ML17263A929
| ML17263A929 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/06/1995 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17263A930 | List: |
| References | |
| NUDOCS 9502100331 | |
| Download: ML17263A929 (28) | |
Text
gy,fl REC0 0
Cy ClO IV)0
~O
++*++
UNITED STATES NUCLEAR REGULATORY COMMlSSION WASHINGTON, D.C. 2055&4001 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
GINNA NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
58 License No.
DPR-18 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 May 13,
- 1994, as supplemented June 24 and September 27,
- 1994, 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; 2.
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.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-18 is hereby amended to read as follows:
9502i0033i 905000244 PDR ADQCK 0 P
(2).
Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
58
, 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 and shall be implemented within 60 days, or upon NRC approval and implementation of the licensee's guality Assurance
- Plan, Revision 20, whichever is later.
FOR THE NUCLEAR REGULATORY COMMISSION g
Z-Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
February 6,
1995
ATTACHMENT TO LICENSE AMENDMENT NO. 58 C
I OPER ING DOCKET NO.
ICENSE NO.
DPR-18 50-244 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 11 3.1-21 3.5-2a 3.6-3 3.16-2 6.2-1 6.2-2 6.2-3 6.2-4 6.4-1 6.5-1 6.5-2 6.5-3 6.5-4 6.5-4a 6.5-5 6.5-6 6.5-7 6.5-8 6.5-8a 6.5-9 6.5-10 6.5-11 6.5-12 6.6-1 6.7-1 6.8-1 6.8-2 6.9-4 6.9-6 6.9-7 6.10-1 6.10-2 6.10-3 6.11-1 6.13-1 6.13-2 6.15-1 6.16-1 6.17-1 Insert ll 3.1-21 3.5-2a 3.6-3 3.16-2 6.2-1 6.2-2 6.2-3 6.4-1 6.5-1 6.6-1 6.7-1 6.8-1 6.9-4 6.9-6 6.9-7 6.10-1 6.11-1 6.13-1 6.13-2 6.15-1 6.16-1 6.17-1
4.8 4.9 4'0 4'1 4.12 4.13 4 14 4.15 4'6 TABLE OF CONTENTS (Cont'd)
Auxiliary Feedwater Systems Reactivity Anomalies Environmental Radiation Survey Refueling Effluent Surveillance Radioactive Material Source Leakage Test Shock Suppressors (Snubbers)
Deleted Overpressure Protection System Pacae 4.8-1 4'-1 4.10-1 4.11-1 4.12-1 4.13-1
- 4. 14-3
- 4. 15-1 4.16-1 5.0 DESIGN FEATURES 5 ~ 1 5.2 5.3 5 '5.5 Site Containment Design Features Reactor Design Features Fuel Storage Waste Treatment Systems 5 ~ 1-1 5.2-1 5.3-1 5.4-1 5.5-1 6.0 ADMINISTRATIVECONTROLS 6.1 6.2 6.3 6 '6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 6.13 6'4 6.15 6.16 6.17 Responsibility Organization 6.2.1 Onsite and Offsite Organization 6.2.2 Facility Staff Station Staff Qualifications Training (Deleted)
(Deleted)
Safety Limit Violation Procedures Reporting Requirements 6.9.1 Routine Reports 6.9.2 Unique Reporting Requirements (Deleted)
(Deleted)
(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 6.6-1 6.7-1 6.8-1 6.9-1 6.9-1 6.9-3
- 6. 10-1 6.11-1 6.12-1 6.13-1 6.14-1 6.15-1 6.16-1 6.17-1 Amendment PPi
$ 9,58
~
~
3.1.4 Maximum Coolant Activit S ecifications 3 ~ 1.4
~ 1
'a ~
b.
co 3.1.4.2 Whenever the reactor is critical or the reactor coolant average temperature is greater than 500oF:
The total specific activity of the reactor coolant shall not exceed 84/E pCi/gm, where E is the average beta and gamma energies per disintegration in Mev.
The I-131 equivalent of the iodine activity in the reactor coolant shall not exceed 0.2 pCi/gm.
The I-131 equivalent of the iodine activity on the secondary side of a steam generator shall not exceed 0.1 pci/gm.
If the limit of 3.1.4.1.a is
- exceeded, then be subcritical with reactor coolant average temperature less than 500'F within.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
3 ~ 1 ~ 4 ~ 3 a ~
If the I-131 equivalent activity in the reactor coolant exceeds the limit of 3.1.4.1.b but is less than the allowable limit shown on Figure 3.1.4-1, operation may continue for up to 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />.
