ML17055B088
| ML17055B088 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/07/1986 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17055B089 | List: |
| References | |
| NUDOCS 8601210237 | |
| Download: ML17055B088 (34) | |
Text
UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.76 License No.
DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated April 26, 1985, 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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-63 is hereby amended to read as follows:
860i2i0237 860i070 PDR ADOCN, 05000220 I
PDR 1
(2) Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
76, 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 the date of its issuance.
FOR T NUCLEAR REGULAT RY C MMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
January Z, 1986 John A. Zwolinski, Director BWR roject Directorate 81 Division of BWR Licensing
ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO.
DPR-63 DOCKET NO. 50-220 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 111 129 130 132 133 134 164a 164b 164c 164d 190 232b 232c 236 237 241ff 241hh 241ii INSERT 111 129 130 132 133 134 190 232b 232c 236 237 24lff 24lhh 24lii
"t
SECTION 3.3.4 Isolation Valves 3.3.5 Access Control 3.3.6 Vacuum Relief 3.3.7 Containment Spray DES CRPTION 4.3.s Isolation Valves 4.3.5 Access Control 4.3.6 Vacuum 'Relief 4.3.7 Containment Spray 144 150 152 158
~
3.4.0 Secondary Containment Limitin Condition for 0 eration 3.4.1 Leakage Rate 3.4.2 Isolation Valves 3.4.3 Access Contro1 3.4.4 Emergency Ventilation 3.4.5 Control Rooy Ventilation Surveillance Re uirements 4.4.1 4.4.2 L'eakage Rate
-Isolation Valves 4.4.3 Access Control 4.4.4 Emergency Ventilation 4.4.5 Control
.".oom Rentil ation 165 166 1.69 171
~ 173 17v 3.5.0 Shutdown and Refueling Limitin Condit' for eration 3.5.1 Source Range Monitoring 3.5.2 Refuelin'g Platform Interlock 3.6.0 General Reactor Plant Limiting Condition for Operation 3.6.1 Station Process Effluents 3.6.2 Pro tecti ve Instrumentati on 3.6.3 Emergency Power Sources Amendment No.
25~
76 Surveillance Re uirements 4.5.1 Source Range Monitoring 4.5.2 Refueling Platform Interlock Survei 1 1 ance Requi rements 4.6.1 Station Process Effluents 4.6.2 P rote'cti ve Instrumentation 4.6.3 Emergency Power Sources 179 100 LB' 185 186 lBS 238
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL 4.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL App 1 icab i 1 ity:
Applies to the interrelated parameters of pressure suppression system pressure and suppression chamber water temperature and level.
Objective:
To assure that the peak suppression chamber pressure does not exceed design values in the event of a loss-of-coolant accident.
App 1 icab i 1 ity:
Applies to the periodic testing of the pressure suppression system pressure and suppression chamber water temperature and level.
Objective:
To assure that the pressure suppression system pressure and suppression chamber water temperature and level are within required limits.
Specif ication:
Specification:
a.
The downcomers in the suppression chamber shall have a minimum submergence of three feet and a
maximum submergence of four and one half feet whenever the reactor coolant system temperature is above 215F.
b.
During normal power operation, the combination of primary containment pressure and suppression chamber
'ulk pool temperature shall be within the shaded area of a.
At least once per day the suppression chamber water level and temperature and pressure suppression system pressure shall be checked.
b.
A visual inspection of the suppression chamber interior, including water line regions, shall be made at each major refueling outage.
129 Amendment No. gg, 76
I
LIMITING CONDITION FOR OPERATION SURVE ILLANCE REQUIREMENT C.
d.
e.
(1) Figure 3.3.2a when downcomer subm.rgence is greater than or equal to 4 feet, or (2) Figure 3.3.2b when downcomer submergence is greater than or equal to 3 feet but less than 4 feet.
If these temperatures are exceeded, pool cooling shall be initiated immediately.
If Specifications a and b above are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be shut down using normal shutdown procedures.
During testing of relief valves which add heat to the torus pool, bulk pool temperature shall not exceed 10F above normal power operation limit specified in b above.
In connection with such tes ting, the pool temperature res t be reduced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to below the normal power operation limit specified in b above.
The reactor shall be scrammed from any operating condition when the suppression pool"bulk temperature reaches 110F.
Operation shall not be resumed until the pool temperature is reduced to below the normal power operation limit specified in b above.
I c.
d.
e.
Whenever heat 'from relief valve operation is being added to the suppression pool, the pool temperature shall be continually monitored and also observed and logged every 5 minutes until the heat addition is terminated.
Whenever operation of a relief valve is indicated and the bulk suppression pool temperature reaches 160F or above while the reactor primary coolant system pressure is greater than 200 psig, an external visual examination of the suppression chamber shall be made before resuming normal power operation.
Whenever there is indication of relief valve operation with the local temperature of the suppression pool reaching 200F or
- more, an external visual examination of the suppression chamber shal'l be conducted before resuming power operation.
