:

ONM 80090&

"B.

Technical Specifications The Technical Specificationsi 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 laaued changes thereto through Change No. 16."

3.

Tais license amandment is effective as of the date of its issuance.

FOR THE ATOMIC ENERGY COMMISSION S

Karl R. Coller, Assistant Director for Operating Reactors Directorate of Licensing i

Attachment:

Change No. 16 to Technical Specifications Date of Issuance:

November 26, 1974 i

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ATTAR MENT TO LICENSE AMENDMENTS AMENDMENT NO. 6 TO FACILITY LICENSE NO. DPR-38, CHANGE NO.16 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 6 TO FACILITY LICENSE NO. DPR-47, CHANGE NO. 11 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 3 TO FACILITY LICENSE NO. DPR-55, CHANGE NO. 3 TO TECHNICAL SPECIFICATIONS; DUKZ POWER COMPANY OCONEE NUCLEAR STATION UNITS 1. 2. AND 3 DOCKET NOS. 50-269. 50-270. AND 50-287 Revise Appendix A as follows:

Insert New Pages Remove Fayg 2.1-1 & 2.1-2 2.1-1 & 2.1-2 2.1-3 2.1-3, 2.1-3a, 2.1-3b &

2.1-4 2.1-4 2.1-4a r

2.1-7 2.1-7 2.1-10 2.1-10 2.3-1 & 2.3.-2 2.3-1 & 2.3-2 2.3-3 & 2.3-4 2.3-3 & 2.3-4 2.3-5 2.3-5 2.3-8 2.3-8 & 2.3-8a 2.3-11 2.3-11 3.5-12 3.5-12 3.5-13 3.5-13 Blank page 3.5-18 3.5-18 3.5-21 3.5-21 N

O 2-Remove Pages Insert New Pages 3.5-24 3.5-24 3.11-1 3.11-1 3.5-6 & 3.5-7 3.5-6 & 3.5-7 i.

3.5-8 & 3.5-9 3.5-8 & 3.5-9 3.5-10 & 3.5-11 3.5-10 & 3.5-11 i

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2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability 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.

Sr cification The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in l

Figure 2.1-1A-Unit 1.

If the actual pressure / temperature point is below 2.1-15-Unit 2 2.1-1C-Unic 3 and to the right of the line, the safety limit is exceeded.

The combination of reactor thermal 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 2.1-25-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases - Unit 1 The safety limits presented for Oconee Unit 1 have been generated using BAW-2 l

critical heat flux (CHF) correlation (1)and the actual measured flow rate at 1

Oconee Unit 1 (2). This development is discussed in the Oconee 1, Cycle 2-Reload Report, reference (2). The flow rate utilized 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 normal q

operating conditions. This is accomplished by operating within the nucleate s

boiling regime of heat transfer, wherein the heat transfer coefficient is large enough 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 f rom 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 not an observable pa-ameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure L

2.1-1 NOV 2 6 1974 J

O-can b'e related to DNB. through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR*f, defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of-the margin 4

to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.32.

A DSBR of 1.32 corresponds to a 95 percent probability at a 99 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was j

4 assumed in reducing the pressure trip setponts to correspond to the elevated l

location where the pressure is actually measured.

The curve preJented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.32 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pu s are operating (minimum reactor coolantflowis107.6percentof131.3x1g0 lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential fuel densifi-cation effects, which result in a more conservative DNBR than any other shape that exista during normal operation.

The curves of Figure 2.1-2A are based on the more restrictive of two thermal g

limits and include the effects of potential fuel densification:

C 1.

The 1.32 DNBR limit produced by the combination of the radial peak, axial u

peak and position of the axial peak that yielda no less than a 1.32 DNBR.

2.

The combinatien of radis1 and axial peak that causes central fuel melting at the hot spot. The limit is 20.15 kw/ft for Unit 1.

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 pcwer peaking.

The specified flow rates fo. Curves 1, 2, 3 and 4 of rigure 2.1-2A correspond co the expected minimum flow rates with four pumps, three pumps, one. pump in s

cacheloop and two pumps in one loop, respectively.

The curve of Figure 2.1-1A is the most restrictive of all possible reactor i aolant pump-maximum thermal power combinations shown in Figure 2.1-3A (because the four-pump pressure - temperature restriction is known to be more li.iting than the 3 and 2 pump combinations, only -the four pump limit has ba n shown on Figure 2.1-3A).

n.. c...:.am tNr:a.51 rower f or :.hree-puue aperat im is 3 7 parce: - d u. *. :

p w r hvel trip produced by the fl'.i:(-t low ra:Le 75 parcent flow x 1.04 =

error.

The v! F r. en t pow r, plus the maximum catt5r it L n nd instrument r.ax inum therm.il power for other coolant pamp conditions are prodated in a s ;ni t.tr sanner.

2.1-2 NOV 2 61974 v-

For Figure 2.1-3A, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than 1.32.

The 1.32 DNBR curve for four-pump operation is more restrictive than any otl.er 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.

L6/11/2 References (1) Correlation of Critical, Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March, 1970.

(2) Oconee 1, Cycle 2 - Reload Report - BAW-1409, Sepetmeber, 1974.

2.1-3 NOV 2 6 B74

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0-s Bases - Units 2 and 3 To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is laric enough so that the clad surface temperatura is only slightly greatec than the coolant temperature. The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DN3). 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 not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the W-3 correlation.(1)

The W-3 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.3.

A DNBR of 1.3 corresponds to a 94.3 percent probability at a 99 percent confidence level that DNB will noc occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

15/. -

The curve presented in Figure 2.1-1B represents the conditions at which a 3

2.1-1C minimum 1:NBR of 1.3 is predicted for the maximum possible thermal power (112%)

when four reactor coolant pumps are operating (minimum reactor coolant flow is 131.3 x 106 lbs/hr). This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:

1.78;F

= 1.50 2.67; F F

=

=

an

\\

The design peaking combination results in a more conservative DNBR than any j

other shape that exists during normal operation.

1 The curves of Figure 2.1-2B are based on the more restrictive of two thermal 16/?1 1 l

2.1-2C 3

limits and include the effects of potential fuel densification:

1.

1he 1.1 !?:P,R lini t produced by a nuclear power peaking factor of Fj' = 2.67 o r t iw co.ab ina t ion a t llc radial peak, axial peak and position of tha 1

axial peak that yields no less than 1.3 DNBR.

J.

The comoination of rad Lal and axial peak that causes central fuel molting fl5/31. j at the hot spot. The limit is 19.8 kw/ft - Unit 2 19.8 kw/ft - Unit 3 3 3 2.1-3a l

NOV 2 6 *a74 j

.. =

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.

10/II!

The specified flow rates for Curves.1, 2, 3, and 4 of Figure 2.1-2B correspond 3

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.

L6/11/

The curve of Figure 2.1-1B is the most restrictive of all possible reactor 2.1-1C coolant pump-maximum thermal power combinations shown in Figure 2.1-38.

2.1-3C The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR

f6W, l3 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 minimus DNBR is equal to 15%,(3) whichever condition is more restrictive.

i 1

i Using a local quality limit of 15 percent at the point of minimus DNBR as a 16/1r basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even 3

2.1-3C though the quality of the exit is higher than the quality at the point of a

minimum DNBR.

The DNBR as calculated by the W-3 correlation continually increases from point of minimum DNBR, so that the exit DNBR is 1.7 or higher, depending on the Extrapolation of the W-3 correlation beyond its published quality pressure.

range of +15 percent is justified on the basis of experimental data.(4) i l

The maximum thermal power for three piap operation is 86% - Unit 2 15/11 86% - Unit 3 due to a power level trip produced by the flux-flow ratio 75% flow x 1.07 = SO%

),

1.07 = 80%

power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

I 16/11 For each curve of Figure 2.1-3B, a pressure-temperature point above and to the i

l 2.1-3C l

lef t of the curve would result in a DNBR greater than 1.3 or a local quality the point of minimum DNBR less than 15 percent for that particular reactor at 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 /

t.mperature point above and to the lef t of the four-pump curve will be nove

.n the'left.sf the o:.hur curves.

1 ITFERENCES

~

(1)- FS A2, Sec tion 3. 2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (3) FSAR, Sec t ion 3.2. 3.1.1.k 2.1-3b NOV 2 6 $74 l

e-

+

< (4) The following papers which were presented at the Winter Annual Meeting, r

ASME, November-18, 1969, during the "Two-phase Flow and Heat Transfer in Rod Bundles Symposium:"

- (a) Wilson, et al.

" Critical Heat Flux in Non-tiniform Heater Rod Bundles" (b) Cellerstedt, et al.

" Correlation of a Critical Heat Flux in a-Bundle Cooled by Pressurized Water" l

l 4

2.1-4 NOV 2 61974

2500 2400

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$oem ae 2200

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1900 560 580 600 620 640 660 Reactor Outlet Temperature, F t ; ;.-

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4 40 20 n

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+40 Reactor Power Imbalance, 7, CURVE REACTOR COOLANT FLOT (LB/4R) 6 1

131.3 x 10 6

2 99.1 x 10 3

3 S4.4 x 10 5

4 60.1 x 10 CORE PROTECilCN SAFETY Llx!TS UnT 1 h OCONEE NUCLEAR STATIOT

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NOV 2 019Y4 ngue 2.1-n f16/1U3 l-i

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1900 560 560 600 620 640 660 Reactor Outlet Temperature, F CORE PROTECTION SAFETY LIMITS

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2.1-3A f16/11/3 rigure NOV 2 e 1974

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.2.3 LIMITING SAPETY SYSTDi SETTINGS, PROTECTIVE INSTRU:fENTAfint; Applicability Appiles to instruments monitoring reactor power, reactor power igbain.t,

reactor coolant system pressure,. reactor coolant outlet temperatere. ::.w,

number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protective action to prevent any combination of.cacess 1

variables from exceeding a safety limit.

Specification The reactor protective system trip setting limits and the permissible cypuas,:s for the instrument channels shall be as stated in Table 2.3-1A - Unic 1 nd 2.3-1B - Unic 2 2.3-lc - Unit 3 16/11/3 Figure 2.3-1A1 } Unit 1 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:

h6/11/3 Loes 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 than 0.0% of rated power.

(Power /RC pump trip setpoint is reset to 55% of rated power for single loop operation. Power /RC pump trip setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)

16/11/'

c.

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

dases The reactor protective system consists of four instrument channela to c.onitor rio it each of several selected plant conditions which will cause a reactar any one of these conditions deviates from a pre-selected operatint; r. u

.o the degree that. safety limit may be reached.

The trip set ting limits for protective system instrumentatioc.,c u 1.

Table 2.3-1A - Unit 1.

The safety analysis has been based upa. u. -.

a t,.

2.3 Unic 2 2.1-1C - Un ir 3 dy:,te;3 instrut.e;'t :s t le.n trip set points plus calibratio7 *nd it -

w & L w i et.

.Lu h u 0.v r..g.,3r 1.

A eactor t cip at htgn power level (neutron flux) is provi; a...

daa. iga to the rual claitding f rom reactivity excursions too ::p!"

a by ptessure and wperature ueasure: tents.

2.3-1 NOV E 6 B74

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

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 react r coolant system flow is based on a power-to-flow ratio which has been established to accommodate the cost severe thermal transient considered in the design, the loss-of-coolant flow accident frem high power. Analysis has demonstrated that the specified

_ 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.

