ML15112A835

Amends 66,66 & 63 to Licenses DPR-38,DPR-47 & DPR-55 Re Proposed Changes Re Control Rod Operability & Support of Unit 2 Full Rated Pwr During Cycle 4 After Core Reload & Removal of Orifice Rod Assemblies from Core

ML15112A835
Person / Time
Site:	Oconee
Issue date:	12/18/1978
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML15112A836	List: ML15112A835 ML15112A837 ML15112A839 ML15112A840
References
NUDOCS 7812280247
Download: ML15112A835 (36)

Text

.<

,o NUCLEARUNITED STATES NUCLEAR REGULATORY COMMISSION SEWASHINGTON,

0. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-38

1. The Nuclear Regulatory Commission (the Conmission) has found that:

A. The application for amendment by Duke Power Company (the licensee) dated September 18, 1978, as supplemented September 25 and November 1, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

781228 0;1

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2. Accordingly, the license is amended by changes to the Technical

-Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility Operating License No.

DPR-38 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 66 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/9.iRobert W. R d, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 15, 1978

.e 1 Rec O UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-47

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

A. The application for amendment by Duke Power Company (the licensee) dated September 18, 1978, as supplemented September 25 and November 1, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility Operating License No.

DPR-47 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 66 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION SRobert W. ReChief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: December 15, 1978

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. 0. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO.3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. DPR-55

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

A. The application for amendment by Duke Power Company (the licensee) dated September 18, 1978, as supplemented September 25 and November 1, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 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. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility Operating License No.

DPR-55 is hereby amended to read as follows:

3.8 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION GRobert W.

d, ief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: December 15, 1978

ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 66 TO DPR-38 AMENDMENT NO. 66 TO DPR-47 AMENDMENT NO. 63 TO DPR-55 Revise Appendix A as follows:

Remove the following pages and insert the revised identically numbered pages.

2.1-3a & 2.1-3b 2.1-8 (Figure 2.1-28) 2.3-9 (Figure 2.3-2B) 3.2-1 & 3.2-2 3.5 3.5-11 3.5-11a & 3.5-11b 3.5-llc* (Table 3.5-1) 3.5-14 (Figure 3.5.2-lBl) 3.5-14a (Figure 3.5.2-182) 3.5-15 3.5-19 (Figure 3.5.2-2B1) 3.5-19a (Figure 3.5.2-282) 3.5-19b 3.5-22 (Figure 3.5.2-381) 3,5-22a (Figure 3.5.2-382) 3.5-22b 3.5-23f (Figure 3.5.2-481) 3.5-23g (Figure 3.5.2-482) 3.5-23h 4.1,-9 Changes on the revised pages are identified by marginal lines. Page 3.5-5 is unchanged and is included for convenience only.

• New Page

Bases - Unit 2 The safety limits presented fo f19conee Unit 2 have been generated using BAW-2 critical heat flux correlation and the Reactor Coolant System flow rate of 106.5 percent of the design flow (design flow is 352,000 gpm for four-pump operation).

The f 2 rate utilized is conservative compared to the actual measured flow rate 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 fail ure. 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 (1).

The BAW-2 correlation has been developed to predict DNB and the loca tion 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 confi dence 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 2.1-1B represents the conditions at which a mini mum DNBR of 1.30 is predictecd for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 374,880 gpm).

This curve is based on the following nuclear power peak ing factors with potential fuel densification and fudl rod bowing effects:

F qN = 2.565; FHN = 1.71 FzN = 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 2.1-23 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:

Amendments Nos. 66, 66 & 63 2.1-3a

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 19.8 kw/ft for Unit 2.

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

The curve of Figure 2.1-13 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.

The maximum thermal power for three-pump operation is 85.3 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.055 =

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.

For each curve of Figure 2.1-3B, 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 particu lar 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 2, Cycle 3 - Reload Report -

BAW-1452, April, 1977.

(3) Oconee 2, Cycle 4 -

Reload Report -

BAW-1491, August, 1978.

