ML20090A666
| ML20090A666 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Monticello |
| Issue date: | 04/30/1980 |
| From: | Cooper J EG&G, INC. |
| To: | |
| Shared Package | |
| ML17139C189 | List:
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| References | |
| FOIA-84-105 EGG-1183-4143, NUDOCS 8006240864 | |
| Download: ML20090A666 (14) | |
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1183-4143 TECHNICAL EVALUATION OF THE ELECTRICAL,.
INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OFTHE.
PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION OFTHE MONTICELLO NUCLEAR GENERATING PLANT, UNIT 1 (DOCK ET NO. 50-263)
__A P RIL.1980 l
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l Work Performed for Lawrence Livermore' Laboratory under U.S. Department of Energy Contract No. DE-ACOS-76 NVQ 1183.
NOTICE This report was prepared as an. account of work sponsored by the United States Government.
Neither the United States nor the United States Department of Energy, nor any of their employees, nor any of their contractors, subcontractors, or their emplojees, makes any warranty, express or impl ied, or assynes any legal liability or responsibility for the accuracy, completeness or useful-ness of any information, apparatus, prcduct or process disclosed, or represents that its use would not infringe privately owned rights.
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NOTICE Reference to a company or product name does not imply approval or recomendation of the product by EG&G, Inc. or the United States Department of Energy to the exclusion of others that may be suitable.
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Energy Measurements Group San Ramon Operations 1183-4143 c
TECHNICAL EVALUATION OF THE ELECTRICAL, INSTRUMENTATION, AND CONTROL DESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT FOlit SINGLE-LOOP OPERATION OF THE MONTICELLO NUCLEAR GENERATING PLANT, UNIT 1 (DOCKET NO. 50 263)
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by.
James H. Cooper Approved for Publication 0
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- dohn R. R'adosevic Department Manager This Document is UNCLASSIFIED Re$b [Wn M Richolas E.'Broderick Department Manager Derivative Classifier
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Work Perforrned for Lawrence Livermore Laboratory under U.S.' Department of Energy Contract No. DE. ACO8-76 NVO 1183.
ABSTRACT This report documents the technical evaluation of the electrical, instrumentation, and control design aspects of the proposed license amendment for single-loop operation of the Monticello nuclear generating plant, Unit 1.
The review criteria are based on IEEE Std-279-1971 re-quirements and the Code of Federal Reculations, Title 10, Part 50.46 and Part 50, Appendices A and K requirements for single-loop operation.
This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the U.
S.
Nuclear Regulatory Commission by Lawrence Livermore Laboratory.
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FOREWORD This report is supplied as part, of the Selected El ectrical,
Instrumentation, and Control Systems Issues (SE.ICSI) Program being conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Operating Beactors, by Lawrence Livermore Laboratory, Field 7 c Systems Division of the El ectronics Engineering Department.
The U. S. Nuclear Regulatory Commission funded the work under the authorization entitled "El ectrical,
Instrumentation and Control System Support," B&R 20 19 04 031, FIN A-0231.
This work was performed by EG&G, Inc., Energy Measurements Group, San Ramon Operations, for Lawrence Livermore Laboratory under U.
S.
Department of Energy contract number DE-AC08-76NV01183.
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TABLE OF CONTENTS Page 1.
INTRODUCTION..................*.....
1 2.
DESCRIPTION AND EVALUATION OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION...........
3 2.1 Description of the Proposed Changes.......
3 2.2 Evaluation of the P.roposed Changes........
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-3.
CONCLUSIONS.......................
7 3.1 Introd uct ion...................
7 3.2 Concl us ion s...................
7 REFERENCES............_.....
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TECHNICAL EVALUATION OF THE f
ELECTRICAL, INSTRUMENTATION, AND' CONTROL DESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION i
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THE MONTICELLO NUCLEAR GENERATING PLANT, UNIT 1 (Docket No. 50-263) b James H. Cooper EG&G, Inc., Energy Measurements Group, San Ramon Operations 1
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1.
INTRODUCTION
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i By 1'etter.1 to the U. S. Nuclear Regulatory Commission (NRC) dated E
September 7, 1976, the Northern States Power Company (NSPC) submitted information to support its proposed license amendment to operate the Monticello nuclear generating plant, Unit 1, with one recirculation loop out of service (i.
e., single-loop operation).
