ML20076E498
| ML20076E498 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/12/1991 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076E503 | List: |
| References | |
| NUDOCS 9108200169 | |
| Download: ML20076E498 (8) | |
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Pt UNITED STATES E ",,,., !
I NUCLEAR REGULATORY COMMISSION t
WASHINoToN o C. 30006 l
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NORTHERN STATES POWER COMPANY DOCKET NO. 50 263 t
MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE I
Amendment No. 80 l
License No. DPR 22 i
1.
The Nuclear Regulatory Comission (the Comissind has found that:
f A.
The application for amendment by Northern States swer Company (the licensee)datedJune 13, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
i and the Commission's rules and regulations set forth in 10 CFR
-[
Chapter I; l
B.
The f acility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations of the i
Comission; i
C.
There is reasonable assurance (1) that the activities authorized by I
this amendment can be conducted without endangering the health and i
safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; L
D.
The issuance of this atiiendment will not be inimical to the common defense and security or to the health and safety of the public; and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 r
of the Commission's regulations and all_ applicable requirements have been satisfied.
[
t 2.
Accordingly, the license is amended by changes to the Technical i
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:
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9108200169 910012 PDR ADOCK 05000263
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2 Technical Specifications The Technical Specifications containen in Appendix A, as revised through Amendment No.
, are here5y incorportted in the license.
The licensee shall operate the facil t;, in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 80 1
Ledyard B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects :ll/IV/ V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
l August 12, 1991 t
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$J'TACHMENT TO LICENSE AMEN 0 MENT NO.-80 FACILITY OPERATING LICENSE NO. DPR-22 D_0CKET NO. 50-263 i
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Revise Appendix A Technical Specifications by. removing the pages identified j
below and inserting the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating _ the area of change.
Remove
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164 164 179_
179
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180 180 202 202
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204 204
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Y 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 4. Pressure Suppression Chamber-Drywell Vacuum 4.
Presnure Suppression Chamber-Drywell Vacuum Breakers Breakers a.
When prima g coatainment is required! all a.
Operability and full closure of the eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum breakers shall be operable and positioned breakers shall be verified by performance in the closed position as indicated by of the following:
the position indication system, except during testing and except as specified in (1)
Monthly each operable dryvell-3.7.A.4.b through 3.7.A.4.d below, suppression chamber vacuum breaker shall be exercised through an b.
Any drywell-suppression chamber vacuum opening-closing cycle.
breaker may be nonfully closed as indicated by the position indication and (2)
Once each operating cycle, dry-alarm system provided that drywell to well to suppression chamber suppression chamber differential pressure leakage shall be demonstrated to decay does not exceed that shown on be less than that equivalent to a Figure 3.7.1 one-inch diameter orifice and each vacuum breaker shall be visually c.
Up to two drywell-suppression chamber intpected.
(Containment access vacuum breakers may be inoperable required) provided that: (1) the vacuum breakers are determined to be fully closed and at (3)
Once each operating cycle, vacuum least one position alarm circuit is breaker position indication and operable or (2) the vacuum breaker is alarm systems shall be calibrated secured in the closed position or and functionally tested.
replaced by a blank flange.
(Containment access required) d.
Drywell-suppression chamber vacuum (4)
Once each operating cycle, the l
breakers may be cycled, one at a time, vacuum breaners shall be tested to during containment inerting and determine that the force required i
deinerting operations to assist in to open each valve from fully purging air or nitrogen from the closed to fully open does not suppression chamber vent header.
exceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc.
(Containment access required) 3.7/4.7 164 I
[
Amendment No. B, 26, 80
)
i s
Bases Continued:
One-inch opening of any one valve or a 1/8-inch opening for all eight valves, measured at the bottom of the disc with the top of the disc at the seat.
The position indication system is designed to detect closure within 1/8 inch at the bottom of the disc.
At each refueling outage and following any sigificant maintenance on the vacuum breaker' valves, positive seating of the vacuum breakers will be verified by leak test.
The leak test is conservatively designed to demonstrate that leakage is less than that equivalent to leakage through a one-inch orifice which is about 31 of the maximum allowable. This test is planned to establish a baseline for valve performance at the start of each operating cycle and to ensure that vacuum breakers are maintained as -
nearly as possible to their design condition. This test is not planned to serve as a limiting condition for operation.
During reactor operation, an exercise test of the vacuum breakers will be conducted monthly. This test will verify that disc travel is unobstructed and will provide verification that the valves are closing fully through the positioa indication systes.
If one or more of the vacuum breakers do not seat fully as determined from the indicating system, a leak test will he conducted to verify that leakage is within the maximum allowable. Since the extreme loser limit of switch detection capability is approximately 1/16", the planned test is designed to strike a balance between the detection switch capability to verify closure and the maximum allowable leak rate.
A special test was performed to establish the basis for this liciting condition. During the first refueling outage all ten vacuum breakers were shimmed 1/16" open at the bottom of the disc.
The bypass area associated with the shimming corresponded to 63% of the maximum allowable.' The results of this test are shown in Figure 3.7.1.
Two of the original ten vacuum breakers have since been removed.
When a drywell-suppression chamber vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights at the remote test panels are designed to function as follows:
Fully Closed 2 Green - On 2 Red
- Off Intermediate Position 2 Green - Off 2 Red
- Off Fully Open 2 Creen - Off
'2 Red
- On The remote test panels consist of indication and controls in the control room and indication in the reactor building. The (
71 room indication and controls for the drywell to suppression chamber vacuum breakers consist c red light and one green light for each of the eight valves, a common j
3.7 BASES 179 i;
Amendment No.
