Text

Amends 105,105 & 102 to Licenses DPR-38,DPR-47 & DPR-55, Revising Tech Specs to Allow Full Power Operation of Unit 1 for Fuel Cycle 7 & Reflecting Completed Mods to High Pressure Injection Sys
	ML15113A049
Person / Time
Site:	Oconee
Issue date:	11/30/1981
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Duke Power Co
Shared Package
ML15113A050	List: ML15113A049; ML15113A051; ML15113A052;
References
	DPR-38-A-105, DPR-47-A-105, DPR-55-A-102 NUDOCS 8112220264
	Download: ML15113A049 (42)
	v • d • e

4 V0 UNITED STATES 4? T.;

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.105 License No. DPR-38

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 29 and August 25, 1981, as supplemented October 16, 1981, comply 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 applications, 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 accordaice 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 3.8 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.105 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(8112220264 8il1l30 PDR ADOCK 05000269 P

PDR

-2

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

FOR THE NUCLEAR REGULATORY COMMISSION Johi F. Stolz, Chieff Op ating Reactors Branch #4 D\\ ision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 30, 1981

-UNITEb STATES NU LEAR REGULATORY COMMISSION WASHINGTN, D.C. 20555 DUKE POWER COMPANY DOCKET NO. 50- 270 OCONEE NUCLEAR STATION, UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.105 License No. DPR-47

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 29 and August 25, 1981, as supplemented October 16, 1981, comply 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 applications, 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 3.8 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. 105 are hereby incorporated in the license. The licensee shall operate the facility i.n accordance with the Technical Specifications.

-2

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

FOR THE NUCLEAR REGULATORY COMMISSION Jo n F. Stolz, Chief r terating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 30, 1981

UNITED STATES Nt EAR 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. 102 License No. DPR-55

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

A. The applications for amendment by Duke Power Company (the licensee) dated May 29 and August 25, 1981, as supplemented October 16, 1981, comply 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 applications, 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 accordaice 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.102 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.

FOR THE NUCLEAR REGULATORY COMMISSION Or 0

F. Stolz, Chief erating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 30, 1981

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 105TO DPR-38 AMENDMENT NO. 105TO DPR-47 AMENDMENT NO. 102TO 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 2.1-7 2.1-7 2.1-10 2.1-10 2.3-2 2.3-2 2.3-3 2.3-3 2.3-8 2.3-8 3.2-1 3.2-1 3.2-2 3.2-2 3.3-1 3.3-1 3.3-2 3.3-2 3.3-3 3.3-3 3.3-4 3.3-4 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7

- 0

-2 REMOVE PAGES INSERT PAGES 3.5-9 3.5-9 3.5-10 3.5-10 3.5-15 3.5-15 3.5-15a 3.5-15a 3.5-15b 3.5-18 3.5-18 3.5-18a 3.5-18a 3.5-18b 3.5-18c 3.5-18d 3.5-18e 3.5-21 3.5-21 3.5-21a 3.5-21a 3.5-21b 3.5-24 3.5-24 3.5-24a 3.5-24a 3.5-24b 3.8-3 3.8-3 4.4-12 4.4-12

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 it a 95 prcrent 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-lA 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 (3).

All plant operating limits are presently based on an original method of cal culating rod bow penalties that are more conservative than those that would be obtained with new approved procedures (3).

For Cycle 7 operation, this sub rogation results in a 10% DNBR margin, which is partially used to offset the reduction in DNBR due to fuel -rod bowing.

Amendments Nos.10 5

,105

, &102

.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 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.

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. 105

,105 & 102 2.1-3

Tnermal Power Level, i 120

(-28,112)

(32, 112)

ACCEPTABLE 110 M2= -0.9375 M =O. 850 I4 PUMP I\\

2=-.97

.00 OPERATION 100

(-48,95) 1 I (48,97)

I(-28,89.899) 90 (32,89.899)

I.

