ML20057D743
| ML20057D743 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/29/1993 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20057D739 | List: |
| References | |
| NUDOCS 9310050290 | |
| Download: ML20057D743 (16) | |
Text
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NUCLEAR REGULATORY COMMISSION t
WASHINGTON. D.C. 20FII41001
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NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Northeast Nuclear Energy Company (the licensee), dated May 25, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
1 9310050290 930929 PDR ADOCK C5000245 P
PDR i
. 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-21 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 64, 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 issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION O
h F. Stolz, Director Pro ect Directorate I-Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 29, 1993
ATTACHMENT TO LICENSE AMENDMENT NO. 64 FACILITY OPERATING LICENSE NO. DPR-21 DOCKET NO. 50-245 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-6 3/4 1-6 3/4 1-7 3/4 1-7 3/4 1-8 3/4 1-8 3/4 2-2 3/4 2-2 3/4 2-9 3/4 2-9 3/4 6-18 3/4 6-18 3/4 6-19 3/4 6-19 3/4 7-17 3/4 7-17 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 B 3/4 1-4a B 3/4 2-3 B 3/4 2-3 8 3/4 2-3a
r TABLE 3.1.1 (Continued)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS Minimum Number of Operable Modes in which Function Inst. Channels Trip Function Trip Level Setting Must Be Operable Action
- per Trip (1)
REFUEL /
STARTUP/ HOT System SHUTDOWN (8,11)
STANDBY RUN 2
Turbine Condenser low 2 23 in. Hg. Vacuum X (3)
X (3)
X A or C Vacuum I
4 (6)
Main Steamline Isolation s 10% Valve Closure X (3)
X (3)
X
- A or C Valve Closure 2
Turbine Control Valve See Section 2.1.2 F X (4)
X (4)
X (4)
A or C Fast Closure 2
Turbine Stop Valve i 10% Valve Closure X (4)
X (4)
X (4)
'A or C Notes:
1.
There shall be two operable or tripped trip systems for each function.
2.
Permissible to bypass, with control rod block, for reactor protection system reset in REFUEL and SHUTDOWN positions of the reactor mode switch.
3.
Bypassed when reactor pressure is < 600 psig.
4.
Bypassed when first stage turbine pressure is less than that which corresponds to 50% rated reactor thermal power.
Millstone Unit 1 3/4 1-4 Amendment No. J/, ff, ES, gy,64 0094 i
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i TABLE 3.1.1 (Continued)
Reactor Protection System (Scram) Instrumentation Requirements Notes:
5.
IRM's are bypassed when mode switch is placed in RUN. The detector for each operable IRM channel shall be fully inserted until the associated APRM channel is operable and indicating at least 3/125 full scale.
6.
The design permits closure of any one valve without a scram being initiated.
7.
May be bypassed when necessary by closing the manual instrument isolation valve for scram of PS-1621 A through 0 during purging for containment inerting or deinerting.
8.
When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:
a.
Mode Switch in SHUTDOWN
~
b.
Manual Scram c.
Iligh Flux IRM d.
Scram Discharge Volume High Level e.
APRM Reduced High Flux 9.
Not required to be operable when primary containment integrity is not required.
- 10. With the mode switch in RUN position an inoperative trip function also requires an associated APRM "downscale alarm."
- 11. Trip functions are not required to be operable if all control rods are fully inserted, and either electrically or hydraulically disarmed in accordance with Specification 4.1.0.