The power level trip set point produced by the poyer-to-flow ratio provides i

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 by the power-to-flow ratio provides overpower DNB pro-j tection for all modes of pump operation. For every flow rate there is a maxi-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:

1 1.

Trip would occur when four reactor coolant pumps are operating if power is 108% and reactor flow race is 100%, or flow race is 93% and power l

level is 100%.

2.

Trip-would occur when three reactor coolant pumps are operating if power is 81.0% and reactor flow rate is 74.7% or flow rate is 69% and power 1evel is 75%.

l 3.

Trip would occur when two reactor coolant pumps are operating in a single loop if power is 59% and the operating loop flow rate is 54.5% or flow rate is 43% and power level is 46%.

i 4.

Trip would occur when one reactor coolant pump is operating in each loop (total of two 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 4

the power level trip were used.

The-power-imbalance boundaries are established in order to prevent reactor

-h.

.s1 limits from being exceeded. These thermal limits are either power v

v... l f t i iti: 3 or N:39 11 :.1.1.

"ie rewtor pover imbalance (pover in top halt of core tr.inus power in the botton halt' of core) reduces the power

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1. trip produced by the power-to-flow ratio such that the boundaries of

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The power-te-ficw ratio reduces the m er 16/11/3

. ;- u - en u

2. 3-r - Unit 2 2.3-2 NOV % 5 1974

level, trip and associated reactor power / reactor power-imbalance boundaries by 1.08% - Unit 1 for a 1% flow reduction.

1.07% - Unit 2 1.07% - Unit 3 Pt=p 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 punp(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.

1 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 16/11/3 (1800) psiS (16.25 T

-7756)

(1800) psig (16.25 tut 7756) setpoints shownin Figure 2.3-1A have been established to maintakn the DNB 2.3-1B 2.3-lc ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.(2,3) i Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (13.77 Tout - 6221)

[6/11/3 (16.25 T

-7796)

(16.25 T "t -7796) ou Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting linit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-13 2.3-1C temperatures in the operating range. Due to calibration and instrumentation s er u s, the safety analysis used a trip set point of 620'F.

.. ?:.22:..-

1 Lt :cac tor oeild ng pressure

.p setting limit (4 psig) provides

_..c a.-

rh t : reae:ct L:..p w.ll cetur in :h; uniikely even: of a 1.

--of sa..m: tecident, evan in the absence of a Ic*. reactor coolant system pn2 nre :.ip.

2. 3-3 1

nov 2 s s74

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htfown 3vnpis In order to provide for control rod drive tasts, zero power physics testing, and atartup procedures, thersia provision for bypassing certain segments of the rc2-tor protection system.

The reactor protection system segments which be hypassed are shown in Table 2.3-1A.

Two conditions are i= posed when

. at 2.3-1B 2.3-1C the bypass is caed:

4 1.

3y administratLve control the nuclear overpower trip set point must be reduced to a value 15.0% of rated power during reactor shutdown.

2.

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

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

This high pressure trip set point is lower than the normal low pfessure trip set point so that the reactor cust be tripped before the bypass is. initiated.

The over power trip set point of 15.0% prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would 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. Af ter shutdown has occurred, the following actions will permit operation with one pump in each loop:

16/11/3 1.

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

2.

(Unit 1) Reset the protective system maximum allowable setpoint as

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shown in Figure 2.3-2A2.

B.

Single Loop Operation Single loop operation is permitted only after the reactor has been tripped.

Af ter the pump contact monitor trip has occurred, the following actions vill permit single loop operation:

'ssec the aucp contact mcnitor power level trip setpoint to 55.0%.

two protective channels receiving outlet temperature i c:p o-or t forc a t. n f rom sensors in the Idle Loop.

, '.' n D 11 Roset the protective system maximum allowable setpoints as

', 7 gure 2.1-2A2.

Tripping one at the two protective channels 15/11/3 o

tecperature information f ree the idle loop asstrca

a= t:tn logic of one out af two.

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i 540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTIVE SYSTEM 9AXINUM ALLOWABLE SET POINTS c::LT -

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2.3-3 (ce=rt:+'i OCONEE NUCLEAR STATION Figure 2. 3-11 f 16/11/3 NGV 2 C ;374

Pacer Level, 5 120 FOUR PUNP

- 10 SET POINTS d

0 66 THREE PUMP SET POINTS 40 20

-40

-20 0

20 40 Reactor Power imbalance, 5 PROTECTIVE SYSTEM MAXIMUM

LLOW' ELE SCT PO!NTS

t (fN%

2.3-

.II' !"'c 'f, OCONEE NUCLEAR STATION Td' ngure 2.3-2nj 16/n/

IdOV 2 61974

O Power Level.

120

-- 100 80 O

TWO PUMP SET POINTS

- 40 TWO PUMPS IN ONE LOOP DNE PUMP IN EACH LOOP

-- 20

-40 20 0

20 40 Reactor Power Imualance, 5 PROTECT!VE SYSTEM MAXIMU9 ALLO *AB.E SET POINTS U. NIT 1

'c[ui r

a".

\\ OCONEE NUCLEAR STATION 2.3-82 Figure 2.3-2A2f16/11/3 NOV 2 61974

m Table 2.3-1A Unit 1 Reactor Pretective System Trip Setting Limith Two Reactor One Reactor Four Reactor Three Reactor Coolant Pumps coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in' operating Operasing Single Loop Each Loop

-(gerating Power (Operating Power (Operating Power (Oper.iting Power Shut.w.n th

-100% Rated)

-751 Rated)

-46% Rated)

-49% Rated)

Bg._

l.

.a v.. e :

105.5 105.5 105.5 105.5 5.0(3)

O tr t 1,

a 2.

Nucle.ar I'..r :. Itamed 1.08 times flow 1.03 times flow 1.08 times flow 1.08 times flow Bypawe.J r n F l.m t.

.ac ! nalance, minus reduction minas reduction minus reduction minus reduction (1 kateJ) due to imbalance due to imbalance due to imbalance due to imbalance i

).

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4 4

Pscrmure, ; e. i.;,

(1) T is ir degr. *, Fahrenheit ("t).

(5) Reactor power level trip set point produced by pump contact monitor reset to 55.0%.

(2) lic. set ar Cuviant 'ystem Flow, Z.

i (6) Specification 3.1.8 applies. Trip one of the

( l;

.Itsc htr.tively untrolled reduction set two protection channels receiving outlet temper-

..a : ;

.a v.

s e... *.. : t h..t. lown.

2!

ature information from sensors in the idle loop.

O (4)

Ant.c s t R:. !:,

.ne. othen ser.r.cnts of the k t!..ee ta gs

..l.

cn W

w b

o 1.

Rod index is the percentage su'3 of the withdra:al of the operating groups.

2. Tnese withdrawal limits are effective only for 250 5 full pcwer dgts of cperation after issuance of kner&ents No. 6, 6 and 3, respectively, of Licenses No. CPR-38, -47, and -55 173 209 100 154 213 945 pa,,, t,,c, Cutoff Restricted Region 80 122 80% #

230 75%

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/'%,**

4, 2

i'*/

60 f

as

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Permissible

.(525 P)

$[p Operating Region f

40 20 0

0 50 100 150 200 250 300 Rod index, 5 Witndrawal 25 50 75 100 9

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0 25 50 75 100 Spf i

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i i

0 25 50 75 100 Gp6 j

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i Sp5 CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 4 PDIP OPERATION UNIT 1 3.5-12 t

OCONEE NUCLEAR STAT!CN 16/11/3 Figure 3.5.2-1A1 NOV 2 61974

e r

me O

BLANK PAGE f

l r

3.5-13

'..i i 1974 u

l 1

{

1 Rod index is the percentage su:r. of the witherawal of the opereting (The applicable power level cutoff is 100% power) groups.

2 Pump Withdrawal Limit x N

100 150 275 3

Restricted g

E Region 3

80 4

S 4

6 Permissible p'

Operating 0

N 4,'

2 60 Region 2

~

8ls E

s 3

4

~

40 _

e d

a 20 _

O i

i i

i i

0 50 100 150 200 250 300 Rod index, % withdrawal CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 3 AND 2 PUMP OPERATION UNIT 1

{g'l OCONEE NUCLEAR STATION 3.5-18 s

Figure 3.5.2-2A 16/31/3 NOV 2 61974

l Power, 5 of 2566 M t

__ 110

-20.4

+14.1 1025 100 1

__ 90 1

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85

- 30, 75 70 60 h

50 I

-31.2,52

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f I

f f

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-30

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10 20 30 Core imaalance, ",

OPERATIONAL POT:.R IFEALANCE ENVELOPE UNIT 1

{

a\\, OCONEE NUCLEAR STATION 3.5-21 Figure 3.5.2-3A : 16/11/2 NOV 2 61974 _ __

o e

20

"*1}

Generic FAC jg Results (BAW-10091)

\\ Batch 2 & 3 Fuel 16 L.-

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3 a

14 3

=

=

E 3

12 g

E 10 0

2 4

6 8

10 12 Distance from i.11st, ft LOCA LIMITED MAXD EI ALL0k?A3LE LINEAR HEAT RATE

',:" t *' t OCONEE NUCLEAR STATION

3. ;.

2,,

sy,

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Figure 3.5.2-4 l16/11/3 i

NGV 2 61974

3.11' E\\XIMUM POWER RESTRICTION Applicability Applies to the nuclear steam supply system of Units 2 and 3 reactors.

Objective To maintain core life margin in reserve until the s'ystem has perforned under operating conditions and design objectives for a significant period of time.

Soeci fica tion l16/11/3 1.11.1 The first reactor core in Unit 2 may not be operated beyond 11,040 effective full power hours until supporting analyst-and data pertinent to fuel clad collapse under fuel deasif i-cation conditions have oeen approved by the Directorate of Licensing.

3.11.2 The first reactor core in Unit 3 may not be operated beyond 10,9'+4 effective full power hours until supporting analysis and o'ata pertinent to fuel clad collapse e-der fuel densifi-cati;n conditions have been approved by the Directorate of Licensing.

Bases The licensing staff has reviewed the effects of fuel densification for the first core in Oconee Units 2 and 3 and concluded that clad collapse will nor 16/11/3 take place within the first fuel cycle (11,040 effective full power hours for Unit 3 and 10,944 effective full power hours for Unit 3).

However, the clad collapse model used is questionable for extrapolation of clad collapse time out beyond the first fuel cycle because of limited experi-mental verification.

3.11-1 NOV E c 1974

3.5.2 Contrel Rod Grrup rnd Pov r Distribution Limit 3, I

Applicability This specification applies to power distribution and operation of control l

rods during power operation.

Obiective To assure an acceptable core pcwer distributioq during power operation, to set a limit on potential reactivity insertion ) rom a hypothetical control rod ejection, and to assure core suberiticality af ter a reactrr trip.

Specification I

The available shutdown margin shall be not less than II Ak/k with 3.5.2.1 the highest worth control rod fully withdrawn.

3.5.2.2 Operation with inoperable rods:

7 If a control rod is misaligned with its group average by more a.

than an indicsted nine (9) inches, the rod shall be declared inoperable. Tim rod with the greatest misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for determining the operability of rods with lesser misalignments.

(

b.

If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.

If a control rod cannot meet the requirements of Specification c.

4.7.1, the rod shall be declared inoperable.

d.

