ML16138A688
| ML16138A688 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/15/1989 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A689 | List: |
| References | |
| NUDOCS 9001020012 | |
| Download: ML16138A688 (20) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-269 OCCNEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATINC LICENSE Amendment No.180 License No.
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Cconee Nuclear Station, Unit 1 (the facility) Facility Operating License No. DPR-38 filed by the Duke Power Company (the licensee) dated September 25,
- 198, as supplemented October 18, 1989, complies with the standards and requirements of the Atornc 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 set forth in 10 CFR Chapter I; D. The issuance of this license amendment 0'i riot 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, ard all applicable requirements have been satisfied.
- 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B. of Facility Operating License No. DFR-38 is hereby amended to read as follows:
9001020012 891215 PDR ADOCK 05000269 P
PNU
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 180, 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 its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David E. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: December 15, 1989
SRGUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-270 OCONEE Nt!CLEAR STATION, UNIT 2 AMFNDMENT TO FACILITY OPEPATING LICENSE Amendment No. 180 License No. DPR-47
- 1. The Nuclear Pegulatory Ccnsission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility) Facility Operatina License No. DPR-47 filed by the Duke Power Company (the licensee3 dated September 25, 1989, as supplemented October 18, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulaticns set forth in 10 CFR Chapter I;
- b. The facility will operate in conformity with the application, the provisions of the Act, End the rules and regulations of the Commission; C. There is reasonable assurance (W) 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 set forth in 10 CFR Chapter 1; U. The issuance of this license 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. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, arid Paragraph 3.B. of Facility Operating License No. DPR-47 is hereby amended to read as follows:
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 180, 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 its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: December 15, 1989
_pUNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. DPR-55 The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility) Facility Operating License No. DPP-55 filed by the Duke Power Company (the licensee) dated September 25, 1989, as supplemented October 18, 1989, complies with the standards and requirements of the Atumic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; E. The facility will operate in conformity with the application, the provisicns of the Act, and the rules and reculations 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 regulaticns set forth in 10 CFR Chapter I;
- 0. The issuance of this license 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..
Accordingly, the license is hereby amended by pace changes to the Technical Specifications as indicated in the attachiment to this license amendment, and Paragraph 3.B. of Facility Operating License No. OPR-55 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 177, 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 its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/11 Office of Nuclear Reactor Regulation Atta chment:
Technical Specification Changes Date of Issuance: December 15, 1989
ATTACHMENT TO LICENSE AMENDMENT NO. 180 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 180 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 177 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Pages Insert Pages 2.1-2 2.1-2 2.1-3 2.1-3 2.1-5 2.1-5 2.3-1 2.3-1 2.3-2 2.3-2 2.3-3 2.3-3 2.3-4 2.3-4 2.3-6 2.3-6 2.3-7 2.3-7 3.1-1 3.1-1 3.1-2 3.1-2 3.5-9 3.5-9 6.9-1 6.9-1
The curve presented in Figure 2.1-1(3) represents the conditions at which the minimum allowable DNBR is predicted to occur for the limiting combination of thermal power and number of operating reactor coolant pumps.
This curve is based upon the design nuclear peaking factors (4,6,7):
N F = 1.714 AH N
F = 1.50 Since power peaking is not a directly measurable quantity, DNBR limited power peaks and fuel melt limitped power peaks are separately correlated to measur able reactor power and power imbalance.
The reactor power imbalance limits, Figure 2.1-2(5),
define the values of reactor power as a function of axial imbalance that correspond to the more rescrictive of two thermal limits MDNBR equal to the DNBR limit or the linear heat rate equal. to the centerline fuel melt mit.
The core protection safety limits are based on an RCS flow less than or equal to 385,440 gpm (4 pump operation).
Three pump operation is analyzed assuming 74.7 percent of four pump flow.
The maximum thermal power for three pump operation is 84.9 percent (Figure 2.1-2) due to a power level trip produced by the flux/flow ratio (74.7 percent flow x 1.07 = 79.9 percent power 84.9 percent power adding the maximum calibration and instrument error).
OCONEE -
UNITS 1, 2, & 3 2.1-2 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
REFERENCES (1) Correlation of Critical Heat Flux in a Bundle cooled by Pressurized Water, BAW-1oo, March 1970.
(2)
Correlation of 15 x 15 Geometry Zircaloy Grid Rod Bundle CHF Data with the BWC Correlation, BAW-10143P, Part 2, August 1981.
