ML19312B891

Proposed Revisions to Tech Specs 2.1,2.3,3.1.9 & 3.5.2 Re Core Safety Limits,Safety Sys Settings Limitation,Low Power Physics Testing Restrictions & Control Rod Group & Power Distribution Limit

ML19312B891
Person / Time
Site:	Oconee
Issue date:	12/01/1975
From:	DUKE POWER CO.
To:
Shared Package
ML19312B882	List: ML19312B880 ML19312B891
References
NUDOCS 7911250080
Download: ML19312B891 (27)

Text

. g -

e ATTACIDENT 1 Proposed Technical Specification Revisions To Support Oconee Unit 1 Cycle 3 Operation l

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i December 31, 1975

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.V 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 c2idding.

Soecification The combination of the reactor system pressure and coolant temps.rature 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 r

• is below 2.1-13-Unit 2 2.1-1C-Unit 3 and to the tight 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 itnit 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 2.1-23-Unit 2 2.1-2C-Unit 3 imbalance point is above the line for the specified flow, the safety limit is exceeded.

?asee - Unit 1 The safety limits presented for Oconee critical heat flux (C'dF) correlation (1) Unit and1 the haveactual been generated using measured flow rate BAW-2 at Oconee Unit 1 (2). This development is discussed in the Oconee 1, Cycle 3 Reload Report, reference (2) . The flow rate utilized is 107.6 percent of the design flow (131.32 x 106 lbs/hr) based on four-pump operation. (2)

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 "deparrure from sueleate boiling" (DNB) . At this point, '

there is a sharp reduction of the heat transfer coefficient, which would result I 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 i 2.1-1 l

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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 1.30. A DNBR of 1.30 corresponds to a 95 percent probability at a 95 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 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 1.30 is predicted for the maxi =um possible thermal power l (112 percent) when four reactor coolant pu=

coolant flow is 107.6 percent of 131.3 x 10gs are operating lbs/hr.). This (minimum reactoron curve is based 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:

1. The 1.30 DNBR limit produced by the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.

2. The combination of radial 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 power peaking.

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

The maximum thermal power for three-pump operation is 35.3 percent due to a i power level trip produced by the flux-flow ratio 74.7 percent flow x 1.055 = l 78.8 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner. )

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i 2.1-2 I i l l

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. 30. The 1. 30DN3R curve for four- l pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of the four pump curve will be above and to the lef t of the other curves.

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

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

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3 2.1-3

2400 -

2300 -

ACCEPTABLE OPERATION

, UNACCEPTABLE 3_ OPERATION n'

E 2100 _

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=

E

= 2000 -

8 1900 -

1900 -

I  ! I I 560 580 500 S20 540 560 Reactor Ou tle t is.rcer atur s. *F LT!IT 1 CCFE FFOTE~TICU S&-iY UMITS 3@h a OCONEE NUCLEAR STATION FIstas2.1-IA 2.1-4

Thermal Power Level. ",

. 120 .

UNACCEPTABLE OPERATION

(-30,i 2)

(11 2 )

(+3a. 12) 110 ACCEPTABLE 4 PUMP OPERATION 100

(-40.35) 90

(-3o.s5 3) (ss.3) (+3s.55.3) \ (+5o. 3s)

ACCEPTABLE 3 & 4

. 80 PUMP OPERATION

(-40.sa.3) -- 70

(-30.sa.2) 60 (5s.2) (+3s.ss.2)

ACCEPTABLE 2.3 h('~so 53 3)

& .i PUMP

-- 30 OPERATION

(-4o.41.2)

. 40 (3 & 4

(+s0.3 .2)

- 30

. 20  ;

. 10

- a

-40 20 0 +20 +40 +60 Reactar Power imcal3nce 5 CUR'IE REACTOR COOLANT FLOW (lachr) 1 141.3 x 10 2 105.6 t 10 3 3 59.3 x 10 4 S4.7 x 106 UNIT 1 CCRE FFOTECTICN hY LIMITS

CCONEE NUCLEAR STATION Fracas 2,1-2A l 2.1- 7 l

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I 2 3 I400 .