Millstone Unit 1 3/4 1-5 Amendment No. J, JJ, JS, 5/,64 0094 m
W TABLE 4.1.1 SCRAM INSTRUMENTATION FUNCTIONAL TESTS HINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Instrument Channel Group (3)
Functional Test Minimum Frecuency (4)
High Reactor Pressure A
Trip Channel and Alarm (1)
High Drywell Pressure A
Trip Channel and Alarm (1)
Low Reactor Water level A
Trip Channel and Alarm (2)
(1) l High Water Level in Scram Discharge A
Trip Channel and Alarm (1)
Condenser Low Vacuum A
Trip Channel and Alarm (8)
(1)
Main Steam Line Isolation Valve Closure A
Trip Channel and Alarm (1)
Turbine Stop Valves Closure A
Trip Alarm (1)
Trip Channel and Alarm (1)
Turbine Control Valve Fast Closure A
Trip Channel and Alarm (6) (8)
(1)
Flow Biased High Flux APRM B
Trip Output Relays (6) (7) (8)
(1)
Reduced High Flux APRM B
Trip Output Relays (8)
Before each startup (5)
IRM C
Trip Channel and Alarm (5) (8)
Before each startup (5) l Mode Switch in SHUTDOWN A
Place Mode Switch in SHUTDOWN Each refueling outage
]1]stoneUnit1 3/4 1-6 Amendment No.64
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TABLE 4.1.1 (Continued)
SCRAM INSTRUMENTATION FUNCTIONAL TESTS HINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTS AND CONTROL CIRCUITS 5
Notes:
1.
Initially once per month until exposure hours (H as defined in Figure 4.1.1) is 2.0 x 10,
thereafter according to Figure 4.1.1, with an interval not less than one month nor more than three months.
2.
An instrument check shall be performed on low reactor water level once per day.
3.
A description of the three groups is included in the bases of this Specification.
4.
Functional tests are not required when the systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systems to an operable status.
5.
Maximum test frequency required is once per week.
6.
This test includes verification of time delay relay performance.
7.
This test includes verification of 90% setdown in 30 seconds or less.
8.
This instrumentation is excepted from the functional test definition.
The functional test will consist of injecting a simulated electrical signal into the measurement channel.
Millstone Unit 1 3/4 1-7 Amendment No. 64 0094
TABLE 4.1.2 SCRAM INSTRUMENTATION CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Calibration Minimum Calibration Instrument Channel Group (1)
Method Er_eguency (2)
APRM Output Signal (4)
B.
Heat Balance Once every 7 days APRM Flow Bias Trip B
Standard Current Source Once every 3 monthr APRM Reduced High Flux Trip B
Standard Current Source Once every 3 montk IRM B
Standard Current Source Refueling High Reactor Pressure A
Pressure Standard Every 3 months High Drywell Pressure A
Pressure Standard Every 3 months Low Reactor Water A
Delta Pressure Standard Every 3 months Condenser low Vacuum A
Vacuum Pump Every refueling Generator load Rejection A
Pressure Standard Every refueling i
High Water Level in Scram Discharge A
Water Level Every 3 months Notes:
1.
A description of the three groups is included in the bases of this Specification.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
If tests are missed, they shall be performed prior to returning the systems to an operable status.
3.
Maximum calibration frequency required is once per week.
4.
The heat balance method serves the calibration of the normal APRM high flux trip and the reduced APRM high flux trip.
Millstone Unit 1 3/4 1-8 Amendment No. 64 0094
TABLE 3.2.1 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS Minimum Number of Operable Instrument Channels Per Trip System (1)
Instruments Trio Level Settina Action (3) 2 Reactor Low Water 2 127 inches above top of active fuel A
2 Reactor low low Water 79 (+4-0) inches above top of active fuel A
2 (4)
High Drywell Pressure 5 2 psig A
2 (2) (5)
High Flow Main Steamline
$ 120% of rated steam flow B
2 of 4 in each of High Temperature Main 2 subchannels Steamline Tunnel 5 200*F B
g 2
Low Pressure Main 1 825 psig B
Steamlines 2
liigh Flow Isolation 164 inches 2 trip setting (water differential C
Condenser Line on steam line) 2150 inches.
44 inches 2 trip setting (water differential on water side) 2 35 inches.
(1) Whenever primary containment integrity is required, there shall be two operable or tripped trip systems for each function, except for low pressure main steamline which only need be available in the RUN position.
(2)
Per each steamline.
(3) Action:
If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriate actions listed below shall be taken:
A.
Initiate an orderly shutdown and have reactor in cold shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have reactor in HOT STANDBY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Close isolation valves in isolation condenser system.
