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| Volume 4: SECTIONS 3.3.1.1 ~- 3.3.4.1 - ; -I Remove Replace B 3.3.1.1 ITS pg B 3.3.1.1-14 Rev 2 B 3.3.1.1 ITS pg B 3.3.1.1-14 Rev 15 B 3.3.1.1 ITS pg B 3.3.1.1-15 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-15 Rev 15 B 3.3.1.1 ITS pg B 3.3.1.1 16 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-16 Rev 15 B 3.3.1.1 NUREG M/U pg B 3.3-15 B 3.3.1.1 NUREG hW pg B 3.3-15 Rev 15 I
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| l Rev 15 09/03/99 P
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| [
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| .-__._.,1 - . - -
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| RPS Instrumentation B 3.3.1.1 I
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| BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued)
| |
| : 5. Main Steam Isolation Valve-Closure HSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve-Closure signal before the MSIVs are completely closed in anticipation of the complete loss l m of the normal heat sink and subsequent overpressurization 4 transient. However, for the overpressurization protection analysis pf Reference 4. the Averagt. Power Range Monitor Neutron Flux-Upscale Function, alcag with the SRVs. limits the peak RPV pressure to less tnan the ASME Code limits.
| |
| y That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.
| |
| The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS. ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| MSIV closure signals are initiated from aosition switches located on each of the eight MSIVs. Eac1 MSIV has one J position switch that provides the originating sensor for two se)arate channels: one inputs to RPS trip system A while the otler inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve-Closure channels, each channel consisting of one position swi ch, u which is shared with one other channel.
| |
| 'f l The logic for the Main Steam Isolation Valve-Closure l Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.
| |
| The Main Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.
| |
| Sixteen channels of the Main Steam Isolation Valve-Closure Function, with eight channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This function is only required in MODE 1 since, with the MSIVs open and the heat generation rate UNIT 2 B 3.3.1.1 - 14 Revision 15 09/03/99 l FERMI
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| - " , .- ,,,,__N--
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| o RPS lnstrumentation B 3.3.1.1 BASES APPLICABLE SAFETY t.NALYSES. LCO, and APPLICABILITY (continued) high, a pressurization transient can occur if the MSIVs l close. In MODE 2. the MSIV closure trip is automatically bypassed, and the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.
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| : 6. Main Steam Line-Hiah Radiation Main Steam Line-High Radiation Function ensures prompt reactor shutdown upon detection of high radiation in the I vicinity of the main steam lines. High radiation in the !
| |
| vicinity of the main steam lines could indicate a gross fuel failure in the core. The scram is initiated to limit the '
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| fission product release from the fuel. This Function is not specifically credited in any accident analysis but is being retained for overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
| |
| Main Steam Line-High Radiation signals are initiated from four radiation monitors. Each monitor senses high gamma ,
| |
| radiation in the vicinity of the main steam line. The Main j Steam Line-High Radiation Allowable Value is selected high l enough above background radiation levels to avoid spurious stran, yet low enough to promptly detect a gross release of fission products from the fuel.
| |
| Four channels of Main Steam Line-High Radiation Function with two channels in each trip system, arranged in a one-out of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid signal. This Function is required in MODES 1 and 2 where considerable energy exists such that steam is being produced at a rate which could release considerable fission products from the fuel.
| |
| The Allowable Value is based on the NRC guidelines of 3.6 times the full )ower background radiation level with nominal full power lydrogen injection rate. This Allowable Value remains fixed at this nominal full-power basis even when operating at reduced power and/or reduced hydrogen injection rates.
| |
| : 7. Drywell Pressure-High High pressure in the drywell could indicate a break in the RCPB. A reactor Scram is initiated to minimize the l FERMI UNIT 2 B 3.3.1.1 - 15 Revision 15. 09/03/99
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. ''). and APPLICABILITY (continued) possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure-High Function is assumed to scram the olant coincident with the Reactor Vessel Water Level-Low. Level 3 function in the analysis of the LOCA inside primary containment. The reactor scram reduces the amount of energy to be absorbed and helps the ECCS ensure that the fuel peak I cladding temperature remains below the limits of 10 CFR l 50.46.
| |
| High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.
| |
| Four channels of Drywell Pressure-High Function. with two i channels in each trip system arranged in a one out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS. resulting in the limiting transients and accidents.
| |
| Sa. 8b. Scram Discharae Volume Water Level-Hiah l The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the j displaced water, control rod insertion would be hindered.
| |
| , Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level-High Functions are an input to the RPS logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the UFSAR. However. they are retained to ensure the RPS remains OPERABLE.
| |
| SDV water level is measured by two diverse methods. The l level is measured by four float type level switches and four l level transmitters for a total of eight level signals. The I
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| outputs of these devices are arranged so that there is a signal from a level switch and a level transmitter to each RPS logic channel. The level mea'surement instrumentation satisfies the recommendations of Reference 8.
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| ] l FERMI - UNIT 2 B 3.3.1.1 - 16 Revision 15. 09/03/99
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 5. Main steam Isolation valve-Closure (continued)
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| SAFETY ANALYSES, LCO, and The reactor scram reduces the amount of energy required to p,L __ APPLICA8ILITY be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the n a,PS*S M {5p O) limits of 10 CFR 50.46.
| |
| gigina41 p 5, aask . MSIV closure signal are initiated from position switches .
| |
| C k
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| * e *\ ; located on each of he eight MSIVs. Each MSIV has cwe
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| _ f.k position switt. one inputs to RPS trip system A whi e e other inputs to RPS trip system B. Thus, each RPS trip system receives an inout from eight Main Steam Isolation g g> Valve-closure channels, eachtconsisting of one position '
| |
| switch) The logic for the Main Steam Isolation W4D5
| |
| 'j valve-Closure Function is arranged such that either the l
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| #6 a#.uj oM inboard or outboard valve on three or more of the main steam c lines must close in order for a scram to occur.
| |
| l g,t The Main Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a )
| |
| significant reduction in steam flow, thereby reducing the ]
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| severity of the subsequent pressure transient.
| |
| l Sixteen channels of the Main Steam Isolation Valve-Closure l Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument l
| |
| failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, heat generation rate is low enough so i that the other divers RPSfunctionsprovide_ sufficient]soradent/7 j protection.
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| NSCAf (33,3. I l- (continued)
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| BWR/4 STS B 3.3-15 Rc.tI,04/07/95 I
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| Rev /r 9' 'w"_-
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| i l
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| }
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| INSERT THIS PAGE IN FRONT OF VOLUME 5
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| . _ . . 7;l re:;; % Volumeb; SECTIONS 3.3.5.1 - 3.3.8.2 4 5 ;
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| Remove Replace B 3.3.5.2 ITS pg B 3.3.5.2 3 Rev 0 B 3.3.5.2 ITS pg B 3.3.5.2-3 Rev 15 B 3.3.5.2 ITS pg B 3.3.5.2-4 Rev 0 B 3.3.5.2 ITS pg B 3.3.5.2-4 Rev 15 I B 3.3.5.2 ITS pg B 3.3.5.2-5 Rev 12 B 3.3.5.2 ITS pg B 3.3.5.2-5 Rev 15 B 3.3.5.2 NUREG M/U pg B 3.3-144 Rev 12 B 3.3.5.2 NUREG M/U pg B 3.3-144 Rev 15 -
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| Rev 15 09/03/99
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| -h, % -.uN
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| I l
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| RClC System Instrue.entation B 3.3.5.2 BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued)
| |
| The individual Functions are required to be OPERABLE in MODE 1. and in MODES 2 and 3 with reactor steam dome pressure > 150 psig since this is when RCIC is required to be OPERABLE. (Refer to LC0 3.5.3'for Applicability Bases.
| |
| for the RCIC System.)
| |
| The specific Applicable Safety Analyses. LCO. and Applicability discussions are listed below on a function by
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| . Function basis.
| |
| : 1. Reactor Vessel Water Level-Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water level above the top of the active fuel.
| |
| Reactor Vessel Water Level-Low Low. Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
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| The Reactor Vessel Water Level-Low Low. Level 2 Allowable Value is set high enough such that for complete loss of .
| |
| feedwater flow. the RCIC System ficw with high pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1.
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| Four channels of Reactor Vessel Water Level-Low Low.
| |
| Level 2 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.
| |
| Refer to LCO 3.5.3 for RCIC Applicability Bases.
| |
| : 2. Reactor Vessel Water Level-Hioh. Level 8 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam supply. steam supply bypass, and cooling water supply valves to prevent overflow into the i .
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| g l FERMI UNIT 2 B 3.3.5.2 - 3 Revision 15. 09/03/99
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| i RCIC System Instrumentation B 3.3.5.2 l
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| BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) main steam lines (MSLs). (The injection valve also closes due to the closure of the steam supply valve.)
| |
| Reactor Vessel Water Level-High. Level 8. signals for RCIC are initiated from two slevel transmitters from the wide range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
| |
| The Reactor Vessel Water Level-High. Level 8 Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.
| |
| Two channels of Reactor Vessel Water Level-High. Level 8 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LC0 3.5.3 for RCIC Applicability Bases.
| |
| : 3. Condensate Storaae Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
| |
| Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal.
| |
| ; water for RCIC injection would be taken from the CST.
| |
| However, if the water level in the CST falls below a l preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve (consistency) automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump. the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
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| ; Two level transmitters and trip units in the HPCI system are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST.
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| h l FERMI UNIT 2 B 3.3.5.2 -4 Revision 15. 09/03/99
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| I l
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| 1 RCIC System Instrumentation B 3.3.5.2 BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued)
| |
| Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source.
| |
| Refer to LC0 3.5.3 for RCIC Applicability Bases. )
| |
| : 4. Manual Initiation l
| |
| The Manual Initiation channel provides manual initiation I capability to individual valves. There is one manual initiation channel for the RCIC System. '
| |
| The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the I
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| : d. Function is retained for overall redundancy and diversity of Q the RCIC function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value fo' this Function since the channel is mechanically actuated based solely on the position of the valve control. One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.
| |
| ~
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3.
| |
| Completion Times, specifies that once a Condition has been ,
| |
| entered, subsequent divisions, subsystems components, or j variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for -
| |
| inoperable RCIC System instrumentation channels provide a)propriate compensatory measures for separate inoperable c1annels. As such a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel. ,
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| l
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| ] FERMI - UNIT 2 B 3.3.5.2 - 5 Revision 15. 09/03/99 l
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| 1 l
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| n- O
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| RCIC System Instrumentation B 3.3.5.2 i
| |
| ,,, BASES APPLICABLE Manual Initiation SAFETY ANALYSES, LCO, and daunc[
| |
| The Manual Initiation paah Lit;; cuit s- hirc i x: - ";-9 l APPLICABILITY Ste th "C' Cy;;;; '-ithtu; h;h tht M 7:t-%-t t: -
| |
| (continued) th ;;;_ _ iis pruiw6U.. ;n.i. oi.i;.,n .od provides manual ! I g initiation capabilit There is one =h ht'n for the d RCIC System.
| |
| ggQ ggg ugd. e 04
| |
| ')g The Nanual Initiation Function is n sumedn.in;
| |
| .;;Hab clan,uf
| |
| - accident or transient analyses in t AR. However, the
| |
| ' Function is retained for overall re ancy and diversity of the RCIC function as required by__the_NRC in the plant j d licensing basis.
| |
| gQfg -
| |
| There is no Allowable Value for this Function since the l channel is mechanically actuated based solely on the -
| |
| position of the E " ' " . One channel of Manual '
| |
| Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.
| |
| ACTIONS Rev te: Certain LC0 Completion Times ar- 4 8 approved top ca In ord consee to use l the times, the license he Completion Times as required afety Evaluat on' 43 l A Note has been provided to modify the ACTIONS related to
| |
| : RCIC System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate l entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide l appropriate compensatory measures for separate inoperable i channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System j instrumentation channel. i (continued)
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| BWR/4 STS B 3.3-144 Rev 1, 04/07/95 rav %
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| . w, -y W
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| INSERT THIS PAGEIN FRONT OF VOLUME 8 6 ,a '^ i / . , , , Valmne 8: SECTION 3.7 .,; . , . .;
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| .g1 Remme Replace 3.7.4 NUREG M/U pg 3.715 Rev 13 3.7.4 NUREG M/U pg 3.7-15 Rev 15 B 3.7.4 NUREG M/U pg B 3.7 29 Rev 13 B 3.7.4 NUREG M/U pg B 3.7 29 Rev 15 3.7.4 JFDs pg i Rev 0 3.7.4 JFDs pg 1 Rev 15 l
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| l t
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| j' Rev 15 09/03/99
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| $ Control oor System (e.ifp c 3.7 ACTIONS (continued) hh}
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| CONDITION REQUIRED ACTION COMPLETION TIME E. Two4contro C$c - NOTE - -
| |
| subsystems inoperable LC0 3.0.3 is not applicable. [ ,",' 7 2 AC I'. J"/
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| during movement of . - -
| |
| irradiated fuel assemblies in the E.1 Suspend movement of Imediately
| |
| ;(secondary}< irradiated fuel containment, during assemblies in the Q.7,2. de feb C,3 ,
| |
| CORE ALTERATIONS, or ;(secondary}:-
| |
| during CPDRVs. containment.
| |
| M E.2 Suspend CORE Imediately ALTERATIONS. l (3.7,2 /7ctI% C. N 1 E.3 Initiate actions to Innediately suspend OPDRVs. /
| |
| (3 ,"/,2. Acis,on C _.:
| |
| SURVEILLANCF REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ; SR 3.7.h.1 Verifygach2::tr:1 ;; s ".C7"; i;,;t a hu I."] e-th:-
| |
| MMty te . -.. U. eee.ud hat 4 t.
| |
| A gs corr lro{ (corrs ae'r kretf aa{vff. Il- l10Vf5
| |
| ; th .6 95'F. - ' i l
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| l
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| - i;;;/4 STS--- 3.7-15 -Rn 1. O'/"/*"
| |
| . RetWll Rev 13/ ;
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| ^_ , _ h* hN
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| O . ,.
| |
| i' M44T fControlRoom.AC) B 3.7.8c- System
| |
| ~
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| BASES
| |
| ~.
| |
| ACTIONS E.1. E.2. and E.3 (continued) require isolation of the control room. This places the unit in a condition that minimizes risk.
| |
| If applicable, CORE ALTERATIONS and handling of irradiated -
| |
| fuel in the } secondary) containment must be suspended issnediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue untti the OPDRVs are suspended.
| |
| h SURVEILLANCE REQUIREMENTS SR 3.7 '
| |
| T M
| |
| This SR verifies that the heat removal capability of the quificofiatd system is sufficient to remove the control room heat load ,
| |
| lb. canhol Itbn4 ,b. Z_ ...L_.I.'.'
| |
| "_'5. 'II.I.I.I5 C "* !,,N* *I. *. . . . . .
| |
| .., ...'_.IIE..i.iE. ..I. _...
| |
| k l .pwp-c/d k/m a N Frequency is appropriate since significant deg,radation of '
| |
| Ithe the( ntrol Epen AC{ System is not expected over this time R Cs & )
| |
| REFERENCES I. uFSAR,Section36.4{.
| |
| .2. O rsAR, .S ee flo., fl. 'l . I l
| |
| 8WR/t !!! B 3.7-29 -Rev-h--04/M/M-l i
| |
| l ..
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| . get15 Il Rw 0 l h =m s e, i W
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| | |
| JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.7.4 CONTROL CENTER AIR CONDITIONING (AC) SYSTEM NON BRACKETED PLANT SPECIFIC CHANGES P.1 These changes are made to NUREG 1433 to reflect Fermi 2 current licensing basis: including design features, existing license requirements and commitments. Additional rewording, reformatting, and revised numbering is made to incorporate these. changes consistent ,
| |
| with Writer's Guide conventions. Refer to CTS Discussion Of Changes k to the related requirements for a detailed justification of changes made to the current licensing basis which are also reflected in the ITS as presented.
| |
| P.2 Bases changes are made to reflect plant specific design details, equipment terminology, and analyses.
| |
| P.3 Bases changes are made to reflect changes made to the Specification.
| |
| Refer to the Specification change (and associated JFD) for additional detail.
| |
| 1 P.4 The reference to the NRC Policy Statement has been replaced with a l
| |
| more appropriate reference to the Improved Technical Specification
| |
| " split" criteria found in 10 CFR 50.36(c)(2)(ii).
| |
| 1 FERMI UNIT 2 1 REVISION 15. 09/03/99l
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| | |
| f:!
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| INSERT THIS PAGE IN FRONT OF VOLUME 10 Jolume: 10 SECTIONS 3.9,3.10,4.0 & 5.0' --
| |
| 4 Remove Replace 3.10.5 ITS pg 3.10-15 Rev 0 3.10.5 ITS pg 3.10-15 Rev 15 3.10.5 NUREG M/U pg 3.10-15 . 3.10.5 NUREG M/U pg 3.10-15 Rev 15 3.10.5 JFDs pg i Rev 0 3.10.5 JFDs pg i Rev 15 1
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| i l
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| Rev 15 09/03/99
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| | |
| Single CPD Removal-Refueling 3.10.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods. Other than the 24 hours control rod withdrawn for the removal of the associated CRD. are fully inserted. j SR 3.10.5.2 Verify all control rods, other than the 24 hours control rod withdrawn for the removal of the associated CRD. in a five by five array centered on the control rod withdrawn for the removal of the associated CRD. are disarmed.
| |
| SR 3.10.5.3 Verify a control rod withdrawal block is 24 hours inserted.
| |
| SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 SR 3.10.5.5 Verify no other CORE ALTERATIONS are in 24 hours l () )) progress.
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| ] FERMI UNIT 2 3.10 15 Revision 15. 09/03/99 1
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| i
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| - .._-_..a
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| 1 Single CRD Removal-Refueling 3.10.5 SURVEILLANCE REOUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.5 Verify no CORE ALTERATIONS are in progress.
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| 24 hours (Doc A,g) fa #v '@
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| - 3.10-15 by 2,- 0 je7,2 mee game e.w m F .e mp -- .a 6A Fe6 +
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| I JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.10.5 - SINGLE CRD REMOVAL REFUELING NON-BRACKETED PLANT SPECIFIC CHANGES P.1 Not used.
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| P.2 Not used.
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| P.3 Bases changes are made to reflect changes made to the Specification.
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| Refer to the Specification change (and associated JFD) for additional detail.
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| P.4 Additional detail added to the Bases to reflect information relocated from CTS. Refer to CTS Discussion Of Changes to the related requirement for a detailed justification of changes made to the current licensing basis, which are reflected in the ITS as presented.
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| P.5 Revisions to NUREG 1433 are made to correct the presentation of ,
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| the appropriate steps for a control rod drive (CRD) removal. The NUREG requirements and associated Bases discussions infer that the control rod is fully withdrawn, and then in conjunction with j
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| disconnecting the position indication probe for the drive a control rod block is inserted and/or valving out of the CRD is l performed. However, after disconnecting the position indication probe the step preceding insertion of a control rod block and/or valving out the CRD is to de couple the control rod blade from the ;
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| drive, which necessitates applying another control rod withdraw j signal. This final withdraw signal can not be applied if a !
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| control rod block is in place or if the CRD is valved out.
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| ITS 3.10.5 presents revised requirements and Bases that are consistent with the actual procedures necessary to perform the applicable CRD maintenance. These changes are also the subject of l a pending generic change to NUREG 1433 (reference BWROG 42).
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| P.6 The reference to the NRC Policy Statement has been replaced with a more appropriate reference to the Improved Technical Specification
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| " split" criteria found in 10 CFR 50.36(c)(2)(ii).
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| P.7 ITS LC0 3.10.5.c and the Bases corresponding to the Bases for SR 3.10.5.5 clearly discuss the need to assurt. that no "other" (other than the control rod being withdrawn / removed) Core Alterations are in progress. NUREG SR 3.10.5.5 is corrected to explicitly add the k "other" phrasing (consistent with the LCO) to correct an editorial oversight.
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| FERMI UNIT 2 1 REVISION 15. 09/03/99l l
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| * mv ew4 A w. m er p r .une A 4 e m g e_pg
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| p INSERT THIS PAGE IN FRONT OF VOLUME 12
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| ' Volume 12: IMPROVED TECHNICAL SPECIFICATIONS ,
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| Remove Replace I 3.10.5 ITS pg 3.10-15 Rev 0 3.10.5 ITS pg 3.10-15 Rev 15
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| Single CRD Removal-Refueling 3.10.5 i
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| SURVEILLANCE REQUIREMENTS l
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| SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the 24 hours control rod withdrawn for the removal of the associated CRD are fully inserted.
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| SR 3.10.5.2 Verify all control rods, other than the 24 hours control rod withdrawn for the removal of the associated CRD, in a five by five array centered on the control rod withdrawn for the removal of the associated CRD are disarmed.
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| ! SR 3.10.b.3 Verify a control rod withdrawal block is 24 hours inserted. 1 i
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| SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 i
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| 24 hours l
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| hl SR 3.10.5.5 Verify no other CORE ALTERATIONS are in progress.
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| l l FERMI UNIT 2 3.10 15 Revision 15. 09/03/99
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| I INSERT THIS PAGE IN FRONT OF VOLUME 13 i
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| l Volume 13:' IMPROVED TECHNICAL SPECIFICATIONS BASES
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| B 3.3.1.1 ITS pg B 3.3.1.1 14 Rev 2 B 3.3.1.1 ITS pg B 3.3.1.1-14 Rev 15 B 3.3.1.1 ITS pg B 3.3.1.1-15 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-15 Rev 15 l
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| B 3.3.1.1 ITS pg B 3.3.1.1-16 Rev 0 B 3.3.1.1 ITS pg B 3.3.1.1-16 Rev 15 B 3.3.5.2 ITS pg B 3.3.5.2-3 Rev 0 B 3.3.5.2 ITS pg B 3.3.5.2-3 Rev 15 .
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| B 3.3.5.2 ITS pg B 3.3.5.2-4 Rev 0 B 3.3.5.2 ITS pg B 3.3.5.2-4 Rev 15 B 3.3.5.2 ITS pg B 3.3.5.2-5 Rev 12 B 3.3.5.2 ITS pg B 3.3.5.2-5 Rev 15 l
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| RPS Instrumentation l B 3.3.1.1 l BASES APPLICABLE SAFETY ANALYSES. LCO and APPLICABILITY (continued)
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| : 5. Main Steam [ solation Valve-Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve-Closure signal before the MSIVs are completely closed in anticipation of the complete loss n of the normal heat sink and subsequent overpressurization 4 transient. However, for the overpressurization protection analysis of Reference 4. the Average Power Range Monitor Neutron Flux-Upscale Function, along with the SRVs. limits the peak RPV pressure to less than the A5ME Code limits.
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| q That is, the direct scram on position switches for MSIV closure events is mt assumed in the overpressurization analysis.
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| The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS. ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has one position switch that provides the originating sensor for two separate channels; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip iL system receives an input from eight Main Steam Isolation g Valve-Closure channels, each channel consisting of one j position switch, which is shared with one other channel.
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| j The logic for the Main Steam Isolation Valve-Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.
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| The Main Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.
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| l Sixteen channels of the Main Steam Isolation Valve-Closure {
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| Function, with eight channels in each trip system, are l required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate i 1
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| i i
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| 'l FERMI - UNIT 2 B 3.3.1.1- 14 Revision 15. 09/03/99
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) high, a pressurization transient can occur if the MSIVs close. In MODE 2. the MSIV closure trip is automatically bypassed. and the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.
| |
| : 6. Main Steam Line-Hiah Radiation Main Steam Line-High Radiation Function ensures prompt reactor shutdown upon detection of high radiation in the vicinity of the main steam lines. High radiation in the vicinity of the main steam lines could indicate a gross fuel failure in the core. The scram is initiated to limit the fission product release from the fuel. This Function is not specifically credited in any accident analysis but is being retained for overall redundancy and diversity of the RPS as 3
| |
| required by the NRC approved licensing basis.
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| Main Steam Line-High Radiation signals are initiated from four radiation monitors. Each monitor senses high gamma radiation in the vicinity of the main steam line. The Main Steam Line-High Radiation Allowable Value is selected high enough above background radiation levels to avoid spurious scrams yet low enotgh to promptly detect a gross release of fission products fr(m the fuel.
| |
| Four channels of Main Steam Line-High Radiation Function with two channels in each trip system. arranged in a one-out of two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid signal. This Function is required in MODES 1 and 2 where considerable energy exists such that steam is being produced at a rate which could release considerable fission products from the fuel.
| |
| The Allowable Value is based on the NRC guidelines of 3.6 times the full ]ower background radiation level with nominal full power lydrogen injection rate. This Allowable Value remains fixed at this nominal full-power basis even when operating at reduced power and/or reduced hydrogen I
| |
| injection rates.
| |
| : 7. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. A reactor Scram is initiated to minimize the l
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| l l FERMI - UNIT 2 B 3.3.1.1 - 15 Revision 15. 09/03/99
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| .A__._._.2._'. re %. p,%,
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILITY (continued) possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure-High Function is assumed to scram the plant coincident with the Reactor Vessel Water Level-Low. Level 3 function in the analysis of the LOCA inside primary containment. The reactor scram reduces the amount of energy to be absorbed and helps the ECCS ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.
| |
| Four channels of Drywell Pressure-High Function, with two channels in each trip system arranged in a one out of two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS. resulting in the limiting transients and accidents.
| |
| 8a. 8b. Scram Discharae Volume Water Level-Hiah The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.
| |
| Therefore, a reactor scram i.; initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level-High Functions are an input to the RPS .
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| logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the UFSAR. However, they are retained to ensure the RPS remains OPERABLE.
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| SDV water level is measured by two diverse methods. The level is measured by four float type level switches and four level transmitters for a total of eight level signals. The outputs of these devices are arranged so that there is a signal from a level switch and a level transmitter to each ,
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| RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8.
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| ] l FERMI UNIT 2 E 3.3.1.1 - 16 Revision 15. 09/03/99 l
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| . . - ~ . ~ .
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| 1 RCZC System Instrumentation B 3.3.5.2 8ASES 1
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| APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued) {
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| The individual Functions are required to be OPERABLE in MODE 1. and in MODES 2 and 3 with reactor steam dome
| |
| )ressure > 150 psig since this is when RCIC is required to
| |
| )e OPERABLE. (Refer to LCO 3.5.3' for Applicability Bases for the RCIC System.)
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| The specific Applicable Safety Analyses. LCO, and Applicability discussions are listed below on a Function by
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| . Function basis.
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| : 1. Reactor Vessel Water Level-Low Low.__ Level 2 Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decreese too far, fuel damage could result. Therefore, the RCIC System is initiated at level 2 to assist in maintaining water level !
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| above the top of the active fuel.
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| Reactor Vessel Water Level-Low Law. Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
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| The Reactor Vessel Water Level-Low Low. Level 2 Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1. l Four channels of Reactor Vessel Water Level-Low Low.
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| Level 2 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.
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| Refer to LCO 3.5.3 for RCIC Applicability Bases.
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| : 2. Reactor Vessel Water level-Hich. Level 8 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam supply, steam supply bypass, and cooling water supply valves to prevent overflow into the I
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| UNIT 2 B 3.3.5.2 - 3 Revision 15. 09/03/99 h l FERMI
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| RCXC System Instrumentation B 3.3.5.2 BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued) main steam lines (MSLs). (The injection valve also closes due to the closure of the steam supply valve.)
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| Reactor Vessel Water Level-High. Level 8. signals for RCIC are initiated from two , level transmitters from the wide range water level measurement instrumentation, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
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| The Reactor Vessel Water Level-High. Level 8 Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.
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| I Two channels of Reactor Vessel Water Level-High. Level 8 '
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| Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.
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| : 3. Condensate Storace Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
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| Normally. the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST. l However, if the water level in the CST f alls below a preselected level. first the suppression pool suction valves automatically open. and then the CST suction valve 1 (consistency) automatically closes. This ensures that an j adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump. the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
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| Two level transmitters and trip units in the HPCI system are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST.
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| i g l FERMI UNIT 2 B 3.3.5.2-4 Revision 15. 09/03/99
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| RCIC System Instrumentation B 3.3.5.2 BASES l
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| APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source.
| |
| Refer to LCO 3.5.3 for RCIC Applicability Bases.
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| ~
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| : 4. Manual Initiation The Manual Initiation channel provides manual initiation capability to individual valves. There is one manual initiation channel for the RCIC System.
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| The Manual Initiation Function is not assumed in any
| |
| - accident or transient analyses in the UFSAR. However, the d Function is retained for overall redundancy and diversity of the RCIC function as required by the NRC in the plant Q licensing basis.
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| There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the valve control. One channel of Manual Initiation is required to be OPERABLE when RCIC is required to be OPERABLE.
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| ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3.
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| Completion Times specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that 1 Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.
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| l l FERMI UNIT 2 B 3.3.5.2 - 5 Revision 15. 09/03/99
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| ATTACHMENT 3 TO ,
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| 1 NRC-99-0079 DETROIT EDISON COMMENTS ON DRAFT ITS SAFETY EVALUATION I
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| 4
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| l I
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| Attachment 3 to NRC-99-0079 Page 2 Detroit Edison's Comments on Draft Safety Evaluation (SE)
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| As discussed with the NRC, these comments do not encompass formatting, references between ITS sections and the SE, and the updating of the SE for the most current information from supporting discussions of changes. These items should be checked in the final SE preparation following comment incorporation.
| |
| The following are specific comments on the Draft Safety Evaluation for the Fermi 2 Improved Technical Specifications:
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| Reference Comment
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| : 1. Page 1,3'd paragraph, There were other conference calls that took place
| |
| ... conference calls and meetings beyond July 8,1999. This statement should be that concluded on July 8,1999." revised to clarify this fact or remove reference to
| |
| ;- July 8.1999.
| |
| : 2. Page 5,2"d paragraph,"The Staff's The dates of the individual supplements should be evaluation..., including the restated here to avoid confusion with the table of supplements listed above...". recent license amendments provided above the statement.
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| : 3. Page 5, 3"' paragraph, "...will make This statement should be revised to indicate that the enforceable... the implementation license condition makes the implementation of new and revised SRs in the ITS". transition provisions enforceable and not the actual SRs, which are enforceable by way of their inclusion in the ITS.
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| : 4. Page 19, Section D -Type 2,3 and These statements should be revised to eliminate 4 paragraphs, "...specified in reference to relocation of details to plant
| |
| . . station procedures..." procedures. Relocations to procedures is not in accordance with the current NRC ITS conversion guidance.
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| : 5. Page 21, Section D,2"d bullet under This statement should be revised to indicate that the 1" paragraph, "UFSAR (which UFSAR includes the TRM."
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| includes the TRM by reference)
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| : 6. Page 32, Section E,1" paragraph This statement should be revised to accommodate after bulin '', 'The ,elocated the fact that the LCO for Decay Time (bullet 21) specifications...do not fall within satisfies Criterion 2.
| |
| the criteria for mandatory inclusion in the TS in 10 CFR 50.36 (c)(2)(ii)."
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| i i
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| Attachment 3 to NRC-99-0079
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| ' Page 3 Reference Comment
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| : 7. Page 33, Section F,1" paragraph, This statement should be removed since it is not
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| " Temporary procedure changes are germane to the Fermi 2 ITS conversion.
| |
| also controlled by 10 CFR 50.54(a)."
| |
| : 8. Page 36, Section G, bullet 3 for ITS This section should be revise to reflect the recent 3.4.1, " CTS 3.4.1.1, Recirculation - revision Fermi 2 ITS that added provisions related System... Operation". to resetting the operating limits for minimum critical power ratio and maximum average planar linear heat generation rate for single loop operation.
| |
| : 9. Page 43, Section IV, states "The For clarity, these sections (i.e., IV and V) should be licensee has been requested to revised to reflect that "... Fermi 2 was requested and submit a license condition to make did submit the requested license conj.itions..."
| |
| this commitment enforceable."
| |
| Page 43, Section V, states "In the letter of ... , the licensee proposed a license condition...".
| |
| : 10. Table A,2.0, A.4 These changes should be indicated as "not used."
| |
| Table A,3.5.2, A.13 Table A,5.7, A.4
| |
| : 11. Table L,3.3.5.2, LC.1 The table should be revised to categorize this change as "C"instead of"G".
| |
| : 12. Table L, 3.6.1.1, L.1 The SE should be revised to define the change category of" Unique" for 3.6.1.1 L.1 as stated in this table. As discussed with the NRC, a global search of the SE should be performed for similar problems following comment incorporation.
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| ATTACIIMENT 4 TO NRC-99-0079 I
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| l FERMI 2 ITS AND BASES 1
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| I L
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| Definitions
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| '1.1 1.0 USE AND APPLICATION 1.1 Definitions
| |
| ................................... NOTE- ---
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| The defined terms of this section a pear in ca italized type and are applicable throughout these Technic 1 Specific tions and Bases.
| |
| Term Definition ACTIONS ACTIONS shall be that part of a Specification that 1 prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
| |
| AVERAGE Pl.ANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE 31anar height and is equal to the sum of the (APLHGR) _HGRs for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at the height.
| |
| CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it O responds within the necessary range and accuracy to known values of the )arameter that the channel monitors. A CHANNEL CA_IBRATION shall encompacs all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST. j Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. A CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.
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| I (continued)
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| FERMI UNIT 2 1.1-1 Amendment No. 134
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| +
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| | |
| Definitions 1.1 1.1 Definitions (continued)
| |
| CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
| |
| CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel 0PERABILITY. A CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.
| |
| CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:
| |
| : a. Movement of source range monitors, local power l
| |
| range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement);
| |
| and
| |
| : b. Control rod movement, provided there are no fuel assemblies in the associated core cell.
| |
| l Suspension of CORE ALTERATIONS shall not preclude j completion of movement of a component to a safe position.
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| '- (continued)
| |
| FERMI - UNIT 2 1.1 2 Amendment No. 134
| |
| | |
| Definitions 1.1 i
| |
| ( 1.1 Definitions (continued)
| |
| CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
| |
| DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131. I-132. 1-133. I-134.
| |
| and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844 AEC.1962. " Calculation of Distance Factors for Power and Test Reactor Sites."
| |
| EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until
| |
| /~T the ECCS equipment is capable of performing its V safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
| |
| ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading l
| |
| ! delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
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| )
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| fl !
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| d (continued)
| |
| FERMI - UNIT 2 1.1-3 Amendment No. 134 e
| |
| | |
| Definitions 1.1 l -.
| |
| f)l s
| |
| 1.1 Definitions (continued)
| |
| LEAKAGE LEAKAGE shall be:
| |
| : a. Identified LEAKAGE -
| |
| : 1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank: or
| |
| : 2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the l operation of leakage detection systems or not to be pressure boundary LEAKAGE:
| |
| : b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE:
| |
| : c. Iptal LEAKAGE i
| |
| /~T Sum of the identified and unidentified V
| |
| LEAKAGE:
| |
| : d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.
| |
| LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. '
| |
| LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.
| |
| /m
| |
| (") (continued)
| |
| FERMI UNIT 2 1.1 4 Amendment No. 134 L
| |
| | |
| ~
| |
| Definitions 1.1 1.1 Definitions (continued)
| |
| MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each type of fuel. The CPR is that power in the assembly that is calculated by application of the a)propriate correlation (s) to cause some point in tie assembly to experience boiling transition, divided by the actual assembly operating power.
| |
| MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
| |
| OPERABLE -0PERABILITY A system, subsystem, division,, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety i function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that s are required for the system, subsystem, division.
| |
| component, or device to perform its specified
| |
| ~
| |
| safety function (s) are also capable of performing their related support function (s).
| |
| PHYSICS TESTS PHYSICS TESTS shall be those tests performed to ,
| |
| measure the fundamental nuclear characteristics of i the reactor core and related instrumentation.
| |
| These tests are:
| |
| : a. Described in Chapter 14. Initial Test Program I of the UFSAR: ,
| |
| : b. Authorized under the provisions of 10 CFR 50.59: or i
| |
| : c. Otherwise approved by the Nuclear Regulatory Commission.
| |
| O (continued)
| |
| FERMI UNIT 2 1.1-5 Amendment No. 134 i
| |
| L
| |
| | |
| Definitions 1.1 1.1 Definitions (continued)
| |
| (
| |
| RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3430 MWt.
| |
| REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the r.hannel sensor until de energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured, SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subtritical assuming that:
| |
| : a. The reactor is xenon free; I
| |
| : b. The. moderator temperature is 68*F: and ,
| |
| l
| |
| : c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
| |
| pJ With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.
| |
| STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillanta Frequency, so that all systems, subsyste s, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, cr other designated components in the associated function.
| |
| l 10 V (continued)
| |
| FERMI UNIT 2 1.1 6 Amendment No. 134 l
| |
| | |
| i Definitions 1.1
| |
| ( 1.1 Definitions (continued)
| |
| THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant, TURBINE BYPASS SYSTEM The TURBINE BYPASS SYSTEM RESPONSE TIME consists RESPONSE TIME of the time from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine by required positions. pass valves travel The response to their time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
| |
| O O
| |
| FERMI UNIT 2 1.1 7 Amendment No. 134
| |
| : t. ,
| |
| | |
| F .
| |
| Definitions 1.1
| |
| 'l Table 1.1-1 (page 1 of 1)
| |
| V H0 DES REACTOR MODE AVERAGE REACTOR
| |
| ! MODE TITLE SWITCH POSITION COOLANT TEMPERATURE
| |
| ! (*F) 1 Power Operation Run NA 2 Startup Refuel (a) or NA Startup/ Hot Standby 3 Hot Shutdown (a) Shutdown > 200 4 Cold Shut'Jown(a) Shutdown s 200 5 Refueling (b) Shutdown or Refuel NA (a) All reactor vessel head closure bolts fully tensioned.
| |
| (b) One or more reactor vessel head closure bolts less than fully tensioned.
| |
| l l
| |
| l l
| |
| FERMI' UNIT 2 1.1 8 Amendment No. 134
| |
| | |
| i
| |
| ! Logical Connectors 1.2 l
| |
| > y~s
| |
| ( ) 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.
| |
| l Logical connectors are used in Technical Specifications (TS) l to discriminate between, and yet connect, discrete l Conditions, Re
| |
| ! Surveillances, quired Actions, Completion and Frequencies. Times,connectors The only logical i i that appear in TS are ANQ and 2 The physical arrangement of these connectors constitutes logical conventions with specific meanings.
| |
| I BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the
| |
| ('_') number of the Required Action). The successive levels of V , logic are identified by additional digits of the Required l
| |
| Action number and by successive indentions of the logical connectors.
| |
| When logical connectors are used to state a Condition, Completicn Time, Surveillance, or Frequency, only the first
| |
| , level of logic is used, and the logical connector is left l justified with the statement of the Condition, Completion Time. Surveillance, or Frequency.
| |
| i i
| |
| f (Continued)
| |
| FERMI - UNIT 2 1.2 1 Amendment No. 134 l
| |
| 1
| |
| | |
| Logical Connectors 1.2 1.2 Logical Connectors (continued)
| |
| EXAMPLES The following examples illustrate the use of logical connectors.
| |
| EXat!PLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LC0 not met. A.1 Verify . . .
| |
| htQ A.2 Restore . . .
| |
| In this example the logical connector 8NQ is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.
| |
| O l
| |
| l i
| |
| I l
| |
| (continued) ;
| |
| FERMI - UNIT 2 1.2 2 Amendment No. 134
| |
| (
| |
| | |
| Logical Connectors 1.2 7w 1.2 Logical Connectors i]
| |
| EXAMPLES EXAMPLE 1.2-2 (continued) '
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LC0 not met. A.1 Trip . . .
| |
| 2 A.2.1 Verify . . .
| |
| AND A.2.2.1 Reduce . . .
| |
| E A.2.2.2 Perform . . .
| |
| O V A.3 Align . . .
| |
| This example represents a more complicated use of logical connectors. Required Actions A.1. A.2 and A.3 are alternative choices. only one of which must be performed as indicated by the use of the logical connector M and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND.
| |
| Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector M indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.
| |
| rN U
| |
| FERMI UNIT 2 1.2 3 Amendment No. 134
| |
| | |
| Completion Times 1.3 O
| |
| v 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
| |
| BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LC0 state Conditions that typically describe the ways in which the requirements of the LC0 can fail to be met. Specified with each stated Condition are Required Action (s) and Completion Times (s).
| |
| DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or varicble not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Q
| |
| V Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LC0 Applicability.
| |
| If situations are discovered that require entry into more than one Condition at a time within a single LC0 (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of j the situation that required entry into the Condition. !
| |
| Once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the l Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.
| |
| 1 (continued)
| |
| FERMI - UNIT 2 1.3 1 Amendment No. 134
| |
| | |
| i l
| |
| Completion Times 1.3
| |
| ( 1.3 Completion Times DESCRIPTION However, when a subseauent division, subsystem, component.
| |
| (continued) or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time (s-) may be extended. To apply this Completion Time extension. two criteria must first be met. The subsequent inoperability:
| |
| : a. Must exist concurrent with the first inoperability:
| |
| and
| |
| : b. Must remain inoperable or not within limits after the first inoperability is resolved.
| |
| The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:
| |
| j
| |
| : a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours; or
| |
| : b. The stated Completion Time as measured from discovery of the subsequent inoperability.
| |
| b The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely seaarate re entry into the Condition (for each division, su) system, component or variable expressed in the Condition) and separate tracking of Completion Times based on this re entry. These exceptions are stated in individual Specifications. j The above Completion Time extension does not apply to a Completion Time with a modified " time zero." This modified !
| |
| " time zero" may be expressed as a repetitive tina (i.e.,
| |
| "once per 8 hours." where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Example 1.3 3 illustrates one use of
| |
| , this type of Completion Time. The 10 day Completion Time i specified for Condition A and B in Example 1.3 3 may not be.
| |
| extended.
| |
| O (continued)
| |
| FERMI - UNIT 2 1.3 2 Amendment No. 134
| |
| | |
| I Completion Times 1.3
| |
| \
| |
| 1.3 Completion Times (continued) l EXAMPLES The following examples illustrate the use of Completion l Times with different types of Conditions and changing Conditions. l EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met.