(3) Oconee Unit 3, Cycle 7 - Reload Report, DPC-RD-2001, Rev.
1, Duke Power Company, July 1982.
(4) Oconee Nuclear Station Reload Design Methodology II, DPC-NE-1002A, Duke Power Company, October 1985.
(5) Oconee Unit 2, Cycle 7 - Reload.Report, DPC-RD-2002, Duke Power Company, September 1983.
(6) Oconee Nuclear Station Core Thermal Hydraulic Methodology using VIPRE-01, DPC-NE-2003A, Duke Power Company, July 1989.
(7)
Oconee Nuclear Station Relnd Design Methodology, NFS-1001A, Duke Power Company. April 1984.
OCONEE -
UNITS 1, 2, & 3 2.1-3 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
THERMAL POWER LEVEL %
120 M I=0(713 1. 1,112. 0)
(31.1,112.0)
Ml =0.71 ACCEPTABLE 4 PUMPOPERATIO M2=-o.71
(-48.0,100.0) 100 -
(801 00 31.1. 84. 9)
(3 1,1 8 4. 9) 80 4 8. 0, 72.9)
ACCEPTABLE 3 & 4 PUMP OPERATIODN (48.0,72.9) 60 40 20
-60
-40
-20 0
20 40 60 REACTOR POWER IMBALANCE CORE PROTECTION SAFETY LIMITS UNITS 1, 2, AND 3 Figure 2.1-2 OCONEE NUCLEAR STATION OCONEE -
UNITS 1, 2, & 3 2.1-5 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature,
- flow, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protective action to prevent any combination of process variables from exceeding a safety limit.
Specification The reactor protective system trip setpoints and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.
The pump monitors shall produce a reactor trip when a loss of two pumps occurs and reactor power level is greater than 0.0% of rated power.
Bases The reactor trip setpoints for reactor protective system (RPS) instrumentation are given in Table 2.3-1.
The trip setpoints have been selected to ensure that the core and reactor coolant system are prevented from exceeding their safety limits.
The various reactor trip circuits automatically open the reactor trip breakers whenever a parameter monitored by the RPS deviates from an allowed range. The RPS consists of four instrument channels for redundancy.
The plant safety analyses are based on the trip setpoints given in Table 2.3-1 plus calibration and instrumentation errors.
Nuclear Overpower A r-actor trip at high power level (nPutron flux) is provided to prevent damage to the fuel cladding from reactivity exciirsions too rapid to be detected by pressure. and temperature measurnmrnts.
Du1ring normal plant operation with all reactor coolant pumps operating, a
rectr-ri inp is initLated whlin the reactor power level reaches i 0 T of rnatod power.
Adding to this the possible variation in the trip setpoint due to
',lihibrain and instrument orrors, the maximum actniit power at which a
t p would le actuated conld 1b 112%,
which is the va hin used in the safr analyvsis.
(1)
OCONEE -
UNITS 1, 2, & 3 2.3-1 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
Overpower Trip Based on Flow and Imbalance Following the loss of one or more reactor coolant pumps, the core is prevented from violating the minimum DNBR criterion by a reactor trip initiated by exceeding the allowable reactor power to reactor coolant flow (flux/flow) ratio setpoint.
Loss of one or more reactor coolant pumps is also detected by the pump monitors.
The power level trip produced by the flux/flow ratio provides DNB protection for all modes of pump operation.
The power level trip setpoint produced by the flux/flow ratio provides both high power level and low flow protection.
For every flow rate there is a
maximum permissible power level, and for every power level there is a minimum permissible flow rate. Typical power level and flow rate combinations for different pump situations are as follows:
- 1.
Trip would occur when four reactor coolant pumps are operating if power is 107,0 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 79.9% and reactor flow rate is 74.7% or flow rate is 70.09% and power level is 75%.
The analysis to determine the flux/flow setpoint accounts for calibration and instriment errors and the variation in RC flow in such a manner as to ensure a conservative setpoint.
Statistical methods are used to determine the combined effects of calibration and instrument uncertainties with the final string uncertainties used in the analysis corresponding to the 95/95 tolerance limits.
The reactor power imbalance (power in the top half of the core minus the power in the bottom half) reduces the power level trip produced by the flux/flow ratio as shown in Figure 2.3-2.
