ML20095J608
| ML20095J608 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/15/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Northern States Power Co |
| Shared Package | |
| ML20095J614 | List: |
| References | |
| DPR-22-A-025, TAC 62137, TAC 62138 NUDOCS 8408290324 | |
| Download: ML20095J608 (18) | |
Text
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NUCLEAR REGULATORY COMMISSION
' UNITED STATES 2
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WASHINGTON, D. C. 20555
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NORTHERN STATES' POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 25 License No. DPR-22 k
1.
TheNuclearRegulatoryCommission(theCommission)hasfoundthat:
A..
The application for amendment by Northern States Power Company (the licensee) dated March 30,1984, complies with the standards and requirements of the Atomic Energy Act of.1954, as amended (the Act),
and the Coquission'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 wi.thout 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specificatiotis 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:
2 Technical Specifications The Technical Specifications contained in Appendix A at revised through Amendment No. 25, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
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This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION en Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes.to the Technical Specifications
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Date of Issuance:
,?.ugust 15,1984 -
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ATTACHMENT TO LICENSE AMENDMENT NO. 25 FACILITY OPERATING LICENSE NO. DPR-22 i-DOCKET NO. 50-263 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 area of change.
Remove Insert 17 17 18 18 19 19-20 20 i
84 84 117
.117 157 157 172 172
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'201
.. 201 209a 227a 227a 229p 229p 242 242 244 244 246b
- -246b l
O O
I 4
Bases Continued:
backed up by.the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawai is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a signific, ant percentage of rated. power, the rate of power rise is very slow.
Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed ~
uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The IRM scram remains active until the mode switch is placed in the run position. This switch occurs when reactor pressure is greater than 850 psig.
The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps. During steady-state operation with one recirculation i
pump operating the equalizer line shall be open. Analysis of transients from this operating condition are less severe than the same transients from the two pump operation.
i The operator will set the APRM neutron flux trip setting no greater than that stated in Specifica-tion 2.3.A.I.
However, the actual setpoint can be as much as 3% greater than that stated in Specification 2.3.A.1 for recirculation driving tiows less than 50% of design and 2% greater than that shown for recirculation driving flows greater, than 50% of design due to the deviations 1
discussed on page 39.
B.
APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a HCPR less than the Safety Limit (T.S.2.1.A).
This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase l
in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow. range. The margin to the Safety Limit I
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I 2.3 BASES t
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4 Bases Continued:
increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored by the in-core LPRM system. When the maximum fraction of limiting power density exceeds the fraction of rated thennal reactor power.
I the rod block setting is adjusted i'n accordance with the fonnula in Specification 2.3.8.
If the APRM rod block setting should require a change due to an abnormal peaking condition, it will be i
done by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.
The operator will set the APRM rod block trip settings no griater than that stated in Specification
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2.3.B.
However, the actual setpoint can be as much as 3% greater than that stated in Specification 2.3.8 for recirculation driving flows less than 50% of design and 2% greater than that shown for
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recirculation driving flows greater than 50% of design due to the deviations discussed on page 39.
C.
Reactor Low Water Level Scram The reactor low water. level scram is set at a point which will l
assure that the water level used,in the bases for the safety limit is maintained.
The operator will set the low water level trip setting no lower than 10'6" above the top'of the active fuel. However, the actual setpoint can be as much as.6 inches lower due to the deviations i
discussed on page 39.
D.
Reactor Low Low Water _ Level ECCS Initiation Tri) Point The emergency core cooling subsystens are designed to provide sufficient cooling to tie core to dissipate the energy associated with the loss 3
of coolant accident and to limit fuel clad temperature to pell below the clad melting temperature to
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assure that core geometry remains intact and to limit any cla,d metal-water reaction to less than 1%.
The design of the ECCS components to meet the above criterion was dependent on three previously i
set parameters; the maximum break size, the low water level scram setpoint, and the ECCS initiation l-
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setpoint. To lower the setpoint for initiation of the ECCS could prevent the ECCS components from if i
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I meeting their criterion. To raise the ECCS initiation setpoint would be in a safe ' direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or i
during normally expected transients.