If a control rod is found to be improperly programmed per Specification 4.7.2. the rod shall be declared inoperable until properly programmed.

Operation with more than one inoperable rod in the safety or e.

regulating rod groups shall not be permitted.

f.

If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position, an evaluation shall be initiated immediately to verify the existance of 17.

Ak/k hot shutdown margin.

Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are fully withdrawn, whichever occurs first.

i Simultaneously, a program of exercising the remaining regulating and saf ety rods shall be initiated to verify operability.

3.5-6 3,;. 2 6 1974

,, - - ~

~

If within one (1) hour of ~ determination of an inoperable rod, g.. it is not determined that a 1%Ak/k hot shutdown margin ex'ists combining the worth of the inoperable rod with each of the other

(.

rods, the reactor shall be brought to the hot standby condition until this margin is established.

h.

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

1.

If a control rod in the regulating or safety rod groups is declared inoperable, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump com-bination.

j. If a control rod in the regulating or axial power shaping groups is declared inoperable, operation above 60 percent of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained 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 worth of a single inserted control rod shall not exceed 0.5%

,Ak/k at rated power or 1.0% Ak/k at hot zero power except for physics testing when the rr,quirements of Specification 3.1.9 shall apply.

3.5.2.4 Quadrant Power Tilt

(

Whenever the quadrant power tilt exceeds 4 percent, except for a.

physics tests, the quadrant tilt shall be reduced to less than 4 percent within two hours or the following actions shall be

)

taken:

' (1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each 1 percent tilt in excess of 4 percent below the power 1

16/11/3 level cutoff (see Figures 3.5.2-1A1, 3.5 0 1B1, 3.5.2-1B2, 3.5.2-1B3, 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-Ic3).

(2) If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each 1 percent tilt below the power allowable for the reactor cool. ant pump combination as defined 'uy 2

Specification 2.3.

(3) Except as provided in 3.5.2.4.b, sthe reactor shall be brought to the hot shutdown condition within four hours if the quadrant tilt is not reduced to less than 4 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. If the quadrant tilt axceeds 4 percent and there is simultaneous indication of a misaligned ' control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60 percent of the thermal power ~ allowable for the reactor coolant 3.5-7 NOV 2 61974

- - - -,,, ~. -

pump combination.

Except for physics tests, if quadrant tilt exceeds 9 percent, a c.

controlled shutdown shall be initiated immedintely and the reactor shall be brought to the hot shutdown condition within four hours.

Whenever the reactor is brought to hot shutdown pursuant to d.

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 thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each I per-cent tilt for the maximum tilt observed prior to shutdown.

Quadrant power tilt shall be monitored on a minimum f'requency of e.

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 (safety rod kithdrawal) does not prohibit a.

the exercising of individual safety rods as required by Table 4.1-2 or app 1v to inoperable safety rod limits in Technical Specification 3.5.2.2.

b'.

Operating rod group overlap shall be 25% + 5% between two sequential groups, except for physics tests.

(

c.

Except for physics tests or exercising control rods, the control rod with-drawal limits are specified on Figures 3.5.2-1A1 (Unit 1),

16/11/*,

3.5.2-1B1, 3.5.2-132 and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump operation and on Figures 3.5.2-2A (Unit 1), 3.5.2-23 (Unit 2), and 3.5.2-2C (Unit 3) for three or two pump operation. If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall then be attained within two hours,

d. Except for physics tests, power shall not be increased above the power lgvel cutoff as shown on Figure 3.5.2-1A1 (Unit 1), 3.5.2-131, l 16/11/:

3.5.2-1B2, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 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 steady-state rated power.

/

3.5-8 NWi 4 61974 o

.1

3.5.2.6 Reactor power imbalance shall be monitored on a frgquincy not to excee'd two ' hours during power operation above 40 p-ecnt rated power.

[

within the Except for physics tests, imbalance shall be maint.

envelope defined by Figures 3.5.2-3A, 3.5.2-3B, and. '.2-3C.

If 16/11/3 the imbalance is not within the envelope defined by Figure 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C, corrective measures shall be taken to achieve an acceptable imbalance.

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

1 1.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.

16/11/3 Y

~~

A

(

(

3.5-9 NGv 2 s 1974

---mw-e

.v--

w-

Bases The power-imbalance envelope defined in Figures 3.5.2-3A, 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 vill 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 situation that would cause the Final acceptance criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while si6ultaneously all other engineering and uncertainty factors are also at their limit's.**

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% 15% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and 1.ower -part of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function 1

Safety 2

Safety 3

Safety

(

4 Safety 5

Regulating 6

Regulating 7

Xenon transient override 8

APSR (axial power shaping bank)

The minimum available rod. worth provides for achieving hot shutdown by reactor trip at any time assuming the highest worth control rod remains in the full out position.(1)

Inserted rod groups during power operation will not contain single rod worths greater than0.5% ak/k.

This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident.(2) A single inserted control rod worth of'1.0% ok/k at beginning of life, hot, zero power would result in the same transient peak thermal power and, therefore, the same environmental consequences as a 0.5% ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with Group 1.

Groups 5, 6, and 7 are overlapped 25 percent.

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

• • Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors.

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

3.5-10 f.;V 2C 371

l Th2 quadrant power tilt limits set forth in Sp3cificatica 3.5.2.4 hsva be:n catablished within 'the thermal analysis design base using the definition of These quadrant power tilt given in Technical Specifications, Section 1.6.

limits in conjunction with the control rod position limits in Specification 3.5.2.5c ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.

The quadrant tilt-and axial imbalance monitoring in Specifications 3.5.2.4 1 5.2.6, respectively, normally will be performed in the process computer, t two-huur frequency for monitoring these quantities will provide adequate

i. -

uiveillance when the computer is out of service.

i

'llowance is provided for withdrawal limits and reactor power imbalance limits be exceeded for a period of two hours without specification violation.

l Sui *ptance rod positions and imbalance must be achieved within the two-hour eo re Lod or appropriate action such as a reduction of power taken.

'8a-4 16/11/3 Opiu it ing restrictions are included in Technical Specification 3.5.2.5d i.o The xenon reactivity prevent excessive power peaking by transient xenon.

must be beyond the "undershoot" region and asymptotically approaching its squilibrium value at rated power.

('

REFERENCES 1Section 3.2.2.1.2 2Section 14.2.2.2 e

J 3.5-11 NOV 2 61974

UNITED STATES g.

ATOMIC ENERGY COMMISSION

' ". l i

,a I

• WASHINGTON. D.C. 20545 i

%, ei DUKE POWER COMPANY _

DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE j

Amendment Na. 6 License No. DPR-47 1.

The Atomic Energy Connaission (the Commission) having found that:

A.

The application for amandment by Duke Power Company (the licensee) dated September 20, 1974, as supplemented October 8 and 31, 1974, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, 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 :.s reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Conunission's regulations; D.

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

Prior public notice of this amendment is not required since the amendment does not involve a significant hazards consideration.

2.

Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license acendment and Paragraph 3.B. of Facility License No. DPR-47 is hereby amended to read as follows:

"B.

Technical Specifications 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. 11."

3.

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

FOR THE ATOMIC ENERGY COMfISSION t

Karl R. Goller, Assistant Director

~

for Operating Reactors Directorate of Licensing

Attachment:

Change No. 11 to Tech M c=1 Specifications Date of Issuance: November 26, 1974

~

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 6 TO FACILITY LICENSE NO. DPR-38, CHANGE NO.16 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 6 TO FACILITY LICENSE NO. DPR-47, CHANGE NO. 11 TO TECHNICAL SPECIFICATIONS; AMENDMENT LO. 3 TO FACILITY LICENSE NO. DPR-55, CHANGE NO. 3 TO TECHNICAL SPECIFICATIONS; DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287 Ravisa Appendix A as follows:

Remove Pages, Insert New Pages 2.1-1 & 2.1-2 2.1-1 & 2.1-2 2.1-3 2.1-3, 2.1-3a, 2.1-3b &

2.1-4 2.1-4 2.1-4a 2.1-7 2.1-7 2.1-10 2.1-10 2.3-1 & 2.k2 2.3-1 & 2.3-2 2.3-3 & 2.3-4 2.3-3 & 2.3-4 2.3-5 2.3-5 2.3-8 2.3-8 & 2.3-8a 2.3-11 2.3-11 3.5-12 3.5-12 3.5-13 3.5-13 Blank page

'3.5-18 3.5-18 3.5-21 3.5-21

O

. Remove Pages Insert New Pages

, 3.5-24 3.5-24 3.11-1 3.11-1 3.5-6 & 3.5-7 3.5-6 & 3.5-7 3.5-8 & 3.5-9 3.5-8 & 3.5-9 3.5-10 & 3.5-11 3.5-10 & 3.5-11

/

f e

I I

l

+

can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.32.

A DNBR of 1.32 corresponds to a 95 percent probability at a 99 percent confidance level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.

Tht, difference in these two-;reasures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.32 is predicted for the maximum fossible thermal power coolant flow is 107.6 percent of 131.3 x 10gs are operating (minimum reactor (112 percent) when four reactor coolant pum lbs/hr.).* This curve is based on the combination of nuclear power peaking factors, with potential fuel densifi-cation effects, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:

h C

1.

The 1.32 DNBR 1imit produced by the combination of the radial peak, axial u

peak and position of the axial peak that yields no less than a 1.32 DNER.

2.

The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.15 kw/f t for Unit 1.

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

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

The urve of Figure 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A (because the four-pump pressure - temperature restriction is known to be more It ! ting than the 3 and 2 pump combinations, only the four pump limit has i'en shown on Figure 2.1-3A).

n. iram t..erra t poser for three pt.:ap operat t au i., M pc.rcer.t due tc.;

p".ier level trip produced by the flux-flow ratio 75 percent flow x 1.03 =

o ene p ar, plus the mav.Lnta cal thr.ttien aad instrument error. The

c. W 'num therma l power for other coolant pdep conditions are pr0duc ci in a s i 2. I i r m.mner.

j 2.1-2 NOV 2 61974

-l Fot-Figure 2.1-3A, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than 1.32.

The 1.32 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 left of the other curves.

16/11/'

References _,

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

(2) Oconee 1, Cycle 2 - Reload Report - BAW-1409, Sepetmeber, 1974.

4-J 2.1-3 NOV 2 61974

O

, Pase s - Units 2 and 3 To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough 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 tesperatures and the possibility of cladding failure.

Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the W-3 correlation.(1)

The W-3 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux disgributions. The local DNB ratio (DNBR), defined as the ratio of the he flux that would cause DNB at a particular core location to the actual heat i

, is.; indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.3.

A DNBR of 1.3 corresponds to a 94.3 percent probability at a 99 percent coulidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor cuolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

15/J' The curve presented in Figure 2.1-1B represents the conditions at which a 1

2.1-1C minimum DNER of 1.3 is predicted for the maximum possible thermal pvaer (112%)

when four reactor coolant pumps are operating (minimum reactor coolant flow is 131.3 x 106 lbs/hr). This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:

1.78 ;F

= 1.50 2.67; F F

=

=

9 AH The design peaking combination results in a more conservative DNBR than any other snape that exists during normal operation.

