NRC FORM 195 U.S. NUCLEAR REGULATORY COMI 'ION DOCKET NUMBER

(2-76) -z 9 I FILE NUMBER NIRC DISTRIBUTION FO 'PART 50 DOCKET MATERIAL

FROM: DATE OF DOCUMENT MakRugcher Duke Power Company 7-21-76

Charlotte, NC

E b Powel1 DATE RECEIVED 7-26-76

ETTER XRNOTORIZED PROP INPUT FORM NUMBER OF COPIES RECEIVED

JMORIGINAL OUNCLASSIFIED .IOPY XXXXXXX .3 signed

DESCRIPTION ENCLOSURE

Ltr requesting withholding of certain sections Proposed.Amdt to 0L/Chanje to Tech Specs: pursuant to 10 CFR 2.790 ..... trans the consisting of the following:

following....notarized 7-21-76.

1. Proposed Tech Specs concerning

Cycle #2 operation within applicable fuel

design-& performance criteria....... 2. Prop version of BAW-1432........

3. Oconee Unit 3, Cycle 2-Reload Report (BAW-1432 dtd June 1976).....non prop version....

PLANT NAME: Oconee #3

(40 sets enet rec d)

Denotes without encl #2 . 6'

SAFETY FOR ACTION/ NFORMAlION ENVTRO 7-27-76 ehf _ ASSIGNED AD: ASSIGNED AD: _ BRANCH CHIEF: 1- BRANCH CHTEF

.PROJECT MANAGER: Z L PROJECT MANAGER: LIC. ASST.: LIC. ASST:

7 _cV__3_31+cs 4 . A 4 . . 0RN CHTE INTERNAL DISTRIBUTION Ma -4t4. 54.A + 0f I .

R SYSTEMS SAFETY PTANT SYSTEMS SITE SAFETY & HEINEMAN ENVIRO ANALYSIS

ISCHRO R E BENAROYA DENTON & MULLER

/. OELD w LAINAS

GOSSICK & STAFF ENGINEERING IPPOLITO ENVIRO TECH MIPC MACCARRY KIRKWOOD ERNST

CASE KNIGHT BALLARD

HANAUER SIHWEIL OPERATING REACTORS SPANGLER HARLESS. PAWLICKI STELL0

STTE TEH,

PROJECT MANAGEMENT REACTOR SAFETY OPERATING JECH. GAMMILL BOYD ROSS EISENHUT STEPP P. COLLINS NOVAK SHAO At Z HULMAN HOUSTON ROSZTOCZY BAER 3 PETERSON CHECK BUTLER AJ4 SITE ANALYSIS MELTZ /____ GRIMES 7 VOLMER

HELTEMES AT & I BUNCH

_SKOVHOLT SALTZMAN J. COLLTNS RUTBERG _KREGER

EXTERNAL DISTRIBUTION CONTROL NUMBER

NAT LAB: BROOKHAVEN NAT LAB TIC: REG. VIE ULRIKSON(ORNL)

NSI_ . LA PDR

ASLB: CONSULTANTS q_?t /RACRS 19#CYS EN'76,

unJC FORM 1995 to.7ml

DUKE POWER COMPANY POWER BUILDING

422 SOUTH CHURCH SfREET, CKARLOTTE, N. C. 28242

E. D. POWELL

ASSISTANT VICE PRESIDENT TELEPHONE: AREA 7Q4

PRODUCTION AND TRANSMISSION 374-4682

July 21, 1976 o.

Mr. Benard C. Rusche, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Re: Oconee Unit 3 Docket No. 50-287

Dear Mr. Rusche:

Pursuant to 10CFR50, §50.90, please find attached proposed changes to the Oconee Nuclear Station Technical Specifications. The purpose of these revisions is to assure operation of the Oconee Unit 3, Cycle 2 core within applicable fuel design and performance criteria. The proposed changes are shown in Attachment 1 as replacement pages for the Oconee Nuclear Station Technical Specifications:

2.1 SAFETY LIMITS, REACTOR CORE 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 3.5.2 Control Rod Group and Power Distribution Limit

Also attached is Babcock and Wilcox Report BAW-1432, "Oconee Unit 3, Cycle 2 Reload Report." This report includes a summary of Cycle 1 operating parameters and contains the safety analyses supporting the operation of Oconee 3, Cycle 2 core at rated power in accordance with the Technical Specifications provided.