(4) May be bypassed when necessary by closing the manual instrument isolation valve for PS-1621, A through D, during purging for containment inerting or deinerting.
(5) Minimum number of operable instrument channels per trip system requirement does not have to be met for a steamline if both containment isolation valves in the line are closed.
l l
l Millstone Unit No. 1 3/4 2-2 Amendment No. J/, ES, EJ 64 0095 m
+
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TABLE 4.2.1
?
MINIMUM TEST AND CALIBRATION FRE0VENCY FOR CORE COOLING INSTRUMENTATION ROD BLOCKS AND ISOLATIONS Instrument Functigl Calibration (2)
Instrugt Instrument Channel Test Check ECCS Instrumentation i
1.
Reactor Low-Low Water Level (1)
Once/3 Months 2.
Drywell High Pressure (1)
Once/3 Months 3.
Reactor Low Pressure (1)
.Once/3 Months (Valve Permissive) 4.
APR LP Core Cooling Pump Interlock (1)
Once/3 Months 3
5.
Containment Spray Interlock (1)
Once/3 Months 6.
Loss of Normal Power Relays Refueling Outage None 7.
Power Available Relays (1) (5)
None' 8.
Reactor High Pressure (1)
Once/3 Months 9.
Isolation Condenser Timer (1)
Once/3 Months
- 10. Auto Blowdown Timer Refueling Outage Refueling Outage Rod Blocks 1.
APRM Downscale (1) (3)
Once/3 Months (1) 2.
APRM Flow Variable (1) (3)
Once/3 Months (1) 3.
IRM Upscale (6)
(6)
(6) 4.
IRM Downscale (6)
(6)
(6)-
5.
RBM Upscale (1) (3)
Once/3 Months (1) 6.
RBM Downscale (1) (3)
Once/3 Months (1)-
7.
SRM Upscale (6)
(6)
(6) 8.
SRM Detector not in Startup Position
.(6)
(6)
(6) 9.
Scram Discharge Volume - Water Quarterly Refueling Outage Level High 10.
Scram Discharge Volume - Scram Trip Quarterly None Bypassed Containment Isolations 1.
Reactor Low Water Level
.(1)
Once/3 Months 2.
Reactor Low-Low Water Level (1)
Once/3 Mon'ths 3.
Drywell High Pressure (1)
Once/3 Months 4.
Steam Tunnel High Temperature Refueling Outage' Refueling Outage 5.
Steam Line High Flow (1)
Once/3 Months Once/ Day 6.
Steam Line Low Pressure (1)
Once/3 Months.
None-
-l Millstone Unit'1 3/4 2-9 Amendment No. JJ, JJ, 64:
0096 E
LIMITING CONDITION FOR OPERATION 3.6 PRIMARY SYSTEM B0UNDARY J.
Andensate Demineralizers 1.
Replacement of a condensate demineralizing resin charge shall occur before the unused capacity of the resin reaches a minimum value of 5 pounds as chloride ions.
2.
When the charge is replaced, the new anion resin shall have a minimum salt-splitting capacity of 1.2 milliequivalents per milliliter in the wet, chloride form.
3.
At least once condensate demineralizer influent conductivity instrument shall be operable.
4.
Whenever a demineralizer is on-line, the conductivity of either its effluent or the condensate-booster pump discharge shall be continuously monitored.
5.
Flow rate and/or integrating flow instrumentation shall be operable and recorded for each demineralizer.
K.
Mechanical Condenser "acuum Pumo 1.
The mechanical condenser vacuum pump shall be capable of being isolated and secured on a signal of main steam line high l
radioactivity whenever tFe main steam line isolation valves are open.
SURVEILLANCE RE0VIREMENT 4.6 PRIMARY SYSTEM BOUNDARY J.
Condensate Demineralizers 1.
The percent of the remaining ion exchange capacity of the anion resins shall be calculated and logged:
a.
Weekly when the influent conductivity is between 0.055 and 0.3 umho/cm; b.