I ACCEPTABLE 13&4 PUMP

'OPERATION 50

(-48,72.899)

(48,74.899) 70

(-28,62.73)

(32,62.73)

ACCEPTABLE 60 1

12,3&4 PUMP OOPERATION 50

(-48,45.73)

(48,47.73) 40 30 UNACCEPTABLE

=

0UNACCPAL OPERATION OPERTIO 10

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 Reactor Power Imnalance,%

CORE PROTECTION SAFETY LIMITS UNIT 1 unEPO OCONEE NUCLEAR STATION Figure 2.1-2A Amendments Nos. 105 105 & 102 2.1-7

(10 (g) 2400 ACCEPTABLE OPERATION 2200 2000 2

3 1800 1600 J

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%)

89.899 3

DNBR 3

183,690 (49.0%)

62.73 2

QUALITY

106.5% OF FIRST CORE DESIGN FLOW CORE PROTECTION SAFETY LIMITS UNIT 1 unoW OCONEE NUCLEAR STATION Figure 2.1-3A Amendments Nos. 105

, 105 & 102 2.1-10

During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9% 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 setpoint produced by the reactor coolant system flow is based on a Dower-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 setpoint 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 setpoint produced by the power-to-flow ratio provides overpower DNB pro tection for all modes of pump operation. For every flow rate there is a 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 situations of Table 2.3-1A are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 107% and reactor flow rate is 100%, or flow rate is 93.46% and power level is 100%.

2. Trip would occur when three reactor coolant pumps are operating if power is 79.92% and reactor flow rate is 74.7% or flow rate is 70.09% and power level is 75%.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.43% and reactor flow rate is 49.0% or flow rate is 45.79% and the power level is 49%.

The flux-to-flow ratios account for the maximum calibration and instrument 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 power 2.3-2B -

Unit 2 2.3-2C -

Unit 3 Amendments Nos. 105

, 105 & 102 2.3-2

level trip and associated reactor power/reactor power-imbalance boundaries by 1.07% - Unit 1 for 1% flow reduction.

1.08% -

Unit 2 1.08% -

Unit 3 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 signzl divorc 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 setpoint is reached before the nuclear over power trip setpoint. 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 (2300 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient. (1)

The low pressure (1800) psig and variable low pressure.(11.14 T t-4 706 ) trip (1800) psig (11.14 Tout-4706) out (1800) psig (11.14 T O-4706) setpoints shown in Figure 2.3-IA have been established to maintain the DNB 2.3-1B 2.3-10 ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction. (2, 3)

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

'11.14 Tout -

4746)

(11.14 Tt 4746) out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (6180F) shown in Figure 2.3-1A has been established to prevent excessive core coolant 2.3-1B 2.3-1C temperature in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 0F.

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 Amendments Nos. 105

, 105 & 102

Thermal Power Level, %

(-10,107)

} 10 (14.5,107) 1

- 100 M =1.000 M =- 1.135 M1 ACCEPTABLE 2=

4 PUMP

.90 OPERATION I

(33,86.0)

(-33,84.0)

I

( 10,79.92)

(14.5,79.92)

I UNACCEPTABLE UNACCEPTABLE OPERATION

-70 OPERATION ACCEPTABLE I

3&4 PUMP OPERATION 60I

(-33,56.92)

I (33,58.92) 1(-10,52.43)

(14.5,52.43) 1 1

50 I 1 40 ACCEPTABLE

(-33,29.43) 2,3&4 PUMP 30 (33,31.43)

OPERATION I

I 20 K, l

-7 10

-n I

IiI II

-40

-30

-20

-10 0

10 20

30.

40 Power Imoalance, %

PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNIT 1 DUEPOWER OCONEE NUCLEAR STATION Figure 2.3-2A Amendments Nos. 105

, 105 & 102 2.3-8

6 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 1020 ft3 of 8700 ppm boron as boric acid solution with a temperature at least 100F 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the reactor shall be placed in a hot shutdown condition and be borated to a shutdown margin equivalent to 1% ak/k at 200 0 F 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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in a cold shutdown condition within an addition al 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendments Nos. 105

,105

& 102 3.2-1

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% dk/k subcritical margin at cold conditions (700 F) 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 5, 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 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months 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 770 F and therefore a temperature requirement of 870 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. 105

, 105 & 102 3.2-2

3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR BUILDING SPRAY, AND LOW PRESSURE SERVICE WATER SYSTEMS Applicability Applies to the emergency core cooling, reactor building cooling, reactor building spray, and low pressure service water systems.

Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building cooling, reactor building spray and low pressure service water systems.

Specification 3.3.1 High Pressure Injection (HPI) System

a. Prior to initiating maintenance on any component of the HPI system, the redundant component shall be tested to assure operability.

b. When the reactor coolant system (RCS), with fuel in the core, is in a condition with temperaturE above 350oF and reactor power -less than 60% FP:

(1) Two independent trains, each comprised of an HPI pump and a flow path capable of taking suction from the borated water storage tank and discharging into the reactor coolant system automatically upon Engineered Safeguards Protective System (ESPS) actuation (HPI segment) shall be operable.

(2) Test or maintenance shall be allowed on any component of the HPI system provided one train of the HPI system is operable.

If the HPI system is not restored to meet the requirements of Specification 3.3.1.b(1) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the requirements of Specification 3.3.1.b(1) are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in.a condition with RCS temperature below 3501F within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. For all Units, when reactor power is greater than 60% FP:

(1) In addition to the requirements of Specification 3.3..1.b(1) above, the remaining HPI pump and valves 3HP-409 and 3HP-410 shall be operable and valves HP-99 and HP-100 shall be open.

(2) Tests or maintenance shall be allowed on any component of the HPI system, provided two trains of HPI system are operable.

If the inoperable component is not restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reactor power shall be reduced below 60% FP within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.3-1 Amendments Nos. 105

,15

& 102

3.3.2 Low Pressure Injection (LPI) System

a. Prior to initiating maintenance on any component of the LPI system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250 0F:

(1) Two independent LPI trains, each comprised of an LPI pump and a flowpath capable of taking suction from the borated water storage tank and.discharging into the RCS automatically upon ESPS actuation (LPI segment), together with two LPI coolers and two reactor building emergency sump isolation valves (manual or remote-manual) shall be operable.

(2) Tests or maintenance shall be allowed on any component of the LPI system provided the redundant train of the LPI system is operable. If the LPI system is not restored to meet the re quirements of Specification 3.3.2.b(1) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the requirements of Specification 3.3.2.b(1) are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 0F within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.3.3 Core Flood Tank (CFT) System When the RCS is in a condition with pressure above 800 psig both CFT's shall be operable with the electrically operated discharge valves open and breakers locked open and tagged; a minimum level of 13 t.44 feet (1040 +/- 30 ft. 3 ) and one level instrument channel per CFT; a minimum concentration of borated water in each CFT of 1835 ppm boron; and pressure at 600 +/- 25 psig with one pressure instrument channel per CF1.

3.3.4 Borated Water Storage Tank (BWST)

When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250 0F:

a. The BWST shall have operable two level instrument channels.

(1) Tests or maintenance shall be allowed on one channel of BWST level instrumentation provided the other channel is operable.

(2) If the BWST level instrumentation is not restored to meet the requirements of Specification 3.3.4.a above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the requirements of Specification 3.3.4.a are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 0F within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Amendments Nos.

105 9 105 &102 3.3-2

b. The BWST shall contain a minimum level of 46 feet of water having a minimum concentration of 1835 ppm boron at a minimum temperature of 50 F. The manual valve, LP-28, on the discharge line shall be locked open. If these requirements are not met, the BWST shall be considered unavailable and action initiated in accordance with Specification 3.2.

3.3.5 Reactor Building Cooling (RBC) System

a. Prior to initiating mainteace on any component of the RBC system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250PF and subcritical:

(1) Two independent R3C trains, each comprised of an RBC fan, associated cooling unit, and associated ESF valves shall be operable.

(2) Tests or maintenance shall be allowed on any component of the RBC system provided one train of the RBC and one train of the RBS are operable. If the RBC system is not restored to meet the requirements of Specification 3.3.5.b(l) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 2500F with in.an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. When the reactor is critical:

(1) In addition to the requirements of Specifications 3.3.5.b(l) above, the remaining RBC fan, associated cooling unit, and associated ESF valves shall be operable.