The curve of Figure 2.1-28 are based on the more restrictive of two thermal 16/?l/

2.1-2C 3

limits and include the ef'ects of potential fuel densification:

N

4 1.1 IrmR limit produced by a nuclear power peaking factor of F = 2.67 q

the cu:.b u...t.lan of the radial peak, axial peak and position of the a>.ial peak that yields no less than 1.3 DNBR.

2.

The combination of radial and axial peak that causes central fuel melting l15/11/,

at the hot spot.

The limit is 19.8 kw/ft - Unit 2 19.8 kw/ft - Unit 3 3

2.1-3a i

NOV 2 6 B74

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

13/'l The specified flow rates for Curves 1, 2, 3, and 4 of Figure 2.1-2B correspond 3

2.1-2C to the expected minimum flow rates with four pumps. three pumps, one pump in each loop arid two pumps in one loop, respectively.

16/11/

The curve oi Figure 2.1-1B is the most restrictive of all possible reactor 3

2.1-1C coolant pump-maximum thermal power combinations shown in Figure 2.1-38.

2.1-3C 16/ 11/

The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR 3

2.1-3C of 1.3 is predicted at the maximum possible thermal power for the number of i

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.

Using a local quality limit of 15 percent at the point of minimum DNBR as a 16/11 basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even 3

2.1-3C though the quality of the exit is higher than the quality at the point of minimum DNBR.

4 The DNBR as calculated by the W-3 correlation continually increases from point of minimum DNBR, so that the exit DRBR is 1.7 or higher, depending on the pressure. Extrapolation of the W-3 correlation beyond its published quality i

range of +15 percent. is justified on the basis of experimental data.(4)

The maximum thermal pw for three pump operation is 86% - Unit 2 15/11 86% - Unit 3 3

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

1.07 = 80%

~

power plus the maximum calibratica and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

l 16/11 For each curve of Figure 2.1-38, a pressure-temperature point above ar.d to the 3

2.1-3C le'* of the curve would result in a DNBR greater than 1.3 or a local quslicy

}-

at one 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 /

emperature point above and to the lef t of the four-punp curve will Se.f.,ove

,i. the left of the Ah~r curves.

1[FyENCES (1) FS AR, S ec t ion - 3. 2.3.1.1 (2) FSAR, Sec tion 3. 2.3.1..l.c 5

(3) FSAR, Section 3.2.3.1.1.k 2.1-3b NOV 2 6 374

.L

(4) The following papers which were presented at the Winter Annual Meeting, ASME, November 18, 1969, during the "Two-phase Flow and Heat Transfer in Rod Bundles Symposium:"

(a) Wilson, et al.

" Critical Heat Flux in Non-Unic 'rm Heater Rod Bundles" (b).Cellerstedt, et al.

" Correlation of a Critical Heat Flux in a Bundle Cooled by Pressurized Water" i

1 2.1-4 NOV 2 61974

2500 2400 i

.5 2300 m

c.

a

_m 1

M.

I 2200

-=a 5

2100 2000 I

i 1900 560 580 600 620 640 660 Reactor Outlet Temperature, F

~cr v ~~cG~ inn :': T4.

'**3

..:r 7,f 'N(\\.? CCOt&E NUCL: R STATIO.N

-? v 3

!igure

.L-;,

l16/11/3 NOV 2 61974

Thermal Power Level, 5

- 120

--100 I

2 80 z

':60 (3AND4}

/

40 20

-40

-20 0

+20

+40 Reactor Power imoalance, s CURVE REACTOR COOLANT FLOW (LB/5iR) 6 1

131.3 x 10 6

2 99.1 x 10 6

3 S4.4 x 10 8

4 S0.1 x 10 CORE PROTECTION SAFETY LIMITS c::T :

\\

-*1 y.it as ao OCONEE NUCLEAR STATION 3

sg NOV 2 81974 ngun 2.1-2A flu 11/3

2500 2400 e

.~d 2300 E

t G

U r

2200 5

Sa 2l00 2000

/

1900 560 580 600 620 G40 660 Reactor Outlet Temperature. F CORE PROTECTION SAFETY LIMITS

. n.

lb $%

- 1

,tv yne rcat a, OCONEE NUCLEAR STATION sg.

Figure 2.1-3!.

f16/11/3 20y 2 e 1974

2.3 LIMITING SAFETY. SYSTEM SETTINGS, PROTECTIVE INSTRUMENTAIIGN Applicability Appites to instrumenen monitoring reactor power, reactor pcaec 2 ca.ca reactor coolant system pressure, reactor coolant outlet tempct<<.tre, r;.,,

aumber of pumps in operation, and high reactor building pressure.

Objective To provide automatic protective action to prevent any combination uC pracess variables f rom exceeding a safety limit.

Specification

'ihe reactor protective system trip setting limits and the parmissible ty p..shes for the instrument channels shall be as statedin Table 2.3-Li - Unit 1 and 2.3-1B - Unit 2 2.3-1C - Unit 3 16/11/3 Figure 2.3-2A1 } Unit 1 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:

a.

Loss of two pumps and reactor power level is greater than SS: (0.0% for h6/11/3 Unit 1) of rated 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 55% of rated power for single loop operation. Power /RC pump ctip setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)

L6/11/-

c.

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

Bases

'1he reactor protective system consists of four instrument channels to s.cnitor each of several selected plant conditions which will enuse a reseror trse if any one of these conditions deviates from a pre-selected operactug r.s e co the legcee that a safety limit may be reached.

l th.t t. rip eetting liwit= f ar protective system instrumentatii-n.c a'abic 2.3 1A - Unit 1.

The safety analysis has been based L.;u; t 1

2.3 Unit 2 2.1 6..: t c 1

.. e 1 f r t em..e at..n e c!p set points plus enlibration "u

s L,e i

-... - - w r.~..... - :-

..c

.i 9.... tor trip it ' tan power level (neutron flux) is providet

.L. a ge T i. the p:4L c'. add.c.g from reactivity excursions tea r..

j

et essucc....e re yorot ut e measurc=ents.

t 2.3-1 l

NOV 161974

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.

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

The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power 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 pu=p operation.

For every flow rate there is a maxi-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:

1.

Trip would occur when four reactor coolant pumps are operating if power is 108% and reactor flow rate is 100%, or flow race is 93% and power level is 100%.

2.

Trip would occur when three reactor coolant pumps are operating if power is 81.0% and reactor flow rate is 74.7% or flow rate is 69% and power level is 75%.

3.

Trip would occur when two reactor coolant pumps are operating in a single loop if power is 59% and the operating loop flow rate is 54.5% or flow rate is 43% and power level is 46%.

4.

Trip would occur when one reactor coolant pump is operating in each loop (total of two 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 the power level trip were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits fro = being exceeded.

These ther=al limits are either power r e J 'r - 1. / ? t licl*

" M;M 11c' t<.

% r..acter power imbalance (power ir 3

tnc top half of core Lnus pawer in the bctcan half of core) reduces the pcwer Icvel trip produced by :he pcwer-to-flew ratio such that the boundaries of Figure i

't'

re produced.

The pe er-to-flow ratio reduces the pcwer 16711/3

2. 3-2r.' 4ni -

2. 5-2n - Unit j

2.i-2c - Unit 3 2.3-2 HOV16 1974

y_

IcveL trip,and associated reactor power / reactor power-imbalance boundaries by 1.08% - Unit 1 for a 1% flow reduction.

1.07% - Unit 2 1.07* - Unit 3 Pumo Monitors The pump monitors prevent the mintzum 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 honitors also restrict the power level for the number of pumps in operation.

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 pressurit (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) psis and variable low pressure (13.77 Tout-618D trip 16/11/3 (1800) psig (16.25 T

-7756)

(1800) psig (16.25T$-7756) setpoints shown in Figure 2.3-1A have been established to maintain the DNB 2.3-1B 2.3-lc ratio greater than or equal to 1.3 for those design accidents

• hat result in a pressure reduction.(2,3)

Due to the calibration and instrumentation errors the safety analysis used a bNI11/3 variable low reactor coolant system pressure trip value of (13.77 Tout -6221)

(16.25 T

-7796)

(16.25T$*-7796)

Coolant Outlet Temperature The nigh reacter coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to pre >ent excessive core coolant 2.3-15 2.3-IC te=peratures in the operating range. Due to calibration and instrumentation the safety analysis used a trip set point of 620*F.

vercre, Pre w:rt "d 4 re r

- ouilding pecsaurb t rip setting limit (4 psig) provides tiut a reactor...y will accur i.. the '_u'1kely :eant of a

.,: vf-. u ' e. t accident, even in LLe absence of a'lcw reactor cociant syste=

> rc.c su re t ;..

2.3-3 NOV 2 6 574 I

Shu p

,3yoty g

In order to provide for control rod drive tests, zero power physics testing, and startup procedures, therais provicion for bypassing certain segments of the reactor protection system. The reactor protection system segments which ctn be bypassed are shown in Table 2.3-1A.

Two conditions are imposed when 2.3-13 2.3-lc the 5 pass is uued:

/

1.

3y administrative control the nuclear overpower trip set point must be reduced to a value 1 5.0% of rated power during reactor shutdown.

2.

A hish reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed. This high pres sure trip net point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The over power trip set point of 15.0% prevents any significant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolane pumps were operating.

Two Puep Goeration A.

Two Loop Operation i

Operation with one pump in each loop will be allowed only following reactor shutdown. Af ter shutdown has occurred, the following actions will permit operation with one pump in each loop:

16/11/3 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 Operation Single loop operation is permitted only af ter the reactor has been tripped.

After the pump contact monitor trip has occurred, the following actions ulli permit afngle loop operation:

2eset the pump contact monitor power level trip setpoint to 55.0%.

i<ip cm of the two protective channels receiving outlet temperature D ' tr

.t ton f ro.s sensors in the Idle Loop.

i.

(Unit li Resec the protective system maximu= allowable setpoints as J.~.r t

7tcure 2.3-2A2.

Tripping one of the two protective channels 15/11/3 mt t"t temperature information frem the idle loop assures ite v.a cn t:1p logic of one cut of two.

t' E '-

l00

.1.2.6 i

s s

. i., r i

'.,i,7,j 2.3-4 NOV 2 6 G74

2400 2300 2

E.

J 2200 a

2 m

e l'

2!00 8

u a

O 2000 1900 I

i i

i 540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS c:::T 1

') OCONEE NUCLEAR STATION 2.3-5 (2

i Figure 2.3-1Af16/Ill3 NOV 2 6 '974

O Pacer Level, 5 120 FOUR PUNP 10 SET POINTS 0

60 THREE PUMP SET POINTS 40 20 4

a 40

-20 0

20 40 Reactor Power imbalance, ",

PROTECTIVE SY STEM M AXIMUM ALLO 913LE SET POINTS CIT 1 l6

.' OCONEE Ni) CLEAR STATION 2.3-3 rita' ",

l 16 /.11/ '

Figure 2.3-2Alj NOV % 61974

Power Level, 1

- 120 0

-- 100 80 i

60 TWO PUNP SET POINTS

-. 40 TWO PUMPS IN ONE LOOP DNE PUNP IN EACH LOOP

-- 20

-40

-20 0

20 40 Reactor Power leoalance. '-

?;GTECTIVE SYSTEM MAXIMUM ALLOWASLE SET POINTS UNIT 1 8'[4cah.\\

. OCONEE NUCLEAR STATIO*.