2300 .

ACCEPTABLE OPERATION 2266 a E

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O 2100 _

f 2.

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E u 2000 _

E a

1900 -

1800 -

550 580 G00 820 S$0 560 Reactor Outlet Temaer3:ure.*F CURVE REACTOR COOLANT Fi.01 (l3.nr) 1 141.3 x 105DNSR Limit 2 105.S x 10 5DNGR Li.mit -

3 59.3 x 103 Quality Limit

nt&mnUllIT CCONEE1 NUCLEAR STATION FMURE 2.1-3A 2.1-10

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability l

Applies to instruments monitoring reactor power, reactor power imbalance,

> reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.

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

Specification The reac or protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1A - Unit 1 and 2.3 Unit 2 Figure 2.3-2A - Unit 1 l 2.3 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 55% 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% for all modes of 2 pump operation.)

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

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

The trip setting limits for protective system instrumentation are listed in Table 2.3-1A - Unit 1. The safety analysis has been based upon these protective 2.3 Unit 2 2.3-lc - Unit 3 system instrumentation trip set points plus calibration and instrumentation erTors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding frem reactivity excursions too rapid to be 3

detected by pressure and temperature measurements.

1 2.3-1

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%uring n:r=al plant opara:1on vi:5 cil reactor coolant pumps opcracing, raccror crip is inicicecd wh:n chs racetor psvar icval racch s 105.5% of rettd povar. Adding to thic tha possibla vnristion in trip soepoints dua .;

to calibra: ion and instrument errors, the saximum actual power at which a *J crip would be actuated could be 112%, which is = ore conservative than the l value used in the safety analysis. (4)

Overeover Trip Based on Flow and Imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DN3R 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 providas both high power level and low flow protection in the even: :he reactor power level ine: eases or the ~eactor coolant flow rate decreases. The power level crip set point produced by the power-to-flow ratio provides overpower DN3 pro-cac:icn for all = odes of pu=p operation. For every flow rate there is a =axi-mus permissible power level, and for every power level there is a sini=um permissible low flow rate. Typical power level and low flow rate combinations for the pump sicucations of Table 2.3-1A are as follows:

1. Trip would occur when four reac:c coolant pu=ps are operating if power is 105.5% and reactor flow ra:e is 100%, or flow race is 94.8% and power level is 100%.

2. Trip would occur when :hree reactor coolant pu=ps are operating if power is 73.3% and reactor flow rate is 74.7% or flow race is 71.1" and power level is 75%.

3. Trip would occur when two reactor coolant pumps are opera:ing in a single loop if power is 51.7% and the operating loop flow r.:e is 54.5% or flow race is 48.5% and power level is 46%.

4. Trip would occur when one reactor coolant pump is operating in each loop (cotal of two pumps operating) if the power is 51.7% and reactor flow rate is 49.0% or flow race is 46.4% and the power level is 49%.

The flux-to-flov ratios for Uni: I account for the =axi=um variation from :he average value of the RC flo" signal in such a =anner :hac the reac:or pro:ec:1ve sys:em receives a conserva:1ve indica:icn of :he RC flow.

For safety calcula: ions the =aximum calibration and instrumenta:icn errors for :he power level : rip were used.

