ML14183A236
| ML14183A236 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 12/07/1992 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A235 | List: |
| References | |
| DPR-23-A-142 NUDOCS 9212140222 | |
| Download: ML14183A236 (13) | |
Text
gt REGU 1, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 142 License No. DPR-23
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Carolina Power & Light Company (the licensee), dated February 21, 1992, as supplemented August 7, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i).
that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulationsand all applicable requirements have been satisfied.
- 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-23 is hereby amended to read as follows:
9212140222 921207 PDR ADOCK 05000261 P
-2 B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 142, are hereby incorporated in the license.- Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
- 3.
License Condition 3.E. of Facility Operating License No. DPR-23 is hereby amended to read as follows:
E. Fire Protection Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Report, dated February 28, 1978, and supplements thereto. Carolina Power & Light Company may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
- 4.
This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION linor G. Adensam, Director Project Directorate II-1 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 7, 1992
ATTACHMENT TO LICENSE AMENDMENT NO. 142 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Pages Insert Pages ii ii iii iii 1-4 1-4.
3.14-1 3.14-1 3.14-2 3.14-3 3.14-4 3.14-5 3.14-5a 3.14-6 3.14-7 3.14-8 4.1-12 4.1-12 4.14-1 4.14-1 4.14-2 4.14-3 4.14-4 6.2-4 6.2-4 6.4-1 6.4-1 6.5-7 6.5-7 6.9-9 6.9-9
Section Title Page 3.9.3 Compliance with 10 CFR.Part 20 - Radioactive Materials in Gaseous Effluents 3.9-3 3.9.4 Compliance with 10 CFR Part 50 - Radionoble Cases 3.9.4 3.9.5 Compliance with 10 CFR Part 50 - Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides other than Radionoble Gases 3.9-5 3.9.6 Compliance with 40 CFR Part 190 - Radioactive Effluents from Uranium Fuel Cycle Sources 3.9-6 3.10 Required Shutdown Margins, Control Rods, and Power Distribution Limits 3.10-1 3.10.1 Full Length Control Rod Insertion Limits 3.10-1 3.10.2 Power Distribution Limits 3.10-2 3.10.3 Quadrant Power Tilt Limits 3.10-7a 3.10.4 Rod Drop Time 3.10-8 3.10.5 Deleted 3.10.6 Inoperable Control Rods 3.10-8 3.10.7 Power Ramp Rate Limits 3.10-9 3.10.8 Required Shutdown Margins 3.10-9 3.11 Movable In-Core Instrumentation 3.11-1 3.12 Seismic Shutdown 3.12-1 3.13 Shock Suppressors (Snubbers) 3.13-1 3.14 DELETED 3.14-1 3.15 Control Room Filter System 3.15-1 3.16 Radioactive Waste Systems 3.16-1 3.16.1 Liquid Radwaste Treatment System 3.16-1 3.16.2 Liquid Holdup Tanks 3.16-1 3.16.3 Gaseous Radwaste and Ventilation Exhaust Treatment Systems 3.16-2 3.16.4 Waste Gas Decay Tanks (Hydrogen and Oxygen) 3.16-3 3.16.5 Waste Gas Decay Tanks (Radioactive Materials) 3.16-5 3.16.6 Solidification of Wet Radioactive Waste 3.16-6 3.17 Radiological Environmental Monitoring Program 3.17-1 3.17.1 Monitoring Program 3.17-1 3.17.2 Land Use Census 3.17-3 3.17.3 Interlaboratory Comparison Program 3.17-4 4.0 Surveillance Requirements 4.1-1 4.1 Operational Safety Review 4.1-2 4.2 Primary System Surveillance 4.2-1 4.3 Primary System Testing Following Opening 4.3-1 4.4 Containment Tests 4.4-1 4.4.1 Operational Leakage Rate Tests 4.4-1 4.4.2 Isolation Valve Tests 4.4-4 ii Amendment No. 142