2.1-3b Amendments Nos. 66, 66 & 63

5 OF RATED THERMAL POWER ON8R LIMIT

-120

(-33.60,112) 112.0 00 (10.84.112.0)/F L MI

-110 KWFT LIMIT

,//

I ACCEPTABLE 4 PUMP 10 KW/FTOPERATION LIMIT

-- 100 (31.0,100.0)

( -52.0,95.0)

-- 090 85.31 C

IACCEPTABLE 3 80 3 &.4 PUMP OPERATION 73.31 68.31 70 58.20 60 60 ACCEPTABLE 2.3 &4 PUMP 0

OPERATION 4

41.20 40 I

I I

I I

30 I

I..

o

-. 20

--I10 I

II I 1 I II II 11I

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 Axial Power Imoalance, %

CURVE REACTOR COOLANT FLOW (GPM) 1 374.880 2

280.035 3

183,690 CORE PROTECTION SAFETY LIMITS NtT 7 (urEP OCONEE NUCLEAR STATION Figure 2.1-2B Amendments Nos.

66, 66 & 63 2.1-8

THERNAL POWER LEVEL, 1 UNACCEPTABLE OPERATION 110 2.463 ACCEPTABLE 100 4 PUMP OPERATION I

1 10

( -42.0,80) 78.81 o9

/,04 ACCEPTABLE OPERATION (20.0,88.31) 42.0,53. 31I c.7c ACCEPTAMLEW 50 PUMP

.. 1 (20.0,41.20)

O NPERATIONN L

"30 (Fu42.0,26.21)

I

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 Power imbalance, %

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UN-IT 2 KPOEOCONEE NUCLEAR STATION Figure 2.3-2B Amendments Nos. 66, 66 &63 2.3-9

3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the high pressure injection and the chemical addition systems.

Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.

Specification The reactor shall not be critical unless the following conditions are met:

3.2.1 Two high pressure injection pumps per unit are operable except as specified in 3.3.

3.2.2 One source per unit of concentrated soluble boric acid in addition to the borated water storage tank is available and operable.

This source will be the concentrated boric acid storage tank contain ing at least the equivalent of 995 ft3 of 8700 ppm boron as boric acid solution with a temperature at least 10*F above the crystalliza tion temperature. System piping and valves necessary to establish a flow path from the tank to the high pressure injection system shall be operable and shall have the same temperature requirement as the concentrated boric acid storage tank.

At least one channel of heat tracing capable of meeting the above temperature requirement shall be in operation. One associated boric acid pump shall be operable.

If the concentrated boric acid storage tank with its associated flow path is unavailable, but the borated water storage tank is available and operable, the concentrated boric acid storage tank shall be re stored to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be placed in a hot shutdown condition and be borated to a shutdown margin equivalent to 1% dk/k at 2000F within the next twelve hours; if the concentrated boric acid storage tank has not been restored to opera bility within the next 7 days the reactor shall be placed in a cold shutdown condition within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

If the concentrated boric acid storage tank is available but the borated water storage tank is neither available nor operable, the borated water storage tank shall be restored to operability within one hour or the reactor shall be placed in a hot shutdown condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in a cold shutdown condition within an addition al 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Amendments Nos.

66, 66 & 63 3.2-1

0 Bases The high pressure injection system and chemical addition system provide con trol of the reactor coolant system boron concentration.(1) This is normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the concentrated boric acid storage tank. An alternate method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank.(2)

The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient to borate thereactor coolant system to a 1% dk/k subcritical margin at cold conditions (70aF) with the maximum worth stuck rod and no credit for zenon at the worst time in core life.

The current cycles for each unit, Oconee 1 Cycle 5, Oconee 2 Cycle 4, and Oconee 3 Cycle 4 were analyzed with the most limiting case selected as the basis for all three units. Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload.

A minimum of 995 ft3 of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1800 ppm boric acid in the borated water storage tank (3) will satisfy the requirements. The volume requirements in clude a 10% margin and in addition allow for a deviation of 10 EFPD in the cycle length.

The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The required amount of boric acid can be added in several ways.

Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would require approximately 12.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> to inject the required boron.

An alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.

The required boric acid can be injected in less than six hours using only one of the makeup pumps.

The concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.

For this reason and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 10aF above the crystallization temperature for the concentration present.

The boric acid concentration of 8,700 ppm in the concentrated boric acid storage tank cor responds to a crystallization temperature of 770F and therefore a temperature requirement of 870F. Once in the high pressure injection system, the concen trate is sufficiently well mixed and diluted so that normal system tempera tures assure boric acid solubility.

REFERENCES (1) FSAR, Section_9.1; 9.2 (2) FSAR, Figure 6.2 (3) Technical Specification 3.3 Amendments Nos. 66, 66 & 63 3.2-2

TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (Cont'd)

(A)

Minimum (B)

(C)

Operable Minimum Operator Action if Condi:ions Analog Degree Of Of Column A and B Functional Unit Channels Redundancy Cannot 38 Met

b. Manual Pushbutton 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (e)

15. Turbine Stop Valves 2

1 Bring to hot shutdown within Closure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (f)

(a) For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours.

(b) When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

(c) When 1 of 2 intermediate range instrument channels is greater than 10-10 amps, hot shutdown is not required.

(d) Single loop operation at power (after testing and approval by the AEC/DOL) is not permitted unless the operating channels are the two receiving Reactor Coolant Temperature from operating loop.

(e) If minimum conditions are not met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after hot shutdown, the unit shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(f) One operable channel with zero minimum degree of redundancy is allowed for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before going to the hot shutdown condition.

3.5-5

0 0

3.5.2 Control Rod Grouo and Power Distribution Limits Applicability This specification applies to 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 from a hypothetical control rod ejec tion, and to assure core subcriticality after a reactor trip.

Specification 3.5.2.1 Shutdown dargin

a.

The available shutdown margin shall be greater than 1% Ak/k with the highest worth control rod fully withdrawn.

b.

If the shutdown margin is less than 1% Ak/k, then within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and continue boration until the required shutdown mar gin is restored.

The requirements of specification 3.5.2.5.c shall be met.

3.5.2.2 Movable Control Assemblies

a.

All control (safety and regulating) rods shall be operable and positioned within nine (9) inches of their group average height.

b.

A control rod shall be declared inoperable if any of the follow ing conditions exist for that rod:

1. The control rod cannot be moved due to excessive friction or mechanical interference, or cannot perform its intended trip function.

2.

The control rod cannot be located by either absolute or re lative position indication or by in or out limit lights.

3.

The control rod is misaligned with its group average by more than nine (9) inches.

4. The control rod does not meet the exercise requirements of Specification 4.1.

5.

The control rod does not meet the rod trip insertion times of Specification 4.7.1.

6. The control rod does not meet the rod program verification of Specification 4.7.2.

Amendments Nos. 66, 66 & 63 3.5-6

c. If a control rod is declared inoperable by being immovable due to excessive friction or mechanical interference or known to be un trippable then:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify that the shutdown margin requirement of Specification 3.5.Z.1 is satisfiedand

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> place the reactor in the hot standby condition.

d. If a control rod is declared inoperable due to causes other than addressed in 3.5.2.2.c above then:

1. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the rod to operable status,or

2. Continue power operation with the control rod declared in operable,and

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify the shutdown margin require meat of Specification 3.5.2.1 with an ad ditional allowance for the withdrawn worth of the inop erable rod,and

b. Either reactor thermal power shall be reduced to less than 60% of the allowable power for the reactor coolant pump combination within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow/imbalance, shall be reduced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of thermal power value allowable for the reactor coolant pump combination,or

c.

Position the remaining rods in the affected group such that the inoperable rod is maintained within allow able group average limits of Specification 3.5.2.2.a and. the withdrawal limits of Specification 3.5.2.5.c.

e. If more than one control rod is inoperable or misaligned, the reactor shall be shut down to the hot standby condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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

3.5.2.4 Quadrant Power Tilt

a. Except for physics tests, the maximum positive quadrant power tilt shall not exceed the Steady State Limit of Table 3.5-1 during power operation above 15% full power.

b. If the maximum positive quadrant power tilt exceeds the Steady State Limit but is less than or equal to the Transient Limit of Table 3.5-1, then:

Amendments Nos.