This infonnation represent-ed the licensee's analysis of significant events, based on a review of accidents and abnormal operational transients associated with power opera-i s
tions in the single-loop mod,e and provided by the nuclear steam supply F
system designer (.Generai Electric Company, Nuclear Energy Division (GE-NED)).
Conservative assumptions were employed in the GE-NED Report NED0-21252,2 dated June 1976, to ensure that the generic analyses for l
y boiling water reactors (BWR 3 and/or 4) were app}icable to the Monticello E
nuclear generating plant, Unit 1.
GE-NED submitted an additional report
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(NED0-20566-2,3 dated July 1978) of an analytical model for a lo s s-of-I coolant accident (LOCA) with one recirculation loop out of. service which is under review by the NRC Reactor Safety Branch (RSB).
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The purpose of this report is to evaluate the electrical, instru-r.
mentation, and control (EI&C) design aspects of the proposed license amend-2 4
ment as presented in f:ED0-21252 and using IEEE Std-279-1971 criteria and the Cede of Federal Regulations, Title 10, Part 50.46,5 and Title 10, Part 6
I 50, Appendix A -and Appendix X criteria.
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DESCRIPTION AND EVALUATION OF THE PROPOSED LICENSE AMENDMENT F'OR SINGLE-LOOP OPERATION 2.1.
DESCRIPTION OF THE PROPOSED CHANGES The licensee stai.es that, from its analysis of NED0-21252,2 the only changes necessary to the reactor protection system (RPS) are modifi--
cations to the:
(1.)
SCRAM trip settings of the average power range monitor (APRM) system.
(2)
Rod-block.setpoints of the rod block monitor (RBM) system.
Because of the different flow quantity and different flow path during single-loop operation, the APRM SCRAM trip settings, which are flow-biased according to the equation in the technical specifications, require re-setting to protact the reactor from overpower.
The rod-block setpoint equation is 'flowI a.;ed in the same way and with the same flow signal as bi the APRM'setpoint, and must also be modified to provide adequate local core protection for the postulated rod withdrawal error.
l The revised technical specifications propose single-loop operation at reduced safety settings for unlimited periods of time.
The revised technical specifications also propose a limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in which to. reduce the safety settings.. Use of Section 3.4.1.1.a of the standard BWR technical specifications will be required.
Section 3.4.1.1.a states that,
"With one recirculation loop not in operation, (reactor) operation may continue; restore both loops to operation within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be i at least hot shutdown ~within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."g n
The numerical values of.the new settings are delineated in the revised technical specifications which accompany the licensee's submittal.1 2.2 EVALUATION OF THE PROPOSED CHANGES The temporary changes in the settings of the trip points for the APRM and RBM must be made in the power-range cabinets in the control room and so must be done with the reactor shut down (i. e...with the mode switch in shutdown or refuel, condition 3, 4, 5) as required by the NRC Branch Technical Position ICSB 12. 9 These adjustments include readjusting the power and flow potentioneters in each of the six APRM channels and the two RBM channels.
One channel of the multichannel systems will be adjusted at a time and then returned to service.
Before all of the channels are re-turned to service, the new trip setpoints will be verified by the instru-ment engineer following the readjustment and testing of the setpoints by the instrument technician.
Two operators will perform ' functional tests to double check the new setpoints and to check the instrument's return to an operable condition.
The sequence outlined above shall be written into the pl ant technical specifications.
A permanent installation of the 'setpoint-change capability must be made in order for the system to satisfy the requirements of Section 4.15 of IEEE Std-279-1971.4 Hard-wire modifications will be required to enable making setpoint changes from the front panel of the power range cabinet by way of control switches.
The licensee's proposed modifications must be submitted to the NRC staff for review prior to this installation.
The recirculation-loop equalizer valves ~ must be verified closed
,and tagged for single-loop power operation, as i;s the case for two-loop 2
power operation.
The safety analysis in NED0-21252 assumes that these valves remain closed as their effect on a LOCA has not been analyzed.
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Recirculation, flow must be manually controlled by the operator, as opposed to autcmatic control, whenever the system is operating in the single-loop mode, since control stability is degraded and manual. control is assumed in the NED0-21252.2 The technical specifications will be changed -
to include this restriction.
Due to the different flow pattern during single-lcop operaticn as described by the licensee, a number of indications in.the control room will change, such as individual jet-pump flow and total summed core flow.
Some indications will be only slightly less accurate, but some others will be erroneous.
The control room indications must be corrected prior to single-loop power operation or they must be tagged out-of-service, as appropriate.