' 3-52 Figure 2.3-2A2 f 16/11/3 NOV 2 61974

~--

4 T.sble 2. 3-I A L'n t t 1 Reactor Protective System Trip Setting I.imits s

Two Reactor One Reactor Four Reactor Three Reactor-Coolant Pumps coolant Pump Cool. int Pumps Coolant Pumps Operating in A Operating in Operating Operating Single Loop Each 1.oop (Operating Power (Operating Power (Operating Power (Operating Iower Shutdown 1'

,e

__100% Rated)

-752 kated)'

-46% Rated)

-492 Rated)

Bypau 1.

..e :.m ID 105.5 105.5 105.5 105.5 5.0

( tra t. i..

2.

.%. l e i r 1. r ::..

ha t.ed 1.03 times flow 1.08 times finw 1.08 times flow 1.08 times flow _

byp w. J -

e n I ! 6.w t c, an.t

.milance, minus reductio.:

minus reductions minus reduction minus reduction C. utrd; due to imbalance due to imbalance due to imbalance due to imbalance 1.

Nuclear 1. e tt i ta :.at NA NA 552 (5)(6) 551 (5)

Bypassed un I uiup M stos.

. ?, kated) 4.

su ch a rac e.r t.u,1 c.t 2355 2355 2355 2355 17200)

w......,. @, m.

s_ 4 -

!~w c.-

t wi.

y 1935 1985 1985 1985 Bypa.esed p

.t...

s.

.n e.,.. 3,, Hin.

[6/,

6 t e r,.n l. '-

s (13.77 T

-6181)II) (13.77 T

- 6181)III (13.77 7

- 6181)II) (13.77 7 - 6181)II)

Bypaned "E

a.. c, t at.t. 4 r-Ir e.sure p s t e:. < n..

7.

Peactor 'i.lant T.np.

619 619 619 (6).

619 619 f..

Mas.

8.

High Re..ct...

8:ul. :iur.

4 4

4 4

4 Picssus....

.,t, (1) r"' is in.try,s.. 11:irenheit ("F).

(5) Reactor power level trip set point produced by pump contact monitor reset to 55.02.

(2)

r..s e t o n. i.. i..n t :yseem ilow, Z.

(6) Specification 3.1.8 applies. Trip one of tf-e (3)

.1:aanlati : Ivel.

ontiolled reduction set

.' nit;.. r:

two protection channels receiving outlet te.jer-se. r t heit.1:2wn.

ature information from sensurs in the idle loop.

O<

(4) Au t < >Li t s.... ? J >

.s when other sesn.onts of tbe RPS.seu hygia..ad.

g cn W

w Sam 1-t

1.

Rod indes is the percentage su2 of tite eithdra:al of the operating groups.

2. These withdrawal limits are effective only for 25015 full pcwer days of operation after issuance of ke.tT.ents No. 6, 6 and 3, respectively, of Licenses No. DPR-38, 47, aa.d -55 d

173 209 100

/

54 213 94%

Power Level Cutoff Restricted Region 230 80 122 80% 4 75%

h 60

<k

[

Permissinle

.(52% P) 3 y

Operating 40 20 0

O 50 100 150 200 250 300 Rod index, S Witndrawal 25 50 75 100 E

i e

i i

0 25 50 75 100 6p7 1

1 1

1 1

0 25 50 75 100 Gp6 t

1 1

1 1

Gp5 CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 4 PUMP OPERATION UNIT 1 i) OCONEE NUCLEAR STATION 3.5-12 16/11/3 Figure 3.5.2-1A1 NOV 2 61974

e e

e e

BLANK PAGE I

i l

3.5-13

.g.,

g'- J t 4Q lej n

, 1.

Rod index is the percentage sum of the withdrawal of the operating groups. (The applicable power level cutoff is 100% power) 2 Pump Withdrawal Limit \\

100 150 275 Restricted j

.E Region 3

W 9

u 80 g

i d

E D

Permissible

• /

o

.[

Operating k,'

2 8'EIOR 60 2

~

a Ea 2=

40 j

J E

E 20 i

0 i

e i

e i

0 50 100 150 200 250 300 Rod index. 5 withdrawal CONTROL ROD CROUP WITHDRAWAL LIMITS F02 3 AND 2 PDT OPERATION UNIT 1 xg; OCONEE NUCLEAR STATION a

3.5-18 Figure 3.5 2-2A 16/11/3 l

NOV 2 61974

0 Power, ",of 2568 ilft 110

-20.4

+14.1 1025

__ 100 90

__ 80

- 30, 7 5 70 60 50 I

I

-31.2,52

+28.1, 52

/

/

40 Ii i

i i

i i

-30

-20 10 0

10 20 30 Core imaalance, ",

OPERATIONAL POWER DGALANCE ENVELOPE l

UNIT 1 l ocONEE NUCLEAR STATION 3.5-21 q

Figure 3.5.2-3A.16/11/3 NOV 2 S 1974 _._.

O 4

20 Batch 4 Fuel]

Generic FAC 18 Results (BAW-10091)

\\ Batch 2.& 3 Fuel J

a:

14 E

li; I

3 12 g

E 10 0

2 4

6 8

10 12 Distance from inlet, ft LOCA LIMITED MAXIMLB ALLOWABLE LINEAR HEAT RATE (b\\

G, OCONEE NUCLEAR STATION pt Na+4, 3.5-14 l'N11/ 3 Figure 3.5.2-4 NOV 2 6 i974

O 3.11 MAXIMUM POWER RESTRICTION Aeplicability Applies to the nuclear steam supply system of Units 2 and 3 reactors.

Objective To maintain core life margin in reserva until the system has performed under operating conditions and design objectives for a significant period of time.

Speci fica tion

\\16/11/3 3.11.1 Tha first reactor core in Unit 2 may not be operated beycad 11,040 effective full power hours until supporting analys!4 and data pertinent to fuel clad collapse under fuel densii t-cation conditions have been approved by the Directorate of Licensing.

3.11.2 The first reactor core in Unit 3 cay not be operated beyond 10,944 effective full power hours until supporting analysis and data pertinent to fuel clad collapse under fuel densifi-cation conditions have been approved by the Directorate of Licensing.

Bases The licensing staff has reviewed the effects of fuel densification for the 16/11/3 first core in Oconee Units 2 and 3 and concluded that clad collapse will not

.take place within the first fuel cycle (11,040 ef fective full power hours for Unit 3 and 10,944 affective full power hours for Unit 3).

However, the clad collapse model used is questionable for extrapolation of clad collapse time out beyond the first fuel cycle because of limited experi-mental verification.

NOV % 6 IS74 3.11-1

i 3.5.2 contral Rod Group and Powir Distributien Liaitn Apolicability This specification applies to power distribution and operation of control rods during power operation.

Obiective To assure an acceptable core power distributionf during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip.

Specification 3.5.2.1 The available shutdown margin shall be not less than 1% Ak/k with the highest worth control rod fully withdrawn.

3.5.2.2 Operation with inoperable rods:

7 a.

If a control rod is misaligned with its group average by more than an indicated nine (9) inches, the rod shall be declared i

inoperable. The rod with the greatese misalignment shall be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing i

the average position of the group for determining the i

operability of rods with lesser misalignments.

i

/

i 1

n j

i b.

If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out 4

limit lights, the rod shall be declared to be inoperable.

c.

If a control rod cannot meet the requirements of Specification 4.7.1, the rod shall be declared inoperable.

d.

If a control rod is found to be improperly programmed per Specification 4.7.2, the rod shall be declared inoperable until

]

properly programmed.

4

)

e.

Operation with more than one incperable rod in the safety or regulating rod groups shall not be permitted.

f.

If a control rod in the regulating or safety rod groups is i

declared inoperable in the withdrawn position, an evaluation shall be initiated immediately to verify the existance of 1%

i ak/k hot shutdown margin. Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are fully withdrawn, whichever occurs first.

Simultaneously, a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.

4 3.5-6 ga, ed B74 9

v

,,..a

--w

-.,--..,-n.,_.,,..--.,.,n------an

g.

If within one (1) hour of' determination of an inoperable rod, it 'is not determined that a 1%4k/k hot shutdown margin ex'ista combining the worth of the inoperable rod with each of the other

(,

rods, the reactor shall be brought to the hot standby condition until this margin is established.

h.

Following the determination of an inoperable rod, all rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until 'the* rod problem is solved.

1.

If a control rod in the regulating or safety rod groups is declared inoperable, power shall be reduced to 60 percent of i

the thermal power allowable for the reactor coolant pump com-bination.

j. If a control rod in the regulating or axial power shaping groups is declared inoperable, operation above 60 percent of rated power may continue provided the rods in the group are positioned such that the rod that was declared intiperable is maintained within allowable group average position limits of Specification 3.5.2.2.s and the withdrawal limits of Specification 3.5.2.5.c.

3.5.2.3 The worth of a single inserted control rod shall not exceed 0.5%

, Ak/k at rated power or 1.0% ak/k at hot zero power except for physies testing when the requirements of Specification 3.1.9 shall apply.

3.5.2.4 Quadrant Power Tilt s.

Whenever the quadrant power tilt exceeds 4 percent, except for physics tests, the quadrant tilt shall be reduced to less than 4 percent within two hours or the following actions shall be 2

taken:

i

~ (1) If four reactor coolant pumps are in operation,, t'ae allowable thermal power shall be reduced by 2 percast of 6hl power for each 1 percent tilt in excess of 4 percent,'131, below the power level cutoff (see Figures 3.5.2-1A1, 3.5.6-16/11/3 3.5.2-132, 3.5.2-133, 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3).

i (2) If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each 1 percent tilt below the power allowable for the reactor coolant pump combination as defined by Specification 2.3.

(3) Except as provided in 3.5.2.4.b, sche reactor shall be brought to the hot shutdown condition within four hours if the quadrant tilt is not reduced to icss than 4 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. If the quadrant tilt exceeds 4 percent and there is simultaneous indication of a misaligned control red per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60 percent of the thermal power' allowable for the reactor coolant 3.5-7 NOV % 61974

=

O pump combination.

Except for physics tests, if quadrant tilt exceeds 9 percent, a c.

controlled shutdown shall be initiated bamediately 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 abovw, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each 1 per-cent tilt for the maximum tilt observed prior to shutdown.

Quadrant power tilt shall be monitored on a minimum frequency of e.

once every two hours during power operation above 15 percent of rated power.

3.5.2.5 Control Rod Positions i

Technical Specification 3.1.3.5 (safety rod withdrasal) does not prohibit s.

the exercising of individual safety rods as required by Table 4.1-2 or i

apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

b'.

Operating rod group overlap shall be' 25% + 5% between two sequential groups, except for physics tests.

drawal limits are specified on Figures 3.5.2-1A1 (Unit 1),

16/11

(

c.

Except for physics tests or exercising control rods, the control rod with-3.5.2-1B1, 3.5.2-1B2 and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump operation and on Figures 3.5.2-2A f

(Unit 1), 3.5.2-23 (Unit 2), and 3.5.2-2C (Unit 3) for three or two pump operation. If the control rod position limits are exceeded, corrective seasures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall then be attained within two hours..

d. Except for physics tests, power shall not be iacreased above the power igvel cutoff as shown on Figure 3.5.2-1A1 (Unit 1), 3.5.2-1B1, l 16/11/~.

3.5.2-1B2, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 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 xe.non reactivity shall be asymptotically approaching the value for operation at steady-state rated power.

i 3.5-8 NON iS 674 o

3.5.2.6 React:r power imbnlanco chall b3 monitsrcd en a frcquency not to exceed two hours during power operation above 40 percent rated power.