66, 66 & 63 3.5-7

1. Either the quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Steady State Limit, or

2. The reactor thermal powershall be reduced below the power level cutoff (as specified in Specification 3.5.2.5) and further reduced 2% thermal power for each 1% of quadrant power tilt in excess of the Steady State Limit, and the Nuclear Overpower Trip Setpoints, based on flux and flux/

flow imbalance, shall be reduced within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by 2%

thermal power for each 1% tilt in excess of the Steady State Limit.

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% for each 1% excess tilt.

c.

Quadrant power tilt shall be reduced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to within its Steady State Limit or

1.

The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the re actor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be re duced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

d.

If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1 and if there is a simultaneous indication of a misaligned control rod then:

1.

Reactor thermal power shall be reduced within 30 minutes at least Z% for each 1% of the quadrant power tilt in ex cess of the Steady State Limit.

2.

Either quadrant power tilt shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within its Transient Limit,or

3. The reactor thermal power shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the allowable power for the re actor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be re duced within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

e.

If the quadrant power tilt exceeds the Transient Limit but is less than the Maximum Limit of Table 3.5-1, due to causes other than simultaneous indication of a misaligned control rod then:

1.

Reactor thermal power shall be reduced within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than 60% of the.allowable power for the reactor coolant pump combination and the Nuclear Overpower Trip Setpoints, based on flux and flux/flow imbalance, shall be reduced within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 65.5% of the thermal power value allowable for the reactor coolant pump combination.

Amendments Nos. 66, 66 & 63 3.5-8

f. If the maximum positive quadrant power tilt exceeds the Maximum Limit of Table 3.5-1, the reactor shall be shut down within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the ther mal power and the Nuclear Overpower Trip Setpoints allowable for the reactor coolant pump combination are restricted by a reduc tion of 2%.of thermal power for each 1% tilt for the maximum tilt observed prior to shutdown.

g. Quadrant power tilt shall be monitored on a minimum frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% full 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.Z.2.

b. Except for physics tests, operating rod group overlap shall be 25%

+/- 5% between two sequential groups. If this limit is exceeded, cor rective measures shall be taken immediately to achieve an acceptable overlap.

Acceptable overlap shall be attained within two hours or the reactor shall be placed in a hot shutdown condition within a.

additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

Position limits are specified for regulating and axial power shap ing control rods.

Except for physics tests or exercising control rods, the regulating control rod insertion/withdrawal limits are specified on figures 3.5.2-lAl and 3.5.2-1A2 (Unit 1); 3.5.2-131, 3.5.2-132 and 3.5.2-133 (Unit 2); 3.5.2-11, 3.5.2-1C2 and 3.5.2 1C3 (Unit 3) for four pump operation, and on figures 3.5.2-2A1 and 3.S.2-2A2 (Unit 1); 3.5.2-231, 3.5.2-232 and 3.5.2-233 (Unit 2);

3.5.2-2C1, 3.5.2-2C2 and 3.5.2-2C3 (Unit 3) for two or three pump operation.

Also, excepting physics tests or exercising control rods, the axial power shaping control rod iasertion/withdrawal limits are specified on figures 3.5.2-4A1, and 3.5.2-4A2 (Unit 1);

3.5.2-4B1, 3.5.2-4B2, and 3.5.2-4B3 (Unit 2); 3.5.2-4C1, 3.5.2 4C2, and 3.5.2-403 (Unit 3).

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

An acceptable control rod position shall then be attained within two hours.

The minimum shutdown margin required by Specifi cation 3.5.2.1 shall be maintained at all times.

