ML20197H850

Proposed Tech Specs Re Control Rod Operability Limitations

ML20197H850
Person / Time
Site:	Prairie Island
Issue date:	11/14/1990
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20197H838	List: ML20197H832 ML20197H840 ML20197H850
References
NUDOCS 9011200205
Download: ML20197H850 (12)

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Prairie Island Nuclear-Generating Plant: ]

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, License Amendment. Request Dated November 14,.1990: #

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e-l Exhibit.B consists'.of existing Technical Specification pages with the proposed  ;}

.- >;; , . changes' written on'those pages. Existing pages affe'cted by this License _;

fAmendment Request are listed below: 4

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- .g \( ' , ' + i. l p 4 >t /~ 9011200205 901114 PDR ADOCK 05000282 :1 1 P PDC s 4 TS-iv RE'.' O ! 10/27/09 . TABLE OF CONTENTS (Continued) L .TS SECTION IlTLE PAGE 4 .3.10 Control. Rod and Power Distribution Limits TS.3.10 1 A. Shutdown Margin TS.3.10 1- l B. Power Distribution Limits TS.3.10 1 C. Quadrant Power Tilt Ratio TS.3.10-4 l D, Rod Insertion Limits TS.3.10 5 E. Rod Misalignment Limitations TS.3.10-6 F. Inoperable Rod Position Indicator Channels TS.3.10-6 i G. Ir. p ::ble Red L1=itatione- TS.3.10 7 H. Rod Drop Time TS.3.10 7- -I. Monitor Inoperability Requirements TS.3.10 8 J. DNB Parameters TS.3.10 8 3,11 Core Surveillance Instrumentation TS.3.11 1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13 1 , A.' Control Room Special Ventilation System TS.3.13 1 B. Chlorine Detection Systems TS.3.13 2 l 3.14 Fire Detection and Protection Systems TS.3.14-1 A. Fire Detection Instrumentation TS.3.14-1 B. Fire Suppression Water System TS.3.14 1- ~ -C. Spray and Sprinkler Systems TS.3.14 2 D. Carbon Dioxide System TS.3.14-3 E. Fire Hose Stations TS.3.14-3 F. Yard Hydrant Hose Houses TS.3.14-4 G. Penetration Fire' Barriers TS.3.14-4' 3.15 Event Monitoring Instrumentation TS.3.15 1 > A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15 1 C. Reactor Vessel Level Instrumentation TS.3.15 2 -h w C.od < o l Rod. O p e m b't1\-h g _L;s e -b a t h o w s i I I l -- 1 TS-x RE" 01 10/27/00 TABLE OF CONTENTS (continued) .i If BASES SECTION IIH,E EAGI 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2 .1- Safety Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, ' Protective B.2;3 1 Instrumentation  ! 3.0, BASES FOR LIMITING CONDITIONS FOR OPERATION-3.0 B.3.0-1 Applicability 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressuce/ Temperature Limits B.3.1-4 l C. Reactor Coolant System Leakage B.3.1-6 l J D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolanc 0xygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature coefficient (ITC) B.3,1-9 3.2  ; Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1~ j 3.4 Steam and Power conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 3.6- Containment System B.3.6-1 3.7 Auxiliary Electrical. System B.3.7-1 - J: _3.8 Refueling and Fuel Handling B.3.8-1 3 '. 9 Radioactive Effluents B.3.9 A. Liquid Effluents B.3.9-1 B. Gaseous Effluents B.3.9-2 C. Solid Radioactive Waste B.3.9-4' D. Dose From All Uranium Fuel Cycle So trees B.3.9-5 E. & F. Effluent Monitoring Instrumentstion B.3.9-5 3.10 Control Rod and Power Distribution Lim!.ts B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10 1 r' . Quadrant Power Tilt Ratio . B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 ,,G ,In per:ble Red Lizitati:n: B.3.10-9 H. Rod Drop Time B.3.10-10

1. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance InstrumentationB.3.11-1 3.12 Snubbers B.3.12-1