Except for physics tests, imbalance shall be maintained within the

(

envelope defined by Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C.

If 16/11/3 the imbalance is not within the envelope defined by Figure 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C, corrective measures shall be taken to achieve an acceptable imbalance.

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

i 3.5.2.7 The control rod drive patch panels shall be lockdd at all times with limited access to be authorized by the superintendent.

16/11/3 3

(

i

(

3.5-9 h0V 26 :974

3 Ba-c7 The power-imbalance envelope defined in Figures 3.5.2-3A, 3.5.2-33, and 3.5.2-3C is g

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 situation that would cause the Final acceptance criteria to be approached should a LOCA occur is highly improbable because all of the power distribution 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 i

by application of:

a.

Nuclear uncertainty factors i

b.

Thennal calibration c.

Fuel densification effects d.

Hot rod manufacturing tolerance factors The 25% i 5% overlap between successive control rod groups is allowed since the a

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 1

Safety 2

Safety 3

Safety k.

4 Safety 5

Regulating 6

Regulating 7

Xenon transient override 8

APSR (axial power shaping bank)

The minimum available rod worth provides for achieving hot shutdown by reactor crip at any time assuming the highest worth control rod remains in the full out position. (1)

Inserted rod groups during power operation will not contain single rod worths greater than0.5% ak/k. This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident.(2) A single inserted control rod worth of 1.0% ok/k at beginning of life, hot, zero power would result in the same transient peak thermal power and, therefore, the same environmental consequences as a 0.5% ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with Group 1.

Groups 5, 6, and 7 are overlapped 25 percent. The normal position at power is for Groups 6 and 7 to be partially inserted.

i

• • Actual operating linits depend on whether or not incore or excore detectors.

are used and their respective instrument and calibration errors.

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

3.5-10 p;.yi 2.

137d

The quadrant power tilt limito cce forth in Sp::cificction 3.5.2.4 hnva bsc:a cotablished within the thermal analysis design base using the definition of i

quadrant power tilt given in Technical Specifications, Section 1.6.

These limits in conjunction with the control rod position limits in Specificar. ion 3.5.2.5c ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.

The iguadrant tilt and axial imbalance monitoring in Specifications 3.5.2.4 1 %.2.6, respectively, normally will be perforued in the process computer, f

s w hour f requency for monitoring these quantities will pro ".e adequate er

. itveillance when the computer is out of service.

'llowanc.e is provided for withdrawal limits and reactor power imbalance limits a. he eneeded for a period of two hours without specification violation.

& wptance rod positions and imbalance must be achieved within the two-hour

' i..

rer Lod or appropriate action such as a reduction of power taken.

16/11/3 opii. iring restrictions are included in Technical Specification 3.5.2.5d to prevent excessive power peaking by transient xenon. The xenon reactivity must be beyond the "undershoot" region and asymptotically a;.proaching its squilibrium value at rated power.

(.

REFERENCES 1Section 3.2.2.1.2 2Section 14.2.2.2 6

3.5-il l

NOV 2 61974

UNITED STATES i.

h ATOMIC ENERGY COMMISSION

% % /[l 7

WASHINGTON D.C. 20545

%n DUKE PO'4ER COMPANY DOCKET No. 50-287 OCONEE NUCLEAR STATION, UNIT 3_

AMENDMENT TO FACILITY OPERATING LICENSE 4

Amendment No. 3 License No. DPR-55 1.

The Atomic Energy Commission (the Commission) having found t%t:

A.

The application for amendment by Duke Power Company (the licensee) dated September 20, 1974, as supplemented October 8 and 31, 1974, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10 CPR Chapter I; B.

The facility will operatsi in conformity with the application, the provisions of the Act, and the rules and regulations of the Cosa:ission; C.

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

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

Prior public notice of this amendment is not required since the amendment does not involve a significant hazards consideration.

2.

Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.3 of Facility License No. DFR-55 is hereby amended to read as follows:

=

.- 2 --

"B.

Technical Specifications 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. 3."

3.

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

FOR THE ATOMIC ENERGY COMMISSION Karl R. Coller, Assistant Director for Operating Reactors Directorate of Licensing

Attachment:

Change No. 3 to Technical Specifications Date of Issuance:

November 26, 1974 i

l I

r ATTACHMENT TO LICENSE AMENDMENTS AMEND}ENT NO. 6 TO FACILITY LICENSE NO. DPR-38, CHANGE NO.16 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 6 70 FACILITY LICENSE NO. DPR-47, CHANGE NO. 11 TO TECHNICAL SPECIFICATIONS; AMENDMENT NO. 3 TO FACILITY LICENSE NO. DPR-55, CHANGE NO. 3 TO TECHNICAL SPECIFICATIONS;

)

DUKE POWER COMPANY l

OCONEE NUCLEAR STATION, UNITS 1, 2. AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287 l

Ravise Appendix A as follows:

Remove Pages Insert New Pages 2.1-1 & 2.1-2 2.1-1 & 2.1-2 2.1-3 2.1-3, 2.1-3a, 2.1-3b &

2.1-4 t

2.1-4 2.1-4a 2.1-7 2.1-7 4

2.1-10 2.1-10 2.3-1 & 2.3-2 2.3-1 & 2.3-2 2.3-3 & 2.3-4 2.3-3 & 2.3-4 2..?-5 2.3-5 2.F8 2.3-8 & 2.3-8a 2.F11 2.3-11 3.5-12 3.5-12 3.5-13 3.5-13 Blank page 3.5-18 3.5-18 3.5-21 3.5-21 1

v

-,,,...._m

x k

7,

Remove Pages lpsert New p ges, 3.5-24

,a 3.5-24 3.11-1 6

3. u-1 3.5-6 & 3.5-7 3.5-6 & 3.5-7 3.5-8 & 3.5-9 3.5-8 5 3.5-9 3.5-10 & 3.5-11 3.5-10 & 3.5-11

.?

-?

i i

l a

i i

l i

1

2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperaturt, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification The combination 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-lC-Unit 3 and to the right of the line, the safety limit is exceeded.

The combination of reactor thermal 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.

pases - Unit 1 The safety limits presented for Oconee Unit I have been generated using BAW-2 critical heat flux (CHF) correlation (1}and the actual measured flow rate at Oconee Unit 1 (2). This development is discussed in the Oconee 1, Cycle 2-Reload Report, reference (2). The flow rate utilized is 107.6 percent of the i

d2 sign flow (131.32 x 106 lbs/hr) based on four-pump operation.(2) 5 To maintain the integrity of the fuel cladding and to prevent fission product 2;

release, it is necessary to prevent overheating of the cladding under normal operating conditions.

This is accomplished by operating within the nucleate s

boiling regime of heat transfer, wherein the heat transfer coefficient is large enough 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 c.1 adding temperatures and the possibility of cladding failure. Although DNB is not'an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure 2.1-1 NOV 2 6 G74

can be related to DNB through the use of the BAW-2 correlation (1). The BAW-2 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.32.

A DNBR of 1.32 corresponds to a 95 percent probability at a 99 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operat ;g canditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference j

in these two pressures is nominally 45 psi; however, only a 30 psi drop was

(

assumed in reducing the pressure trip setponts to correspond to the elevated location where the pressure is actually measured.

l The curve presented in Figure 2.1-1A represents the conditions at which a minimum DNBR of 1.32 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pum coolant flow is 107.6 percent of 131.3 x 10gs are operating (minimum reactor lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential fuel densifi-cation effects, which result in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2A are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:

f)

C 1.

The 1.32 DNBR limit produced by the combination of the radial peak, axial u

peak and position of the axial peak that yields no less than a 1.32 DNBR.

2.

The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 20.15 kw/f t for Unit 1.

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

The specified flow rates for Curves 1, 2, 3 and 4 of Figure 2.1-2A correspond 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 2.1-1A is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3A (because the fcur-pump pressure - temperature restriction is known to _ a more limiting than the 3 and 2 pump combinations, only the four pump limit has been shown on Figure 2.1-3A).

~

n e u.a Laum ace:ai power for three-pu:ap operat i.m L. 57 percent' d n to a

. peyer levoi,3jip produced by the flux-flow ratio 75 percent flow x 1.08 =

9? pe rcent power, plus the naximum calibratian and instrument errne.

The maximur. therea t power for othe r coolant pump cand itiona are produced in a l

similar manner.

l 2.1-2

.t NOV 2 61974

O l

For Figure 2.1-3A, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.32.

The 1.32 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.

16/11/2 References

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

(1) Oconee 1, Cycle 2 - Reload Report - BAW-1409, Sepetmeber,1974.

2.1-3 NOV 2 61974 t

~

~

Bases - Units 2 and 3 To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough 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 not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the W-3 correlation.(1)

The W-3 correlation has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operati.:n, normal operational transients, and anticipated trans'.ents is limited to 1.3.

A DNBR of 1.3 corresponds to a 94.3 percent probability at a 99 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. 'lhe difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.

15/7 l The curve presented in Figure 2.1-1B represents the conditions at which a 3 ;

2.1-1C minimum DNBR of 1.3 is predicted for the maximum possible thermal power (112%)

when four reactor coolant pumps are operating (minimum reactor coolant flow is 131.3 x 106 lbs/hr). This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:

1.78;F

= 1.50 2.67; F F

=

=

9 AH The design peaking combination results in a more conservative DNBR than any other shape that exists during normal operation.

The curves of Figure 2.1-2B are based on the more restrictive of two thermal 1o/31, 2.1-2C 3

limits and include the ef fects of potential fuel densification:

N

i. 'i he L. 3 DNP.R limit produced by a nuclear powar peaking factor of F'

= 2.67 the radial peak, axial peak and position of kho a th. co. oina tiun ot

.txlat peak that yields no less than 1.3 DNBR.

2.

The combination of radial and axial peak that causes central fuel melting l15/.11 at the hot spot.

The limit is 19.8 kw/ft - Unit 2 19.8 kw/ft - Unit 3 3

l 2.1-3a l

l l

NOV 2 61974 l

i

~

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

The specifieu flos tates for Curves 1, 2, 3, and 4 of Figure 2.1-2B correspond 15/31 2.1-2C 3

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 of all possible reactor 16/11/

2.1-1C 3

coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.

2.1-3C

'L6/ll/

The curves of Figure 2.1-3B represent the conditions at which a minimum DNBR 2.1-3C 3

of 1.3 is predicted at the maximum possible thermal power for the number af 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.

Using a local quality limit of 15 percent at the point of minimum DNBR as a 16/11 basis for Curves 2 and 4 of Figure 2.1-3B is a conservative criterion even 2.1-3C 3

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 continually 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% - Unit 2 15/11 86% - Unit 3 3

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

l.07 = 80%

power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

16/11 For each curve of Figure 2.1-3B, a pressure-temperature point above and to the 2.1-3C

~

lef t of the curve would result in a DNBR greater than 1.3 or a local qualicy 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

d n the left of the other curves.

REFERENCES (1) FSAR, Sec tion 3. 2.3.1.1 (2) FSAR, Section 3.2.3.1.1.c (1) FSAR, Section 3.2.3.1.1.k 2.1-3b NDV 2 6 $74

g, i

(4) -The following papers which were presented at the Winter Annual Meeting, ASME, November 18 1969, during the "Two-phase Flow and Heat Transfer in Rod Bundles Symposium:"

(a) Wilson, et al.