During reactor isolation conditions, the reactor pressure vessel shall be depressurized to 'less than 200 psig at.
normal cooldown rates if the pool bulk temperature reaches 120F.
Amendment No.
25 76 130
, ~
FIGURE 3.3.2 a
ALLOWABLE PRESSURE SUPPRESSION SYSTEM 4
FOOT DOWNCOMER SUBMERGENCE 5.0 4.0 C)
I C?:
La>
OC) 3.0 2.0 1.0 0
68 72 76 80 84 88 92 96 '00 SUPPRESSION CHAMBER WATER OPERATIONAL TEMPERATURE F
Amendment No. gg, 76 132
FIGURE 3.3.2 b
ALLOWABLE PRESSURE SUPPRESSION SYSTEM 3
FOOT OOWNCOMER SUBMERGENCE 5.0 CO L/l O
I LLI CJ L/l L/)
LJJ CL CL
<<C I-CY LJJ 0C) 4.0 3.0 2.0 LJJ CY L/)
L/l LJJ CL 0
1.0 0
68
?2
?6 80 84 88 92 96 100 SUPPRESSION CHAt1BER HATER OPERATIONAL TEMPERATURE F
Amendment No. NJ 76 133
%l
BASES FOR 3.3.2 AND 4.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL The values specified for suppression chamber water temperature, maximum downcomer submergence, and system pressures are based on the effect these parameters have on the short-term post-accident system pressure following a loss-of-coolant accident.
The combinations shown on Figures 3.3.2 a
and b and the water level required are based on maintaining the post-accident pressure below the design value of 35 psig and the maximum suppression chamber water temperature below 140F in the containment design basis loss-of-coolant accident (Appendix E-11.2.2.3).*
The calculational basis for the pressure suppression system initial conditions, Figures 3.3.2 a and b are presented in the Fifth Supplement.*
The three foot minirmm and the four and one-half foot maxirmm submergence are a result of the Mark I Containment Long Term Program.
The 215F li'mit for the reactor is specified, since below this temperature the containment can tolerate a blowdown without exceeding the 35 psig desi gn pressure of the suppression chamber without condensation.
Actually, for reactor temperatures up to 312F the containment can tolerate a blowdown without exceeding the 35 psig design pressure of the suppression
- chamber, without condensation.
Some experimental data suggests that excessive steam condensing loads might be encountered if the bulk temperature of the suppression pool exceeds 160F during any period of relief valve operation with sonic conditions at the discharge exit.
This can result in local pool temperatures in the vicinity of the quencher of 200F.
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 act~
to be taken in the event of a relief valve inadvertently opens or sticks open.
As a mininam, this action would include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor
- shutdown, and (4') if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat
- addition, the temperature trends will be closely followed so that appropriate action can be taken.
The requirement for an external visual examination following any event where potentially high loadings
- FSAR 134 Amendment No. 2. 76
h
~
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREMENT (8)
(9)
(10)
(12)
(13)
Off-Gas and Vacuum Pump Isolation-The respective system shall be isolated orthe instrument channel shall be considered inoperable and Specification 3.6.1 shall be applied.
Diesel Generator Initiation The diesel generator shall be considered inoperable and Specification 3.6.3 shall be applied.
Emergency Venti 1 ati on Initiati on The emergency venti.lation system shall be considered inoperable and Specification 3.4.4 shall be applied.
High Pressure Coolant Injection Initiation The high pressure coolant injection system shall be considered inoperable and Specification 3.1.8.c shall be applied.
Primary Containment Monitoring The primary containment moni toring instrumentation shall be considered inoperable and Specification 3.3.2.
shall be applied.
Control Room Ventilation The control room ventilation system shall be considered inoperable and Specification 3.4.5 shall be applied.
b.
During operation with a Maximum Total Peaking Factor (MTPF) greater than the design value, either:
Amendment No. 78, 76 190
TABLE 3.6.21 PRIMARY CONTAINMENT MONITORING Limitin Condition for Operation Parameter Minimum No.
of Tripped or Operable Trip Systems Minimum No. of Operable Instrument Channels Per Operable Trip System Setpoint Reactor Mode Switch Position in Which Function Must Be Oper able
( 1) Suppression Chamber Water Level Specification 3.3.2 X
X
~
Amendment f3o. g5,76 232b
TAOLE 4.6.21 PRIMARY CONTAINMENT MONITORING Survei 1 1 ance Requirement Parameter (1) Suppression Chamber Water Level Sensor Check once/day Instrument Channel Test N/A Instrument Channel Calibration Once Every Six Months Amendment Ho. g5, 76 232c
0
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points on the generator load rejection and turbine stop valve closure scram trips are set to anticipate and minimize the consequences of turbine trip with failure of the turbine bypass system as described in the bases for Specification 2.1.2.
Since the severity of the transients is dependent on the reactor operating power level, bypassing of the scrams below the specified power level is permissible.
The primary containment monitoring system is provided to alert the operator of conditions which could reduce safety i
margi'ns during a postulated Loss of Coolant Accident.
