ML20199L917
| ML20199L917 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/01/1986 |
| From: | Rajender Auluck Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20199L921 | List: |
| References | |
| NUDOCS 8607100057 | |
| Download: ML20199L917 (12) | |
Text
p ter
'o UNITED STATES g
NUCLEAR REGULATORY COMMISSION 3
o h
WASHINGTON, D. C. 20555
\\...../
NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 45 License No. DPR-22 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (tr.e licensee)datedSeptember 24, 1982, complies with the standards t.nd requirements of the Atomic Energy Act of 1954, as amended (the /.ct),
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 Commission; C.
There is reasonable assurance (1) 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 comon 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.
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-22 is hereby amended to read as follows:
8607100057 860701 PDR ADOCK 05000263 P
_2_
2 Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 45, 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 THE NUCLEAR REGULATORY COMMISSION k
W i
Rajender Auluck, Project Manager BWR Project Directorate #1 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: July 1, 1986
ATTACHMENT TO LICENSE AMENDMENT N0. 45 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET N0. 50-263 Revise 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 INSERT i
i vi vi 60 60 62 62 69 69 71 71 129 129 172 172 238 238 I
l i
TABLE OF CONTENTS Page 1.0 DEFINITIONS 1
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM CETTINGS 6
2.1 and 2.3 Fuel Cladding Integrity 6
2.1 Bases 10 2.3 Bases 14 2.2 and 2.4 Reactor Coolant System 21 2.2 Bases 22 2.4 Bases 24 3.0 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS 25a 4.0 Surveillance Requirements 25a 3.1 and 4.1 Reactor Protection System 26 7
3.1 Bases 35 4.1 Bases 41 3.2 and 4.2 Protective Instrumentation 45 A.
Primary Containment Isolation Functions 45 B.
Emergency Core Cooling Subsystems Actuation 46 C.
Control Rod Block Actuation 46 D.
Other Instrumentation 46a E.
Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation 47 F.
Recirculation Pump Trip Initiation and Alternate Rod Injection Initiation 48 G.
Safeguards Bus Voltage Protection 48 H.
Instrumentation for S/RV Low-Low Set Logic 48 y
3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 Control Rod System 76 A.
Reactivity Limitations 76 B.
Control Rod Withdrawal 77-C.
Scram Insertion Times 81 D.
Control Rod Accumulators 82 83 E.
Reactivity Anomalies F.
Scram Discharge Volume 83A 83A G.
Required Action 3.3 and 4.3 Bases 84 1
Amendment No. 30, 37, 45
-~
LIST OF TABLES Tcblo No.
Psga 3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional 32 Test Frequencies for Safety Instrumentation and Control Circuits 4.1.2 Scram Instrument Calibration - Minimum Calibration 34 Frequencies for Reactor Protection Instrument Channels 3.2.1 Instrumentation that Initiates Primary Containment 49 Isolation Functions 3.2.2 Instrumentation that Initiates Emergency Core Cooling Systems 52 3.2.3 Instrumentation that Initiates Rod Block 57 3.2.4 Instrumentation that Initiates Reactor Building Ventilation 59 Isolation and Standby Gas Treatment System Initiation 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip 60 and Alternate Rod Injection l
3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and 60a Loss of Voltage Protection 7
3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 3.2.8 Other Instrumentation 60d 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation 61 3.7.1 Primary Containment Isolation 172 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 1891 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198m Surveillance Requirements 4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198n Surveillance Requirements 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p 4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s vi Amendment No. 37, 39, 44,45 l
Table 3.2.5 Instrumentation that Initiates a Recirculation Pump Trip and Alternate Rod Injection l
Minimum No. of Oper-Minimum No. of able or Operating 1
Operable or Total No. of Instru-Instrument Channels Required Operating Trip ment Channels Per Per Trip System Conditions
- Function Trip Setting Systems (1)
Trip System (1)
- 1. High Reactor Dome Pressure
<1150 psig 2
2 2
A
- 2. Low-Low Reactor Water
>6' 6" above the Level top of the active fuel.