" Critical Heat Flux in Non-Uniform Heater Rod Bundles" (b) Gellerstedt, et al.

" Correlation of a Critical Heat Flux in a Bundle Cooled by Pressurized Water" e

i l

f 2.1-4 NOV 2 61974

2500 2400

.5 2300 via a

u, 2

2200 3o e

-a 2100 2000

/

1900 560 580 600 620 640 660 Reactor Outlet Temperature, F pt s,-rFcilon 3.'.Eis ' i,i t S r;

h..: r:a' :,OCOTJ'-E NUCLEAR STATION l

.%.7 l

F i gu r e 2. '. - 1!. llA/11/3 NOV 2 61974

Thermal Power level, t,

- 120

--100 1

2)

" 80 (3AND4}

=

60 g

x

- 40 l

20

-40

-20 0

+20

+40 Reactor Poser imbalance, 5 CURVE REACTOR COOLANT FLOW (LB/4R) 6 1

131.3 x 10 6

2 98.1 x 10 6

3 84.4 x 10 '

6 4

60.1 x 10 CORE PROTECTION SAFETY LIMITS t:NTT 1 2,1_7

[

= OCONEE NUCLEAR STATION Figure 2.1-2A f16/11/3 MV 2 019N

g 2500 2400

?

2300 E.

a; em U

I 2200 I

8 5a 2100 2000

/

I 1900 j

560 580 600 620 640 660 Reactor Outlet Temperature, F CORE PROTECTION SAFETY LIMITS C::L:.

2.1-10

/

A'; OCONEE NUCLEAR STATION rigure 2.1-3;.

l16/11/3 NOV 2 61974

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRU tENTA11N Applicability Applies to instru=ents monitoring reactor power, reactor power i..cCc.:

reactor coolant systes pressure, reactor coolant outlet teurer:3.r.

number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protective action to prevent any combinatiec ut

r....a variables from exceeding a safety limit.

Specification The reactor protective system trip setting limits and the permisaible c,f,. :m.4 f or the instrument channels shall be as stated in Table 2.3-LA - Unlu 1 ar.d 2.3-1B - Unit 2 2.3-1C - Unit 3 16/11/3 Figure 2.3-2Al } Unit 1 2.3-2A2 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% (0.0% for h6/11/3 a.

Unit 1) of rated 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 55% of rated power for single loop operation. Power /RC pump trip L6/11/2' setpoint is reset to 55% for all modes of 2 pump operation for Unit 1.)

c.

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

Bases

'1he reactor protective system consists of four instrument channels to mitor each of several selected plant conditions which will cause a reacb..

any one of these conditions deviates from a pre-selected operat!ua

.r.

salsty limit may be reached.

the degree that a

the t.r ip ret t.ing I nuito for protective system instru=eutatlos

.r. i lable 2.J-1A - Unit 1.

The safety analysis has been based

,a 2.3 Unit 2 2.1-1C - Un tt i

, wree irstoa.e.st ar a a er 'p set points plus calibratica ar

... c c..

L.. i n. r w t r...u a A ce ctor tcip it ate. power ie. vel (neucron flux) is prr.i.

.!ac:.1ge to the rt:sl c hidding from reactivity excursions t.a :

.7

,/ pressuse a.ut t.*.vero:uru measurecents.

2.3-1 NOV E6 G74

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.

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 a

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

The power level trip set point produced by the poyer-to-flow ratio provides both tigh power level and low 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 DNB pro-tection for all modes of pump operation. For every flow rate there is a maxi-mum permissible power level. ad for every power level there is a minimum permissible low flow rats. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1A are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if power is 108% and reactor flow rate is 100%, or flow rate is 93% and power level is 100%.

2.

Trip would occur when three reactor coolant pumps are oper:ating if power is 81.0% and reactor flow rate is 74.7% or flow rate is 63% and power level is 75%.

3.

Trip would occur when tvc reactor coolant pumps are operating in a single loop if power-is 59% and the operating loop flow rate is 54.5% or flow rate is 43% and power level is 46%.

4.

Trip would occur when one reactor coolant pump is operating in each loop (total of two 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 maxi =um calibration and instrumentation errors for the power level trip were used.

The power-i= balance boundaries are established in order to prevent reactor ther al li=its from being exceeded. These thermal limits are either power ir. k..lft l in : 1, e r DN33 timits.

Se ra eror power imbalance (power in tep half or core cinus power in the bottoa half of core) reduces the pcwer reel trip produced by the power-to-ficw ratio such that the boundaries of

. are produced.

he rm:er-te-flow ratio reduces the pcwer 16/11/3 2.

v..

1 itu.a

.3.a.

2.1-2n - Unit 2

2. 3-:c - Unit 3 l

2.3-2 NOV 2 6 $74

lev' l trip and associated reactor power / reactor power-imbalance boundaries Sy e

1.0Pi - Unit 1 for a 1% flow reduction.

1.07% - Unit 2 1.07% - Unit.3 Pump Monitors The pump conitors prevent the minimu= 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 The pump monitors also restrict the power level for the number of ratio.

pumps in operation.

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-lC - 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) psis and variable low pressure (13.77 Tout-618D trip 16/11/3 (16.25TU[7756)

(1800) psig (16.25 T (1800) psig 7756) setpoints shown in Figure 2.3-1A have been established to maintain the DNS 2.3-1B 2.3-lc 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 bI/11/2 variable low reactor coolant system pressure trip value of (13.77 Tout - 6221)

(16.25 T

-7796)

(16.25I[]*-7796)

Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting li=it (619 F) shown in Figure 2.3-1A has bee.t established to prevent excessive core coolant 2.1-13 2.1-IC tenperatures in the operating range. Due to calibration and instrumentation s - ra r,, '_ he.:afety analysis used a trip set point of 620*F.

, n e :m u n..

building pressure trip setting lici (4 psig) provides

't: 0 r.: ::

.e...

. c ;..x.

reaccet u i; will es.ur in :Se ';clikely cvent af a

. 3 - c.' -

' me accident, even in the absence cf a low reactor coolant system

,rr.s v are t ::..

2.3-3 NOV 2 6 E;4

O

".j:JI.7,,3 m ss In order to provide for control rod drive tests, zero power physics testing, and startup procedures, thersjWsprovision for bypassing certain sessants of the reactor protection system.

The reactor protection systen segments which can be bypassed are shown in Table 2.3-1A.

Two conditions are imposed when 2.3-13 2.3-1C the bypass is used:

A 1.

By administrative control the nuclear overpower trip set point must be reduced to a value 1 5.0% of rated power during reactor shutdown.

2.

A high reactor coolant system pressure trip setpoint of 1720 peig is autocatically imposed.

Tha pucpose of the 1720 psig high pressure trip set point is to prevent normal operaticn with port of the reactor protection system bypassed.

This high pressure trip set point is lower than the normal low pt' essure trip set point so that the reactor must be tripped before the bypass is.. initiated.

The over power trip set point of 15.0% prevents any,significant reactor power f rom being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0% of. rated power if none of the reactor coolant pumps were operating.

Two Pump Operatic 1 A.

Two Loop Operation Operation with one pump in each loop will be allowed only following reactor shutdown. Af ter shutdown has occurred, the following actions will permit operation with one pump in each loop:

16/11/3 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 Operation Single loop operation is permitted only after the reactor has been tripped.

Af ter the pcmp contact monitor trip has occurred, the following actions u!'l pernit single loop operation:

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

-i:> cne of the two protectiv channels receiving outlet tempersture

'av

.t. ton frc. sensors in the Idle Loop.

(Unit 1) Reset the protective system maximum allowable setpoints as

' '. vr n 7'gure 2.3-2A2.

Tripping one of the two protective channels 13/11/3

t t.n t u-nperature inf omu ton f ree the idle loop assures

r.

<t 'i t. P. Et ip Icdic Of One Out of t%).

'. ?. r..

b "o

..b e

. i'

..c m...i.

1 2. )- a MOV 2 6 E74

e e

2400 d

2300

.?

E.

J 2200 E

U

+

S a.

Z

." 2100 8

o a

t; 3

2000 1900 i

i i

i 540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SET POINTS UNIT 1 h

i\\ OCONEE NUCLEAR STATION 2.1-5 Figure ~.3-1A (16/11/3 I

NOV 2 61974

Po er Level, 5 120 FOUR PUNP 10 SET FOINTS 0

S d' THREE PUNP SET POINTS 40 20 40

-20 0

20 40 Reactor Power imbalance, 5 PRO TECTI V E SY S TEM M A.t i MUM ALLOC.9tE SET POINTS c:::T 1

.A f" %

. 3-5 li# li: OCONEE NUCLEAR STATION

'b75f' l 16/31/~I Figure 2.3-2A1j f40V % B 1974 if

O Power Level..

120

-- 100 80 60 TWO PUMP SET POINTS 40 TWO PUMPS IN ONE LOOP ONE PUMP IN EACH LOOP 20 40

-20 0

20 40 Reactor Power imoalance 5 PROTECT!VE SYSTEM.9AXivuy ALLONASLE SET POINTS UNIT 1

([hu:*'s.\\

2.3-Sa n

. OCONEE NUCLEAR STATION Figure 2.3-2A2 f 16/11/3 NOV 2 61974 1

Table 2.3-1A t! nit 1 Reactor Protective System Trip Setting Limits Two Reactor one Reactor Four Reactor Three Reacto' Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A operating in l

Operating Operating Single Loop F.ach Loop (Operating Power (Operating Power (Operating Power (Operating Power Shutdown F P's ?.-m.

-100% Itated)

-75% Rated)

' -46% Rated)

-492 Rated)

Bypas t 1.

.uc lea r

. r P-105.5 105.5 105.5 105.5 5.0(3)

F cat i 6.

J. ' to s lito I.- I*.. ~ tia.ed 1.08 times flow 1.08 times flow 1.08 times flow 1.08 times flow bypassed on 51aw (J, and.a.alance.

.minus reduction minus reduction minus reduction minus reduction (2 Nat..!)

due to imbalance due to imbalance due to imbalance due to imbalance 1.

?:ucie.n i.C. r Pe t.v.ed NA NA 55% (5)(6) 55% (5)

Bypassed un hme %ritorn. D. R.ited) 4.

liigti s...s c o Co.

2355 2355 2355 2355 17200)

%. r *. - '.

.. re.

. i y, ?t.n.

p 5.

Law k... t r L..u i..'

1985 1985 1985 1985 Bypassed.

h sys t a. i r. :..u r u, d e;. m n.

16/

6.

Va r i..ui., :. w Ke..

.:r (13.77 7*"* -6181)(I) (13.77 T - 6181)(1) (13.77 7

- 6181)(1) (13.77 T

- 6181) D)

Bypassed

• "t ut

....o 4... t :,

.. r u p.: g. F i ri.

7.

<<.ac t o r t:.

..p.

619 619 619 (6) 619

'19 s

6...dax.

ft. 11igh I.e.ct or I at ing 4

4 4

4 4

'Prenustr.,.>r,..

(I) T is a n.!cgr.. Fahrenheit ("F).

(5) Reactor power level trip set point pruduced by pump contact monitor reset to 55.0%.

j

(.')

rear t or t..Iane ? ystsa Flow, 2.