The operator will set the low low water level ECCS initiation trip setting > 6'6"
< 6'10" above. the top of the active fuel.
However, the actual setpoint can be as much as 3 inches lower than the,
6'6" setpoint and 3 inches greater than the 6'10" setpoint due to the deviations discussed on page 39._
i E.
Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided j
to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the
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turbine control valves due to a load rejection and subsequent failure of the bypass. This transient is less severe than the turbine stop valve closure with bypass failure and therefore adequate margin exists.
t F.
Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of110% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the Safety Limit (T.S.2.1.A) even during the worst case I
t rans ient that assumes the turbine-liypass.is closed.
C.
Main Steam Line Isolation Valve Closure Scram The main. steam line isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation closure. With the scram' set at 10% valve closure there is no increase in neutron flux.
H.
Main Steam Line Low Pressure Initistes Main Steam Isolation Valve Closure The low pressure isolation g
of the main steam lines at 825 psig was provided to give protection against rapid reactor depressurization r
f and the resulting rapid cooldown of the vessel.
- Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressure does not occur,* thus providing protection for the fuel p'
g cladding integrity safety limit. Operation at steamline pressures lower than 825 psig requires I
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Bases Continued:
that the reactor mode switch be in the startup position where protection of the fuel cladding f
integrity safety limit is provided by the IRM high neutron flux scram.
Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the avallability i
of the neutron scram protection over the entire range of applicability of the fuel cladding integrity l
safety limit.
I The operator will set this pressure trip at greater than or equal to 825 psig. However, the actual trip setting can be as much as 10 psi lower due to the deviations discussed on page 39.
References 2
1.
Linford, R.
B.,
" Analytical Methods of Plant Transient Evaluations for the General Electric 5
Boiling Water Reactor". NEDO-10802, Feb., 1973.
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Bases Continued 3.3 and 4.3:
A.
Reactivity Limitations
,i 1.
Reactivity Margin - core loading The core reactivity limitation is a restriction to be applied principally to the design of new fuel' which may.be loaded in the core or into a particular refueling pattern.
Satisfaction of the limitation can only.
be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. The generalized form is.that the reactivity of the corp loading will be limited so the core can be made suberitical by at least R + 0.25% Ak at the beginning of the cycle, with the strongest' control rod fully withdrawn and all others fully inserted. The value of R in % Ak is the amount by which the core reactivity, at any time in the operating cycle, is calculated to be greater than at the time of - the check; i.e., at the beginning of the cycle.
R must be a positive quan,tity or zero. A core which'contains temporary control or other burnable neutron absorbers may have a reactivity characteristic which increases with core lifetime, goes through a maximum and then decraases thereaf ter.
See Figure 3.3.2 of the FSAR for such a curve.
3 The value of R is the difference between the calculated core reactivity at the beginning of ' the operating cycle and the calculated value of core reactivity any time later in the cycle where it would be greater than at the beginning.' The value of R shall include the potential shutdown margin loss assuming full B C 4
settling in all inverted poison tubes present in the core.
New values of R must be calcualted for each new fuel cycle.
The 0.25% Ak in the expression R + 0.'25% Ak is provided as a finite, demonstrable, sub-criticality margin.
This margin is demonstrated by full withdrawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to inser't at least R + 0.25% ak in reactivity. Observation of sub-criticality in this condition assures sub-criticality with not only the strongest rod fully withdrawn but at'least a R + 0.25% ak margin beyond this, f
2.
Reactivity margin - stuck control rods 5.
2 Specification 3.3.A.2 requires that a rod be taken ou't of service if it cannot be moved 5
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Bases Continued 3.5:
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C.
The containment heat removal ~ portion of the RHR system is provided to remove heat energy frot.the
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containment in the event of a loss of coolant accident. RFor the flow specified, the containment longterm pressure is limited to less than 5 psig and, therefore, is more than ample to-provide the required heat removal capability.
Reference Section 6.2.3.2.3.