| |
| l Condition B has two Required Actions. Each Required Action I has its own separate Completion Time. Each Completion Time
| |
| 'O is referenced to the time that Condition B is entered.
| |
| The Required Actions of Condition B are to be in MODE 3 within 12 hours AND in MODE 4 within 36 hours. A total of 12 hours is allowed for reaching MODE 3 and a total of l 36 hours (not 48 hours) is allowed for reaching MODE 4 from !
| |
| the time that Condition B was entered. If MODE 3 is reached within 6 hours. the time allowed for reaching MODE 4 is the next 30 hours because the total time allowed for reaching MODE 4 is 36 hours.
| |
| If Condition B is entered while in MODE 3. the time allowed for reaching MODE 4 is the next 36 hours.
| |
| l n
| |
| U (continued) :
| |
| l FERMI UNIT 2 1.3 3 Amendment No. 134 i
| |
| | |
| Completion Times 1.3 O
| |
| LJ 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued) {
| |
| ACTIONS )
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i
| |
| A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE status.
| |
| B. Required B.1 Be in MODE 3. 12 hours Action and !
| |
| associated AND Completion Time not B.2 Be in MODE 4. 36 hours met. ,
| |
| , When a pump is declared inoperable. Condition A is entered.
| |
| If the pump is not restored to OPERABLE status within 7 days. Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered. Conditions A and B are exited, and therefore. the Required Actions of Condition B may be terminated.
| |
| When a second pump is declared inoperable while the first pump is still inoperable. Condition A is not re-entered for the second pump. LC0 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump.
| |
| The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.
| |
| While in LCO 3.0.3 if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired. LCO 3.0.3 may be exited and operation continued in accordance with Condition A.
| |
| ,( p) (continued)
| |
| FERMI - UNIT 2 1.3 4 Amendment No. 134 L
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)
| |
| While in LC0 3.0.3 if one of the inoperable pumps is -
| |
| restored to OPERABLE status and the Com)letion Time for Condition A has expired, LC0 3.0.3 may )e exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.
| |
| On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour extension to the stated 7 days is allowed, provided this ;
| |
| does not result in the second pump being inoperable for
| |
| > 7 days. i i
| |
| O l
| |
| l (continued)
| |
| FERMI UNIT 2 1.3-5 Amendment No. 134
| |
| | |
| Completion Times 1.3 4 1
| |
| 0 1.3 co a'et4eo T4 es
| |
| -EXAMPLES EXAMPLE 1.3-3 -
| |
| (continued) -
| |
| 4 ACTIONS 1
| |
| COWITION REQUIRED ACTION COMPLETION TIME A. One. A.1 Restore 7 days Function X Function X subsystem subsystem to E
| |
| : inoperable. OPERABLE status. i discovery of failure to meet j the LC0 B. One ' B.1 Restore 72 hours Function Y Function Y subsystem subsystem to- E !
| |
| inoperable.' OPERABLE status.
| |
| O 10 days from discovery of failure to meet the LC0 C. One C.1 Restore 72 hours Function X Function X subsystem subsystem to inoperable. OPERABLE status.
| |
| M QB One C.2 Restore 72 hours Function Y Function Y subsystem subsystem to
| |
| . inoperable. OPERABLE status.
| |
| :e h,-. (continued)
| |
| FERMI - UNIT 21 1.3 6 Amendment No. 134
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)
| |
| When one Function X subsystem and one Function Y subsystem are ino>erable. Condition A and Condition B are concurrently applica)le. The Com)letion Times for Condition A and Condition B are tracted separately for each subsystem, starting from the time each subsystem was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second subsystem was declared inoperable (i.e., the time the situation described in Condition C was discovered).
| |
| If Required Action C.2 is completed within the specified Completion Time. Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired.
| |
| operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected subsystem was declared inoperable
| |
| -(i.e., initial entry into Condition A).
| |
| The Completion Times of Conditions A and B are modified by a
| |
| ' logical connector, with a separate 10 day Completion Time O- measured from the time it was discovered the LC0 was not met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A. B.
| |
| and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO.
| |
| The separate Completion Time modified by the phrase "from discovery of failure to meet the LC0" is designed to prevent indefinite continued operation while not meeting the LCO.
| |
| This Completion Time allows for an exception to the normal
| |
| " time zero" for beginning the Completion Time " clock". In this instance, the Completion Time " time zero" is specified as commencing at the time the LC0 was initially not met, instead of at the time the associated Condition was entered.
| |
| t (continued) 1 FERMI UNIT 2 1.3-7 Amendment No. 134 I l
| |
| | |
| Completion Times 1.3
| |
| ( 1.3 Completion Times EXAMPLES EXAMPLE 1.3 4 (continued)
| |
| ACTIONS -
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve (s) 4 hours s valves to OPERABLE inoperable. status.
| |
| B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met.
| |
| / n U A single Completion Time is used for any number of valves '
| |
| ino)erable at the same time. The Completion Time associated wit 1 Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis.
| |
| Declaring subsequent valves inoperable, while Condition A is still in effect. does not trigger the tracking of separate Completion Times.
| |
| Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The I Completion Time may be extended if the valve restored to l OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours provided this does not result in any subsequent valve being inoperable for > 4 hours.
| |
| If the Completion Time of 4 hours (plus the extension) expires while one or more valves are still inoperable.
| |
| Condition B is entered.
| |
| O)
| |
| % (continued)
| |
| FERMI - UNIT 2 1.3-8 Amendment No. 134
| |
| | |
| Completion Times 1.3 1.3 Completion Times V(3 EXAMPLES EXAMPLE 1.3 5 (continued) '
| |
| ACTIONS
| |
| .. ......................... NOTE -- --------
| |
| Se arate Condition entry is allowed for each inoperable j va ve.
| |
| ............................................................ )
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours valves OPERABLE status.
| |
| inoperable.
| |
| B. Required B.1 Be in MODE 3. 12 hours Action and associated AND O' Completion Time not B.2 Be in MODE 4. 36 hours met.
| |
| The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.
| |
| The Note allows Condition A to be entered separately for each inoperable valve. and Completion Times tracked on a per valve basis. When a valve is declared inoperable.
| |
| Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable. Condition A is (
| |
| entered for each valve and separate Completion Times start [
| |
| and are tracked for each valve. s (continued)
| |
| FERMI UNIT 2 1.3 9 Amendment No. 134
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued)
| |
| If the Completion Time associated with a valve in
| |
| * Condition A expires. Condition B is entered for that valve.
| |
| If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve.
| |
| Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times. Completion Time extensions do not apply.
| |
| O (continued)
| |
| FERMI UNIT 2 1.3 10 Amendment No. 134
| |
| | |
| Completion Times 1.3 1.3_ Completion Times EXAMPLES EXAMPLE 1.3 6 (continued)
| |
| ACTIONS -
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform Once per inoperable. SR 3.x.x.x. 8 hours
| |
| .08 A.2 Reduce THERMAL 8 hours POWER to 5 50% RTP.
| |
| B. Required B.1 Be in MODE 3. 12 hours Action and associated Completion t,_s) Time not v met.
| |
| Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" i Completion Time, which qualifies for the 25% extension, per SR 3.0.2 to each performance after the initial performance.
| |
| The initial 8 hour interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be coalplete within the first 8 hour interval. If Required Action A.1 is followed and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered.
| |
| If Required Action A.2 is followed and the Completion Time of 8 hours is not met. Condition B is entered.
| |
| If after entry into Condition B. Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.
| |
| (continued)
| |
| FERMI - UNIT 2 1.3 11 Amendment No. 134 i
| |
| l
| |
| | |
| ~
| |
| l Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3 7 (continued)
| |
| ACTIONS .
| |
| COEITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour subsystem subsystem inoperable, isolated. Ngl Once per 8 hours thereafter AND A.2 Restore subsystem 72 hours to OPERABLE status.
| |
| O V B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met.
| |
| Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.1.
| |
| If after Condition A is entered Required Action A.1 is not met within either the initial I hour or any subsequent 8 hour interval from the previous performance (plus the extension allowed by SR 3.0.2). Condition B is entered. The
| |
| ! Completion Time clock for Condition A does not stop after l Condition B is entered, but continues from the time I
| |
| Condition A was initially entered. If Required Action A.]
| |
| i (continued)
| |
| FERMI UNIT 2 1.3 12 Amendment No. 134 l
| |
| i
| |
| | |
| Completion Times 1.3 1.3- Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A.
| |
| provided the Completion Time for Required Action A.2 has not expired.
| |
| IMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
| |
| O FERMI - UNIT 2 1.3 13 Amendment No. 134
| |
| | |
| Frequency 1.4 Q
| |
| U l.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define 't proper use and application of Frequency requirements.
| |
| DESCRIPTION Each Surveillance Requirement (SR) has a specified frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified frequency is necessary for compliance with the SR.
| |
| The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.
| |
| Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated v as clarifying Notes in the Surveillance, as part of the Surveillance, or both. Example 1.4-4 discusses these special situations.
| |
| Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the 1 associated LC0 is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only " required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.
| |
| The use of " met" or " performed" in these instances conveys specific meanings., A Surveillance is " met" only when the !
| |
| acceptance criteria are satisfied. Known failure of the !
| |
| requirements of a Surveillance, even without a Surveillance specifically being " performed." constitutes a Surveillance i not " met." " Performance" refers only to the requirement to ;
| |
| specifically aetermine the ability to meet the acceptance '
| |
| L (continued)
| |
| FERMI - UNIT 2 1.4 1 Amendment No. 134
| |
| | |
| Frequency 1.4 1.4 Frequency DESCRIPTION criteria. SR 3.0.4 restrictions would not apply if both the (continued) following conditions are satisfied: ,
| |
| : a. The Surveillance is not required to be performed; and
| |
| : b. The Surveillance is not required to be met or, even if required to be met, is not known to be failed.
| |
| EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LC0 (LC0 not shown) is MODES 1. 2.
| |
| and 3.
| |
| l EXAMPLE 1.4-1 j SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| '( Perform CHANNEL CHECK. 12 hours l
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| Exam)le 1.41 contains the type of SR most often encountered in t1e Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time.
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| Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the interval specified in the Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met )er SR 3.0.1 (such as when the equipment is inoperaale, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO. and the performance of the Surveillance is not (continued)
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| FERMI UNIT 2 1.4 2 Amendment No. 134
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| Frequency 1.4 n
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| () 1.4 Frequency EXAMPLES EXAMPLE 1.4 1 (continued) otherwise modified (refer to Examples 1.4 3 and 1.4-4), then SR 3.0.3 becomes applicable.
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| !/ the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LC0 for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.
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| 1 EXAMPLE 1.4 2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within
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| /l 12 hours after ;
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| V a 25% RTP AND 24 hours thereafter Example 1.4 2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "6N_D" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to a 25% RTP. the Surveillance must be performed within 12 hours.
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| The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2.
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| ()
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| o (continued)
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| FERMI UNIT 2 1.4 3 Amendment No. 134
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| Frequency 1.4 m
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| Q 1.4 Frequency EXAMPLES EXAMPLE 1.4 2 (continued) l "Thereafter" indicates future performances must be -
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| established per SR 3.0.2. but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP. the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
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| l' EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
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| ...............N0TE -- -- - --- ---
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| Not required to be performed until 12 hours after = 25% RTP.
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| tO V
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| Perform channel adjustment. 7 days The interval continues whether or not the unit operation is
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| < 25% RTP between performances, i l
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| As the Note modifies the required performance of the l Surveillance, it is construed to be part of the "specified l Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after :
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| power reaches = 25 RTP to perform the Surveillance. The l Surveillance is still considered to be within the "specified Frequency." Themf ore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES. even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power a 25% RTP.
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| 'O Q (continued)
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| FERMI UNIT 2 1.4 4 Amendment No. 134
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| 1 Frequency 1.4 m
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| (j 1.4 Frequency ,
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| EXAMPLES EXAMPLE 1.4 3 (continued)
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| Once the unit reaches 25% RTP,12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
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| I 1
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| EXAMPLE 1.4 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
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| ..................N0TE - - ----+ ----
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| Only required to be met in MODE 1.
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| Verify leakage rates are within limits. 24 hours O
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| Example 1.4 4 specifies that the requirements of this ,
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| Surveillance do not have to be met until the unit is in i MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4 1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.
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| Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2),
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| but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the :
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| 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that th, bour Frequency were not met). SR 3.0.4 would require se.fisfying the SR.
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| F'.MI - UNIT 2 1.4 5 Amendment No. 134
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| r SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 1
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| 2.1.1 Reactor Core SLs -
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| 1 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:
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| THERMAL POWER shall be s 25% RTP.
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| 2.1.1.2 With the reactor steam dome pressure = 785 psig and core flow = 10% rated core flow:
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| MCPR shall be = 1.11 for two recirculation loop operation or = 1.13 for single recirculation loop operation.
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| 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
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| 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig.
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| U .
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| 2.2 SL Violations E
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| With any SL violation, the following actions shall be completed within 2 hours:
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| 2.2.1 Restore compliance with all SLs: and 2.2.2 Insert all insertable control rods.
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| I I
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| O FERMI - UNIT 2 2.0 1 Amendment No. 134
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| i LC0 Applicability 3.0 l
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| 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LC0 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LC0 3.0.2 and LC0 3.0.7.
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| LC0 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LC0 3.0.5 and LC0 3.0.6.
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| If the LC0 is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise stated.
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| {
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| J LC0 3.0.3 When an LC0 is not met and ch.a associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS. the unit shall be ) laced in a MODE or other specified condition in which the _C0 is not i applicable. Action shall be initiated within I hour to l A place the unit, as applicable, in: l U. {
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| : a. MODE 2 within 7 hours:
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| : b. MODE 3 within 13 hours: and
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| : c. MODE 4 within 37 hours.
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| Exceptions to this Specification are stated in the individual Specifications.
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| Where corrective measures are completed that permit operation in accordance with the LC0 or ACTIONS. completion of the actions required by LC0 3.0.3 is not required.
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| LC0 3.0.3 is only applicable in MODES 1. 2, and 3.
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| O (continued)
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| FERMI UNIT 2 3.0-1 Amendment No. 134 I
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| c LC0 Applicability 3.0 0 3.o 'C0 ^ee'ica81'1Tv <comt4"oeo) .
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| LC0 3.0.4 When an LC0 is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued
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| * operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
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| Exceptions to this-Specification are stated in the individual Specifications.
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| LC0 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2.
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| and 3.
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| LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other
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| [\ equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
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| LC0 3.0.6 When a supported system LC0 is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LC0 ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.11. " Safety Function Determination Program (SFDP)." If a loss of safety function is octermined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.
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| () (continued)
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| FERMI UNIT 2 3.0 2 Amendment No. 134
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| i l
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| LC0 Applicability )
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| 3.0 l l
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| ,-~ \
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| ' 3.0 LC0 APPLICABILITY LCO 3.0.6 When a_ support system's Required Action directs a supported (continued) system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.
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| LC0 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed to permit performance of special tests and operations. Unless
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| , otherwise specified, all other TS requirements remain unchanged. Compliance with Special 0)erations LCOs is optional. When a Special Operations C0 is desired to be met but is not met, the ACTIONS of the Special Operations LC0 shall be met. When a Special Operations LC0 is not desired to be met entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.
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| o V
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| FERMI - UNIT 2 3.0 3 Amendment No. 134
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| SR Applicability 3.0 t
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| 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs. unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.
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| Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
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| SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
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| For Frequencies specified as "once," the above interval extension does not apply.
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| C%, If a Completion Time requires >eriodic performance on a V "once per . . . " basis, the a)ove Frequency extension applies to each performance after the initial performance.
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| Exceptions to this Specification are stated in the individual Specifications.
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| 2
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| (~)
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| (continued)
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| FERMI UNIT 2 3.0 4 Amendment No. 134 1
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| I
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| SR Applicability 3.0 3.0 SR APPLICABILITY (continued) 0 SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LC0 not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.
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| If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition (s) must be entered.
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| When the Surveillance is performed within the delay period and the Surveillance is not met, the LC0 must immediately be declared not met, and the applicable Condition (s) must be entered.
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| SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be made unless the LC0's Surveillances have been met within their specified Frequency. This provision shall not prevent entry into i
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| MODES or other specified conditions in the Applicability x that are required to comply with Actions or that are part of a shutdown of the unit.
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| SR 3.0.4 is only applicable for entry into a' MODE or other specified condition in the Applicability in MODES 1. 2. and 3.
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| c, O
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| b FERMI UNIT 2 3.0 5 Amendment No. 134
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| o
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| .;; . SDM 3.1.1 l 3.1 REACTIVITY CONTROL SYSTEMS-3.1.1- SHUTDOWN MARGIN (SDM)
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| -LC0 3.1.1- SDM shall be: -
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| : a. = 0.38% Ak/k, with the highest worth control rod analytically determined; or
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| : b. = 0.28% Ak/k. with the highest worth control rod determined by test.
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| -APPLICABILITY:- MODES 1, 2, 3, 4, and 5.
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| }
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| ACTIONS-C0lOITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits A.1 Restore SDM to within 6 hours in MODE 1 or 2. limits.
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| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion
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| ~ Time.of_ Condition AL not met.
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| C. SDM not within limits C.1 Initiate action to Immediately
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| , in MODE 3. fully insert all insertable control rods.
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| (continued)
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| )
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| O i
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| ~
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| FERMI UNIT 2- 3.1-1 Amendment No. 134
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| SDM 3.1.1 ACTIONS (continued)
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| CONDITION COMPLETION TIME REQUIRED ACTION D. SDM not within limits D.1 Initiate action to Immediately-in MODE 4. fully insert all insertable control rods.
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| AND D.2 Initiate action to 1 hour restore secondary containment to OPERABLE status.
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| AND D.3 Initiate action to 1 hour '
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| restore one standby gas treatment (SGT) subsystem to OPERABLE l
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| status.
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| O AND V
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| D.4 Initiate action to I hour restore isolation l capability in each required secondary !
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| containment penetration flow path not isolated.
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| L I
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| (continued)
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| O FERMI UNIT 2 3.1 2 Amendment No. 134 i
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| SDM 3.1.1 i
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| i ACTIONS (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME E. SDM not within limits E.1 Suspend CORE Immediately-in MODE 5. ALTERATIONS except for control rod l insertion and fuel assembly removal.
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| E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies.
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| AND E.3 Initiate action to 1 hour restore secondary containment to
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| (
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| l(3.)
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| M
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| -OPERABLE status.
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| E.4 Initiate action to I hour restore one SGT subsystem to OPERABLE '
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| status.
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| AND E.5 Initiate action to I hour restore isolation capability in each required secondary containment penetration flow path ;
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| not isolated.
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| ~
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| l fx FERMI - UNIT 2 3.1 3 Amendment No. 134 l
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| f l
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| l SDM 3.1.1 l O suavet'teace aeouineasu1s SURVEILLANCE FREQUENCY
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| !- 'SR 3.1.1.1 Verify SDM is: Prior to each in vessel fuel
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| : a. = 0.38% Ak/k with the highest worth movement during control rod analytically determined; fuel loading or sequence
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| : b. = 0.28t Ak/k with the highest worth ANQ control rod determined by test.
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| Once within 4 hours after criticality following fuel movement within the reactor pressure vessel O
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| O FERMI - UNIT 2 3.1 4 Amendment No. 134 l
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| l i
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| Reactivity Anomalies !
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| 3.1.2 O'3.1REACTIVITYCONTROLSYSTEMS 3.1.2 Reactivity Anomalies LCO 3.1.2 The reactivity difference between the monitored reactivity and the predicted reactivity shall be within i lt Ak/k.
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| APPLICABILITY: MODES 1 and 2.
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| _ ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. Core reactivity A.1 Restore core 72 hours difference not within reactivity difference limit, to within limit.
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| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
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| i l
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| FERMI - UNIT 2 3.1 5 Amendment No. 134
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| Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity difference between Once withiri the monitored reactivity and the predicted 24 hours after reactivity is within i 15 Ak/k. reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel AND 1000 MWD /ST thereafter during operations in MODE 1 O
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| O FERMI UNIT 2 3.1 6 Amendment No. 134
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| Control Rod OPERABILITY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3- Control Rod 0PERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.
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| APPLICABILITY: MODES 1 and 2.
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| ACTIONS
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| ..................................... NOTE--- - --- - - --------- - - - --- -
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| Separate Condition entry is allowed for each control rod.
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| CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control -----
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| - - NOTE -- --- - -
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| rod stuck, Rod worth minimizer (RWM) may be bypassed as allowed by N'
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| LC0 3.3.2.1, " Control Rod
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| ' Block Instrumentation." if required, to allow continued operation.
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| A.1 Verify stuck control Immediately rod separation criteria are met.
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| 8NQ A.2 Disarm the associated 2 hours control rod drive (CRD).
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| AND (continued) m U
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| FERMI UNIT 2 3.1-7 Amendment No. 134
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| 1 Control Rod OPERABILITY 3.1.3 pd ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i'
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| A. (continued) A.3 Perform SR 3.1.3.2 24 hours from and SR 3.1.3.3 for discovery of each withdrawn Condition A l OPERABLE control rod, concurrent with (
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| THERMAL POWER greater than the low power setpoint (LPSP) l of the RWM j 8NQ A.4 Perform SR 3.1.1.1. 72 hours B. Two or more withdrawn B.1 Be in MODE 3. 12 hours control rods stuck.
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| O C. One or more control ------
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| - NOTE - - - -- -
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| rods inoperable for RWM may be bypassed as reasons other than allowed by LC0 3.3.2.1. if Condition A or B. required. to allow insertion of inoperable control rod and continued operation.
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| C.1 Fully insert 3 hours inoperable control i rod.
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| AND C.2 Disarm the associated 4 hours CRD. j i
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| l (continued)
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| I O
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| FERMI UNIT 2 3.1 8 Amendment No. 134
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| Control Rod OPERABILITY 3.1.3 I ACTIONS (continued)
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| COMPLETION TIME CONDITION REQUIRED ACTION D. -
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| - NOTE .......- D.1 Restore compliance 4 hours .
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| Not ap)licable when with the prescribed THERMA. POWER withdrawal sequence.
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| > 10% RTP.
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| ...................... g One or more inoperable D.2 Restore control rod 4 hours control rods not in to OPERABLE status.
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| compliance with the prescribed withdrawal sequence.
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| E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A.
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| C. or D not met.
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| /~')
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| k/ Nine or more control rods inoperable.
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| l l
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| I I
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| )
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| ,)
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| i t
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| FERMI UNIT 2 3.1 9 Amendment No. 134 i
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| I l
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| Control Rod OPERABILITY 3.1.3 Il y/
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| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod. 24 hours SR 3.1.3.2 - - - -- - NOTE - - -- - -- --
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| Not required to be performed until 7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM.
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| Insert each fully withdrawn control rod at 7 days least one notch.
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| SR 3.1.3.3 - - ----- -
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| NOTE -- - - -- --
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| Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of t
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| c the RWM.
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| 's ...........................................
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| Insert each partially withdrawn control rod 31 days at least one notch.
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| SR 3.1.3.4 Verify each control rod scram time from In accordance fully withdrawn to notch position 06 is with s 7 seconds. SR 3.1.4.1.
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| SR 3.1.4.2.
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| SR 3.1.4.3. and SR 3.1.4.4 (continued)
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| FERMI UNIT 2 3.1 10 Amendment No. 134
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| Control Rod OPERABILITY 3.1.3 O' SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY I 5R 3.1.3.5 Verify each control rod does not go to the Each time the withdrawn overtravel position. control rod is withdrawn to
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| " full out" position AND Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling O
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| i l
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| i I
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| 1 I
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| i O l I
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| FERMI UNIT 2 3.1 11 Amendment No. 134
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| Control Rod Scram Times l 3.1.4 i 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 13 OPERABLE control rods shall be "sidw "
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| in accordance with Table 3.1.4 1: and
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| : b. No more than 2 OPERABLE control rods that are " slow" shall occupy adjacent locations.
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| APPLICABILITY: MODES 1 and 2.
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours LC0 not met.
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| (3 LJ SURVEILLANCE REQUIREMENTS
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| ..................................... NOTE --- - -- ----- ---- - -- --- -
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| During single control rod scram time Surveillances. the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator.
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| SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify each control rod scram time is Prior to within the limits of Table 3.1.41 with exceeding reactor steam dome pressure a 800 psig. 40% RTP after each reactor shutdown
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| > 120 days (continued) )
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| FERMI UNIT 2 3.1 12 Amendment No. 134
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| Control Rod Scram Times 3.1.4 SURVEILLANCE REWIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify. for a representative sample, each 120 days -
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| tested control rod scram time is within the cumulative limits of Table 3.1.41 with reactor steam - operation in dome pressure = 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor-steam dome pressure. control rod OPERABLE after work on control rod or CRD System that could affect scram time 3.1.4.4 Verify each affected control rod scram time Prior to O ' SR is within the limits of Table 3.1.4-1 with reactor steam dome pressure = 800 psig.
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| exceeding 40% RTP after fuel movement within the associated core l cell l AND Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time O
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| FERMI UNIT 2 3.1 13 Amendment No. 134 t
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| Control Rod Scram Times 3.1.4 h" Table 3.1.41 (page 1 of 1)
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| Control Rod Scram Times
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| .....................................yoyg5....................................
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| : 1. OPERABLE control rods with scram times not within the limits of this Table are considered " slow." -
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| : 2. Enter applicable Conditions and Required Actions of LCO 3.1.3. " Control Rod OPERABILITY." for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.4. and are not considered " slow.
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| SCRAMTIMESwhenREACTORSTEApag NOTCH POSITION PRESSURE = 800 psig (seconds) 46 0.457 36 1.084 26 1.841 O 06 V 3.361 (a) Maximum scram time from fully withdrawn position, based on de.energization of scram pilot valve solenoids at time zero.
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| (b) When reactor steam dome pressure is < 800 psig established scram time limits apply.
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| FERMI . UNIT 2 3.1 14 Amendment No. 134 1 I
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| l
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| Control Rod Scram Accumulators i 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5' Each control' rod scram accumulator shall be OPERABLE.
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| APPLICABILITY: MODES 1 and 2.
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| ACTIONS
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| ..................................... NOTE- - <-- ------ - - -- - --- --- - --
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| Separate Condition entry is allowed for each control. rod scram accumulator.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One control rod scram A.1 -- .- --NOTE.-- --- -
| |
| accumulator inoperable Only applicable if with reactor steam the associated O dome pressure
| |
| = 900 psig.
| |
| control rod scram time was within the limits of Table 3.1.4 1 during the last scram time Surveillance.
| |
| Declare the 8 hours associated control rod scram time
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| " slow."
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| A.2 Declare the 8 hours associated control
| |
| -rod inoperable.
| |
| )
| |
| (continued)
| |
| O FERMI UNIT 2 3.1 15 Amendment No. 134 l
| |
| )
| |
| | |
| Control Rod Scram Accumulators 3.1.5
| |
| )
| |
| ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| B. Two or more control B.1 Restore charging 20 minutes from rod scram accumulators water header pressure discovery of inoperable with to a 940 psig. Condition B reactor steam dome concurrent with pressure = 900 psig. charging water header pressure
| |
| < 940 psig 8h!Q B.2.1 ------- NOTE --------
| |
| Only applicable if the associated control rod scram time was within the limits of Table 3.1.4 1 during the last scram time Surveillance.
| |
| Declare the 1 hour associated control ,
| |
| rod scram time l
| |
| " slow." j 1
| |
| B.2.2 Declare the I hour associated control rod inoperable.
| |
| (continued)
| |
| O l 1
| |
| FERMI - UNIT 2 3.1 16 Amendment No. 134
| |
| | |
| I Control Rod Scram Accumulators 3.1.5 i
| |
| j ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. One or more control C.1 Verify all control Immediately'upon
| |
| , rod scram accumulators rods associated with discovery of
| |
| ! inoperable with inoperable charging water reactor steam dome accumulators are header pressure l pressure < 900 psig. fully inserted. < 940 psig AND C.2 Declare the 1 hour associated control rod inoperable.
| |
| D. Required Action and D.1 ---
| |
| NOTE ---- -
| |
| associated Completion Not applicable if all Time of Required inoperable control Action B.1 or C.1 not rod scram met. accumulators are associated with fully
| |
| [] inserted control rods.
| |
| Place the reactor Immediately mode switch in the shutdown position.
| |
| SURVEILLANCE REQUIREMENTS I SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each control rod scram accumulator 7 days pressure is a 940 psig.
| |
| i O
| |
| FERMI - UNIT 2 3.1 17 Amendment No. 134 l
| |
| l
| |
| | |
| r-l Rod Pattern Control 3.1.6 i
| |
| l h? 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control I
| |
| : LC0 3.1.6 OPERABLE control rods shall comply with the requirements of I the prescribed withdrawal sequence.
| |
| APPLICABILITY: MODES 1 and 2 with THERMAL POWER s 10% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OPERABLE A.1 ---
| |
| ---NOTE---------
| |
| control rods not in Rod worth minimizer compliance with the (RWM) may be bypassed prescribed withdrawal as allowed by sequence. LC0 3.3.2.1, " Control Rod Block n Instrumentation."
| |
| Move associated 8 hours control rod (s) to correct position.
| |
| E A.2 Declare associated 8 hours control rod (s) inoperable.
| |
| (continued) o ;
| |
| . FERMI UNIT 2 3.1-18 Amendment No. 134 l
| |
| l L
| |
| | |
| i Rod Pattern Control j 3.1.6 l l 't ACTIONS (continued) I R,j' :
| |
| CONDITION REQUIRED ACTION COMPLETION TIME I
| |
| B. Nine or more OPERABLE B.1 -
| |
| --- -NOTE-- ---
| |
| l control rods not in Rod worth minimizer '
| |
| compliance with the (RWM) may be bypassed prescribed withdrawal as allowed by s sequence. LC0 3.3.2.1.
| |
| Ruspend withdrawal of Immediately control rods.
| |
| 1 AND B.2 Place the reactor 1 hour mode switch in the shutdown position.
| |
| (~) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply 24 hours with the prescribed withdrawal sequence.
| |
| ()g
| |
| /
| |
| FERMI - UNIT 2 3.1-19 Amendment No. 134
| |
| | |
| SLC System 3.1.7
| |
| (~J v
| |
| 3.1 . REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LC0 3.1.7 Two SLC subsystems shall be OPERABLE.
| |
| ~
| |
| APPLICABILITY: MODES I and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable. to OPERABLE status.
| |
| B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable. subsystem to OPERABLE
| |
| .j status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. .
| |
| I
| |
| )
| |
| O FERMI - UNIT 2 3.1 20 Amendment No. 134
| |
| | |
| SLC System 3.1.7 O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours pentaborate solution is within the limits of Figure 3.~ .7-1.
| |
| SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is = 48'F.
| |
| I SR 3.1.7.3 Verify temperature of pump suction piping 24 hours is = 48'F.
| |
| SR 3.1.7.4 Verify continuity of explosive charge. 31 days O SR 3.1.7.5 Verify the concentration of boron in 31 days solution is within the limits of Figure 3.1.7 1. AND Once within 24 hours after water or boron is added to solution AND Once within 24 hours after solution temperature is restored a 48*F (continued) !
| |
| FERMI - UNIT 2 3.1-21 Amendment No. 134
| |
| | |
| I SLC System 3.1.7 O suavetu^~ce acouineaeaTs <coet4""ed)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each SLC subsystem manual valve in 31 days '
| |
| the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
| |
| SR 3.1.7.7 Verify each pump develops a flow rate In accordance
| |
| = 41.2 gpm at a discharge pressure with the
| |
| = 1215 psig. Inservice Testing Program l
| |
| SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS O SR 3.1.7.9 Verify all piping between storage tank and 18 months explosive valve is unblocked.
| |
| AND Once within 24 hours after solution temperature is restored e 48 F SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to a 65 atom percent B-10. addition to SLC tank
| |
| _O l l
| |
| FERMI UNIT 2 3.1-22 Amendment No. 134 I
| |
| | |
| O 15 -
| |
| I O
| |
| i h a
| |
| is s
| |
| II*E!
| |
| g i
| |
| El ws
| |
| =
| |
| 5 x
| |
| 8 $
| |
| u a
| |
| lN - l- $
| |
| lt 1
| |
| I_
| |
| I$E . t B[' -
| |
| /
| |
| l5 y<
| |
| ~
| |
| )
| |
| O' -
| |
| ei!
| |
| IlB z~ -
| |
| * ec
| |
| >w Ell
| |
| ~
| |
| 'N i 5
| |
| ~
| |
| / ll-l gr e
| |
| 3 E3 8E L*
| |
| IE
| |
| ~ g8 g$
| |
| I ;
| |
| I.*g I I meiam As NOuMMIDNOD EWOWm3d nocOS MEDWEd FERMI UNIT 2 3.1 23 Amenament No. 134
| |
| | |
| SDV Vent and Drain Valves 3.1.8 h 3.1. REACTIVITY CONTROL SYSTEMS 3.1.8 . Scram Discharge Volume (SDV) Vent and Drain Valves LC0 3.1.8 Each SDV vent and drain valve shall be OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS
| |
| ..................................... NOTE - ---- -- --------- ---- ----- -
| |
| Separate Condition entry is allowed for each SDV vent and drain line.
| |
| 1 CONDITION REQUIRED ACTION COMPLETION TIME
| |
| ' A. One or more SDV vent A.1 Restore valve to 7 days or drain lines with OPERABLE status.
| |
| one valve inoperable.
| |
| O B. One or more SDV vent B.1 -
| |
| ----NOTE---- - -
| |
| or drain lines with An isolated line may both valves be unisolated under inoperable. administrative control to allow draining and venting of the SDV.
| |
| Isolate the 8 hours associated line.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met..
| |
| i i
| |
| FERMI - UNIT 2 3.1 24 Amendment No. 134
| |
| | |
| SDV Vent and Drain Valves 3.1.8 o
| |
| SURVEILLANCE REQUIREMENTS
| |
| !]
| |
| SURVEILLANCE FREQUENCY SR 3.1.8.1 ---NOTE-- ---- -- --
| |
| Not required to be met on vent and drain valves closed intermittently for testing under administrative control.
| |
| Verify each SDV vent and drain valve is 31 days open.
| |
| SR 3.1.8.2 Verify each SDV vent and drain valve: 18 months
| |
| : a. Closes in s 30 seconds after receipt of an actual or simulated scram l signal: and !
| |
| : b. Opens when the actual or simulated scram signal is reset.
| |
| O L,J i FERMI - UNIT 2 3.1 25 Amendment No. 134
| |
| | |
| APLHGR 3.2.1
| |
| ( 3.2 POWER DISTRIBLITION LIMITS 3.2.1 AVERAGE PLANfR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| LC0 3.2.1 All APLHGRs shall be less than or equal to the limits '
| |
| specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER = 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours limits. within limits. ;
| |
| i I
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours o associated Completion to < 25% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal Once within to the limits specified in the COLR. 12 hours after a 25% RTP AND 24 hours thereafter O
| |
| FERMI - UNIT 2 3.2-1 Amendment No. 134
| |
| | |
| MCPR ;
| |
| 3.2.2
| |
| )
| |
| O 3.2 eowea ois>a18u11oa 'intTs !
| |
| 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| LC0 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified in the COLR.
| |
| APPLICABILITY: THERMAL POWER = 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within A.1 Restore MCPR(s) to 2 hours !
| |
| limits. within limits.
| |
| i B. Required Action and B.1 Reduce THERMAL POWER 4 hours p associated Completion to < 25% RTP.
| |
| V Time not met. j l
| |
| I
| |
| - \_./
| |
| FERMI UNIT 2 3.2 2 Amendment No. 134
| |
| | |
| i- . 1 MCPR 3.2.2 L
| |
| SURVEILLANCE REQUIREMENTS
| |
| . SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal Once within to the limits specified in the COLR. 12 hours after l
| |
| = 251 RTP AND 24 hours thereafter SR 3.2.2.2 Determine the MCPR limits. Once within 72 hours after each completion of SR 3.1.4.1 AND Once within i 72 hours after O- each completion of SR 3.1.4.2 8NQ Once within 72 hours after each completion of SR 3.1.4.4 O
| |
| FERMI . UNIT 2 3.2 3 Amendment No. 134 l l
| |
| | |
| K 3 ..
| |
| LHGR 3.2.3 h 3.2. POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| :LCO 3.2.3- - All LHGRs shall be less than or equal to the limits
| |
| * specified 'in the COLR.
| |
| APPLICABILITY: : THERMAL POWER = 25% RTP. I ACTIONS.
| |
| -CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to 2 hours limitst within limits.
| |
| 1 B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to'< 25% RTP.
| |
| 'O Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| .SR. 3.2.3.1- . Verify all LHGRs are less than or equal to Once within the limits specified in the COLR. 12 hours after
| |
| = 25% RTP
| |
| , ANQ 24 hours thereafter 1
| |
| 0 i
| |
| FERMI UNIT.2- 3.2 4 Amendment No. 134 l
| |
| | |
| RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LC0 3.3.1.1- The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.1.1 1.
| |
| ACTIONS
| |
| .....................................NTE----------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours channels inoperable. trip.
| |
| gg A.2 ---------NOTE--------
| |
| Not applicable for Functions 2.a. 2.b.
| |
| 2.c. and 2.d.
| |
| Place associated trip 12 hours system in trip.
| |
| (continued)
| |
| A V
| |
| l FERMI UNIT 2 3.3 1 Amendment No. 134
| |
| | |
| RPS Instrumentation 3.3.1.1
| |
| 'O actioas <ce#14##eo)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. .
| |
| .......N0TE------ -
| |
| B.1 P1 ace channel in one 6 hours -
| |
| -Not applicable for trip system in trip.
| |
| Functions 2.a. 2.b.
| |
| 2.c, and 2.d. IE One or more Functions in trip.
| |
| with one or more required channels inoperable in.both trip systems.
| |
| C. 'One or more Functions C.1 Restore RPS trip 1 hour with RPS trip capability.
| |
| capability not maintained.
| |
| O D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A. Table 3.3.1.1-1 for B, or C not met. the channel.
| |
| E. As required by E.1 Reduce THERMAL POWER 4 hours Required Action D.1 to < 30% RTP.
| |
| and referenced in Table 3.3.1.1 1.
| |
| F. As required by F.1 Be in MODE 2. 6 hours Required Action D.1 and referenced in
| |
| . Table 3.3.1.1-1.
| |
| (continued)
| |
| O FERMI - UNIT 2 3.3 2 Amendment No. 134
| |
| | |
| RPS Instrumentation 3.3.1.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Be in MODE 3. 12 hours -
| |
| Required Action D.1 and referenced in Table 3.3.1.1-1.
| |
| H. As required by H.1 Isolate all main 12 hours Recuired Action D.1 steam lines.
| |
| anc referenced in Table 3.3.1.1-1. @
| |
| H.2 Be in MODE 3. 12 hours I. As required by I.1 Initiate action to Immediately Required Action D.1 fully insert all 3 and referenced in insertable control Table 3.3.1.1-1. rods in core cells
| |
| ] containing one or more fuel assemblies.
| |
| O FERMI UNIT 2 3.3 3 Amendment No. 134 i
| |
| | |
| RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS
| |
| :{m')
| |
| ..................................... NOTES-- ---- -- ------------------- -
| |
| l
| |
| : 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours SR 3.3.1.1.3 ------------------NOTE ----- ----- ------
| |
| (A)
| |
| Not required to be performed until 12 hours after THERMAL POWER = 25% RTP.
| |
| Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power is s 2% RTP, while operating at = 25% RTP.
| |
| SR 3.3.1.1.4 ----- ----- -
| |
| NOTE ----------- ---
| |
| Not required to be performed when entering MODE 2 from MODE 1 until ;
| |
| 12 hours after entering MODE 2. '
| |
| Perform CHANNEL FUNCTIONAL TEST. 7 days i SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days i
| |
| I (continued) g
| |
| .x FERMI - UNIT 2 3.3 4 Amendment No. 134
| |
| | |
| . r RPS Instrumentation 3.3.1.1 O suavetu^ ace acou'aeae"'s (ce"t4"#ed)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to fully intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the core SR 3.3.1.1.7 ------- ------- --NOTE --------- -- ---
| |
| Only re uired to be met during entry into MODE 2 rom MODE 1.
| |
| Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 MWD /T average core exposure O
| |
| SR 3.3.1.1.9- Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1;10 Verify the trip unit setpoint. 92 days
| |
| -SR 3.3.1.1.11 - - -- - - --
| |
| -NOTES -- -- - --
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. . For Function 1.a not required to be performed when entering MODE 2 from ;
| |
| MODE 1 until 12 hours after entering '
| |
| MODE 2.
| |
| Perform CHANNEL CALIBRATION. 184 days .
| |
| l (continued)
| |
| O l FERMI' UNIT 2 3.3 5 Amendment No. 134
| |
| | |
| RPS Instrumentation 3.3.1.1 h SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.12 - ---- ------ - NOTE - --- -- - - -- -
| |
| For Function 2.a. not required to be
| |
| :Herformed when entering MODE 2 from
| |
| ,100E 1 until 12 hours after entering MODE 2.
| |
| Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 18 months 1
| |
| SR 3.3.1.1.14 Perform CHANNEL CALIBRATION. 18 months
| |
| ( .SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and 18 months Turbine Control Valve Fast Closure Functions are not bypassed when THERMAL POWER is a 30t RTP.
| |
| 1 (continued) 1 FERMI UNIT 2 3.3-6 Amendment No. 134 l
| |
| | |
| i i
| |
| RPS Instrumentation 3.3.1.1
| |
| '( SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.1.1.17 --
| |
| ----- ---- ---NOTES --- --- -- - -
| |
| )
| |
| : 1. Neutron detectors are excluded.
| |
| : 2. For Functions 3 and 4 channel sensor response times are not required to be measured.
| |
| : 3. For Function 5 "n" equals 4 channels for the purpose of determining the {
| |
| STAGGERED TEST BASIS Frequency. j Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS SR 3.3.1.1.18 - ------ ----
| |
| ---NOTE-------------------
| |
| Neutron detectors are excluded.
| |
| Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.19 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months
| |
| ('
| |
| (_)s FERMI UNIT 2 3.3 7 Amendment No. 134
| |
| | |
| RPS Instrumentation 3.3.1.1 A)/
| |
| r\, Table 3.3.1.1 1 (page 1 of 3)
| |
| Reactor Protection System Instrtmentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM .
| |
| l SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREENTS VALUE
| |
| : 1. Intermediate Range Monitors
| |
| : a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 s 122/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.6 full scale SR 3.3.1.1.7 SR 3.3.1.1.11 J SR 3.3.1.1.15 5(a) 3 I SR 3.3.1.1.1 s 122/125 i SR 3.3.1.1.5 divisions of i SR 3.3.1.1.11 full scale i SR 3.3.1.1.15 l
| |
| : b. Inop 2 3 G SR 3.3.1.1.4 NA SR 3.3.1.1.15 5(a) 3 I SR 3.3.1.1.5 NA !
| |
| SR 3.3.1.1.15 1
| |
| : 2. Average Power Range
| |
| . Monitors
| |
| : a. Neutron Flux - Upscale 2 3(C) G SR 3.3.1.1.2 s 20% RTP (g (Setdown) SR 3.3.1.1.7 t 4 SR 3.3.1.1.8 V .
| |
| SR 3.3.1.1.12 SR 3.3.1.1.18
| |
| : b. Simulated Thermal 1 3(c) F SR 3.3.1.1.2 s 0.63 (W aW)
| |
| Power - Upscale SR 3.3.1.1.3 + 64.31 RTP SR 3.3.1.1.8 and s SR 3.3.1.1.12 RTP(b)115.51 SR 3.3.1.1.18 (continued)
| |
| (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. l (b) 4W = 83 when reset for single loop operation per LCO 3.4.1, " Recirculation Loops Operating."
| |
| Otherwise 4W = 08.
| |
| (c) Each APRM channel provides inputs to both trip systems.
| |
| O V
| |
| FERMI UNIT 2 3.3 8 Amendment No. 134 I
| |
| | |
| !I RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 2 of 3)
| |
| Reactor Protection System Instrtmentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED StRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE I
| |
| : 2. Average Power Range i Monitors (continued)
| |
| : c. Neutron 1 3(C) F SR 3.3.1.1.2 s 1201 RTP Flux - Upscale SR 3.3.1.1.3 SR 3.3.1.1.8 SR 3.3.1.1.12 SR 3.3.1.1.18 i
| |
| : d. Inc3 1.2 3(C) G SR 3.3.1.1.12 NA
| |
| : e. 2-out of 4 Voter 1.2 2 G SR 3.3.1.1.2 NA SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.19
| |
| : 3. Reactor Vessel Steam 1.2 2 G SR 3.3.1.1.1 s 1113 psig Dome Pressure - High SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17
| |
| : 4. Reactor Vessel Water 1.2 2 G SR 3.3.1.1.1 = 171.9 inches A Level - Low. Level 3 SR 3.3.1.1.9 l
| |
| (/ ) SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17
| |
| : 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 122 closed Valve - Closure SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17
| |
| : 6. Main Steam Line 1.2 2 H SR 3.3.1.1.1 s 3.6 X full Radiation - High SR 3.3.1.1.9 power SR 3.3.1.1.14 background SR 3.3.1.1.15
| |
| : 7. Drywell Pressure - High 1.2 2 G SR 3.3.1.1.1 s 1.88 psig SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 (continued)
| |
| (c) Each APRM channel provides inputs to both trip systems, i
| |
| I O
| |
| FERMI - UNIT 2 3.3-9 Amendment No. 134
| |
| | |
| RPS Instrumentation 3.3.1.1 Table 3.3.1.11 (page 3 of 3)
| |
| Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM .
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| SPECIFIED PER TRIP REQUIRED StRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREENTS VALUE
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| : 8. Scram Discharge Volume Water Level-High
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| : a. Level 1.2 2 G SR 3.3.1.1.1 s 596 ft.
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| Transmitter SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.1 s 596 ft.
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| SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15
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| : b. Float Switch 1.2 2 G SR 3.3.1.1.9 s 596 ft.
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| SR 3.3.1.1.14 0 inches SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.9 s 596 ft.
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| SR 3.3.1.1.14 0 inches SR 3.3.1.1.15
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| : 9. Turbine Stop a 301 RTP 4 E SR 3.3.1.1.9 s 71 closed Valve - Closure SR 3.3.1.1.14 f.- SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17
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| : 10. Turbine Control Valve a 301 RTP 2 E SR 3.3.1.1.9 NA Fast Closure SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17
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| : 11. Reactor Mode Switch- 1.2 2 G SR 3.3.4.1.13 NA Shutdown Position SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.13 NA SR 3.3.1.1.15
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| : 12. Manual Scram 1.2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 5(a) 2 1 SR 3.3.1.1.5 NA SR 3.3.1.1.15
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| .(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
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| [
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| \
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| FERMI - UNIT 2 3.3 10 Amendment No. 134
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| SRM Instrumentation 3.3.1.2 0 3.3 INSTRUMENT ^TIO" 3.3.1.2 Source Range Monitor (SRM) Instrumentation 1
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| LCO 3.3.1.2 The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.
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| i APPLICABILITY: According to Table 3.3.1.2 1.
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
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| A. One or more required A.1 Restore required SRMs 4 hours SRMs inoperable in to OPERABLE status.
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| MODE 2 with i intermediate range '
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| monitors (IRMs) on Range 2 or below, p
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| b B. Three required SRMs B.1 Sus)end control rod Immediately inoperable in MODE 2 wit 1drawal.
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| with IRMs on Range 2 or below.
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| i l
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| C. Required Action and C.1 Be in MODE 3. 12 hours i associated Completion i Time of Condition A or B not met.
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| (continued) i
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| .p, (j-FERMI UNIT 2 3.3 11 Amendment No. 134 L
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| SRM Instrumentation 3.3.1.2 ACTIONS (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME D. One or more required D.1 Fully insert all I hour SRMs inoperable in insertable control MODE 3 or 4. rods.
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| AND D.2 Place reactor mode I hour switch in the shutdown position.
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| E. One or more required E.1 Suspend CORE Immediately SRMs inoperable in ALTERATIONS except MODE 5. for control rod insertion.
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| ANQ E.2 Initiate action to Immediately fully insert all O~ insertable control rods in core cells containing one or more fuel assemblies.
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| O FERMI UNIT 2 3.3 12 Amendment No. 134
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| SRM Instrumentation 3.3.1.2 O suaveit'aace atouiarac Ts
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| ...................................-. NOTE .....--- --- - .- ------ -- -
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| Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.
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| SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.2.2 --- --- -- -
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| - NOTES -- - - - ---
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| : 1. Only required to be met during CORE l ALTERATIONS.
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| : 2. One SRM may be used to satisfy more than one of the following.
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| Verify an OPERABLE SRM detector is 12 hours located in:
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| : a. The fueled region;
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| : b. The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region: and
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| : c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.
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| SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours (continued)
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| O FERMI UNIT 2 3.3 13 Amendment No. 134
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| . n. _ - - _ - - _ _
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| 1 SRM Instrumentation 3.3.1.2
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| / SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY l
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| SR 3.3.1.2.4 -- -- ----
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| ------NOTE- -- - .--. -- -.-- l Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
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| Verify count rate is: 12 hours during CORE
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| : a. = 3.0 cps: or ALTERATIONS
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| : b. = 0.7 cps when signal-to noise ratio AND is = 20:1.
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| 24 hours SR 3.3.1.2.5 ------- -- -
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| --NOTE- - ------ - - ---
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| Signal-to noise ratio not required to be determined when SRM count rate is O = 3.0 cps Perform CHANNEL FUNCTIONAL TEST and 7 days determination of signal-to noise ratio.
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| SR 3.3.1.2.6 ---- - - - -
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| --NOTES -- - - -- - ---
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| : 1. Signal-to noise ratio not required to be determined when SRM count rate is
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| = 3.0 cps
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| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
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| Perform CHANNEL FUNCTIONAL TEST and 31 days determination of signal to-noise ratio.
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| (continued)
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| O FERMI UNIT 2 3.3-14 Amendment No. 134
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| SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.3.1.2.7 - - ---
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| ---- - NOTES - -- - - ------- '
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| : 1. Neutron detectors are excluded.
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| : 2. Not required to be performed until 12 hours after IRMs on Range 2 or below.
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| Perform CHANNEL CALIBRATION. 18 months O
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| FERMI UNIT 2 3.3 15 Amendment No. 134
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| SRM Instrumentation 3.3.1.2 Table 3.3.1.2 1 ( ge 1 of 1)
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| / Source Range Monitor Instrumentation APPLICABLE MODES OR OTER REQUIRED SLRVEILLANCE FUNCTION SPECIFIED CONDITIONS CMNNELS REQUIREMENTS
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| : 1. Source Range Monitor 2(a) 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 3,4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1,2.7 5 2(b)(c) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 (a) With IRMs on Range 2 or below.
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| (b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled region includes only that SRM detector.
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| (c) Special movable detectors may be used in place of SRMs if connected to normal SRM circuits.
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| l l
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| t I
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| l i
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| l
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| (
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| FERMI UNIT 2 3.3-16 Amendment No. 134
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| ControlRodBlockInstrum$tation 3.3.2.1 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation
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| -LC0 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.
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| APPLICABILITY: According to Table 3.3.2.11.
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel 24 hours (RBM) channel to OPERABLE status, inoperable.
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| B. Required Action and B.1 Place one RBM channel I hour
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| *n associated Completion in trip.
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| Time of Condition A not met.
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| M Two RBM channels inoperable.
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| l C. Rod worth minimizer C.1 Suspend control rod Immediately (RWM) inoperable movement except by during reactor scram.
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| startup.
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| 2 (continued)
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| (
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| , V) !
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| FERMI - UNIT 2 3.3-17 Amendment No. 134
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| Control Rod Block Instrumentation 3.3.2.1 ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1.1 Verify = 12 rods Immediately-withdrawn.
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| .QB C.2.1.2 Verify by Immediately administrative methods that startup with RWM inoperable has not been performed in the current calendar year.
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| AND C.2.2 Verify movement of During control control rods is in rod movement compliance with the prescribed withdrawal sequence by a second O. licensed operator or other qualified member of the technical staff.
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| D. RWM inoperable during D.1 Verify movement of During control reactor shutdown. control rods is in rod movement accordance with the prescribed withdrawal sequence by a second licensed operator or other qualified member of the technical staff.
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| (continued)
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| O FERMI UNIT 2 3.3 18 Amendment No. 134
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| Control Rod Block Instrumentation I 3.3.2.1-ACTIONS (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Sus and control rod Immediately-Mode Switch-Shutdown wit 1drawal.
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| Position channels-inoperable. Mil l
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| l E.2 Initiate action to Immediately fully insert all i insertable control rods in core cells j l containing one or more fuel assemblies.
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| I SURVEILLANCE REQUIREMENTS
| |
| .....................................N0TES - - --- -- - - - - --------
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| .1. - Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
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| : 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances entry into associated Conditions i and Required Actions may be delayed for up to 6 hours provided the '
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| associated Function maintains control rod block capability.
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| SURVEILLANCE FREQUENCY l
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| SR 3.3.2.1.1 - ---
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| ---- NOTE ----- ----- -- --
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| Not required to be performed until I hour after any control rod is withdrawn at s 10% RTP in MODE 2.
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| Perform CHANNEL FUNCTIONAL TEST. 92 days (continued)
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| O FERMI UNIT 2 3.3-19 Amendment No. 134 l
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| Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)
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| SURVEILLANCE FREQUENCY SR 3.3.2.1.2 -- ----- - - NOTE- - --- - -- - - -
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| j Not required to be performed until I bour after THERMAL POWER is s 10% RTP in MODE 1.
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| Perform CHANNEL FUNCTIONAL TEST. 92 days i
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| SR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.2.1.4 -- - ----- ----NOTE - - -- - --------
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| Not required to be performed until 1 hour after reactor mode switch is in the shutdown position.
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| Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.2.1.5 Verify the RBM is not bypassed when 24 months THERMAL POWER is = 30% RTP.
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| SR 3.3.2.1.6 -- --
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| ---.-------NOTE --- ----- - .--
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| Neutron detectors are excluded.
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| I Perform CHANNEL CALIBRATION. 24 months j l
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| SR 3.3.2.1.7 Verify control rod sequences input to the Prior to RWM are in conformance with the declaring RWM prescribed withdrawal sequence. OPERABLE following loading of sequence into RWM O
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| FERMI UNIT 2 3.3 20 Amendment No. 134 I l
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| l i
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| l Control Rod Block Instrumentation 3.3.2.1 8
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| Table 3.3.2.1 1 (page 1 of 1)
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| Control Rod Block Instrimentation APPLICABLE MODES OR
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| * OTER SPECIFIED REQUIRED SLRVEILLANCE ALLOWABLE RmCTION COWITIONS CHANNELS REQUIREENTS VALUE
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| : 1. Rod Block Monitor
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| : a. Upscale (a) 2 SR 3.3.2.1.3 As specified in SR 3.3.2.1.5 the COLR SR .3.3.2.1.6
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| : b. Inop. (a) 2 SR 3.3.2.1.3 NA
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| : c. Downscale (a) 2 SR 3.3.2.1.3 As specified in SR 3.3.2.1.6 the COLR
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| : 2. Rod Worth Minimizer 1(b) 2(b)
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| . 1 SR 3.3.2.1.1 NA SR 3.3.2.1.2 SR 3.3.2.1.7
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| : 3. Reactor Mode Switch - Shutdown (c) 2 SR 3.3.2.1.4 NA Position O (a) THERMAL POWER = 302 RTP.
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| .(b) With TERMAL POWER s 101 RTP.
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| (c) Reactor mode switch in the shutdown position.
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| O FERMI - UNIT 2 3.3-21 Amendment No. 134
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| Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation LC0 3.3.2.2 Four channels of feedwater and main turbine high water level trip instrumentation shall be OPERABLE.
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| APPLICABILITY: THERMAL POWER a: 25% RTP.
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| ACTIONS
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| .....................................N0TE --- --
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| Separate Condition entry is allowed for each channel.
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| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more fL Jwater A.1 Place channel in 7 days and main turbine high trip.
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| water level trip O-channel (s) inoperable.
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| I B. Feedwater and main B.1 Restore feedwater and 2 hours turbine high water main turbine high level trip capability water level trip not maintained. capability.
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| C. Required Action and C.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP.
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| Time not met.
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| I I
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| i O
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| FERMI - UNIT 2 3.3 22 Amendment No. 134
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| Feedwater and Main Turbine High Water Level Trip Instrumentation 3.3.2.2
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| ( SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.2.2.2 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.2.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Value shall be 5 219 inches.
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| SR 3.3.2.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including valve actuation.
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| i l
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| 4 l
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| (3 '
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| V FERMI UNIT 2 3.3 23 Amendment No. 134
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| PAM Instrumentation 3.3.3.1 O 3.3 INSTRUMENTATION V
| |
| 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LC0 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1 1 shall be OPERABLE.
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| APPLICABILITY: MODES 1 and 2.
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| ACTIONS
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| .................................- - NOTES - ---- ---- --- --- --- -- - --
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| : 1. LC0 3.0.4 is not applicable.
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| : 2. Separate Condition entry is allowed for each Function.
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| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE G
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| V channel inoperable. status. 4 l
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| i B. Required Action and B.1 Initiate action in Immediately associated Completion accordance with Time of Condition A Specification 5.6.7.
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| not met.
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| C. ----- - NOTE- ------- C.1 Restore one required 7 days l Not applicable to channel to OPERABLE i 3rimary containment status.
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| lydrogen and primary j containment oxygen 4 concentration channels.
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| One or more Functions with two required channels inoperable.
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| g(3 (continued)
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| FERMI UNIT 2 3.3-24 Amendment No. 134
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| PAM Instrumentation 3.3.3.1 ACTIONS (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME D. Two required primary D.1 Restore one required 72 hours containment hydrogen primary containment concentration channels hydrogen inoperable. concentration channel to OPERABLE status.
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| E AND Two required primary containment oxygen D.2 Restore one required 72 hours concentration channels primary containment inoperable. Oxygen concentration channel to OPERABLE status.
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| E. Required Action and E.1 Enter the Condition Immediately associated Completion referenced in Time of Condition C Table 3.3.3.1-1 for f or D not met. the channel.
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| (
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| F. As required by F.1 Be in MODE 3. 12 hours Required Action E.1 and referenced in Table 3.3.3.1 1.
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| G. As required by G.1 Initiate action in Immediately Required Action E.1 accordance with and referenced in Specification 5.6.7.
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| Table 3.3.3.1-1.
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| O FERMI - UNIT 2 3.3-25 Amendment No. 134
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| PAM Instrumentation 3.3.3.1 r
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| ( SURVEILLANCE REQUIREMENTS
| |
| - - - -- ----- -- ---~~~--
| |
| ...................................- NOTE----
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| These SRs apply to each Function in Table 3.3.3.1-1.
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| SURVEILLANCE FREQUENCY SR 3.3.3.1.1 Perform CHANNEL CHECK. 31 days I
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| SR 3.3.3.1.2 --- -.---- -
| |
| - NOTES - --- ---- -- -
| |
| : 1. Only applicable to Functions 7 and 8.
| |
| : 2. Not required to be performed until 72 hours for one channel and 7 days for the second channel after a 15% RTP.
| |
| Perform CHANNEL CALIBRATION. 92 days SR 3.3.3.1.3 -- - -
| |
| --- NOTES------ -- ----- -
| |
| : 1. Not applicable to Functions 7 and 8.
| |
| : 2. Radiation detectors are excluded.
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| Perform CHANNEL CALIBRATION. 18 months O
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| l FERMI UNIT 2 3.3 26 Amendment No. 134
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| | |
| PAM Instrumentation 3.3.3.1 Table 3.3.3.1-1 (page 1 of 1)
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| Post Accident Monitoring Instrumentation CONDITIONS REFERENCED REQUIRED FRON REQUIRED FUNCTION- CHANNELS ACTION E.1
| |
| : 1. Reactor Vessel Pressure 2 F
| |
| : 2. Reactor Vessel Water Level - Fuel Zone 2 F
| |
| : 3. Reactor Vessel Water Level - Wide Range 2 F
| |
| : 4. Suppression Pool Water Level 2 F
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| : 5. Suppression Pool Water Temperature 2 F
| |
| : 6. Drywell Pressure Wide Range 2 F
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| : 7. Primary Containment 0, Concentration 2 F
| |
| : 8. Primary Containment H, Concentration 2 F
| |
| : 9. Primary Containment High Range Radiation Monitor 2 G
| |
| : 10. PCIV Position 2g tgtg F (a) Not required for itolation valves whose associated penetration flow path is isolated by at least one closed and deactivated automatic valve closed manual valve. blind flange. or check valve with flow through the valve secured.
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| (b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
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| O FERMI UNIT 2 3.3 27 Amendment No. 134
| |
| | |
| Remote Shutdown System 3.3.3.2
| |
| -O L)- 3.3 INSTRUMENTATION
| |
| )
| |
| 3.3.3.2 Remote Shutdown System LC0 3.3.3.2 The Division I Remote Shutdown System Functions in Table 3.3.3.2 1 shall be OPERABLE. .,
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS
| |
| .................................... NOTES------
| |
| : 1. LC0 3.0.4 is not applicable.
| |
| : 2. Separate Condition entry is allowed for each function.
| |
| CONDITIOd REQUIRED ACTION COMPLETION TIME O
| |
| O 'A. One or more required functions inoperable.
| |
| A.1 Restore required 30 days function to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| l p
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| (_)
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| l FERMI- UNIT 2 3.3 28 Amendment No. 134 J
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| l
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| | |
| 4 Remote Shutdown System 3.3.3.2 l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel.
| |
| SR 3.3.3.2.2 Verify each required control circuit and 18 months transfer switch is capable of performing the intended-function.
| |
| SR 3.3.3.2.3 Perform CHANNEL CALIBRATION for each 18 months required _ instrumentation channel.
| |
| O O
| |
| FERMI - UNIT 2 3.3-29 Amendment No. 134
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| '\.
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| | |
| Remote Shutdown System 3.3.3.2 Table 3.3.3.2 1 (page 1 of 1)
| |
| Remote Shutdown System Instrumentation INSTRUMENT FUNCTION
| |
| : 1. Reactor Vessel Pressure
| |
| : 2. Reactor Vessel Water Level
| |
| : 3. Suppression Chamber Water Temperature
| |
| : 4. Drywell Pressure
| |
| : 5. RHR Heat Exchanger Discharge Flow
| |
| : 6. RCIC Flow CONTROL FUNCTION
| |
| : 1. Control Rod Drive Pump A
| |
| : 2. Control Rod Drive Pump B
| |
| : 3. RHR Valve E1150-F009
| |
| : 4. RHR Valve E1150-F008
| |
| : 5. RHR Valve E1150-F006A O
| |
| V
| |
| : 6. Recirc Pump A Valve B3105-F023A
| |
| : 7. Main Steam Line (D) Relief Valve B2104 F013A
| |
| : 8. Main Steam Line (C) Relief Valve B2104 F0138
| |
| : 9. RHR Valve E1150 F015A
| |
| : 10. RHR Valve E1150 F017A
| |
| : 11. RHR Valve E1150 F004A
| |
| : 12. RHR Pump A
| |
| : 13. RHR Valve E1150 F024A
| |
| : 14. RHR Valve E1150-F028A
| |
| : 15. RHR Valve E1150 F048A
| |
| : 16. RHR Valve E1150 F068A
| |
| : 17. RHR Service Water Pump A
| |
| : 18. RHR Service Water Pump C
| |
| : 19. Cooling Tower Fan A
| |
| : 20. Cooling Tower Fan C
| |
| : 21. RCIC Valve E5150-F059
| |
| : 22. RCIC Valve E5150 F045
| |
| : 23. RCIC Initiate
| |
| : 24. Division II DC Transfer
| |
| : 25. B0P Transfer
| |
| : 26. Division I DC Transfer
| |
| : 27. Division I AC Transfer
| |
| : 28. Swing Bus Transfer I (3
| |
| NY.
| |
| l FERM1 UNIT 2 3.3 30 Amendment No. 134
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS RPT) Instrumentation LC0 3.3.4.1 Two channels per trip system for each ATWS RPT 1 instrumentation Function listed below shall be OPERABLE: j
| |
| )
| |
| : a. Reactcr Vessel Water Level-Low Low. Level 2: and
| |
| : b. Reactor Vessel Pressure-High.
| |
| l APPLICABILITY: MODE 1. l ACTIONS
| |
| .....................................N0TE - ----- - - -- - ------ - - -- - l Separate Condition entry is allowed for each channel. l CONDITION REQUIRED ACTION' COMPLETION TIME l
| |
| A. One or more channels A.1 Restore channel to 14 days inoperable. OPERABLE status.
| |
| 2 A.2 -------NOTE- -- -- -
| |
| Not applicable if inoperable channel is the result of an inoperable breaker.
| |
| Place channel in 14 days trip.
| |
| (continued)
| |
| C)
| |
| 't FERMI UNIT 2 3.3 31 Amendment No. 134
| |
| | |
| ATWS RPT Instrumentation 3.3.4.1 w
| |
| ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B, One Function with B.1 Restore ATWS RPT trip 72 hours -
| |
| AlWS RPT trip capability.
| |
| capability not maintained.
| |
| C. Both Functions with C.1 Restore AlWS RPT trip 1 hour ATWS RPT trip capability for one capability not Function.
| |
| maintained.
| |
| D. Required Action and D.1 Remove the associated 6 hours associated Completion recirculation pump Time not met. from service.
| |
| D.2 Be in MODE 2. 6 hours o
| |
| V FERMI - UNIT 2 3.3-32 Amendment No. 134
| |
| | |
| I ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS
| |
| .....................................N0TE.............................-.......
| |
| When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 2 hours provided the associated Function maintains ,
| |
| A1WS RPT trip capability. l
| |
| .............................................................................. l SURVEILLANCE FREQUENCY
| |
| {
| |
| SR 3.3.4.1.1 Perform CHANNEL CHECK. 12 hours 1
| |
| SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST. 31 days i
| |
| SR 3.3.4.1.3 Perform CHANNEL CALIBRATION. The 18 months Allcwable Values shall be-I
| |
| : a. Reactor Vessel Water Level-Low Low, I Level 2: a 103.8 inches; and
| |
| : b. Reactor Vessel Pressure-High:
| |
| s 1153 psig.
| |
| 1 SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
| |
| G U
| |
| FERMI UNIT 2 3.3 33 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 3.3 INSTRUMENTATION 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation LC0 3.3.5.1 The ECCS instrumentation for each function in Table 3.3.5.11 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.5.1 1.
| |
| ACTIONS
| |
| ..................................... NOTE -- -- - --
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in r i Table 3.3.5.1 1 for s the channel. ,
| |
| 1 (continued)
| |
| O FERMI UNIT 2 3.3-34 Amendment No. 134 1
| |
| l l
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPI.ETION TIME B. As required by B.1 - - --- NOTES - - - -
| |
| Required Action A.1 1. Only applicable !
| |
| and referenced in in MODES 1, 2. I Table 3.3.5.1-1. and 3. I
| |
| : 2. Only applicable for Functions 1.a. 1.b. 2.a.
| |
| 2.b. 2.d and 2.g.
| |
| Declare supported 1 hour from feature (s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature (s) in both divisions AND B.2 .........N0Te........
| |
| Only applicable for Functions 3.a and 3.b.
| |
| Declare High Pressure 1 hour from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable, initiation capability AND B.3 Place channel in 24 hours trip.
| |
| (continued) l
| |
| )
| |
| O V i 1
| |
| FERMI ' UNIT 2 3.3 35 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 (3 ACTIONS (continued)
| |
| (/
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 - - NOTES- --- -
| |
| Required Action A.1 1. Only applicable and referenced in in MODES 1, 2.
| |
| Table 3.3.5.1 1. and 3.
| |
| : 2. Only applicable for Functions 1.c, 2.c. 2.e, ,
| |
| and 2.f.
| |
| Declare supported I hour from feature (s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable, feature (s) in both divisions )
| |
| AND C.2 Restore channel to 24 hours OPERABLE status.
| |
| (continued) 4 O
| |
| FERMI UNIT 2 3.3-36 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 1
| |
| ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 - --- NOTE - ---- -
| |
| Required Action A.1 Only applicable if and referenced in HPCI pump suction is Table 3.3.5.1-1. not aligned to the suppression pool.
| |
| )
| |
| Declare HPCI System 1 hour from inoperable. discovery of loss of HPCI initiation capability AN_Q D.2.1 Place channel in 24 hours trip.
| |
| 2 D 2.2 Align the HPCI pump 24 hours
| |
| / suction to the suppression pool.
| |
| (continued) i l
| |
| l l
| |
| V'O FERMI UNIT 2 3.3 37 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS (continued)
| |
| . CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 Declare Automatic 1 hour from' Required Action A.1 Depressurization discovery of and referenced in System (ADS) valves loss of ADS Table 3.3.5.1 1. inoperable. initiation capability in both trip systems AND E.2 Place channel in 96 hours from trip. discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable i AND 8 days I
| |
| (continued) p
| |
| ()
| |
| FERMI - UNIT 2 3.3-38 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 i ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| l F. As required by F.1 - ---
| |
| NOTE- -- -
| |
| Required Action A.1 Only applicable for and referenced in Functions 4.c. 4.e.
| |
| Table 3.3.5.1 1. 4. f. 4.g. 5.c, 5.e.
| |
| 5.f. and 5.g.
| |
| I Declare ADS valves 1 hour from l inoperable. discovery of l loss of ADS
| |
| : initiation I
| |
| capability in l both trip systems AND F.2 Restore channel to 96 hours from OPERABLE status. discovery of inoperable channel
| |
| 's concurrent with HPCI or RCIC inoperable AND 4 8 days i
| |
| G. Required Action and G.1 Declare associated Immediately associated Completion supported feature (s) l Time of Condition B. inoperable.
| |
| l C. D. E. or F not met.
| |
| l l
| |
| l fh) l FERMI UNIT 2 3.3 39 Amendment No. 134 l
| |
| | |
| ECCS Instrumentation 3.3.5.1 (m) SURVEILLANCE REQUIREMENTS
| |
| .................................... NOTES-- --------------- - --- - - ****
| |
| : 1. Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function. ,
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows: (a) for up to 6 hours for Function 3.c: and (b) for up to 6 hours for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains ECCS initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days d SR 3.3.5.1.3 Verify the trip unit setpoint. 92 days SR 3.3.5.1.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.5.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months SR 3.3.5.1.6 Perform CHANNEL FUNCTIONAL TEST. 18 months l r
| |
| 1
| |
| ()
| |
| ,m FERMI - UNIT 2 3.3 40 Amendment No. 134
| |
| | |
| ECCS Instrumentation 3.3.5.1 O Table 3.3.5.1 1 (page 1 of 5)
| |
| Emergency Core Cooling system Instrimentation APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OROTER CHANNELS FROM SPECIFIED PER REQUIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
| |
| : 1. Core Spray System
| |
| : a. Reactor Vessel Water 1.2.3. 4(b) B SR 3.3.5.1.1 a 24.8 inches Level - Low Low Low. SR 3.3.5.1.2 Level 1 4(a). 5(a) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : b. Drywell 1.2.3 4(b) B SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : c. Reactor Steam Dome 1.2.3 4 C SR 3.3.5.1.1 a 441 psig Pressure - Low SR 3.3.5.1.2 (Injection Permissive) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 4(a). 5(a) 4 B Sh 3.3.5.1.1 a 441 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : d. Manual Initiation 1.2.3. 2(C) C SR 3.3.5.1.6 NA 4(a). 5(a)
| |
| (continued)
| |
| (a) When associated subsystem (s) of LCO 3.5.2 are required to be OPERABLE.
| |
| (b) Also required to initiate the associated emergency diesel generator (EDG).
| |
| (c) Individual component controls.
| |
| FERMI - UNIT 2 3.3 41 Amendment No. 134 l l
| |
| | |
| ECCS Instrumentation 3.3.5.1 6
| |
| Table 3.3.5.1 1 (page 2 of 5)
| |
| Emergency Core Cooling System Instrumentation l
| |
| l l APPLICABLE CONDITIONS MODES REQUIRED REFERENCED OROTER CHANNELS FROM .
| |
| SPECIFIED PER REQUIRED SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE
| |
| : 2. Low Pressure Coolant Injection (LPCI) System
| |
| : a. Reactor Vessel Water 1.2.3. 4 B SR 3.3.5.1.1 e 24.S inches Level - Low Low Low. SR 3.3.5.1.2 Level 1 4(a), $(a) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : b. Drywell 1.2.3 4 B SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : c. Reactor Steam Dome 1.2.3 4 C SR 3.3.5.1.1 m 441 psig Pressure - Low SR 3.3.5.1.2 (Injection Permissive) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5 4(a), 5(a) 4 B SR 3.3.5.1.1 = 441 psig SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.4
| |
| ( SR 3.3.5.1.5
| |
| : d. Reactor Vessel Water 1.2,3 4 B SR 3.3.5.1.1 e 103.8 Level - Low Low. Level SR 3.3.5.1.2 inches 2 (Loop Select Logic) 4(a). 5(a) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : e. Reactor Steam Dome 1.2.3. 4 C SR 3.3.5.1.1 m 886 psig Pressure - Low (Break SR 3.3.5.1.2 Detection Logic) 4(a)5(a) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : f. Riser Differential 1.2.3 4 C SR 3.3.5.1.1 s 0.927 psid Pressure - High (Break SR 3.3.5.1.2 Detection) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : g. Recirculation Pinp 1.2.3 4 per punp B SR 3.3.5.1.1 s 1.927 psid Differential SR 3.3.5.1.2 Pressore - High (Break SR 3.3.5.1.3 Detection) SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : h. Manual Initiation 1.2.3 2(C) C SR 3.3.5.1.6 NA 4(a). 5(a)
| |
| (continued) j (a) When associated subsystem (s) of LCO 3.5.2 are required to be OPERABLE.
| |
| (c)- Individual conponent controls.
| |
| FERMI - UNIT 2 3.3 42 Amendment No. 134 l
| |
| i l
| |
| | |
| ECCS Instrumentation 3.3.5.1 s
| |
| Table 3.3.5.1 1 (page 3 of 5)
| |
| Emergency Core Cooling System Instrunentation APPLICABLE CONDITIONS MODES OR REQUIRED PIFERENCED OTER CHANNELS FROM SPECIFIED PER REQUIRED SLRVEILLANCE ALLDWABLE CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE FUNCTION i
| |
| : 3. High Pressure Coolant l i
| |
| In,)ection OPCI) System
| |
| : a. Reactor Vessel Water 1. 4 B SR 3.3.5.1.1 = 103.8 Level - Low Law. SR 3.3.5.1.2 inches Level 2 2(d), 3(d) SR 3.3.5.1.3 l SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : b. Drywell 1. 4 8 SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 2Id)3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : c. Reactor Vessel Water 1. 2 C SR 3.3.5.1.1 s 219 inches Level - High. Level 8 SR 3.3.5.1.2 2(d). 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : d. Condensate Storage 1. 2 D SR 3.3.5.1.1 a 0 inches Tank Level - Low SR 3.3.5.1.2 2(d). 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 y SR 3.3.5.1.5
| |
| : e. Suppression Pool Water 1. 2 D SR 3.3.5.1.1 s 5.0 inches Level - High SR 3.3.5.1.2 2(d), 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : f. Manual Initiation 1, 1(c) C SR 3.3.5.1.6 NA 2(d), 3(d)
| |
| (continued)
| |
| (c) Individual conponent controls.
| |
| (d) With reactor steam dome pressure > 150 psig.
| |
| t l
| |
| (d FERMI UNIT 2 3.3 43 Amendment No. 134
| |
| | |
| i
| |
| {
| |
| ECCS Instrumentation 3.3.5.1 e
| |
| /
| |
| Table 3.3.5.1 1 (page 4 of 5)
| |
| Emergency Core Cooling System Instr o entation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER REQ'JIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREENTS VALUE
| |
| : 4. Automatic Depressurization System (ADS) Trip System A
| |
| : a. Reactor Vessel Water 1, 2 E SR 3.3.5.1.1 = 24.8 inches Level - Low Low Low. SR 3.3.5.1.2 Level 1 2Id). 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : b. Drywell 1, 2 E SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 2Id). 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : c. Automatic 1. 1 F SR 3.3.5.1.2 s 117 seconds Depressurization SR 3.3.5.1.4 System Initiation 2(d), 3(d) SR 3.3.5.1.5 {
| |
| Timer
| |
| : d. Reactor Vessel Water 1. 1 E SR 3.3.5.1.1 = 171.9 Level - Low, Level 3 SR 3.3.5.1.2 inches (Confirmatory) 2(d)3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 N SR 3.3.5.1.5
| |
| : e. Core Spray P mp 1. I per pinp F SR 3.3.5.1.1 = 125 psig Discharge SR 3.3.5.1.2 Pressure - High 2(d),3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : f. Low Pressure Coolant 1. 2 per pump F SR 3.3.5.1.1 = 115 psig Injection Puip SR 3.3.5.1.2 Discharge 2Id)3(d) SR 3.3.5.1.3 Pressure - High SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : g. Drywell 1. 2 F SR 3.3.5.1.2 s 450 seconds Pressure - High Bypass SR 3.3.5.1.3 2(d), 3(d) SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : h. Manual Inhibit 1. 1 F SR 3.3.5.1.5 NA 2(d), 3(d)
| |
| : 1. Manual Initiation 1. I r F SR 3.3.5.1.6 NA 2(d), 3(d)
| |
| (continued)
| |
| (d) With reactor steam dome pressure > 150 psig.
| |
| FERMI UNIT 2 3.3 44 Amendment No. 134
| |
| | |
| r ECCS Instrumentation 3.3.5.1
| |
| . Table 3.3.5.1 1 (page 5 of 5)
| |
| V Emergency Core Cooling System Instrisnentation l
| |
| l APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FRON SPECIFIED PER REQUIRED SlRVEILLANCE ALLOWABLE
| |
| ' FUNCTION CONDITIONS FUNCTION ACTION A.1 REQUIREMENTS VALUE I 5. ADS Trip System B
| |
| : a. Reactor Vessel Water 1. 2 E SR 3.3.5.1.1 e 24.8 inches Level - Low Low Low. SR 3.3.5.1.2 Level 1 2(d), 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : b. Drywell 1, 2 E SR 3.3.5.1.1 s 1.88 psig Pressure - High SR 3.3.5.1.2 2(d). 3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : c. Automatic 1, 1 F SR 3.3.5.1.2 s 117 seconds Depressurization SR 3.3.5.1.4 System Initiation 2(d), 3(d) SR 3.3.5.1.5 Timer
| |
| : d. Reactor Vessel Water 1. 1 E SR 3.3.5.1.1 e 171.9 Level- Low. Level 3 SR 3.3.5.1.2 inches (Confirmatory) 2(d),3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : e. Core Spray Ptap 1. 1 per pump F SR 3.3.5.1.1 = 125 psig Discharge SR 3.3.5.1.2 Pressure - High 2(d),3(d) SR 3.3.5.1.3 SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : f. Low Pressure Coolant 1. 2 per pimp F SR 3.3.5.1.1 e 115 psig Injection Ptap SR 3.3.5.1.2 Discharge 2(d), 3(d) SR 3.3.5.1.3 Pressure - High SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : g. Drywell 1, 2 F SR 3.3.5.1.2 s 450 seconds Pressure - High Bypass SR 3.3.5.1.3 2(d),3(d) SR 3.3.5.1.4 SR 3.3.5.1.5
| |
| : h. Manual Inhibit 1, 1 F SR 3.3.5.1.5 NA 2(d), 3(d)
| |
| : 1. Manual Initiation 1. I r F SR 3.3.5.1.6 NA va ve 2(d), 3(d)
| |
| (d) With reactor steam dome pressure > 150 psig.
| |
| , .vn FERMI UNIT 2 3.3-45 Amendment No. 134 l
| |
| | |
| RCIC System Instrumentation 3.3.5.2 I' 3.3 INSTRUMENTATION 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation LC0 3.3.5.2 The RCIC System instrumentation for each Function in Table 3.3.5.2-1 shall be OPERABLE.
| |
| APPLICABILITY: MODE 1.
| |
| MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS
| |
| .................................... NOTE- --
| |
| Separa37,gondj$jon 7ntry !s anoygd for eaSh ?hannel
| |
| )
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.2 1 for the channel. I B. As required by B.1 Declare RCIC System I hour from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.2 1. initiation capability AND B.2 Place channel in 24 hours trip.
| |
| C. As required by C.1 Restore channel to 24 hours Required Action A.1 OPERABLE status. I and referenced in Table 3.3.5.2 1.
| |
| (continued)
| |
| G
| |
| .N)
| |
| FERMI - UNIT 2 3.3 46 Amendment No. 134
| |
| | |
| RCIC System Instrumentation 3.3.5.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 - ----- NOTE----- ---
| |
| Required Action A.1 Only applicable if and referenced in RCIC p g suction is Table 3.3.5.2-1. not aligned to the suppression pool.
| |
| Declare RCIC System I hour from inoperable. discovery of loss of RCIC initiation capability 8NQ D.2.1 Place channel in 24 hours trip.
| |
| 2 D.2.2 Align RCIC pump 24 hours o suction to the Q suppression pool.
| |
| E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable.
| |
| Time of Condition B, C, or D not met.
| |
| O
| |
| ' FERMI UNIT 2 3.3 47 Amendment No. 134 j l
| |
| l
| |
| | |
| RCIC System Instrumentation 3.3.5.2 O suavei uaace aeouineae#1s
| |
| .....................................N0TES...........
| |
| : 1. Refer to Table 3.3.5.2 1 to determine which SRs apply for each RCIC Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required i Actions may be delayed as follows: (a) for up to 6 hours for Function 2:
| |
| . and (b) for up to_6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.5.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.5.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days b
| |
| .a SR 3.3.5.2.3 Verify the trip unit setpoint. 92 days SR 3.3.5.2.4 Per form CHANNEL CALIBRATION. 18 months SR 3.3.5.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months SR 3.3.5.2.6 Perform CHANNEL FUNCTIONAL TEST. 18 months l
| |
| FERMI . UNIT 2 3.3 48 Amendment No. 134 ;
| |
| i
| |
| | |
| RCIC System Instrumentation 3.3.5.2 l
| |
| Table 3.3.5.2 1 (page 1 of 1) y Reactor Core Isolation Cooling System Instrunentation CONDITIONS REQUIRED StRVEILLANCE REFERENCED CHANNELS ALLOWABLE FROM REQUIRED REQUIREMENTS PER FUNCTION VALUE FUNCTION ACTION A.1
| |
| : 1. Reactor Vessel Water 4 B SR 3.3.5.2.1 e 103.8 inches Level - Low Low.. Level 2 SR 3.3.5.2.2 l SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5
| |
| : 2. Reactor Vessel Water 2 C SR 3.3.5.2.1 s 219 inches Level - High. Level 8 SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5
| |
| : 3. Condensate Storage Tank 2 D SR 3.3.5.2.1 a 0 inches Level - Low SR 3.3.5.2.2 SR 3.3.5.2.3 SR 3.3.5.2.4 SR 3.3.5.2.5
| |
| : 4. Manual Initiation 1 per valve C SR 3.3.5.2.6 NA O(%
| |
| 1
| |
| (*)
| |
| %.)
| |
| FERMI - UNIT 2 3.3 49 Amendment No. 134 ,
| |
| l
| |
| | |
| Pri ary Containment Isolation Instrumentation 3.3.6.1 n
| |
| ! 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LC0 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE. ,
| |
| I i
| |
| l APPLICABILITY: According to Table 3.3.6.1-1. I l
| |
| ACTIONS l
| |
| .................................... NOTE -- ---- --- - --------- ------ -
| |
| Separate Condition entry is allowed for each channel.
| |
| 1 CONDITION REQUIRED ACTION COMPLETION TIME 1 A. One or more required A.1 Place channel in 12 hours for channels inoperable. trip. Functions 1.f.
| |
| O' V
| |
| 2.a. 2.c.
| |
| and 6.b MD )
| |
| 24 hours for Functions other than Functions
| |
| : 1. f. 2. a . 2. c.
| |
| and 6.b B. One or more automatic B.1 Restore isolation 1 hour Functions with capability.
| |
| isolation capability not maintained.
| |
| I (continued) l
| |
| ]J FERMI UNIT 2 3.3 50 Amendment No. 134
| |
| | |
| )
| |
| Primary Containment Isolation Instrumentation 3.3.6.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Enter the Condition Immediately- )
| |
| associated Completion referenced in l Time of Condition A Table 3.3.6.1-1 for 1 or B not met, the channel.
| |
| l I
| |
| D. As required by D.1 Isolate associated 12 hours l Required Action C.1 main steam line and referenced in (MSL).
| |
| Table 3.3.6.1 1.
| |
| D.2.1 Be in MODE 3. 12 hours AND D.2.2 Be in MODE 4. 36 hours O
| |
| E. As re uired by E.1 Be in MODE 2. 6 hours Recui ed Action C.1 anc referenced in Table 3.3.6.1-1.
| |
| F. As required by F.1 Isolate the affected I hour Recuired Action C.1 penetration flow anc referenced in path (s).
| |
| Table 3.3.6.1-1. {
| |
| G. As required by G.1 Isolate the affected 24 hours Required Action C.1 penetration flow and referenced in path (s).
| |
| Table 3.3.6.1 1.
| |
| (continued) 1 FERMI - UNIT 2 3.3 51 Amendment No. 134
| |
| | |
| p Primary Containment Isolation Instrumentation 1 3.3.6.1 I O ac'ioas- (ce"t4 #ed)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME H. As required by . H.1 Be in MODE 3. 12 hours -
| |
| Required Action C.1 and referenced in atD Table 3.3.6.1-1.
| |
| H.2 Be in MODE 4. 36 hours E
| |
| Required Action and-associated Completion Time for Condition F or G not met.
| |
| I. As required by I.1 Declare associated I hour Required Action C.1 standby liquid and referenced in control subsystem Table 3.3.6.1-1. (SLC) inoperable.
| |
| E I.2 Isolate the Reactor 1 hour Water Cleanup System.
| |
| J. As required by J.1 Initiate action to Immediately Required Action C.1 restore channel to and referenced:in OPERABLE status.
| |
| Table 3.3.6.1-1. -
| |
| 3 E )
| |
| J.2 Initiate action to Immediately isolate the Residual Heat Removal (RHR)
| |
| Shutdown Cooling System.
| |
| O FERMI UNIT 2 3.3 52 Amendment No. 134
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 b)
| |
| G SURVEILLANCE REQUIREMENTS
| |
| ..................................... NOTES-----------
| |
| : 1. Refer to Table 3.3.6.11 to determine which SRs apply --- for each Primary Containment Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to:
| |
| : a. 2 hours for Function 5.a when testing non redundant circuitry that results in loss of isolation capability associated with this Function, provided Functions 5.b. 5.c, and 5.e are OPERABLE:
| |
| : b. 6 hours for Functions 1. 2, 5 (other than non-redundant circuitry of 5.a). and 6, provided the associated Function maintains isolation capability; and
| |
| : c. 8 hours for Functions 3 and 4. provided the associated Function maintains isolation capability.
| |
| 1 SURVEILLANCE FREQUENCY (3
| |
| v' SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days i
| |
| l l
| |
| SR 3.3.6.1.3 Verify the trip unit setpoint. 92 days SR 3.3.6.1.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months l
| |
| SR 3.3.6.1.6 Perform CHANNEL FUNCTIONAL TEST. 18 months n
| |
| (continued)
| |
| {}
| |
| FERMI UNIT 2 3.3-53 Amendment No. 134
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.1.7 -- - -
| |
| --- -NOTES -- - -- -----
| |
| : 1. Radiation detectors may be excluded.
| |
| : 2. Channel sensor response times are not required to be measured.
| |
| Verify the ISOLATION SYSTEM RESPONSE TIME 18 months on a is within limits. STAGGERED TEST BASIS
| |
| (
| |
| l O
| |
| FERMI UNIT 2 3.3 54 Amendment No. 134
| |
| [.
