ML15112B053

Amends 111,111 & 108 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Revising Tech Specs to Support Full Power Operation of Unit 2 During Fuel Cycle 6,per Util 811113 & 820324 Requests

ML15112B053
Person / Time
Site:	Oconee
Issue date:	04/08/1982
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Duke Power Co
Shared Package
ML15112B054	List: ML15112B053 ML15112B055 ML15112B057
References
DPR-38-A-111, DPR-47-A-111, DPR-55-A-108 NUDOCS 8205100043
Download: ML15112B053 (33)

Text

4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 111 License No. DPR-38

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

A. The application for amendment by Duke Power Company (the licensee) dated November 13, 1981, 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 pro visions 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 satis fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B 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. 111 are hereby incorporated in the license. The licensee shiall operate the facility in accordance with the Technical Specifications.

8~205100043 820408 PDR ADOCK 05000269 P

-~

PDR

-2

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

FOR THE NUCLEAR REGULATORY COMMISSION Joh F Stolz, Chief Ope ating Reactors Branch.4 D ision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 8, 1982

R oUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.

50- 270 OCONEE NUCLEAR STATION, UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 11 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 November 13, 1981, 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 pro visions 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 reoulations and all applicable requirements have been satis fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No. DPR-4 7 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.111 are hereby inrorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

-2 3.. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Joh -F. Stolz, Chief rating Reactors Branch #4 ivision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 8, 1982

oUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.

50- 287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.108 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 November 13, 1981, 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 pro visions 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 satis fied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No.

DPR-55 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.108 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

-2

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

OR THE NUCLEAR REGULATORY COMMISSION Jo n--F. Stolz, Chief erating Reactors Br nch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 8, 1982

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 111 TO DPR-38 AMENDMENT NO. 111 TO DPR-47 AMENDMENT NO.

108 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 2.1-2 2.1-2 2.1-3 2.1-3 2.1-3b 2.1-3b 2.1-8 2.1-8 2.1 -11 2.1-11 2.3-9 2.3-9 2.3-11 2.3-11 3.2-2 3.2-2 3.5-9 3.5-9 3.5-10 3.5-10 3.5-16 3.5-16 3.5-16a 3.5-16a 3.5-16b 3.5-19 3.5-19 3.5-19a 3.5-19a 3.5-19b 3.5-19c 3.5-19d 3.5-19e

-2 REMOVE PAGES INSERT PAGES 3.5-22 3.5-22 3.5-22a 3.5-22b 3.5-25 3.5-25 3.5-25a 3.5-25a 3.5-25b

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 setpoints to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-lA 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 106.5 percent of 131.3 x 106 lbs/hr.). This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, 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 and rod bowing:

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

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

The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup independent DNBR rod bow penalty for the ap plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis.

All plant operating limits are based on a minimum DNBR criteria of 1.30 plus the amount necessary to offset the reduction in DNBR due to fuel rod bow. (3)

Amendments Nos. 111 111, & 108 2.1-2

The maximum thermal power for three-pump operation is 89.899 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 =

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

For each curve of Figure 2.1-3A 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 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.

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

(2) Oconee 1, Cycle 4 -

Reload Report - BAW-1447, March, 1977.

(3) Oconee 1, Cycle 7 - Reload Report -

BAW-1660, March, 1981 Amendments Nos.

&11

, 111 108 2.1-3

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

The magnitude of the rod bow penalty applied to each fuel cycle is equal to or greater than the necessary burnup independent DNBR rod bow penalty for the ap plicable cycle minus a credit of 1% for the flow area reduction factor used in the hot channel analysis. All plant operating limits are based on a minimum DNBR critieria of 1.30 plus the amount necessary to offset the reduction in DNBR due to fuel rod bow. (3)

The maximum thermal power for three-pump operation is 90.606 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.08 =

80.68 percent power plus the maximum calibration and instrument error. The maximum thermal power for other coolant pump conditions is 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 particular reactor coolant pump situation. The curve of Figure 2.1-1B is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.

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

(2) Oconee 2, Cycle 4 - Reload Report -

BAW-1491, August, 1978.

(3) Oconee 2, Cycle 6 - Reload Report -

BAW-1691, August 1981.