(6) Specification 3.1.8 applies. Trip one of the (3) 1.dmint tr:.ttvely controlled reduction set two protection channels receiving outict temper-l onl y. tor i g s e...

t <.r ths.e down, 2

ature infornaation from sensors in the idle loop.

,)

a<

(4) Au t os.a t i.,. ! !, -.: when other segments of 1:

the RPS a r e byp..,wd.

g cn dw A

i 4

~

I.

Rod inder is the percentage suz of the eithdra al of the operating groups.

2. These withdrawal limits are effective only for 25015 full power days of cperation after issuance of Amer &ents No. 6, 6 and 3, respectively, of Licenses No. DPR-38, -47, and -55 173 209 100 154 213 g45 Power Level Cutoff Restricted Regfon 80 122 230 80% 4 75%

~

60 ae

[

Permissinle

.(52% P) 3

~

$[p Operating Region j

40 i

20 0'

O 50 100 150 200 250 300 Red index, 5 tithdrawal 25 50 75 100 f

f f

i 0,

,25

,50

,75 100 Gp7 0

25 50 75 100 Go6 i

I t

i I

Cp5 CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 4 PDT OPERATION UNIT 1 l

nub') OCONEE NUCLEAR STATION 3.5-12 16/11/3 Figure 3.5.2-1A1 i

l NOV 2 61974

.w..ww,-rw o

S

_g em O

e s

3 e

i e'

BLANK PAGE i

3.5-13 N'M 2 6 1974

1 Rod index is the percentage sum of the citndra:al of the operating

'g r oup s.

(The applicable power level cutoff is 100% power) 2 Pump Withdrawal Limit x 3-N 100 150 g

275

-3 Restricted j

f Region s

80 S

d a

E 6

Permissible

• e

[

Operating k

3 Region 60 E

2 2

i 2

E 40 -

.=

I d'

20 0

i i

e i

0 50 100 150 iLO 250 300 Roa index. $ withdrawal CONTROL ROD GROUP WITHDRAWAL LIMITS FOR 3 AND 2 PUMP OPERATION UNIT 1 3.5-18 OCONEE NUCLEAR STATION Figure 3.5.2-2A 16/11/3 NOV 2 61974

Power, 5 of 2568 M t 110

-20.4

+14.1 1025 100 90 l

80

- 30, 75 70 60 t

4 i

-- 50

-31.2,52

+28.1, 52 40 l

9 f

f I

f f

-30

-20

-13 0

10 20 30 Core imoalance, %

OPEPa\\TIONAL P0b'ER IMSALANCE ENVELOPE UNIT 1

(,a) OCONEE NUCLEAR S 3.5-21 figure 3.5.2-3A '16/11/:

NOV 2 61974 _ _

20 9atch4 Fuel; Generic FAC IO Results (BAW-10091)

\\ Batch 2 & 3 Fuel 1

3a 14 0

E i

12 4

g 10 0

2 4

6 8

10 12 Distance from inlet, ft I

1 LOCA LIMITED Y.AXIFD! ALLOVASLE LINEAR HEAT RATE

{::=0cu s, OCONEE NUCLEAR STATION

3. 5 -2 :,

Qg

~

Figure 3.5.2-4 16/11/3 NOV % 61974

3.11 MAXDIUM POWER RESTRICTION Aoplicability Applies to the nuclear steam supply system of Units 2 and 3 reactors.

Objective To maintain core life margin in reserve until the s'ystem has performed under operating conditions and design objectives for a significant peried of time.

Saecifica tion l16/11/3 3.11.1 The first reactor core in Unit 2 may not be operated beyon.!

4 11,040 effective full power hours until supporting analyst and data pertinent to fuel clad collapse under fuel densif;-

cation conditions have been approved by the Directorate of Licensing.

3.

4 The first reactor core in Unit 3 may not be operated beyond 10,944 effective full power hours until supporting analysis and data pertinent to fuel clad collapse under fuel densifi-cation conditions have been approved by the Directorate of Licensing.

Bases The licensing staff has reviewed the effects of fuel densification for the first core in Oconee Units 2 and 3 and concluded that clad collapse will not 16/11/3 take place within the first fuel cycle (11,040 effective full power hours for Unit 3 and 10,944 effective full power hours for Unit 3).

However, the clad collapse model used is questionable for extrapolation of clad collapse time out beyond the first fuel cycle because of limited experi-mental verification.

3.11-1 NOV 2 61974

O 3.5.2 Control Rod Group and Power Distribution Limits Applicability This specification applies tc power distribution and operation of control rods during power operation.

Objective To assure an acceptable core power distribution' during power operation, to set a limit on potential reactivity insertion,Yrom a hypothetical control rod ejection, and to assure core suberiticality af ter a reactor trip.

Specification The available shutdown margin shall be not less than ik Ak/k with 3.5.2.1 the highest worth control rod fully withdrawn.

3.5.2.2 Operation with inoperable rods:

7 If a control rod is misaligned with its group average by more a.

than an indicated nine (9) inches, the rod shall be declared inoperable. The rod with-the greatest misalignment shall 1

be evaluated first. The position of a rod declared inoperable due to misalignment shall not be included in computing i

i the average position of the group for determining the operability of rods with lesser misalignments.

i

\\

b.

If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indiestions or in or out limit lights, the rod shall be declared to be inoperable.

If a control rod cannot meet the requirements of Specification c.

4.7.1, the rod shall be declared inoperable.

d.

If a control rod is found to be improperly programmed per Specification 4.7.2, the rod shall be declared inoperable until j

properly programmed.

Operation with more than one inoperable rod in the safety or e.

regulating rod groups shall not be permitted.

f.

If a control rod in the regulating or safety rod groups is declared iaoperable in the withdrawn position, an evaluation shall be initiated immediately to verify the existance of 17.

ak/k hot shutdown margin.

Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are fully withdrawn, whichever occurs firs t.

Simultaneously, a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.

3.5-6 314 g., ;,

a If 'within one (1) hour of' determination of an inoperable red, g.

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

h.

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

1.

If a control rod in the regulating or safety rod groups is 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 gr'oups j.

ir dec?ared ineperable, operation above 60 percent of rated powrc mcy continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained 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 worth of a single inserted control rod shall not exceed 0.5%

, ak/k at rated power or 1.0% ak/k at hot zero power except for i

physics testing when the requirements of Specification 3.1.9 shall apply.

3.5.2.4 Quadrant Power Tilt Whenever the quadrant power tilt exceeds 4 peteent, except for a.

physics tests, the quadrant tilt shall be reduced to less than j

4 percent within two hours or the following actions shall be taken:

(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each 1 percent tilt in excess of 4 percent,below the power 1evel cutoff (see Figures 3.5.2-1A1, 3.5.S-131, 16/11/3 3.5.2-132, 3.5.2-133, 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3).

(2) If less than four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 percent of full power for each I percent tilt below the power allowable for the reactor coolant pump combination as defined by Specification 2.3.

(3) Except as ' provided in 3.5.2.4.b, >.the reac tor shall be brought to the hot shutdown condition within four hours if the quadrant tilt is not reduced to less than 4 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. If the quadrant tilt exceeds 4 perc ant and there is simultaneous indication of a misaligned contro'. r cd per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60 percent of the thermal power' allowable for the reactor coolant 3.5-7 NOV :: 31974 s

pump combination.

(

Except for physics tests, if quadrant tilt exceeds 9 percent, a c.

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 tc 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 thercal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a reduction of 2 percent of full power for each 1 per-cent tilt for the maximum tilt observed prior to shutdown.

Quadrant power tilt shall be monitored on a minimum f'requency of e,

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 (safety rod'vithdrawal) does not prohibit a.

the exercising 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.

b'.

Operating rod group overlap shall be 25% i 5% between two sequential groups, except for physics tests.

(.

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

drawal limits are specified on Figures 3.5.2-1A1 (Unit 1),

16/11/'

3.5.2-1B1, 3.5.2-1B2 and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump operation and on Figures 3.5.2-2A (Unit 1), 3.5.2-2B (Unit 2), and 3.5.2-2C (Unic 3) for three or two pump operation.

If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod posicions shall then be attained within i

two hours.

l l

d. Except for physics tests, power shall not be increased above the power igvel cutoff as shown on Figure 3.5.2-1A1 (Unit 1), 3.5.2-1B1, l 16/11/ ' ;

3.5.2-1B2, and 3.5.2-133 (Unit 2), and 3.5.2-1C1, 3.5.2-1C2, and 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 chall be asymptotically approaching the value for operation at steady-state rated power.

l l

i l

-i 3.5-8

,9f," :. ', IU r

.l 3.5.2.6' Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.

Except for physics tests, imbalance shall be maint.ained within the envelope defined by Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C.

If 16/11/3 the imoalance is not within the envelope defined by Figure 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C, corrective measures shall be taken to achieve an acceptable imbalance.

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

1.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintend-;

16/11/3

(

e 3.5-9 NOV 2 61974

'Basas

[

The power-imbalance envelope defined in Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C is based on LOCA analyses which hcve 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 situation that would cause the Final acceptance criteria to be approached should a LOCA occur is highly improbable because all of the power distribution 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.

ruel densification effects d.

Hot rod manufacturing tolerance factors The 25% i 5% overlap between successive conErol 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 1

Safety 2

Safety 3

Safety

(

4 Safety 5

Regulating 6

Regulating 7

Xenen transient override 8

APSR (axial power shaping bank)

The minimum available rod worth provides for achieving hot shutdown by reactor trip at any time assuming the highest worth control rod remains in the full out position.(1)

Inserted rod groups during power operation will not contain single rod worths greater than0.5% ak/k.

This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident.(2) A single inserted control rod worth of 1.0% ak/k at beginning of life, hot, zero power would result in the same transient peak thermal power and, therefore, the same environmental consequences as a 0.5% ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with Group 1.

Groups 5, 6, and 7 are overlapped 25 percent.

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

• • Actual operating limits der 2nd on whether or not incore or excore detectors.

are used and their respective instrument and calibration errors.

The method i

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

3.5-10

. yi 2 : 374

The quadrant power tilt limits set forth in Specification 3.5.2.4 hava b2cn established within the thermal analysis design base using the definition of quadrant power tilt given -in Technical Specifications, Section 1.6.

These limits.in conjunction with the control rod position limits in Specification 3.5.2.5c ensure that design peak heat rate criteria are not exceeded during s

normal operation when including the effects of potential fuel densification.

4 i

tilt and axial imbalance monitoring in Specifications 3.5.2.4 The quadrant i

J 1 ' 2.6, respectively, normally will be performed in the process computer.

iw. hour frequency for monitoring these quantities will provide adequate i.

)

isivelliance when the computer is out of service.

'llowance is provided for withdrawal limits and reactor power imbalance limits be exceeded for a period of two hours without specification violation.

1 4

so A ciptance rod positions and imbalance must be achieved within the two-hour Lod or appropriate action such as a reduction of power taken.

' i.+ p t 16/11/3 4

op i. iring restrictions are included in Technical Specification 3.5.2.5d to The xenon reactivity prevent excessive power peaking by transient xenon.

must be beyond the "undershoot" region and asymptotically approaching its squilibrium value at rated power.

[

g REFERENCES ISection 3.2.2.1.2 Section 14.2.2.2 4

4 s

l I

i i

t 3.5-11 fl0V % 61974 f

m e

g

-wr-y w-

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e 5

-y4-y ww-py-y w