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 (m
| |
| Nd
| |
| ) Table 3.3.6.11 (page 1 of 4)
| |
| Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
| |
| : 1. Main Steam Line Isolation
| |
| : a. Reactor Vessel Water 1.2.3 2 D SR 3.3.6.1.1 a 24.8 inches Level - Low Low Low. SR 3.3.6.1.2 Level 1 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7
| |
| : b. Main Steam Line 1 2 E SR 3.3.6.1.1 m 736 psig Pressure - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : c. Main Steam Line 1.2.3 2 per D SR 3.3.6.1.1 s 118.4 psid Flow - High MSL SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7
| |
| : d. Condenser 1. 2 D SR 3.3.6.1.1 s 7.05 psia Pressure - High SR 3.3.6.1.2 i g 2(a), 3(a) SR 3.3.6.1.3 V SR 3.3.6.1.4 SR 3.3.6.1.5 ;
| |
| : e. Main Steam Tunnel 1.2.3 2 per D SR 3.3.6.1.1 s 206*F i Temperature - High trip SR 3.3.6.1.2 string SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : f. Main Steam Line 1.2.3 2 D SR 3.3.6.1.1 s 3.6 x full Radiation - High SR 3.3.6.1.2 power SR 3.3.6,1.4 background SR 3.3.6.1.5
| |
| : g. Turbine Building Area 1.2.3 4 D SR 3.3.6.1.1 s 206*F Temperature - High SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : h. Manual Initiation 1.2.3 1 r G SR 3.3.6.1.6 NA va e )
| |
| (continued)
| |
| (a) Except when bypassed during reactor shutdown or for reactor startup under adninistrative control.
| |
| %)
| |
| FERMI - UNIT 2 3.3-55 Amendment No. 134
| |
| | |
| Pricary Containment Isolation Instrumentation 3.3.6.1
| |
| -.m (y) Table 3.3.6.1 1 (psge 2 of 4)
| |
| Primary Containment Isol'. tion Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED StRVEILLANCE ALLOWABLE FUNCTION CONDITIOKS SYSTEM ACTION C.1 REQUIREENTS VALUE
| |
| : 2. Primary Containment Isolation
| |
| : a. Reactor vessel Water 1.2.3 2 H SR 3.3.6.1.1 = 171.9 inches Level-Low. Level 3 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : b. Reactor Vessel Water 1.2.3 2 H SR 3.3.6.1.1 = 103.8 inches Level - Low. Level 2 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3,3.6.1.4 SR 3.3.6.1.5
| |
| : c. Drywell Pressure - High 1.2.3 2 H SR 3.3.6.1.1 s 1.88 psig SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : d. Manual Initiation 1.2.3 1 r G SR 3.3.6.1.6 NA va ve
| |
| : 3. High Pressure Coolant
| |
| (_/ Injection (WCI) System
| |
| . Isolation
| |
| : a. HPCI Steam Line 1.2.3 1 F SR 3.3.6.1.1 s 410 inches Flow - High SR 3.3.6.1.2 of water with SR 3.3.6.1.3 time delay SR 3.3.6.1.4 e 1 second, and SR 3.3.6.1.5 s 5 seconds
| |
| : b. WCI Steam Supply Line 1.2.3 2 F SR 3.3.6.1.1 a 90 psig Pressure - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : c. HPCI Turbine 1.2.3 2 F SR 3.3.6.1.1 s 20 psig Exhaust Diaphragn SR 3.3.6.1.2 Pressure -High SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : d. HPCI Equipment Room 1.2.3 1 F SR 3.3.6.1.1 s 162'F Temperature - High SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : e. Drpell Pressure - High 1.2.3 1 F SR 3.3.6.1.1 s 1.88 psig SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : f. Manual Initiation 1.2.3 1 per G SR 3.3.6.1.6 NA valve
| |
| . (continued)
| |
| FERMI - UNIT 2 3.3 56 Amendment No. 134
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.11 (page 3 of 4)
| |
| (s Primary containment Isolation Instrumentation APPLICABLE CONDITIONS NODE $ OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE
| |
| : 4. Reactor Core Isolation Cooling (RCIC) System Isolation
| |
| : a. RCIC Steam Line 1.2.3 1 F SR 3.3.6.1.1 s 95.0 inches Flow - High SR 3.3.6.1.2 of water with SR 3.3.6.1.3 time delay SR 3.3.6.1.4 = 1 second and SR 3.3.6.1.5 s 5 seconds
| |
| : b. RCIC Steam Supply 1.2.3 2 F SR 3.3.6.1.1 m 53 psiD Line Pressure- Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : c. RCIC Turbine 1.2.3 2 F SR 3.3.6.1.1 s 20 psig Exhaust Diaphragm SR 3.3.6.1.2 Pressure - High SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : d. RCIC Equipment Room 1.2.3 1 F SR 3.3.6.1.1 s 162'F Tenperature - High SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : e. Dr)wil 1.2.3 1 F SR 3.3.6.1.1 s 1.88 psig Pressure - High SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : f. Manual Initiation 1.2.3 1 per G SR 3.3.6.1.6 NA valve j l
| |
| (continued) l l
| |
| I l
| |
| i
| |
| [\
| |
| FERMI UNIT 2 3.3 57 Amendment No. 134
| |
| | |
| Prirary Containment Isolation Instrumentation 1 3.3.6.1 O Table 3.3.6.1 1 (page 4 of 4)
| |
| Primary Containment Isolation Instrtmentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREENTS VALUE 1
| |
| l l 5, Reactor Water Cleanup (RWCU) System Isolation
| |
| : a. Differential 1,2,3 1 F SR 3.3.6.1.1 s 63.4 gpm Flow - High SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : b. Area 1.2.3 1 per F SR 3.3.6.1.1 s 183'F Tenperature - High area SR 3.3.6.1.2 SR 3.3,6.1.4 SR 3.3.6.1.5
| |
| : c. Area Ventilation 1.2.3 1 per F SR 3.3.6.1.1 s 53'F Differential room SR 3.3.6.1.2 Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : d. SLC System Initiation 1.2 2(b) I SR 3.3.6.1.5 NA
| |
| : e. Reactor Vessel Water 1.2.3 2 F SR 3.3.6.1.1 a 103.8 inches Level - Low Low. SR 3.3.6.1.2 Level 2 SR 3.3,6.1.3 l SR 3.3.6.1.4
| |
| ( SR 3.3.6.1.5
| |
| : f. Manual Initiation 1.2.3 1 per G SR 3.3.6.1.6 NA valve
| |
| : 6. Shutdown Cooling System Isolation
| |
| : a. Reactor Steam Dome 1.2.3 1 F SR 3.3.6.1.1 s 95.5 psig Pressure - High SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : b. Reactor Vessel Water 3.4.5 2(c) J SR 3.3.6.1.1 a 171.9 inches Level - Low. Level 3 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5
| |
| : c. Manual Initiation 1.2.3 1 r G SR 3.3.6.1.6 NA va{ve (b) SLC System Initiation only inputs into one of the two trip systems.
| |
| (c) Only one trip system required in MODES 4 and 5 when RFR Shutdown Cooling System integrity maintained.
| |
| O FERMI UNIT 2 3.3 58 Amendment No. 134
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 1
| |
| 0 3.3 ins >nuaeaTarioa 3.3.6.2 Secondary Containment Isolation Instrumentation LC0 3.3.6.2 The secondary containment isolation instrumentation for ech Function in Table 3.3.6.2-1 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.6.2-1.
| |
| ACTIONS
| |
| ..................................... NOTE------ ----------- ---------.------ -
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Function 2 AND 24 hours for Functions other than Function 2 B. One or more automatic B.1 Restore secondary 1 hour Functions with containment isolation secondary containment capability.
| |
| isolation capability not maintained.
| |
| (continued) ,
| |
| i 7N V_
| |
| FERMI - UNIT 2 3.3-59 Amendment No. 134
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1.1 Isolate the Secondary 1 hour -
| |
| associated Completion Containment.
| |
| Time.
| |
| E C.1.2 Deciare associated 1 hour secondary containment isolation valves inoperable.
| |
| AND C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation.
| |
| C.2.2 Declare associated 1 hour O SGT subsystem (s) inoperable.
| |
| 1 l
| |
| l
| |
| .. -0 FERMI UNIT 2 3.3-60 Amendment No. 134
| |
| | |
| Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS
| |
| ................................. NOTES-------- -- -------- --- - -- -
| |
| : 1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment-Isolation Function.
| |
| : 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability.
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3- Verify the trip unit setpoint. 92 days O
| |
| SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months I l
| |
| i FERMI - UNIT 2- 3.3 61 Amendment No. 134 i
| |
| | |
| I Secondary Containment Isolation Instrumentation 3.3.6.2 O, Table 3.3.6.2 1 (page 1 of 1)
| |
| Secondary Containment Isolation Instrtamentation APPLICABLE MODES OR REQUIRED OTER CHANNELS SPECIFIED PER SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEN REQUIREENTS VALUE
| |
| : 1. Reactor Vessel Water 1.2.3. 2 SR 3.3.6.2.1 = 103.8 inches Level - Low Low. Level 2 (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 SR 3.3.6.2.5
| |
| : 2. Drywell Pressure - High 1.2.3 2 SR 3.3.6.2.1 s 1.88 psig SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.4 SR 3.3.6.2.5
| |
| : 3. Fuel Pool Ventilation 1.2.3 2 SR 3.3.6.2.1 s 6 nR/hr Exhaust Radiation-High (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5 4, Manual Initiation 1.2.3 1 SR 3.3.6.2.5 NA (a),(b)
| |
| (a) During operations with a potential for draining the reactor vessel.
| |
| (b) During CORE ALTERATIONS and during movement of irradiated fuel assemblies in secondary containment.
| |
| (
| |
| FERMI UNIT 2 3.3 62 Amendment No. 134
| |
| | |
| LLS Instrumentation 3.3.6.3
| |
| ; 3.3 INSTRUMENTATION.-
| |
| ~
| |
| 3.3.6.3 Low Low Set (LLS) Instrumentation LC0 3.3.6.3 The LLS valve instrumentation for each Function in Table 3.3.6.31 shall be OPERABLE.
| |
| APPLICABILITY:- MODES 1. 2.:and 3.
| |
| ACTIONS C0f0ITION REQUIRED ACTION COMPLETION TIME A. One LLS valve A.1 Restore channel (s) to 14 days inoperable due'to OPERABLE status, inoperable channel (s).
| |
| (continued) 4 O
| |
| ' FERMI - UNIT 2 3.3 63 Amendment No. 134
| |
| | |
| i LLS Instrumentation '
| |
| 3.3.6.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. - -
| |
| NOTE - -
| |
| B.1 Restore one tailpipe 24 hours -
| |
| Separate Condition pressure switch for entry is allowed for 11 OPERABLE SRVs to each SRV. OPERABLE status.
| |
| AND l One or more safety /
| |
| relief valves (SRVs) B.2 Restore one tailpipe 24 hours with one or more pressure switch in '
| |
| Function 3 channel (s) each Division for an 1 inoperable. OPERABLE SRV in the lowest setpoint group, to OPERABLE status.
| |
| AND
| |
| ............N0TE- -- --
| |
| LCO 3.0.4 is not applicable.
| |
| B.3 Restore both tailpipe Prior to - l pressure switches for entering MODE 2 11 OPERABLE SRVs. or 3 from MODE 4 including 4 of 5 OPERABLE SRVs with the lowest relief setpoints, to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met.
| |
| C.2 Be in MODE 4. 36 hours l
| |
| Two LLS valves inoperable due to inoperable channels. ,
| |
| I O
| |
| FERMI - UNIT 2 3.3 64 Amendment No. 134
| |
| | |
| LLS Instrumentation 3.3.6.3 n
| |
| Q SURVEILLANCE REQUIREMENTS
| |
| ..................................... NOTE- . -
| |
| Refer to Table 3.3.6.3 1 to determine which SRs apply for each Function.
| |
| SURVEILLANCE FREQUENCY SR 3.3.6.3.1 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.6.3.2 Perform CHANNEL FUNCTIONAL TEST for 31 days portion of the channel outside primary containment.
| |
| SR 3.3.6.3.3 Perform CHANNEL CALIBRATION. 18 months
| |
| ,SR 3.3.6.3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months
| |
| ]w G
| |
| V FERMI - UNIT 2 3.3-65 Amendment No. 134 l
| |
| | |
| I J
| |
| LLS Instrumentation 3.3.6.3 r-( Table 3.3.6.3 1 (page 1 of 1)
| |
| Lw Lw Set Instrunentation REQUIRED CHANNELS PER SLRVEILLANCE ALLOWABLE FUNCTION FUNCTION REQUIREENTS VALUE.
| |
| : 1. ' Reactor Steam Dome Pressure- High 1 per LLS valve SR 3.3.6.3.1 s 1113 psig SR 3.3.6.3.3 SR 3.3.6.3.4
| |
| : 2. Lw Lw Set Pressure Setpoints 2 per LLS valve SR 3.3.6.3.1 Lw:
| |
| SR 3.3.6.3.3 Open s 1037 psig SR 3.3.6.3.4 Close (a)
| |
| High:
| |
| Open s 1067 psig l Close (a)
| |
| : 3. Tailpipe Pressure Switch 2 per SRV SR 3.3.6.3.2 = 25 psig and SR 3.3.6.3.3 s 35 psig SR 3.3.6.3.4 (a) = 100 psi belw actual opening setpoint.
| |
| i
| |
| )
| |
| FERMI - UNIT 2 3.3 66 Amendment No. 134
| |
| | |
| i CREF System Instrumentation ]
| |
| 3.3.7.1 '
| |
| 3.3 INSTRUMENTATION 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrunientation LC0 3.3.7.1 The CREF System instrumentation for each Function in -
| |
| Table 3.3.7.1 1 shall be OPERABLE.
| |
| I APPLICABILITY: According to Table 3.3.7.1-1.
| |
| ACTIONS i
| |
| ..................................... NOTE-- -- --- ----- - --------- --- --- -
| |
| -Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Enter the Condition Immediately channels inoperable. referenced in Table 3.3.7.1 1 for ON the channel.
| |
| B. As required by B.1 Declare associated 1 hour from Required Action A.1 CREF subsystem discovery of and referenced in inoperable. loss of CREF Table 3.3.7.1 1. initiation capability in both trip systems AND B.2 Place channel in 24 hours trip. ,
| |
| i l
| |
| (continued)
| |
| O FERMI UNIT;2 3.3-67 Amendment No. 134 ,
| |
| i
| |
| | |
| . 1 CREF System Instrumentation 3.3.7.1 ACTIONS (continued)
| |
| REQUIRED ACTION COMPLETION TIME CONDITION C. As required by C.1 Declare associated I hour from Required Action A.1 CREF subsystem discovery of and referenced in inoperable. loss of CREF Table 3.3.7.1 1. initiation capability in both trip systems AND C.2 Place channel in 6 hours downscale trip.
| |
| D. Required Action and D.1 Place the CREF System Immediately associated Completion in the recirculation Time of Condition B mode of operation.
| |
| or C not met.
| |
| 08 D.2 Declare associated Immediately CREF subsystem 3 inoperable. '
| |
| i l
| |
| l
| |
| [G FERMI UNIT 2 3.3-68 Amendment No. 134
| |
| | |
| CREF System Instrumentation 3.3.7.1
| |
| . SURVEILLANCE REQUIREMENTS
| |
| ..................................... NOTES- --
| |
| : 1. Refer to Table 3.3.7.1-1 to determine which SRs apply for each CREF Function.
| |
| : 2. For Functions 1, 2,. and 3. when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to
| |
| : 6. hours provided the associated Function maintains CREF initiation I capability.
| |
| I SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 12 hours l
| |
| SR '3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. 31 days
| |
| , ' SR 3.3.7.1.3 Perform CHANNEL FUNCTIONAL TEST. 92 days SR' 3.3.7.1.4 Verify the trip unit setpoint.- 92 days SR 3.3.7.1.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.7.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months O I
| |
| -FERMI UNIT 2 3.3 69~ Amendment No. 134 i
| |
| | |
| CREF System Instrumentation 3.3.7.1 7m \
| |
| ! Table 3.3.7.1 1 (page 1 of 1) '
| |
| \._) Control Room Emergency Filtration System Instrunentation l APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED StRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE
| |
| : 1. Reactor Vessel Water 1.2.3 (a) 2 B SR 3.3.7.1.1 e 103.8 inches Level - Low Low. SR 3.3.7.1.3 Level 2 SR 3.3.7.1.4 SR 3.3.7.1.5 SR 3.3.7.1.6 1
| |
| : 2. Drywell Pressure - High 1.2.3 2 B SR 3.3.7.1.1 s 1.88 psig SR 3.3.7.1.3 !
| |
| l SR 3.3.7.1.4 SR 3.3.7.1.5 SR 3.3.7.1.6
| |
| : 3. Fuel Pool Ventilation 1.2.3. 2 B SR 3.3.7.1.1 s 6 d/hr Exhaust SR 3.3.7.1.3 Radiation - High (a).(b) SR 3.3.7.1.5 SR 3.3.7.1.6
| |
| : 4. Control Center Normal 1.2.3. 1 C SR 3.3.7.1.1 s 5 d/hr Makeup Air SR 3.3.7.1.2 Radiation - High (a).(b) SR 3.3.7.1.5 I '
| |
| V<
| |
| (a) During operations with a potential for draining the reactor vessel.
| |
| (b) During CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment.
| |
| /*\
| |
| V FERMI - UNIT 2 3.3 70 Amendment No. 134 .
| |
| l
| |
| | |
| LOP Instrumentation 3.3.8.1 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LC0 3.3.8.1 The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| When the associated emergency diesel generator (EDG) is required to be OPERABLE by LCO 3.8.2. "AC Sources -Shutdown. "
| |
| ACTIONS
| |
| ..................................... NOTE---- -- ------- -- - -- - --- ---
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more buses with A.1 Restore channel to 72 hours one or more channels OPERABLE status, inoperable.
| |
| B. Required Action and B.1 Declare associated Immediately associated Completion EDG inoperable.
| |
| Time of Condition A not met.
| |
| M One or more buses with LOP trip capability not maintained.
| |
| O.t V
| |
| FERMI UNIT 2 3.3-71 Amendment No. 134
| |
| | |
| i LOP Instrumentation j 3.3.8.1 l i
| |
| O suave > uaace aeouineae 1s
| |
| .....................................N0TE--
| |
| Refer to Table 3.3.8.11 to determine which SRs apply for each LOP Function.
| |
| SURVEILLANCE FREQUENCY SR 3.3.8.1.1 Perform CHANNEL FUNCTIONAL TEST. 31 days i
| |
| l l SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. 18 months l
| |
| l SR 3.3.8.1.3 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months i
| |
| l l
| |
| 1 i
| |
| l l
| |
| t
| |
| )
| |
| I
| |
| )
| |
| l l
| |
| l l
| |
| .O FERMI UNIT 2 3.3 72 Amendment No. 134 I
| |
| | |
| r LOP Instrumentation 3.3.8.1 Table 3.3.8.1 1 (page 1 of 1)
| |
| Loss of Power Instrumentation REQUIRED CHANNELS SLRVEILLANCE FUNCTION PER SUS REQUIREENTS ALLDWA8LE VALUE
| |
| : 1. 4.16 kV Emergency Bus Undervoltage-(Loss of Voltage)
| |
| : a. Bus Undervoltage 4 SR 3.3.8.1.1 (a)
| |
| SR 3.3.8.1.2 SR 3.3.8.1.3
| |
| : b. Time Delay 4 SR 3.3.8.1.1 (b)
| |
| SR 3.3.8.1.2 SR 3.3.8.1.3
| |
| : 2. 4.16 kV Emergency Bus undervoltage (Degraded Voltage)
| |
| : a. Bus Undervoltage 4 SR 3.3.8.1.1 (c)
| |
| SR 3.3.8.1.2 SR 3.3.8.1.3
| |
| : b. Time Delay 4 SR 3.3.8.1.1 (d)
| |
| SR 3.3.8.1.2 SR 3.3.8.1.3 (a) Division I: = 2972.3 V and s 3093.7 V Divi: ion II: a 3016.4 V and a 3139.6 V (b) Division I: a 1.9 see and s 2.1 see Division II: = 1.9 see and a 2.1 see (c) Division I: a 3873.0 V and s 4031.0 V Division II: a 3628.0 V and s 3776.0 V (d) Division I: e 41.8 see and s 46.2 see
| |
| - Division II: a 20.33 sec and s 22.47 see i
| |
| i I
| |
| i FERMI - UNIT 2 3.3-73 Amendment No. 134
| |
| | |
| . 1 1
| |
| RPS Electric Power Monitoring 3.3.8.2
| |
| ( 3.3 INSTRUMENTATION 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring LC0 3.3.8.2 Two RPS electric power monitoring assemblies shall be -
| |
| OPERABLE for each inservice RPS motor generator set or alternate power supply.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| MODES 4 and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, or with both residual heat removal shutdown cooling (RHR SDC) isolation valves open.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or both inservice A.1 Remove associated 72 hours power supplies with inservice power O-V one electric power monitoring assembly supply (s) from service.
| |
| inoperable.
| |
| B. One or both inservice B.1 Remove associated 1 hour
| |
| )ower supplies with inservice power
| |
| )oth electric power supply (s) from monitoring assemblies service, inoperable.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met in MODE 1. 2. or 3. C.2 Be in MODE 4. 36 hours (continued)
| |
| (~
| |
| C)/
| |
| FERMI UNIT 2 3.3 74 Amendment No. 134 1
| |
| | |
| I i
| |
| RPS Electric Power Monitoring 3.3.8.2 O actions (ce"t4#"ee>
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate action to Immediately.
| |
| associated Completion fully insert all Time of Condition A insertable control or B not met in MODE 4 rods in core cells or 5 with any control containing one or rod withdrawn from a more fuel assemblies.
| |
| core cell containing one or more fuel egl assemblies or with both RHR-SDC isolation D.2.1 Initiate action to Immediately valves open. restore one electric power monitoring assembly to OPERABLE status for inservice power supply (s) supplying required instrumentation.
| |
| E D.2.2 Initiate action to Immediately
| |
| ( '
| |
| isolate the Residual Heat Rernoval Shutdown Cooling System.
| |
| _ l l
| |
| l O
| |
| FERMI UNIT 2 3.3 75 Amendment No. 134
| |
| | |
| F i
| |
| 1 t
| |
| l RPS Electric Power Monitoring
| |
| < 3.3.8.2 L
| |
| i l[ SURVEILLANCE REQUIREMENTS _
| |
| : SURVEILLANCE FREQUENCY SR 3.3.8.2.1 -- ---- ---- -- NOTE---- -- - ---- ---
| |
| Only required to be performed prior to entering MODE 2 or 3 from MODE 4 when in MODE 4 for = 24 hours.
| |
| l .........................................
| |
| Perform CHANNEL FUNCTICNAL TEST. 184 days l
| |
| l l SR 3.3.8.2.2 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:
| |
| : a. Overvoltage 5 132 V.
| |
| : b. Undervoltage = 108 V.
| |
| : c. Underfrequency = 57 Hz.
| |
| O t
| |
| G SR 3.3.8.2.3 Perform a system functional test. 18 months l
| |
| l l
| |
| l i
| |
| i O
| |
| FERMI - UNIT 2 3.3 76 Amendment No. 134
| |
| | |
| Recirculation Loops Operating 3.4.1
| |
| ( 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| '3.4.1 Recirculation Loops Operating LC0 3.4.1 The reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or " Exit" Regions.
| |
| NG
| |
| : a. Two recirculation loops with matched recirculation loop jet pump flows shall be in operation:
| |
| : b. One recirculation loop may be in operation provided the following limits are applied when the associated LC0 is applicable:
| |
| : 1. LCO 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)." single loop operation limits I specified in the COLR:
| |
| : 2. LC0 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)."
| |
| single loop operation limits specified in the COLR:
| |
| : 3. LC0 3.3.1.1. " Reactor Protection System (RPS)
| |
| Instrumentation." Function 2.b (Average Power Range Monitors Simulated Thermal Power-Upscale) Allowable Value of Table 3.3.1.1-1 is reset for single loop operation, when in MODE 1: and
| |
| : 4. THERMAL POWER is s 67.2% RTP.
| |
| ............................N0TE-- - ---- -- --- -- -----
| |
| Application of the required limitations for single loop operation may be delayed for up to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. l I
| |
| APPLICABILITY: MODES 1 and 2. I i
| |
| i l
| |
| o i
| |
| '9) l FERMI UNIT 2 3.4 1 Amendment No. 134 i
| |
| | |
| Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| A. Recirculation jet pump A.1 Declare recirculation 2 hours j
| |
| loop flow mismatch not loop with lower flow: l within limits. "not in operation." l i
| |
| B. Reactor core operating ----
| |
| ---NOTE---- ---- ---
| |
| l in the " Exit" Region. Restart of an idle i recirculation loop or l resetting a recirculation j flow limiter is not allowed.
| |
| l B.1 Initiate action to Immediately insert control rods or increase core flow I to restore operation I outside the " Exit" Region.
| |
| C. No recirculation loops C.1 Be in MODE 3. 6 hours operating while in MODE 2.
| |
| (continued) l t
| |
| FERMI UNIT 2 3.4-2 Amendment No. 134
| |
| | |
| Recirculation Loops Operating 3.4.1 i
| |
| ACTIONS (continued) t] COMPLETION TIME l CONDITION REQUIRED ACTION
| |
| : 0. No recirculation loops D.1 Place the reactor Immediately.
| |
| operating while in mode switch in the MODE 1. shutdown position.
| |
| Reactor core operating in the " Scram" Region.
| |
| E Core thermal hydraulic j
| |
| instability evidenced.
| |
| SURVEILLANCE REQUIREMENTS
| |
| ) SURVEILLANCE FREQUENCY SR 3.4.1.1 - - - -
| |
| -NOTE- - -- -
| |
| Only required to be performed when operating in the " Stability Awareness" Region.
| |
| Verify the reactor core is not exhibiting 1 hour core thermal hydraulic instability.
| |
| (continued)
| |
| ~
| |
| FERMI UNIT 2 3.4 3 Amendment No. 134 I
| |
| i
| |
| | |
| Recirculation Loops Operating 3.4.1 O sVRvE1't>NCe aEouraE N1s <comt4nued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.1.2 ...-.........-......N0TE... ... .- -...--
| |
| Not required to be performed until 24 hours after both recirculation loops are in operation.
| |
| Verify recirculation loop jet pump flow 24 hours mismatch with both recirculation loops in operation is:
| |
| : a. s 10% of rated core flow when operating at < 70% of rated core flow; and
| |
| : b. .s 5% of rated core flow when operating at = 70% of rated core flow.
| |
| O 1
| |
| l o
| |
| FERMI UNIT 2 3.4 4 Amendment No. 134
| |
| | |
| Jet Pumps 3.4.2 3.4 REACTOR _' COOLANT SYSTEM (RCS) 3.4.2- Jet Pumps LCO' 3.4.2 , All. jet pumps shall be OPERABLE. -
| |
| APPLICABILITY: -MODES 1 and 2. ,
| |
| ACTIONS C0lOITION. REQUIRED ACTION COMPLETION TIME 1
| |
| A. One or more jet pumps A.1 Be in MODE 3. 12 hours <
| |
| inoperable. l I
| |
| i O
| |
| FERMI UNIT 2 3.4 5 ' Amendment No. 134
| |
| | |
| m Jet Pumps 3.4.2 SURVEILLANCE REQUIREMENTS _
| |
| SURVEILLANCE FREQUENCY )
| |
| l SR 3.4.2.1 ...............- N0TES- -- --- ------ - '
| |
| : l. Not required to be performed until.
| |
| 4 hours after associated recirculation loop is in operation. .
| |
| : 2. Not required to be performed until 24 hours after > 25% RTP.
| |
| Verify at least two of the following 24 hours criteria (a. b or c) are satisfied for each operating recirculation loop: ,
| |
| : a. Recirculation loop drive flow versus recirculation pump speed differs by s 10% from established patterns.
| |
| : b. Recirculation loop drive flow versus total core flow differs by 510% from established patterns.
| |
| t
| |
| : c. Each jet pu p diffuser to lower plenum differentia )ressure differs by 5 20%
| |
| from establisled patterns. or each jet pump flow differs by s 10t from established patterns. ,
| |
| )
| |
| 1 O !
| |
| FERMI UNIT 2 3.4 6 Amendment No. 134
| |
| | |
| SRVs 3.4.3 3.4 - REACTOR COOLANT SYSTEM (RCS)-
| |
| 3.4.3 Safety Relief Valves (SRVs)
| |
| ' LC0 3.4.3 The. safety function of 11 SRVs shall be OPERABLE. -
| |
| APPLICABILITY: MODES 1. 2..and 3.
| |
| I ACTIONS C0lOITION REQUIRED ACTION COMPLETION TIME l
| |
| A. One or more required A.1 Be in MODE 3. 12 hours SRVs inoperable.
| |
| AND A.2 Be in MODE 4. 36 hours O
| |
| I I
| |
| O FERMI - UNIT 2 3.4 7 Amendment No. 134 l
| |
| | |
| SRVs 3.4.3 4
| |
| O suavetuaace aeouiaeaeats SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift'setpoints In accordance of the required SRVs are as follows: with the Inservice Number of- Setpoint Testing Program SRVs (osia) .
| |
| 5 1135
| |
| * 34.05 5 1145
| |
| * 34.35 5 1155 i 34.65 Following testing, lift settings shall be
| |
| .within i it.
| |
| SR 3.4.3.2 - ----
| |
| - ------ NOTE---- -- -- --- -
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify each required SRV opens when 18 months manually actuated. j i
| |
| i
| |
| ~O ;
| |
| i FERMI UNIT 2 3.4 8 Amendment No. 134
| |
| | |
| RCS Operational LEAKAGE 3.4.4
| |
| / 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO '3.4.4 RCS operational LEAKAGE shall be limited to: -
| |
| : a. No pressure boundary LEAKAGE:
| |
| : b. s 5 gpm unidentified LEAKAGE:
| |
| : c. s 25 gpm total LEAKAGE averaged over the previous 24 hour period: and
| |
| : d. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.
| |
| APPLICABILITY: MODES 1. 2. and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to 4 hours not within limit. within limits.
| |
| 2 Total LEAKAGE not within limit.
| |
| (continued)
| |
| .\
| |
| FERMI UNIT 2 3.4 9 Amendment No. 134
| |
| | |
| RCS Operational LEAKAGE 1 3.4.4 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i
| |
| B. Unidentified LEAKAGE B.1 Reduce LEAKAGE to 4 hours !
| |
| increase not within within limits. '
| |
| limit.
| |
| 0B B.2 Verify source of 4 hours unidentified LEAKAGE increase is not service sensitive type 304 or type 316 )
| |
| austenitic stainless steel.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours !
| |
| associated Completion Time of Condition A AND or B not met.
| |
| g . C.2 Be in MODE 4. 36 hours V
| |
| Pressure boundary j LEAKAGE exists, l
| |
| SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE 8 hours and unidentified LEAKAGE increase are within limits.
| |
| t' FERMI - UNIT 2 3.4-10 Amendment No. 134
| |
| | |
| RCS PIV Leakage 3.4.5 l l
| |
| 3.4 REACTOR'C00LANT SYSTEM (RCS) !
| |
| 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage LC0 3.4.5 The leakage from each RCS PIV shall be within limit. -
| |
| l APPLICASILITY: MODES I and 2.
| |
| MODE 3, except valves in the residual heat removal (RHR) ,
| |
| shutdown cooling flow {
| |
| transition to or from, path when in, or the shutdown during cooling the of mode ;
| |
| operation. '
| |
| I ACTIONS
| |
| .....................................N0TES - - ---- --- -- -- - ---- ---- --
| |
| : 1. Separate Condition entry is allowed for each flow path.
| |
| : 2. Enter applicable Conditions and Required Actions for systems made inoperable by PIVs.
| |
| O CONDITION REQUIRED ACTION COMPLETION TIME A. One or more flow paths --- - .---- NOTE - ---- -- --
| |
| with leakage from one Each check valve used to or more RCS PIVs not satisfy Required Action A.1 within limit. must have been verified to meet SR 3.4.5.1 at the last refueling outage or after the last time the valve was disturbed, whichever is more recent.
| |
| ............................. J A.1 Isolate the high 4 hours pressure portion of the affected system from the low pressure portion by use of one other closed manual. j de activated J automatic. or check '
| |
| valve. !
| |
| /G (continued)
| |
| O l
| |
| FERMI - UNIT'2 3.4 11 Amendment No. 134 l
| |
| | |
| RCS PIV Leakage 3.4.5
| |
| /'"x ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours l
| |
| l
| |
| {
| |
| SURVEILLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY SR 3.4.5.1 -------
| |
| -----------NOTE --- -- ---- - ----
| |
| Not required to be performed in MODE 3.
| |
| Verify equivalent leakage of each RCS PIV, In accordance at an RCS pressure a 1035 and s 1055 psig: with the g')
| |
| i
| |
| \'
| |
| Inservice
| |
| : a. For PIVs other than LPCI loop A and B Testing Program injection isolation valves is s 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm: ,
| |
| i
| |
| : b. For LPCI loop A and B outboard injection isolation valves is s 0.4 gpm through-seat, and 5 5 ml/ min external leakage: and
| |
| : c. For LPCI loop A and B inboard injection isolation testable check l
| |
| valves is s 10 gpm. l j
| |
| FERMI - UNIT 2 3.4 12 Amendment No. 134
| |
| | |
| 4 RCS Leakage Detection Instrumentation 3.4.6 i
| |
| 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Leakage Detection Instrumentation LC0 3.4.6 The following RCS leakage detection instrumentation shall be OPERABLE:
| |
| : a. Drywell floor drain sump flow monitoring syster;
| |
| : b. The primary containment atmosphere gaseous radioactivity monitoring system channel: and
| |
| : c. Drywell floor drain sump level monitoring system.
| |
| APPLICABILITY: MODES 1, 2. and 3.
| |
| ACTIONS
| |
| .................................. NOTE----- -- ----- ----------- ----- ---
| |
| LC0 3.0.4 is not applicable. ,
| |
| O CONDITION I REQUIRED ACTION COMPLETION TIME i
| |
| l A. Drywell floor drain A.1 Restore drywell floor. 30 days sump flow monitoring drain sump flow system inoperable, monitoring system to OPERABLE status.
| |
| B. Required primary B.1 Analyze grab samples Once per i containment atmosphere of primary 24 hours :
| |
| gaseous radioactivity containment I monitoring system atmosphere. l inoperable.
| |
| (continued)
| |
| C v
| |
| FERMI UNIT 2 3.4-13 Amendment No. 134
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.6 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell floor drain C.1 ....... NOTE----- -- -
| |
| sump level monitoring Not applicable when system inoperable. primary containment atmosphere gaseous radioactivity monitoring system is inoperable.
| |
| Perform SR 3.4.6.1. Once per 8 hours .
| |
| D. Primary containment D.1 Restore primary 30 days atmosphere gaseous containment I radioactivity atmosphere gaseous '
| |
| monitoring system radioactivity inoperable. monitoring system to OPERABLE status.
| |
| 8!Q O
| |
| V M Drywell floor drain sump level monitoring D.2 Restore drywell floor 30 days system inoperable. drain sump level monitoring system to OPERABLE status.
| |
| E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A. AND B. C. or D not met.
| |
| E.2 Be in MODE 4. 36 hours F. All required leakage F.1 Enter LC0 3.0.3. Immediately detection systems inoperable.
| |
| FERMI - UNIT 2 3.4 14 Amendment No. 134 l l
| |
| )
| |
| | |
| RCS Leakage Detection Instrumentation 3.4.6
| |
| ] SURVEILLANCE' REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Perform a CHANNEL CHECK of required primary 12 hours -
| |
| containment atmosphere gaseous radioactivity monitoring system.
| |
| SR 3.4.6.2 Perform a CHANNEL FUNCTIONAL TEST of 31 days required leakage detection instrumentation.
| |
| SR 3.4.6.3 Perform a CHANNEL CALIBRATION of required 18 months leakage detection instrumentation.
| |
| .)
| |
| l 0
| |
| FERMI UNIT 2 3.4-15 Amendment No. 134
| |
| | |
| 1 RCS Specific Activity -
| |
| 3.4.7 O 2.4 atacToa coo'aa' svsrea <acs) 3.4.7 RCS Specific Activity LC0 3.4.7 The specific activity of the reactor coolant shall be -
| |
| limited to DOSE EQUIVALENT I 131 specific activity 5 0.2 yC1/gm.
| |
| APPLICABILITY: MODE 1.
| |
| MODES 2 and 3 with any main steam line not isolated.
| |
| 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor coolant --
| |
| ---- - -- NOTE------ - -
| |
| specific activity LC0 3.0.4 is not applicable.
| |
| > 0.2 pCilgm and ---------------- -- ------ -
| |
| s 4.0 Ci/gm DOSE EQUIVALENT I 131. A.1 Determine DOSE Once per 4 hours EQUIVALENT I 131.
| |
| O ,
| |
| A.2 Restore DOSE 48 hours f EQUIVALENT I-131 to within limits.
| |
| (continued) 1
| |
| .O FERMI UNIT 2 3.4 16 Amendment No. 134
| |
| | |
| RCS Specific Activity 3.4.7 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Determine DOSE Once per 4 hours associated Completion EQUIVALENT I 131.
| |
| Time of Condition A not met. 6@
| |
| E B.2.1 Isolate all main 12 hours steam lines.
| |
| Reactor Coolant specific activity 3
| |
| > 4.0 pCi/gm Dose EQUIVALENT I-131. B.2.2.1 Be in MODE 3. 12 hours AND B.2.2.2 Be in MODE 4. 36 hours
| |
| , SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 --- --- -
| |
| ---- - NOTE---- - - ----- ---
| |
| Only required to be performed in MODE 1.
| |
| Verify reactor coolant DOSE EQUIVALENT 7 days I 131 specific activity is s 0.2 Ci/gm.
