ML20070B336
| ML20070B336 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/21/1994 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070B338 | List: |
| References | |
| NUDOCS 9406300064 | |
| Download: ML20070B336 (16) | |
Text
.=.
Y"%
i S
UNITED STATES 5
i!
NUCLEAR REGULATORY COMMISSION y...../
t WASHINGTON, D.C. 20555 4 01 i
BOSTON EDISON COMPANY DOCKET NO. 50-293 4
PILGRIM NUCLEAR POWER STATION I
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.154 License No. DPR-35 l.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
s A.
The application for amendment filed by the Boston Edison Company (the licensee) dated October 19, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
l and the Commission's rules and regulations; B.
The facility will operate in conformity with the application, the-i provisions of the Act, and the rules and regulations of the Commission; I
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 4
4 E.
The issuance of this amendment is in accordance with 10 CFR Part 51 1
of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-l tions as indicated in the attachment to this license amendment.
1 J
4 J
4 P
. 3.
This license amendment will be implemented coincidental with the corresponding plant design change or within 90 days following receipt of the Amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Walter. R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II l
Office of Nuclear Reactor Regulation
Attachment:
[
Changes to the Technical Specifications l
Date of Issuance: June 21, 1994 l
ATTACHMENT TO LICENSE AMENDMENT N0.154 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET N0. 50-293 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 27 27 29 29 30 30 32 32 33 33 39 39 45 45 46 46 46a 46a 67 67 69 69 70 70 193f 193f
PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Operable Inst.
Modes in Which Function Channels per Trip Function Trip Level Setting Must Be Operable Action (1)
Trio System (1)
Refuel (7) Startup/ Hot Run MinimumlAvail.
Standbv 1
1 Mode Switch in Shutdown X
X X
A 1
1 Manual Scram X
X X
A IRM 3
4 High Flux
$120/125 of full scale X
X (5)
A 3
4 Inoperative X
X (5)
A APRM 2
3 High Flux (15)
(17)
(17)
X A or B 2
3 Inoperative (13)
X X(9)
X A or B 2
3 High Flux (15%)
515% of Design Power X
X (16)
A or B 2
2 High Reactor Pressure
$1063.5 psig X(10)
X X
A 2
2 High Drywell Pressure
$2.22 psig X(8)
X(8)
X A
2 2
Reactor low Water Level 211.7 In. Indicated X
X X
A Level SDIV High Water Level:
538 Gallons X(2)
X X
A 2
2 East 2
2 West 1
4 4
Main Steam Line Isolation Valve Closure
$10% Valve Closure X(3)(6)
X(3)(6)
X(6)
A or C 2
2 Turbine Control Valve 2150 psig Control Oil Fast Closure Pressure at Acceleration Relay X(4)
X(4)
X(4)
A or D 4
4 Turbine Stop Valve
$10% Valve Closure X(4)
X(4)
X(4)
A or D Closure 15 -42,-86,-921-117, 133, 147, 151, 152, 154 27 Amendment No.
1
i.
l NOTES FOR TABLE 3.1.1 (Cont'd) 2.
Permissible to bypass, with control rod block, for reactor protection system reset in refuel and shutdown positions of the reactor mode switch.
3.
Permissible to bypass when reactor pressure is $576 psig.
4.
Permissible to bypass when turbine first stage pressure is $112 psig.
5.
IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position.
6.
The design permits closure of any two lines without a scram being initiated.
7.
When the reactor is subcritical, fuel is in the reactor vessel 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.
High flux IRM D.
Scram discharge volume high level E.
Not required to be operable when primary containment integrity is not required.
9.
Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
- 10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 11. Deleted
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of LPRM's to an APRM.
- 14. Deleted
- 15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal power.
startup/ hot standby modes.
- 18. Deleted.
Amendment No. 6, 15, 27, 42, 86, 117, 118, 133, 147, 151, 152,154 29
TABLE 4.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION AND CONTROL CIRCUITS Functional Test Minimum Freauency (3_1 Mode Switch in Shutdown Place Mode Switch in Shutdown Each Refueling Outage Manual Scram Trip Channel and Alarm Every 3 Months RPS Channel Test Switch (5)
Trip Channel and Alarm Once per week IRM High Flux Trip Channel and Alarm (4)
Once Per Week During Refueling and Before Each Startup Inoperative Trip Channel and Alarm Once Per Week During Refueling and Before Each Startup APRM High Flux Trip Output Relays (4)
Every 3 Months (7)
Inoperative Trip Output Relays (4)
Every 3 Months Flow Bias Trip Output Relays (4)
Every 3 Months High Flux (15%)
Trip Output Relays (4)
Once Per Week During Refueling and Before Each Startup High Reactor Pressure Trip Channel and Alarm (4)
Every 3 Months High Drywell Pressure Trip Channel and Alarm (4)
Every 3 Months Reactor low Water Level Trip Channel and Alarm (4)
Every 3 Months High Water Level in Scram Discharge Tanks Trip Channel and Alarm (4)
Every 3 Months Main Steam Line Isolation Valve Closure Trip Channel and Alarm Every 3 Months Turbine Control Valve Fast Closure Trip Channel and Alarm Every 3 Months Turbine First Stage Pressure Permissive Trip Channel and Alarm (4)
Every 3 Months Turbine Stop Valve Closure Trip Channel and Alarm Every 3 Months Reactor Pressure Permissive Trip Channel and Alarm (4)
Every 3 Months i
79 -99, 117, 147, 152, 154 30 Amendment No.