FSAR. The repair periods specified were arrived at as in 3.5.B above.
The containment cooling subsystem consists of two sets of 2; service water pumps, I heat exchanger, and 2 RHR pumps.
Either set of equipment is capable of performing ~the containment cooling func-tion.
Loss of one Ri!R service water pump does not seriousl'y jeoperdige the containment cooling -
capability as two of the remaining threet pumps can satisfy the' cooling requirements.. Since there is some redundancy lef t, a 30 day repair period is adequate. Ioss of I containment' cooling subsystem leaves one remaining system to perfors the containment cooling function.
The operable system is demonstrated to be operable cach day when the above condition occurs. Based on the fact that when one containment cooling subsystem becomes inoperable only one system remains
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which is tested daily. A 7 day repair period was specified.
The RilR service water system provides cooling for the RHR, heat exchangers and ' can thus maintain
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the suppression pool water within 11mits. With the -flow specified, the pool temperature limits are maintained as specified in Spec,1fication 3.7.A.1.
D.
High Pressure Coolant Injection The high pressure coolant injection system is provided tg-adequately-cool the core for all' pipe breaks smaller thsn those for which the LPCI or core spray subsystens can protect the core.
The HPCI meets this requirement without1the use of off-site AC power; For the' pipe breaks for which the HPCI is intended to function, the core never uncovers and is continuously cooled and thus no clad damage occurs. Reference Section 6.2.4.3 FSAR.
E The llPCI system is backed up by the automatic pressure relief system and either of two core spray S-systems or the LPCI system. Therefore, when the HPCI system is out of service, the automatic pressure relief and core spray systems and LPCI system are required to be operable.
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3.5 BASES 4
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLAhCE REQUIREMENTS d.
During reactor isolation conditions d.
Whenever there is indication of relief the reactor pressure vessel shall be valve operation with a suppression pool depressurized to <200 psic st normal temperature >160*F and the primary cooldown rates if the suppression c6olant system pressure >200 psig, an pool temperature exceeds 120*F.
extended visual examination of the suppression chamber shall be conducted e.
The suppression chamber water volume before resuming power operation.
shall be >68,000 and <77,970 cubic feet.
e.
The suppression chamber water volume shall be checked once per day, f.
Two channels of torus water level instru-mentation shall be. operable.
From and -
f.
The' suppression' chamber water volume after the date that one channel is made indicators shall be calibrated semi-or found to be inoperable for any reason, annually.
reactor operation is permissible only 2.
Primary Containment Integrity during the succeeding 30 days unless such channel is sooner made operable.
If both channels are made or found to be a.
Integrated Primary Containment Leak Test (IPCLT) inoperable for any reason, reactor opera-tion is permissible only during the The containment leakage rates shall be succeeding six hours unless at least demonstrated at the following test schedule one channel is sooner made operable, and shall be determined in conformance with the' criteria specified in Appendix J of 10 2.
Primary Containment Integrity CFR 50 using the methods and provisions of ANS! N45.4-1972:
Primary containment integrity, as defined in Section 1, shall be maintained at all
- 1. ; Three Type A Overall Integrated Containment times when the reactor is critical or when the reactor water temperature is above Leakage Rate tests shall be conducte'd at 212*F and fuel is in the reactor vessel
-40 + 10 month intervals during shutdown service (41 psig) during each 10-year except while performing low power physics at P i 37 tests at atmospheric pressure during or pe riod.
The third test of each
, @Et after refueling at power levels not to set shall be conducted during the shut-exceed 5 Mw(t).
down for the 10-year plant inservice inspection.*
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- The third test of the first 10-year service period shall be conducted during the 1980 Dj refueling shutdown. Thi;first test of the second 10-year period shall be conducted during the 1984 refueling shutdown.