Amendment No.
g7,58 3.1-21
3.5.5.2 If the setpoint for a radioactive effluent monitor alarm and/or trip is found to be higher than required, one of the following three measures shall be taken immediately:
(i) the setpoint shall be immediately corrected (ii) without declaring the channels inoperable; or immediately suspend the release of effluents 3.5.5 '
3'.6 3.5.6.1 monitored by the effected channels or (iii) declare the channel inoperable.
If the number of channels which are operable is found to be less than required, take the action shown in Table 3.5-5.
Exert best efforts to return the instruments to OPERABLE status within 31 days and, if unsuccessful,.
explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
Control Room HVAC Detection Systems During all modes of plant operation, detection systems for chlorine gas, ammonia gas and radioactivity in the control room HVAC intake shall be operable with setpoints to isolate air intake adjusted as follows:
Amendment No. pp,58 3.5-2a
~Bas s:
The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the reactor coolant system ruptures.
The shutdown margins are selected based on the type of activities that are being carried out.
The (2000 ppm) boron concentration provides shutdown margin which precludes criticality under any circumstances.
When the reactor head is not to be removed, a cold shutdown margin of 14~k/k precludes criticality in any occurrence.
Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major steam break accident were as much as 1 psig."'he containment is designed to withstand an internal vacuum of 2s5 psig.+
The 2.0 psig vacuum is specified as an operating limit to avoid any difficulties with motor cooling.
In order to minimize containment leakage during a
design basis accident involving a
significant fission product
- release, penetrations not required for accident mitigation are provided with isolation boundaries.
These isolation boundaries consist of either passive devices or active automatic valves and are listed in a
procedure under the control of the Quality Assurance Program.
Closed manual valves, deactivated automatic valves secured in their closed position (including check valves with flow through the valve secured),
blind flanges and closed systems are considered passive devices.
Automatic isolation valves designed to close following an accident without operator action, are considered active devices.
Two isolation devices are provided for each mechanical penetration, such that no single credible fai'lure or malfunction of an active component can cause a loss of isolation, or result in a
leakage rate that exceeds limits assumed in the safety analyses+.
In the event that one isolation boundary is inoperable, the affected penetration must be isolated with at least one boundary that is not affected by a single active failure.
Isolation boundaries that meet this criterion are a closed and deactivated automatic containment isolation valve, a closed manual valve, or a blind flange.
The opening of closed containment isolation valves on an intermittent basis under administrative control includes the following considerations:
(1) stationing an individual qualified in accordance with station procedures, who is in constant communication with the control room, at the valve controls, (2) instructing this individual to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to isolate the boundary and that this action will prevent the release of radioactivity outside the containment.
Amendment No. gp,
$ $,58
- 3. 6-3
6.9-2 when averaged over any calendar quarter, a Special Report shall be submitted to the Commission within thirty days which includes an evaluation of any release conditions, environmental factors or other aspects which caused the reporting levels of Table 6.9-2 to be exceeded.
When more than one of the radionuclides in Table 6.9-2 are detected in the sampling medium, this report shall be submitted if:
concentration 1
+
concentration 2
+
....> 1.0 limit level (1) limit level (2)
When radionuclides other than those in Table 6.9-2 are detected and are the result of plant effluents, this'eport shall be submitted if the potential annual dose to an individual is greater than the calendar year limit of Specifications 3.9.1.2.a or 3.9.2.2.b. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
3.16.1.4 If milk or fresh leafy vegetable'amples are unavailable for more than one sample period from one or more of the sampling locations indicated by the
- ODCM, a discussion shall be included in the Radioactive Effluent Release Report which identifies the cause of the unavailability of samples and identifies locations for Arnnnknan+
Mn CA
6.2 6 ' '
0 G NIZATION site a d Offsite Or a 'zatio An onsite and an offsite organization shall be established for unit operation and corporate management.
The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.
a ~
Lines of authority, responsibility and b.
communication shall be established and defined from the highest management levels through intermediate levels to and including all Plant management positions.
Those relationships shall be documented and
- updated, as appropriate, in the form of organization charts.
These organization charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71.
The Senior Vice President, Customer Operations, shall have corporate responsibility for overall Plant nuclear safety, and shall take any measures needed to assure acceptable performance of the staff in operating, maintaining, and providing technical support in the Plant so that continued nuclear safety is assured..
An alternate title may be designated for this position in accordance with 10 CFR 50.54(a)(3).