Certain data contained on pages 4-5 and 4-6 of Report BAW-1432 have been obtained from independently conducted studies investigating cladding creep collapse. Attachment 2 provides an affidavit of Mr. Kenneth E. Suhrke, Manager of Licensing in the Nuclear Power Generation Division of Babcock and Wilcox, supporting a request that this information be considered proprietary and withheld from public disclosure pursuant to 10CFR2, 92.790.

Very truly yours,

E. D. Powell

MST:vr Attachment(s)

Mr. Benard C. Rusche Page 2 July 21, 1976

E. D. POWELL, being duly sworn, states that he is Assistant Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this request for amendment of the Oconee Nuclear Station Facility Operating Licenses DPR-38, DPR-47, and DPR-55; and that all statements and matters set forth therein are true and correct to the best of his knowledge.

E. D. Powell, Assistant Vice President

ATTEST:

#John C. Goodman, Jr. Assistant Secretary

Subscribed and sworn to before me this 20th day of July, 1976.

Notary Public

My Commission Expires:

c2C- ,7

ATTACHMENT 1

PROPOSED TECHNICAL SPECIFICATION REVISIONS

TO SUPPORT

OCONEE 3, CYCLE 3 OPERATION

Jy .76

July 19, 1976

Bases - Unit 3

The safety limits presented for Oconee.Unit 3 have been generated using BAW-2 critical heat flux correlation(1) and the Reactor Coolant System flow rate of 107.6 percent of the design flow (131.32 x 106 lbs/hr for four-pump operation). The flow rate utilized is conservative compared to the actual measured flow rate..(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 "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 BAW-2 correlation(l). 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 setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1C represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 141.3 x 106 lbs/hr.). This curve is based on the following nuclear power peaking factors with potential fuel densification and fuel rod bowing effects: N N N

F = 2.67; FAH = 1.78; Fz = 1.50. The design peaking

combination results in a more conservative DNBR than any other power shape that exists during normal operation.

The curves of Figure 2.1-2C are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing.

1. The 1.30 DNBR limit produced by a nuclear peaking factor of Fq = 2.67 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than a 1.30 DNBR.

2 .1-3c

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 3.

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 and 3 of Figure 2.1-2C correspond to the expected minimum flow rates with four pumps, three pumps and one pump in each loop, respectively.

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

The maximum thermal power for three-pump operation is 86.4 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 = 79.9 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.

For each curve of Figure 2.1-3C a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor coolant pump situation. The 1.30 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 left of the four-pump curve will be above and to the left of the other curves.

References

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

(2) Oconee 3, Cycle 2 - Reload Report - BAW-1432, Juno 1976.

2.1-3d

2600

2400

2200

2000 0

0

1800

1600 I I 1 I 560 580 600 620 640 660

Reactor Outlet Temperature, F

CORE PROTECTION SAFETY LIMITS UNIT 3

OCONEE NUCLEAR STATION

2.1-6 Figure 2.1-1C

120

(-1, 112) (30 112)

Acceptable

(-40, 100) 100 4-Pump coo, Operation 43, 100)

(-2 86.4) (30.8. 86.4)

80 (-40, 74.4) Acceptable (43, 74.4)

3 & 4 Pump Operation

(-21 58.9) -- 60 (30.8, 58.9).

(-40, 46.9) Acceptable 43, 46.9) 2, 3 & 4 Pump Operation 40

20

II I I I I -60 -40 -20 20 40 60

Reactor Power Imbalance, %

Curve Reactor Coolant Flow (lb/h) 1 141.3 x 106 2 105.6 x 106 3 69.3 x 106

CORE PROTECTION SAFETY LIMITS UNIT 3

2.1-9 OCONEE NUCLEAR STATION

Figure 2.1-2C

2600

2400

2200

2000

0.