3.7/4.7 157
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TABLE 3.7.1 PRIMARY CONTAIMENT ISOLATION t
Number of Maximum Isolation Valve Valves Operating Normal Croup Tdantification Inboard Outboard Time (Sec)
-Position 1
Main Steam Line Isolation 4
4 5*
Open 1
Main Steam Line Drain 1
1 60 Closed
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1 Recirculation Loop Sampic~Line 1
1 60 Closed 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
6C Closed Air Makeup 2
Suppression Chamber to Drywell 1
60 Open N Recirculation 2
2 Suppression Chamber Vent 2
60 Closed k
2 Suppression Chamber Vent Bypass 1
60 Open 5.
g 2
Shutdown Cooling System 1
1 120 Closed 5
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- Minimum closure time shall be >3 seconds ru 3.7/4.7 172 m
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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILI.ANCE REQUIREMElfrS f
service providing both the emergency diesel generators are operable.
l 2.
Reserve Transformers j
During power oper'ation one. reserve trans-former may be out of service for main-tenance if the second reserve transformer is operational and available for. automatic operation on loss of normal auxiliary power.
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3.
Standby Diesel Generators B.
3.
Standby Diesel Generators a.
From and af ter the date that one of a.
Each diesel generat'or shall be the diesel generators is made or found manually started and loaded once to be inoperable for any reason, reac-every month to demonstrate opera-tor operation is permissible only tional readiness. The test shall during the succeeding seven days unless continue until both the diesel such diesel generator is sooner mede engine and the generator are at operable, provided that during such i
equilibrium conditions of tempera-seven days the operable diesel genera-ture while full load output is tor shall be demonstrated to be opera-maintained.
ble immediately and daily thereafter.
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b.
If both diesel generators become b.
During the monthly generator test, E
inoperable during power operation, the the diesel starting air compressor reactor shall be placed in the cold shall be checked for operation and I
I shutdown condition.
their ability to recharge air receivers, o
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m.m.=
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g Bases (continued):
si D.
Minimum Shutdown Period A minimum shutdown period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is specified prior to movement of fuel within the reactor since analysis of refueling accidents assume a 24-hour decay time following extended operation at power.
Since the reactor must be shut down, depressurized, and the head removed prior to moving fuel, it is not expected that fuel could actually be moved in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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J 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMElffS E.
Sprinkler Systems E.
Sprinkler Systems,
1.
The following spray or sprinkler systems l.
Each of the spray or sprinkler systems shall be cperable whenever equipment in the Itsted in specification 3.13.E.1 shall protected area (s) is required to be operable:
be demonstrated operable as follows:
I a.
Diesel Generator and Day Tank Rooms a.
Each valve (manual, power operated.
l l
b.
Lube Oil Drum Storage or automatic) in the flow path that c.
Lube Oil Storage Tank Sprinkler
, sealed or otherwise secured in position, is not electrically supervised, Jocked, I
d.
Hydrogen Seal Oil Unit Sprinkler i
e.
Lube Oil Pipiag System Sprinkler shall be verified to be in its correct f.
Lube Oil Reservoir position every month.
g.
Recirc HG Set Sprinklers l
h.
Intake Structure b.
Cycle each testable valve in the flow path through at least one complete 2.
If Specification 3.13.E.1 cannot be met, cycle of full travel once each year.
)
within one hour establish a continuous fire
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watch with backup fire suppression equipment c..
Perform a system functional test every for the unprotected area (s).
Restore the 18 months which includes, where appli-system to operable status within 14 days or.
cable, simulated automatic actuation submit a 30-day written report outlining of the system and verification that the i
the cause of the inoperability and the plans automatic valves in the flow path and schedule for restoring the system to actuate to their correct positions on a operable status.
i test signal.
i' d.' At' least once per 5 years by perfor, ming an l
air flow test through each open head sprinkler f((
header and verifying each open head sprinkler is unobstructed.
E 5
e.
At least once per 18 months by a visual examination of system piping and sprinkler
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heads. An air flow test shall be per-s%
formed upon evidence of obstruction of any 4
open head sprinkler.