Amendments Nos, 66, 66 & 63 3.5-9

0l 3.5.2.6 Xenon Reactivity Except for physics tests, reactor power shall not be increased above the power level-cutoff shown in Figures 3.5.2-1A1, and 3.5.2-1A2 for Unit 1; Figures 3.5.2 131, 3.5.2-1B2, and 3.5.2-133 for Unit Z; and Figures 3.5.2-1i1, 3.5.2-1C2, and 3.5.2-1C3 for Unit 3 unless one of the following conditions is satisfied:

1. Xenon reactivity did not deviate more than 10 percent from the equi librium value for operation at steady state power.

2. Xenon reactivity deviated more than 10 percent but is now within 10 percent of the equilibrium value for operation at steady state rated power and has passed its final maximum or minimum peak during is ap proach to its equilibrium value for operation at the power level cut off.

3. Except for xenon free startup (when 2. applies), the reactor has oper ated within a range of 87 to 92 percent of rated thermal Power for a period exceeding I hours.

3.5.2.7 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.Z-231, 3.5.2-332, 3.5.2-383, 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.

3.5.2.3 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager or his designated alternate.

Amendments Nos.

66, 66 & 63 3.5-10

Bases Operation at power with an inoperable control rod is permitted within the limits provided. These limits assure that an acceptable power distribution is maintained and that the potential effects of rod misalignment on associ ated accident analyses are minimized. For a rod declared inoperable due to misalignment, the rod with the greatest misalignment shall be evaluated first.

Additionally, the position of the rod declared inoperable due to misalignment shall not be included in computing the average position of the group for deter mining the operability of rods with lesser misalignments. When a control rod is declared inoperable, boration may be initiated to achieve the existence of 1% ak/k hot shutdown margin.

The power-imbalance envelope defined in Figures 3.5.2-3Al and -3A2, 3.5.2-3M1,

-3B2 and -3B3, 3.5.2-3C1, -3C2 and -3C3 is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-5) such that the maximum clad temperature will not exceed the Final Acceptance Criteria. Cor rective 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 distribu tion 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 power spike factors (Units 1 and 2 only)

d.

Hot rod manufacturing tolerance factors

e. Fuel rod bowing power spike 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. Con trol rods are arranged in groups or banks defined as follows:

Growo Function 1

Safety 2

Safety 3

Safety 4

Safety 5

Regulating 6

Regulating 7

Xenon transient override 8

APSR (axial power shaping bank)

• • Actual operating limits depend on whether or not incore or excore.detectors are used and their respective instrument calibration errors. The method used to define the operating limits is defined in plant operating procedures.

Amendments Nos. 66, 66 & 63 3.5-11

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(l).

The rod position limits also ensure that in serted rod groups will not contain single rod worths greater than 0.65% Ak/k at rated power.

These values have been shown to be safe by the safety analysis (2,3,4,5) of hypothetical rod ejection accident.

A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position 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 0.65% 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.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 7.50% for Unit 1.

The limits shown in Specification 3.5.2.4 7.50% for Unit 2 7.50% 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. 7, 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 reductior% of power taken.

Operating restrictions are included in Techqical Specification 3.5.2.6 to prevent excessive power peaking by transient xenon.

For Unit 1, a 5%

peaking increase is applied to calculated peaks at equilibrium conditions for powers above the power level cutoff.

For Units 2 and 3, an 8% peaking increase is applied.

These values conservatively bound the peaking effects of transient xenon once the applicable requirement of 3.5.2.6 has been satisfied.

Amendments Nos, 66, 66 & 63 3.5-11a

0 0

REFERENCES 1FSAR, Section 3.2.2.1.2 2FSAR, Section 14.2.2.2 3FSAR, SUPPLEMENT 9 B&W FUEL DENSIFICATION REPORT BAW-1409 (UNIT 1)

BAW-1396 (UNIT 2)

BAW-1400 (UNIT 3) 5Oconee 1, Cycle 4 -

Reload Report -

BAW 1447, March 1977, Section 7.11 3.5-11b Amendments Nos.