The power-imbalance boundaries are es:ablished in order :o preven: reac:or

hermal limits from being exceeded. These ther=al 1121:3 are ei:her power "eaking kw/f: lisi:s or TGR limi:s. The reac:or power imbalance (pewer in

.e :op half of core sinus power in the be::cs half of : ore) reduces :he power 'l level : rip produced by :he power-:o-flow ra:io such :ha: he boundaries of Figure 2.3-2A - Tni: 1 are produced. The power-to-flow ra:io reduces the pcwerl ,

2.3 Uni: 2 2.3-2C - Uni: 3 i 2.3-2 l

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level trip and associated reactor power / reactor power-imbalance boundaries by 1. 0 5 5 %- Unit 1 for a 1% flow reduction. I 1.07% - Unic 2 1.07% - Unit 3 For Unit 1, the power-to-flow reduction ratio is 0.949, and for Units 2 and 3, the power-to-flow reduction factor is 0.961 during single loop operation.

Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The circuitry monitoring pump operational status provides redundant trip protection for DNB by tripping the reactor on a signal diverse from that of the power-to-flow ratio. The pump monitors also restrict the power level for the number of pumps in operation.

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 over-power trip set point. The trip setting limit shown in Figure 2.3-1A - Unit 1 2.3 Unit 2 2.3-1C - Unit 3 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)

The low pressure (1800) psig and variable low pressure (11.14 T -4706) trip l (1800) psig (16.25 T "*-7756)

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

Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of (11.14 T -4746) l (16.25 T "* -7796)

(16.25 Iout

"* -7796)

Coolant Outlet Temperature The high reactor coolant outlet temperature trip settlng limit (619 F) showa 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 errors, the safety analysis used a trip set point of 620 F.

Reactor Building Pressure The high reactor building pressure trip setting limit (4 3sig) provides positive assurance that a reactor trip will occur in the talikely event of a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

l 2.3-3 l

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r Shutdown Bypass In order to provide for cottrol rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reac* > protection system segments which can be bypassed are shown in Table 2.2>1A. Two conditions are imposed when 2.3-13 2.3-1C the bypass is used:

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

2. A high 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 pressure trip set 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 ,5.0% prevents any significant reactor power from being produced when p- .orning the physics tests. Sufficient natural circulation (5) woulc. se available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.

Two Pump Operation A. Two Loop ~.'peration Opa- tion with one pump in each loop will be allowed only following reactor shutdown. After shutdown has occurred, reset the pump contact moaitor power level trip setpoint to 55.0%.

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 will permit single loop operation:

1. Reset the pump contact monitor power level trip setpoint ts 55.0:.

2. Trip one of the two protective channels receiving outlet temperature information from sensors in the Idle Loop. 1

3. Reset flux-flow setpoint to 0.949 (Unit 1). l 0.961 (Units 2,3)

REFERENCES (1) FSAR, Section 14.1.2.2 (4) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.7 (5) FSAR, Section 14.1.2.6 (3) FSAR, Section 14.1.2.8 2.3-4 1

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. -l 2400 P = 2355 psig T = SIS F 2300 -

i 2200 -

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ACCEPTABLE f

5 m 2100 OPERATION l u i I

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1 2000 M  !

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,' UNACCEPTABLE M ~ OPERATION

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A P = 1800 psig i I i i 540 550 580 500 620 640 Reactor Outlet Tincars:ura,*F 1

UNIT 1 l PROTECTIVE SYSTEM MAXI?UM -l

'g ALLOWAELE SET POINTS

,sut Mm CCONEE NUCLEAR STATION l

Freuas 2.3-1A l

l 2.3-5 i

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Power Level, ",

120 UNACCEPTABLE OPERATION

,110(105.5) 4-go ACCEPTA8LE g w 100 4 PUNP s' h OPERATION -

is s "a e

+' -

90 1 1

. 80 (7a.s) l ACCEPTABLE 3 & 4 PUMP PERATicN l

. 70 I

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2.3 &

• PUMP 0'E**7'0" l

- 40 l 30 20 o o 10 ll ll

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m m" 40 - 2' O O 2U i0 Reactor Power imoalanca, 5

'THE FLUX. FLOW SETPOINI FOR 2/0 ? UMP JPERAIl0N LI!II 1

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MUST SE SET AT 0.949 EFO ELIICN SYSI E G XI M I

?g ALLOWE SEi?OliiTS

nt!?*o CCONEE NUCLEAR STATION Fretas 2.3-2.a 2.3-8

Table 2.3-IA tini t i Neuctor Psotect ive System Trly Set t ing 1.imit u .