This is a requirement of Section 4.20 of IEEE Std-279-1971.4 The nonnal plant configuration as described in the final safety analysis report (FSAR )10 includes recirculation-pump start interlocks to prevent an inadvertent cold-water injection into the reactor.
Any recircu-lation loop that is out of service and whose water has cooled must be run in the bypass mode to preheat the water to within 100 F of the reactor cooling water before the water may be valved back to the reactor pressure vessel.
The recirculation pump start is interlocked to permit start-up only if the pump discharge valve is closed, the bypass valve is open, and the suction valve is open.
This configuration will limit the amount of cold water which can be transported through the reactor vessel from a cold-lcop startup, thereby limiting the effect of a cold-water slug event. -
Al though interlocks are provided, no credit is taken for their safety 2
function in NED0-21252 for single-loop operation since this is not the limiting transient.
The instrument setpoints can be set down to enable operation in the single-loop mode' for unrestricted periods.
This mode of operation is desired by the licensee to facilitate more extensive unscheduled mainte-i i
nance without the requirement of keeping the reactor shut down.
It is stipulated that single-loop operation will not be a planned mode of operation.
1.
The new rod-bl' ck and trip setpoints vary' linearly as a function o
of recirculation flow rate.
For power increases by rod withdrawal, the RBM rod block must be set to the next higher trip level by manual operator acticn.
The APRM, flow-biased, SCRAM trip.follows the r.cw trip curve automatically for both power increases and decreases.
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CONCLUSIONS
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3.1 INTRODUCTION
The initial licensee submittal analyzed in this report was based 2
on the analysis in NED0-21252 perfomed by the nuclear steam supply system manufacturer (GE-NED).
The manufacturer, however, 'had not analyzed the perfomance of the emergency core cooling system (ECCS) during single-loop operation.
A new analysis was performed by GE-NED for a LOCA with one recirculation loop out of service.
The analysis, reported in NEDO 20566-2,3 includes the ECCS single loop. analysis.
This NEDO report was published subsequent to the initial licensee submittal.
The licensee will submit mere detailed information based on the new analysis.
3.2 CONCLUSION
S We conclude that the concept outlined in the Northern States l
Power canpany's initial license amendment submittal is in accordance with the intent of IEEE-Std-271-1971.
At the conclusion of this review, the specific design criteria noted herein were ccmunicated to the licensee, as required by the NRC.
The licensee has agreed to address that criteria. in the detailed license amendment subnittal which was necessitated by the revised GE report.
We recomm'end that the NRC review the El&C aspects of the forthcoming detailed license amendment submittal for confomance to these criteria.
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O REFERENCES 1.
NSPC Letter to NRC (V. Stello), dated September 7,1976.
2.
General Electric Canpany, Nuclear Energy Division, License Amend-ment Submittal for Single-Looo Operation - Monticello Nuclear 3
Generating Plant, Unit 1, NE00-21252 (June 1976).
3.
General Electric Company Nuclear Energy Division, An Analytical Model for loss-of-Coolant Accident (LOCA) with One Recirculation Loop Out-Of-Service, NE00-20566-2 (July 1978).
<4.
IEEE Std-279-1971:
Criteria for Protection Systems for Nuclear Power Generating Stations (n. d.).
5.
Code of Federal Regulations, Title 10, Part 50.46:
Acceptance Criteria for Energency Core Cooling Systems for Light Water Nuclear Power Reactors (January 1976).
6.
Code of' Federal Regulations, Title 10, Part 50, Appendix A:
General Design Criteria for Nuclear Power Plants (January 1978).
7.
Code of Federal Regulations, Title 10, Part 50, Appendix K:
ECCS Evaluation Mcdels (January 1,1978).
8..
General Electric Company, Standard Boiling Water Reactor Tech-nical Specifications (April 1978).
9.
NRC/RSB, Protection System Trio Point changes for Operation With Reactor Coolant Puce Out of Service, Branch Technical Position ICSB 12 (n. d.).
f 10.
- NSPC, Final S~afety Analysis Report for Monticello Nuclear Power Station (FSAR) (n. d.).
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ENCLOSURE SAFETY EVALUATION REPORT N-1 LOOP OPERATION MONTICELLO NUCLEAR GENERATING PLANT
1.0 INTRODUCTION
The current Monticello Technical Specifications do not allow plant operation beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if an idle recirculation loop can not be returned to service.
The ability to operate at reduced power with a single loop is highly desirable from an availability / outage planning standpoint in the event that maintenance or. component unavailability renders one loop inoperable.