The flux/flow ratio reduces the power level trip and associated power-imbalance boundaries by 1.07% for a 1% flow reduction.
The power-imbalance boundaries shown in Figure 2.3-2 are established to prevent fuel thermal limits, DNBR and centerline fuel melt limits, from being exceeded.
Pump Monitors The pump monitors trip the reactor due to the loss of reactor coolant pump(s) to ensure the DNBR remains above the minimum allowable DNBR.
The pump monitors provide redundant trip protection of DNB; tripping the reactor on a signal diverse from that of the flux/flow trip. The pump monitors also res.trict the power level depending on the number of operating reactor coolant pumps.
OCONEE -
UNITS 1, 2, & 3 2.3-2 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
Reactor Coolant System Pressure During a startup accident from low power or a slow rod withdraw from high
- power, the reactor coolant system (RCS) high pressure setpoint is reached before the nuclear overpower trip setpoint.
The high RCS pressure trip setpoint (2355 psig) ensures that the pressure remains below the safety limit (2750 psig) for any design transient. (2) The low pressure (1800 psig) and variable low pressure (11.14 Tout -
4706) trip setpoints shown in Figure 2.3-1 ensure that the minimum DNBR is greater than or equal to minimum allowable DNBR for those accidents that result in a reduction in pressure. (3,4)
The limits shown in Figure 2.3-1 bound the pressure-temperature curves calculated for 4 and 3 pump operation.
Accounting for calibration and instrumentation errors, the safety analyses used a variable low RCS pressure trip setpoint of (11.14 T 4756).
out Coolant Outlet Temperature The high reactor coolant outlet temperature trip setpoint (618 0 F) shown in Figure 2.3-1. has been established to prevent excessive core coolant temperatures.
Accounting for calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 0 F.
Reactor Building Pressure The high reactor building pressure trip setpoint (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.
Shutdown Bypass In order to startup the reactor and to he able to perform control rod drive tests and zero power physics tests (see Technical Specification 3.1.9), there is provision for bypassing certain segments of the reactor protective system (RPS).
The RPS segmenLs which can be bypassed are given in Table 2.3-1.
Two conditions are imposed when the RPS is bypassed:
- 1.
By administrative control the nuclear overpower trip setpoint is reduced to a value of < 5.0% of rated power.
- 2.
The high reactor coolant system pressure trip setpoint is automatically lowered to 1720 psig.
The high RCS pressure trip setpoint is reduced to prevent normal operation with part of the RPS bypassed. The reactor must be tripped before the bypass is initiated since the high pre sure trip setpoint is lower than the normal low pressure trip setpoint (1800 psig).
The overpower trip setpoint of < 5.0, prevents any significant reactor power from being produced when performing physics tests.
If no reactor coolant pumps are oerating, sufficient natural circulation would be available to remove 5.0' of rated power.(S)
OCONEE - UNITS 1, 2, & 3 2.3-3 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
REFERENCES (1) FSAR, Section 15.3 (2) FSAR, Section 15.2 (3) FSAR, Section 15.7 (4) FSAR, Section 15.8 (5) FSAR, Section 15.6 OCONEE -
UNITS 1, 2, & 3 2.3-4 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
THERMAL POWER LEVEL %
120
(-17.0.107.01 (17.0.107.0)
Ml =0.944 ACCEPTABLE 4 P UMP OPERATI9 M2=-0.944 100 9
(-35.0,90.0)
(35.0,90.0)
- 17. 0. 9. 9)
(17:0.79.9)
ASEPTABLE 3 & 4 PUMP OPERA ON
(-35.0,62.9) 60-(35.0,62.9) 40 20
-40
-20 0
20 40 REACTOR POWER IMBALANCE, %
PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS UNITS 1. 2. AND 3 Fi-ure 2.3-2 OCONEE NUCLEAR STATION 2.3-6 OCONEE UNITS 1, 2, & 3 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
TABLE 2.3-1 Reactor Protective System Trip Setting Limits Shutdown RPS Trip RPS Trip Setpoint Bypass
- 1.
Nuclear Overpower 105.5% Rated Power 5.0%
Rated Power (1)
- 2.
Flux/Flow/Imbalance 1.07 Bypassed
- 3.
Pump Monitors
> 0% Rated Power loss Bypassed of two pumps
- 4.
High Reactor Coolant 2355 psig 1720(2)
System Pressure
- 5.