Amendments Nos. 111 111

, & 108 2.1-3b

9 9

THERMAL POWER LEVEL, S FP UNACCEPTABLE 120 OPERATION

(-41.0,112.0)

(42.0,112.0)

(-42.0,107.3)

ACCPAL ACCEPTABLE 4 PUMP OPERATION 100

(-41.0,90.806)

(42.0,90.606)

(-42.0,85.906)1)

ACCEPTABLE 3 & 4 PUMP OPERATION--8 UNACCEPTABLE O

OPERATION I

(59.0,71.906)

II

(-41.0.63.150)

(42.0,63.150)

(-42.0,58.45)

ACCEPTABLE 60 2,3 & 4 PUMP OPERATION N~(59.0,44.45)

-- 40 20

-60

-40

-20 0

+20

+40

+60 Reactor Power Imnalance. %

CORE PROTECTION SAFETY LIMITS UNIT 2 emIpow OCONEE NUCLEAR STATION FIGURE 2.1-28 Amendments Nos. 111

,111, & 108 2.1-8

2400

/

ACCEPTABLE OPERATION 2200 2000 1

3 1800 1600 560 580 600 620 640 Reactor Coolant Outlet Temperature, F CURVE COOLANT FLOW, GPM POWER, %

PUMPS OPERATING TYPE OF LIMIT 1

374,880 (100%)

112 4

ONBR 2

280,035 (74.7%)

90.606 3

ONBR 3

183,690 (49.0%)

63.150 1 PER LOOP QUALITY CORE PROTECTION SAFETY LIMITS UNIT 2 u

OCONEE NUCLEAR STATION FIGURE 2.1-38 Amendments Nos.111

, 11

, &108 2.1-11

THERMAL POWER LEVEL, 7 FP (17,108.0)

(-17,108.0)

= +1.0 ACCEPTABLE 2

4 PUMP 100 UNACCEPTABLE

(-28,97.0)

OOPERATION I

I OPERATION I

I I

I UNACCEPTABLE j

1(-17,80.6) 80 (17,80.6)

(42,83.0)

OPERATION ACCEPTABLE 3&4 PUMP

(-28,69.6)

OPERATION II 1(-17,52.9)

I(17,52.9)

(42,55.6)

ACCEPTABLE 2.3&4

(-28,41.9)

PUMP 40 OPERATION (42,27.9)

- 20 I

I -I w

II I

I I

I I

-40

-20 0

+20

+40 Reactor Power lImoalance, *%

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 2 omWs OCONEE NUCLEAR STATION FIGURE 2.3-2B Amendments Nos. 111

, 111

, & 108 2.3-9

Table 2.3-1A Unit I 93 Reactor Protective System Trip Setting Limits One Reactor o

Four Reactor Three Reactor Coolant Pump 0

Coolant Pumps Coolant Pumps Operating in Operating Operating Each Loop (Operating Power (Operating Power (Operating Power Shutdown RPSSement

-100% Rated)

-75%. Rated)

-49% Rated)pss

1.

Nuclear Power Max.

104.9 104.9 104.9 (f Rated)

2.

Niclear Power Hax. Based 1.07 times flow 1.07 times flow 1.07 times flow Bypassed on Flow (2) and Imbalance, minus reduction minus reduction minsim red ction (7.

Rated) due to imbalance due to imbalance due to imbalance

3.

Nuclear Power Hax. Based NA NA 55%

Bypasse on Pump Honitors, (f Rated)

4.

Hfigh Reactor Coolant System 2300 2300 2300 1720(4)

Pressure, psig, lax.

5.

Low Reactor Coolant System 1800 1800 1800 Bypassed Pressure, psig, Min.

6.

Variable Low Reactor Coolant (11.14 T 4706)(1)

(11.14 T 4706)(1)

(11.14 To 4706)(1)

RypassP4 System Pressure psig, Hin.

out out out

7.

Reactor Coolant Temp. F., Max.

618 618 61R 618

8.

High Reactor Building 4

4 4

4 Pressure, psig, Hax.

UlT is in degrees FahrenheitoF).

(2) Reactor Coolant System Flow, %.

(3) Administratively controlled reduction set only during reactor shutdown.