By letter dated July 2,1982, Northern States Power Company (NSP) (the licensee) requested changes to the Technical Specification for Single Loop Operation of Monticello.
The requested changes would permit Monticello to' operate at up to 50% of rated power with one recirculation loop out of service for unlimited time.
While analyses indicate that it may be safe to operate BWRs on a single loop in the range higher than 50% cf rated power, the experience (reference letter from L. M. Mills, TVA dated March 17, 1980 to H. Denton, NRC) at Browns Ferry Unit I has caused concern.about flow and power oscillations. 'However, because single loop operation at 50%' rated power at several plants, including
-Browns Ferry Plant Unit-1, has shown to result-in acceptable flow and power characteristics, we will permit NSP to operate at power levels up to.50% of rated with one loop out of service, during Cycle 10.
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-c-If requested, we will reconsider operation at a higher power level for Monticello with one recirculation loop out of service after staff concerns stemming frcm Browns Ferry - Unit 1 s. ingle loop operation are satisfied.
-2 EVALUATION 2.1 Accidents (Other than Loss of Coolant Accident (LOCA)) and Transients Affected by One Recirculation Looo Out of Service 2.1.1 One Pumo Seizure Accident The licensee states that the one-pump seizure accident is a relatively mild event during two recirculation pump operation.
Similar analyses were performed to determine the impact this accident would have on one recirculation pump operation.
These analyses were performed using NRC approved moo ~els for a large core BWR/4 plant.
The analyses were conducted from steady-state operation at the following initial conditions, with the added condition of one inactive recirculation loop.
Two sets of initial conditions were assumed:
Thermal Power = h5% and core flow = 58% of rated a.
b.
Thermal Power = 82% and core flow = 56% of rated These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing Maximum Average Linear Heat Generation Rate (MAPLHGR) and Minimum Critical Power Ratio (MCPR) limits at the same maximum pump speed.
Pump seizure was simulated by setting"the single operating pump speed to zero instantaneously.
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3 The anticipated sequence of events following a ' recirculation pump seizure which occurs during plant operation with the alternate recirculation loop out of service is as:follows:
's a.
The recirculation loop flow in the loop in which the pump seizure occurs drops instantaneously to zero.
b.
Core voids increase which resu.lts.in a negative reactivity insertion and sharp decrease in neutron fiux.
c.
Heat flux drops more slowly bec~ause of the fuel time constent.
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Neutron flux, heat flux, reactor water level, steam flow, and feedwater flow all exhibit transient behaviors.
However, it is not anticipated that the increase in water level will cause a turbine trip and result in scram.
It is expected that the transient will terminate at a condition of natural circulation and an orderly reactor shutdown will be accomplished.
There will also be a small decrease in system pressure.
a The licensee concludes that the MCPR for the pump seizure accident for g
-j the large core BWR/4 plant was determined to be greater than the fuel cladding integrity safety limit; therefore, no fuel failures were postulated to occur as a result of this analyzed e' vent.
These results are applicable to Monticello.
2.1.2 Abnormal Transients 2.1.2.1 a.
Idle Looo Startuo 4
The idle loop startup transient was anlayzed, in the Monticello FSAR, 4
with an initial power of 60%.
The licensee is to operate at no greater d>
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'than 50% power with one loop out of service.
Additionally, the Technical Specifications are being modified to require that, during single loop. operation, the suction valve in the idle loop be shut and electrically disconnected.
These measures are being taken to preclude startup of an idle loop.
b.
Flow Increase For single-loop operation, the rated condition steady-state MCPR limit is increased by 0.01 to account for increased uncertainties in the core total flow and Traversing In-core Probe (TIP) readings.
The 11CPR will vary depending on flow conditions.
This leads to the possibility of a large inadvertent flow increase which could cause the MCPR to decrease below the Safety Limit for a low initial MCPR at reduced flow conditions.
Therefore, the required MCPR must be increased at reduced core flow by a flow factor K The K factors are derived assuming both f,
g recirculation loop pumps increase speed to the maximum permitted by the scoop tube position set screws.
This condition maximizes the power increase and hence maximum-A MCPR for transients initiated from less than rated conditions.
When operating on one loop the flow and power i
increase will be less than associated with two pumps increasing speed, therefore, the K factors derived from the two-pump' assumption are f
conservative for single-loop operation.
c.
Rod Withdrawal Error The rod withdrawal error a.t rated power is gi.ven in the FSAR for the inital core and in cycle dependent reicad supplemental submittals.