66, 66 & 63

TABLE 3.5-1 Quadrant Power Tilt Limits Steady State Transient Maximum Limit Limit Limit Unit 1 5.00 9.44 20.0 Unit 2 5.00 9.44 20.0 Unit 3 5.00 9.44 20.0 Amendments Nos. 66, 66 & 63 3.5-11k

110 100 (108,102)

(171,102)

(206,102)

OPERATION IN THIS REGION IS NOT (171,92)

(206.92)

POWER LEVEL ALLOWED CUTOFF 80 -

(151.4,80)

(225.5,80) 70 SHUTDOWN RESTRICTED RESTRICTED SHUTOON REGION REGION CO MARGIN 60 LIMIT 5 (37.50)

(125.5,50 (251.4,50) 40 PERMISSIBLE OPERATING 30 REGION 20 10 0.0) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index. 5 WO I

I IIII I

I I

0 25 50 75 100 0

25 50 75 100 Group 5 Group 7 I

I I

0 25 50 75 100 Group 6 ROD POSITION LIMITS FOR FOUR-PUMP OPERATION FROM 0 TO 250 + EFD OCONEE 2 oUKPOWE OCONEE NUCLEAR STATION Figure 3.5.2-1B Amendments Nos.

66, 66 & 63 3.5-14

OPERATION IN THIS (133,102)

(259,102)

(300.102) 100 - REGION IS NOT ALLOWED

(

(253.7,92) 90 80 -

(235.80)

POWER LEVEL SHUTDOWN MARGIN CUTOFF 70 -

LIMIT RESTRICTED REGION 50 o

50 (52,50)

(168.5p)

PERMISSIBLE OPERATING REGION 40 30 30 -(0,27) 20 10 0

20 40 60 80 100 120 140 160 180 200 220 240 250 280 300 Roo Index, 5 WO I

I I

I I

I I

I 0

25 50 75 100 0

25 50 75 100 Group 5 Group 7 11 1I I

0 25 50 75 100 Group 6 ROD POSITION LIMITS FOR FOUR-PUMP OPERATION AFTER 250 + EFPD OCONEE 2 DHOWE). OCONEE NUCLEAR STATION Figure 3.5.2-1B2 Amendments Nos. 66, 66 & 63 3.5-14a

Figure 3.5.2-1B3 Deleted During Oconee Unit 2, Cycle 4 Operation Amendments Nos.

66, 66 & 63 3.5-15

OPERATION IN THIS REGION IS NOT ALLOWED WITH 2 OR (130102) (151.4,102) (225.5,102) 3 PUMPS (108, 102)

(246.5,102) 100 (125.5.92)

(251. 4 92) 90 RESTRICTED 80 g

REGION FOR 80 -

as 1 vt2&3 PUMP

• ^. F,.OPER.

70 SHUTDOWN (300,74)

MARGIN (125.5,63)

(251.4.63)

.60 LIMIT 50 37,50)

(300,51) 40 4

PERMISSIBLE 30 (0,28)

OPERATING REGION 20 10

'a 0

20 40 SO 80- 100 120 140 160 180 200 220 240 260 280 300 320 Roa Index, 5 WO I

II I

I I

0 25 50 75 100 0

25 50 75 100 Group 5 Group 7 1

1 I

I I

O 25 50 75 100 Group 6 ROD POSITION LIITS FOR TWO AND THREE PUMP OPERATION FROM 0 TO 250 + 10 EFPD OCONEE 2 OCONEE NUCLEAR STATION Figure 3.5.2-2B1 Amendments Nos. 66, 66 & 63 3.5-19

110 OPERATION IN THIS (133.102)

(180 102)

(235 102) 100. REGION IS NOT (300.102)

ALLOWED WITH 90 - 2 0R 3 PUMPS (168.92) 80 70 LIMIT (188.6-3) 60 -

o';

50 -

(52, 50) 4 40 1

PERMISSIBLE 30 - (0,27)

OPERATING REGION 20 10 0

I f

I I

I I

I I

I 0

20 40 60 80 100 120 140 180 180 200 220 240 250 280 300 Roa Index, % WO p

I I

I I

0 25 50 75 100 0

25 50 75 100 Group 5 Group 7 0

25 50 75 100 Group 6 ROD POSITION LIMITS' FOR TWO AND THREE PUMP OPERATION AFTER 250 + 10 EFPD OCONEE 2 turowE OCONEE NUCLEAR STATION Figure 3.5.2-2B2 Amendments Nos.

