ML20086J810
| ML20086J810 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/12/1995 |
| From: | Hannon J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20086J812 | List: |
| References | |
| NUDOCS 9507190251 | |
| Download: ML20086J810 (15) | |
Text
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UNITED STATES j
j NUCLEAR REGULATORY COMMISSION t
WASHINoTON, D.C. 3066lM1001 49 * * * * *,o NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 93 License No. DPR-22
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated February 12, 1993, as supplemented March 22, 1993, and August 25, 1994, 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-22 is hereby amended to read as follows:
9507190251 950712 PDR ADOCK 05000263 p
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 93, 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.
FOR THE NUCLEAR REGULATORY COMMISSION e
ohn N. Hannon, Director Project Directorate III-l Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 12, 1995 I
I'
ATTACHMENT TO LICENSE AMENDMENT NO. 93 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
REMOVE INSERT 52 52 53 53 54 54 55 55 60d 60d 101 101 107 107 110 110 113 113 127 127 151 151 156 156 t
Table 3.2.2 Iustrumentation That Initiates Emergency Core Cooling Systems Minimum No. of Oper-Minimum No. of able or Operating Operable or Total No. of Instru-Instrument Channels Operating Trip ment Channels Per Per Trip System Required Function Trio Settinz Systems (3)
Trio System (3)
Conditions
- A.
Core Sorav and LPCI 1.
Pump Start a.
Iow low Reactor 26'6"s6'10" 2
4(4) 4 A.
Water Level and b.
1.
Reactor Low 2450 psig 2
2(4) 2 A.
Pressure Permissive or
- 11. Reactor Low 2011 min 2
1 1
B.
Pressure Permissive Bypass Timer c.
High Drywell 52 psig 2
4(4) 4 A.
Pressure (1) l 2.
Low Reactor Pressure 2450 psig 2
2(4) 2 A.
(Valve Permissive) 2 2(2) 2 A.
3.
Loss of Auxiliary Power 3.2/4.2 52 Amendment No. $2,93
i s
?
Table 3.2.2 Instrumentation That Initiates Emergency Core Cooling Systems Minlaum No. of Oper-t Minimum No. of able or Operating Operable or Total No. of Instru-Instrument Channels Operating Trip ment Channels Per Per Trip System Required Function Trio Settine Systems (3)
Trio System (3) Conditions
- f B.
HPCI System 1.
High Drywell 52 psig 1
4 4
A.
Pressure (1)
I 2.
Low-Low Reactor 26'6"56'10" 1
4 4
A.
I Water Level i
C.
Automatic Deores-surization 1.
Low-Low Reactor 26'6"s6'10" 2
2 2
B.
Water Level and 2.
Auto Blowdown s120 seconds 2
1 1
.B.
Timer and 3.
Low Pressure Core 5100 psig 2
12(4) 12(4)
B.
j Cooling Pumps Dis-
{
Charge Pressure i
Interlock l
I 1
i 3.2/4.2 53 haendinent No. $2,93
.~.
Table 3.2.2 - Continued Instrumentation That Initiates Emerzency Core Cooling System Min. No. of Oper-Min. No.
able or Operating of Operable Total No. of Instru-Instrument Channels or Operating ment Channels Per Per Trip System Required Function Trio Settine Trio Systems (3)
Trio System (3)
Conditions
- D.
Diesel Generator 1.
Degraded or Loss of Voltage Essential Bus (5) 2.
Low low Reactor 26'6"s6'10" 2
4(4) 4 C.
Water Level 3.
High Drywell Press 52 psig 2
4(4) 4 C.
NOTES:_
High drywell pressure may be bypassed when necessary only by closing the manual containment isolation valves during 1.
purging for containment inerting or de-inerting. Verification of the bypass condition shall be noted in the control room log. Also need not be operable when primary containment integrity is not required.
2.
One instrument channel is a circuit breaker contact and the other is an undervoltage relay.