Appropriate operator corrective action is described in Specification 3.3.2, should Limiting Conditions for Operation be exceeded.
This monitoring instrumentation does not automatically initiate engineered safeguards systems.
Although the operator wi 11 set the setpoints at the values indicated in Tables 3.6.2.a-l, the actual values of the various set points can differ appreciably from the value the operator is attempting to set.
The deviations include inherent instrument error, operator setting error and drift of the set point.
These errors are compensated for in the transient analyses by conservatism in the controlling parameter assumptions as discussed in the bases for Specification
- 2. 1.2.
The deviations associated with the set'oints for the safety systems used to mitigate accidents have negligible effect on the initiation of these systems.
These safety systems have initiation times which are orders of magnitude greater than the difference in time between reaching the nominal set point and the worst set point due to error.
The maximum allowable set point deviations are listed below:
Neutron Flux APRM, +2.7X of rated neutron flux
- IRM,
+2.5X of rated neutron. flux Recirculation Flow,
+1% of rated recirculation flow Reactor Pressure,
+15.8 psig Containment Pressure,
+0.053 psig Reactor Water Level, +2.6 inches of water Main Steam Line Isolation Valve Position,
+2.5X of stem position Scram Oischarge
- Volume,
<0 and -1 gallon Condenser Low Vacuum, +0.5 inches of mercury Amendment Ho. gS, 76 236
4
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION High Flow-Main Steam Line,
+1 psid High Flow-Errergency Cooling Line,
+1 psid High Area Temperature-Main Steam Line,
+10F High Area Temperature-Clean-up and Shutdown,
+6F High Radiation-Main Steam Line, +100K and -50K of set point value High Radiation-Emergency Cooling System Vent,
+100K and -50K of set point High Radiation-Reactor Building Vent, +10(C and -5(C of set point High Radiation-Refueling
- Platform,
+100% and -50K of set point High Radiation-Offgas
- Line,
+50K of set point, (Appendix D)*
I Suppression Chamber Water Level, +1.8 inches The test intervals for the trip systems result to calculated failure probabilities 410-4 which corresponds to the proposed IEEE Criteria for System Failure Probability.
(IEEE SG-3, Information Doc%et1 - Protection System Reliability, April 24, 1968).
The test intervals for the trip systems result in calculated failure probabilities ranging from 6.7 x 10 to 1.76 x 10-10 (Fifth Supplement,
- p. 115).*
The more frequent sensor checks result in even less probability that the particular system wi 11 fail.
Because of local high radiation, testing instrumentation in the area of the main ste~
line isolation valves can only be done during periods of Station shutdown.
These functions include high area temperature isolation, high radiation isolation and isolation valve position scram.
Testing of the scram associated with the shutdown position of the mode switch can be done only during periods of Station shutdown since it always involves a scram.
"FSAR Amendment No. gg, 76 237
TABLE 3.6.11-1 ACC IDENT MONITORING I NS RUMENTATION Parameters Total Number of Channels Minimum Number of Operable Sensors or Channels Action (See Table 3.6.11-2) 1)
Relief Valve Position Indication 2)
Safety Valve Position Indication 3)
Drywe11 Pressure Monitor 5)
Suppression Chamber Water Level 6)
Containment Hydrogen Monitor 7)
Containment High Range Radiation Monitor 8)
Suppression Chamber Water Temperature 2/Valve 2/Val ve 1/Valve 1/Val ve Amendment Ho.
7g>
76 241ff
TABLE 4.6.11 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENT Parameter Instrument Channel Test Instrument Channel Calibration (1) Relief valve position indicator (Primary - Acoustic)
Relief valve position indicator (Backup - Thermocouple)
(2) Safety valve position indicator (Primary - Acoustic)
Once per month Once per month Once per month Once during each major refueling outage Once during each major refueling outage Once during each major refueling outage Safety valve position indicator (Backup - Thermocouple)
(3) Reactor vessel water level (4) Drywell Pressure Monitor (5) Suppression Chamber Water Level Monitor (6) Containment Hydrogen Monitor (7) Containment High Range Radiation Monitor Once per month Once per month Once per month Once per month Once per month Once per month Once during each major refueling outage Once during each major refueling outage Once during each major refueling outage Once during each major refue1ing outage Once per quarter Once during each major refueling outage (8) Suppression Chamber Water Temperature Once per month Once during each major refueling outage Amendment No. 7g, 76 24lhh
t
BASES 3.6.11 and 4.6.11 ACCIOENT MONITORING INSTRUMENTATION Accident monitoring instrumentation ensures that sufficient information is available on'elected plant parameters to monitor and assess these variables during and following an accident.
This capability is consistent with the recoranendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" and/or NUREG-0737, "Clarification of TMI Action Plan Requirements,"
November 1980 and NUREG-0661, "Safety Evaluation Report Mark I Containment Long Term Program."
Amendment Ho. 7g, 76 241i i