All requirements of these Technical Specifications apply to the position with the alternate title as, apply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report.
Amendment No. )P,)g,gg,58 6.2-1
c ~
d.
The Plant
- Manager, Ginna Station shall have responsibility for overall unit operation and shall have control over those resources necessary for safe operation and maintenance of the Plant.
The persons responsible for the training, health physics and quality assurance functions may report to an appropriate manager
- onsite, but shall have direct access to responsible corporate management at a
level where action appropriate to the mitigation of training, health physics and quality assurance concerns can be accomplished.
6'.2 ac't Staf The Facility organization shall include the following:
a 0 b.
co An auxiliary operator shall be assigned to the shift crew with fuel in the reactor.
An additional auxiliary operator shall be assigned to the shift crew above Cold Shutdown.
At least one licensed operator shall be present in the control room when fuel is in the reactor.
In
- addition, above Cold
- Shutdown, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
Shift crew composition may be less than the minimum requirements of 10 CFR 50.54(m)(2)(i) and Specifications 6.2.2.a and 6.2.2.f 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 crew members provided immediate action is taken to restore shift crew composition to within the minimum requirement.
Amendment No. gp, g8
- 6. 2-2
d.
e.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
Adequate, shift coverage shall be maintained without routine heavy use of overtime.
Administrative procedures 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.
The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.
The STA shall be assigned to the shift crew above Cold Shutdown.
An alternate title may be designated for this position.
All requirements of these Technical Specifications apply to the position with the alternate title as a'pply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report.
Amendment No.
6.2-3
6.4 6.4.1 6.4 '
Q~INING A retraining and replacement training program for the facilitystaff shall be maintained under the direction of the Division Training Manager 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.
The training program shall meet or exceed NFPA No.
27, 1975 Section 40, except that (1) training for salvage operations need not be provided and (2) the Fire Brigade training sessions shall be held at least quarterly.
Drills are considered to be training sessions.
An alternate title may be designated for this position.
All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report.
Amendment No.
pg >8 6.4-1
- 6. 5 (Deleted)
(Intentionally Left Blank)
Amendment No.
$,$ P,Pg,jl@
PPiPPiPP'
6.6 (Deleted)
(Intentionally Left Blank)
Amendment No. ),PP,PP,58
6.7 6.7.1 SA ETY LIMIT VIOLATION The following actions shall be taken in the event a
Safety Limit is violated:
a ~
b.
c ~
d.
The provisions of 10 CFR 50.36(c) (1) (i) (A) shall be complied with immediately.
The Safety Limit violation shall be, reported to the Senior Vice President, Customer Operations, to the offsite review
- function, and to the NRC immediately.
A Safety Limit Violation Report shall be prepared.
The report shall be reviewed by the onsite review function. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
The Safety Limit Violation Report shall be submitted to the NRC, the offsite review function, and the Senior Vice President, Customer Operations within two weeks of the violation.
An alternate title may be designated for this position in accordance with 10 CFR 50.54(a)(3).
All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report.
Amendment No. Pg,gg
~~8
- 6. 7-1
6.8 6.8.1 PROCEDURES Written procedures shall be established, implemented, and maintained covering the following activities:
a ~
b.
c ~
d.
e.
The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Fire Protection Program implementation.
The radiological environmental monitoring program.
Offsite Dose Calculation Manual implementation.
Process Control Program implementation.
Amendment No. )$,58
- 6. 8-1
and directions from the reactor, and the results of the participation in an interlaboratory comparison program.
6.9.1.4 adioactive Effluent Release Re ort Routine radioactive effluent release reports covering the operation of the unit during the previous twelve months of operation shall be submitted by May 1 of each year.
This report shall include a
- summary, on a quarterly
- basis, of the quantities of radioactive liquid and gaseous effluents and solid waste released as outlined in Regulatory Guide 1.21, Revision 1.
This report shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each of the previous four calendar quarters as outlined in Regulatory Guide 1.21, Revision 1.
Zn addition, the site boundary maximum noble gas gamma air and beta air doses shall be evaluated.
The assessment of radiation doses shall be performed in accordance with the ODCM.
This same report shall include an annual summary of hourly meteorological data collected over the previous calendar year.
Alternatively, the licensee has the option of retaining this summary on site in a file that shall be provided to the NRC upon request.
Also, the report shall include any nearby location(s) identified by the land use census which Amendment No.
58 6.9-4
6.9 '
6 ' ' '
U ue Re o tin Re uirements Annually: Results of required leak test performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.