| |
| b FERMI - UNIT 2 3.4 17 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown 3.4.8 3.4 REACTOR COOLANT' SYSTEM (RCS) 3.4.8~ Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.
| |
| ............................N0TES- - - -- - --
| |
| : 1. Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours per 8 hour period.
| |
| : 2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours for the performance of Surveillances.
| |
| APPLICABILITY: MODE 3, with reactor steam dome pressure less than the RHR cut in permissive pressure.
| |
| ACTIONS
| |
| .....................................N0TES - - - - -
| |
| : 1. LC0 3.0.4 is not applicable.
| |
| : 2. Separate Condition entry is allowed for each RHR shutdown cooling subsystem.
| |
| REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Initiate action to Immediately RHR shutdown cooling restore required RHR subsystems inoperable. shutdown cooling subsystem (s) to OPERABLE status, l
| |
| AND l 1
| |
| (continued) '
| |
| I O
| |
| FERMI - UNIT 2 3.4 18 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown
| |
| ; 3.4.8 ACTIONS 4
| |
| CONDITION REQUIRED ACTION COMPLETION TIME f
| |
| A. (continued) A.2 Verify an alternate 1 hour method of decay heat removal is available for each inoperable RHR shutdown cooling subsystem. l AND A.3 Be in MODE 4. 24 hours I
| |
| B. No RHR shutdown B.1 Initiate action to Immediately cooling subsystem in restore one RHR operation. shutdown cooling I subsystem or one l AND recirculation pump to operation.
| |
| f No recirculation pump
| |
| ; in operation. AND B.2 Verify reactor 1 hour from coolant circulation discovery of no by an alternate reactor coolant method. circulation AND B.3 Monitor reactor Once per hour coolant temperature and pressure.
| |
| ; GL)
| |
| FERMI UNIT 2 3.4-19 Amendment No. 134 L'
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown 3.4.8
| |
| -s (j SURVEILLANCE REQUIREMENTS 5URVEILLANLL FREQUENCY SR 3.4.8.1 - - -- -
| |
| - - - NOTE --- ----- - --
| |
| Not required to be met until 4 hours after reactor steam dome pressure is less than the RHR cut 1n permissive pressure.
| |
| Verify one RHR shutdown cooling subsystem 12 hours or recirculation pump is operating.
| |
| ,/ \
| |
| L) 1 l
| |
| l LJ FERMI UNIT 2 3.4 20 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Cold Shutdown 3.4.9 O 34 atactoa coo'^at svs't" <acs) 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown LC0 3.4.9 Two RHR shutdown cooling subsystems shall be OPERABLE, and, '
| |
| with no recirculation pump in operation at least one RHR shutdown cooling subsystem shall be in operation.
| |
| ...........................N0TES-------- --- ------ ----
| |
| : 1. Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours per 8 hour period.
| |
| : 2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours for the performance of Surveillances.
| |
| APPLICABILITY: MODE 4.
| |
| ............................N0TE- --- --- -- --- - - --
| |
| Not applicable when heat losses to the ambient are greater than or equal to beat input to the reactor coolant.
| |
| ACTIONS
| |
| .................................... NOTE-- ------ --- - - - ----- ---------
| |
| . Separate Condition entry is allowed for each shutdown cooling subsystem.
| |
| i i
| |
| CONDITION REQUIRED ACTION COMPLETION TIME j A. One or two required A.1 Verify an alternate I hour RHR shutdown cooling method of decay heat subsystems inoperable. removal is available AND for each inoperable ,
| |
| RHR shutdown cooling Once per subsystem. 24 hours thereafter (continued)
| |
| O FERMI UNIT 2 3.4 21 Amendment No. 134
| |
| | |
| l RHR Shutdown Cooling System-Cold Shutdown 3.4.9
| |
| - ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown B.1 Initiate action to Immediately cooling subsystem in restore one RHR operation. shutdown cooling subsystem or one 8ND recirculation pump to operation.
| |
| No recirculation pump in operation. AtiD B.2 Verify reactor 1 hour from coolant circulating discovery of no by an alternate reactor coolant method. circulation AND B.3 Monitor reactor Once per hour coolant temperature.
| |
| O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify one RHR shutdown cooling subsystem 12 hours or recirculation pump is operating. I l
| |
| l l
| |
| FERMI UNIT 2 3.4 22 Amendment No. 134
| |
| | |
| RCS P/T Limits 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 RCS Pressure and Temperature (P/T) Limits LC0 3.4.10 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits.
| |
| APPLICABILITY: At all times.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---
| |
| NOTE -------- A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits.
| |
| shall be completed if this Condition is ANQ entered.
| |
| A.2 Determine RCS is 72 hours O acceptable for D Requirements of the - continued operation.
| |
| LC0 not met in MODES 1, 2, and 3.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A N 6_NQ not met.
| |
| B.2 Be in MODE 4. 36 hours (continued) i FERMI UNIT 2 3.4 23 Amendment No. 134
| |
| | |
| RCS P/T Limits 3.4.10 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| C. -----
| |
| - NOTE- ------- C.1 Initiate action to Immediately-Required Action C.2 restore parameter (s) shall be completed if to within limits.
| |
| this Condition is I entered. AND i C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 i LC0 not met in other operation, or 3 i than MODES 1. 2 and 3.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 ---- ----- -
| |
| --NOTE------ --- ---
| |
| Only required to be performed as applicable during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
| |
| Verify: 30 minutes
| |
| : a. RCS pressure and RCS temperature are to the right of the limits specified in Figure 3.4.10 1: and
| |
| : b. RCS heatup and cooldown rates are limited to:
| |
| : 1. s 100*F in any 1 hour period: and j
| |
| : 2. s 20*F in any 1 hour period during inservice hydrostatic and leak testing operations above the 4 heatup and cooldown limit curves.
| |
| (continued)
| |
| O O
| |
| FERMI - UNIT 2 3.4 24 Amendment No. 134
| |
| | |
| 1 RCS P/T Limits 3.4.10
| |
| ( ,
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes Figure 3.4.10 1. prior to control rod withdrawal for )
| |
| the purpose of achieving ,
| |
| criticality I i
| |
| i SR 3.4.10.3 ----- ---
| |
| - NOTE--- -- ----------- l Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
| |
| Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) steam space coolant prior to each temperature is s 145*F. startup of a recirculation pump SR 3.4.10.4 - --- - --
| |
| ---NOTE----- - --- ----
| |
| Only required to be met in MODES 1, 2. 3, and 4 during recirculation pump startup.
| |
| Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes i loop to be started and the RPV coolant prior to each l temperature is s 50"F. startup of a i recirculation {
| |
| pump
| |
| )
| |
| (continued) ,
| |
| i v
| |
| FERMI UNIT 2 3.4 25 Amendment No. 134 l
| |
| | |
| RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.5 --.- - ------ - - NOTE------------- -- - -
| |
| Only required to be met during a THERMAL POWER increase or recirculation flow increase in MODES 1 and 2 with one idle recirculation loop when THERMAL POWER is s 30t RTP or when operating loop flow is s 50% rated loop flow.
| |
| Verify the difference between the bottom Once within head coolant temperature and the RPV steam 15 minutes space coolant temperature is s 145'F. prior to a THERMAL POWER increase or recirculation flow increase 1
| |
| SR 3.4.10.6 - ---- - ---- ---
| |
| NOTE----------- --- - -
| |
| Only required to be met during a THERMAL POWER increase or recirculation flow increase in MODES 1 and 2 with one non.
| |
| isolated idle recirculation loo) when )
| |
| THERMAL POWER is s 30% RTP or w1en operating loop flow is s 50% rated loop flow.
| |
| Verify the difference between the reactor Once within coolant temperature in the idle 15 minutes ,
| |
| recirculation loop and the RPV coolant prior to a temperature is s 50*F.
| |
| THERMAL POWER increase or recirculation flow increase (continued)
| |
| /%
| |
| U FERMI - UNIT 2 3.4-26 Amendment No. 134
| |
| | |
| RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.4.10.7 --.---.....-- ---- NOTE -- - ------ - ----
| |
| Only required to be performed when tensioning the reactor vessel head bolting studs.
| |
| Verify reactor vessel flange and head 30 minutes flange temperatures are a 71*F when the reactor vessel head bolt st9ds are under
| |
| . tension.
| |
| SR 3.4.10.8 ----- -
| |
| ---NOTE- - ---- -- ---- -
| |
| Not required to be performed until 30 minutes after RCS temperature s 80*F in MODE 4.
| |
| Verify reactor vessel flange and head 30 minutes f9 V
| |
| flange temperatures are a 71*F.
| |
| SR 3.4.10.9 ---- - -- --
| |
| --- NOTE -- -- -- -- ---
| |
| Not required to be performed until 12 hours after RCS temperature s 100"F in MODE 4.
| |
| Verify reactor vessel flange and head 12 hours I flange temperatures are a 71*F.
| |
| FERMI UNIT 2 3.4 27 Amendment No. 134
| |
| | |
| . RCS P/T Limits 3.4.10 1600 O
| |
| V I
| |
| uoo -
| |
| / A I
| |
| P A' 8'& C' I - ,
| |
| f f I ii I 'l I
| |
| =
| |
| ? I I
| |
| $ 1o00 ,
| |
| l !
| |
| j 5 J I N
| |
| e i i E
| |
| ya l i e A',F,C'.COREBELTUNE AFTER E j ASStad!D 11cF SWT FROM AN j
| |
| NTIAL WELD RTpey OF -erF -
| |
| @ Goo ) A . SYSTEM HYDROTEST LMTWTH F12L N VESSEL E
| |
| B . NON-NUCLEAR EATLF/ _
| |
| COOLDOM LMT g
| |
| 101*F l/ l C . NUCLEAR (CORE OtmCAL)LMTS _
| |
| l -VESSEL DISCONTNUITY LMTS
| |
| ~
| |
| 2oe - E
| |
| / ;
| |
| --CORE BELTLNE WTH 11rF
| |
| ~
| |
| 7 32 EFPY CURVES A', B', C'NOT -
| |
| l !
| |
| LMTNG. NFORMATION CN.Y
| |
| ' l l CURVES A, B, C ARE VALD FOR 32
| |
| [ EFPY OF OPERATION
| |
| , I i e i e i i e too 2co soo ano soo soo 64N4WUW REAC704 VESSEL hETAL TEMPERATURE ('F)
| |
| .,, FIGURE 3.4.10-1 HINIMUM REACTOR PRESSURE VESSEL METAL T'EMDERATURE VS. R FERMI UNIT 2 3.4-28 Amencment No. 134
| |
| | |
| Reactor Steam Dome Pressure 3.4.11 O 3.4 aeacToa coo'^at svs'ea (acs)
| |
| '3.4.11 Reactor Steam Dome Pressure LC0 3.4.11' The reactor steam dome pressure shall be s 1045 psig. -
| |
| -APPLICABILITY: MODES 1 and 2.
| |
| 1 ACTIONS-COM)ITION REQUIRED ACTION COMPLETION TIME
| |
| 'A. Reactor steam. dome A.1 Restore reactor steam 15 minutes pressure not within dome ressure to limit. withi limit.
| |
| 1 B. ~
| |
| Required Action and B.1 Be in MODE 3. 12 hours
| |
| : g. associated Completion Time not met.
| |
| JJRVEILLANCE REQUIREMENTS' SURVEILLANCE FREQUENCY SR '3.4.11.1 Verify reactor. steam dome pressure is 12 hours ,
| |
| s 1045 psig. l i
| |
| l i
| |
| O FERMI ! UNIT 2 3.4 29 Amendment No. 134
| |
| | |
| ECCS -Operating 3.5.1 0 3.5 eseaceacv cone coo'taa SvSreas <eces) ^~o aeacToa coas iso'^Tio" coo'1"o (RCIC) SYSTEM 3.5.1 ECCS-Operating LCO 3.5.1 Each ECCS injection / spray subsystem and the Automatic Depressurization System (ADS) function of five safety / relief valves shall be OPERABLE.
| |
| APPLICABILITY: MODE 1.
| |
| MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection / spray ECCS injection / spray subsystem inoperable. subsystem to OPERABLE O' status.
| |
| B. One LPCI pump in both B.1 Restore both LPCI 7 days LPCI subsystems pumps to OPERABLE inoperable. status.
| |
| C. One CSS subsystem C.1 Restore CSS subsystem 72 hours inoperable. to OPERABLE status.
| |
| b_ND - QB One LPCI subsystem C.2 Restore LPCI 72 hours inoperable. subsystem to OPERABLE status.
| |
| (continued)
| |
| O FERMI UNIT 2 3.5 1 Amendment No. 134
| |
| | |
| ECCS-Operating 3.5.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. ' Required Action'and- D.1 Be in MODE 3. 12 hours -
| |
| associated Completion Time of Condition A. E B, or C not met.
| |
| D.2 Be in MODE 4. 36 hours E. HPCI System E.1 Verify by Immediately inoperable, administrative means RCIC System is OPERABLE.
| |
| M E.2 Restore HPCI System 14 days to OPERABLE status.
| |
| 'F. HPCI System F.1 Restore HPCI System 72 hours inoperable.
| |
| to OPERABLE status.
| |
| Condition A.~or F.2 Restore low pressure 72 hours Condition B. or ECCSinjection/ spray Condition C entered. subsystem (s) to OPERABLE status.
| |
| G. One ADS ~ valve G.1 Restore ADS valve to 14 days
| |
| -inoperable. OPERABLE status.
| |
| (continued)
| |
| FERMI UNIT 2 3.5 2 Amendment No. 134 I
| |
| | |
| ECCS-Operating 3.5.1
| |
| -i ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME 1
| |
| H. One ADS valve H.1 Restore ADS valve to 72 hours -
| |
| inoperable. OPERABLE status. ,
| |
| i 8NQ @ l l'
| |
| Condition A or H.2 Restore low pressure 72 hours Condition B entered. ECCSinjection/ spray subsystem (s) to OPERABLE status. 1 I. Two or more ADS valves I.1 Be in MODE 3. 12 hours inoperable.
| |
| AND- l 2 l I.2 Reduce reactor steam 36 hours ]
| |
| Required Action and dome pressure to I associated Completion s 150 psig. l Time of Condition E. I F. G. or H not met.
| |
| J. Two or more low J.1 Enter LC0 3.0.3. Immediately pressure ECCS injection / spray subsystems ino)erable for reasons otler than Condition B or C.
| |
| 2 HPCI System and one or more ADS valves inoperable.
| |
| Condition C and Condition G entered.
| |
| 1 O l FERMI UNIT 2 3.5 3 Amendment No. 134
| |
| | |
| ECCS-Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify correct voltage and breaker 7 days -
| |
| alignment to the LPCI swing bus.
| |
| SR 3.5.1.2 ----- ----
| |
| ------ NOTE- - -- - ------- -
| |
| When LPCI is placed in an ino erable status solely for performance of thi SR, or when the LPCI swing bus automatic throwaver scheme is inoperable due to EDG-12 being paralleled to the bus for required testing, entry into associated Conditions and Required Actions may be delayed up to 12 hours for completion of the required testing.
| |
| Perform a functional test of the LPCI swing 31 days bus automatic throwaver scheme.
| |
| O V
| |
| SR 3.5.1.3 Verify. for each ECCS injection / spray 31 days subsystem, the piping is filled with water from the pump discharge valve to the ,
| |
| injection valve.
| |
| (continued) l l
| |
| 1 l
| |
| O FERMI UNIT 2 3.5 4 Amendment No. 134
| |
| | |
| . 1 L
| |
| ECCS-Operating 3.5.1 l l
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.4 - - --- -- -
| |
| -- NOTE -- -- ------ -
| |
| * Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the Residual Heat Removal (RHR) cut in permissive pressure in MODE 3. and for 4 hours after exceeding the RHR cut-in permissive pressure in MODE 3.
| |
| if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each ECCS injection / spray subsystem 31 days ,
| |
| manual, power o>erated and automatic valve l in the flow pat 1. that is not locked, l sealed, or otherwise secured in position, !
| |
| is in the correct position. l O SR 3.5.1.5 Verify primary containment pneumatic supply pressure is a 75 psig.
| |
| 31 days SR- 3.5.1.6 Verify the RHR System power operated cross 31 days tie valve is open.
| |
| SR 3.5.1.7 Verify each recirculation pump discharge 18 months valve cycles through one complete cycle of full travel or is de energized in the closed position.
| |
| (continued)
| |
| O FERMI UNIT 2 3.5 5 Amendment No. 134
| |
| | |
| l ECCS-Operating 3.5.1 l
| |
| ) SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.8 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor Inservice pressure. Testing SYSTEM HEAD Program NO. CORRESPONDING 0F TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF Core Spray a 6350 gpm 2 = 100 psig LPCI = 10,000 gpm 1 = 20 psig SR 3.5.1.9 -- ---- --
| |
| --- NOTE- ---- -- ----
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify, with reactor pressure s 1045 and In accordance
| |
| = 945 psig, the HPCI pump can develop a with the flow rate a 5000 gpm against a system head Inservice corresponding to reactor pressure. Testing Program l
| |
| SR 3.5.1.10 ---- -
| |
| - -- NOTE--- - ---- --- -
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify, with reactor pressure s 215 psig. 18 months the HPCI pump can develop a flow rate l
| |
| = 5000 gpm against a system head !
| |
| corresponding to reactor pressure. l (continued)
| |
| {
| |
| 1-FERMI - UNIT 2 3.5 6 Amendment No. 134
| |
| | |
| E ;
| |
| i I
| |
| E ECCS- Operating !
| |
| 3.5.1 <
| |
| (~%
| |
| V SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.1.11 --- - -
| |
| - -- -- NOTE---- - --- ----- -- -
| |
| -Vessel injection / spray may be excluded.
| |
| Verify each ECCS injection / spray subsystem 18 months '
| |
| actuates on an actual or simulated automatic initiation signal.
| |
| SR 3.5.1.12 ------ - -- ---
| |
| -NOTE --- --- ------- - i Valve actuation may be excluded.
| |
| ........................................... j l
| |
| Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal.
| |
| I SR 3.5.1.13 - -----
| |
| -- ----NOTE - - - ------- ---
| |
| ( Not required to be_ performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. !
| |
| Verify each ADS valve opens when manually 18 months ,
| |
| actuated.
| |
| I SR 3.5.1.14 - ---- -
| |
| - -- NOTE - --- - ---
| |
| ECCS instrumentation response times are not required to be measured.
| |
| Verify ECCS RESPONSE TIME is within limits. 18 months i
| |
| O ,
| |
| -FERMI - UNIT 2 3.5 7 Amendment No. 134
| |
| | |
| ECCS-Shutdown 3.5.2 1
| |
| l L O 3 s taeascacv coat coo'iaa svs'tas <eccs) aao aeac'oa coat 'so'au oa COOLING-(RCIC) SYSTEM.
| |
| 3.5.2 ECCS-Shutdown l -
| |
| l LCO 3.5.2 Two low pressure ECCS injection / spray subsystems shall be OPERABLE.
| |
| 1 APPLICABILITY: MODE 4 MODE 5. except with the spent fuel storage pool gates removed and water level = 20 ft 6 inches over the top of .
| |
| the reactor pressure vessel flange.
| |
| ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME l'
| |
| i A. One reguired ECCS A.1 Restore required ECCS 4 hours injection / spray injection / spray subsystem inoperable. subsystem to OPERABLE
| |
| , status.
| |
| B. Required Action and B.1 Initiate action to Immediate1v associated Completion suspend operations Time of Condition A with a potential for not met. draining the reactor vessel (0PDRVs). {
| |
| j
| |
| )
| |
| C. Two required ECCS C.1 Initiate action to Immediately l
| |
| injection / spray suspend OPDRVs.
| |
| l subsystems inoperable.
| |
| AND C.2 Restore one ECCS 4 hours injection / spray subsystem to OPERABLE status.
| |
| i (continued) i
| |
| : FERMI UNIT 2 3.5 8 Amendment No. 134
| |
| | |
| ECCS - Shutdown 3.5.2 i A*,TIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 D.1 Initiate action to Immediately and associated restore secondary Completion Time not containment to met. OPERABLE status.
| |
| AND D.2 Initiate action to Immediately restore one standby gas treatment I subsystem to OPERABLE status. :
| |
| 6ND D.3 Initiate action to Immediately restore isolation I capability in each required secondary containment penetration flow path '
| |
| O)
| |
| (, not isolated.
| |
| )
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify. for each required low pressure 12 hours coolant injection (LPCI) subsystem, the suppression pool water level is a 66 inches.
| |
| (continued)
| |
| O FERMI UNIT 2 3.5 9 Amendment No. 134
| |
| | |
| ECCS-Shutdown i 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify, for each required core spray (CS) 12 hours .
| |
| subsystem. the:
| |
| : a. Suppression pool water level is
| |
| = -66 inches: or
| |
| : b. .................N0TE---- ----- ----
| |
| Only one required CS subsystem may take credit for this option during OPDRVs.
| |
| Condensate storage tank water level is
| |
| = 19 ft.
| |
| SR 3.5.2.3 Verify correct voltage and breaker 7 days alignment to the LPCI swing bus.
| |
| m SR 3.5.2.4 Verify, for each required ECCS injection / 31 days spray subsystem. the piping is filled with water from the pump discharge valve to the injection valve.
| |
| (continued) l FERMI UNIT 2 3.5 10 Amendment No. 134 l l l
| |
| I
| |
| ( j
| |
| | |
| ~
| |
| ECCS-Shutdown 3.5.2 O Suave' u^"c" aeoutaeaea's (co#t4##ed)
| |
| SURVEILLANCE FREQUENCY SR 3.5.2.5 ...................N0TE.-.....---.-- ---..-
| |
| LPCI subsystem (s) may be considered OPERABLE during alignment and operation for-decay heat removal if capable of being manually realigned and not otherwise inoperable.
| |
| Verify each required ECCS injection / spray 31 days
| |
| ' subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| 1 SR 3.5.2.6 Verify each required ECCS pump develops the In accordance !
| |
| specified flow rate against a system head with the l corresponding to the specified reactor Inservice pressure. Testing O NO.
| |
| OF SYSTEM HEAD CORRESPONDING TO A REACTOR Program SYSTEM FLOW RATE. PUMPS PRESSURE OF CS = 6350 gpm 2 = 100 psig LPCI = 10,000 gpm 1 = 20 psig l l
| |
| SR 3.5.2.7 ...................N0TE. --.--.. -......- !
| |
| Vessel injection / spray may be excluded.
| |
| Verify each required ECCS injection / spray 18 months subsystem actuates on an actual or simulated automatic initiation signal.
| |
| O b
| |
| FERMI . UNIT 2 3.5 11 Amendment No. 134
| |
| | |
| RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ANO REACTOR CORE ISOLATION COOLING A
| |
| (RCIC) SYSTEM 3.5.3 RCIC System LC0 3.5.3 The RCIC System shall be OPERABLE.
| |
| APPLICABILITY: MODE 1.
| |
| MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by Immediately inoperable. administrative means High Pressure Coolant Injection System is OPERABLE.
| |
| O =
| |
| A.2 Restore RCIC System 14 days to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Reduce reactor steam 36 hours dome pressure to s 150 psig.
| |
| I l
| |
| l O
| |
| V FERMI - UNIT 2 3.5 12 Amendment No. 134 s
| |
| | |
| RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS I)
| |
| SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled 31 days -
| |
| with water from the pump discharge valve to the injection valve.
| |
| SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.5.3.3 -- -- - - -- -- -NOTE------ -- -- --- --
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| O ver4rr. ~4th reector pressere and = 945 psig, the RCIC pump can develop a 1o4s as49 92 e xs flow rate m 600 gpm against a system head corresponding to reactor pressure.
| |
| SR 3.5.3.4 -- - -
| |
| -- - NOTE--- ----- - ----
| |
| Not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify with reactor pressure s 200 psig, 18 months the RCIC pump can develop a flow rate a 600 gpm against a system head corresponding to reactor pressure.
| |
| (continued)
| |
| FERMI - UNIT 2 3.5-13 Amendment No. 134
| |
| | |
| i l
| |
| RCIC System 3.5.3 l l
| |
| () SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY i
| |
| l SR 3.5.3.5 -- - -- --
| |
| - --NOTE--- --- -- - - -
| |
| Vessel injection may be excluded.
| |
| Verify the RCIC System actuates on an 18 months actual or simulated automatic initiation signal .
| |
| O l
| |
| 1 l
| |
| I i
| |
| O l
| |
| . FERMI' UNIT 2 3.5 14 . Amendment No. 134
| |
| | |
| a ,
| |
| Primary Containment 3.6.1.1
| |
| ?3.6 -CONTAINMENT SYSTEMS-3.6.1.1 Primary Containment LCO 3.6.1.1 Primary containment shall be OPERABLE. .
| |
| APPLICABILITYi MODES 1, 2. and 3.
| |
| i ACTIONS COWITION REQUIRED ACTION COMPLETION TIME
| |
| 'A. Primary containment A.1 Restore primary 1 hour inoperable. containment to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion O Time not met. egl B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREENTS' SURVEILLANCE FREQUENCY SR 3.6.1.1.1 Perform required visual examinations and In accordance I leakage rate testing except for primary with the l containment air lock testing, in Primary accordance with the Primary Containment Containment
| |
| : Leakage Rate Testing Program. Leakage Rate Testing Program (continued)
| |
| O
| |
| ' FERMI UNIT 2: 3.6 1 Amendment No. 134
| |
| | |
| Primary Containment 3.6.1.1 O suave' u^"ce aeoutae"e"'s (c "t4""ed)
| |
| SURVEILLANCE FREQUENCY SR 3.6.1.1.2 Verify drywell to suppression chamber 18 months -
| |
| differential pressure does not decrease I at a rate > 0.2 inch water gauge per AND minute tested over a 10 minute period at an initial differential pressure of ----
| |
| NOTE- -
| |
| 1 psid. Only required after two consecutive tests fail and continues until two consecutive tests pass i 9 months I i
| |
| SR 3.6.1.1.3 ----- -- -----
| |
| -NOTE---- ---- -- --- --
| |
| Only required to be performed after Q
| |
| l safety / relief valve operation with the
| |
| \s suppression chamber average water temperature a 160*F and reactor coolant system pressure > 200 psig.
| |
| Perform an external visual examination of Once prior to the suppression chamber. entry into MODE 2 or 3 from MODE 4
| |
| ~%
| |
| (O FERMI - UNIT 2 3.6-2 Amendment No. 134
| |
| | |
| Pricary Containment Air Lock 3.6.1.2 (7
| |
| v 3.6 CONTAINMENT SYSTEMS 3.6.1.2 Primary Containment Air Lock LCO 3.6.1.2 The primary containment air lock shall be OPERABLE. -
| |
| ~ APPLICABILITY: MODES 1, 2. and 3.
| |
| ACTIONS
| |
| .....................................N0TES------------
| |
| : 1. Entry and exit is permissible to perform repairs of the air lock components.
| |
| : 2. Enter applicable Conditions and Required Actions of LC0 3.6.1.1. " Primary Containment." when air lock leakage results in exceeding overall containment leakage rate acceptance criteria.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME O A. One primary - ------
| |
| -NOTES - -- ---
| |
| containment air lock 1. Required Actions A.1 door inoperable. A.2, and A.3 are not applicable if both doors in the air lock are inoperable and Condition C is entered.
| |
| : 2. Entry and exit is permissible for 7 days under administrative controls.
| |
| A.1 Verify the OPERABLE 1 hour door is closed.
| |
| 8ND
| |
| , (continued)
| |
| O
| |
| ' ..V FERMI UNIT 2 3.6 3 Amendment No. 134
| |
| | |
| E l l
| |
| Primary Containment Air Lock l 3.6.1.2 l
| |
| ACTIONS
| |
| ) _
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Lock the OPERABLE 24 hours .
| |
| door closed.
| |
| ANQ A.3 -- -- --NOTE-- ------
| |
| Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by administrative means.
| |
| Verify the OPERABLE Once per 31 days door is locked '
| |
| closed.
| |
| , O B. Primary containment ----- -
| |
| ---NOTES ----------- i air lock interlock 1. Required Actions B.1, mechanism inoperable. B.2, and B.3 are not i applicable if both doors !
| |
| in the air lock are l inoperable and Condition C is entered.
| |
| , 2. Entry into and exit from containment is permissible under the control of a dedicated individual.
| |
| 1 B.1 Verify an OPERABLE 1 hour door is closed.
| |
| AND (continued)
| |
| G
| |
| .L.)
| |
| FERMI' UNIT 2 3.6 4 Amendment No. 134 L
| |
| | |
| 1 1
| |
| Prinary Containment Air Lock l 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Lock an OPERABLE door 24 hours '
| |
| closed.
| |
| 8NQ B.3 -- NOTE --- -
| |
| Air lock doors in high radiation areas or areas with limited access due to j inerting may be verified locked closed by I administrative means.
| |
| Verify an OPERABLE Once per 31 days door is locked closed.
| |
| O C. Primary containment C.1 Initiate action to Immediately air lock inoperable evaluate primary for reasons other than containment overall Condition A or B. leakage rate per LC0 3.6.1.1. using current air lock test results.
| |
| AND C.2 Verify a door is 1 hour l closed.
| |
| AND i C3 Restore air lock to 24 hours OPERABLE status.
| |
| (continued)
| |
| O.
| |
| FERMI UNIT 2 3.6 5 Amendment No. 134
| |
| | |
| Pritary Containment Air Lock 3.6.1.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours .
| |
| associated Completion Time not met. 8NQ D.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 - ----
| |
| --- NOTES --- -- -- -
| |
| : 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
| |
| : 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1.
| |
| Perform required primary containment air In accordance lock leakage rate testing in accordance with the with the Primary Containment Leakage Rate Primary Testing Program. Containment Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary 24 months containment air lock can be opened at a time.
| |
| l 1
| |
| )
| |
| O FERMI - UNIT 2 3.6 6 Amendment No. 134
| |
| | |
| PCIVs l{ .:
| |
| 3.6.1.3 3.6 CONTAINENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs).
| |
| I
| |
| ^ LCO 3.6.1.3 Each PCIV, except reactor building to suppression chamber vacuum breakers, shall be OPERABLE. 1 APPLICABILITY:- MODES 1, 2, and 3. .
| |
| When associated ir.strumentation is required to be OPERABLE per LC0 3.3.6.1, '" Primary Containment Isolation Instrumentation."
| |
| ACTIONS i
| |
| ................................ .... NOTES -- --- - - -- - --- --- --
| |
| : 1. Penetration flow paths may be unisolated intermittently under administrative controls.
| |
| : 2. Separate Condition entry is allowed for each penetration flow path.
| |
| : 3. -Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
| |
| -(~N 4 Enter applicable Conditions and Required Actions of LC0 3.6.1.1. " Primary V Containment," when PCIV leakage results in exceeding overall containment
| |
| '. leakage rate acceptance criteria in MODES 1, 2. and 3.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. ----
| |
| NOTE - - - - A.1 Isolate the affected 4 hours except Only applicable to >enetration' flow path for main steam penetration flow paths )y use of at least line with two PCIVs. one closed and de activated E
| |
| . automatic valve, One or more closed manual valve. 8 hours for main penetration flow paths .
| |
| blind flange, or steam line with one PCIV- check valve with flow inoperable,-except due through the valve l to leakage not within secured.
| |
| limit.
| |
| E )
| |
| (continued)
| |
| .O FERMI - UNIT 2- 3.6 7 Amendment No. 134
| |
| | |
| 1 PCIVs 3.6.1.3 l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 -- -
| |
| NOTES---- - - -
| |
| : 1. Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| : 2. Isolation devices that are locked.
| |
| sealed, or otherwise secured may be verified by use of administrative means.
| |
| Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside O primary containment 8NQ Prior to entering MODE 2 or 3 from MODE 4. if primary containment was de inerted while in MODE 4. if l not performed within the previous 92 days. for isolation i devices inside primary containment ,
| |
| l (continued)
| |
| FERMI - UNIT 2 3.6 8 Amendment No. 134 I
| |
| i I
| |
| | |
| PCIVs 3.6.1.3 ;
| |
| ACTIONS (continued) l C0WITION REQUIRED ACTION COMPLETION TIME B. - --
| |
| NOTE- - --
| |
| B.1 Isolate the affected 1 hour -
| |
| Only applicable to >enetration flow path penetration flow paths )y use of-at least with two PCIVs. one closed and de-activated automatic valve, One or more- closed manual valve, penetration flow paths or blind flange.
| |
| with two PCIVs inoperable, except due to leakage not within limit.
| |
| (continued)
| |
| O l
| |
| l-O
| |
| -FERMI UNIT 2 3.6 9 Amendment No. 134 l
| |
| | |
| l PCIVs 3.6.1.3 i
| |
| ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. - -NOTE --- -
| |
| C.1 Isolate the affected 4 hours except Only applicable to >enetration flow path for excess flow penetration flow paths )y use of at least check valves with only one PCIV. one closed and (EFCVs) and de activated penetrations automatic valve. with a closed One or more closed manual valve, system penetration flow paths or blind flange, with one PCIV ANQ inoperable except due to leakage not within 72 hours for limit. EFCVs and penetrations with a closed system AND C.2 ---- NOTES -- --
| |
| : 1. Isolation devices in high radiation O
| |
| (J areas may be verified by use '
| |
| of administrative means.
| |
| : 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
| |
| Verify the affected Once per 31 days penetration flow path is isolated.
| |
| (continued) n U
| |
| FERMI UNIT 2 3.6 10 Amendment No. 134
| |
| | |
| f PCIVs 3.6.1.3
| |
| (, ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. One or more D.1 Restore leakage rates 4 hours for-penetration flow paths to within limit. leakage on with one or more PCIVs hydrostatically inoperable due to tested line secondary containment without a closed bypass leakage rate, system MSIV leakage rate, purge valve leakage ANQ rate, hydrostatically tested line leakage 4 hours for rate. or EFCV leakage secondary rate not within limit. , containment bypass leakage AND 8 hours for MSIV leakage AND 24 hours for purge valve leakage 6N_Q 72 hours for leakage on hydrostatically tested line on a closed system and EFCV leakage l
| |
| AND (continued)
| |
| {
| |
| O FERMI UNIT 2 3.6 11 Amendment No. 134
| |
| | |
| PCIVs 3.6.1.3
| |
| ~"s O COWITION REQUIRED ACTION COMPLETION TIME D.- (continued) - NOTES --- - - -
| |
| : 1. Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| : 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
| |
| : 3. Only applicable to penetration flow paths isolated to restore leakage to within limits.
| |
| D.2 Verify the affected Once per 31 days penetration flow path for isolation devices outside O
| |
| is isolated, primary z
| |
| containment j
| |
| Prior to entering MODE 2 or 3 from l MODE 4. if !
| |
| primary containment was de inerted while i in MODE 4. if not performed within the previous 92 days. for isolation devices inside primary containment (continued)
| |
| O FERMI UNIT 2 3.6 12 Amendment No. 134
| |
| | |
| PCIVs 3.6.1.3 O ac'ioas (co"t4""ea)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i
| |
| j E. Required Action and' E.1 Be in MODE 3. 12 hours -
| |
| I associated Compietion Time of Condition A, egl B. C, or D not met in MODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours F. Required Action and F.1 Initiate action to Immediately associated Completion isolate RHR Shutdown Time of Condition A, Cooling System. ,
| |
| B C. or D not met for l RHR-SDC PCIV(s) @
| |
| required to be OPERABLE during MODE 4 F.2 Initiate action to Immediately ,
| |
| or 5. restore valve (s) to !
| |
| OPERABLE status. j 1
| |
| O FERMI UNIT 2- 3.6 13 Amendment No. 134
| |
| | |
| PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 -- - .
| |
| ---- NOTE-- - -- --- -- -
| |
| Not required to be met when the isolation valves for one purge or containment pressure control supply line and one l purge or containment pressure control exhaust line are open for inerting, de-inerting, pressure control. ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.
| |
| Verify each drywell and suppression 31 days chamber purge system and containment pressure control isolation valve is closed.
| |
| SR 3.6.1.3.2 - -- ----
| |
| -NOTES - - ------ --- -
| |
| : 1. Valves and blind flanges in high 0_ radiation areas may be verified by I
| |
| use of administrative means.
| |
| : 2. Not required to be met for PCIVs that ,
| |
| are open under administrative I controls.
| |
| Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment and is not locked sealed, or otherwise secured and is required to be closed during accident conditions is closed.
| |
| (continued)
| |
| O
| |
| - - - FERMI UNIT 2 3.6 14 Amendment No. 134
| |
| | |
| PCIVs 3.6.1.3
| |
| (] SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY j
| |
| l SR 3.6.1.3.3 - - - -
| |
| - NOTES - ---- - - -- - -
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for PCIVs that are open under administrative controls.
| |
| Verify each primary containment isolation Prior to I manual valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 from not locked. sealed. or otherwise secured MODE 4 if and is required to be closed during primary accident conditions is closed. containment was de inerted while in '
| |
| MODE 4. if not
| |
| , performed rs within the j previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve explosive charge.
| |
| i SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated automatic PCIV. except for with the MSIVs. is within limits. Inservice Testing Program (continued)
| |
| FERMI UNIT 2 3.6 15 Amendment No. 134
| |
| | |
| PCIVs 3.6.1.3
| |
| ) SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILUWCE FREQUENCY I
| |
| SR 3.6.1.3.6 Perform leakage rate testing for each 184 days -
| |
| primary containment purge valve with l resilient seals. 6NQ 1 Once within 92 days after I opening the valve SR 3.6.1.3.7 Verify the isolation time of each MSIV is In accordance
| |
| = 3 seconds and s 5 seconds. with the l Inservice I Testing Program j l
| |
| l SR 3.6.1.3.8 Verify each automatic PCIV actuates to 18 months the isolation position on an actual or i simulated isolation signal.
| |
| SR 3.6.1.3.9 Verify each reactor instrumentation line 18 months EFCV actuates on a simulated instrument line break to restrict flow.
| |
| SR 3.6.1.3.10 Remove and test the explosive squib from 18 months on a each shear isolation valve of the TIP STAGGERED TEST System. BASIS (continued) i f
| |
| FERMI UNIT 2 3.6 16 Amendment No. 134 1
| |
| | |
| PCIVs 3.6.1.3 l
| |
| l e j SURVEILLANCE REQUIREMENTS (continued) b SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Verify the combined leakage rate for all In accordance secondary containment bypass leakage with the paths that are not provided with a seal Primary system is s 0.04 L, when pressurized to Containment
| |
| = 56.5 psig. Leakage Rate Testing Program and Inservice Testing Program SR 3.6.1.3.12 Verify combined MSIV leakage rate for all In accordance four main steam lines is s 100 scfh when with the tested at = 25 psig. Primary Containment Leakage Rate Testing Program
| |
| ('
| |
| 'SR 3.6.1.3.13 --- - - -- -- -- -NOTE - -- - - -------.
| |
| Only required to be met in MODES 1, 2.
| |
| and 3.
| |
| Verify combined leakage rate through In accordance hydrostatically tested lines that with the penetrate the primary containment is Primary within limits. Containment Leakage Rate Testing Program I
| |
| i i
| |
| rm V
| |
| FERMI - UNIT 2 3.6 17 Amendment No. 134
| |
| | |
| 1 i
| |
| l Pricary Containment Pressure l
| |
| 3.6.1.4
| |
| , 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Primary Containment Pressure LC0 3.6.1.4' Primary containment pressure shall be = -0.10 psig and l 5 +2.0 psig. I APPLICABILITY: MODES 1. 2. and 3.
| |
| ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 Restore primary 1 hour pressure not within containment pressure I limit. to within limit.
| |
| l B. Required Action and B.1 Be in MODE 3. 12 hours I g associated Completion g Time not mat. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1
| |
| SR 3.6.1.4.1 Verify rimary containment pressure is 12 hours within imit.
| |
| l i
| |
| ,s~
| |
| U 1 FERMI - UNIT 2 3.6-18 Amendment No. 134 ;
| |
| I 1
| |
| | |
| Drywell Air Temperature 3.6.1.5 3.6 CONTAINMENT SYSTEMS f) 3.6.1.5 Drpell Air Temperature LC0 3.6.1.5 Drpell average air temperature shall be s 145'F. -
| |
| APPLICABILITY: MODES-1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drpell average air A.1 Restore drpell 8 hours temperature not within average air limit. temperature to within limit.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours
| |
| . (N.l associated Completion v Time not met. MQ l
| |
| B.2 Be in MODE 4. 36 hours i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify drpell average air temperature is 24 hours within limit.
| |
| O.
| |
| .U FERMI UNIT 2 3.6 19 Amendment No. 134
| |
| | |
| LLS Valves 3.6.1.6
| |
| ( 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low Low Set (LLS) Valves LC0 3.6.1.6 The LLS function of two safety / relief valves shall be -
| |
| OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1
| |
| A. One LLS valve A.1 Restore LLS valve to 14 days inoperable. OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours fs associated Completion Time of Condition A
| |
| (/ )
| |
| ( ANQ not met.
| |
| B.2 Be in MODE 4. 36 hours Both LLS valves inoperable.
| |
| I i
| |
| i l
| |
| l 2
| |
| i O
| |
| FERMI UNIT 2 3.6 20 Amendment No. 134
| |
| | |
| LLS Valves 3.6.1.6 1
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1 - . - .
| |
| .- NOTE- - - ------ ---
| |
| Not required to be performed until !
| |
| 12 hours after reactor steam pressure and flow are adequate to perform the test.
| |
| Verify each LLS valve opens when manually 18 months j actuated.
| |
| SR 3.6.1.6.2 - -
| |
| - - -- NOTE- - - --- - ------
| |
| Valve actuation may be excluded.
| |
| Verify the LLS System actuates on an 18 months actual or simulated automatic initiation signal.
| |
| .o l
| |
| p.
| |
| FERMI - UNIT 2 3.6 21 Amendment No. 134
| |
| | |
| Reactor Building to Suppression Chamber Vacuum Breakers 3.6.1.7 3.6 CONTAINMENT SYSTEMS 3.6.1.7 Reactor Building to. Suppression Chamber Vacuum Breakers LC0 3.6.1.7 Each reactor building to-suppression chamber vacuum breaker shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2 and 3.
| |
| ACTIONS
| |
| .................................... NOTE---- -------- -- - ----- --- ------
| |
| Separate Condition entry is allowed for each line.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more lines with A.1 Close the open vacuum 72 hours one reactor building- breaker.
| |
| O V
| |
| to suppression chamber vacuum breaker not I closed. l l
| |
| l B. One or more lines with B.1 Close one open vacuum 2 hours l two reactor building- breaker. ;
| |
| to suppression chamber vacuum breakers not closed.
| |
| C. One line with one or C.1 Restore the vacuum 72 hours more reactor building- breaker (s) to to-suppression chamber OPERABLE status.
| |
| vacuum breakers inoperable for opening.
| |
| (continued)
| |
| )
| |
| FERMI UNIT 2 3.6 22 Amendment No. 134
| |
| | |
| Reactor Building to Suppression Chamber Vacuum Breakers 3.6.1.7
| |
| ' ACTIONS (continued)
| |
| C0lOITION REQUIRED ACTION COMPLETION TIME D. Two lines with one or D.1 Restore all vacuum 1 hour -
| |
| more reactor building- breakers in one line to-suppression chamber to OPERABLE status.
| |
| vacuum breakers l
| |
| inoperable for opening.
| |
| I E. Required Action and E.1 Be in MODE 3. 12 hours
| |
| . Associated Completion Time not met. ANQ E.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 -- - --
| |
| NOTES-- - -- - ----
| |
| : 1. Not required to be met for vacuum breakers that are open during Surveillances.