.3.13 Control Room Air Treatment System B.3.13-1 3.14 Fire Detection and Protection Systems B.3.14 1 3.15 Event Monitoring Instrumentation B.3.15-1 _ , v v - ~ ~v N - oWh% } O d. Op a r as'? \ 's t $ Usv-sst v E'so w s - ~ ~ A# 4, , . Control Rod Operability Limitations TS.3.10-7 REY 7: 3/10/00 kt 1 o3.10.G. Inc;; rabic Rod Limitation;

l. An inoperable rod is a rod which (a) does not trip, (im) is

' declared inoperable under specification 3.10.Ex or 3.10.H. es- j cannot be moved by ;r.;; b; s its driv; ;;;har.i__;; ;.d _m _. - M s'a~ result of~ excessive friction or

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(me_chanicalinterference,or- ' 2. The reactor shall be brought to the HOT SHUTDOWN condition within

6. hours should more than one inoperable rod be discovered during POWER OPERATION.

3. If the. inoperable rod is located below the 200 step level and is capable of being tripped, or if the rod is located below the 30 step level whether or not it is capable of being tripped, then the  !

insertion limits'specified in the CORE OPERATING LIMITS REPORT' l ' apply. 4 If the inoperable rod cannot be located, or if the inoperable rod is. located above the 30 step level and cannot be tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT .l apply.

5. If POWER OPERATION is continued with-one inoperable rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is earlier made OPERABLE. The analysis shall' include due allowance.for nonuniform fuel depletion in.the neighborhood of the inoperable rod. If the analysis results in.a more limiting hypothetical- transient than the cases ,

reported in the safety analysis, THERMAL POWER shall be reduced to a level consistent with.the safety analysis. i H. Rod Droo' Time At operating temperature and full flow, the drop time of each RCCA shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to daahpot entry. If the. time is greater than 1.8 seconds, the rod shall be declared inoperable. )

6. With one or more rod (s) trippable, but immovable due to an electrical problem in the rod control system, within one hour verify that control rod position is within the rod insertion 1Lmits specified in section 3.10.D. Restore the Rod Control System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the affected rod (s) inoperable and apply the limitations specified in sections 3.10.G.2 through 3.10.G.S. _ ,__ l l

l l 1 ,y , l L . B.3.10 9 REV ^1 10/27/b> I 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS ER111 continued 1 D. Rod Insertion Limits (continued) as stated above. Therefore, this specification has been written to further minimize _the likelihood of any hypothesized event during the l performance of these tests later in life. This is accomplished by , limiting to two hours per year the time the reactor can be in this type of configuration, and requiring that a rod drop test is performed on , the rod-to be measured prior to performance of test. L Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special_ precautions are taken during the test. E. Rod Misalignment Limitation l Rod misalignment requirements are specified to ensure that power distributions more severe than those assumed in the safety analyses do not occur. F. Inoperable Rod Position Indicator Channels The_ rod position indicator channel is sufficiently accurate to detect a' l rod 17 inches away from its demand position. A misalignment less than 15 inches does not lead to over-limit power peaking factors. If the rod position-indicator channel is not operable, the operator will be fully aware of the inoperability of the channel, and special surveil-lance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or core thermocouples, and/or movable incore detectors,1will be used to verify power distribution symmetry. These indirect measurements do not have the same resolution if the bank is near either end of.the core, because a.15-inch misalign-ment would have no effect on power distributions. Therefore,Jit is necessary to apply the indirect checks following significant rod motion " - - - - - - - ,_ n. G. __ . . .. r . . .n. . . .m. . . . . . . . . . . i One inoperable control rod is acceptable provided that the power i distribution limits are met, trip shutdown capability is available, and provided the potential hypothetical ejection of the inoperable rod is. , not worse than _the ' cases analyzed in the safety analysis report. The i rod ejection accident for an isolated fully-inserted rod will be worse if the residence time of the rod is long enough to cause significant non-uniform fuel depletion. The four-week period is short compared with the time interval required to achieve a significant non uniform fuel depletion. - B.3.10 10 1:'l 2:  :/ /?O 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Bases continued H. Rod Drop Time The required drop'tima to dashpot entry is consistent with the safety analysis. I. Monitor Inoperability Requirements If either the rod bank insertion limit monitor or rod position devia-tion monitor are inoperable, additional surveillance is required to ensure adequate shutdown margin is maintained. If thetrod position deviation monitor and quadrant power tilt monitor (s) are inoperable, the overpower reactor trip setpoint is reduced (and also power) to ensure that adequate core protection is provided in the event that unsatisfactory conditions arise that could affect radial power distribution. Increased surveillance is required, if the quadrant power tilt monitors are inoperable and a load change occurs, in order to confirm satisfac-tory power distribution behavior. The automatic alarm functions related to QUADRANT POWER TILT must be considered incapable of alerting the operator to unsatisfactory power distribution conditions. J. DNB Parameters The RCS flow rate, T av and Pressurizer Pressure requirements are based ontransient-analyses $s,sumptions. The flow rate shall be veri fied by calorimetric flow data and/or elbow caps. Elbow taps are usem in the ~ reactor coolant system'as_an instrument device that indicates the status  ! of the reactor coolant flow. The basic function =of this device is to provide information as to whether or not a reduction in flow rate has occurred. If a reduction in flow rate is indicated below the value specified in the CORE OPERATING LIMITS REPORT, shutdown is required to investigate adequacy of core cooling during operation. In most cases, when more than one rod is found to be trippable but immovable, the malfunction can be traced to the Rod Control System. Since the majority of Rod Contral System malfunctions can be repaired  ! without reactor shutdown and since the unit conditions are not outside any accident analysis assumptions,-there is time available to locate the malfunction and restore the rods to an OPERABLE status. The rod insertion and power distribution limitations in specifications 3.10.D. 3.10.G.3, 3.10.G.4 and 3.10.G.5 ensure that core design limits are not exceeded. , . y. - . . . o - - .y, o 3 ...h , , + 3 .  ; , ,t x. . 73 - . g , , e ~~;5