1800

16001II 560 580 600 620 640 660

Reactor Outlet Temperature, F

Reactor coolant flow Curve (1bs/h) Power Pumps operating (type of limit)

1 141.3 . 106 (100%) 112% Four pumps (DNBR limit)

2 105.6 x 106 (74.7%) 86.4% Three pumps (DNBR limit)

3 69.3 x106 (49.0%) 58.9% One pump in each loop (quality limit)

CORE PROTECTION SAFETY LIMITS 2.1-12

o OCONEE NUCLEAR STATION (U9 EPDO ER Figure 2.1-3C

During normal plantWeration with all reactor coola 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 actuated could be 112%, which is more conservative than the value used in the safety analysis. (4)

Overpower Trip Based on Flow and Imbalance

The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant 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 by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum 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 situtations of Table 2.3-1A are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 105.5% and reactor flow rate is 100%, or flow rate is 94.8% and power level is 100%.

2. Trip would occur when three reactor coolant pumps are operating if power is 78.8% and reactor flow rate is 74.7% or flow rate is 71.1% and power level is 75%.

3. Trip would occur when two reactor coolant pumps are operating in a single loop if power is 51.7% and the operating loop flow rate is 54.5% or flow rate is 48.5% 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 51.7% and reactor flow rate is 49.0% or flow rate is 46.4% and the power level is 49%.

The flux-to-flow ratios account for the maximum calibration gand instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

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 from being exceeded. These thermal limits are either power peaking kw/ft limits or DNBR limits. The reactor power imbalance (power in

the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio such that the boundaries of

Figure 2.3-2A - Unit 1 are produced. The power-to-flow ratio reduces the powerj 2.3-2B - Unit 2 2.3-2C - Unit 3

2.3-2

level trip and associated reactor power/reactor power-imbalance boundaries by 1.055%-Unit 1 for a 1% flow reduction,

1.07% - Unit .2 1.07% - Unit 3

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

Pump Monitors

The pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the 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 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 (1800) psig and variable low pressure (11.1.4 T -4706) trip out

(1800) psig (10.79 T -4539) (1800) psig (10.79 T ~- 4 5 3 9 )

setpoints shown in Figure 2.3-1A have been established to mainta n 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 out -4746)

(10.79 T -4579) (10.79 Tout -4579)

Coolant Outlet Temperature

The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1A has been established to prevent excessive core coolant

2.3-1B 2.3-1C

temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 6200

Reactor Building Pressure

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

2.3-3

2400

P = 2355 psig T = 619F,

2300

2200

Acceptable

w 2100 Operation

4

O 0

0

2000

I I II

U

0)0

Unacceptable

Qo Operation 1900

1800 P =1800 psig

(587.5)

540 560 580 600 620 640

Reactor Outlet Temperature, F

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS

2.3-7 UNIT 3

OCONEE NUCLEAR STATION

Figure 2.3-1C

Power level, %

120

(-11, 107) (18, 107)

Four Pump T3

Setpoint -- 100

(-28, 93)

(30, 90)

Three Pump Setpoin An

(11, 79.9) (18, 79.9)

(-28, 65.9)

(30, 62.9)

*Two Pump 60

Setpoints

-11, 52.4) (18, 52.4)

(-28, 38.4).- 40

(30, 35.4)

o a - -20 C o

-. 44

C'~l 0

I I I 0 I.I -60 -40 -20 0 20 40 60

Power Imbalance, %

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS

2.3-10 UNIT 3

ou3own OCONEE NUCLEAR STATION

Figure 2.3-2C

Table 2.3-1C

Unit 3

Reactor Protective Sy stem Trip Setting Limits

Two Reactor One Reactor Four Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pn:uIps Operat ing in A Operating in Operating Operating Single Loop Each Loop (Operattog Power (Operating Power (Operating Power (Operating Shutdown

RPS Segment -100% Rated) -75% Rated) --46% Rated) -49% Rated) Bypass

1. Nuclear Power Max. 105.5 105.5 105.5 105.5 5.0(3)

(% Rated) .

2. Nuclear Power Max. Based 1.07 times flow 1.07 times flow 0.961 times flow 1.07 times flow. Bypaused

on Flow (2) and Imbalance, minus reduction minus reduction minus reduction minus reduction (% Rated) due to imbalance due to imbalance due to imbalance due to imbalance

3. Nuclear Ponar Max. Based NA NA 55% (5) (6) 55% Bypassed

on Pump Monitors, (% Rated)

4. High Reactor Coolant 2355 2355 2355 2355 1720(4)

System Pressure, psig, Max.