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended function.
| |
| Verify each vacuum breaker is closed. 14 days SR 3.6.1.7.2 Perform a functional test of each vacuum 31 days breaker.
| |
| (continued)
| |
| O l FERMI UNIT 2 3.6 23 Amendment No. 134 l I
| |
| | |
| Reactor Building to Suppression Chamber Vacuum Breakers 3.6.1.7 SURVEILLANCE REQUIREENTS (continued)
| |
| SURVEILLANCE FREQUENCY l
| |
| l SR 3.6.1.7.3 Verify the opening setpoint of each 18 months -
| |
| vacuum breaker is s 0.5 psid.
| |
| l 1
| |
| O l
| |
| )
| |
| l l
| |
| i A
| |
| i FERMI . UNIT 2 3.6 24 Amendment No. 134
| |
| | |
| 4 Suppression Chamber to Dr.Well Vacuum Breakers 3.6.1.8 3.6 CONTAINMENT SYSTEMS
| |
| (]
| |
| 3.6.1.8 Suppression Chamber to Drywell Vacuum Breakers LC0 3.6.1.8 Twelve su)pression chamber to drywell vacuum breakers shall be OPERAB_E.
| |
| APPLICABILITY: MODES 1. 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One suppression A.1 Restore vacuum 72 hours chamber to drywell breaker to OPERABLE vacuum breaker status.
| |
| inoperable for opening.
| |
| (3 kJ B. One or more B.1 Close the open vacuum 2 hours suppression chamber- breaker (s).
| |
| to-drywell vacuum breaker not closed.
| |
| I C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours
| |
| /'~}
| |
| '% J FERMI UNIT 2 3.6 25 Amendment No. 134
| |
| | |
| Suppression Chamber to Drywell Vacuum Breakers 3.6.1.8 4
| |
| SURVEILLANCE REQUIREMENTS j SURVEILLANCE FREQUENCY SR 3.6.1.8.1 - - -- -
| |
| - NOTES --- - --
| |
| : 1. Not required to be met for vacuum breakers that are open during Surveillances. '
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended function.
| |
| Verify each vacuum breaker is closed. 7 days SR 3.6.1.8.2 Perform a functional test of each vacuum Prior to breaker. entering MODE 2 or 3 from MODE 4 if not g- performed in
| |
| ( the previous 92 days 8NQ Within 12 hours !
| |
| after any discharge of steam to the suppression chamber from the SRVs SR 3.6.1.8.3 Verify the opening setpoint of each 18 months vacuum breaker is s 0.5 psid.
| |
| a O
| |
| V FERMI - UNIT 2 3.6 26 Amendment No. 134
| |
| | |
| MSIV LCS 3.6.1.9 3.6 CONTAINMENT SYSTEMS 3.6.1.9 . Main Steam Isolation Valve (MSIV) Leakage Control System (LCS)
| |
| LC0 3.6.1.9 Two MSIV LCS subsystems shall be OPERABLE. -
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. One MSIV LCS subsystem A.1 Restore MSIV LCS 30 days inoperable. subsystem to OPERABLE status.
| |
| B. Two MSIV LCS B.1 Restore one MSIV LCS 7 days subsystems inoperable, subsystem to OPERABLE status.
| |
| N)
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. A_N.Q C.2 Be in MODE 4. 36 hours l
| |
| i rm FERMI UNIT 2 3.6 27 Amendment No. 134
| |
| | |
| l MSIV LCS l 3.6.1.9 l
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.9.1 Verify each MSIV LCS valve, testable 31 days -
| |
| during operation, cycles through at least j one complete cycle of full travel.
| |
| SR 3.6.1.9.2 Verify each MSIV LCS valve, not testable Prior to during operation, cycles through at least entering MODE 2 one complete cycle of full travel. or 3 from MODE 4 if not performed in the previous i
| |
| 31 days SR 3.6.1.9.3 Perform a system functional test of each 18 months MSIV LCS subsystem.
| |
| O l
| |
| l i
| |
| .O i FERMI UNIT 2 3.6 28 Amendment No. 134 l
| |
| i
| |
| | |
| m Suppression Pool Average Temperature 3.6.2.1 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LCO' 3.6.2.1 Suppression pool average temperature shall be: -
| |
| : a. s 95*F with THERMAL POWER > lt RTP and no testing that
| |
| ~a dds heat to the suppression pool is being performed: j
| |
| .b. s 105*F with THERMAL POWER > lt RTP and testing that adds heat to the suppression pool is being performed: I and j
| |
| : c. s 110*F with THERMAL POWER s it RTP.
| |
| I APPLICABILITY: MODES 1, 2, and 3.
| |
| ' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
| |
| , A. Suppression pool A.1 Verify suppression Once per hour
| |
| -average temperature pool average
| |
| > 95'F but s 110*F. temperature 5 110*F.
| |
| M M THERMAL POWER A.2 Restore suppression 24 hours
| |
| > lt RTP. pool average temperature to M s 95*F.
| |
| Not performing testing
| |
| .that adds heat to the suppression pool.
| |
| (continued)
| |
| O
| |
| . FERMI UNIT 2 3.6 29 Amendment No. 134 r
| |
| | |
| Suppression Pool Average Temperature 3.6.2.1
| |
| ( ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Reduce THERMAL POWER 12 hours -
| |
| associated Completion to s 1% RTP.
| |
| Time of Condition A not met.
| |
| C. Suppression pool C.1 Suspend all testing Immediately average temperature that adds heat to the
| |
| > 105*F. suppression pool.
| |
| AND THERMAL POWER > lt RTP.
| |
| AND Performing testing that adds heat to the suppression pool.
| |
| D. Suppression pool D.1 Place the reactor Immediately average temperature mode switch in the
| |
| > 110*F but s 120"F. shutdown position.
| |
| AND i
| |
| D.2 Verify suppression Once per pool average 30 minutes temperature s 120*F.
| |
| AND D.3 Be in MODE 4. 36 hours (continued) m U
| |
| FERMI UNIT 2 3.6-30 Amendment No. 134
| |
| | |
| Suppression Pool Average Temperature l 3.6.2.1 (O.,1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. Suppression pool E.1 Depressurize the 12 hours I average temperature reactor vessel to
| |
| > 120*F. < 200 psig.
| |
| MQ E.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1.1 Verify suppression pool average 24 hours temperature is within the applicable
| |
| /~'N 1imits. AND G
| |
| 5 minutes when performing testing that adds heat to the suppression pool n
| |
| _s/
| |
| FERMI UNIT 2 3.6 31 Amendment No. 134
| |
| | |
| Suppression Pool Water Level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be a -2 inches and -
| |
| 5 +2 inches.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool water A.1 Restore suppression 2 hours level not within pool water level to limits, within limits.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours
| |
| , associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.2.1 Verify suppression pool water level is 24 hours within limits.
| |
| -. j (3
| |
| G 1 l
| |
| FERMI.- UNIT 2 3.6 32 Amendment No. 134
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LC0 3.6.2.3 Two RHR suppression pool cooling subsystems shall be -
| |
| OPERABLE.
| |
| APPLICABILITY: MODES 1, 2 and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days pool cooling subsystem suppression pool inoperable. cooling subsystem to OPERABLE status.
| |
| B. Two RHR suppression B.1 Restore one RHR 8 hours O pool cooling subsystems inoperable, suppression pool cooling subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours FERMI - UNIT 2 3.6 33 Amendment No. 134 J
| |
| | |
| 7 RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling 31 days -
| |
| subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct' position or can be aligned to the correct position.
| |
| SR 3.6.2.3.2 Verify each RHR pum) develops a flow rate In accordance
| |
| = 10,000 gpm throug1 the associated heat with the exchanger while operating in the Inservice suppression pool cooling mode. Testing Program O
| |
| O FERMI UNIT 2 3.6 34 Amendment No. 134
| |
| | |
| RHR Suppression Pool Spray 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray
| |
| .LC0 3.6.2.4 ~ Two Rm suppression pool spray subsystems shall be OPERABLE.
| |
| APPLICABILITY:- MODES 1. 2. and 3.
| |
| ACTIONS-COWITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days pool spray subsystem suppression pool-inoperable. spray subsystem to OPERABLE status.
| |
| -B. Two RHR suppression B.1 Restore one RHR 8 hours pool spray subsystems suppression pool
| |
| - inoperable. spray subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be.in MODE 3. 12 hours associated Completion
| |
| ' Time not' met. AN_Q C.2 Be in MODE 4. 36 hours D
| |
| FERMI - UNIT 2 3.6 35 Amendment No. 134
| |
| | |
| i RHR Suppression Pool Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS-SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RtR suppression pool spray 31 days -
| |
| subsystem manual, power operated, and
| |
| ' automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the. correct position or can be aligned to the correct position.
| |
| l l SR 3.6.2.4.2 Verify each R}R pum) develops a flow rate In accordance l = 500 gpm through t1e heat exchanger and with the
| |
| ! suppression pool spray sparger while Inservice operating in the suppression pool spray Testing Program m&.
| |
| O i
| |
| q l
| |
| l l
| |
| l i
| |
| O FERMI UNIT 2. 3.6 36 Amendment No. 134
| |
| | |
| P Primary Containment Hydrogen Recombiners 3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Hydrogen Recombiners LCO--3.6.3.1 Two primary containment hydrogen recombiners shall be .
| |
| OPERABLE.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One primary A.1 ---- ---NOTE- -------
| |
| containment hydrogen LC0 3.0.4 is not recombiner inoperable. applicable.
| |
| Restore primary 30 days containment hydrogen recombiner to O. OPERABLE status.
| |
| B. Two primary B.1 Verify by 1 hour containment hydrogen administrative means recombiners that the hydrogen ANQ inoperable. control function is maintained. Once per 12 hours thereafter 8!E B.2 Restore one primary 7 days containment hydrogen recombiner to OPERABLE status.
| |
| (continued)
| |
| O FERMI' UNIT 2 3.6 37 Amendment No. 134 i
| |
| | |
| I Primary Containment Hydrogen Recombiners 3.6.3.1 I ACTIONS (continued) ;
| |
| CONDITION REQUIRED ACTION COMPLETION TIME j l
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Perform a system functional test for each 18 months primary containment hydrogen recombiner.
| |
| I
| |
| { SR 3.6.3.1.2 Visually examine each primary containment 18 months I
| |
| (' hydrogen recombiner enclosure and verify there is no evidence of abnormal conditions.
| |
| SR 3.6.3.1.3 Perform a resistance to ground test for 18 months each heater phase.
| |
| O FERMI UNIT 2 3.6 38 Amendment No. 134
| |
| | |
| Prirary Containment Oxygen Concentration !
| |
| 3.6.3.2 I 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxygen Concentration LC0 3.6.3.2 The primary containment oxygen concentration shall be l
| |
| < 4.0 volume percent.
| |
| APPLICABILITY: MODE 1 during the time period: l
| |
| : a. From 24 hours after THERMAL POWER is > 15% RTP following ,
| |
| startup, to I
| |
| : b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next reactor shutdown. ;
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 Restore oxygen 24 hours l q oxygen concentration concentration to !
| |
| b not within limit. within limit. l B. Required Action and B.1 Reduce THERMAL POWER 8 hours associated Completion to 5 15% RTP.
| |
| Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 Verify primary containment oxygen 7 days concentration is within limits.
| |
| O L/
| |
| FERMI UNIT 2 3.6-39 Amendment No. 134
| |
| | |
| l Secondary Containment i 3.6.4.1 h 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment '
| |
| LCO 3.6.4.1 The secondary containment shall be OPERABLE. -
| |
| I i APPLICABILITY: MODES 1. 2. and 3.
| |
| During movement of irradiated fuel assemblies in the secondary containment.
| |
| During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).
| |
| , ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| l A. Secondary Containment A.1 Restore railroad bay 7 days inoperable due to one door to OPERABLE railroad bay access status.
| |
| door inoperable.
| |
| (~')\
| |
| w 1 4
| |
| B. Secondary containment B.1 Restore secondary 4 hours inoperable in MODE 1. containment to i 2. or 3 for reasons OPERABLE status.
| |
| l other than l Condition A.
| |
| l C. Required Action and C.1 Be in MODE 3. 12 hours I
| |
| associated Completion Time of Condition A or AND B not met.
| |
| C.2 Be in MODE 4. 36 hours (continued) r~)
| |
| NJ FERMI UNIT 2 3.6 40 Amendment No. 134
| |
| | |
| i Secondary Containment 3.6.4.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| -NOTE -- --- -- --
| |
| D. Secondary containment ---------
| |
| inoperable during LC0 3.0.3 is not applicable.
| |
| movement of irradiated --- ---- -------------------
| |
| fuel assemblies in the secondary containment, D.1 Suspend movement of Immediately during CORE irradiated fuel ALTERATIONS, or during assemblies in the OPDRVs. secondary containment.
| |
| 8N_Q D.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND D.3 Initiate action to Immediately suspend OPDRVs.
| |
| O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S.3 3.6.4.1.1 Verify secondary containment vacuum is 24 hours
| |
| = 0.125 inch of vacuum water gauge.
| |
| (continued) l C~J FERMI UNIT 2 3.6 41 Amendment No. 134
| |
| | |
| Secondary Containment 3.6.4.1
| |
| 'O suavet u^ ace aeouineae"Ts <co"t4meee>
| |
| SURVEILLANCE FREQUENCY SR 3.6.4.1.2 -- -
| |
| -- - ----NOTE ----- - - -- --- - -
| |
| Not required to be met for one railroad bay access door until:
| |
| : a. 4 hours after opening for entry, exit, or testing: and
| |
| : b. 12 hours after opening for new fuel receipt activities 3rovided the other door remains OPERAB_E and closed.
| |
| Verify all secondary containment 31 days equipment hatches. pressure relief doors and railroad bay access doors are closed and sealed.
| |
| SR 3.6.4.1.3 Verify one secondary containment access 31 days door in each access opening is closed.
| |
| m b
| |
| SR 3.6.4.1.4 Verify steam tunnel blowout panels are Prior to closed. entering MODE 2 l or 3 from
| |
| , MODE 4 if not performed in the previous 31 days SR 3.6.4.1.5 Verify each standby gas treatment (SGT) 18 months on a subsystem will draw down the secondary STAGGERED TEST containment to a 0.25 inch of vacuum BASIS water gauge in 5 567 seconds.
| |
| SR 3.6.4.1.6 Verify each SGT subsystem can maintain 18 months on a
| |
| = 0.25 inch of vacuum water gauge in the STAGGERED TEST secondary containment for 1 hour at a BASIS flow rate s 3000 cfm.
| |
| [v "
| |
| FERMI UNIT 2 3.6 42 Amendment No. 134
| |
| | |
| SCIVs 3.6.4.2
| |
| _3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment' Isolation' Valves (SCIVs)
| |
| LCO 3.6.4.2 Each SCIV shall be OPERABLE.
| |
| APPLICABILITY: MODES 1. 2. and 3.
| |
| During movement of irradiated fuel assemblies in the secondary containment.
| |
| During. CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ,
| |
| ACTIONS l
| |
| .....................................N0TES------------ ------ ---- -- - - ----
| |
| : 1. Penetration flow paths may be unisolated intermittently under administrative controls.
| |
| : 2. Separate Condition entry is allowed for each penetration flow path.
| |
| /~''s 3. Enter applicable Conditions and Required Actions for systems made V inoperable by SCIVs.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours penetration flow paths )enetration flow path with one SCIV )y use of at least inoperable. one closed and de activated automatic valve, closed manual valve.
| |
| or blind flange.
| |
| AND (continued)
| |
| O FERMI UNIT 2 3.6 43 Amendment No. 134
| |
| | |
| SCIVs 3.6.4.2
| |
| /
| |
| ( ACTIONS s_
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --
| |
| ---NOTES-------- '
| |
| : 1. Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| : 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative 1 means.
| |
| Verify the affected Once per 31 days penetration flow path
| |
| ''T is isolated.
| |
| (G B. --
| |
| - - NOTE----- --- B.1 Isolate the affected 4 hours Only applicable to penetration flow path ,
| |
| penetration flow paths by use of at least !
| |
| with two isolation one closed and valves. de activated
| |
| ...................... automatic valve, closed manual valve.
| |
| One or more or blind flange.
| |
| penetration flow paths with two SCIVs inoperable.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met in MODE 1, 2. or 3. C .;2 Be in MODE 4. 36 hours
| |
| () (continued)
| |
| FERMI UNIT 2 3.6-44 Amendment No. 134
| |
| | |
| SCIVs 3.6.4.2 O ^c" o"s (co"t4""ed)
| |
| CONDITION COMPLETION TIME i
| |
| REQUIRED ACTION l l
| |
| D. Required Action and - ------ ---NOTE -- -- - -
| |
| associated Completion LCO 3.0.3 is not applicable.
| |
| Time of Condition A ---- -------- ---------
| |
| or B not met during movement of irradiated D.1 Suspend movement of Immediately fuel assemblies in the irradiated feel secondary containment, assemblies in the during CORE secondary ALTERATIONS, or during containment.
| |
| OPDRVs.
| |
| D.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND D.3 Initiate action to Immediately suspend OPDRVs.
| |
| O i
| |
| O FERMI - UNIT 2 3.6 45 Amendment No. 134
| |
| | |
| E ,
| |
| SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------------ --
| |
| -NOTES--------- -- - --
| |
| : 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
| |
| : 2. Not required to be met for SCIVs that are open under administrative controls.
| |
| Verify each secondary containment 31 days isolation manual valve and blind flange not locked, sealed, or otherwise secured that is required to be closed during accident conditions is closed.
| |
| SR 3.6.4.2.2 Verify the isolation time of each power In accordance operated automatic SCIV is within limits. with the t Inservice Testing Program SR 3.6.4.2.3 Verify each automatic SCIV actuates to 18 months !
| |
| the isolation position on an actual or '
| |
| simulated actuation signal.
| |
| l O
| |
| FERMI UNIT 2 3.6 46 Amendment No. 134
| |
| | |
| SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LCO 3.6.4.3 Two SGT subsystems shall be OPERABLE.
| |
| * APPLICABILITY: MODES 1, 2, and 3.
| |
| During movement of irradiated fuel assemblies in the j secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor i vessel (0PDRVs).
| |
| I l
| |
| ACTIONS
| |
| {
| |
| l CONDITION REQUIRED ACTION COMPLETION TIME j A. One SGT subsystem A.1 Restore SGT 7 days inoperable. subsystem to OPERABLE status.
| |
| I B. Required Action and B.1 Be in MODE 3. 12 hours i associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours (continued)
| |
| )
| |
| I O
| |
| FERMI - UNIT 2 3.6 47 Amendment No. 134
| |
| | |
| i I
| |
| l SGT System j l 3.6.4.3 l l
| |
| l '/ O. ACTIONS (continued)
| |
| ! V CONDITION REQUIRED ACTION COMPLETION TIME l 1
| |
| 1 C. Required Action and ---
| |
| -- -- -NOTE -- - ----- -
| |
| associated Completion LC0 3.0.3 is not applicable.
| |
| Time of Condition A -- ----- ---- -- ------- --
| |
| not met during movement of irradiated C.1 Place OPERABLE SGT Immediately ,
| |
| fuel assemblies in the subsystem in secondary containment, operation.
| |
| during CORE ALTERATIONS, or during @ ;
| |
| OPDRVs.
| |
| C.2.1 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.
| |
| AND C.2.2 Suspend CORE Immediately ALTERATIONS.
| |
| l O
| |
| V AND C.2.3 Initiate action to Immediately suspend OPDRVs.
| |
| i D. Two SGT subsystems D.1 Enter LC0 3.0.3. Immediately inoperable in MODE 1, l
| |
| : 2. or 3.
| |
| I (continued) l l l O
| |
| I FERMI UNIT 2 3.6 48 Amendment No. 134 l
| |
| l
| |
| | |
| L 1 SGT System 3.6.4.3
| |
| (]. ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| -- ------- NOTE---
| |
| E. Two SGT subsystems -- - -
| |
| inoperable during LC0 3.0.3 is not applicable.
| |
| movement of irradiated - -- - ------ --- --
| |
| fuel assemblies in the l secondary E.1 Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS or assemblies in i during OPDRVs. secondary containment.
| |
| AND E.2 Suspend CORE Immediately ALTERATIONS.
| |
| AND E.3 Initiate action to Immediately suspend OPDRVs.
| |
| O 1
| |
| i l
| |
| l l
| |
| l l
| |
| l v
| |
| FERMI UNIT 2 3.6 49 Amendment No. 134 1
| |
| l J
| |
| | |
| c ..
| |
| )
| |
| SGT System 3.6.4.3
| |
| '{} SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
| |
| SR 3.6.4.3.1 Operate each SGT subsystem for a 10 31 days continuous hours with heaters operating.
| |
| l SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
| |
| SR 3.6.4.3.3 Verify each SGT subsystem actuates on an 18 months actual'or simulated initiation signal.
| |
| 1 SR 3.6.4.3.4 Verify each SGT filter cooler bypass 18 months damper can be opened and the fan started.
| |
| 1 i
| |
| l I
| |
| I l
| |
| O FERMI UNIT 2 3.6-50 Amendment No. 134 L
| |
| | |
| RHRSW System 3.7.1
| |
| & 3.7 PLANT SYSTEMS v
| |
| 3.7.1 Residual Heat Removal Service Water (RHRSW) System LC0 3.7.1 Two RHRSW subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1. 2. and 3. i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHRSW pump A.1 Restore RHRSW pump to 30 days inoperable. OPERABLE status.
| |
| B. One RHRSW pump in each B.1 Restore one RHRSW 7 days subsystem inoperable. pump to OPERABLE status.
| |
| C. One RHRSW subsystem C.1- ----- - NOTE-- --- --
| |
| inoperable for reasons Enter applicable other than Conditions and !
| |
| i Condition A. Required Actions of !
| |
| LC0 3.4.8, " Residual Heat Removal (RHR) 3 Shutdown Cooling :
| |
| System- Hot '
| |
| Shutdown." for RHR shutdown cooling made inoperable by RHRSW System.
| |
| Restore RHRSW 7 days subsystem to OPERABLE status.
| |
| (continued)
| |
| FERMI UNIT 2 3.7 1 Amendment No. 134
| |
| | |
| RHRSW System
| |
| , 3.7.1 l
| |
| '4 ACTIONS (continued) 1 (G COMPLETION TIME CONDITION REQUIRED ACTION l
| |
| D. Both RHRSW subsystems D.1 --- -NOTE----- --- -
| |
| ; inoperable for reasons Enter applicable l other than Conditions and Condition B. Required Actions of LC0 3.4.8 for RHR shutdown cooling made inoperable by RHRSW System.
| |
| * j Restore one RHRSW 8 hours i
| |
| subsystem to OPERABLE status.
| |
| l E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time not met. AND E.2 Be in MODE 4. 36 hours l l
| |
| l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ;
| |
| L SR 3.7.1.1 Verify each RHRSW manual, power operated. 31 days i I
| |
| and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct position.
| |
| c lO FERMI UNIT 2 3.7 2 Amendment No. 134
| |
| | |
| EECW/EESW System and UHS ,
| |
| 3.7.2 I
| |
| 3.7 PLANT SYSTEMS 3.7.2 Emergency Equipment Cooling Water (EECW)/ Emergency Equipment Service Water (EESW) System and Ultimate Heat Sink (VHS) ;
| |
| i LC0 3.7.2 Two EECW/EESW subsystems and UHS shall be OPERABLE. !
| |
| APPLICABILITY: MODES 1, 2 and 3. ;
| |
| i l
| |
| ACTIONS l 1
| |
| .......................................N0TES--- -------- --------------- ----
| |
| : 1. Enter applicable Conditions and Required Actions of LC0 3.8.1. "AC Sources-0perating." for diesel generator made inoperable by UHS.
| |
| : 2. Enter applicable Conditions and Required Actions of LC0 3.4.8. " Residual I Heat Removal (RHR) Shutdown Cooling System Hot Shutdown." for RHR shutdown cooling made inoperable by EECW/EESW or UHS.
| |
| l O CONDITION COMPLETION TIME
| |
| .() REQUIRED ACTION A. UHS ino>erable due to A.1 Restore VHS cross tie 8 hours inopera)le cross-tie lines to OPERABLE line(s).
| |
| status.
| |
| B. One reservoir B.1 Restore reservoir to 72 hours inoperable. OPERABLE status.
| |
| C. One EECW/EESW C.1 Restore the EECW/EESW 72 hours subsystem inoperable subsystem to OPERABLE for reasons other than status.
| |
| Conditions A and B.
| |
| (continued)
| |
| O FERMI UNIT 2 3.7 3 Amendment No. 134
| |
| | |
| i EECW/EESW System and UHS !
| |
| 3.7.2
| |
| / ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 12 hours -
| |
| associated Completion Time of Condition A. AND B, or C not met.
| |
| D.2 Be in MODE 4. 36 hours E
| |
| Both EECW/EESW subsystems ino>erable for reasons otler than Condition A.
| |
| UHS inoperable for reasons other than Conditions A and B.
| |
| O l b ,
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level of each UHS 24 hours reservoir. and the average water level of each of the two reservoirs, are a 25 ft.
| |
| SR 3.7.2.2 Verify the average water temperature of 24 hours each reservoir, and combined average water temperature of the two reservoirs, are s 80*F. j i
| |
| (continued) !
| |
| i U,,
| |
| FERMI - UNIT 2 3.7 4 Amendment No. 134
| |
| | |
| i EECW/EESW System and UHS 3.7.2
| |
| (]~ SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.7.2.3 - - --- - -
| |
| NOTE- - - ----------
| |
| Fast speed testing not required to be performed during icing periods.
| |
| Operate each cooling tower fan on slow 31 days speed and on fast speed. each for {
| |
| = 15 minutes.
| |
| l SR 3.7.2.4 -- - --- -
| |
| - --- NOTE-- --------------- -
| |
| Isolation of EECW flow to individual components does not render EECW System inoperable.
| |
| Verify each EECW/EESW subsystem and UHS 31 days manual, power o)erated. and automatic valve.
| |
| . in the flow pat 1s servicing safety related systems or components, that is not locked.
| |
| sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.7.2.5 Verify each EECW/EESW subsystem actuates on 18 months I an actual or simulated initiation signal. '
| |
| 1 O
| |
| FERMI UNIT 2 3.7 5 Amendment No. 134
| |
| | |
| r CREF System '
| |
| 3.7.3 I 3.7 PLANT SYSTEMS 3.7.3 Control Room Emergency Filtration (CREF) System LC0 3.7.3 The CREF System shall be OPERABLE.
| |
| * APPLICABILITY: MODES 1, 2, and 3.
| |
| During movement of irradiated fuel assemblies in the !
| |
| secondary containment. l During CORE ALTERATIONS, I During operations with a potential for draining the reactor vessel (0PDRVs).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I
| |
| A. One CREF subsystem A.1 Restore CREF 7 days inoperable. subsystem to OPERABLE status. i O. ;
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours ,
| |
| associated Completion Time of Condition A ANJ not met in MODE 1, 2.
| |
| or 3. B.2 Be in MODE 4. 36 hours (continued) f3 U
| |
| FERMI UNIT 2 3.7-6 Amendment No. 134
| |
| | |
| CREF System 3.7.3 f) ACTIONS (continued) v CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and -- --.-
| |
| NOTE -- - -- -
| |
| associated Completion LC0 3.0.3 is not applicable.
| |
| Time of Condition A -- -- ------ ----- --- --
| |
| not met during !
| |
| movement of irradiated C.1 Place OPERABLE CREF Immediately I fuel assemblies in the subsystem in secondary containment, recirculation mode.
| |
| during CORE ALTERATIONS. or during 2 OPDRVs.
| |
| C.2.1 Initiate action to Immediately suspend OPDRVs.
| |
| AN._Q
| |
| ..............N0TE------- ---
| |
| Not required for a CREF System or subsystem inoperable for performance of
| |
| 'SR 3.7.3.6 due to failure to O provide the required filtration efficiency. or due U
| |
| to replacement of charcoal l filtration media. l C.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment.
| |
| AND C.2.3 Suspend CORE Immediately ALTERATIONS.
| |
| (continued)
| |
| .O FERMI UNIT 2 3.7-7 Amendment No. 134
| |
| | |
| CREF System 3.7.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. Two CREF subsystems or D.1 Enter LCO 3.0.3. Immediately-a non redundant component or portion of the CREF System inoperable in MODE 1, 2, or 3.
| |
| E. Two CREF subsystems or ----- -
| |
| ---NOTE--------- - -
| |
| a non redundant LCO 3.0.3 is not applicable.
| |
| component or portion ---------- -------- - --- ---
| |
| of the CREF System inoperable during . E.1 Initiate action to Immediately movement of irradiated suspend OPDRVs.
| |
| fuel assemblies in the secondary containment, 8@
| |
| during CORE ALTERATIONS, or during --
| |
| -- --NOTE - -- -----
| |
| OPDRVs. Not required for a CREF
| |
| /'h System or subsystem V inoperable for performance of SR 3.7.3.6 due to failure to provide the reguired filtration efficiency. or due to replacement of charcoal filtration media.
| |
| E.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment.
| |
| AND E.3 Suspend CORE Immediately ALTERATIONS.
| |
| O FERMI - UNIT 2 3.7 8 Amendment No. 134
| |
| | |
| CREF System 3.7.3 SURVEILLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY
| |
| .SR 3.7.3.1 Operate one CREF subsystem for = 10 31-days on a continuous hours with the heater operating STAGGERED TEST and operate the other CREF subsystem BASIS for e 75 minutes.
| |
| SR 3.7.3.2 ----- - ----- -- NOTE--------- -- - ----
| |
| When the CREF system is made inoperable in '
| |
| MODE 1. 2. or 3 solely for VFTP required surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours.
| |
| Perform required CREF filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
| |
| O SR 3.7.3.3 Visually inspect silicone sealant on 12 months accessible portions of CREF system duct work outside the control room that are at negative pressure during accident l conditions and for which potential in-leakage would not receive full filtration.
| |
| SR .3.7.3.4 Verify each CREF subsystem actuates on an 18 months actual or simulated initiation signal.
| |
| (continued) l 1
| |
| 'O i V
| |
| FERMI - UNIT 2 3.7 9 Amendment No. 134
| |
| | |
| 1 CREF System 3.7.3 t SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY l SR 3.7.3.5 Verify each CREF subsystem can maintain a 18 months .
| |
| positive pressure of = 0.125 inches water on a gauge relative to the outside atmosphere STAGGERED during the recirculation mode of operation TEST BASIS at a makeup flow rate of s 1800 cfm.
| |
| SR 3.7.3.6 Verify that unfiltered inleakage from CREF 36 months system duct work outside the Control Room envelope that is at negative pressure during accident conditions is within limits.
| |
| (3 V 1 (9
| |
| 'J FERMI UNIT 2 3.7 10 Amendment No. 134
| |
| | |
| Control Center AC System 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Control Center Air Conditioning (AC) System LC0 3.7.4 Two control center AC subsystems shall be OPERABLE. -
| |
| APPLICABILITY: MODES 1. 2, and 3.
| |
| During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One control center AC A.1 Restore control 30 days subsystem inoperable. center AC subsystem to OPERABLE status.
| |
| B. Required Action and B.1 Be in MODE 3. 12 hours -
| |
| associated Completion.
| |
| Time of Condition A 8N_Q not met in MODE 1, 2.
| |
| or 3. B.2 Be in MODE 4. 36 hours (continued)
| |
| FERMI UNIT 2 3.7 11 Amendment No. 134 t
| |
| Lj .. . . .. . .
| |
| | |
| l Control Center AC System 3.7.4 ACTIONS (continued) _ _ _ _
| |
| CONDITION REQUIRED ACTION COMPLETION TIME !
| |
| C. Required Action and --- - NOTE-- -- --- ---
| |
| * associated Completion LC0 3.0.3 is not applicable. I Time of Condition A ---- -- -- -- - -- --
| |
| not met during movement of irradiated C.1 Place OPERABLE Immediately fuel assemblies in the control center AC secondary containment, subsystem in !
| |
| I during CORE operation. '
| |
| l ALTERATIONS. or during OPDRVs. QB l
| |
| C.2.1 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment. l
| |
| ! I E i C.2.2 Suspend CORE Immediately )
| |
| ALTERATIONS.
| |
| m C.2.3 Initiate action to Immediately ;
| |
| suspend OPDRVs.
| |
| l D. Two control center AC D.1 Enter LC0 3.0.3. Immediately l subsystems inoperable l in MODE 1, 2. or 3.
| |
| (continued) l l
| |
| .O FERMI UNIT 2 3.7 12 Amendment No. 134
| |
| | |
| Control Center AC System 3.7.4 ACTIONS (continued)
| |
| C0WITION REQUIRED ACTION COMPLETION TIME E. Two control center AC - ---
| |
| - NOTE---- - --- -
| |
| subsystems inoperable LC0 3.0.3 is not applicable.
| |
| during movement of - - -- - --- -- --- --
| |
| irradiated fuel assemblies in the E.1 Suspend movement of Immediately secondary containment, irradiated fuel during CORE assemblies in the ALTERATIONS or during secondary 0PORVs. containment.
| |
| M E.2 Suspend CORE Immediately ALTERATIONS.
| |
| M l E.3 Initiate actions to Immediately suspend OPDRVs.
| |
| 1
| |
| ' SURVEILLANCE REQUIREMENTS l
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.7.4.1 Verify the control room air temperature is 12 hours s 95*F. ,
| |
| i
| |
| .O FERMI - UNIT 2 3.7 13 Amendment No. 134
| |
| | |
| I L Main Condenser Offgas l 3.7.5 !
| |
| } 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas l
| |
| l LCO 3.7.5 The gross radioactivity rate of the noble gases measured at the discharge of the 2.2 minute delay piping shall be 5 340 mC1/second after decay of 30 minutes.
| |
| APPLICABILITY: MODE 1.
| |
| MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.
| |
| 1 ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. Gross radioactivity A.1 Restore gross 72 hours rate of the noble radioactivity rate of gases not within the noble gases to ;
| |
| limit, within limit. !
| |
| l l
| |
| I B. Required Action and B.1 Isolate all main 12 hours !
| |
| associated Completion steam lines Time not met.
| |
| 2 B.2 Isolate SJAE. 12 hours E
| |
| B.3.1 Be in MODE 3. 12 hours AND B.3.2 Be in MODE 4. 36 hours
| |
| [v FERMI UNIT 2 3.7 14 Amendment No. 134
| |
| | |
| Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 -
| |
| --- ---- -- - -NOTE--- ---- -- -------
| |
| Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
| |
| Verify the gross radioactivity rate of the 31 days noble gases is s 340 mci /second after decay of 30 minutes.
| |
| SR 3.7.5.2 Verify the gross radioactivity rate of the Once within noble gases is s 340 mci /second 4 hours after a after decay of 30 minutes. = 50% increase in the nominal
| |
| , steady state fission gas release after c factoring out increases due to changes in THERMAL POWER level 1
| |
| I FERMI UNIT 2 3.7-15 Amendment No. 134
| |
| | |
| Main Turbine Bypass System and Moisture Separator Reheater 3.7.6 3.7 PLANT SYSTEMS 3.7.6 The Main Turbine Bypass System and Moisture Separator Reheater LC0 3.7.6 The Main Turbine Bypass System and Moisture Separator -
| |
| Reheater shall be OPERABLE.
| |
| DB LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System and Moisture Separator Reheater, as specified in the COLR, are made applicable.
| |
| APPLICABILITY: THERMAL POWER = 25% RTP.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 2 hours A LC0 not met. requirements of the V LCO.
| |
| B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to < 25% RTP.
| |
| Time not met.
| |
| O FERMI UNIT 2 3.7 16 Amendment No. 134
| |
| | |
| l Main Turbine Bypass System and Moisture Separator Reheater 3.7.6 l
| |
| rD V SURVEILLANCE REQUIREMENTS 1
| |
| l SURVEILLANCE FREQUENCY
| |
| .SR 3.7.6.1 Verify one complete cycle of each main 92 days turbine bypass valve.
| |
| l Once after each entry into MODE 4 l SR 3.7.6.2 Perform a system functional test. 18 months i I
| |
| l SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE 18 months TIME is within limits. I l
| |
| l c
| |
| b FERMI UNIT 2 3.7 17 Amendment No. 134 i
| |
| | |
| Spent Fuel Storage Pool Water Level 3.7.7
| |
| '(3 V 3.7 PLANT SYSTEMS 3.7.7 Spent Fuel Storage Pool Water Level LC0 3.7.7 The spent fuel storage pool water level shall be = 22 ft over the top of irradiated fuel assemblies seated in the 1 spent fuel storage pool racks.
| |
| APPLICABILITY: During movement of irradiated fuel assemblies in the spent l fuel storage pool. !
| |
| 1 1
| |
| ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage A.1 ------- NOTE - -- ---
| |
| pool water level not LC0 3.0.3 is not within limit. applicable.
| |
| Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the spent fuel storage pool water 7 days !
| |
| level is a 22 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.
| |
| i I
| |
| FERMI UNIT 2 3.7 18 Amendment No. 134
| |
| | |
| L l l
| |
| EDGSW System 3.7.8 i
| |
| /3 3.7 PLANT SYSTEMS i U l 3.7.8 Emergency Diesel Generator Service Water (EDGSW) System LC0 3.7.8 Four FDGSW subsystems shall be OPERABLE. -
| |
| APPLICABILITY: When associated EDG is required to be OPERABLE. ,
| |
| ACTIONS {
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more EDGSW A.1 Declare associated Immediately subsystems inoperable. EDG(s) inoperable.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify each EDGSW subsystem manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
| |
| SR 3.7.8.2 Verify each EDGSW subsystem pump starts 18 months automatically when the associated EDG <
| |
| starts.
| |
| FERMI UNIT 2 3.7 19 Amendment No. 134
| |
| | |
| AC Sources-Operating 3.8.1
| |
| ~
| |
| 3.8 ELECTRICAL POWER ~ SYSTEMS-3.8.1 AC Sources-Operatings LC0 3.8.1 The. following AC electrical power sources shall be OPERABLE:
| |
| -a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and
| |
| : b. Two emergency diesel generators (EDGs) per division.
| |
| A APPLICABILITY: MODES 1, 2. and 3.
| |
| ACTIONS:
| |
| . CONDITION REQUIRED ACTION COMPLETION TIME A. One or both EDGs in. A.1 Perform SR 3.8.1.1 1 hour
| |
| , one division for OPERABLE offsite inoperable. circuit (s). AN12 Once per 8 hours thereafter 8NQ A.2 Declare required 4 hours from
| |
| . feature (s), supported discovery of an by the inoperable inoperable EDG EDGs. inoperable when concurrent with the redundant inoperability of required feature (s) redundant are inoperable, required feature (s) 8N_Q A.3 Verify the status of Once per 8 hours CTG 11-1.
| |
| (continued)
| |
| O
| |
| ' FERMI - UNIT 2. 3.8 1 Amendment No. 134
| |
| | |
| AC Sources-Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4.1 Determine OPERABLE 24 hours -
| |
| EDG(s) are not inoperable due to common cause failure.
| |
| 2 A.4.2 Perform SR 3.8.1.2 24 hours for OPERABLE EDG(s).
| |
| AN_Q A.5 Restore availability 72 hours from of CTG 11 1. discovery of Condition A concurrent with CTG 11-1 not available AND D
| |
| (/ A.6 Restore both EDGs in 7 days j the division to OPERABLE status.
| |
| l 1
| |
| B. One or both EDGs in B.1 Restore both EDGs in 2 hours both divisions one division to inoperable. OPERABLE status. i l
| |
| C. One or two offsite C.1 Be in MODE 3. 12 hours circuits inoperable.
| |
| % MD Required Action and C.2 Be in MODE 4. 36 hours Associated Completion Time of Condition A or B not met.
| |
| !3 V
| |
| FERMI UNIT 2 3.8 2 Amendment No. 134
| |
| | |
| AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker . alignment and 7 days indicated power availability for each offsite circuit.
| |
| I l
| |
| SR 3.8.1.2 -
| |
| - - -- ---- ----NOTES-- -- - --- ---
| |
| : 1. All EDG starts may be receded by an engine prelube period nd followed by a warmup period prior to loading. )
| |
| {
| |
| .2. A modified EDG start involving idling i and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.
| |
| Verify each EDG starts and achieves steady 31 days state voltage = 3740 V and s 4580 V and frequency = 58.8 Hz and s 61.2 Hz.
| |
| SR 3.8.1.3 --- --- ---- -
| |
| - NOTES - - - - - -
| |
| : 1. EDG loadings may include gradual loading as recommended by the manufacturer.
| |
| : 2. Momentary transients below the iced limit do not invalidate this test.