(2) Tests or maintenance shall be allowed on one RBC train under either of the following conditions:

(a) One RBC train may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(b) One RBC train may be out of service for 7 days provided both RBS trains are operable.

(c) If the inoperable RBC train is not restored to meet the requirements of Specification 3.3.5.c(l) within the time permitted by Specification 3.3.5.c(2)(a) or (b), the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the requirements of Specification 3.3.5.c(l) are not met within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be ?laced in a condition with RCS 0pressure below 350 psig and RCS temperature below 250 F.within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.3-3 S Amendments Nos. 105

,105 & 102

0-)

01) 3.3.6 Reactor Buildihg Spray (RBS) System

a. Prior to initiating maintenance on any component of the RBS system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 2500F and subcritical:

(1) One RBS train, comprised of an RBS Dumo and a flowpath capable of taking suction from the LPI system and discharging through.

the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.

(2) Tests or maintenance shall be allowed on any component of the RBS system under the following conditions:

(a) One RBS train may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided two RBC train are operable.

(b) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.b(1) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a condition with the RCS pressure below 350 psig and RCS temperature below 2500?

within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

When the reactor is critical:

(1) In addition to the requirements of Specifications 3.3.6.b(l)

L above, the other RBS train comprised of an R3S pump and a flowpath capable of taking suction of the LPI system and discharging through the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.

(2) Tests or maintenance shall be allowed on one RBS train under either of the following conditions:

(a) One RBS train may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(b) One RBS train may be out of service for 7 days provided all three RBC trains are operable.

(c) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.c(1) above within the time permitted by Specification 3.3.5.c(2)(a) or (b),

the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the requirements of Specification 3.3.6.c(l) are not met within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temp erature below 250 0F within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

,3.3-4 Amendments Nos. 105

, 105 & 102

3.3.7 Low Pressure Service Water '(LPSW)

a. Prior to initiating maintenance on any component of the LPSW system, the redundant component shall be tested to assure operability.

b. When the RCS, with fuel in the core, is in a condition with pres sure equal to or greater than 350 psig or temperature equal to or greater than 250 F:

(1) Two LPSW pumps for the shared Unit 1, 2 LPSW system and two LPSW pumps for the Unit 3 LPSW system shall be operable with valves LPSW-108, 2LPSW-108, and 3LPSW-108 locked open.

(2)

Tests or maintenance shall be allowed on any component of the LPSW system provided the redundant train of the LPSW system is operable. If the LPSW system is not restored to meet the requirements of Specification 3.3.7.b(l) above within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If the requirements of Specification 3.3.7.b(l) are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig O

and RCS temperature below 250 within an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Bases Specification 3.3 assures that, for whatever condition the reactor coolant system is in, adequate engineered safety. feature equipment is operable.

For operation up to 60% FP, two high pressure injection pumps are specified.

(

Also, two low pressure injection pumps and both core flood tanks are required.

In the event that the need for emergency core cooling should occur, func tioning of one high pressure injection pump, one low pressure injection pump, and both core flood tanks will protect the core, and in the event of a main coolant loop severence, limit the peak clad temperature to less than 2,2000?

and the metal-water reaction to that representing less than 1 percent of the clad. (1) Both core flooding tanks.are required as a single core flood tank has insufficient inventory to reflood the core.

The requirement to have three HPI pumps and two HPI flowpaths operable during power operation above 60% FP is based on considerations of potential small breaks at the reactor coolant pump discharge piping for which two HPI trains (two pumps and two flow paths) are required to assure adequate core cooling.(2)

The analysis of these breaks indicates that for operation at or below 60% F?

only a single train of the HPI system is needed to provide the necessary core cooling.

The borated water storage tanks are used for two purposes:

(a) As a supply of borated water for accident conditions.

