ML20134B798
| ML20134B798 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/28/1997 |
| From: | Stang J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134B800 | List: |
| References | |
| NUDOCS 9701310091 | |
| Download: ML20134B798 (12) | |
Text
_._.
e
@ City p
UNITED STATES
}
NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 2006H1001 o
4.....g COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE
]
Amendment No. 152 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated January 13, 1997, resubmitted January 17, 1997 and supplemented January 22, 1997, 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 t'e 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 t',s 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 Specifi-cations as indicated in the attachment to this license amendment and paragraphs 2.C.(2) and 2.C.(6) of Facility Operating License No. DPR-19 are hereby amended to read as follows:
- License pages 3 and 3a are provided, for convenience, for the composite license to reflect this change.
9701310091 970128 PDR ADOCK 05000237 P
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 152, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(6)
By Amendment No. 152, the license is amended to allow the licensee to change the Updated Final Safety Analysis Report to allow credit for two psig containment pressure to compensate for a slight increase in the amount of Net Positive Suction Head (NPSH) deficiency during the first 10 minutes following a design basis accident (DBA).
3.
This license amendment is effective as of thc date of its issuance and shall be implemented within 30 days.
FOR HE NUCLEA REGULATORY COMMISSION r--
ahn F. Stang, Senior Project Manager roject Directorate Ill-2 ivision of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attachments:
1.
License pages 3 and 3a 2.
Changes to the Technical Specifications Date of Issuance: January 23, 1997
l 1 '
1 j ;
(5) Comed, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.
i C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, i
regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level i
The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2527 megawatts thermal i
.(100 percent rated power) in accordance with the conditions specified herein.
(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 152, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
4 (3) Operation in the coastdown mode is permitted to 40% power.
i (4) The valves in the equalizer piping between the recirculation loops
{
shall be closed at all times during reactor operation.
1
{
(5) The licensee shall maintain the commitments made in response to the i
March 14,1983, NUREG-0737 Order, subject to the following provision:
The licensee may make changes to commitments made in response to the i
March 14,1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC i
approval, pursuant to the requirements of 10 CFR 50.59. Consistent j
with this regulation, if the change results in an Unreviewed Safety i
Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.
(6) By Amendment No.152, the license is amended to allow the licensee to change the Updated Final Safety Analysis Report to allow credit for two psig containment pressure to compensate for a slight increase in the amount of Net Positive Suction Head (NPSH) deficiency during the first 10 minutes following a design basis accident (DBA).
Amendment No. 152
i
- 3a -
4 D.
The facility has been granted certain exemptions from the requirements of Section III.G of Appendix R to 10 CFR Part 50, " Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage.
These exemptions were granted and sent to the licensee in letters dated February 2, 1983, September 28, 1987, July 6, 1989, and August 15, 1989.
I In addition, the facility has been granted certain exemptions from Sections II and III of Appendix J to 10 CFR Part 50, " Primary Reactor J
Containment Leakage Testing for Water-Cooled Power Reactors." This section contains leakage test requirements, schedules and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment.
These exemptions were granted and sent to the licensee in a letter dated June 25, 1982.
i a
i Amendment No. 152
p it It UNITED STATES y;
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20ssNm01'
\\...../
COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-249 4
DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 147 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated January 13, 1997, resubmitted January 17, 1997 and supplemented January 22, 1997, 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 regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraphs 3.B. and 3.0. of Facility Operating License No. DPR-25 are hereby amended to read as follows:
- License page 6 is provided, for convenience, for the composite license to reflect this change.
. B.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 147, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
1 0.
By Amendment No.147, the license is amended to allow the licensee to change the Updated Final Safety Analysis Report to allow credit i
a for two psig containment pressure to compensate for a slight increase in the amount of Net Positive Suction Head (NPSH) deficiency during the first 10 minutes following a design basis accident (DBA).
t 3.
This license amendment is effective as of the date of its issuance and l
shall be implemented within 30 days.
4 F0 THE NUCLEAR REGULATORY COMMISSION M
& F2
'l ohn F. Stanf, Senior Project Manager Project Direc.torate III-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attachments:
- 1. License page 6
- 2. Changes to the Technical Specifications Date of Issuance: January 28, 1997 l
d 6
i W
. - 4
~
L.
Deleted.
[Amdt. 87, 7-24-86]
M.
Deleted.
[Amdt. 85, 12-12-85]
N.