5. Low Reactor Coolant 1800 1800 1800 1800 Bypassed

System Pressure, psig, Min.

6. Variable Low Reactor (10.79 T -4539)(1) (10.79 T 4539)(1) (10.79 T -4539)(1) (10.79 T u-4539)(1) Bypassed

Coolant System Pressure psig, Min.

7. Reactor Coolant Temp. 619 619 619 (6) 619 619

F., Max.

8. High Reactor Building 4 4 4 4 4

Pressure, psig, Max.

(1) Tou is in degrees Fahrenheit (oF). (5) Reactor power level trip set point produced Tt by pump contact monitor reset to 55.0%.

(2) Reactor Coolant System Flow, %. (6) Specification 3.1.8 applies. Trip one of the

(3) Administratively controlled reduction set two protection channels receiving outlet

only during reactor shutdown. temperature information from sensors in the idle loop.

(4) Automatically set when other segments of the RiS are bypassed.

g. If within one (1) hour of determination of an inoperable rod, it is not determined that a 1%Ak/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 hours and exercised 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 combination.

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 worths of_ single inserted controlrods during criticality are limited by the restrictions of Specification 3.1.1-5.and the_ control rod position limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Power Tilt

a. Except for physics tests, if the maximum positive quadrant power tilt exceeds +3.41% Unit 1, either the quadrant power tilt shall

3.41% Unit 2 3.41% Unit 3

be reduced to less than +3.41% Unit 1 within two hours or the 3.41% Unit 2 3.41% Unit 3

following actions shall be taken:

(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced below the power level cutoff (as identified in specification 3.5.2.5) and further reduced by 2% of full power for each 1% tilt in excess of 3.41% Unit 1.

3.41% Unit 2 3.41% Unit 3 1

(2) If less than four reactor coolant pumps are in operation, the allowable thermal power for the reactor coolant pump combination shall be reduced by 2% of full power for each 1% tilt.

3.5-7

(3) Except as provided in specification 3.5.2.4.b, the reactor shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 3.41% Unit 1 within 24 hours. 3.41% Unit 2 3.41% Unit 3

b. If the quadrant tilt exceeds +3.41% Unit 1 and there is simultaneous 3.41% Unit 2 3.41% Unit 3

indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60% of the thermal power allowable for the reactor coolant pump combination.

c. Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, 9.44% Unit 2 9.44% Unit 3

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

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

a. Technical Specification 3.1.3.5 does not prohibit 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% + 5% between two sequential groups, except for physics tests.

c. Except for physics tests or exercising control rods, the control rod withdrawal limits are specified on Figures 3.5.2-lAl and 3.5.2-1A2, (Unit 1), 3.5.2-ll, 3.5.2-1B2 and 3.5.2-1B3 (Unit 2), and 3.5.2-lCl, 3.5.2-1C2, and 3.5.2-1C3 (Unit 3) for four pump operation and on Figures 3.5.2-2A1, 3.5.2-2A2 (Unit 1), 3.5.2-2B1, 3.5.2-2B2 and 3.5.2-2B3 (Unit 2), and 3.5.2-2C1, 3.5.2-2C2, and 3.5.2-2C3 (Unit 3) for three or two pump

3.5-8

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 position shall then be attained within two hours. The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.

d. Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-lAl, 3.5.2-1A2 (Unit 1), 3.5.2-lB1, 3.5.2-1B2, and 3.5.2-1B3 (Unit 2), and 3.5.2-1i1, 3.5.2-1C2, 3.5.2-103 (Unit 3), unless the following requirements are met.

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

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

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, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2, and 3.5.2-3C3. If the imbalance is not within the envelope defined by these figures, 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 rod drive patch panels shall be locked at all times with limited access to be authorized by the manager.