(b) As a supply of borated water for flooding the fuel transfer canal during refueling operation.(3)

Amendments Nos. 105 105 & 102 3.3-5

Three-hundred and fifty thousand (350,000) gallons of borated water (a level of 46 feet in the BWST) are required to supply emergency core cooling and reactor building spray in the event of a loss-of-core cooling accident. This amount fulfills requirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based on refueling volume require ments. Heaters maintain the borated water supply at a temperature above 500F to lessen the potential for thermal shock of the reactor vessel during high pressure injection system operation. The boron concentration-is set at the amount of boron required to maintain the core 1 percent subcritical at 700F without any control rods in the core.

The minimum value speuified in the tanks is 1835 ppm boron.

It2has been shown for the worst design basis loss-of-coolant accident (a 14.1 ft hot leg break) that the Reactor Building design pressure will not be exceeded with one spray and two coolers operable. (4) Therefore, a maintenance period of seven days is acceptable for one Reactor Building cooling fan and its associated cooling unit provided two Reactor Building spray systems are operable for seven days or one Reactor Building spray system provided all three Reactor Building cooling units are operable.

Three low pressure service water pumps serve Oconee Units 1 and 2 and two low pressure service water pumps serve Oconee Unit 3. There is a manual cross connection on the supply headers for Units 1, 2, and 3. One low pressure service water pump per unit is required for normal operation. The normal oper ating requirements are greater than the emergency requirements following a loss-of-coolant accident.

Prior to initiating maintenance on any of the components, the redundant component (s) shall be tested to assure operability. Operability shall be based on the results of testing as required by Technical Specification 4.5.

The maintenance period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable if the operability of equipment redundant to that removed from service is demonstrated i=edi ately prior to removal. The basis of acceptability is a likelihood of fail ure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following such demonstration.

REFERENCES (1) ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock &

Wilcox, Lynchburg, Virginia, June 1975.

(2) Duke Power Company to NRC letter, July 14, 1978, "Proposed Modifications of High Pressure Injection System".

(3) FSAR, Section 9.5.2 (4) FSAR, Supplement 13 3.3-6 Amendments Nos. 105

, 105 & 102

0 0-)

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 3.3.2.2.

b.

Except for physics tests, operating rod group overlap shall be 25% t 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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, 3.5.2-1A2, and 3.5.2-1A3 (Unit 1),

3.5.2-131, and 3.5.2-132 (Unit 2); 3.3.2-1C1, 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, and 3.5.2-2A3 for three pump operation and 3.5.2-2A4, 3.5.2-2A5, and 3.5.2-2A6 for two pump operation (Unit 1); 3.5.2-231, and 3.5.2-2B2 (Unit 2); and 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 3.5.2-4A1, and 3.3.2-4A2 (Unit 1); 3.5.2-431, and 3.5.2-4B2, (Unit.2); 3.5.2-4G1, 3.5.2-4C2, and 3.5.2-4C3 (Unit 3).

If the control rod position limits are exceeded, corrective mea sures 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. 105 105 & 102 3.5-9

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-lA1, 3.5.2-1A2, and 3.5.2-1A3 for Unit 1; Figures 3.5.2-lBl, and 3.5.2-1B2, for Unit 2; and Figures 3.5.2-lCl, 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 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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, 3.5.2-3B1, 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.

3.5.2.8 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.

Operation beyond the nominal design cycle length is.permitted provided 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. 105

, 105 & 102 3.5-10

(144.5. 102)

(284,102)

(300,102) 100 f-t (281,92)

POWER 0

LEVEL 80 (271.80)

CUTOFF OPERATION RESTRICTED

= 100% FP 60 SHUTDOWN MARGIN X3 LIMIT ca (88.50) *(200,50)

S40 OPERATION 40 NOT ALLOWED OPERATION ACCEPTABLE 20 (37,15)*

(0.11.8) <

(0.2.5) 0 100 200 300 GR 5 I

Rod Index (wi tndrawn) 0 75 100 GR06 I

I I

0 25 7

100 GR 7 0

25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 0 to 50 (+10,

-0)

EFPD UNIT 1 u

OCONEE NUCLEAR STATION FiGURE 3.5.2 -

IAl

100 -

(144.5,102 (277.5.102)

. (300,102)

(214.5,92).