Section Title 4.4.3 Post Accident Recirculation Heat Removal System 4.4-4 4.4.4 Operational Surveillance Program 4.4-5 4.5 Emergency Core Cooling, Containment Cooling and Iodine Removal Systems Tests 4.5-1 4.5.1 System Tests 4.5-1 4.5.2 Component Verification 4.5-2 4.6 Emergency Power System Periodic Tests 4.6-1 4.6.1 Diesel Generators 4.6-1 4.6.2 Diesel Fuel Tanks 4.6-2 4.6.3 Station Batteries 4.6-2 4.7 Secondary Steam and Power ConversioniSystem 4.7-1 4.8 Auxiliary Feedwater System 4.8-1 4.9 Reactivity Anomalies 4.9-1 4.10 Radioactive Effluents 4.10-1 4.10.1 Radioactive Liquid Effluents 4.10-1 4.10.2 Radioactive Gaseous Effluents 4.10-2 4.10.3 Radionoble Gases 4.10-2 4.10.4 Radioiodines, Radioactive Materials in Particulate Form, and Radionuclides Other Than Radionoble Gases 4.10-3 4.10.5 Radioactive Effluents From Uranium Fuel Cycle Sources 4.10-3 4.11 Reactor Core 4.11-1 4.12 Refueling Filter Systems 4.12-1 4.13 Shock Suppressors (Snubbers) 4.13-1 4.14 DELETED 4.14-1 4.15 Control Room Filter System 4.15-1 4.16 Radioactive Source Leakage Testing 4.16-1 4.19 Radioactive Effluent Instrumentation 4.19-1 4.19.1 Radioactive Liquid Effluent Instrumentation 4.19-1 4.19.2 Radioactive Gaseous Effluent Instrumentation 4.19-1 4.20 Radioactive Waste Systems 4.20-1 4.20.1 Liquid Radwaste Treatment System 4.20-1 4.20.2 Liquid Holdup Tanks 4.20-1 4.20.3 Gaseous Radwaste and Ventilation Exhaust Treatment System 4.20-2 4.20.4 Waste Gas Decay Tanks (Hydrogen and Oxygen) 4.20-2 4.20.5 Waste Gas Decay Tanks (Radioactive Material) 4.20-3 4.20.6 Solidification of Wet Radioactive Waste 4.20-3 4.21 Radiological Environmental.Monitoring Program 4.21-1 4.21.1 Monitoring Program 4.21-1 4.21.2 Land Use Census 4.21-1 4.21.3 Interlaboratory Comparison Program 4.21-2 5.0 Design Features 5.1-1 5.1 Site 5.1-1 5.2 Containment 5.2-1 5.2.1 Reactor Containment 5.2-1 5.2.2 Penetrations 5.2-1 5.2.3 Containment Systems 5.2-2 iii Amendment No. 142
- a.
All non-automatic containment isolation valves not required for normal operation are closed and blind flanges are properly installed where required.
- b.
The equipment door is properly closed and sealed.
- c.
At least one door in the personnel air lock is properly closed and sealed.
- d.
All automatic containment isolation trip valves required to be closed during accident conditions are operable or are secured closed except as stated in Specification 3.6.3. Manual valves qualifying as automatic containment isolation valves are secured closed.
- e.
The uncontrolled containment leakage satisfies Specification 4.4.
1.8 QUADRANT POWER TILT The quadrant power tilt is defined as the ratio of maximum to average of the upper excore detector currents or the lower excore detector currents, whichever is greater. If one excore is out of service, the three in-service units are used in computing the average.
1.9 DELETED 1.10 STAGGERED TEST BASIS A Staggered Test Basis shall consist of:
- a.
A test schedule for n systems, subsystems, trains or designated components obtained by dividing the specified test interval into n equal subintervals.
1-4 Amendment No.
142
PAGES 3.14-1 THROUGH 3.14-8 HAVE BEEN DELETED.