1
~
TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Calibration Test (5)
Minimum Freauency (2)
IRM High Flux Comparison to APRM on Controlled Note (4)
Shutdowns Full Calibration Once per Operating APRM High Flux Output Signal Heat Balance Once every 3 Days Flow Bias Signal Calibrate Flow Comparator and At least once Every Flow Bias Network 18 Months Calibrate Flow Bias Signal (1)
Every 3 Months LPRM Signal TIP System Traverse Every 1000 Effective Full Power Hours High Reactor Pressure Note (7)
Note (7)
High Drywell Pressure Note (7)
Note (7)
Reactor Low Water Level Note (7)
Note (7)
High Water Level in Scram Discharge Tanks Note (7)
Note (7)
Main Steam Line Isolation Valve Closure Note (6)
Note (6)
Turbine First Stage Pressure Permissive Note (7)
Note (7)
Turbine Control Valve Fast Closure Standard Pressure Source Every 3 Months Turbine Stop Valve Closure Note (6)
Note (6)
Reactor Pressure Permissive Note (7)
Note (7)
Amendment No. 147, 151, 152, 154 32
f7ES FOR TABLE 4.1.2 I
Adjust the flow bias trip reference, as necessary, to conform to a calibrated 1
flow signal.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
3.
Deleted l
4.
Maximum frequency required is once per week.
5.
Response time is not a part of the routine instrument channel test, but will be checked once per operating cycle.
6.
Physical inspection and actuation of these position switches will be performed during the refueling outages.
7.
Calibration of these devices will be performed during refueling outages.
To verify transmitter output, a daily instrument check will be performed.
Calibration of the associated analog trip units will be performed concurrent with functional testing as specified in Table 4.1.1.
I Revision No. 166 Amendment No. 79, 99, 147,154 33 I
3.1 BASES (Cont'd) range of applicability of the fuel cladding integrity safety limit.
transients that occur during normal or inadvertent isolation v In With the scrams set at 10 percent of valve closure, neutron flux does not increase.
Hiah Reactor Prsssure The high reactor pressure scram setting is chosen slightly above the maximum yet provide a wide margin to the ASME Section III allowable r system pressure (1250 psig, see Bases Section 3.6.D).
Bioh Drywell Pressure Instrumentation for the drywell is provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the Core Standby Cooling Systems during a(loss of coolant accident and to prevent return to crit CSCS) initiation to minimize the energy that must be accommodated instrumentation is a backup to the reactor vessel water level instrumentation Reactor Mode Switch The reactor mode switch actuates or bypasses the various scram functions 7.2.3.9). appropriate to the particular plant operating status (Reference FSAR Section Manual Scram 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 Amendment No. 6, 79, 133, 147, 154 39
PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Operable Instrument Channels Per Trip System (1)
Minimum Available Instrument Trio Level Setting Action (2) 2(7) 2 Reactor Low Water Level 211.7" indicated 1cvel (3)
A and D 1
1 Reactor High Pressure
$110 psig D
2 2
Reactor Low-Low Water Level at or above -46.3 in.
A indicated level (4) 2 2
Reactor High Water Level
$45.3" indicated level (5)
B 2(7) 2 High Drywell Pressure
$2.22 psig A
l 2
2 Low Pressure Main Steam Line 2810 psig (8)
B 2(6) 2 High Flow Main Steam Line
$136% of rated steam flow B
2 2
Main Steam Line Tunnel Exhaust Duct High Temperature
$170 F B
2 2
Turbine Basement Exhaust Duct High Temperature
$150 F B
1 1
Reactor Cleanup System High Flow
$300% of rated flow C
2 2
Reactor Cleanup System High Temperature
$150 F C
45 Amendment No. 86. 147, 150, 151,154
NOTES FOR TABLE 3.2.A 1.
Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
An instrument channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter; or, where only one channel exists per trip system, the other trip system shall be operable.
2.
Action 4
If the minimum number of operable instrument channels cannot be met for one of the trip systems of a trip function, the appropriate conditions listed below shall be followed:
If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable channel (s) and/or that trip system shall be placed in the tripped condition within one hour (twelve hours for Reactor Low Water Level and High Drywell l
Pressure), or initiate the action required by Table 3.2.A for the affected trip functions.
If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channel (s) shall be restored to operable status within two hours (six hours for Reactor Low Water l
Level and High Drywell Pressure) or initiate the Action required by Table 3.2. A for the affected trip function.