54 3.2/4.2 l
Amendment No. 3,93
Table 3.2.2 - Continued Notes:
3.
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:
(a) Satisfy the requirements by' placing appropriate channels or systems in the tripped condition, or (b) Place the plant under the specified required conditions using normal operating procedures.
4.
All instrument channels are shared by both trip systems.
5.
See table 3.2.6.
Required conditions when minimum conditions for operation are not satisfied.
A.
Comply with Specification 3.5.A.
3.
Reactor pressure 5150 psig.
C.
Comply with Specification 3.9.B.
55 3.2/4.2 Ateendment No. 3,93
Table 3.2.8 Other Instrumentation Minimum No. of Minimum No. of Oper-Operable or Total No. of Instru-able or Operating Required Function Trip Setting Operating Trip ment Channels Per Instrument Channels Conditions
- System (1)
Trio System Per Trio System (1)
A.
RCIC Initiation
- 1. Low-Low Reactor Level 26'6"& 56'10" 1
2 2
B above top of active fuel B.
HPCI/RCIC Turbine Shutdown
- a. High Reactor Level 514'6" above 1
2 2
A top of active fuel C.
HPCI/RCIC Turbine Suction Transfer
- a. Condensate Storage 22'0" above 1
2 2
C Tank Low Level tank bottom NOTE: Upon discovery that minimum requirements for the number of operable or operating trip systems or instrument channels 1.
are not satisfied, action shall be initiated to:
a.
Satisfy the requirements by placing the appropriate channels or systems in the tripped condition (Turbine /Feedwater Trip only), or b.
Dlace 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.
B.
Comply with Specification 3.5.D.
C.
Align HPCI and RCIC suction to the suppression pool. Restore channels to operable status within 30 days or place the plant in Required Condition A.
60d 3.2/4.2 Amendment No. 37, 93
3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS 4.5 CORE AND CONTAINMENT SPRAY / COOLING SYSTEMS Apolicability:
Aeolicabilitv:
Applies to the operational status of the emergency Applies to the periodic testing of the emergency cooling systems.
cooling systems.
Obiective:
Obiective:
To insure adequate cooling capability for i. cat removal To verify the operability of the emergency cooling in the event of a loss of coolant accident or systems.
isolation from the normal reactor heat sink.
Soecification:
Specification:
A. ECCS Systems A. ECCS Systems 1.
Except as specified in section 3.5.A.3, both 1.
Demonstrate the Core Spray Pumps develop a Core Spray subsystems and the Low Pressure 2,800 gpm flow rate against a-system head l
Coolant Injection (LPCI) Subsystem (LPCI Mode corresponding to a reactor pressure of 130 of RHR System) shall be operable whenever psi greater than containment pressure, when irradiated fuel is in the reactor vessel and tested pursuant to Specification 4.15.B.
the reactor water temperature is greater than 212*F.
2.
Demonstrate the LPCI Pumps develop a 3,870 gpm flow rate against a system head 2.
Except as specified in section 3.5.A.3, the corresponding to two pumps delivering 7,740 High Pressure Coolant Injection (HPCI) System gpm at a reactor pressure of 20 psi greater and the Automatic Depressurization System than containment pressure, when tested (ADS) shall be operable whenever the reactor pursuant to Specification 4.15.B.
pressure is greater than 150 psig and irradiated fuel is in the reactor vessel 3.
Demonstrate the HPC1 Pump develops a 2700 except during reactor vessel hydrostatic or gpm flow rate against a reactor pressure leakage tests.
range of 1120 psig to 150 psig, when tested pursuant to Specification 4.15.B.
101 3.5/4.5 Amendment No.'77, /$,93
3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMEh'rS F.
Recirculation System F.
Recirculation System 1.