3.5-9

Bases

The power-imbalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-3B3, 3.5.2-3C1, 3.5.2-3C2 and 3.5.2-3C3 is j 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 by application of:

a. Nuclear uncertainty factors b. Thermal calibration c. Fuel densification effects d. Hot rod manufacturing tolerance factors

The 25% + 5% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function

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, consistent 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 rated power. These values have been shown to be safe by the safety analysis (2,3,4) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod positions limits at hot zero power. A single inserted control rod worth of 1.0% Ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, 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

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 with consideration of potential effects of rod bowing and fuel densification to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 5.10% for Unit 1. The limits shown in Specification 3.5.2.4

5.10% for Unit 2 5.10% for Unit 3

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

The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4 and 3.5.2.6, respectively, normally 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 limits to be exceeded for a period of two hours without specification violation. Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.

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

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

REFERENCES

1FSAR, Section 3.2.2.1.2

2FSAR, Section 14.2.2.2

3FSAR, SUPPLEMENT 9

4B&W FUEL DENSIFICATION REPORT

BAW-1409 (UNIT 1)

BAW-1396 (UNIT 2)

BAW-1400 (UNIT 3)

3.5-11

100 - 170, 102 202.5, 102

90 170, 91 202.5, 91 Level 90 - Cutoff 161, 85

206.7, 85 80 - Restricted Restricted

Region Region

S 70 25.8, 65 300, 64 0 60129, 62

S60

50 0 50

Permissible 40 - Operating

o Region 30

20

100, 0

0 I I I I I I I I I I I 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100 1 I I I I I I I I

Group 5 Group 7

0 25 50 75 100 I I I I I

Group 6

ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 TO 115 (+ 10) EFPD UNIT 3

3.5-16 DUKEPOWER OCONEE NUCLEAR STATION

Figure 3.5.2-1C1

115 ,102 170, 10 209.4, 102 100- Operation in this

Region is Not 90- Allowed 170, 9 209.4, 91 Power

150, 858

80- Shutdown Margin Restricted

70- Limit Region 70- Region Region

129, 62 300. 64,

6C- 225. 8, 65

50- 18, 50 Permissible

S 4C- Operating Region

3C

2C 0, 15

10

0 0 'i I I I I I I I I I I I I I

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100 I I I iI I I I I

Group 5 Group 7

0 25 50 75 100 II I I I

Group 6

ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 115 (+ 10) EFPD TO 226 (+ 10) EFPD UNIT 3

DUKEPOWE OCONEE NUCLEAR STATION

3.5-16a Figure 3.5.2-1C2

Operation in this 140 102 255 102

Region is Not Allowed Power Level

90 - Cutoff 240, 91

80 - Restricted

Region Shutdown 70

0 70 Margin Limit 96, 67

604-4 0

50 - 54, 50

40 -Permissible 0 Operating

Region 30

20

10

0, O 0 O 20 40 66 80 100 120 140 160 180 200 220 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100 t I I I I I I I I

Group 5 Group 7

0 25 50 75 100 I I I I I

Group 6

ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 226 (+ 10) EFPD UNIT 3

urEPO OCONEE NUCLEAR STATION

3.5-17 Figure 3.5.2-1C3

Restricted Region 114 102 1 4,102 2 ,102 Restricted a 100 - for 2 and 3 Pump Region for

0 Operation 3 Pump cc 90 - Operation

26,83 300,82 o 80 80 129,79

70 -d

60

0

50 Permissible Operating Region

40 0

Co 30 '4-4 0

20

10 00

20 40 60 80 100 120 140 160 180 210 12 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100

L 1 I I I I I I I I

Group 5 Group 7

0 25 50 75 100 II I I I

Group 6

ROD POSITION LIMITS FOR TWOAND THREE-PUMP OPERATION FROM 0 TO 115 (+ 10) EFPD UNIT 3

3.5-20 DUKEPOWE OCONEE NUCLEAR STATION

Figure 3.5.2-2C1

Operation in this 115 102 46,102 218 02 Restricted 100- Region is Not Region for

Allowed 3 Pump 90 9Operation 90 4

o 80- Shutdown 226,8_i80 oo 129,79 300,82 Margin--4 Limit

70

60- a Permissible Operating Region

0 50

50- 8,50'

o'o

40 30 1

00

15 o 10

P.4

0 ,g I I 1 I I I I I I I ) 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100I II I I I I I|

Group 5 Group 7

0 25 50 75 100 I I I I

Group 6

ROD POSITION LIMITS FOR TWOAND THREE-PUMP OPERATION FROM 115 (+ 10) TO 226 (+ 10) EFPD UNIT 3

3.5-20a OCONEE NUCLEAR STATION

Figure 3.5.2-2C2

\ 140,102 33,102 a 100- Operation in this Restricted Region . Region is Not for 3 Pump

Allowed Operation 90- 226,87

o 80- Shutdown Margin 20,85 Limit A

: 70

60- Permissible Operating Region

0

a 50- 54,5

400

30. 3, '4-4 0 30

20.

