ML17262B007
| ML17262B007 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/21/1992 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17262B008 | List: |
| References | |
| NUDOCS 9209280280 | |
| Download: ML17262B007 (15) | |
Text
~S RE0y fy I
flO c
n +w*w+
l UNITED STATES t
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
E.
GINNA NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No.
DPR-18 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Rochester Gas and Electric Corporation (the licensee) dated April 21,
- 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2),
and 2.C.(3) of Facility Operating License No.
DPR-1& are hereby amended to read as follows*:
"Pages 3 and 4 of Facility Operating License No.
DPR-18 are also attached for convenience for the composite license to reflect these changes.
Please remove pages 3 and 4 of the existing license and replace with the attached pages.
9209280280 92092l PDR
- DOCK 05000244 P
.'DR +
~
~
(2)
Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
49
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Fire Protection (a)
The licensees shall implement and maintain in effect all fire protection features described in the licensee's submittals referenced in and as approved or modified by the NRC's Fire Protection Safety Evaluation (SE) dated February 14, 1979 and SE supplements dated December 17,
- 1980, February 6,
- 1981, June 22,
- 1981, February 27, 1985 and March 21, 1985 or configurations subsequently approved by the NRC, subject to provision (b) below.
(b)
The licensee 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.
(c)
Deleted
- 3. This license amendment is effective as of its date of issuance, and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION Attachments:
1.
Changes to the Technical Specifications 2.
Pages 3 and 4 of Facility Operating License NPR-18 Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Date of Issuance:
September 21, 1992
ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO.
DPR-18 DOCKET NO. 50-244 Replace the following pages of the License with the attached pages.
The revised pages contain vertical lines indicating the area of change.
Remove Insert Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 1ll 1-5 3.14-1 3.14-2 3.14-3 3.14-4
'3.14-5 3.14-6 3.14-7 3.14-8 3.14-9 3.14-10 4.15-1 4.15-2 4.15-3 4.15-4 4.15-5 6.2-3 6.5-4 1ll
~
1-5 3.14-1 4.15-1 6.2-3 6.5-4
(5) pursuant to the Act and 10-CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and. 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
(1)
Maximum Power Level RG&E is authorized to operate the facility at steady-state power levelsup to a maximum of 1520 megawatts
( thermal).
(2)
Technical S ecifications The Technical Specifications contained in Appendix A,are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Fire Protection (a)
The licensee shall implement and maintain in
(
effect all fire protection features described in the licensee's submittals referenced in and as approved or modified by the NRC's Fire Protection Safety Evaluation (SE) dated February 14, 1979 and SE supplements dated December 17, 1980, February 6, 1981, June 22, 1981, February 27, 1985 and March 21, 1985. or configurations subsequently approved by the NRC, subject to provision (b) be1ow.
(b)
The licensee 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.
Amendment No.
49
(c)
Deleted (4)
Secondar Water Chemistr Monitorin Pro ram The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation.
This program shall be described in the plant procedures and shall includ'e:
(a)
Identification of a sampling schedule for the critical parameters and control points for these parameters;,
(b)
Identification of the procedures used to measure the values of the critical parameters; (c)
Identification of process sampling points; (d)
Procedure for the recording and management of data; (e)
Procedures defining corrective actions for off control point chemistry conditions; and (f)
A procedure identifying (i) the authority responsible for the interpretation of the
- data, and (ii) the sequence and timing of administrative events required to initiate corrective action.
(5)
S stems Inte rit The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as, low as reasonably achievable levels.
This program shall include the following:
Amendment No.
49
f
1.0 DEFINITIONS 2 '
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 2.2 2.3 Safety Timit, Reactor Core Safety Limit, Reactor Coolant System Pressure Limiting Safety System Settings, Protective Instrumentation 3.0 LIMITING CONDITIONS FOR OPERATION PAGE 2.1-1 2.1-1 2.2-1 2.3-1 3.0 3.1 3.2 3.3 3.4 3.5 3.6 3.7 3.8 3.9 3.10 3'1 3'2 3.13 3.14 3.15 3.16 Applicability Reactor Coolant System 3.1. 1 Operational Components 3.1.2 Heatup and Cooldown 3.1.3 Minimum Conditions for Criticality
- 3. 1. 4 Maximum Coolant Activity
- 3. 1. 5 Leakage 3.1.6 Maximum Reactor Coolant Oxygen,
- Fluoride, and Chloride Concentration Chemical and Volume Control System Emergency Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan Coolers, Containment Spray and Charcoal Filters Turbine Cycle Instrumentation System Containment System Auxiliary Electrical Systems Refueling Plant Effluents Control Rod and Power Distribution Limits Fuel Handling in the Auxiliary Building Movable In-Core Instrumentation Shock Suppressors (Snubbers)
Deleted Overpressure Protection System Radiological Environmental Monitoring 3.0-1 3.1-1 3.1-1 3.1-5 3.1-19 3.1-21 3.1-25 3.1-31 3.2-1 3.3-1 3.4-1 3.5-1 3.6-1 3.7-1 3.8-1 3.9-1 3.10-1 3.11-1 3.12-1 3.13-1 3.14-1 {
3.15-1 3.16-1 4.0 SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review 4.2 Inservice Inspection 4.3 Reactor Coolant System 4.4 Containment Tests 4.5 Safety In)ection, Containment Spray and Iodine Removal Systems Tests 4.6 Preferred and Emergency Power System.