The reactor may be started and operated, or
- 1. See Specification 4.6.C operation may continue with only one recirculation loop in operation provided that:
- 2. The following baseline noise levels will be obtained prior to operation with only one The following changes to setpoints and recirculation pump in operation at a core a.
safety limit settings will be made within thermal power greater than that specified 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiating operation with in Figure 3.5.1 or with a core flow greater only one recirculation loop in operation.
than 45% provided that baseline values have not been established since the last core
- 1. The Operating Limit MCPR (MCPR) will be refueling. Baseline values will be taken changed per Specification 3.11.C.
with only one recirculation pump running.
- 2. The Maximum Average Planar Linear Heat a.
Establish a baseline core plate AP noise Generation Rate (MAPLHGR) will be level.
changed as noted in Table 1 of the Core Operating Limits Report.
b.
Establish a baseline APRM and LPRM neutron flux noise level.
- 3. With only one recirculation loop in in Specification 2.3.A and Table 3.2.3.
operation at a core thermal power greater than that specified in Figure 3.5.1 or with b.
Total core flow will be maintained greater a core flow greater than 45%, determine the than 39% when core thermal po'wer is above following noise levels at least once per 8 the limit specified in Figure 3.5.1.
hour period and within 30 minutes after a core thermel power increase of greater than 5% of rated thermal power.
Core plate AP noise levels.
a.
b.
APRM and LPRM neutron noise levels.
107 3.5/4.5 Amendment No. 27, 77, /$,93
Bases 3.5/4.5 A.
ECCS Systems The core spray system is provided to assure that the core is adequately cooled following a loss-of-coolant accident and, together with the LPCI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS).
The Core Spray System is a primary source of emergency core cooling after the reactor vessel is depressurized dnd a source for flooding of the core in case of accidental draining. The Core Spray pump is designed to deliver greater than or equal to 3020 gpm (the SAFER /CESTR-lDCA safety analysis assumed a Core Spray Pump flow of 2,800 gpm, or 2,700 gpm flow into the core + 100 gpa to account for ECCS bypass leakage) against a system head corresponding to a reactor pressure of 130 psi greater than containment pressure.
The surveillance requirements provide adequate assurance that the Core Spray System will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The l
pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Four pumps are available to provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
LPCI Loop Selection Logic determines which Recirculation loop the four RHR pumps will pump into.
Each RHR pump was designed to deliver greater than or equal to 4000 gpm (the safety analysis assumed two pumps delivering 7,740 gpm) against a system head corresponding to a reactor pressure 6f 20 psi greater than containment pressure.
The allowed out-of-service conditions (Section 3.5.A.3) are determined from ECCS analysis cases analyzed.
l Only one of these conditions is permitted to exist.
If more than one condition exists, an orderly shutdown shall be initiated. A LPCI injection path consists of the two motor operated injection valves on that path.
The surveillance requirements provide adequate assurance that the LPCI system will be operable when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The 3.5/4.5 110 Amendment No. $3, 77, 7),93
Bases 3.5/4.5 Continued:
The RHR service water system provides cooling for the RHR heat exchangers and can thus maintain the suppression pool water within limits. With the flow specified, the pool temperature limits are maintained as specified in Specification 3.7.A.l.
D.
RCIC The RCIC system is provided to supply continuous makeup water to the reactor core when the reactor is isolated from the turbine and when the feedwater system is not available. The pumping capacity of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts.
The system may also be manually initiated at any time.
The HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailable. Therefore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.
The surveillance requirements provide adequate assurance that the RCIC system will be operable when required. All active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
E.
Cold Shutdown and Refueling Requirements The purpose of Specification 3.5.E is to assure that sufficient core cooling equipment is available at all times.
It is during refueling outages that major maintenance is performed and during such time that all core and containment spray / cooling subsystems may be out of service. This specification allows all core and containment spray / cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel. Thus events requiring core cooling are precluded.
Specification 3.5.E.2 recognizes that concurrent with control rod drive maintenance during the l
1 refueling outage, it may by necessary to drain the suppression chamber for maintenance or for the j
inspection required by Specification 4.7.A.l.
In this situation, a sufficient inventory of water is I
maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.
113 3.5/4.5 Bases Amendment No. $7,77,79, 93 L
3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS E.