Daily when the influent conductivity is equal to or greater than 0.3 umho/cm.
4 Millstone Unit 1 3/4 6-18 Amendment No. J, JE, ff, 64 0097
SURVEILLANCE REOUIREMENT (Continued) i 4.6 PRIMARY SYSTEM BOUNDARY 4.6.J.2 New samples of anion resin shall be analyzed for salt-splitting capacity as follows:
At least once per year or at each replacement, whichever is a.
longer, if resin is replaced with material of the same type.
b.
Prior to use in the condensate demineralizers if the type of anion resin changed.
K.
Mechanical Condenser Vacuum Pumo 1.
At least once during each operating cycle, verify automatic securing and isolation of the mechanical condenser vacuum pump.
2.
The main steam line radiation monitors shall be calibrated with a radioactive source at least once each operating cycle.
b Millstone Unit 1 3/4 6-19 Amendment No. J, U, M,64 0097
FOOTNOTES FOR TABLE 3.7.1 Key: 0 - Open C - Closed SC - Stays Closed GC - Goes Closed NA - Not Applicable Note:
Isolation groupings are as follows:
GROUP 1: The valves in Group 1 are closed upon any one of the following conditions:
1.
Reactor low-low water level (This signal also trips the reactor recirculation pumps.)
2.
Main steam line high flow.
3.
Main steam line tunnel high temperature.
4.
Main steam line low pressure.
GROUP 2: The actions in Group 2 are initiated by any one of the following conditions:
1.
Reactor low water level.
2.
High drywell pressure.
GROUP 3: Reactor low water level alone initiates the following:
1.
Shutdown cooling system isolation.
2.
Reactor head cooling isolation.
GROUP 4:
Isolation valves associated with the isolation condenser are closed upon indication of either high isolation condenser steam or condensate flow.
GROUP 5: Reactor low-low water level alone initiates the following:
1.
Clean-up demineralizer system isolation.
Millstone Unit 1 3/4 7-17 Amendment No. g,64
3.1 REACTOR PROTECT 10N SYSTEM BASES Additional IRM channels have also been provided to allow for bypassing of one such channel.
The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water, generator load rejection, and turbine stop valve closure are discussed in Section 2 of these specifications.
Instrumentation (pressure switches) in the drywell is provided to detect a loss of coolant accident and initiate the emergency core cooling equipment.
This instrumentation is a backup to the water level instrumentation which is discussed in Specification 3.2.
A scram is provided at the same setting i
as the emergency core cooling system (ECCS) initiation to minimize the energy which must be accomodated during a loss of coolant accident and to prevent the reactor from going critical following the accident.
The control rod drive scram system is designed so that all of.the water that is discharged from the reactor by a scram can be accomodated in the discharge piping.
A part of this piping is an instrument volume which is the low point in the piping.
No credit was taken for the volume contained in the piping below a point which is 26 inches above the lower cap to the SDIV pipe weld when calculating the amount of water which must be accomodated during a scram.
During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated which would result in slow scram times or partial or no control rod insertion.
To preclude this occurrence, level switches have been provided in the instrumented volume which alarm and scram the reactor while there is still greater than 3.34 gallons per drive available to accept scram water. As indicated above, there is sufficient volume in the piping to accomodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accomodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.
l Loss of condenser vacuum occurs when the condenser can no longer handle the heat input.
Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the i
condenser.
Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux.
To prevent the clad safety limit from being exceeded if this occurs, a reactor j
scram occurs on turbine stop valve closure.
The turbine stop valve closure j
scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient with bypass closure.
Ref.
Section 7.2 of the UFSAR.
The condenser low vacuum scram is a back-up to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe.
Scram occurs at 23" Hg vacuum; stop valve closure occurs at 20" Hg vacuum; and bypass closure at 7" Hg vacuum.
1 Millstone Unit 1 B V41-3 Amendment No. g64 0099
3.1 REACTOR PROTECT 10N SYSTEM BASES Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors which cause an isolation of the main condenser off-gas line, provided the limit for a 15-minute period specified in Specification 3.8 is not exceeded.