POWER LEVEL 0

80 (264.5,80)

CUTOFF 100% FP OPERATION C

-RESTRICTED E

SHUTDOWN MARGIN 60 LIMIT

~'

a) oso (850).

(200,50).

OPERATION OPERATION ACCEPTABLE NOT ALLOWED 40 20 (31,15) 90,15)

(0,11.8)<

(0,2.5) 0 100 200 300 GR 5 I

I Rod Index (witidrawn) 0 75 100 GR+6 I

I I

0 25 GR I 15 1op 0

25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION FROM 50 (+10,

-0) to 200 + 10 EFPD UNIT 1 O ONi E NUCLEAFI Ed ATION

I,~

y 1 11 1 1

9 9

(220,102)

(274.5.102)

(300,10.)

100 (271,92) m POWER Ct LEVEL OPERATION NOT ALLOWED (264.5,80)

CUTOFF = 100% FP o80 C)

OPERATION RESTRICTED 60 X31 u 4(160,50)

(200.50)

SHUTDOWN MARGIN 40

-OPERATION ACCEPTABLE ILIMIT a,

39 a-20 (82,15)

(90.15)

(0,8.5)

(0,2.5 2

0 100200 300 GR 5 I IRod Index (witnrawn) 0 75 10o GR 6 I-0 25 R5 100 GR I 0

25 100 ROD POSITION LIMITS FOR FOUR PUMP OPERATION AFTER 200 + 10 EFPD UNIT I

'~r 1 OONEE NUCLEAR STATION vFTORp

.S.2 -

1A3

IOWN 100 m

CD Pt OPERATION NOT ALLOWED 0

MARGIN77

247)

(200,50) e

.LIMI cn TOPERATION RESTRICTED OD 20 01 SHUTDOWN MARGIN (0,0 P13 c

LIMIT 40 0 9(38538) ca OPERATION ACCEPTABLE 00

a. 20 (0,9. 35),

(0,2.5 I

0 100 200 300 GR 5 I

I I

Roa Indlex (witnirawn) 0 GR6 715 I0 0

25 75 1P0 GR 7P 0

25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 0 to 50 (+10,

-0)

EFPD UNIT I OOONEE NUCLER.9 :STATION FIGURE 3.5.2 -

2Al

100 03 80 o

(144.5,71)

(258,77)

(300,77)

C)

OPERATION NOT ALLOWED OPERATION RESTRICTED 60 SHUTDOWN MARGIN r-LIMIT (200,50) r3 01 40 (88,38)

OPERATION ACCEPTABLE 0091 20 (3,11.5)015)

(0.9.35)

(0,2.5) 0 100 200 300 CR 5 10Ro Index (withdrawn) 0 7R6100 GR 6 II 0

25 75 100 GR 1 t

0 25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT I OCONEE NUCLEAR STATION iGlURdE 3.J.2 -

2A2

100 (D

rt 80 (220,71)

(258,17)

(300,77)

OPERATION NOT ALLOWED.

OPER.

RESTO 60 Ln SHUTPOWN X3 MARGIN

(200, 50)

LIMIT NCa 40 (160.38)

OPERATION ACCEPTABLE co

133, 29) 39 20 a a OPERATION RESTRICTED (0,6.875)

(54.10)

(0,2.5) 0 100 200 300 GR 5 Ron Inuex (wi tnufawn) 0 5

100 GR6 JI 025 75 100 GR 7 I

I 0

25 100 ROD POSITION LIMITS FOR THREE PUMP OPERATION AFTER 200 + 10 EFPD UNIT I OCONEE NUCLFAR STATION FIGURE 3.5.2 -

2A3

100 O

m o

0_

C)

C) 60 OPERATION CD (300,52)

NOT ALLOWED (205,52) ca(144.5,52).O C)

Q SHUTDOWN PERATION 40 MARGIN LIMIT RESTRICTED Ln (88, 26) 3c OPERATION ACCEPTABLE Go Q

20

(90, 15)

(37.8.5)

(0.6.9)

(0.2.5)0 S100 200 300 GR 5 Rod Index (wilildrawn) 0 7!