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. y q ' L R, , u. Exhibit'C-  ; ' 1 ' Prairie Island Nuclear Generating' Plant. , ) 4- - License Amendment Request Dated November 14, 1990- ^ ' , g  ;] m , 4 ] Revised.. Technical Specification'~Pages < i .. Exhibit C' consists'of revised'pages'for-the Prairie Island Nuclear Generatin'g [ l Plant' Technical = Specification with the proposed changes incorporated.' The 3

• - lri.. revised' pages'are / listed below:

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• , l :L C. Quadrant' Power Tilt Ratio TS.3.10-4 (

Di Rod Insertion Limitss -TS.3.10 5- i E! Rod Misalignment Limitations TS.3.10 6' 3 , , F. :: Inoperable Rod Position Indicator Channels ; .TS,3'10 -

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. Gl Control Rod Operability Limitations TS.3.10-7 = . l!.: Rod Drop Time-. TS. 3 i10 7 // 3 ,I. . Monitor Inoperability Requirements -TS.3'.10 8' l' , y' _ JJ. DNB Parameters TS.3.-10;81 3.11' Core Surve111anceLInstrumentation' TS.3'11-1 . Li - 3 .12. Snubbers- :TS.3.12il. ' V', 3.13 Control Room Air Treatmant System TS.3.13L1 I e .A. Control; Room.Specia1' Ventilation System TS.3.13 l' W , B~. Chlorine Detection. Systems- TS.3013 2:- , Q ' 3.14 ' Fire Detection and Protection Systems TS.3.14-1 N! ' Ai-- Fire Detection. Instrumentation TS 3.14.1

• @"" uB'. Fire Suppression Water - System , TS.3.14L1L. -A s .t

.C. Spray-and Sprinkler systems TS.3~.14 2 , ,D; Carbon Dioxide. System: -TS.3;1423' I 'E. Fire llose Stations-- TS.3.14'3= Li -  : F. Yard Ilydrant flose..llouses TS.3.14 4  ! e -G.' Penetration-Fire Barriaars TS.3.14'-4 4 3.15 Event . Monitoring Instrumentation : TS.3il5 1 .A.' Process Monitors- .TS.3.15-1' B. itadiation Monitors TS.3.15-1 1 O. Reactor Vessel Level Instrumentation ~TS'3.15-2' d1 g -?? s a

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, o e T .i 3 .- c) f ,4 : if W 7 , l4 i , - , a a .- TS-x TABLE OF CONTENTS (continued) TS BASES SECTION. IITJS f.6fGl .2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2 1 2.3 Limiting Safety System Settings, Protective B.2.3 1 Instrumentation ?c 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION

3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 n' C. Reactor Coolant System Leakage B.3.1-6

'_ D. Maximum Coolant Activity B.3.1-7 E. Maxi:aum Reactor Coolant Oxygen, Chloride B.3.1 8 ( and Fluoride Concentration t- F. Isottermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3,2-1 3.3 : Engineered Safety Features B.3.3-1 T' 3.4 - fteam and Power Conversion Systems B.3,4-1 3.5 Instrumentation System B.3.5 1 s 3.6 : Containment System B.3.6-1 .= 3.7 Auxiliary Electrical System B.3.7 a '3.8 Refueling and Fuel Handling B.3.8-1 t 3.9 Radioaccive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 _- B.: Gaseous Effluents .B'3.9-2 'C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 'T E. & F. Effluent Monitoring Instrumentation B.3.9-5 .a.10 Control Rod and Power Distribution Limits B.3.10 1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control .B.3.10-1 I C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10 8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Control Rod Operability Limitations B.3.10-9 H. Rod Drop Time. B.3.10 10 l I. Monitor Inoperability Requirements B.3.10 10 t J. DNB Parameters B.3.10-10 I 3.11 Core Surveillance Instrumentation B.3.ll 1 3.12 Snubbers B.3.12 1 3.13 Control Room Air Treatment System B.3.13 1 3.14- Fire Detection and Protection Systems B.3.14 1 3.15 Event Monitoring Instrumentation B.3.15-1 L E y> I

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. .-- .. -[ TS.3.10 7 ^ P :la f ~ ' '3'.10.G. Control Rod Operability Limitations ] , -1. An inoperable rod is a rod which (a) does not trip, (b) cannot be-moved as'a result of excessive friction or mechanical interference, d or (c) is declared inoperable under specification 3.10.E or 3.10.H. . t

2. The-reactor shall be brought to the HOT SHUTDOWN condition within. d p 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> should more than one inoperable rod be discovered during. i POWER OPERATION. y

3. If the -inoperable rod is located.below the 200 step level and -is --

capable of being tripped, or if the rod is located below.the.30 , .." i step level whether or not it is capable'of'being. tripped, then the insertion limits specified in the CORL OPERATING LIMITS REPORT apply. .1

4. If the inoperable rod cannot be located, or if the inoperable' rod is located above the 30 step level -and -cannot be tripped, then the insertion limits specified in the CORE OPERATING LIMITS REPORT. ,

apply. * '5 . If POWER OPERATION is con'tinued with one: inoperable rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis' j s 'within.30 days unless the rod is earlier made OPERABLE. The . analysis shall' include due allowance for nonuniform fuel depletion  ; e' in the neighborhoodcof the inoperable rod. If the' analysis 1 i Eresults in a more limiting hypothetical-transient than the cases reported in the safety analysis; THERMAL POWER shall be. reduced to 1 a level consistent with the safety analysis, i i

6. Withione'or.more rod (s)-trippable, but immovable due to an electrical problem in the rod control system, within.one' hour 'l verify that control rod position,is within the rod insertion limits i specified in section 3.10.D. Restore.the. Rod Control ^ System to-OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the affected: rod (s)- ,

inoperable and apply the limitations specified'in sections 3.10.012 .j through 3.10.G.5. H. Rott'Dron Time 1 . At operating temperature and full flow, the drop time of each RCCA shall be.no greater than 1.8 seconds from loss of stationary gripper  ; .y ' ' 4 ;" 4 coil voltage to dashpot entry. If the time :is greater than 1.8 , ' seconds, the rod shall be declared inoperable. j ,k.. N ~ > iU s i.;  ; ; . - fi 3 , _. s B.3.10-9 -! .y , , ;t: , , 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS l a- . i1 Bases continued j g l , , e D. Rod Insertion Limits (continued) l J 'as stated above. .Therefore, this specification has been written to~ 4 L further minimize- the likelihood of any hypothesized event during the l , performancerof these tests later in life. This is accomplished by .. j m_* 1 limiting to. two hours per year the' time the reactor can be :In 'this type. , of configuration, and requiring ~that a rod drop-test is performed on ~ .the rod to he measured prior to performanceLof test, i Operation with abnormal, rod cor. figuration during low power-and zero 3 power testing is permitted because of the brief period of the test and :7 , ;7L . .because.special precautions are taken during the test. p - , . lE. Rod Misalignment Limitation