17, 80,
Bases Continued:
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vacuum breaker selector switch, and a common test switch. The reactor building vacuum breaker panel contains one red light and one green light for each of the eight valves. There are four independent limit switches on each valve. The two switches controlling the green lights are adjusted to provide an indication of disc opening of less than 1/8" at the bottom of the disc. These switches are also used to activate the valve position alarm circuits. The two switches controlling the red lights are adjusted to provide indication of the disc very near the full open position.
The control room alarm circuits are reduradant and fail safe. This assures that no simple failure will defeat alarming to the control room when a valve is open beyond allowable and when power to the switches fails. The alarm is needed to alert the operator that action must be taken to correct a malfunction or to investigate possible changes in valve position status, or both.
If the alarm cannot be cleared due to the insbility to establish indication of closure of one or more valves, additional testing is required. The alarm system allows the operator to make this evaluation on a timely basis. The frequency of the testing of the alarms is the same as that required for the position indication system.
Operability of a vacuum breaker valve and the four associated indicating light circuits shall be established by cycling the valve. The sequence of the indicating lights will be observed to be that previously described. If both green light circuits are inoperable, the valve shall be considered inoperable and a pressure test is required immediately and upon indication of subsequent operation.
If both red light circuits are inoperable, the valve shall be considered inoperable, however, no pressure test is required if positive closure indication is present.
Oxygen concentration is limited to 4% by volume to minimize the possibility of hydrogen combustion following a loss of coolant accident. Significant quantities of hydrogen could be generated if the core cooling systems failed to sufficiently cool the core.
The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is more probable than the occurrence of the loss of coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in term of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be sufficient to perform the leak inspection and establish the required oxygen concentration. The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the containment but air could not leak in to increase oxygen concentration. Once the containment is filled with nitrogen to the required concentration, no monitoring of oxygen concentration is necessary. However, at least once a week the oxygen concentration will be determined as added assurance.
3.7 BASES 180 b
l Amendment No. 17, 35, 80 g
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I 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS c.
For the diesel generators to be t
c.
A t. least considered operable, there shall be once each Operating Cycle l
a minimum of 34,500 gallons of diesel durin5 shutdown simulate a loss of offsite.
power in conjunction with an ECCS actuation fuel (7 days supply for 1 diesel gen-test signal, and:
erator at full load @ 2500 KW) in the diesel oil storage tank.
1.
Verify de-energization of the emergency busses and load shedding from the emer-gency busses.
2.
Verifying diesel starts from ambient conditions on the auto-start signal and is ready to accept emergency loads within ten seconds, energizes the escrgency busses with permanently connected loads, energizes the auto-connected emergency loads in proper time sequence, and operates for' greater than five minutes while its generator is loaded with the emergency loads.
d.
During the monthly generator test, the diesel fuel oil transfer pump and diesel oil service pump shall be operated.
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g Once a month the quantity of diesel fuel I
e.
available shall be logged.
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f.
Once a month a sample of diesel fuel shall be E,
taken and checked for quality.
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3.9/4.9 202
Bases 3.9:
The general objective is to assure an adequate supply of power with at least one active and one standby source of power available for operation of equipment required for a safe plant shutdown, to maintain the plant in a safe shutdown condition, and to operate the required engineered safeguards equipment following an accident.
AC for shutdown requirements and operation of engineered safe 6uards equipment can be provided by either of the two standby sources of power (the diesel generators) or any of the three active sources of power (No.
1R, No. 2R, or No. lAR transformers). Refer to Section 8 of the USAR.
To provide for maintenance and repair of equipment and still have redundancy of power sources, the requirement of one active and one standby source of power was established. The plant's main generator is not given credit as a source since it is not available during shutdown.
The plant 250 V de power is supplied by two batteries. Most station 250 V loads are supplied by the original station 250 V battery. A new 250 V battery has been installed for HPCI loads and may be used for other station loads in the future.
Each battery is maintained fully charged by two associated chargers which also supply the normal de requirements with the batteries as a standby source during emergency conditions. The plant 125 V de power is normally supplied by two batteries, each with an associated charger.
Backup chargers are available.
The minimum diesel fuel supply of 34,500 gallons will supply one diesel generator for a minimum of seven days of full load (2500 KW) operation. Actual fuel consumption during tLis period would be 33,096 gallons, but the minimum tank level has been established at the higher 34,500 gallon value to allow for instrument inaccuracy, tank volume uncertainties, and the location of the suction piping within the tank.
Additional diesel fuel can normally be obtained within a few hours. Maintaining at least 7 days supply is therefore cons e rva tive.
In the normal mode of operation, power is available from the off-site sources. One diesel may be allowed out of service based on the availability of off-site power provided that the remaining diesel generator is demonstrated to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This test is required even if the inoperable diesel is restored to operability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Thus, though one diesel generator is temporarily out of service, the off-site sources are available, as well as the remaining diesel generator. Based on a monthly testing period (Specification 4.9), the seven day repair period is justified. (1)
(1)
" Reliability of Engineered Safety Features as a Function of Testing Frequency", l.M. Jacobs, Nuclear Safety, Volume 9 No. 4, July - August 1968.
3.9/4.9 BASES 204 Amendment No. U, 51, 75, 77, 80
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