If the minimum number of operable instrument channels cannot be met for both trip systems, place at least one trip system (with the most inoperable channels) in the tripped condition within one hour or initiate the appropriate Action required by Table 3.2.A. listed below for the affected trip function.
A.
Initiate an orderly shutdown and have the 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 Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
holate Shutdown Cooling.
Amendment No. 86, 105, 119, 147, 154 46
3.
Instrument set point corresponds to 128.26 inches above top of active fuel.
4.
Instrument set point corresponds to 77.26 inches above top of active fuel.
l 5.
Not required in Run Mode (bypassed by Mode Switch).
\\
6.
Two required for each steam line.
7.
These signals also start SBGTS and initiate secondary containment isolation.
1 8.
Only required in Run Mode (interlocked with Mode Switch).
9.
Deleted.
l i
I i
l l
l Amendment No. 147, 151,154 46a i
NOTES FOR TABLES 4.2.A THROUGH 4.2.G 1.
Initially once per month until exposure hours (M as defined on Figure 5
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.
Functional tests, calibrations and instrument checks are.not required when these instruments are not required to be operable or are tripped.
Functional tests shall be performed before each startup with a required frequency not to exceed once per week.
Calibrations of IRMs and SRMs shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week.
Instrument checks shall be performed at least once per day during those periods when the instrument:; are required to be operable.
3.
This instrumentation is excepted from the functional test definition.
The functional test will consist of injecting a simulated electrical i
4 signal into the measurement channel.
4.
Simulated automatic actuation shall be performed once each operating cycle.
Where possible, all logic system functional tests will be performed using the test jacks.
5.
Reactor low water level and high drywell pressure are not included on Tabie 4.2.A since they are tested on Tables 4.1.1 and 4.1.2.
6.
The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.
7.
Calibration of analog trip units will be performed concurrent with functional testing. The functional test will consist of injecting a simulated electrical signal into the measurement channel.
Calibration of associated analog transmitters will be performed each refueling outage.
Amendment No. 147, 154 67 1
e 3.2 BASES (Cont'd)
The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 isolation valves.
For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also.
The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of Group 1 isolation valves.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a break in the main steam line.
For the worst case accident, main steam line break outside the drywell, the steam flow trip setting in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures remain approximately 1000 F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.
Temperature monitoring instrumentation is provided in the main steam line tunnel and the turbine basement to detect leaks in these areas.
Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.
The setting of 170 F for the main steam line tunnel detector 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, the high steam flow instrumentation is a backup to the temperature instrumentation.
Amendment No. 34, 113, 151, 154 69
4 b
3.2 BASES (Cont'd)
)
i l
4 l
i l
i Pressure instrumentation is provided to close the main steam isolation valves in RyN mode before the reactor pressure drops below 785 psig.
This function is primarily intended to prevent excessive vessel depressurization in.the event of a malfunction of the nuclear system pressure regulator.
This function also provides automatic protection of the low-pressure core-thermal-power safety limit (25% of rated core thermal power for reactor pressure < 785 psig).
In the Refuel or Startup Mode, the inventory loss associated with such a malfunction would be limited by closure of the Main Steam Isolation Valves due to either high or low reactor water level; no fuel would be uncovered.
This function is not required to satisfy any safety design bases.
The.HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping.
Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic, and all sensors are required to be operable.
Temperature is monitored at three (3) locations with four (4) temperature sensors at each location.
Two (2) sensors at each location are powered by "A" direct current control bus and two (2) by "B" direct current control bus.
Each pair of sensors, e.g., "A" or "B", at each i
location are physically separated and the tripping of either "A" or "B" bus sensor will actuate HPCI isolation valves.
The trip settings of f 300% of design flow for high flow and 200 F or 170 F, depending on sensor location, for high temperature are such that core uncovery is prevented and fission product release is within limits.
The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI.
The trip setting of 1300% for high flow and 200 F,170*F and 150 F, depending on sensor location, for. temperature are based on the same criteria as the HPCI.
The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar as that for the HPCI.
The trip settings are such that core uncovery is prevented and fission product release is within limits.
The instrumentation which initiates CSCS 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.
An exception to this is when logic functional testing is being performed.
Amendment No. 140, 150, 15A 70
BASES 3/4.8.G Main Condenser (Continued)
Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line.
Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip.
There is a fifteen minute delay before the air ejector off-gas isolation valve is closed. This delay is accounted for by the 30-minute holdup time of the off-gas before it is released to the stack.
Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip.
The trip settings of the instruments are set so that the instantaneous stack release rate limit given in Specification 3.8 is not exceeded.
H.
Mechanical Vacuum Pumo The purpose of isolating the mechanical vacuum pump line is to limit the release of activity from the main condenser.
During a Control Rod Drop Accident, fission products would be transported from the reactor through the main steam lines to the condenser.
The fission product radio-activity would be sensed by the main steam line radioactivity monitors, initiating isolation of the mechanical vacuum pump.
Amendment No. E8F,154 193f