2 2
2 A
i NOTE:
I 1.
Upon discovery that minimum requirements for the number of operable or operating trip systems or instrument channels are not satisfied, action shall be initiated to:
i l
a.
Satisfy the requirements by placing the appropriate channels or systems in the tripped condition, or b.
Place the plant under the specified required condition using normal operating procedures.
- Required conditions when minimum conditions for operation are not satisfied:
A.
Reactor in startup, Refuel or Shutdown mode.
4
.l e
3.2/4.2 60 Amendment No. 45
.9 Table 4.2.1 - Continued Minimum Test and Calibration Frequency For Core Cooling Rod Block and Isolation Instrumentation Instrument Channel Test (3)
Calibration (3)
Sensor Check (3) 3.
Steam Line Low Pressure Note 1 Once/3 moftths None 4.
Steam Line High Radiation once/ week (5)
Note 6 Once/ shift HPCI ISOLATION 1.
Steam Line High Flow Once/ month once/3 months None 2.
Steam Line High Temperature Once/ month once/3 months None RCIC ISOLATION 1.
Steam Line High Flow once/ month once/3 months None 2.
Steam Line High Temperature Note 1 Once/3 months None REACTOR BUILDING VENTILATION 1.
Radiation Monitors (Plenum)'
.Once/ month once/3 months Once/ day 2.
Radiation Monitors (Refueling Floor)
Note 1 Once/3 months (4)
See Table 4.8.2 3.
Wide Range Gas Monitors RECIRCULATION PUMP TRIP AND ALTERNATE ROD INJECTION 1.
Reactor High Pressure Note 1 Once/ Operating Cycle-Once/ Day
,8 Transmitter Once/3 Months-Trip Unit
'E 2.
Reactor Low Low Water Level (Note 7)
Once/ month Once/ Operating Cycle-Once/ shift Transmitter f
I Once/3 Months-Trip Unit h
SHUTDOWN COOLING SUPPLY ISOLATION f
1.
Reactor Pressure Interlock Note 1 Once/3 Months None 3.2/4.2 62
t I
Bases Continued:
3.2 increases core voiding, a negative reactivity feedback. High pressure sensors initiate the pump trip in the event of an isolation transient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative.
The ATWS high reactor pressure and low-low water level logic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve to vent the header and causes rod insertion.
This greatly reduces the long term consequences of an ATWS event.
Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate source when a loss of voltage condition or a degraded voltage condition is sensed. On loss of i
voltage this transfer occurs immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage for starting and running loads during a loss of coolant accident. An allowance for relay tolerance is included.
Safety / relief valve low-low set logic is provided to prevent any safety / relief valve from opening when there isan elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when j
steam condenses in the line. If the valve reopens before the discharge line vacuum breakers act to return water level to normal, water clearing thrust loads on the l
discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve actuations is provided assuming one valve fails to i
'a Et
's r
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$5 3.2 BASES 69
l l
Trip Function Deviation Instrumentation That Initiates Emergency Low-Low Reactor Water Level
-3 Inches Core Cooling Systems Table 3.2.2 Reactor Low Pressure (Pump
-10 psi Start) Permissive High Drywell Pressure
+1 psi Low Reactor Pressure (Valve
-10 psi Permissive) i l
i Instrumentation That Initiates IRM Downscale
-2/125 of Scale Rod Block IRM Upscale
+2/123 of Scale Table 3.2.3 APRM Downscale
-2/125 of Scale APRM Upscale See Basis 3.2 l
RBM Downscale
-2/125 of Scale RBM Upscale
+2/125 of Scale Scram Discharge Volume-High
+ 1 gallon Level i
Instrumentation That Initiates High Reactor Pressure
+ 12 psi I
Recirculation Pump Trip and Low-Low Reactor Water Level
-3 Inches Alternate Rod Injection
]f Instrumentation for Safeguards Degraded Voltage 23897 volts (trip)
Bus Protection 53975 volts (reset) 25 sec 510 see (delay)
{
5 Loss of Voltage
<3000 volts >2000 volts 5
)
0; 3.2 BASES 71
)
i
,I, A
A s'*
m 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRENENTS f
H.