| |
| : 3. This Surveillance shall be conducted on only one EDG at a time.
| |
| Verify each EDG is synchronized and loaded 31 days and operates for = 60 minutes at a load a 2500 kW.
| |
| (continued)
| |
| ' FERMI UNIT 2 3.83 Amendment No. 134 l
| |
| l
| |
| | |
| p AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.4 Verify each day tank contains = 210 gal of 31 days -
| |
| . fuel oil.
| |
| SR 3.8.1.5 Check for and remove accumulated water from 31 days each day tank.
| |
| SR 3.8.1.6 Verify each fuel oil transfer system 31 days operates to automatically transfer fuel oil from storage tanks to the day tanks.
| |
| SR 3.8.1.7 -- --- -- - - -
| |
| - NOTE --- - - -------
| |
| All EDG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
| |
| ( ...........................................
| |
| Verify each EDG starts from standby 184 days condition and achieves: I
| |
| : a. In s 10 seconds, voltage = 3740 V and frequency = 58.8 Hz: and
| |
| : b. Steady state voltage = 3740 V and s 4580 V and frequency = 58.8 Hz and 5 61.2 Hz. ,
| |
| SR 3.8.1.8 Verify each EDG rejects a load greater than 18 months or equal to its associated single largest post accident load, and following load rejection, the frequency is s 66.75 Hz.
| |
| (continued) k .
| |
| FERMI UNIT 2 3.8 4 Amendment No. 134
| |
| | |
| 1 AC Sources-Operating I 3.8.1 !
| |
| I SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY f SR 3.8.1.9 Verify each EDG does not trip and voltage 18 months -
| |
| is maintained s 4784 V during and following a load rejection of a 2850 kW.
| |
| I SR 3.8.1.10 -- ------- ------ -NOTE-- ------------ ---
| |
| All EDG starts may be preceded by an engine prelube period. j
| |
| ........................................... i Verify on simulated loss of offsite power 18 months j signal: I
| |
| : a. De-energization of emergency buses;
| |
| : b. Load shedding from emergency buses; and )
| |
| : c. EDG auto starts and:
| |
| : 1. energizes permanently connected loads in s 10 seconds,
| |
| : 2. energizes auto-connected shutdown loads through load sequencer.
| |
| : 3. maintains steady state voltage a 3740 V and s 4580 V.
| |
| : 4. maintains steady state frequency a 58.8 Hz and 5 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto-connected shutdown loads for a 5 minutes.
| |
| (continued)
| |
| O FERMI UNIT 2 3.8 5 Amendment No. 134
| |
| | |
| 1 l
| |
| AC Sources-Operating !
| |
| 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.11 ---- - --- -- .--- NOTE---------- - --- - -
| |
| All EDG starts may be preceded by an engine prelube period.
| |
| Verify on an actual or simulated Emergency 18 months Core Cooling System (ECCS) initiation signal each EDG auto starts and:
| |
| : a. In s 10 seconds after auto start and during tests, achieves voltage
| |
| = 3740 V and frequency = 58.8 Hz:
| |
| : b. Achieves steady state voltage = 3740 V and s 4580 V. and frequency = 58.8 Hz and s 61.2 Hz: and ;
| |
| i
| |
| : c. Operates for a 5 minutes. I SR 3.8.1.12 Verify each EDG's automatic trips are 18 months bypassed on an actual or simulated emergency start signal except:
| |
| : a. Engine overspeed:
| |
| : b. Generator differential current:
| |
| : c. Low lube oil pressure;
| |
| : d. Crankcase overpressure: and
| |
| : e. Failure to start. l (continued)
| |
| (' s FERMI UNIT 2 3.8 6 Amendment No. 134
| |
| : n. .
| |
| AC Sources-Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.13. - - -- - -- -
| |
| NOTE- -- -- - - - ---- -
| |
| Momentary transients outside the load range.
| |
| do not invalidate this test.
| |
| Verify each EDG operates for = 24 hours: 18 months
| |
| : a. For all but the final = 2 hours loaded
| |
| = 2500 kW and s 2600 kW: and )
| |
| : b. For the final = 2 hours of the test loaded a 2800 kW and s 2900 kW.
| |
| SR 3.8.1.14 ----- -- --
| |
| - - NOTES ------ -- -- - ---
| |
| : 1. This Surveillance shall be performed within 5 minutes of shutting down the EDG after the EDG has operated
| |
| = 2 hours loaded = 2500 kW or until O.
| |
| operating temperatures have stabilized.
| |
| Momentary transients below the load
| |
| ' limit do not invalidate this test.
| |
| : 2. All EDG starts may be preceded by an engine prelube period.
| |
| Verify each EDG starts and achieves: 18 months
| |
| : a. In s 10 seconds, voltage = 3740 V and frequency = 58.8 Hz: and
| |
| : b. Steady state voltage = 3740 V and s 4580 V and frequency = 58.8 Hz and 5 61.2 Hz.
| |
| (continued) l
| |
| (~h A.J
| |
| ' FERMI UNIT 2 3.87 Amendment No. 134
| |
| | |
| 1 AC Sources-Operating 3.8.1
| |
| ( ' SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.15 Verify each EDG: 18 months .
| |
| : a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;
| |
| : b. Transfers loads to offsite power source: and
| |
| : c. Returns to standby status.
| |
| SR 3.8.1.16 Verify interval between each sequenced 18 months ,
| |
| load block is within i 10% of design l interval for each load sequencer timer.
| |
| (continued)
| |
| (-
| |
| J
| |
| \
| |
| 1
| |
| %J FERMI UNIT 2 3.8 8 Amendment No. 134
| |
| | |
| AC Sources-Operating i 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.1.17 -- ---
| |
| -- -----NOTE - -- -- -- - --- -
| |
| All EDG starts may be preceded by an engine prelube period. l Verify, on simulated loss of offsite power 18 months signal in conjunction with an actual or simulated ECCS initiation signal:
| |
| : a. De energization of emergency buses;
| |
| : b. Load shedding from emergency buses:
| |
| and
| |
| : c. EDG auto starts and:
| |
| : 1. energizes permanently connected loads in s 10 seconds, !
| |
| l
| |
| : 2. energizes auto-connected i O
| |
| V emergency loads through load sequencer.
| |
| : 3. achieves steady state voltage
| |
| = 3740 V and s 4580 V.
| |
| : 4. achieves steady state frequency
| |
| = 58.8 Hz and s 61.2 Hz, and
| |
| : 5. supplies permanently connected and auto connected emergency loads for a 5 minutes.
| |
| I SR 3.8.1.18 -
| |
| - - -- -- -- -NOTE ---- - -- - -
| |
| All EDG starts may be preceded by an engine prelube period.
| |
| Verify, when started simultaneously each 10 years EDG achieves, in s 10 seconds, frequency
| |
| = 58.8 Hz.
| |
| FERMI - UNIT 2 3.8 9 Amendment No. 134
| |
| | |
| n AC Sources-Shutdown 3.8.2 O 3.8 e'eC1a1Ca' eowea Sv5TEMs 3.8.2 AC Sources-Shutdown LC0 3.8.2 The following AC electrical power sources shall be OPERABLE:
| |
| : a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem (s) required by LC0 3.8.8.
| |
| " Distribution Systems-Shutdown"; and
| |
| : b. Two emergency diesel generators (EDGs) capable of supplying one division of the onsite Class 1E AC electrical power distribution subsystem (s) required by LCO 3.8.8.
| |
| APPLICABILITY: MODES 4 and 5.
| |
| During movement of irradiated fuel assemblies in the secondary containment.
| |
| ACTIONS
| |
| .................................... NOTE ----- ---- - -- - - ------ ------ -
| |
| LC0 3.0.3 is not applicable.
| |
| n ..............................................................................
| |
| V CONDITION REQUIRED ACTION COMPLETION TIME I A. One required offsite -- ------ -NOTE- --- -- -
| |
| circuit inoperable. Enter applicable Condition and Required Actions of LC0 3.8.8. with one required division de-energized as a result of Condition A.
| |
| A.1 Declare affected Immediately required feature (s),
| |
| with no offsite power available, inoperable.
| |
| (continued) {
| |
| .o FERMI - UNIT 2 3.8 10 Amendment No. 134
| |
| | |
| l AC Sources-Shutdown 3.8.2
| |
| ( ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Suspend CORE Immediately- 1 ALTERATIONS.
| |
| M A.2.2 Suspend movement of Immediately l irradiated fuel I assemblies in the secondary containment.
| |
| E A.2.3 Initiate action to Immediately sus)end operations wit 1 a potential for draining the reactor vessel (OPDRVs).
| |
| AND A.2.4 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.
| |
| (continued)
| |
| O FERMI UNIT 2 3.8 11 Amendment No. 134
| |
| | |
| AC Sources-Shutdown 3.8.2
| |
| []
| |
| v ACTIONS (continued)
| |
| CONDITION REQllIRED ACTION COMPLETION TIME B. One or both required B.1 Suspend CORE Immediately-EDGs inoperable. ALTERATIONS.
| |
| M B.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.
| |
| M B.3 Initiate action to Immediately suspend OPDRVs.
| |
| AND B.4 Initiate action to Immediately restore required EDGs
| |
| (]
| |
| %.)
| |
| to OPERABLE status.
| |
| SURVEILLANCE REQUIREME,NTS SURVEILLANCE FREQUENCY SR 3.8.2.1 -- ---------- -
| |
| NOTE - ---- - ----- --
| |
| The following SRs are not required to be performed: SR 3.8.1.2, SR 3.8.1.3, and SR 3.8.1.7 through SR 3.8.1.17.
| |
| For AC sources required to be OPERABLE In accordance SR 3.8.1.1 through SR 3.8.1.17, are with applicable applicable. SRs I
| |
| FERMI UNIT 2 3.8 12 Amendment No. 134
| |
| | |
| 1 Diesel Fuel Oil and Starting Air 3.8.3 Q(~'\ 3.8 ELECTRICAL POWER SYSTEMS 3.8.3 Diesel Fuel Oil and Starting Air LC0 3.8.3 The stored diesel fuel oil and starting air subsystem shall be within limits for each required emergency diesel generator (EDG).
| |
| APPLICABILITY: When associated EDG is required to be OPERABLE.
| |
| ACTIONS
| |
| ..................................... NOTE- --- - - -- - ------ - - --------
| |
| Separate Condition entry is allowed for each EDG.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| . A. One or more required A.1 Restore fuel oil 48 hours EDGs with fuel oil level to within v . level < 35.280 gal and limits.
| |
| > 30,240 gal in storage tank.
| |
| B. One or more required B.1 Restore fuel oil 7 days EDGs with stored fuel total particulates to oil total particulates within limit.
| |
| not within limit, j
| |
| I C. One or more required C.1 Restore stored fuel 30 days !
| |
| EDGs with new fuel oil oil properties to l properties not within within limits. i limits.
| |
| (continued)
| |
| D i O
| |
| FERMI UNIT 2 3.8-13 Amendment No. 134
| |
| | |
| F Diesel Fuel Oil and Starting Air 3.8.3 O' ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i 1
| |
| D. Required Action and D.1 Declare associated Immediately associated Completion EDG inoperable.
| |
| Time not met.
| |
| E One or more required EDGs with diesel fuel oil, or starting air subsystem not within limits for reasons other than Condition A, B, or C.
| |
| ('] SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each required EDG fuel oil storage 31 days tank contains = 35,280 gal of fuel.
| |
| SR 3.8.3.2 Verify each required EDG fuel oil In accordance properties of new and stored fuel oil are with the tested in accordance with, and maintained Emergency within the limits of, the Emergency Diesel Diesel Generator Fuel Oil Testing Program. Generator Fuel Oil Testing Program (continued)
| |
| O FERMI UNIT 2 3.8 14 Amendment No. 134
| |
| | |
| Diesel Fuel Oil and Starting Air 3.8.3 i t SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY l
| |
| SR 3.8.3.3 Verify each required EDG air start receiver pressure.is = 215 psig.
| |
| 31 days -
| |
| )
| |
| SR 3.8.3.4 Check for and remove accumulated water from 31 days each required EDG fuel oil storage tank.
| |
| i O
| |
| i O ;
| |
| FERMI UNIT 2 3.8 15 Amendment No. 134
| |
| =
| |
| | |
| DC Sources-Operating 3.8.4
| |
| . 3.8 ELECTRICAL POWER' SYSTEMS 3.8.4 DC Sources-Operating
| |
| .LCO '3.8.4 The Division I and Division II DC electrical power -
| |
| . subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2. and 3.
| |
| ACTIONS C0WITION REQUIRED ACTION COMPLETION TIME i I
| |
| A. One battery charger A.1 Restore battery 4 hours inoperable. charger to OPERABLE )
| |
| status.
| |
| 4
| |
| 'B. One DC electrical B.1 Restore DC electrical 2 hours power subsystem power subsystem to inoperable for reasons OPERABLE status. 4 other than Condition A. j l
| |
| C. Re' quired Action and- C.1 Be in MODE 3. 12 hours
| |
| . Associated Completion Time not met. A!gl ,
| |
| 1 C.2 Be in MODE 4. 36 hours 4 l
| |
| l l
| |
| 1 i
| |
| O FERMI ' UNIT 2' 3.8 16 Amendment No. 134
| |
| | |
| DC Sources-Operating 3.8.4 O suavetu ^"ce aeou'aeaea's SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is = 130 V 7 days -
| |
| for Division I and = 125.7 V for Division II on float charge.
| |
| SR 3.8.4.2 Verify no visible corrosion at battery 92 days terminals and connectors.
| |
| DB j Verify each battery cell to cell and terminal connection resistance is s 1.5E-4 ohm.
| |
| SR 3.8.4.3 Verify battery cells, cell plates, and 18 months ,
| |
| racks show no visual indication of physical I damage or abnormal deterioration that could O degrade battery performance.
| |
| SR 3.8.4.4 Remove visible corrosion and verify battery 18 months cell to cell and terminal connections are coated with anti-corrosion material.
| |
| SR 3.8.4.5 Verify each battery cell-to-cell and 18 months terminal connection resistance is '
| |
| s 1.5E-4 ohm.
| |
| SR -3.8.4.6 Verify each required battery charger 18 months supplies for Division I: a 100 amps r; a 129 V for a 4 hours; and for Divisiv II:
| |
| a 100 amps at = 124.7 V for a 4 hours.
| |
| (continued)
| |
| .O FERMI - UNIT 2 3.8 17 Amendment No. 134
| |
| | |
| DC Sources-Operating 3.8.4
| |
| ' SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.4.7- ....................N0TE-----.---..-.-. --- -
| |
| The performance discharge test in SR 3.8.4.8 may be performed in lieu of the service test in SR 3.8.4.7 once per 60 months.
| |
| Verify battery capacity is adequate to 18 months supply. and maintain in OPERABLE status, the actual or simulated emergency loads for !
| |
| the design duty cycle when subjected to a i battery service test.
| |
| SR 3.8.4.8 -.-.- .-.....- .N0TE-..------- . --.----
| |
| This Surveillance shall not be performed in MODE 1. 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.
| |
| Verify battery capacity is = 80% of the 60 months manufacturer's rating when subjected to a performance discharge test. ANQ 18 months when battery shows degradation or has reached 85%
| |
| of expected life O
| |
| FERMI UNIT 2 3.8 18 Amendment No. 134 i i
| |
| I
| |
| | |
| DC Sources-Shutdown 3.8.5 l
| |
| .3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown LCO 3.8.5 The following shall be OPERABLE:
| |
| : a. One DC electrical power subsystem capable of supplying one division of the onsite Class 1E electrical power distribution subsystem (s) required by LC0 3.8.8.
| |
| " Distribution System Shutdown":'and
| |
| : b. One DC electrical power subsystem battery or battery charger. other than that required by LC0 3.8.5.a. !
| |
| capable of supplying the remaining onsite Class 1E -
| |
| electrical power distribution subsystem (s) when required by LC0 3.8.8.
| |
| APPLICABILITY: MODES 4 and 5. l During movement of irradiated fuel assemblies in the 1 secondary containment.
| |
| ACTIONS
| |
| .....................................N0TE -- - - - -------- -------- - ----
| |
| LC0 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| :A. 'One or more required A.1 Declare affected Immediately DC electrical power required feature (s) subsystems inoperable, inoperable.
| |
| .08 (continued)
| |
| O FERMI. UNIT 2 3.8 19 Amendment No. 134
| |
| | |
| DC Sources-Shutdown 3.8.5 O ac'ioas C0WITION. REQUIRED ACTION COMPLETION TIME
| |
| )
| |
| j A. (continued) A.2.1 Suspend CORE Immediately.
| |
| ALTERATIONS.
| |
| m A.2.2 Suspend movement of Immediately ;
| |
| irradiated fuel !
| |
| assemblies in the ,
| |
| secondary j containment. j 1
| |
| I A.2.3 Initiate action to Immediately )
| |
| sus and operations wit 1 a potential for draining the reactor ,
| |
| I vessel.
| |
| AND
| |
| ', l A.2.4 Initiate action to Immediately l restore required DC electrical power subsystems to OPERABLE status.
| |
| l l
| |
| O FEPJil UNIT 2 3.8 20 Amendment No. 134
| |
| | |
| DC Sources-Shutdown 3.8.5
| |
| 'h SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
| |
| SR 3.8.5.1 --- ------------ -NOTE- - --- - -- --- ---
| |
| The following SRs are not required to be performed: SR 3.8.4.6 SR 3.8.4.7, and SR 3.8.4.8.
| |
| For DC sources required to be OPERABLE the In accordance following SRs are applicable: with applicable SRs SR 3.8.4.1 SR 3.8.4.4 SR 3.8.4.7 SR 3.8.4.2 SR 3.8.4.5 SR 3.8.4.8.
| |
| SR 3.8.4.3 SR 3.8.4.6 O
| |
| V i
| |
| ,O FERMI - UNIT 2 3.8-21 Amendment No. 134
| |
| | |
| Battery Cell Parameters 3.8.6
| |
| / 3.8 ELECTRICAL POWER S GTEMS 3.8.6 Battery Cell Parameters l LC0 3.8.6 Battery cell parameters for the Division I and Division II batteries shall be within liinits.
| |
| APPLICABILITL W.".ar, associated DC electrica'i oower subsystems are required to be OPERABLE.
| |
| ACTIONS )
| |
| i
| |
| .....................................N0Tr-- --- - - -- --- --------- ----
| |
| Separate Condition entry is allowed for each battery.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i
| |
| A. One or more batteries A.1 Verify pilot cells 1 hour l p/
| |
| w.
| |
| , with one rc more battery ce l electrolyte level and float voltage meet parameters not within Table 3.8.6 1 4
| |
| -Table 3.8.6-1 Category Category C limits. '
| |
| A or B limits. l ANQ A.2 Verify battery cell 24 hours parameters meet Table 3.8.6 1 AND Categcry C limits.
| |
| Once per 7 days thereafter AND A.3 Restore battery cell 31 days parameters to Table 3.8.6 1 Category A and B limits.
| |
| (coittinued)
| |
| (V FERMI UNIT 2 3.8 22 Amendment No. 134
| |
| | |
| Battery Cell Parameters 3.8.6 O actions (comt4""eo)
| |
| CONDITION REQUIRED Ar'. inn COMPLETION TIME
| |
| .B. Required Action ar.d B.1 Declare associated Immediately associated Ccapletion battery inoperable.
| |
| Time of Condition A not met.
| |
| 0B One or more batteries with average electrolyte temperature of the representative cells not within limits.
| |
| O_R.
| |
| One or more batteries with one or more battery cell parameters not within D Table 3.8.6-1 (s)' Category C values.
| |
| .* = gy -
| |
| SURVEILLANCE REQUIREMENTS _
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify oattery cell parameters meet 7 days Table 3.8.6-1 Category A limits.
| |
| (continued)
| |
| FERMI - UNIT 2 3.8 23 Amendment No. 134 m
| |
| | |
| Battery Cell Parameters 3.8.6
| |
| {
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet 92 days Table 3.8.6 1 Category B limits.
| |
| AN_Q ,
| |
| 1 Once within I 24 hours after battery discharge
| |
| < 105 V 8ND Once within 24 hours after battery overcharge
| |
| > 150 V for Division I and
| |
| > 145 V for
| |
| , Division II k
| |
| SR 3.8.6.3 Verify average electrolyte temperature of 92 days representative cells is > 60*F. l I
| |
| O FERMI - UNIT 2 3.8 24 Amendment No. 134
| |
| | |
| l Battery Cell Parameters 3.8.6 f3 V Table 3.8.6 1 (page 1 of 1)
| |
| Battery Cell Parameter Requirements {
| |
| I CATEGORY A: CATEGORY C:
| |
| LIMITS FOR EACH CATEGORY B: ALLOWABLE -
| |
| DESIGNATED PILOT LIMITS FOR EACH LIMITS FOR EACH PARAMETER CELL CONNECTED CELL CONNECTED CELL l
| |
| Electrolyte > Minimum level > Minimum level Above top of Level indication mark, and indication mark, plates, and not s V inch above and s V inch above overflowing maximum level maximum level indication mark (a) indication mark (a) .
| |
| Float Voltage = 2.13 V a 2.13 V > 2.07 V i Specific = 1.195(c) = 1.190 Not more than Gravity (b) 0.020 below average of all ANJ n , connected cells Average of all connected cells ANA
| |
| > 1.200 Average of all connectid
| |
| = 1.190 c)cel.5 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum level durir.g equalizing charges provided it is not overflowing.
| |
| (b) Corrected for electrolyte temperature and level.
| |
| '(c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits.
| |
| th U
| |
| FERMI.- UNIT 2 3.8-25 Amendment No. 134
| |
| | |
| = ~ ~
| |
| g Distribution Systems-Operating 3.8.7 l
| |
| h 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution Systems-Operating i
| |
| LC0 3.8.7 The following Division I and Division II AC and DC -
| |
| electrical power distribution subsystems shall be OPERABLE:
| |
| : a. AC electrical power distribution subsystems: 1 Divirion I p_ivision II
| |
| : 1. 4160 V Buses 11EA 12EB 13EC. 14ED 64B. 64C 65E. 65F
| |
| : 2. 480 V Buses 72EA 72EB 72EC 72ED 72B, 72C 72E 72F
| |
| : 3. 120 V MPU 1 MPU 2
| |
| : b. DC electrical power distribution subsystems: I Division I Division II
| |
| : 1. 130 V Distribution 2PA-2 2PB-2
| |
| , Cabinet b 2. 260 V MCC 2PA-1 2PB-1 APPL 7CABILITY: MODES 1, 2 and 3. (
| |
| SW CONDITION REQUIRED ACTION COMPLETION TIME I A. One or more required A.1 Restore AC electrical 8 hours AC electrical power power distribution distribution subsystem (s) to AND subsystems inoperable. OPERABLE status.
| |
| 16 hours from discovery of failure to meet LC0 l
| |
| j (continued)
| |
| ,O )
| |
| ..V FERMI - UNIT 2 3.8-26 Amendmant No. 134 l
| |
| | |
| 1 Distribution Systems-Operating 3.8.7 ACTIONS (continurgi)
| |
| CONDITi0ti REQUIRED ACTION COMPLETION TIME
| |
| , B. -One or more required B.1 Restore DC electrical 2 hours DC electrical power power distribution distribution subsystem (s) to ale subsystems inoperable. OPERABLE status.
| |
| 16 hours from discovery of failure to meet LC0 s
| |
| l C. Required Action and C.1 Be in MODE 3. 12 hours i associated Completion Time of Condition A 8lQ or B not met.
| |
| C.2 Be in MODE 4. 36 hours Two or more required D.1 Enter LC0 3.0.3. Immediately O electrical power
| |
| ' D. distribution
| |
| -subsystems inoperable that result in a loss
| |
| .of function.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct breaker alignments and 7 days
| |
| ' voltage to required AC and DC electrical power distribution subsystems.
| |
| l O
| |
| FERMI UNIT-2; 3.8 27 Amendment No. 134
| |
| | |
| Distribution Systems-Shutdown 3.8.8
| |
| - (m) 3.8 P/OF: CAL POWER SYSTEMS 3.8.8 Distribution Systems-Shutdown LC0 3.8.8 The necessary portions of the AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.
| |
| APPLICABILITY.: MODES 4 and 5.
| |
| During movement of irradiated fuel assemblies in the secondary containment.
| |
| ACTIONS
| |
| .......................................N0TE----------------- ------- - ----- -
| |
| LC0 3.0.3 is not applicable.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required power distribution feature (s) subsystems inoperable. inoperable.
| |
| E A.2.1 Suspend CORE Immediately ALTERATIONS.
| |
| AND A.2.2 Suspend handling of Immediately irradiated fuel assemblies in the secondary containment.
| |
| AND (continued) j l
| |
| l V(3 i i
| |
| FERMI UNIT 2 3.8 28 Amendment No. 134 j i
| |
| )
| |
| | |
| 1 Distribution Systems-Shutdown 3.8.8 l
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.3 Initiate action to Immediately-suspend operations with a potential for draining the reactor vessel.
| |
| 6ND A.2.4 Initiate actions to Immediately restore required AC j and DC electrical l power distribution subsystems to OPERABLE status.
| |
| AND A.2.5 Declare associated Immediately required shutdown cooling subsystem (s) inoperable and not in 5O operation. .
| |
| l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.
| |
| 1%
| |
| ,V FERMI - UNIT 2 3.8 29 Amendment No. 134
| |
| | |
| Refueling Equipment Interlocks 3.9.1 r
| |
| 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks LC0 3.9.1 The refueling equipment interlocks associated with the refuel msition of the reactor mode switch shall be OPERABLE.
| |
| APPLICABILITY: During in-vessel fuel movement with equipment associated with the interlocks when the reactor mode switch is in '
| |
| the refuel position. ,
| |
| ACTIONS I
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Suspend in-vessel Immediately refueling equipment fuel movement with interlocks inoperable. eguipment associated with the inoperable interlock (s).
| |
| 1 l
| |
| O FERMI UNIT 2 3.9 1 Amendment No. 134
| |
| | |
| p-Refueling Equipment Interlocks 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of 7 days )
| |
| the following required refueling equipment interlock inputs:
| |
| : a. All-rods in,
| |
| : b. Refuel platform position, l l
| |
| : c. Refuel platform fuel grapple, fuel loaded,
| |
| : d. . Refuel platform fuel grapple not fully retracted position,
| |
| : e. Refuel platform frame mounted hoist, fuel loaded, and
| |
| : f. Refuel platform monorail mounted hoist, fuel loaded.
| |
| l l
| |
| l
| |
| /'N FERMI UNIT 2 3.9 2 Amendment No. 134
| |
| | |
| Refuel Position One Rod Out Interlock 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Refuel Position One Rod Out Interlock LC0- 3.9.'2' The refuel position one rod out interlock shall be OPERABLE.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
| |
| ACTIONS CONDITION' REQUIRED ACTION COMPLETION TIME 1
| |
| ' A. Refuel position one- A.1 Suspend control rod Immediately rod out interlock withdrawal.
| |
| inoperable.
| |
| NLD A.2 Initiate action to Immediately fully insert all O insertable control V rods in core cells containing one or more fuel assemblies.
| |
| I i
| |
| O FERMI UNIT 2 3.9 3 Amendment No. 134
| |
| | |
| Refuel Position One Rod Out Interlock .
| |
| 3.9.2 I O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked in Refuel 12 hours position.
| |
| SR 3.9.2.2 ----- - - -- -
| |
| NOTE ----- -- ---- ---
| |
| l Not required to be performed until 1 hour after any control rod is withdrawn.
| |
| Perform CilANNEL FUNCTIONAL TEST. 7 days O
| |
| t I
| |
| ._O FERMI UNIT 2 3.9-4 Amendment No. 134
| |
| | |
| D i p Control Rod Position 3.9.3
| |
| '3.9 REFUELING OPERATIONS' 3.9.3 . Control Rod Position L
| |
| LC0''3.9.3' All control rods shall be fully inserted.
| |
| * APPLICABILITY: When loading fuel' assemblies into the core. .
| |
| ]
| |
| l ACTIONS l
| |
| l 'C0WITION REQUIRED ACTION COMPLETION TIME A. One or more control A.1 Suspend loading fuel Immediately -
| |
| rods not fully assemblies into the inserted. core. )
| |
| l O
| |
| l' l.
| |
| l v SURVEILLANCE REQUIREMENTS l- SURVEILLANCE ~ FREQUENCY l
| |
| 'SR 3.9.'3.1 Verify all' control rods are fully inserted. 12 hours l
| |
| l b
| |
| (
| |
| L i
| |
| 4 O
| |
| FERMI . UNIT 2. 3.9-5 Amendment No. 134 e
| |
| | |
| Control Rod Position Indication 3.9.4 (j 3.9 REFUELING OPERATIONS j 1
| |
| 3.9.4 Control Rod Position Indication l LC0 3.9.4 The control rod " full-in" position indication channel for each control rod shall be OPERABLE.
| |
| APPLICABILITY: MODE 5.
| |
| i ACTIONS j
| |
| ..................................... NOTE ---- - --------- -- - --------------
| |
| Separate Condition entry is allowed for each required channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more control A.1.1 Suspend in vessel Immediately rod " full-in" position fuel movement.
| |
| f)
| |
| N/
| |
| indication channels inoperable. I 6NQ
| |
| )
| |
| A.1.2 Sus)end control rod Immediately wit 1drawal.
| |
| AND A.1.3 Initiate action to Immediately I fully insert all insertable control rods in core cells I containing one or more fuel assemblies.
| |
| QR (continued) ,
| |
| i A
| |
| V_
| |
| FERMI UNIT 2 3.9 6 Amendment No. 134 L
| |
| | |
| Control Rod Position Indication 3.9.4 I
| |
| P G ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Initiate action to Immediately fully insert the control rod
| |
| . associated with the l inoperable " full in" position indicator.
| |
| ANQ l A.2.2 Initiate action to Immediately disarm the control I
| |
| rod drive associated with the fully inserted control rod.
| |
| l y SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY l
| |
| SR 3.9.4.1 Verify the required channel has no Each time the
| |
| " full-in" indication on each control rod control rod is that is not " full in." withdrawn from l the " full in" position 1
| |
| 1 1
| |
| FERMI UNIT 2 3.9 7 Amendment No. 134 L
| |
| | |
| Control Rod OPERABILITY-Refueling 3.9.5 m
| |
| (d 3.9 REFUELING OPERATIONS
| |
| _3.9.5 _ Contr01 Rod OPERABILITY-Refueling LC0 3.9.5 Each withdrawn control rod shall be OPERABLE.
| |
| I APPLICABILITY: MODE 5.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to Immediately control rods fully insert inoperable. inoperable withdrawn-control rods.
| |
| 1 O S_URVEILLANCE REQUIREMENTS I
| |
| J SURVEILLANCE FREQUENCY l SR 3.9.5.1 -
| |
| --------- --NOTE--- --- ------ -----
| |
| Not required to be performed until 7 days after the control rod is withdrawn.
| |
| Insert each withdrawn control rod at least 7 days one notch.
| |
| SR-'3.9.5.2~ ' Verify each; withdrawn control rod scram
| |
| - 7 days accumulator pressure is a 940 psig.
| |
| O FERMI -. UNIT 2' 3.9 8 Amendment No. 134 L .,
| |
| | |
| RPV Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Reactor Pressure-Vessel (RPV) Water Level LCO .3.9.6' RPV water level shall be = 20 ft 6 inches above the top' of the RPV flange. l l
| |
| APPLICABILITY: During movement of irradiated fuel assemblies within the I RPV,-
| |
| During movement of new fuel ' assemblies or handling of I control rods within the RPV, when irradiated fuel <
| |
| assemblies are seated within the RPV.
| |
| ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not A.1 Suspend movement of Immediately within limit. fuel assemblies and b
| |
| d handling of control !
| |
| rods within the RPV.
| |
| l 4
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is a 20 ft 6 inches 24 hours above the top of the RPV flange.
| |
| O FERMI UNIT-2 3.9 9 Amendment No. 134
| |
| | |
| RHR-High Water Level 3.9.7 1
| |
| m (j 3.9 REFUELING OPERATIONS l 3.9.7 Residual Heat Removal (RHR)-High Water Level l
| |
| l LC0 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE. -
| |
| l I
| |
| APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV), the water level = 20 ft 6 inches above the top of the RPV flange, and heat losses to ambient not greater than or equal to heat input to reactor coolant.
| |
| l
| |
| ! ACTIONS
| |
| : CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| l l
| |
| A. Required RHR shutdown A.1 Verify an alternate 1 hour ;
| |
| cooling subsystem method of decay heat 1 inoperable. removal is available. AND A Once per V 24 hours thereafter l
| |
| l (continued) l l
| |
| l l
| |
| I h l L/
| |
| 1 -
| |
| 1 FERMI UNIT 2 3.9 10 Amendment No. 134
| |
| | |
| 7.
| |
| RHR-High Water Level J 3.9.7 i
| |
| [} ACTIONS (continued) v CONDITION REQUIRED ACTION COMPLETION TIME ;
| |
| l B. Required Action and B.1 Suspend loading Immediately ,
| |
| associated Completion irradiated fuel ;
| |
| Time of Condition A assemblies into the !
| |
| not met. RPV. j AND B.2 Initiate action to Immediately restore secondary containment to OPERABLE status. j AND B.3 Initiate action to Immediately l restore one standby gas treatment subsystem to OPERABLE status.
| |
| ^
| |
| ( AND U) l B.4 Initiate action to Immediately j restore isolation !
| |
| capability in each required secondary containment l penetration flow path i not isolated. ;
| |
| l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify the RHR shutdown cooling subsystem 12 hours is capable of decay heat removal.
| |
| v FERMI - UNIT 2 3.9 11 Amendment No. 134
| |
| | |
| 1 i
| |
| RHR-Low Water Level 3.9.8 U' 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-Low Water Level LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, one RHR shutdown cooling subsystem shall be in operation. 1 1
| |
| ............................N0TES -----
| |
| l
| |
| : 1. The required operating RHR shutdown cooling subsystem I may be removed from operation for up to 2 hours per f 8 hour period. ,
| |
| 1
| |
| : 2. One RHR shutdown cooling subsystem may be inoperable I for up to 2 hours for the performance of Surveillances. j
| |
| ............................................................ j APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV), the water level < 20 ft 6 inches above the top of the RPV flange, and heat losses to ambient not greater than or equal to heat input to reactor coolant.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
| |
| l A. One or two required A.1 Verify an alternate 1 hour i RHR shutdown cooling method of decay heat I subsystems inoperable, removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter (continued) b, , .
| |
| FERMI - UNIT 2 3.9 12 Amendment No. 134
| |
| | |
| p-RHR-Low Water Level 3.9.8 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Initiate action to Immediately associated Completion restore secondary l Time of Condition A containment to not met. OPERABLE status.
| |
| AND B.2 Initiate action to Immediately restore one standby gas treatment subsystem to OPERABLE status.
| |
| AND 4 I
| |
| B.3 Initiate action to Immediately i restore isolation capability in each required secondary I containment i G penetration flow path C not isolated. !
| |
| C. No RHR shutdown C.1 Initiate action to Immediately cooling subsysi P in restore one RHR operation. shutdown cooling subsystem or one AND recirculation pump to operation.
| |
| No recirculation pump in operation. AND l C.2 Verify reactor 1 hour from l coolant circulation discovery of no by an alternate reactor coolant method. circulation AND C.3 Monitor reactor Once per hour coolant temperature.
| |
| O FERMI UNIT 2 3.9 13 Amendment No. 134
| |
| | |
| r . j t
| |
| RHR-Low Water Level 3.9.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours or recirculation pump is operating.
| |
| SR 3.9.8.2 Verify each RHR shutdown cooling subsystem 12 hours is capable of decay heat removal. l l
| |
| l O !
| |
| i iO FERMI'- UNIT 2- 3.9 14 Amendment No. 134
| |
| | |
| Inservice Leak and Hydrostatic Testing Operation i 3.10.1 1 3.10 SPECIAL OPERATIONS 3.10.1- Inservice Leak and Hydrostatic Testing Operation LC0 3.10.1 The average reactor coolant tem)erature specified in -
| |
| Table 1.11 for MODE 4 may be clanged to "NA," and operation considered not to be in MODE 3: and the requirements of 4 LC0 3.4.9, " Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown," may be suspended, ~ to allow !
| |
| performance of an inservice leak or hydrostatic test j provided the following MODE 3 LCOs are met: ;
| |
| 1
| |
| : a. LC0 3.3.6.2, " Secondary Containment Isolation '
| |
| Instrumentation," Functions 1, 3, and 4 of Table 3.3.6.21:
| |
| : b. LC0 3.6.4.1, " Secondary Containment":
| |
| : c. LC0 3.6.4.2. " Secondary Containment Isolation Valves (SCIVs)"; and
| |
| : d. LCO 3.6.4.3, " Standby Gas Treatment (SGT) System."
| |
| /~'
| |
| G APPLICABILITY: MODE 4 with average reactor coolant temperature > 200*F.
| |
| I
| |
| [
| |
| FERMI - UNIT 2 3.10 1 Amendment No. 134
| |
| | |
| i Inservice Leak and Hydrostatic Testing Operation 3.10.1 O
| |
| J ACTIONS
| |
| .....................................N0re.....................................
| |
| Separate Condition entry is allowed for each requirement of the LCO.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME J
| |
| A. One or more of the A.1 - .
| |
| .--NOTE-- ------
| |
| above requirements not Required Actions to met. be in MODE 4 include reducing average l reactor coolant temperature to s 200*F.
| |
| Enter the applicable Immediately Condition of the l affected LC0.
| |
| QB
| |
| ,e A.2.1 Suspend activities Immediately
| |
| \ that could increase the average reactor coolant temperature or pressure.
| |
| AND A.2.2 Reduce average 24 hours reactor coolant temperature to s 200*F.
| |
| t v
| |
| )
| |
| FERMI - UNIT 2 3.10 2 Amendment No. 134
| |
| | |
| r _..;
| |
| Inservice Leak and Hydrostatic Testing Operation 3.10.1 SURVEILLANCE REQUIREMENTS
| |
| (
| |
| l SURVEILLANCE FREQUENCY l
| |
| t l SR 3.10.1.1 Perform the applicable SRs for the required According to MODE 3 LCOs. the applicable SRs l
| |
| l O
| |
| l l
| |
| l O
| |
| FERMI UNIT 2 3.10 3 Amendment No. 134
| |
| | |
| r
| |
| . 1 Reactor Mode Switch Interlock Testing 3.10.2 h 3.10 SPECIAL OPERATIONS 3.10.2 Reactor Mode Switch Interlock Testing LC0 3.10.2 The reactor mode switch position specified in Table 1.1:1 for MODES 3, 4..and 5 may be changed to include the run, startup/ hot standby, and refuel position, and operation considered not to be in MODE 1 or 2, to allow testing of-instrumentation associated with the reactor mode switch.
| |
| interlock functions, provided:
| |
| : a. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and
| |
| : b. No CORE ALTERATIONS are in progress. ,
| |
| 1
| |
| ' APPLICABILITY: MODES 3 and 4 with the reactor mode switch in the run.
| |
| startup/ hot standby, or refuel position, MODE 5.with the reactor mode switch in the run or startup/ hot standby position.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME i
| |
| A. One or more of the A.1 Suspend CORE Immediately above requirements not ALTERATIONS except met, for control rod insertion.
| |
| AN_Q A.2 Fully' insert all 1 hour insertable control rods in core cells containing one or more fuel assemblies.
| |
| 8NQ (continued)
| |
| O FERMI.-LUNIT 2
| |
| . 3.10 4 Amendment No. 134
| |
| | |
| Reactor Mode Switch Interlock T'esting 3.10.2 19 v
| |
| ACTIONS CONDITION REQUIRED ACTION COMPL'dTION TIME A. (continued) A.3.1 Place the rE2Ctor 1 hour -
| |
| mode switch in the shutdown position. 4 E
| |
| A.3.2 ------- NOTE- --- ---
| |
| Only applicable in MODE 5.
| |
| Place the reactor 1 hour mode switch in the refuel position.
| |
| N SURVEILLANCE REQUIREMENTS (d.
| |
| SURVEILLANCE FREQUENCY SR 3.10.2.1 Verify all control rods are fully inserted 12 hours in core cells containing one or more fuel assemblies.
| |
| SR 3.10.2.2 Verify no CORE ALTERATIONS are in progress. 24 hours n
| |
| l
| |
| . fN \
| |
| L) . l l
| |
| FERMI - UNIT 2 3.10 5 Amen er t No. 134
| |
| | |
| n 1
| |
| Single Control Rod Withdrawal-Hot Shutdovo 3.10.3 ,
| |
| h 3.10 SPECIAL OPERATIONS 3.10.3 Single Control Rod Withdrawal-Hot Shutdown LC0 3.10.3 The reactor mode switch position specified in Table 1.1-1 l for MODE 3 may oe changed to include the refuel position.
| |
| ardi .,aratic; considered not to be in MODE 2, to allow w)thdrawal of a single control rod, provided the following requirements are met:
| |
| : a. LC0 3.9.2, " Refuel Position One Rod-Out Interlock";
| |
| : b. LC0 3.9.4, " Control Rod Position Indication":
| |
| : c. All other control rods are fully inserted; and
| |
| : d. 1. LC0 3.3.1.1. " Reactor Protection System (RPS)
| |
| Instrumentation." MODE 5 requirements for Functions 1.a. 1.b. 8.a. 8.b, 11. and 12 of Table 3.3.1.1 1, and LC0 3.9.5, " Control Rod OPERABILITY-Refueling."
| |
| E O
| |
| b 2. All other control rods in a five by five array !
| |
| centered on the control rod being withdrawn are j disarmed; at which time LC0 3.1.1, " SHUTDOWN MARGIN '
| |
| (SDM)." MODE 3 requirements, may be changed to allow the single control rod withdrawn to be assumed to be l the highest worth control rod. J APPLICABILITY: MODE 3 with the reactor mt de switch in the refuel position.
| |
| i FERMI UNIT 2 3.10 6 Amendment No. 134
| |
| | |
| Single Control Rod Withoi ,,i -Hat Shutdown 3.10.3 ACTIONS
| |
| .....................................N01E.............................. ...... !