Two kcactor Due keactor Four kemctor Tlaree ke.setur Cootent rumpu Coulant Pump Cool.snt Pumps Coolant Puolis Operatisig in A Operdting in Operating Operating Single loop Eacle Imop (Operating Power (Operating Power (Oper. sting Power (Operating Power Stoutdown krS Segment -1001 kated) -751 kated) -462 k.atcJ) -492 kated) typass

1. Nclear Power Nu. 105.5 105.5 105.5 105.5 5.0 I3)

(Z NatcJ)

2. Nc!cas Puwer N a. based 1.0% t imes flow I .0% t ime u t iow 0.949 times tiow I .0% t imem flow typanacJ on Flow (2) and Iml.alance, manus reduction stuum seductico minum seduction minum reduction (I kated) Jue to imbalance Jue to imbalance Jue to Imbalance Jue to imbalance . g.

3. Nuclean- Power N s. BancJ NA NA 55Z (5)(6) 55Z (5) typammed on rump Neit tos a, (Z, kated)

4. Higli kcactor Coolant 2355 2355 2355 2355 1720(4)

System Psumbuse, pulg H.sa.

S. 1.uw kenctus Cool.snt lH00 1800 8800 1800 bypammed l

[ *.

System Psemuuse, pulg Hin.

6. Vas-imble low keactos ( ll .14 T 410f9(II ( 11. 4 4 T - 4106 )( } (11. 3 4 T,,g- 4 706 )I I (ll.14 T out

- 4706 ) hypammeJ t.colaut System t*genmuse pulg Hin.

1. ke ctor Coolant Temp. 619 619 619 (6) 619 619 F., H.sm.

8. Higli kescaua building 4 4 4 4 4 Psemmune, pulg, Ns.

..._____..____._____.._... ___ _ g (1) T""" in in Jegaeca Falirenheit ("F).

l:0 (5) kcactor power level trip met point produced by pump contact monitor remet to 55.02.

(2) kesctor Coulant System Flow, I.

(3) Administsatively controlled seduction met

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(6) Specification 3.1.8 applies. Trip one of Elie two piutection cliannels receiving outlet temper-only Juring seactos ut.u t down . b mus===

atuse information from meumorm in Elie idle loop.

(4) Au t oss.s t ic a l l y me t ubesa utlier mugments of a l.c krS aie I.yp.macJ. p I

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

3.1.9.1 Reactor Protective System Requirements

a. Below 1720 psig shutdown bypass trip setting limits shall apply in accordance with Table 2.3-1A - Unit 1.

2.3 Unit 2.

2.3-1C - Unit 3.

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

2.3 Unit 2.

2.3-1C - Unit 3.

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

3.1.9.3 Shutdown margin may not be reduced below 1.0% ak/k as required by Specification 3.5.2.1 with the exception that the stuck rod worth criterion does not apply during rod worth measurements.

Bases Technical Specification 3.1.9.2 will apply to both the source and Later=ediate ranges.

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

l 3.1-20

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g. If within one (1) hour of' determination of an inoperable rod, it is not determined that a luk/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the 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.

i. 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.i.2.3 .___The worths of_ single inserted control. rods during criticality - -

are limited by the restrictions of Specification 3.lma.5 and the -

control rod position limits defined in Specif t. cation 3.5.2.5.

3.5.2.4 Quadrant Power Tilt

a. 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 taken:

(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced by 2 perc<nt of full power for each 1 percent tilt La excess of 4 percent below the power level cutoff (see Figures 3.5.2-1A1, 3.5.2-1A2, 3.5.2-131, 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 ' percent tilt below the power allowable for the reactor coolant pu=p conbination as d 2 fined by Specification 2.3.