These analyses are performed to demonstrate that, even if the operator
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Correction of the rod block equation and lower initial power for single-loop operation assures that the MCPR safety limit is nort violated.
One-pump operation results in backflow through 10 of the 20 jet pumps while flow is being supplied to the lower plenum from the active jet pumps.
Because of this backflow through the inactive jet pumps the present rod-block equation and APRM settings must be modified.
The licensee has modified the two-pump rod block ecuation and APRM settings i
that exist in the Technical Specification for one-pump operation and the staff has found them acceptable.
The staff finds that one, loop transients and accidents other than LOCA, which is discussed below, are bounded by the two loop operation analysis and are therefore acceptable.
i 2.2 Loss of Coolant Accident (LOCA) i l
l ihe licensee has contracted General Electric Co. (GE) to perform single loop operation analysis for Monticello LOCA.
The licensee states that i
evaluation of these calculations (that are performed according to the l
procedure outlined in NEDC-20566-2, Rev.1) indicates that a muliplier I'
' cf 0.85 '(8x8 fuel), 0.85 (8x8R Fuel) 0.85 (P8x8R fuel) (Ref:
NEDE l
24271 June 1980) should be applied to the MAPLHGR limits for sincie loop
' operation of Monticello,
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We, find the use of these MAPLHGR multipl ers to be, accept THERMAL HYDRAULICS _
3.
The licensee has confirmed that analysis uncertainties are independent of whether flow is provided by two loops or s The only exceptions to this are core total flow and TIP loop.
The effect of the'se uncertainties is an increa reading.
factor required at MCPR by.01, which is more than offset by the K f
The steady state operating MCPR with single-loop low flows.
li the K operation will be conservatively established by multip y ng 7
factor to the rated flow MCPR limit.
STABILITY ANALYSIS _
operating 4
As indicated in the applicant's submittal NED0-24271_,
ning along the minimum forced recirculation line with one pump ru l
at minimum speed is more stabie than operating with natura i h both circulation flow only, but is less stable than operating w t pumps operating at minimum speed.
d The licensee will be required to operate in master manua 1
k The staff the effects of instabilities due to contro 1er feedbac ht has accepted previous stability analyses results as eviden l
tion of BWR the core can be operated safely while_our generic eva ua The
. stability characteristics and analysis methods continues.
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ion previous stability analysis results include natural c rcu at In addition, conditions and thus bound the single loop operation.
llo shows the decay ratio (.62) predicted for cycle 10 cf Mentice
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.s margin relative to Browns Ferry #1 (.83) which had the flow noise oscillations during SLO.
We conclude that with appropriate limitations to recognize and avoid operating instabilities, that the reactor can be operated safely in the single loop mode.
Our evaluation of the flow / power oscillations evidenced in Browns Ferry will continue and any pertinent conclusions resulting from thi's study will be applied to Monticello.
5.
SUMMARY
ON SINGLE LOOP OPERATION 1.
Steady State Thermal Power Level will not exceed 50%
Operating at 50% power with appropriate T-S changes was approved on a long term basis for the Duane Arnold Plant, Peach Bottom Units 2 and 3, Pilgrim-1 and Cooper Nuclear Station (Safety Evaluation Reports (SER)datedNovember 19, 1981, May 15, 1981, December 15, 1981, and December 10, 1981 respectively).
Authorization for single loop operation for extended periods was also given to Dresden Unit 2 and 3 and Quad Cities Units 1 and 2 (SER July 9, 1981).
It was concluded that for operation at 50%, power transient and accident bounds would not be exceeded for these plants.
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2.
Minimum Critical power Ratio (MCPR) Safety Limit will be Increased by 0.01 to 1.08 The MCPR Safety Limit will be increased by 0.01 to account for increased uncertainties in core flow and Traversing Incore Probe (TIP) readings. 'The licensee has repor,ted that this increaso in l
l the MCPR Safety Limit was addressed in GE reports specifically for i
Monticello for one loop operation.
On the basis of previous staff L
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reviews for Pilgrim-1, Cooper, Duane Arnold'and Peach Bottom and our review of plant comparisons we find this analysis acceptable for Monticello.
3.
Minimum Critical Power Ratio (MCPR) Limiting Condition for Ooeration (LCO) will be Increased by 0.01.
The staff require that the operating limit MCPR be increased by 0.01 and multiplied by the appropriate two loop K factors that are f
in the Monticello T-S.