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

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 the reactor coolant system to a 1% Ak/k subcritical margin at cold conditions (700F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life. The current cycles for each unit, Oconee 1, Cycle 7, Oconee 2, Cycle 6, and Oconee 3, Cycle 6 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 1020 ft3 of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1835 ppm boric acid in the borated water storage tank (3) will satisfy the requirements. The volume requirements include 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 con dition. 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 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-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 77oF and therefore a temperature requirement of 87*F. Once in the high pressure injection system, the concen trate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES (1) FSAR, Section 9.1; 9.2 (2) FSAR, Figure 6.2 (3) Technical Specification 3.3 Amendments Nos.

111 111, & 108 3.2-2

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 reduction 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 3.5.2.2.

b. Except for physics tests, operating rod group overlap shall be 25% + 5% between two sequential groups. If this limit is ex ceeded, corrective 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 an 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 shaping 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, 3.5.2-LA2, and 3.5.2-LA3 (Unit 1); 3.5.2-131, 3.5.2-132, and 3.5.2-1B3 (Unit 2); 3.5.2-lCl, 3.5.2-1C2 and 3.5.2-1C3 (Unit 3) for four pump operation, on figures 3.5.2-2A1, 3.5.2-2A2, and 3.5.2-ZA3 (Unit 1); 3.5.2-2B1, 3.5.2-232, and 3.5.2-233 (Unit 2) for three pump operation, on figures 3.5.2-2A4, 3.5.2-2A5, and 3.5.2-2A6 (Unit 1); 3.5.2-234, 3.5.2-2B5, and 3.5.2-2B6 (Unit 2) for two pump operation, and on figures 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 insertion/

withdrawal limits are specified on figures 3.5.2-4Al, 3.5.2-4A2, and 3.5.2-4A3 (Unit 1); 3.5.2-431, and 3.5.2-432, and 3.5.2-433, (Unit 2); 3.5.2-4C1, 3.5.2-4C2, and 3.5.2-4C3 (Unit 3).

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

Amendments Nos. 111

, 111

, & 108

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-lAl, 3.5.2-1A2, and 3.5.2-1A3 for Unit 1; Figures 3.5.2-131, 3.5.2-132, and 3.5.2-133, for Unit 2; and Figures.3.5.2-1C1, 3.5.2-1C2, and 3.5.2-103 for Unit 3 unless one of the following conditions is satisfied:

1. Xenon reactivity did not deviate more than 10 percent from the equilibrium 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 its approach to its equilibrium value for operation at the power level cutoff.

3. Except for xenon free startup (when 2. applies), the reactor has operated within a range of 87 to 92 percent of rated thermal power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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, 3.5.2-3A3, 3.5.2-31, 3.5.2-332, 3.5.2-333, 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, cor rective 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.

3.5.2.8 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.

3.5.2.9 The operational limit curves of Technical Specifications 3.5.2.5.c and 3.5.2.7 are valid for a nominal design cycle length, as defined in the Safety Evaluation Report for the appropriate unit and cycle.

Operational beyond the nominal design cycle length is permitted pro vided that an evaluation is performed to verify that the operational limit curves are valid for extended operation. If the operational limit curves are not valid for the extended period of the operation, appropriate limits will be established and the Technical Specification curves will be modified as required.

Amendments Nos. 111 111, & 108 3.5-10

SHUTDOWN MARGIN LIMIT 100 -

(117,102)

(281,102)

POWER OPERATION IN THIS LEVEL REGION NOT ALLOWED (277.5,92)

CUTOFF = 100:

80 RESTRICTED (264.5,80) 80-OPERATI ON K F 31E KW FT LIMIT 60 (58,50)

(200,50) 40 PERMISSIBLE OPERATING 20 REGION (0,2.5) 0 0

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

25 50 75 100 Rod Index, % Witndrawn Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 I

I Group 7 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 2 neo~

OCONEE NUCLEAR STATION FIGURE 3.5.2-11 Amendments Nos.