10~

GR6 0 25 75 100 0~

Ig GR 7 0

25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION FROM 0 to 50 (+10, -0) EFPD UNIT 1 aOCONEE NUCLEAR STATION FIGkE 3.5.2 -

2A4

100 rt o

cm 80 CD3 0 C)

C E

OPERATION 60 NOT ALLOWED 144.5.52)

(204.52)

(300.52) cu cc, a

( 200, 50)

SHUTOOWN PERATION 40 RESTRICTED LIMIT OPERATION o

(88,26).

ACCEPTABLE 20 (0.6.9)

(0. 2.5) 0 0

100 200 300 Roo innlex (we tndarawn)

GR 5l 0

75 '100 GR 6 IIII O

25 GR7 5

1 0 0

25 100 ROD POSITION LIMITS FOR TWO PUJMP OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1

OMME NCLEAR STATOONl0

100 (D

o 80 OPERATION NOT ALLOWED 60

.(J1 (220, 52) cc (300.52),

C))

SHUToOWN

'U

+-

MARGIN 04 40 LIMIT C.3 Ln w

(160.26)

OPERATION 20 ACCEPTABLE OPERATION RESTRICTED (0.5.25)(

(0.2.5) <(286.5) 0 100 200 300 Roil Index (wI tindrawil)

GR 5 0

75 100 GR 6 I

0 25 15 1010 GR I 0

25 100 ROD POSITION LIMITS FOR TWO PUMP OPERATION AFTER 200 + 10 EFPD UNIT 1 OCONEE NUCtEAR STATION FIGURE 3.5.2 -

2A6

OP76ATION RESTRICTED

(-14,102) 100

.5,102)

(15,92) 17.5,92)

(-25,80) 80 - -

(20, 80' OPERATION 80 ACCEPTABLE 40 20-

-30

-20

-10 0

10 20 Axial Power Imoalance (%)

POWER DMALANCE LIMITS FOR OPERATION FROM 0 to 50 (+10, -0) EFD UNIT 1 Dunaw OCONEE NUCLEAR STATION FIGURE 3.5.2 -

3A1 Amendments Nos. 105

, 105 & 102 3.5-21

OPERATION RESTRICTED

-17.5,102)

(17.5,102)

(-

2

(17.5,92)

(-30.5,80) 80 (20,80)

OPERATION 60 ACCEPTABLE 40 20

-30

-20

-10 0

10 20 Axial Power ImDalance (0)

POWER IMBALANCE LIMITS FOR OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT 1 DMPiOWE OCONEE NUCLEAR STATION FIGURE 3.5.2 -

3A2 Amendments Nos. 105

, 105 & 102 3.5-21a

OPERATION RESTRICTED

(-23.5,102)

(15.5102) 100

(-24.5, 92)..

(17,92)

(-31,80) 80 -

(20,80) a OPERATION ACCEPTABLE 40 20

-30

-20

-10 0

10 20 Axial Power Imnalance (5)

POWER IMBALANCE LIMITS FOR OPERATION AFTER 200 + 10 EFPD LNIT 1 4OCONEE NUCLEAR STATION FIGURE 3.5.2 -

3A3 Amendments Nos. 105

, 105 & 102 3.5-21b

(5. 102)

(35, 102) 100 S9. 5,92)

35. 92)

OPERATION RET IiCTECu 30 ~0 80 OP E RATIO0N A C CE?T A 8L E 0H

(

Z40 A

~1

~ 20 APSR Posi -non (Parcent

nrfr

.AP.S R 0 S 110 N L 11 S FOR 0PERA0N 7ROX 0 to 50 (+!O, -0) T7?D MT71

~BIPWROCONEE NUCLEAR STATION FlGLURZ 3.5.2

-A!1 Amendments Nos. 105

,105

& 102 3-2

(8.5,102)

(32.5,102) 100 OPERATION RESTRICTED (8. 5, 92)

(34,92) 80 (0 810)

,41

80) 60 OPERATION (100,50 ACCEPTABLE b ~

40 20 0

0 20 40 60 80 100 APSR Position (% withdrawn)

APSR POSITION LIMITS FOR OPERATION FROM 50 (+10, -0) to 200 + 10 EFPD UNIT I w

OCONEE NUCLEAR STATION FIGURE 3.5.2 -

4A2 Amendments Nos.105

,105

&102 3.5-24a

(. 5,102)

(34.5,102) 100 T

(8.