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227a 3.13/4.13
Table 4.16.2 (Page 1 of 2)
MAXINUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)***
Airborne Particulate Water or Gag Fish Hilk Food Products Sediment Analysis (pC1/1)
(pci/m )
(pCi/kg, wet)
(pCi/1)
(pC1/kg, wet)
(pCf/kg, dry) b
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gross beta 4
1 x 10 b 3
2000(1000 )
g 54 15 130 g
59 30 260 Fe 58, 60 15 130 C0 65 30 260 Zn c
95 15 Zr-Nb b
d l,d 7 x 10-2 y
60
'131 g-g b
134,137 15(10 ), 18 1 x 10-2 130 15 60 150 Cs c
c 14 0 15 15 8a-La a
5 3.16/4.16 229p
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f.
All events which are required by regulation or technical specifications to be reported to NRC in writing f
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
g.
Drills on emergency procedures (including plant evacuation) and adequacy of communication with of f-site support groups, h.
All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan and the Security Plan (except as, exempted in Section 6.5.F), shall be reviewed with a frequency commensurate with their safety significance but at an interval of not more than two years.
i.
Perform special reviews and investigations, as requested by the Safety Audit Committee.
J.
Review of investigative reports.of unplanned releases of radioactive noterial to the environs.
k.
All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM).
5.
Authority The OC shall be advisory to the Plant Manager.
In the event of disagreement between the recommendations of '.
the DC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the General Manager Nuclear Plants and the Chairman of the SAC for review.
6.
Records Hinutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed.
The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Manager Nucle.ar Plants and others designa,ted by DC Chairman or Vice Chairman.
7.
Procedures E
e A written charter for the OC shall bei prepared that contains:
E.
a.
Responsibility and authority of the group.
b.
Content and method of submission of presentations to the Operations Committee.
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4.
'6.5 Plant Operating Procedures Detailed written procedures, including the applicable check-off lists and instructions, covering areas listed below shall be prepared aild followed.
These procedures and changes thereto, except as specified in 6.5.G shall be reviewed by the Operation Committee and approved by a member of plant management designated by.the Plant Manager.
A.
Plant Operations 1.
Integrated and system procedures for nonnal startup, operation and shutdown of the reactor and all systems and components involving nuclear safety of the facility.
l l
2.
Fuel handling operations.
3.
Actions to be taken to correct specific and foreseen potential or actual malfunction of systems or components including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up actions required after plant protective system actions have initiated.
I 4.
Surveillance and testing requirements that could have an effect on nuclear safety.
l S.
Implementing procedures of the energency plan, including procedures for coping with energency conditions involving potential or actual releases of radioactivity.
6.
Implementing procedures of the fire protection program..
7.
Implementing procedures for the Process Control Program and Offsite Dose Calculation Manual including quality control measures.
i l
Drills on the procedures specified in A.3 above shall be conducted as a part of the retraining program.
Drills on the procedures specified in A.6 above shall' be conducted' at least semi-annually, including a check of consnunications g
with offsite support groups.
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2 6.5 244 b
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E.
Offsite Dose Calculation Manual (ODCM)
The ODCM shall be approved by the Commission prior to initial implementation.
Changes to the ODCM shall satisfy the following requirements:
1.
Shall be submitted to the Commission with the Semi-Annual Radioactiv'e Effluent release report for the period in which the change (s) were made effective. This submittal shall contain:
a.
sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information.
Information submitted should consist of 'a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).
1 b.
a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.
documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.
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2.
Shall become ef fective upon review and acceptance by the Operations Committec.
F.
~ Security i
Procedures shall be developed to in.plement the requirements of the Security Plan and the Security Contingency Plan. These implementing procedures, with the exception of those non-safety related procedures governing work activities exclusively applicable to pr performed by security personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager.
Security proceduresi not reviewed by 'the Operations Committee shall be reviewed and approved by the Superintendent, Security and Services.
EL G.
Temporary Changes to Procedures R
5 Temporary changes to procedures described in A, B, C, D, E and F above, which do not change the intent i
of the original procedures may be made with the concurrence of two individuals holding senior ' operator
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Licenses.
Such changes should be documented, reviewed by the Operations Committee and approved by a member of plant management designated by the Plant NMnager within one month, to 6.5 246b 1
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