By Amendment No. 144, the license is t. mended to allow, on a i
one time temporary basis, operation of Dresden, Unit 3, with the corner room structural steel members in the Low Pressure i
Coolant Injection Corner Rooms outside the Updated Final i
Safety Analysis Report (UFSAR) design parameters. Operation
~
under these conditions is allowed up to and including the next scheduled refueling outage (D3R14).
i i
The repairs to Dresden, Unit 3, corner room structural steel shall restore the steel design margins to the current UFSAR (updated through Revision IA) design criteria.
The design of the modifications to the Dresden, Unit 3, corner room 1
structural steel members will be based on use of elastic section modules and the structural steel stresses will be limited to 1.6 of the American Institute of Steel Construction (AISC allowables). The modifications to i
Dresden, Unit 3, corner room structural steel will be implemented during the upcoming D3R14 refueling outage.
During this interim period of operation, should vibratory ground motion exceeding the UFSAR Operating Basis Earthquake (OBE) design parameters, Dresden, Unit 3, will be shut down for inspection and will not start up without prior NRC approval.
0.
By Amendment No. 147, the license is amended to allow the licensee to change the Updated Final Safety Analysis Report to allow credit for two psig containment pressure to compensate for a slight increase in the amount of Net Positive Suction Head (NPSH) deficiency during the first 10 minutes following a design basis accident (DBA).
4.
This license is effective as of the date of issuance and shall expire at Mid-night January 12, 2001.
FOR THE ATOMIC ENERGY COMMISSION Original Signed By:
Peter A. Morris, Director Division of Licensing
Enclosures:
Appendix A - Technical Specifications Date of Issuance: January 12, 1971 Amendment No. 147
ATTACHMENT TO LICENSE AMENDMENT NOS. 152 AND 147 FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
REMOVE IN.Sffi 3/4.7-16 3/4. -16 3/4.7-17 3/4.7-17 B 3/4.7-6 B 3/4.7-6 3/4.8-5 3/4.8-5
i CONTAINMENT SYSTEMS Suppression Chamber 3/4.7.K 3.7 - LIMITING CONDITIONS FOR OPERATION 4.7 - SURVElLLANCE REQUIREMENTS K.
Suppression Chamber K.
Suppression Chamber The suppression chamber shall be The suppression chamber shall be j
OPERABLE with:
demonstrated OPERABLE:
(
1.
The suppression pool water level 1.
By verifying the suppression pool water i
between 14' 6.5" and 14' 10.5",
level to be within the limits at least 4
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 2.
A suppression pool maximum everage
{
\\
water. temperature of s75*F during 2.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying OPERATIONAL MODE (s) 1 or 2, except the suppression pool average water that the maximum average temperature temperature to be s75'F, except:
l I
may be permitted to increase to; a.
At least once per 5 minutes during l
a.
s85*F during testing which testing which adds heat to the adds heat to the suppression suppression pool, by verifying the pool.
suppression giool average water temperature to be sB5*F.
l l
b.
s100*F with THERMAL POWER s1% of RATED b.
At least once per hour when THERMAL POWER.
suppression pool average water temperature is 2 75'F, by verifying: l l
c.
s110*F with the main steam line isolation valves closed
- 1) Suppression pool average following a scram.
water temperature to be s 100*F, and l
3.
A total leakage between the suppression chamber and drywell of
- 2). THERMAL POWER to be s 1%
less than the equivalent leakage of RATED THERMAL POWER through a 1 inch diameter orifice at a after suppression pool average differential pressure of 1.0 psid.
water temperature has exceeded 75'F for more than l
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
APPLICABILITY:
c.
At least once per 30 minutes with OPERATIONAL MODE (s) 1,2 and 3.
the main steam isolation valves i
closed following a scram and suppression pool average water ACTION:
temperature > 75'F, by verifying l
suppression pool average water 1.
With the suppression pool water level temperature to be s110*F.
l outside the above limits, restore the water level to within the limits DRESDEN - UNITS 2 & 3 3/4.7-16 Amendment Nos. 152 & 147
CONTAINMENT SYSTEMS Suppression Chamber 3/4.7.K 3.7 - LIMITING CONDITIONS FOR OPERATION 4.7 - SURVEILLANCE REQUIREMENTS within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT 3.
Deleted.
SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the 4.
Deleted.
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
At least once per 18 months by 2.
In OPERATIONAL MODE (s) 1 or 2 with conducting a drywell to suppression the suppression pool average water chamber bypass leak test at an initial l
temperature > 75*F, except as differential pressure of 1.0 psid and permitted above, restore the average verifying that the measured leakage is l
temperature to 475'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> within the specified limit. If any or reduce THERMAL POWER to 51%
drywell to suppression chamber bypass RATED THERMAL POWER within the leak test fails to meet the specified next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, limit, the test schedule for subsequent tests shall be reviewed and approved 3.