This will preclude an inadvertent flow increase from causing the MCPR to drop below the safety limit MCPR.
This was also approved by the staff for Peach Botte-2 and 3.
4 The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Limits will be Reduced by-Acorooriate Multioliers The licensee proposed reducing the T-S MAPLHGR by 0.85 (8x8 fuel);
0.85 (8x8R) and 0.85 (P8x8R) for single loop operation.
These reductions were based on analyses by General Electric (GE) in reports NEDE 24011-P-A-1 and NEDO 24271.
The Pea.ch Bottom units were allowed to operate with their MAPLHGR values reduced by factors of 0.71, 0.83, and 0.81 for an unlimited period of time for the first three types of fuel listed above.
l 5.
The APRM Scram and Rod Block Setcoints will be Reduced The licensee proposed to modify the two loop APRM Scram, Rod Block and Rod Block Monitor (R3M) se: points to, account for back flow through half the jet pumps.
The changes were based on piant specific anlayses by GE.
These setpoints equations will be changed
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The above changes are similar to the Peach Bottom T-S changes and are acceptable to the staff.
6.
The Suction Valve in the Idle Looo is Closed and Electrically Isolated The licensee will close the recirculation pumo suction valve and remove power from the valve.
In the event of a loss of coolant accident this would preclude partial loss of LPCI flow through the recirculation loop degrading the intended LPCI performance.
The removal of power also helps to preclude a start up of an idle loop transient.
7.
The Ecualizer line between the locos will be Isolated The licensee will close appropriate valves in the cross-tie (equalizer) line between the loops.
The previously discussed analysis assumed the two loops were isolated.
iherefore, itis required that the cross-tie valve be closed.
8.
The Recirculation Control will be in Manual Control The staff requires that the licensee operate the recirculatiun system in the manual mode to eliminate the neeb for control system analyses and to reduce the effects of potential flow instabilities.
This was also required of Peach Bottom.
9.
Surveillance Recuirements The staff requires that the licensee perform daily surveillance on the jet pumps to ensure that the pressure drop for one jet pump in
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.a loop does not vary from the mean of all j t pumps in that loop by
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more than 5%.
- 10. Provisions to Allow Ooeration with One Recirculation Looo out of Service 1.
.The steady-state thermal power. level will not exceed 50% of rated 2.
The Minimum Critical Power Ratio (MCPR) Safety Limit will be increased by.01 to 1.08 (T.S. 3.11C) 3.
The MCPR Limiting Condition for Operation (LCO) will be increased by 0.01 4.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits will be reduced by multiplying 0.85 for 8x8, 8x8R, P8x8R fuel respectively (T.S. reference Table 3.11.1) 5.
The APRM Scram and Rod Block Setpoints and the REM Setpoints, shall be reduced to read as follows:
T.S. 2.3.A.1 S ;c (.65W + 55% -0.65a W)
T.S. 2.3.A.1*
S ;< (.65W + 55% -0.65a W) FRP/MFLPD T.S. 2.3.2 S ;5 (.65W + 43%
.65a W)
T.S. 2.3.B*
S ;c (.65W + 43% -0.65a W)'
APRM Upscale jf.65 + 43% -0.65a W)
RBM Upscale < (.65W +4-3% - 0.65a W)
'In the event that MFLPD exceeds FRP.
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3 6.
The suction valve in the idle loop is closed and
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electrically isolated until the idle loop is being prepared for return to service.
7.
APRM flux noise will be measured once per shift and the recirculation pump soeed will be. reduced if the flux noise exceeds 5-percent peak to peak.
8.
The core plate delta noise be measured once per shift and "
the recirculation pump speed will be reduced if the noise exceeds 1 psi peak to peak.
Therefore, based upon the above evaluation and a history of successful operation of other BWRs of the same type as Monticello we conclude that single-loop operation of Monticello up to a power level of 50% and in accordance with the proposed TSs, will not exceed the accident and transient bounds previously found acceptable by the NRC staff and is therefore acceptable.
The approval for single loop operation up to a power level of 50%
is authorized during cycle 10 only.
1 We have concluded, based on the considerations discussed above, that: (1) because the enendment does not invols e a significant increase in the probability or consequences of accidents previously considered or create the possibility of an accident of a type different from any evaluated previously, and does not involve a
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. significant decrease in a safety margin, thh amendment does not involve a significant hazards consideration, (2) there is '
1 reasonable essurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of this. amendment will not be inimical to the common defense and security or to the health and safety of the public.
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