Safety / Relief Valves E.
Safety / Relief Valves 1.
During power operating conditions and whenever 1.
a.
Safety / relief valves shall be tested reactor coolant pressure is greater than 110 or replaced each refueling outage psig and temperature is greater than 345'F the pursuant to Specification 4.15.B.
safety valve function (self actuation) of The nominal self-actuation setpoints seven safety / relief valves shall be operable are specified in Section 2.4.B.
(note: Low-Low Set and ADS requirements are located in Specifications 3.2.H and 3.5.A, respectively).
b.
At least two of the safety / relief valves shall be disassembled and 2.
If Specification 3.6.E.1 is not met, initiate inspected each refueling outage.
an orderly shutdown and have reactor coolant c.
The integrity of the safety / relief pressure and temperature reduced to 110 psig or less and 345*F or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
valve bellows shall be continuously monitoreo.
d.
The operability of the bellows monitoring system shall be demonstrated at least nnce every three months.
2.
Low-Low Set Logic surveillance shall be performed in accordance with Table 4.2.1.
1 127 3.6/4.6 Amendment No. 30,EZ,7%,92, 93
Bases Continued 3.6 and 4.6:
The safety / relief valves have two functions; 1) over-pressure relief (self-actuated by high pressure), and 2) Depressurization/ Pressure Control (using air actuators to open the valves via ADS, low-Low Set system, or manual operation). The Low-Low Set and ADS functions are discussed further in Sections 3.2 and 3.5.
The safety function is performed by the same safety / relief valve with self-actuated integral bellows and pilot valve causing main valve operation. Article 9 of the ASME Pressure Vessel Code Section III Nuclear Vessels requires that these bellows be monitored for failure since this would defeat the safety function of the safety / relief valve.
{
Provision also has been made to detect failure of the bellows monitoring system. Testing of this system quarterly provisions assurance of bellows integrity.
When the setpoint is being bench checked, it is prudent to disassemble one of the safety / relief valves to examine for crud buildup, bending of certain actuator members or other signs of possible deterioration.
Low-14w Set Logic has been provided on three non-Automatic Pressure Relief System valves.
This logic is discussed in detail in the Section 3.2 Bases. This logic, through pressure sensing instrumentation, reduces the cpening setpoint and increases the blowdown range of the three selected valves following a scram to eliminate the discharge line water leg clearing loads resulting from multiple valve openings.
I.
Deleted l
l i
l
\\
151 3.6/4.6 BASES Amendment No. 30,76, 93 l
3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 4
Applicability:
Applicability:
Applies to the operating status of the primary Applies to the primary and secondary and secondary containment systems.
containment integrity.
Obiective:
Obiective:
To assure the integrity of the primary and To verify the integrity of the primary and secondary containment systems.
Specification:
Specification:
A.
1.
Suppression Pool Volume and 1.
Suppression Pool Volume and Temperature Temperature When irradiated fuel is in the reactor
- a. The suppression chamber water vessel and either the reactor water temperature shall be checked once temperature is greater than 212*F or work per day.
- b. Whenever there is indication of is being done which has the potential to drain the vessel, the following relief valve operation which adds heat to the suppression pool, the i
requirements shall be met, except as permitted by Specification 3.5.E.2:
pool temperature shall be continually monitored and also
- a. Water temperature during normal observed and logged every 5 minutes until the heat addition is operation shall be 590*F.
l terminated.
- b. Water temperature during test operation which adds heat to the
- c. A visual inspection of the suppression pool shall be $100*F and suppression chamber interior shall not be >90*F for more than 24 including water line regions and the interio* painted surfaces above hours.
the wa. -
ae shall be made at
- c. If the suppression chamber water each rt eling outage, temperature is >110 F, the reactor shall be scrammed immediately. Power operation shall not be resumed until the pool temperature is 590*F.
156 3.7/4.7 Amendment No. $3. 93 I
I n.
+ -, - -
.