Q) 15 105 Restricted for 2 and

o4 3 Pump Operation

0,01

( 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

Rod Index, % withdrawn 0 25 50 75 100 0 25 50 75 100 I I I I I I I I I I

Group 5 Group 7

0 25 50 75 100 I I I II

Group 6

ROD POSITION LIMITS FOR TWOAND THREE-PUMP OPERATION AFTER 226 (+ 10) EFPD UNIT 3

3.5-20b OCONEE NUCLEAR STATION

Figure 3.5.2-2C3

Power, % of 2568 MWt

Restr cted Region -110

-7.96,102 16.83,102

-100

-7.78,91 90 15.22,91

-9.67,8 )20.12,86

-80

-70

-23.64,64 .60

Permissible Operating -50 Region

40

-30

-20

-10

I I I I I

-50 -40 -30 -20 -10 0 10 20 30 40 50

Axial Power Imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 115 (+ 10) EFPD UNIT 3

3.5-23 U OCONEE NUCLEAR STATION

Figure 3.5.2-3C1

Power, % of 2568 MWt

Restricted Region

-16.47,102 100 16.83,102

-15.72,91 -90 16.83,91

-16.94,8 17.00,85

.80

70

-23.64,6

60

Permissible Operating 50 Region

40

30

20

10

-50 -40 -30 -20 -10 0 10 20 30 40 50

Axial Power Imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 115 (+) EFPD to 226 (+ 10) EFPD

3.5-23a NIT

DUKEPOWER OCONEE NUCLEAR STATION

Figure 3.5.2-3C2

Power, % of 2568 MWt Restricted Region

-25.5,102 .100 .71,102

-27.86,91 -.90 23.96,91

--80

-70

--60

Permissible Operating 50 Region

-40

--30

--20

__10

I I I I I I I I I

-50 -40 -30 -20 -10 0 10 20 30 40 50

Axial Power Imbalance, %

OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 226 (+ 10) EFPD UNIT 3

3.5-23b KEPOWER OCONEE NUCLEAR STATION Figure 3.5.2-3C3

20 1 g I IIII

18 14-4

0

Generic FAG BAW 10103 **-

Unit 1, Batch 4 -

12 12

10 0 2 4 6 8 10 1

Axial Location of Peak Power From Bottom of Core, ft

LOCA LIMITED MAXIMUM ALLOWABLE 3.5-24LINEAR HEAT RATE

UNIT 1

DEPWROCONEE NUCLEAR STATION F

Generigur 3.5.2-4010

Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY,

Item Test Frequency

(1) : 1. Control Rod Movement ( Movement of Each Rod Bi-Weekly

2. Pressurizer Safety Valves Setpoint 50% Annually

3. Main Steam Safety Valves Setpoint 25% Annually

4. Refueling System Interlocks Functional Prior to Refueling

5. Main Steam Stop Valves(1) Movement of Each Stop Monthly Valve

6. Reactor Coolant System(2) Evaluate Daily Leakage

7. Condenser Cooling Water Functional Annually System Gravity Flow Test

8. High Pressure Service Functional Monthly Water Pumps and Power Supplies

9. Spent Fuel Cooling System Functional Prior to Refueling

10. Hydraulic Snubbers on Visual Inspection Annually Safety-Related Systems

11. High Pressure and Low(3 ) Vent Pump Casings Monthly and Prior Pressure Injection System to Testing

12. Reactor Coolant System Flow Validate Flow to be Once Per Fuel at least: Cycle

6 Unit 1 141.30 x 106 lb/hr

6 Unit 2 141.30 x 10 6 lb/hr Unit 3 141.30 x 10 lb/hr

(1) Applicable only when the reactor is critical

(2) Applicable only when the reactor coolant is above 2000F and at a steadystate temperature and pressure.

(3) Operating pumps excluded.

4.1-9