(3) Except as provided in 3.5.2.4.b, the reactor shall be brought to the hot shutdown condition within four hours if the quadrant tilt is not reduced to less than 4 percent af ter 24 houra,

b. If the quadrant tilt exceeds 4 percent and there is simultaneous indication of a sisaligned 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

J - _

i

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pump combination.

c. Except for physics tests, if quadrant tilt exceeds 9 percent, a controlled shutdown shall be initiated immediately and the reactor shall be brought to the hot shutdown condition within four hours.

d. Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and

' corrective action provided the 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 percent tilt for the maximum tilt observed prior to shutdown.

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

3.5.2.5 Control Red Positions

a. Technical Specification 3.1.3.5 does not prohibit the exercising of individual safety rods as required by Tr.ble 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.

c. Except for physics tests or exercising control rods, the control rod withdrawal li=tts are specified on Figures 3.5.2-1A1 and 3.5.2-1A2, (Unit 1), 3.5.2-1B1, 3.5.2-132 and 3.5.2-133 (Unit 2), l and 3.5.2-1C1, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump s operation and on Figures 3.5.2-2A1,3.5.2-2A2 (Unit 1), 3.5.2-2B l (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. The minimum shutdown margin required by specification 3.5.2.1 shall be maintained at all tires.

d. Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 I (Unit 1) 3.5.2-131, 3.5.2-132, and 3.5.2-133 (Unic 2), and 3.5.2-1C1, 3.5.2-1C2, 3.5.2-1C3 (Unit 3), unless the following requirements are met.

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

3.5-8

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

0 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 maintained within the envelope defined by Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-33, and 3.5.2-3C. If the imbalance is not within the envelope defined by Figure 3.5.2-3 A1, 3.5.2-3A2, 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.

3.5.2.7 The control red drive patch panels shall be locked at all times with limited access to be authorized by the =anager.

3.5-9

Bases The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, l 3.5.2-3B, and 3.5.2-3C is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4) such that the maximum clad temperature will not exceed the Final Acceptance Criteria. Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary. Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distri-bution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.** Conservatism is introduced by application of:

a. Nuclear uncertainty factors

b. Thermal calibration *

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors The 25% 1 5% overlap between successive control rod groups is allowed sicce 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 Xenon transient override 8 APSR (axial power shaping bank)

The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consis-cent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1) . The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.5% ak/k (Unit 1) or 0.65% ak/k (Units 2 and 3) at raced pcwer. These values have been shown to be safe by the safety analysis (2,3,4) of the hypothetical rod ej ection accident. A maximum single inserted control rod worth of 1.0% ak/k is allcwed by the rod positions limits at hot zero power.

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

• • Actual 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 i

l 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.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using the definition of quadrant power tile 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 normal operation when including the effects of potential fuel densification.

The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally will be performed in the process computer. The two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.

Allowance is provided for withdrawal limits and reactor power imbalance linics to be exceeded for a period of two hours withour specification violation. Acceptable rod positions and imbalance must be achieved within the two-hour eine period or appropriate action such as a reduction of power taken.

Operating restrictions are included in Technical Specification 3.5.2.5d to prevent excessive power peaking by transient xenon. The xenon reactivity must be beyond the "undershoot" region and asymptotically approaching its equilibrium value at the power level cutoff.

REFERENCES FSAR, Section 3.2.2.1.2 2

FSAR, Section 14.2.2.2 FSAR, SUPPLDiENT 9 B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1) 3AW-1396 (UNIT 2) 3AW-1400 (UNIT 3) 3.5-11

RCD FOSITICN LIMITS FCR 4 PUS CPERATICN APFLICABLE TO THE -

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(196.92) 0 - * '

• OPENATION IN THIS REGION RESTRICTED 15 NOT ALLOWED REGION 50 -

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$ 50 - . (155.50) m 2 MINIMUM SHUT 001N 7 40 -

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