I The main steam line isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open in three out of four lines.
This scram anticipates the pressure and flux transient which would occur when the valves close.
By scramming at this setting all thermal margins and pressure limits are met during the resultant transient.
Ref. Section 7.2 of the UFSAR.
A reactor mode switch is provided which actuates, or bypasses, the various scram functions appropriate to the particular plant operating, status.
Ref. Section 7.2 of the UFSAR.
The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.
The IRM and APRM systems provide protection against excessive power levels and short reactor periods in the REFUEL and STARTUP/ HOT STANDBY modes.
A source range monitor (SRM) system is also provided to supply additional neutron level information during startup but has no scram functions. Thus the IRM and APRM systems are required in the REFUEL and STARTUP/ HOT STANDBY modes.
In the power range, the APRM provides the required protections; thus, the IRM system is not required in the RUN mode.
The high reactor pressure, high drywell pressure, reactor low water level, and scram discharge volume high level scrams are required for STARTUP/ HOT STANDBY and RUN modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have all scram functions except those listed in Note 8 of Table 3.1.1 operable in the REFUEL and SHUTDOWN mode is to assure that shifting to the REFUEL mode during reactor power operation does not diminish the need for the reactor protection system.
As indicated in Note 11 of Table 3.1.1, no trip functions are required to be operable if all control rods are fully inserted, and either electrically or hydraulically disarmed, since this condition assures maximum negative reactivity insertion.
The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit. At low power conditions, a turbine stop valve closure does not result in a transient which could not be handled safely by other scrams, such as the APRM.
The requirement that the IRM's be inserted in the core when the APRM's read 3/125, or lower, of full scale assures that there is proper overlap in the neutron monitoring systems and thus, that adequate coverage is provided for all ranges of reactor operation.
Millstone Unit 1 B 3/4 1-4 Amendment No. EE, EJ, fA 0099
3.2 PROTECTIVE INSTRUMENTATION BASES 200*F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks.
For large breaks, it is back-up to high steam flow instrumentation discussed above, and for small breaks with the resultant small release of radioactivity, gives isolation before the guidelines of 10 CFR 100 are exceeded.
I Pressure instrumentation is provided which trips when main steamline pressure at the turbine drops below 825 psig. A trip of this instrumentation results in closure of Group 1 isolation valves.
In the " REFUEL," " Shutdown,"
and "STARTUP/ HOT STANDBY" mode this trip function is bypassed.
This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open. With the trip set at 825 psig, inventory loss is limited so that fue] is not uncovered and peak clad temperatures are much less than 1500*F; thus, there is no release of fission products other than those in the reactor water.-
High pressure actuation of the Isolation Condenser (IC) will be a backup to direct activation on Low-Low level; similar to other ECCS systems. Activa-tion is based on the high pressure signal (1085 PSIG for 15 seconds) which occurs after HSIV closure on Low-Low water level, SRV actuation, and subse-quent repressurization. The activation of the IC requires only the opening of normally closed valve IC-3 in the condensate return line.
This valve is powered by the safety-grade DC battery.
All valves in the system are powered by safety-grade AC or DC power and are also used for containment isolation.
All are normally in the open position (other than IC-3).
The IC system is safety Class 2 and is seismically qualified. The shell side water volume is sufficient for approximately 30 minutes of operation at rated conditions without makeup.
Two sources of makeup are available.
For small break mitiga-tion, less than 10 minutes of operation is required, and generally at less than rated conditions.
1 Two sensors on the isolation condenser supply and return lines are provided to detect line failure and actuate isolation action.
The sensors on the supply and return sides are arranged in a 1 out of 2 logic and to meet the single failure criteria, all sensors and instrumentation are required to be operable.
The isolation settings and valve closure times are such as to prevent core uncovery or exceeding site limits.
The instrumentation which initiates ECCS action is arranged in a dual bus system.
As for other vital instrumentation arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.
l Millstone Unit 1 B 3/4 2-3 Amendment No. fy,64 0100