Periodic Tests 4.7 Hain Steam Stop Va1ves Amendment No. gg,g7, 49
-i-4.1-1 4.2-1 4.3-1 4.4-1 4'-1 4.6-1 4.7-1
TABLE OP CONTENTS (CONT' 4.8 4.9 4.10 4.11 4.12 4.13 4.14 4.15 4.16 Auxiliary Feedwater System Reactivity Anomalies Environmental Radiation Survey Refueling Effluent Surveillance Radioactive Material Source Leakage Test Shock Suppressors (Snubbers)
Deleted Overpressure Protection System PAGE 4.8-1 4.9-1 4.10-1 4.11-1 4.12-1 4.13-1 4.14-1 4.15-1
)
4.16-1 5.0 DESIGN FEATURES 5.1 Site 5.2 Containment Design Features 5.3 Reactor Design Features 5.4 Fuel Storage 5.5 Waste Treatment Systems 6.0 ADMINISTRATIVECONTROLS
- 5. 1-1 5.2-1 5.3-1 5.4-1 5.5-1 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 6.13 6.14 6.15 6.16 6.17 Responsibility Organization 6.2.1 Onsite and Offsite Organization 6.2.2 Facility Staff Station Staff Qualification Training Review and Audit 6.5.1 Plant Operation Review Committee (PORC) 6.5.2 Nuclear Safety Audit and Review Board (NSARB) 5.5.3 Quality Assurance Group Reportable Event Action Safety Limit Violation Procedures Reporting Requirements 6.9.1 Routine Reports 6.9.2 Unique Reporting Requirements Record Retention Radiation Protection Program Deleted High Radiation Area Deleted Offsite Dose Calculation Manual Process Control Program Major Changes to Radioactive Waste Treatment Systems 6.1-1 6.2-1 6.2-1 6.2-2 6.3-1 6.4-1 6.5-1 6.5-1 6.5-5 6.5-11 6.6-1 6.7-1 6.8-1 6.9-1 6.9-1 6.9-3 6.10-1 6.11-1 6.13-1 6.15-1 6.16-1 6 ~ 17-1 Amendment No.
4, 88; 49
1.10 Hot Ch 1 Factors
~
~
F~,
Heat Flux Hot Channel
- Factor, is defined as the maximum local heat flux, on the surface of a fuel rod divided by the average fuel rod heat flux allowing for manufacturing tolerances on fuel pellets and rods.
F ~, Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density assuming nominal fuel pellet and rod dimensions.
F ~, Engineering Heat Flux Hot Channel factor, is defined as the ratio between F~ and F ~ and is the allowance on heat flux required for manufacturing tolerances.
F Nuclear Enthalphy Rise Hot Channel
- Factor, is defined as the ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.
1.11 (DELETED)
Amendment No.
49 1-5
~
~
3.14 (V LETE INTENTXONALLYLEFT BLANK Amendment No. g, 4g 3.14-1
~
~
~
~
1 5
! w~-, ETET '~
INTENTIONALLYLEFT BLANK Amendment No.
4g 4.15-1
(DELETED
~
t Adequate shift coverage shall be maintained without routine heavy use of overtime.
Administrative procedures I
I shall be developed and implemented. to limit the working hours of unit staff who perform safety-related functions including senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.
Changes to the guidelines for the administrative procedures shall be submitted to the NRC for review.
Amendment No.
7'g-,gg,gg -q9 6.2-3
~
~
PESPONSIBILI~IES (Continued) h.
Review of the Plant Security Plan and shall submit recommended changes to the Chairman of I
the Nuclear Safety Audit and Review Board.
i.
Review of the Radiation Emergency Plan
- and, shall submit recommended changes to the Chairman of the Nuclear Safety Audit and Review Board.
j.
'Review of implementing procedures for the Plant Security Plan and the Radiation Emergency Plan and proposed changes thereto.
k.
Review of all Reportable Events.
1.
Review of the Fire Protection Program and.
Implementing Procedures and submittal of recommended Program changes to the Chairman of the Nuclear Safety Audit and Review Board (NSARB).
AUTHORITY 6.5.1.7 The PORC shall:
a ~
b.
Recommend in writing to the Plant
- Manager, Ginna Station approval or disapproval of items considered under 6.5.1.6(a) through (d) and (l) above.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (d) and (1) above constitutes an unreviewed safety question as defined in 10 CFR Section 50.59.
Amendment No.
C.,88,, 49 6.5-4