5,92)

(36,92)

OPERATION RESTRICTED 80 (0

80)

(42.5,80) 60 OPERATION (100,50) 40 ACCEPTABLE 40 20 0

r I

0 20 40 60 80 100 APSR Position (% witnarawn)

APSR POSITION LIMITS FOR OPERATION AFTER 200 + 10 EFPD UNIT 1 oUKwE5ow OCONEE NUCLEAR STATION FIGURE 3.5.2 -

4A3 Amendments Nos.10 5

,105

& 102 3.5-24b

Continuous -monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The low pressure injection-pump is used to maintain a uniform boron concentration. (1) The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core.

(2) The boron concentration will be maintained above 1835 ppm. Although this concentration is sufficient to maintain the core k

<0.99 if all the control rods were removed from the core, only a few con eff=

trol rods will be removed at any one time during fuel shuffling and replace ment. The keff with all rods in the core and with refueling boron concen tration is approximately 0.9.

Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending uiiafe condition detected from the main control board indicators during fuel move ment.

The specification requiring testing of the Reactor Building purge isolation is to verify that these components will function as required should -a fuel hand ling accident occur which resulted in the release of significant fission products.

Specification 3.8.11 is required, as the safety analysis for the fuel handling accident was based on the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .(3)

The off-site doses for the fuel handling accident are within the guidelines of IOCFR100; however, to further reduce the doses resulting from this acci dent, it is required that the spent fuel pool ventilation system be operable whenever the possibility of a fuel handling accident could exist.

Specification 3.8.13 is required as the safety analysis for a postulated cask handling accident was based on the assumptions that spent fuel stored as indicated has decayed for the amount of time specified for each spent fuel pool.

Specification 3.8.14 is required to prohibit transport of loads greater than a fuel assembly with a control rod and the associated fuel handling tool(s).

REFERENCES (1) FSAR, Section 9.7 (2) FSAR, Section 14.2.2.1 (3) FSAR, Section 14.2.2.1.2 Amendments Nos. 105

, 105 & 102 3.8-3

(D TABLE 4.4 (NOTES)

C+

NOTE 1 All veited systems shall be drained of water or other fluids to the extent necessary to assure exposu re of the system coinLaLimiit isolation valves to containment atmosphere aid to assire they will he subjected to the te.;t differential pressure.

NOTE 2 Fluid sysiii that is part of tle reactor coolant pressure boundary and open direcdly to the con Laililent atiiospheret 1sunder post-accideiit conditions (vented to containmeiit atmosph're duri iig Type A test).

NOTE 3 Closed system inside containmeit that penetrates containment and postulated to ru)pture as a result o) of' a loss of coolait accident (vented to containment atmosphere during Type A LesL).

NOTE 4 System reiired to maintain Lie plant in a safe condition during the test (need n:ot be vented).

NOTE 5 System normally filled with water and operating under post-accident condition (need not be vented).

NOTE 6

a.

Coiitainment peiitLration whose design incorporates resilient seals, gaskets, or sealant compounds, piping penetrationii filled with expansion bellows, and electrical penetrations fitted with flexible ie talI seal assembl ies

b.

Air lock door seals includigi door operating mechanismswwhich are part of the containment pressire botuidary.

c.

.D)oors with resilient seals or gaskets except for seal welded doors.

d.

Compoients other than those above which must meet the acceptance criteria of Type B tests.

NOTE 7

a.

Isoliation vives provide a direct connection between the inside and outside atmospheres of the primary reactor contaiuniiLt under normal operation, siclh as purge and ventilation, vacuumi relief, and instrumen t valves.

b.

Istflation valves are repaired to close automatically upon receipt of a containment isolation signal in response to coLtrols intended to affect containment isolation.

ML15113A049

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