111 111, & 108 3.5-16

(184.102)

(274,102)

POWER LEVEL (271,92)LE L

SHUTDOWN

.7,2CUTOFF

= 100 FF MARGIN LIMIT 80 (258,80) 30 OPERATION IN THIS RESTRICTED KW/FT OPERATION LIMIT REGION NOT ALLOWED 60 (118,50)

(200,50) 40 PERMISSIBLE OPERATING REGION 20 (0,8.1(57:15) 0

-- j 0

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

25 50 75 100 Rod Index, S Witndrawn Group 5 0

25 50 75 100 1

I II, Group 6 0

25 50 75 100 1

I I

I Group 7 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 50 (+10, -0) to 225 +10 EFPD UNIT 2 hus OCONEE NUCLEAR STATION FIGURE 3.5.2-1B2 Amendments Nos. 111-

, 111

, & 108 3.516a

120 (274, 102) 100 (243,102)

POWER LEVEL SHUTDOWN (264.5,92)

CUTOFF = 1001 F MARGIN LIMIT 80 OPERATION IN THIS IE (255,80)

REGION NOT ALLOWED KWFT LIMIT 60 (176,50)

(2050) 40 PERMISSIBLE OPERATING REGION 20 (101,15)

(0,5.4) 0 1

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

25 50 75 100 Rod Index, Z Witnarawn Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 225 +10 EFPO

• SSE NEE NUCLEAR STATION FIGURE 3.5.2-113 Amendments Nos. 111 1l,

& 108 3.5-16b

100 80

-SHUTDOWN 77)(258,77)

MARGIN LIMIT RESTRICTED OPERATION 60 OPERATION IN LIMIT THIS REGION NOT ALLOWED (200,50)

S 40 40 -

(58,38) 20 PERMISSIBLE OPERATING REGION 0,11.98)

(90,15)

(0,2.

0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, % Witnarawn 0

25 50 75 100 Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 0 to 50 C+10, -0) EFPO UNIT 2 Pumu OCONEE NUCLEAR STATION FIGURE 3.5.2-281 3.5-19 Amendments Nos.

111

,111

, & 108

100 80 (184,77)

(252,77) 80 SHUTDOWN RESTRICTED WF

60 SHTONLIMIT OPERATION IN MARGIN OPERATION OPERATION IN LIMIT THIS REGION o

NOT ALLOWED (200.50) 40 (118, 38)

PERMISSIBLE OPERATING 20

-REGION (57,11.75) 0 0

20 40 60 80 100 120 140 160 180 200 220 240.260 280 300 0

25 50 75 100 Rod Index, 01 Witndrawn Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM.50 (+10, -0) to 225 +10 EFPO UNIT 2 Dnma OCONEE NUCLEAR STATION FIGURE 3.5.2-282 Amendments Nos.

1,

111, & 108 3.5-19a

100 80 (243.77)

SHUTDOWN MARGIN (250,77)

LIMIT 60 RESTRICTED LIMIT OPERATION IN THIS REGION NOT ALLOWED (200,50)

(176,38) 20

-PERMISSIBLE OPERATING REGION (0,4.5)

(101,11.75) 0 0

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

25' 50 75 100 Rod Index, z Witnarawn I

I I

I Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR THREE PUMP OPERATION AFTER 225 +10 EFPD UNIT 2 OCONEE NUCLEAR STATION FIGURE 3.5.2-283 Amendments Nos.11

,1l

, & 108 3.5-19b

100 80 SHUTDOWN MARGIN LIMIT 60 -

(204.52)

OPERATION IN THIS REGION NOT

/2(17552 ALLOWED 40 RESTRCTEDKW/FT LIMIT OPERATION 20 (0.8.6 (58,26)

PERMISSIBLE OPERATING REGION 0,2.

(90,15)

O 20 40 60 80 100 120 140 160 1.80 200 220 240 260 280 300 0

25 50 75 100 Rod Index, % Witndrawn 0

25 50 75 100 GrouG 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 0 to 50.(+10, -0) EFPD UNIT 2 ouow OW OCONEE NUCLEAR STATION FIGURE 3.5.2-2B4 Amendments Nos.

111

, 111

, & 108 3.5-19c

100 80 HE RESTRICTED 60 OPERATION 1

60 OPERATION IN THIS (204,52)

REGION NOT (184,52)

ALLOWED

( O0,50 ALEDSHUTDOWN 005)

KW/FT 40 MARGIN LIMIT LIMIT (118,26) 20 PERMISSIBLE OPERATING REGION (57,8.5) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0

25 50 75 0Rod index, % Witndrawn Group 5 0

25 50 75 100 Group 6 0

25 50 75 100 Group 7 ROD POSITION LIMITS FOR.TWO PUMP OPERATION FROM 50 (+10, -0) to 225 +10 EFPO UNIT 2

~USEPOW OCONEE NUCLEAR STATION FIGURE 3.5.2-285 Amendments Nos. 111

,111

, & 108 3.5-19d

100 80 ME 60 CD (243,52)