Low Reactor Coolant 1800 psig Bypassed System Pressure
- 6.
Variable Low Reactor P (psig) = (11.14 T Bypassed Coolant System Pressure 4706)(3) out
- 7.
High Reactor Coolant 618 0 F 618 0 F Temperature
- 8.
High Reactor Building 4 psig 4 psig Pressure (1) Administratively controlled reduction set only during reactor shutdown.
(2) Automatically set when other segments of the RPS are bypassed.
(3) TOut i.s in degrees Fahrenheit (OF).
OCONEE -
UNITS 1, 2, & 3 2.3-7 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system.
Objective To specify those limiting conditions for operation of the reactor coolant system components which must be met to ensure safe reactor operation.
Specification 3.1.1 Operational Components
- a.
Reactor Coolant Pumps
- 1.
Whenever the reactor is critical, one and two pump operation shall be prohibited, single-loop operation shall be restricted to testing, and other pump combinations permissible for given power levels shall be as shown in Table 2.3-1.
- 2.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one low pressure injection pump is circulating reactor coolant.
- b.
- 1.
One steam generator shall be operable whenever the reactor coolant average temperature is above 250 0F.
- c.
Pressurizer Safety Valves
- 1.
All pressurizer code safety valves shall be operable whenever the reactor is critical.
- 2.
At least one pressurizer code safety valve shall be operable whenever all reactor coolant system openings are closed, except for hydrostatic tests in accordance with the ASME Section III Boiler and Pressure Vessel Code.
OCONEE -
UNITS 1, 2, & 3 3.1-1 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
Bases A reactor coolant pump or low pressure injection pump is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One low pressure injection pump will circulate the equivalent of the reactor coolant system volume in one-half hour or less.
(1)
The low pressure injection system suction piping is designed for 300oF and 370 psig; thus the system with its redundant components can remove decay heat when the reactor coolant system is below this temperature.
(2,3)
One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.
(4) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities.
The code safety valves prevent overpressure for a rod withdrawal accident at hot shutdown. (5) The pressurizer code safety valve lift setpoint shall be set at 2500 psig +/-1% allowance for error and each valve shall be capable of relieving 300,000 lb/hr of saturated steam at a pressure no greater than 3% above the set pressure.
REFERENCES (1) FSAR, Section 6.3.3.2, and Tables 5.3-1, 5.4-2, 5.4-3, 5.4-6, 5.4-7, 5.4-8 and 6.3-2.
(2) FSAR, Sections 5.4.7-1 and 9.3.3.2.3.
(3) FSAR, Sections 5.4.7.4 and 6.3.3.2 (4) FSAR, Sections 5.2.3.10.4 and 5.4.6.
(5)
FSAR, Sections 5.2.3.7 and 15.2.3.
OCONE UNTS 1
& 33.1-2.
OCONEE UNITS 1, 2, & 3 32Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
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 combina tion.
- 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 thermal 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 exceeded, 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 shall be maintained within acceptable operating limits for regulating rod position provided in the CORE OPERATING LIMITS REPORT for the particular number of operating reactor coolant pumps (4,3).
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.
3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics tests, imbalance shall be maintained within the acceptable operating limits for reactor power imbalance provided in the CORE OPERATING LIMITS REPORT.
OCONEE -
UNITS 1, 2, & 3 3.5-9 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)
6.9 CORE OPERATING LIMITS REPORT Specification 6.9.1 Core operating limits shall be established prior to each reload cycle, or prior to any remaining part of a reload cycle, for the following:
(1) Power Dependent Rod Insertion Limits for Specifications 3.1.3.5, 3.5.2.2.d.2.c, 3.5.2.3, and 3.5.
2.5.c.
(2)
Power Imbalance Limits for Specification 3.5.2.6 and shall be documented in the CORE OPERATING LIMITS REPORTS.
6.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically:
(1) DPC-NE-1002A, Reload Design Methodology II, October 1985.
(2) NFS-1001A, Reload Design Methodology, April 1984.
(3) DPC-NE-2003A, Oconee Nuclear Station Core Thermal Hydraulic Methodology Using VIPRE-01, July 1989.
6.9.3 The core operating limits shall be determined such that all applic able limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
OCONEE -
UNITS 1, 2, and 3 6.9-1 Amendment No. 180 (Unit 1)
Amendment No. 180 (Unit 2)
Amendment No. 177 (Unit 3)