6.9.2.2 Annually: A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions, e.g.,
reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance),
waste processing, and refueling.
The dose assignment to r
various duty functions may be estimates based on pocket dosimeter,
- TLD, or film badge measurements.
Small exposures totalling less than 204 of the individual total dose need not be accounted for.
Zn the aggregate, at least 804 of the total whole body dose received from external sources shall be assigned to specific major work functions.
(NOTE:
This tabulation supplements the requirements of Section 20.407 of 10CFR Part 20) 6.9.2.3 (Deleted)
Amendment No.
$ 7~~8 6'-6
6.9.2.4 Reactor Overpressure Protection System Operation In the event either the PORVs or the RCS vent(s) are used to mitigate a
RCS pressure transient, a Special Report shall be prepared and submitted to the Commission within thirty days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent(s) on the transient and any other corrective action necessary to prevent recurrence.
Amendment No.
$ 7 58
- 6. 9-7
(Deleted)
(Intentionally Left Blank)
Amendment No. g,gP,)7>58
(Deleted)
(Intentionally Left Blank)
Amendment No.
58
- 6. 13 6.13 F 1 G
IO Zn lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR Part 20:
a ~
Each High Radiation Area in which the intensity of radiation is 1000 mrem/hr or less shall be barri-caded and conspicu'ously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a
Radiation Work Permit (RWP).
Any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:
(1)
A
. radiation monitoring device which con-tinuously indicates the radiation dose rate in the area.
(2)
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received.
Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.
Radiation Protection personnel shall be exempt: from the RWP issuance requirement during the performance of their assigned radiation protection
- duties, providing they are following plant radiation protection procedures for entry into high radiation areas.
An alternate title may be designated for this position.
All requirements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report.
Amendment No.
58 6 ~ 13-1
(3)
A Qualified health physicist (i.e., qualified in radiation protection procedures) with a
radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and who" will perform periodic radiation surveillance at the frequency specified in the HPHP.
The surveillance frequency will be established by a plant Pealthghysicist b.
Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.13.1 a.
- above, and in addition locked doors shall be provided to prevent unauthorized entry into these areas and the keys to unlock these locked doors shall be maintained under the administrative control of the Shift Supervisor on duty.
An alternate title may be designated for this position.
All requi'rements of these Technical Specifications apply to the position with the alternate title as apply with the specified title.
Alternate titles shall be specified in the Updated Final Safety Analysis Report..
Amendment No. )g,58
- 6. 13-2
Offsite Dose Calculation Manual (ODCM)
Any changes to the ODCM shall be made by the following method:
6.15.1.a Licensee initiated changes shall be submitted to the Commission with the Radioactive Effluent Release Report for the period in which the change(s) was made and shall contain:
(i) sufficiently detailed information to support the rationale for the change.
(ii) a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and (iii) documentation of the fact that the change has been reviewed and found acceptable by the onsite review function.
6.15.1.b Licensee initiated changes shall become effective after review and acceptance by the onsite review function on a date specified by the licensee.
Amendment No.
58
6.16 6'6 F 1 ocess Control Pro ram (PCP)
Any changes to the PCP shall be made by the following method:
6.16.1.a Licensee initiated changes shall be submitted to the Commission with the Radioactive Efflu'ent Release Report for the period in which the change(s) was made and shall contain:
(i) sufficiently detailed information to support the rationale for the change; (ii) a determination that the change will not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and (iii) documentation of the fact that the change has been reviewed and found acceptable by the onsite review function.
6.16.1.b Licensee initiated changes shall become effective after review and acceptance by the onsite review function on a date specified by the licensee.
Amendment No.
58 6.16-1
6.17 es to Radioactive Waste Treatment S ste s
(Liquid, Gaseous and Solid) 6'7 2
The radioactive waste treatment systems (liquid, gaseous and solid) are those systems defined in Technical Specification 5.5.
Major changes to the radioactive waste systems (liquid and gaseous) shall be reported by the following method.
For the purpose of this specification, "major changes" is defined in Specification 6.17.3 below.
6.17.2.1 The Commission shall be informed of all major changes by the inclusion of a suitable discussion or by reference to a suitable discussion of each change in the Radioactive Effluent Release Report for the period-in which the changes were made.
The discussion of each" change shall contain:
a) a summary of the evaluation that led to the determination that the change could be made (in accordance with 10 CFR 50.59);
b) sufficient detailed information to support the reason for the change; c) a detailed description of the equipment, components and processes involved and the interfaces with other plant systems; Amendment No.
58