SHUTDOWN MARGIN 40 4LIMIT OPERATION IN THIS REGION 176,26) 20 NOT ALLOWED PERMISSIBLE OPERATING REGION (0,3.7) 101,8.5) 0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, S Witndrawn 0

25 50 75 100 Group 5 0

25 50 75 100 roupI I

Group 6 0 25 50 75 100 I

I I

I Group 7 ROD POSITION LIMITS FOR TWO PUMP OPERATION AFTER 225 +10 EFPD UNIT 2 enrown OCONEE NUCLEAR STATION FIGURE 3.5.2-ZB6 Amendments Nos. 111 111, & 108 3.5-19e

(-17.5, 102)

(+17.5,102)

..100 RESTRICTED REGION

(-17. 5,92)

(+20,92)

-90

(-28.80)

PERMISSIBLE 80

(+22,80)

OPERATING REGION

- 70

.60 00

-50 50 40 30 20 10

-40

-30

-20

-10 0

10 20 30 40 Axial Power Imbalance, Z OPERATIONAL POWER IMBALANCE LIMITS FOR OPERATION FROM 0 to 50 (+10, -0) EFPO UNIT 2 Nuo OCONEE NUCLEAR STATION FIGURE 3.5.2-381 Amendments Nos. Il 111

,& 108 352

-20,102)

(+17.5,102)

RESTRICTED 100 REGION

(-24,92) 90

(+20,92)

(-34.80)

Z PERMISSIBLE 80

(+22,80)

OPERATING REGION 70 60 50 40

- 30 20 10

-40

-30

• 20

-10 0

10 20 30 40 Axial Power Imoalance, s OPERATIONAL POWER IMBALANCE LIMITS FOR OPERATION FROM 50 (+10, -0) to 225 +10 EFPD UNIT 2 E

OCONEE NUCLEAR STATION FIGURE 3.5.2-382 Amendments Nos.

111

, 111

, &108 3.5-ZZN

RESTRICTED

(-20,102)

(+17.5,102)

REGION 100

(-22,92)

(+20.92) 90

(-32,80) 80

(+22,80)

PERMISSIBLE OPERATING REGION 70 4

-1 60

~50 40 30 20 10

-40

-30

-20

-10 0

10 20 30 40 Axial Power Imbalance, o OPERATIONAL POWER IMBALANCE LIMITS FOR OPERATION AFTER 225 +10 EFPD UNIT 2 WE w

OCONEE NUCLEAR STATION FIGURE 3.5.2-313 Amendments Nos.111 111, & 108 3.5-22b

(4,102)

(35,102) 100 (9,92)

(36,92) 80 e (0.80)

(42,80)

RESTRICTED REGION

" 60 (100,50) 40 I

PERMISSIBLE OPERATING REGION 20 01 0

20 40 60 80 100 APSR Position, 5 Witndrawn APSR POSITION LIMITS FOR OPERATION FROM 0 to 50 (+10, -0) EFPO UNIT 2 NUSome OCONEE NUCLEAR STATION FIGURE 3.5.2-481 Amendments Nos.111

,111

& 108 3.5-25

(9,102)

(35,102) 100 (9,92)

(37,92) 80 c (0,80)

(44,80)

RESTRICTED REGION S 60 (100,50) 40 PERMISSIBLE OPERATING REGION 20 0

0 20 40 60 80 100 APSR Position, % Witnorawn APSR POSITION LIMITS FOR OPERATION FROM 50 (+10, -0) to 225+10 EFPO

& 2 DunaE N

ONEE NUCLEAR STATION FIGURE 3.5.2-42 Amendments Nos. 111

,111

, & 1o8 3.5-25a

(8.5,102)

(33.5,102) 100 (8.5,92)'

(42,92) 80 c(080)

(45,80)

RESTRICTED REGION ZME 60 (100,50) 40 PERMISSIBLE OPERATING REGION 20 0

0 20 40 60 80 100 APSR Position, % Witndrawn APSR POSITION LIMITS FOR OPERATION AFTER 225 + 10 EFPO UNIT 2 In OCONEE NUCLEAR STATION FIGURE 3.5.2-4B3 Amendments Nos. 111

,111

, & 108 3.5-25b