Snubbers H.
Snubbers 4
1.
Except as permitted below, all safety the following survalliance requirements apply to related snubbers shall be operable all safety related snubbers.
whenever the supported system is required to be operable.
1.
Visual inspection of snubbers shall be conducted in accordance with the following 2.
With one or more snubbers unde or found to schedule be inoperable for any reason when Operability is required, within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s:
No. of Snubbers Found Next Required Inoperable per Inspection Period a.
Replace or restore the inoperable snubbers inspection Period to Operable status and perform an engineer-0 18 months i 251 3
ing evaluation or inspection of the 1
12 months i 25%
l supported components, or 2
6 months i 25X j
3.4 124 days i 251 b.
Determine through engineering evaluation 5,6,7 62 days 1 25%
that the as-found condition of the snubber 8 or more 31 days i 25%
had no adverse effect on the supported components and that they would The required inspection interval shall not be J
retain their structural integrity in lengthened more than one step at a time.
i the event of design basis seismic event, or Snubbers may be categorized in two groups, j
" accessible" or " inaccessible" based on their jf c.
Declare the supported system inoperable accessibility for inspection during reactbr g
and take the action required by the operation. These two groups may be inspected St Technical Specifications for inoper-independently according to the above schedule.
ability of that system.
E.I
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U I=
129 3.6/4.6
.pr TABLE 3.7.1 PRIMARY CONTAINMENT ISOLATION Number of Maximum Isolation Valve Valves Operating Normal Group Identification Inboard Outboard Time (Sec)
Position 1
Main Steam Line Isolation 4
4 5*
Open 1
Main Steam Line Drain 1
1 60 Closed 1
Recirculation Loop Sample Line 1
1 60 Open 2
Drywell Floor Drain 2
60 Open 2
Drywell Equipment Drain 2
60 Open 2
Drywell Vent 2
60 Closed 2
Drywell Vent Bypass 1
60 Closed 2
Drywell Purge Inlet 2
60 Open 2
Drywell and Suppression Chamber 1
60 Closed Air Makeup 2
Suppression Chamber to Drywell 1
60 Open N Recirculation 2
2 Suppression Chamber Vent 2
60 Closed N
2 Suppression Chamber Vent Bypass 1
60 Open
~
1 120 Closed 2
Shutdown Cooling System 1
c+
5
- Minimum closure time shall be >3 seconds S3
'a 3.7/4.7 172
.. ~
b.
When the nature of a particular problem dictates, special consultants will be utilized, as necessary, to provide expert advice to the SAC.
3.
Meeting Frequency The SAC shall meet on call by the Chairman but not less frequently than twice a year.
4.
Quorum No less than a majority of the permanent members or their alternates, including the a.
l SAC Chairman or Vice Chairman.
b.
No more than a minority of the quorum shall be from groups holding line responsibility for the operation of the plant.
5.
Responsibilities - The following subjects should be reported to or reviewed by the SAC:
Written safety evaluations of (1) changes in the facility, (2) changes to procedures, l
a.
and (3) tests or experiments completed without prior Commission approval under I
the provisions of 10 CFR 50.59 to verify that such changes, tests or experiments did not involve a change in the Appendix A Technical Specifications or an unreviewed safety question as defined in 10 CFR 50.59.
b.
Proposed changes to procedures, changes in the facility, and tests and experiments which may involve a change in the Appendix A Technical Specifications or an unreviewed safety question as defined in 10 CFR 50.59.
Matters of this kind shall be referred to the SAC following their review by the onsite operating organization.
Proposed changes in Appendix A Technical Specifications or proposed license amendments c.
relating to nuclear safety.
Y Violations of applicable codes, regulations, orders, Appendix A Technical Specifications, d.
m El and license requirements or internal procedures or instructions having nuclear safety I
5 significance.
5 Significant operating abnormalities or deviations from normal and expected performance m
e.
P of plant safety-related structures, systems, or components.
6.2 238