(NEXT PAGE IS 3.15-1) 3.14-1 Amendment No. 142
TABLE 4.1-3 FREQUENCIES FOR EQUIPMENT TESTS Maximum Time Between Check Frequency Tests
- 1.
Control Rods Rod drop times of Each refueling NA*
all full length shutdown rods
- 2.
Control Rod Partial movement Every 2 weeks during 20 days of all full length reactor critical rods operations
- 3.
Pressurizer Set point Each refueling shutdown NA Safety Valves 4.,
Main Steam Set point Each refueling shutdown NA Safety Valves
- 5.
Containment Iso-Functioning Each refueling shutdown NA lation Trip
- 6.
Refueling System Functioning Prior to each refueling NA Interlocks shutdown
- 7.
Service Water Functioning Each refueling shutdown NA System
- 8.
DELETED
- 9.
Primary System Evaluate Daily when reactor NA Leakage coolant system is above cold shutdown condition
- 10.
Diesel Fuel Fuel Inventory Weekly 10 days Supply
-11.
Critical Headers. 100 Psig Hydro-Every five years 6 years of Auxiliary static Test Coolant System
- 12.
Turbine Steam Closure Monthly during power 45 days Stop, Control, operation and prior Reheat Stop, to startup and Interceptor Valves 4.1-12 Amendment No. 142
PAGES 4.14-1 THROUGH 4.14-4 HAVE BEEN DELETED.
(NEXT PAGE IS 4.15-1) 4.14-1 Amendment No. 142
f)
ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised'by either a licensed Senior Reactor Operator or Senior Reactor Operator limited to fuel handling who has no other concurrent responsibilities during this operation.
g)
DELETED h)
The shift complement may be one less than the minimum requirement of Section 6.2.3.a and 6.2.3.b for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift members provided immediate action is taken to restore the shift complement to within the minimum requirements of Section 6.2.3.a and 6.2.3.b. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member being late or absent.
6.2-4 Amendment No. 142
6.4 TRAINING 6.4.1 A retraining and replacement training program for the plant staff shall be maintained under the direction of the Manager - Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 DELETED 4
6.4-1 Amendment No.
142
6.5.1.6.5 A quorum of the PNSC shall consist of the Chairman, and four members, of which two may be alternates.
6.5.1.6.6 The PNSC activities shall include the following:
a)
Perform an overview of Specifications 6.5.1.1 and 6.5.1.2 to assure that processes are effectively maintained.
b)
Performance of special reviews, investigations, and reports thereon requested by the Manager - Nuclear Assessment Department.
c)
Annual review of the Security Plan and Emergency Plan.
d)
Perform reviews of Specifications 6.5.1.1.6, 6.5.1.2.4, 6.5.1.3.1, and 6.5.1.4.1.
e)
Perform review of all reportable events.
f)
Review of facility operations to detect potential nuclear safety hazards.
g)
Review of every unplanned on site release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrences to the Vice President - Robinson Nuclear Project, Manager - Nuclear Assessment Department.
h)
Review of changes to the Process Control Program and the Offsite Dose Calculation Manual.
i)
Review-of major. changes to radioactive liquid, gaseous, and solid waste.
treatment systems.
j)
Review of changes to the CORE OPERATING LIMITS REPORT.
k)
Annual review of the Fire Protection Program, including Program changes.
6.5-7 Amendment No. 142
6.9.2 Deleted 6.9.3 Special Reports 6.9.3.1 Spe-cial reports shall be submitted to the NRC within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
Area Reference Submittal Date a)
Containment Leak Rate 4.4 Upon completion of each test.
Testing b)
Containment Sample 4.4 Upon completion of the Tendon Surveillance inspection at 25 years of operation.
c)
Post-Operational 4.4 Upon completion of the test Containment Structural at 20 years of operation.
Test d)
DELETED:
e)
Overpressure Protection 3.1.2.1.e Within 30 days of operation.
System Operation f)
Auxiliary Feedwater 3.4 Within 30 days after becoming Pump inoperable.
6.9-9 Amendment No. 142