Rod misalignment requirements are specified to ensure that powe- . _

distributions more' severe-than those assumed in the safety l analyses do g; Enot occur. it W 7M < F4 Inoperable' Rod Position Indica' tor Channels 4 .w N, zThe-rod position indicator channe1~is sufficiently accurate to detect a-h' irod 7. inches /away from its demand position. ,A misalignment less than 4 "i!' 15; inches -does not lead to over-limit power peaking factors. If the- ' rod position indicator channel .is not operable,' the operator will'be : A 'l fully aware:of the inoperability of the channelt and special' surveil 4 j , lance of. core power tilt indications, using established procedures.and <

relying on excore nuclear. detectors,'and/or. core thermocouples, and/or movable incore-detectors, will:be used toiverify power distribution.

, syrnetry, f These _. indirect measurements -do;not have the- same - resolution . iflhe bank is near either end of the core,1because a ?15 inch misalign- 'q" ' ment would have no effect on power distributions, . Therefore, it is

necessary to apply the indirect checks following significant rod J motion.; '

l G. Control Rod Operability Limitations .

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q' ; s .s . . .One . inoperable control rod'is acceptable provided that the. power ~ .+ 'dtstribution limits are met,: trip shutdown. capability is available,' ond N" provided the' potentini hypothetical ejection ofithe . inoperable rodvis- , .. not worse thanathe cases analyzed in the safety analysis report. The M,G rod ejection l accident- for an isolated fully-inserted rod will' be, worse . q if the. residence time of the. rod is lon6 enough to cause significant t l! w non uniform fuel depletion. The four-week period-is short compared j u with :the time interval required to achieve.a significant nou-uniform fuel depletion. p { \ s a k a .O L .- 3 f - 3,3,10 10 4 =3110. CONTROL; ROD AND POWER DISTRIBUTION LIMITS: f Bases continued' [ 'In'most cases, when more than one rod is found to be trippable'but immovable , thefmalfunction can be traced to the Rod Control System. Since 'the' majority of Rod Control System malfunctions can be repaired , 'without reactor shutdown and since the unit conditions are not outside ' any cecident analysis assumptions, there is._ time-available'to locate d the malfunction and restore the rods to an OPERABLE status. The rod insertion and power distribution limitations in specifications 3.10.D, '3.10.G.3, 3.10.G.4 and '3.10.G.5 ensure that core design limits are .nc,t exceeded. ] . > .H. Rod Drop Time. x ~ The required drop time to dashpot entry is consistent with the safety. l ' analysis. j Monitor Inoperability Requirements ~ I. If'either the rod bank insertion limit monitor or rod position davia. 1 tion. monitor are inoperable, additional 1 surveillance.is required to ensure-adequate. shutdown margin-.is maintained. If the rod position deviation monitor and quadrant _ power tilt monitor (s) c are-inoperable, the overpower reactor trip setpoint is-reduced (and also " power) to ensure'that adequate core protection is provided in the event j that 3unsatisfactory conditions.arise that couldLaffect radial power .i distribution.- I q Increased' surveillance is. required,_'if the' quadrant _ power tilt monitors 'j i are inoperable.and a~1oad change occurs, in. order to confirm'satisfac- 1 tory power distribution behavior. The; automatic alarm functions related ,1 1to QUADRANT POWER TILT must be considered incapable of-alerting the- ~4 o -operator'to unsatisfactory power distribution conditions. >J. DNB Parameters 3 -l " The RCS flow rate,=T av and Pressurizer Pressure requirements are based ontransientanalyses$s,sumptions. The flow rate.shall be verified by' calorimetric flow data and/or elbow taps. Elbow taps are used in'the reactor' coolant' system as an instrument device that indicates the status. of the reactor coolant flow. The basic function of this' device is to 4 provide information as to whether or not a.. reduction in flow rate has. 1 J occurred. If a reduction in flow rate is' indicated below the value 'specified in the CORE OPERATING LIMITS REPORT, shutdownsis required to ]; investigate adequacy of core cooling during operation, q

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