66, 66 & 63 3.5-19a

Figure 3.5.2-2B3 Deleted during Oconee Unit 2, Cycle 4 Operation Amendments Nos. 66, 66 & 63 3.5-19b

Power % of 2568 MWt RESTRICTED REGION 110

(-13.2,102)

T"0 (8,102)

(-.102

(-14.2,92) 90 (8.6.92)

(-202.80) 80 70 PERMISSIBLE 60 OPERATING REGION 50

-- 40 30 20 10 I

0

-30

-20

-10 0

10 20 30 Core Imbalance. $

OPERATIONAL POWER T BALANCE ENVELOPE FOR OPERATION FROM 0 TO 250 + 10 EFPD OCONEE1Z QUKEPOWR OCONEE NUCLEAR STATION Figure 3.5.2-3B1 Amendments Nos.

66, 66 & 63 3.5-22

Power af 2568 MWt

-110 RESTRICTED REGION

(-22.8.102)

(9.7.102

-100

(-20.7.92)

(17.4.92)

PERMISSIBLE 80 OPERATING REGION 70 60 50 40 30 20

- -10

-30

-20

-10 0

10 20 30 Imnalance, %

OPERATIONAL POWER IALANCE ENVELOPE FOR OPERATION AFTER 250 + 10 EFPD OCONEE 2 OCONEE NUCLEAR STATION Figure 3.5.2-3B2 Amendments Nos.

66, 66 & 63 3.5-22a

Figure 3.5.2-3B3 Deleted During Oconee Unit 2, Cycle 4 Operation Amendments Nos.

66, 66 & 63 3.5-22b

110 RESTRICTED REGION (6.1,102)

(25.5,102) 100 90 (3.2.92)

(27.5,92) 80 -

(0,80)

(29.3.80) 70 -

RESTRICTED REGION 60 50 (64.4,50)

(100,40.)

-L 40 PERMISSIBLE OPERATING 30 -

REGION 20 10 0

10 20 30 40 50 60 70 80 90 OO APSR, % WO APSR POSITION LIMITS FOR OPERATION FROM 0 TO 250 + 10 EFPD, OCONEE 2 noown.OCONEE NUCLEAR STATION Figure 3.5.2-4B1 Amendments Nos. 66, 66 & 63 3.5-23f

RESTRICTED REGION 110 10 -

(61,102)

(32,102) 10 80 (0 80o (33.7,80) 70 RESTRICTED REGION PERMISSIBLE OPERATING o

REGION 50 RE IN64.4,50)

S40 40--

(100,40) 30 20 10 0

10 20 30 40 50 60 70 80 90 100 APSR %,

WO APSR POSITION LIMITS FOR

-OPERATION AFTER 250 + 10 EFPD OCONEE 2 nown OCONEE NUCLEAR STATION Figure 3.5.2-4B2 Amendments Nos. 66, 66 & 63 3.5-23g

Figure 3.5.2-4B3 Deleted during Oconee Unit 2, Cycle 4 Operation Amendments Nos. 66, 66 & 63 3.5-23h

Table 4.1-2 MINIMUM EQUIPMNT TEST FREQUENCY Item Test Frequency

1.

Control Rod Movement (1)

Movement of Each Rod Monthly

2.

Pressurizer Safety Valves Setpoint 50% Annu ally

3. Main Steam Safety Valves Setpoint 25% Annu ally

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. High Pressure and Low (3)

Vent Pump Casings Monthly and Pressure Injection System Prior to Testing (1) Applicable only when the reactor is critical (2) Applicable only when the reactor coolant is above 200 7 and at a steady state temperature and pressure.

(3) Operating pumps excluded.

Amendments Nos. 66, 66 & 63 4.1-9