With the suppression pool average by the Commission. If two consecutive l
water temperature > 85'F during tests fail to meet ~the specified limit, a testing which adds heat to tha test shall be performed at least every suppression pool, except as permitted 9 menths until two consecutive tests above, stop all testing which adds heat meet the specified limit, at which time to the suppression pool and restore the the 18 month test schedule may be l
average temperature to s75'F within resumed.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce THERMAL POWER to 51% RATED THERMAL POWER within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.
With the suppression pool average l
water temperature > 100*F, immediately place the reactor mode switch in the Shutdown position and operate at least one low pressure coolant injection loop in the suppression pool cooling mode.
5.
With the suppression pool average l
water temperature > 110*F, depressurize the reactor pressure vessel to < 150 psig (reactor steam dome pressure) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
DRESDEN - UNITS 2 & 3 3/4.7-17 Amendment Nos. 152 & 147
I
+
CONTAINMENT SYSTEMS B 3/4.7 i
BASES discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
Under full power operating conditions, blowdown to the suppression chamber with an initial water temperature of 95'F results in a water temperature of approximately 145'F. This peak temperature is low enough to provide complete condensation via T-quencher devices. However, a maximum average suppression pool temperature of 75'F and approximately 2 psi of containment l
pressure is required to assure adequate net positive suction pressure for the ECCS pumps during l
the first 10 minutes followin0 certain analyzed accidents. No positive containment pressure is l
required to assure adequate net positive suction pressure for the ECCS pumps after the first 10 l
minutes.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak -
temperature of the suppression poolis maintained sufficiently low during any period of safety relief valve operation for T-quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings. In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety or relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety or relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety or relief valve to assure mixing and uniformity of energy insertion to the pool.
in conjunction with the Mark l Containment Shu.t Term Program, a plant unique analysis was i
performed which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping. The mai enance of a drywell-suppression chamber differential pressure and a suppression chamber water level corresponding to a downcomer submergence range of 3.67 to 4.00 feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
3/4.7.L Sunnressinn Chamber and Drvwell Sorav Following a Design Basis Accident (DBA), the suppression chamber spray function of the low pressure coolant injection (LPCI)/ containment cooling system removes heat from the suppression chamber air space and condenses steam. The suppression chamber is designed to absorb the i
sudden input of heat from the primary system from a DBA or a rapid depressurization of the reactor pressure vessel through safety or relief valves. There is one 100% capacity containment spray header inside the suppression chamber.
Periodic operation of the suppression chamber and drywell sprays may also be used following a DBA to assist the natural convection and diffusion mixing of hydrogen and oxygen when other ECCS requirements are met and oxygen concentration exceeds 4%. Sinco the spray system is a function of the LPCl/ containment cooling system, the loops will not be aligned for the spray function during normal operation, but all components required to operate for proper alignment must be OPERABLE.
DRESDEN - UNITS 2 & 3 8 3/4.7-6 Amendment Nos. 152 & 147
PLANT SYSTEMS UHS 3/4.8.C,
3.8 - LIMITING CONDITIONS FOR OPERATION 4.8 - SURVEILLANCE REQUIREMENTS C.
Ultimate Heat Sink The ultimate heat sink shall be OPERABLE The ultimate heat sink shall be determined with:
OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the average water temperature 1.
A minimum water level at cr above and water level to be within their limits.
elevation 500 ft Mean Sea Level, and 2.
An average water temperature of l
s75'F.
APPLICABILITY; OPERATIONAL MODE (s) 1, 2, 3, 4, 5 and *.
ACTION:
With the requirements of the above specification not satisfied:
1.
In OPERATIONAL MODE (s) 1,2 or 3, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
In OPERATIONAL MODE (s) 4 or 5 declare the diesel generator cooling water system inoperable and take the I
ACTION required by Specification l
3.8.B 4
3.
In OPEF.ATIONAL MODE *, declare the die %I generator cooling water system ir. operable and take the ACTION required by Specification 3.8.B. The provisions of Specification 3.0.C are not applicable.
When handling irradiated fuelin the secondary containment, during CORE ALTERATION (s), and operations with a potential to drain the reactor vessel.
DRESDEN - UNITS 2 & 3 3/4.8 5 Amendment Nos. 152 & 147