| |
| Separate Condition entry is allowed for each requirement of the LCO.
| |
| l CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more of the A.1 ..- --..N0TES........ !
| |
| above requirements not 1. Required Actions met. to fully insert all insertable !
| |
| control rods i include placing !
| |
| the reactor mode switch in the shutdown position. ,
| |
| i
| |
| : 2. Only applicable if the requirement not met is a required LCO.
| |
| O V Enter the applicable Immediately Condition of the affected LCO.
| |
| OB A.2.1 Initiate action to Immediately fully insert all insertable control rods.
| |
| AND l
| |
| A.2.2 Place the reactor 1 hour i mode switch in the shutdown position.
| |
| O
| |
| %)
| |
| FERMI . UNIT 2 3.10 7 Amendment No. 134 l 4
| |
| L.
| |
| | |
| 1 Single Control Rod Withdrawal-Hot Shutdown 3.10.3 )
| |
| th SURVEILLANCE REQUIREMENTS !
| |
| SURVEILLANCE FREQUENCY i
| |
| SR 3.10.3.1 Perform the applicable SRs for the required According to LCOs. the applicable SRs SR 3.10.3.2 - .-----.--..- - . NOTE----- ---.---- ---
| |
| Not required to be met if SR 3.10.3.1 is satisfied for LC0 3.10.3.d.1 requirements.
| |
| l Verify all control rods, other than the 24 hours control rod being withdrawn, in a five by I five array centered on the control rod being withdrawn, are disarmed.
| |
| SR 3.10.3.3 Verify all control rods, other than the 24 hours O control rod being withdrawn, are fully V , inserted.
| |
| i i
| |
| _C/
| |
| FERMI - UNIT 2 3.10 8 Amendment No. 134
| |
| | |
| Single Control Rod Withdrawal-Cold Shutdown 3.10.4 h 3.10 SPECIAL OPERATIONS 3.10.4 Single Control Rod Withdrawal-Cold Shutdown
| |
| ]
| |
| LC0 3.10.4 The reactor mode switch position specified in Table 1.1' 1 i for MODE 4 may be changed to include the refuel position. l and operation considered not to be in MODE 2, i:o allow '
| |
| withdrawal of a single control rod, and subsecuent removal of the associated control rod drive (CRD) if cesired, provided the following requirements are met:
| |
| : a. All other control rods are fully inserted:
| |
| : b. 1. LC0 3.9.2, " Refuel Position One Rod-Out Interlock,"
| |
| and LC0 3.9.4, " Control Rod Position Indication."
| |
| : 2. A control rod withdrawal block is inserted;
| |
| : c. 1. LC0 3.3.1.1, " Reactor Protection System (RPS)
| |
| Instrumentation." MODE 5 requirements for Functions -1.a.1.b, 8.a. 8.b.11, and 12 of (O./ Table 3.3.1.1-1, and LC0 3.9.5, " Control Rod OPERABILITY-Refueling,"
| |
| 1
| |
| : 2. All other control rods in a five by five array centered on the control rod being withdrawn are .
| |
| disarmed; at which time LC0 3.1.1, " SHUTDOWN MARGIN I (SDM) " MODE 4 requirements, may be changed to allow the single control rod withdrawn to be assumed to be the highest sorth control rod.
| |
| APPLICABILITY: MODE 4 with the reactor mode switch in the refuel position.
| |
| A V.
| |
| FERMI UNIT 2 3.10 9 Amendment No. 134
| |
| | |
| c ,
| |
| i Single Control Rod Withdrawal-Cold Shutdown 3.10.4 l
| |
| ACTIONS
| |
| .....................................N0TE -
| |
| : Separate Condition entry is allowed for each requirement of the LCO. {
| |
| CONDITION REQUIRED ACTION COMPLETION TIME h A. One or more of the A.1 - ---NOTES - - ---
| |
| above requirements not 1. Required Actions met with the affected to fully insert control rod- all insertable insertable. control rods include placing the reactor mode switch in the shutdow:
| |
| position.
| |
| : 2. Only applicable if the requirement not met-is a required Enter the applicable Immediately i Condition of the affected LC0.
| |
| E A.2.1 Initiate action to Immediately fully insert all
| |
| ' insertable control rods.
| |
| AND A.2.2 Place the reactor 1 hour mode switch in the shutdown position.
| |
| I (continued)
| |
| O. :
| |
| FERMI UNIT 2 3.10 10 Amendment No. 134
| |
| | |
| Single Control Rod Withdrawal-Cold Shutdown 3.10.4 ACTIONS (continued)
| |
| U(9 C0WITION REQUIRED ACTION COMPLETION TIME
| |
| >=
| |
| B. One or more of the B.1 Suspend withdrawal of Immediately-above requirements not the control rod and met with the affected removal of associated control rod not CRD.
| |
| insertable.
| |
| 8ND B.2.1 Initiate action to Immediately fully insert all control rods.
| |
| 08 B.2.2 Initiate action to Immediately satisfy the' requirements of this LCO.
| |
| O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.4.1 Perform the applicable SRs for the required According to LCOs. the applicable SRs SR 3.10.4.2 --
| |
| -- ---- - - - NOTE -- ---- - ----
| |
| . Not required to be met if SR 3.10.4.1 is satisfied for LCO 3.10.4.c.1 requirements.
| |
| Verify all control rods, other than the 24 hours control rod being withdrawn. in a five by five array centered on the control rod being withdrawn, are disarmed.
| |
| (continued)
| |
| FERMI - UNIT 2 3.10 11 Amendment No. 134 1
| |
| J
| |
| | |
| r . 1 Single Control Rod Withdrawal-Cold Shutdown 3.10.4 SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.10.4.3 Verify all control rods, other than the 24 hours !
| |
| control rod being withdrawn, are fully '
| |
| inserted.
| |
| SR -3.10.4.4 -- ------------ NOTE ------ ----- ---
| |
| Not required to be met if SR 3.10.4.1 is satisfied for LC0 3.10.4.b.1 requirements.
| |
| Verify a control rod withdrawal block is 24 hours inserted.
| |
| l FERMI - UNIT 2 3.10 12 Amendment No. 134 L
| |
| | |
| Single CRD Removal-Refueling 3.10.5 l 3.10' SPECIAL OPERATIONS 3.10.5' Single Control Rod Drive (CRD) Removal-Refueling
| |
| . LCO 3.10.5 The' requirements of LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation": LC0 3.3.8.2, " Reactor Protection System (RPS) Electric Power Monitoring": and LCO 3.9.5,
| |
| " Control Rod OPERABILITY-Refueling," may be suspended in MODE 5 to allow withdrawal of a' single control rod, and subsequent. removal of the associated CRD from a core cell containing one or more fuel assemblies, provided the following requirements are met:
| |
| : a. All other control rods are fully inserted: and
| |
| : b. All' other control rods in a five by five array centered on the withdrawn control rod are disarmed; at which time LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," MODE 5 requirements may be changed to allow the single control rod withdrawn to be assumed to be the highest worth control rod.
| |
| AND In conjunction with a. and b. above, the requirements of LC0 3.9.1, " Refueling Equipment Interlocks": LC0 3.9.2.
| |
| O- " Refuel Position One Rod Out Interlock"; and LC0 3.9.4,
| |
| " Control Rod Position Indication": may be suspended provided the following requirements are met:
| |
| : c. No other CORE ALTERATIONS are in progress; and
| |
| : d. A control rod withdrawal block is inserted.
| |
| APPLICABILITY: MODE 5 with LC0 3.9.5 not met. !
| |
| O_
| |
| ' FERMI - UNIT 2 3.10 13 Amendment No. 134
| |
| | |
| q 1 ,
| |
| Single CRD Removal-Refueling l 3.10.5 ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 Suspend removal of Immediateli above requirements not the CRD mechanism.
| |
| met.
| |
| 81 0
| |
| - A.2.1 - Initiate action to' Immediately fully insert all control rods.
| |
| QB A.2.2 Initiate action to Immediately satisfy the requirements of this LCO.
| |
| O l 1
| |
| FERMI,' UNIT 2 3.10 14 Amendment No. 134
| |
| | |
| r 3
| |
| Singl.e CRD Removal-Refueling 3.10.5
| |
| ' SURVEILLANCE REQUIREMENTS _
| |
| SURVEILLANCE FREQUENCY SR 3.10.5.1 Verify all control rods, other than the 24 hours control rod withdrawn for the removal of the associated CRD. are fully inserted.
| |
| SR 3.10.5.2 ' erify all control rods, other than the V 24 hours control rod withdrawn for the removal of the associated CRD. in a five by five array centered on the control rod withdrawn for the removal of the associated CRD, are 1 disarmed.
| |
| SR 3.10.5.3 Verify a control rod withdrawal block is 24 hours inserted.
| |
| SR 3.10.5.4 Perform SR 3.1.1.1. According to SR 3.1.1.1 SR 3.10.5.5 Verify no other CORE ALTERATIONS are in 24 hours progress.
| |
| 1 4
| |
| FERMI UNIT 2 3.10 15 Amendment No. 134
| |
| | |
| 7 q
| |
| Multiple Control Rod Withdrawal-Refueling 3.10.6
| |
| ( 3.10 SPECIAL OPERATIONS 3.10.6' Multiple Control Rod Withdrawal-Refueling LCO 3.10.6 The requirements of LC0 3.9.3, " Control Rod Position": -
| |
| LC0 3.9.4, " Control Rod Position Indication"; and LC0 3.9.5.
| |
| " Control Rod OPERABILITY-Refueling," may be suspended, and the " full in" position indicators may be bypassed for any number of control rods in MODE 5. to allow withdrawal (withdrawal only, or withdrawal including removal) of these control rods, removal of associated control rod drives (CRDs), or both, provided the following requirements are met:
| |
| : a. The four fuel assemblies are removed from the associated core cells; and
| |
| : b. All other control rods in core cells containing one or more fuel assemblies are fully inserted.
| |
| APPLICABILITY: MODE 5 with LC0 3.9.3 LC0 3.9.4, or LC0 3.9.5 not met.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME >
| |
| A. One or more of the A.1 Suspend withdrawal of Immediately above requirements not control rods and met, removal of associated CRDs.
| |
| AND A.2 Suspend loading fuel Immediately assemblies. j AND )
| |
| (continued) i t
| |
| i j
| |
| (O_/ i FERMI - UNIT 2 3.10 16 Amendment No. 134 i
| |
| | |
| ~
| |
| l Multiple Control Rod Withdrawal-Refueling 3.10.6 ACTIONS C]'
| |
| L CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3.1 Initiate action to Immediately fully insert all control rods in core cells containing one or more fuel assemblies.
| |
| E A.3.2 Initiate action to Immediately satisfy the requirements of this l LCO.
| |
| 1 SURVEILLANCE REQUIREMENTS l
| |
| () SURVEILLANCE FREQUENCY SR 3.10.6.1 Verify the four fuel assemblies are removed 24 hours from core cells associated with each control rod or CRD removed.
| |
| SR 3.10.6.2 Verify all other control rods in core cells 24 hours containing one or more fuel assemblies are fully inserted.
| |
| ~)
| |
| FERMI UNIT 2 3.10-17 Amendment No. 134
| |
| | |
| p .
| |
| ) i SDM Test-Refueling 3.10.7 h 3.10 SPECIAL OPERATIONS-3.10.7 SHUTDOWN MARGIN (SDM) Test-Refueling LCO 3.10.7 The reactor mode switch position specified in Table 1.1'1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:
| |
| -a. LCO 3.3.1.1, " Reactor Protection System Instrumentation," MODE 2 requirements for Functions 2.a.
| |
| 2.d. and 2.e of Table 3.3.1.1 1:
| |
| : b. 1. LC0 3.3.2.1, " Control Rod Block Instrumentation "
| |
| MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with the prescribed withdrawal sequence requirements of SR 3.3.2.1.7 changed to require'the control rod sequence to conform to t.he SDM test sequence.
| |
| 08
| |
| : 2. .Conformance to the approved control rod sequence for the SDM test is verified by a second licensed O.
| |
| operator or other qualified member of the technical l staff:
| |
| : c. Each withdrawn control rod shall be coupled to the associated CRD:
| |
| : d. All control rod withdrawals during local critical testing shall be made in notch out mode:
| |
| : e. No other CORE ALTERATIONS are in progress; and
| |
| : f. CRD charging water header pressure m 940 psig.
| |
| APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position, j i
| |
| n i d i FERMI - UNIT 2 -3.10 18 Amendment No. 134 j i
| |
| | |
| l l l
| |
| l SDM Test-Refueling i l 3.10.7 l l
| |
| ( ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. -- ---- NOTE- -- --- ---- -- ---- NOTE -- -- --
| |
| Separate Condition Rod worth minimizer may be entry is allowed for bypassed as allowed by each control rod. LC0 3.3.2.1 " Control Rod Block Instrumentation." if required, to allow insertion One or more of inoperable control rod and control rods not continued operation.
| |
| coupled to its --------- ----------- ------
| |
| l' associated CRD.
| |
| A.1 Fully insert the 3 hours uncoupled control rod.
| |
| AND A.2 Disarm the associated 4 hours l CRD.
| |
| t U B. One or more of the B.1 Place the reactor Immediately above requirements not mode switch in the met for reasons other shutdown or refuel than Condition A. position.
| |
| A
| |
| _N)
| |
| FERMI - UNIT 2 3.10 19 Amendment No. 134 l
| |
| l
| |
| | |
| SDM Test-Refueling 3.10.7 O suave 1 uauce aeou1aeae"'s SURVEILLANCE FREQUENCY SR 3.10.7.1 Perform the MODE 2 applicable SRs for LC0 According to 3.3.1.1, Functions 2.a. 2.d. and 2.e of the applicable Table 3.3.1.1 1. SRs i
| |
| SR 3.10.7.2 ---- ----- --
| |
| - NOTE - - --------- -- -
| |
| Not required to be met if SR 3.10.7.3 satisfied.
| |
| Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable SRs 1
| |
| SR 3.10.7.3 --- - -- ----
| |
| --NOTE--------- ---------
| |
| l Not required to be met if SR 3.10.7.2 i
| |
| ,r m satisfied. l
| |
| (- .
| |
| i Verify movement of control rods is in During control '
| |
| compliance with the approved control rod rod movement sequence for the SDM test by a second licensed owrator or other qualified member of the tec1nical staff.
| |
| SR 3.10.7.4 Verify no other CORE ALTERATIONS are in 12 hours progress.
| |
| (continued) n b
| |
| FERMI - UNIT 2 3.10 20 Amendment No. 134
| |
| | |
| 1 SDM Test-Refueling 3.10.7 4
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY l
| |
| SR 3.10.7.5 Verify each withdrawn control rod does not Each time the 90 to the withdrawn overtravel position. control rod is withdrawn to
| |
| " full out" position AND Prior to i satisfying LCO 3.10.7.c '
| |
| requirement after work on control rod or CRD System that l could affect coupling I
| |
| j fT O SR 3.10.7.6 Verify CRD charging water header pressure
| |
| = 940 psig.
| |
| 7 days !
| |
| I l
| |
| I I
| |
| 1 O
| |
| FERMI UNIT 2 3.10 21 Amendment No. l'34
| |
| | |
| . 1 Design Features 4.0
| |
| ( 4.0 DESIGN FEATURES 4.1- Site Location The Fermi 2 site is located on the western shore of Lake Erie in Frenchtown Township, Monroe County, Michigan, approximately 8 miles east northeast of the city of Monroe, Michigan.
| |
| 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,)
| |
| as fuel material and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs.that have been analyzed with NRC staff approved codes and methods and have been shown by tests cr analyses to comply with all safety design bases. A limited number of lead test assemblies
| |
| .r that have not completed representative testing may be placed in nonlimiting core regions.
| |
| 4.2.2 Control Rod Assemblies The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide and/or hafnium metal as approved by the NRC.
| |
| 4.3 Fuel Storage 4.3.1 Criticality i
| |
| The spent fuel storage racks are designed and shall be maintained I with: , I
| |
| : a. Fuel assemblies having a maximum k-infinity of 1.31 in the normal reactor core configuration at cold conditions:
| |
| e
| |
| .( (continued)
| |
| FERMI UNIT 2 4.0 1 Amendment No. 134
| |
| | |
| 9 Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Stora9e (continued) b.
| |
| k,,,ludes inc an allowance for uncertainties as described ins 0.95 if fully floo Section 9.1 of the UFSAR: and f i
| |
| : c. A nominal 6.22 inch center to center distance between fuel !
| |
| assemblies placed in the high density storage racks and a nominal 11.9 x 6.6 inch center to center distance between ,
| |
| fuel assemblies placed in the low density storage racks.
| |
| 4.3.2 Drainaoe The spent fuel sturage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 660 ft 11,5 inches.
| |
| 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2414 fuel O, assemblies.
| |
| .- 1 i
| |
| l D
| |
| FERMI UNIT 2 4.0 2 Amendment No. 134
| |
| | |
| 5.1 5.0 ADMINISTRATIVE CONTROLS
| |
| .................................... NOTE .. -- ---
| |
| Plant specific titles are designated in the UFSAR for each organizational position listed or described in this Section.
| |
| 5.1 Responsibility 5.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during any absence.
| |
| The Plant Manager or designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment tnat affect nuclear safety.
| |
| 5.1.2 The Nuclear Shift Supervisor (NSS) shall be responsible for the control room command function. During any absence of the NSS from the control room while the unit is in MODE 1, 2. or 3, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function.
| |
| During any absence of the NSS from the control room while the unit r~s is in MODE 4 or 5. an individual with an active SR0 license or V Reactor Operator license shall be designated to assume the control room command function.
| |
| I 1
| |
| O FERMI UNIT 2 5.0 1 Amendment No. 134
| |
| | |
| l 1
| |
| Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Oraanizations
| |
| * Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of tae nuclear power plant.
| |
| : a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all o>erating organization positions. These relationships shall )e documented and updated, as appropriate,'in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall .be documented in the UFSAR:
| |
| : b. The Plant Manager shall be responsible for overall safe operation of the-plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant:
| |
| : c. The Senior Vice President Nuclear Generation shall have j corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety: and
| |
| : d. The individuals who train the operating staff, carry out radiation protection, or perform quality assurance functions may report to the appropriate onsite manager: however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.
| |
| O
| |
| -(continued)
| |
| FERM1 UNIT 2 5.0 2 Amendment No. 134 L
| |
| | |
| q l
| |
| Organization 5.2
| |
| - 5.2 Organization (continued) 5.2.2 Unit Staff The unit staff organization shall include the following: '
| |
| : a. At least two non licensed operators shall be assigned while operating in MODE 1, 2 or 3 and at least one non-licensed operator shall be assigned whenever the reactor contains 4 fuel,
| |
| : b. - At least one licensed Reactor Operator (RO) shall be present
| |
| 'in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2 or 3, at least one <
| |
| licensed Senior Reactor Operator (SRO) shall be present in l the control room. 1
| |
| : c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.g a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew member.s ;
| |
| provided immediate action is taken to restore the shift crew ,
| |
| composition to within the minimum requirements.
| |
| I d. A Radiation Protection Technician shall be on site when fuel
| |
| . is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
| |
| : e. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety related functions (e.g., licensed SR0s, licensed R0s, radiation protection technicians, auxiliary operators, and key maintenance personnel). The controls shall include guidelines on working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals.
| |
| ' Any deviation from the established guidelines shall be authorized in advance by the Plant Manager or designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. ;
| |
| J i
| |
| (continued)
| |
| , / FERMI . UNIT-2 '5.0 3 Amendment No. 134
| |
| | |
| 1 Organization 5.2 5.2 Organization- (continued) 5.2.2 Unit Staff (continued)
| |
| Controls shall be included in the procedures such that '
| |
| individual overtime shall be reviewed monthly by the Plant Manager or designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines I is not authorized. l
| |
| : f. The Superintendent 0perations. Assistant Superintendent-Operations, or the Operations Engineer shall hold an SRO-license.
| |
| : g. An STA shall be assigned whenever the reactor is operating in MODES 1, 2. and 3. The Shift Technical Advisor (STA) shall provide advisory technical support to the Nuclear Shift Supervisor (NSS) in the areas of thermal hydraulics.
| |
| reactor engineering, snd plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
| |
| l O
| |
| l n
| |
| V (continued)
| |
| FERMI - UNIT 2 5.0 4 Amendment No. 134
| |
| | |
| l l Unit Staff Qualifications l 5.3 n 5.0 ADMINISTRATIVE CONTROLS
| |
| ^~
| |
| 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum (
| |
| qualifications of ANSI N18.11971 for comparable positions. bxcept !
| |
| for the Superintendent - Radiation Protection or his designee who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.
| |
| l 1
| |
| I l
| |
| fM L) .
| |
| i I
| |
| i
| |
| ,O FERMI - UNIT 2 5.0 5 Amendment No. 134
| |
| | |
| r q Procedures 5.4
| |
| ,-)
| |
| (
| |
| V-5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established. implemented. and maintained covering the following activities: -
| |
| : a. The applicable procedures recommended in Regulatory Guide 1.33. Revision 2, Appendix A. February 1978:
| |
| : b. The emergency operating procedures required to implement the requirements of NUREG 0737 and to NUREG 0737. Supplement 1.
| |
| as stated in Generic Letter 82 33:
| |
| : c. Quality assurance for effluent and environmental monitoring:
| |
| : d. Fire Protection Program implementation: and
| |
| : e. All programs specified in Specification 5.5.
| |
| O o.
| |
| FERMI UNIT 2 5.0 6 Amendment No. 134 u i
| |
| | |
| p .
| |
| Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented and maintained. ,
| |
| l 5.5.1 Offsite Dose Calculation Manual (0DCM) l
| |
| : a. The ODCM shall contain:
| |
| : 1. the methodology and parameters used in the calculation of l offsite doses resulting from radioactive gaseous and l liquid effluents, in the calculation of gaseous and
| |
| ; liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and l 2. the' radioactive effluent controls and radiological i environmental monitoring activities and descriptions of l
| |
| the information that should be included in the Annual l Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 i
| |
| and Specification 5.6.3.
| |
| : b. Licensee initiated changes to the ODCM:
| |
| : 1. Shall be documented and records of reviews erformed shall be retained. This documentation shal contain:
| |
| : i. sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s), and l 11. a determination that the change (s) maintain the
| |
| ! levels of radioactive effluent control required by l 10 CFR 20.1302, 40 CFR 190. 10 CFR 50.36a and l / 10 CFR 50, Appendix I. and not adversely impact the I accuracy or reliability of effluent, dose, or
| |
| ( setpoint calculations:
| |
| : 2. Shall become effective after the approval of the Plant l Manager; and l
| |
| l (continued)
| |
| FERMI UNIT 2 5.0 7 Amendment No. 134 l
| |
| t
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals
| |
| []
| |
| 5.5.1 Offsite Dose Calculation Manual (0DCM) (continued)
| |
| : 3. Shall be submitted to the NRC in the form of a complete.
| |
| l legible copy of the entire ODCM as a part of or l concurrent with the Radioactive Effluent Release Report l for the period of the report in which any change in the ODCM was made.
| |
| Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the i page that was changed, and shall indicate the date (i.e.,
| |
| month and year) the change was implemented.
| |
| ! 5.5.2 Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to l levels as low as practicable. The systems include Core Spray.
| |
| l High Pressure Coolant Injection. Residual Heat Removal. Reactor Core Isolation Cooling, reactor water sampling, Post Accident e Sampling, reactor water cleanup. Hydrogen Recombiners, Primary l Containment Monitoring, control rod drive discharge headers, and Standby Gas Treatment. The program shall include the following:
| |
| l a. Preventive maintenance and periodic visual inspection requirements; and
| |
| : b. Integrated leak test requirements for each system at refueling cycle intervals or less.
| |
| 5.5.3 Post Accident Samolina This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive iodines, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:
| |
| : a. Training of personnel:
| |
| (continued)
| |
| FERMI - UNIT 2 5.0 8 Amendment No. 134 l
| |
| | |
| r: ]
| |
| Programs and Manuals 5.5 i 1
| |
| 5.5 Programs and Manuals 5.5.3 Post Accident Samolina (continued) l
| |
| : b. Procedures for sampling and analysis: and .
| |
| .c. Provisions for maintenance of sampling and analysis equipment.
| |
| 5.5.4 Radioactive Effluent Controls Prooram i
| |
| This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to inembers of the public from. radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM. shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program l
| |
| shall include the following elements:
| |
| l l a. Limitations on the functional capability of radioactive
| |
| ! liquid and gaseous monitoring instrumentation including i surveillance tests and setpoint determination in accordance with the methodology in the ODCM:
| |
| : b. Limitations 'on the concentrations of radioactive' material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in 10 CFR 20.1001 - 20.2402, Appendix B. Table 2 Column 2:
| |
| : c. Monitoring, sampling, and analysis of radioactive liquid anV gaseous effluents in accordance with 10 CFR 20.1302 andwith the methodology and parameters in the ODCM:
| |
| : d. Limitations on the annual. and quarterly doses or dor .
| |
| commitment to a member of the public from radioacti',a <
| |
| l materials in liquid effluents released to unrestricted l areas, conforming to 10 CFR 50 Appendix I:
| |
| 1
| |
| : e. Determination of cumulative and projected dose contributions i from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days:
| |
| l (continued)
| |
| L-FERMI UNIT 2 5.0 9 Amendment No. 134 4
| |
| | |
| T Programs and Manuals 5.5 O
| |
| V 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Proaram (continued)
| |
| : f. Limitations on the functional capability and use of the ,
| |
| liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I:
| |
| : g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas at or beyond
| |
| , the site boundary conforming to the following:
| |
| : 1. For noble gases: s 500 mrem /yr to the total body and s 3000 mrem /yr to the skin: and
| |
| : 2. For Iodine-131, for Iodine-133. for tritium, and for all radionuclides in particulate form with half lives
| |
| > 8 days: s 1500 mrem /yr to any organ:
| |
| : h. Limitations on the annual and quarterly air doses resulting
| |
| ,e . from noble gases released in gaseous effluents to areas beyond the site boundary, conforming to 10 CFR 50,
| |
| ( Appendix I:
| |
| : i. Limitations on the annual and quarterly doses to a member of the public from iodine 131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to areas beyond the site boundary, confe ming to 10 CFR 50, Appendix I
| |
| : j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190: and i i
| |
| : k. Limitations on venting and purging of the Mark I containment {
| |
| through the Standby Gas Treatment System or the Reactor l Building Ventilation System to maintain releases as low as reasonably achievable, i
| |
| (3 V (continued)
| |
| FERMI UNIT 2 5.0-10 Amendment No. 134
| |
| | |
| F .
| |
| 1 Programs and Manuals 5.5 t
| |
| b v
| |
| 5.5 Programs and Manuals (continued) 5.5.5 Comoonent Cyclic or Transient Limit I
| |
| )
| |
| This program provides controls to track the UFSAR Section 5.2.1.2 cyclic and transient (,ccurrences to ensure that components are maintained within the design limits.
| |
| 5.5.6 Inservice Testina and Insoection Proaram These programs provide controls for inservice testing and
| |
| ! inspection of ASME Code Class 1, 2, and 3 components. The program l shall include the following:
| |
| i a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:
| |
| ! ASME Boiler and Pressure Vessel Code and applicable Addenda Required Frequencies terminology for for performing inservice
| |
| , inservice testing and testing and inspection insoection activities activities l
| |
| Weekly At least once per 7 days l Monthly At least once per 31 days Quarterly or every l 3 months At least once per 92 days ;
| |
| Semiannually or !
| |
| l every 6 months At least once per 184 days l Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 day:
| |
| ! b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice tasting and ;
| |
| inspection activities: <
| |
| l c. The provisions of SR 3.0.3 are applicable to inservice testing and inspection activities; and
| |
| : d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.
| |
| f3 (continued) r FERMI - UNIT 2 5.0 11 Amendment No. 134 l
| |
| | |
| Programs and Manuals 5.5 h '5.5 ~ Programs and Manuals- (continued) 5.5.7 Ventilation Filter Testino Prooram (VFTP)
| |
| The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems.
| |
| : a. The following tests shall be performed:
| |
| : 1. Once per 18 months:
| |
| : 2. After each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank:
| |
| : 3. After Fry structural maintenance on the HEPA filter or charcoal adsorber housing: and
| |
| : 4. Following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation.
| |
| Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass <
| |
| specified below when tested in accordance with Regulatory Guide 1.52. Revision 2. and ASME N5101980 at the system flowrate specified below
| |
| * 10%.
| |
| Penetration and ESF Ventilation System Flowrate (cfm) System Byoass Standby Gas Treatment 3800 0.05%
| |
| Control Room Emergency 1800 (makeup filter) 1.0% ]
| |
| Filtration 3000 (recirculation filter)
| |
| I
| |
| : b. The following tests shall be performed: {
| |
| : 1. Once per 18 months: I
| |
| : 2. After each complete or partial replacement of the HEPA I filter bank or charcoal adsorber bank:
| |
| : 3. After any structural maintenance on the HEPA filter or ,
| |
| charcoal adsorber housing: and
| |
| : 4. Following painting, fire, or chemical release in any
| |
| ~
| |
| ventilation zone communicating with the subsystem while it is in operation.
| |
| i (continued)
| |
| FERMI.- UNIT 2 5.0 12 Amendment No. 134
| |
| | |
| l .
| |
| 1 Programs and Manuals 5.5 O s.s ero9ee s eed "em e,e 5.5.7 Ventilation. Filter Testina Procram (VFTP) (continued) f Demonstrate for each of the ESF systems that an inplace test l of the charcoal adsorber shows a penetration and system bypass < specified below when tested in accordance with Regulatory Guide 1.52. Revision 2, and ASME N510 1980 at the
| |
| ; system flowrate specified below i 10%.
| |
| l Penetration and ESF Ventilation Systs Flowrate (cfm) System Byoass l
| |
| Standby Gas Treatment 3800 0.05%
| |
| Control Room Emergency 1800 (makeup filter) 1.0%
| |
| Filtration 3000 (recirculation filter) l
| |
| : c. The following tests shall be performed:
| |
| : 1. Once per 18 months: !
| |
| : 2. After 720 hours of system operation O 3. After a1y structural maintenance on the HEPA filter or V charcoal adsorber housing: and !
| |
| : 4. Following paintirig. fire. or chemical release in any 1 ventilation zone communicating with the subsystem while it is in operation.
| |
| Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52. Revision 2. shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803 1989 at a temperature of 30*C and at the relative humidity specified below.
| |
| ESF Ventilation System Penetration BB Standby Gas Treatment 0.100* 70%
| |
| Control Room Emergency 1.0% 70%
| |
| Filtration (continued)
| |
| FERMI - UNIT 2 5.0 13 Amendment No. 134
| |
| | |
| Programs and Manuals 5.5 h 5.5 Programs and Manuals 5.5.7 Ventilation Filter Testina Proaram (VFTP) (continued)
| |
| : d. The following tests shall be performed once per 18 months.
| |
| Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the )refilters (CREF only), and the charcoal adsorbers is less t1an the value specified below when tested in accordance with Regulatory Guide 1.52 Revision 2, and ASME N510-1980 at the system flowrate specified as follows i 10t:
| |
| Delta P Flowrate ESF Ventilation System (inches water aauae) (cfm)
| |
| Standby Gas Treatment 11.0 3800 Control Room Emergency 6.0 (makeup train) 1800 Filtration (CREF) 8.0 (recirculation 3000 train)
| |
| : e. The following tests shall be performed once per 18 months.
| |
| Demonstrate that the heaters for each of the ESF system dissipate the value specified below when testcd in accordance with ASME N510 1980:
| |
| ESF Ventilation System Wattaae (kW)
| |
| Standby Gas Treatment a: 24 Control Room Emergency 12.0
| |
| * 2.0 Makeup Inlet Air The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
| |
| l (continued)
| |
| FERMI UNIT 2 5.0 14 Amendment No. 134 l l
| |
| | |
| 7 l
| |
| Programs and Manuals 5.5 l
| |
| 5.5 Programs and Manuals - (continued) l- 5.5.8 Exolosive Gas and Storaoe Tank Radioactivity Monitorino Prooram l
| |
| This program provides controls for potentially explosive gas-i mixtures contained in the Main Condenser offgas treatment system.
| |
| ! 'and the quantity of radioactivity contained in temporary outdoor L
| |
| storage tanks.
| |
| The. program shall included
| |
| : a. A limit of s 4% by volume for concentration of hydrogen in the main condenser offgas treatment system and a surveillance program to ensure the limit is maintained,
| |
| : b. A surveillance program to ensure that the quantity of radioactivity contained in any outdoor liquid radwaste tank that is not surrounded by liners, dikes, or walls, capable l of holding the tank's contents and that does not have tank overflows and surrounding area drains connected to the liquid radwsste treatment system is s 10 curies, excluding i tritium and dissolved or entrained noble gases.
| |
| ; The provisions of SR 3.0.2 and SR 3.0.3 are a)plicable to the Explosive Gas and Storage Tank Radioactivity ionitoring Program
| |
| : l. ( surveillance. frequencies.
| |
| 5.5.9 Emeroency Diesel Generator Fuel Oil Testino Prooram An emergency diesel generator fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
| |
| : a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
| |
| : 1. an API gravity or an absolute specific gravity within limits.
| |
| : 2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
| |
| : 3. a clear and bright appearance with proper color:
| |
| (continued)
| |
| FERMI UNIT 2 5.0 15 Amendment No. 134 i
| |
| | |
| 1 Programs and Manuals j 5.5 1 (l
| |
| LJ 5.5 Programs and Manuals 5.5.9 Emeroency Diesel Generator Fuel Oil Testina Prooram (continued)
| |
| : b. Within 31 days following addition of new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil: and
| |
| : c. Total particulate concentration of the fuel oil is s 10 mg/l when tested every 31 days in accordance with ASTN D-2276.
| |
| Method A.
| |
| The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Emergency Diesel Generator Fuel Oil Testing Program testing frequencies.
| |
| 5.5.10 Technical Soecifications (TS) Bases Control Proaram This program provides a means for processing changes to the Bases of these Technical Specifications.
| |
| : a. Changes to the Bases of the TS shall be made under Q appropriate administrative controls and reviews.
| |
| LJ
| |
| : b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
| |
| : 1. a change in the TS incorporated in the license: or
| |
| : 2. a change to the UFSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
| |
| : c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
| |
| : d. Proposed changes that meet the criteria of Specification 5.5.10b above shall be reviewed and ap) roved by the NRC prior to implementation. Changes to t1e Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
| |
| p
| |
| (/' (continued)
| |
| FERMI - UNIT 2 5.0-16 Amendment No. 134
| |
| | |
| Programs and Manuals I 5.5 I
| |
| ( 5.5 Programs and Manuals (continued) 5.5.11 Safety Function Determination Proaram (SFDP)
| |
| This program ensures loss of safety function is detected and.
| |
| appropriate actions taken. Upon entry into LC0 3.0.6. an evaluation shall be made to determine if loss of safety function exists. Additionally. Other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LC0 3.0.6.
| |
| : a. The SFDP shall contain the following:
| |
| : 1. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go i undetected: l
| |
| : 2. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists:
| |
| Provisions to ensure that an inoperable supported
| |
| : 3. 4 A system's Completion Time is not inappropriately b , extended as a result of multiple support system inoperabilities: and
| |
| : 4. Other appropriate limitations and remedial or compensatory actions.
| |
| : b. A loss of safety function exists when. assuming no concurrent single failure. a safety function assumed in the accident analysis cannot be performed. For the pur)ose of this program. a loss of safety function may exist w1en a support system is inoperable, and:
| |
| : 1. A required system redundant to system (s) supported by the inoperable support system is also inoperable: or
| |
| : 2. A required system redundant to system (s) in turn supported by the inoperable supported system is also inoperable; or
| |
| : 3. A required system redundant to support system (s) for the supported systems (a) and (b) above is also inoperable.
| |
| t
| |
| ( (continued)
| |
| FERMI UNIT 2 5.0 17 Amendment No. 134
| |
| | |
| c Programs and Manuals 5.5 l
| |
| 5.5 Programs and Manuals 5.5.11 Safety Function Determination Proaram (SFDP) (continued)
| |
| The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in
| |
| 'which the loss of safety function exists are required to be entered.
| |
| 5.5.12 Primary Containment Leakaae Rate Testina Proaram
| |
| : a. A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163.
| |
| " Performance Based Containment Leak-Test Program." dated September, 1995, with the exception of approved exemptions to 10 CFR 50, Appendix J.
| |
| : b. The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 56.5 psig.
| |
| ( c. The maximum allowable containment leakage rate L, at P,,
| |
| shall be 0.5% of containment air weight per day.
| |
| : d. Leakage Rate acceptance criteria are:
| |
| : 1. Containment leakage rate acceptance criterion is s 1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 L for the required Type B and C tests and s 0.7$ L, for Type A tests.
| |
| : 2. Air lock testing acceptance criteria are: I Overall air lock leakage rate is s 0.05 L, when i) l
| |
| ; tested at a P., l
| |
| , ii) For each door, leakage rate is s 5 scf per hour
| |
| ! when the gap between the door seals is i pressurized to a P,.
| |
| i (continued)
| |
| FERMI -' UNIT 2 5.0 18 Amendment No. 134 r
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals
| |
| (]
| |
| u 5.5.12 Primary Containment Leakaoe Rate Testino Procram (continued)
| |
| : e. The provisions of SR 3.0.2. do not apply to the test frequencies in the Primary Containment Leakage Rate Testing Program.
| |
| : f. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
| |
| 5.5.13 Hioh Density Soent Fuel Racks A program shall be provided which will assure that any unanticipated degradation of the high density spent fuel racks will be detected and will not compromise the integrity of the racks.
| |
| O i
| |
| l l
| |
| O.
| |
| 9 FERMI - UNIT 2 5.0-19 Amendment No. 134 t-
| |
| | |
| Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting R2quirements The following reports shall be submitted in accordance with 10 CFR 50.4.
| |
| 5.6.1 Occupational Radiation Exposure Reoort l l
| |
| A tabulation on an annual basis of the number of )lant, utility, and other personnel (including contractors) for w1om monitoring was required to be performed receiving exposures > 100 mrem /yr and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance describe maintenance, waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on i pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge I measurements. Small exposures totalling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. The report j shall be submitted by April 30 of each year. )
| |
| 5.6.2 Annual Radioloaical Environmental Ooeratina Reoort The Annual Radiological Environmental Operating Report covering the o)eration of the unit during the previous calendar year shall 1 be su)mitted by May 15 of each year. The report shall include ,
| |
| summaries interpretations, and analyses of trends of the results )
| |
| of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with !
| |
| the objectives outlined in the Offsite Dose Calculation Manual (0DCM), and in 10 CFR 50, Appendix I. Sections IV.B.2, IV.B.3, and IV.C.
| |
| 5.6.3 Radioactive Effluent Release Reoort i The Radioactive Effluent Release Re) ort covering the operation of i the unit during the previous year s1all be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I. Section IV.B.1.
| |
| (n) v (continued)
| |
| FERMI UNIT 2 5.0 20 Amendment No. 134 l
| |
| L
| |
| | |
| r q I
| |
| Reporting Requirements 5.6
| |
| ( ') 5.6 Reporting Requirements (continued) 5.6.4 Monthly Doeratina Reoorts Routine reports of operating statistics and shutdown experience .
| |
| shall be submitted on a monthly basis no later than the 15th of l each month following the calendar month covered by the report. l 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
| |
| : a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
| |
| LC0 3.2.1 " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)":
| |
| LC0 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)"; l LC0 3.2.3. " LINEAR HEAT GENERATION RATE (LHGR)"; and i LCO 3.3.2.1, " Control Rod Block Instrumentation."
| |
| : b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by r] the NRC. specifically those described in the following )
| |
| (/ ,
| |
| documents: j
| |
| : 1. NEDE-24011 P A " General Electric Standard Application for Reactor Fuel." (latest approved version); and
| |
| : 2. NEDE 23785-1-PA. "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss of-Coolant Accident -
| |
| SAFER /GESTR Application Methodology." (the approved version at the time the reload analyses are performed).
| |
| : c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits. Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM transient l analysis limits, and accident analysis limits) of the safety analysis are met.
| |
| : d. The COLR. including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
| |
| iO
| |
| 'd (continued)
| |
| FERMI - UNIT 2 5.0 21 Amendment No. 134
| |
| | |
| I Reporting Requirements 5.6
| |
| ) 5.6 Reporting Requirements (continued) 5.6.6 Safety Relief Valve Challence ReDort The main steam line Safety Relief Valve (SRV) Report documenting all challenges to SRVs during the previous calendar year shall be submitted by April 30 of each year.
| |
| 5.6.7 PAM Reoort When a report is required by Condition B or G of LC0 3.3.3.1.
| |
| " Post Accident Monitoring (PAM) Instrumentation." a report shall be submitted within the following 14 days. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation char.nels of the Function to OPERABLE status.
| |
| O V
| |
| C FERMI UNIT 2 5.0 22 Amendment No. 134 l 1
| |
| | |
| n .
| |
| High Radiation Area 5.7 L
| |
| 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area i 5.7.1 Pursuant to 10 CFR 20. paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as.
| |
| defined .in 10 CFR 20, in which the intensity of radiation is l
| |
| > 100 mrem /hr but.< 1000 mres/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto l shall be controlled by requiring issuance of a Radiation Work l l . Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Radiation Protection personnel) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates s 1000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
| |
| Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
| |
| : a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.
| |
| : b. A radiation monitoring device that continuously integrates O, the radiation dose rate in the area and alarms when a preset ,
| |
| integrated dose is received. Entry into such areas with !
| |
| this monitoring device may be made after the dose rate ,
| |
| levels in the area have been established and personnel have l been made knowledgeable of them,
| |
| : c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Protection Supervisor in the RWP.
| |
| I I
| |
| (continued)
| |
| FERMI - UNIT 2 5.0 23 Amendment No. 134
| |
| | |
| c High Radiation Area 5.7 5.7' High' Radiation Area (continued)
| |
| Si7.2 In addition to the requirements of Specification 5.7.1. areas accessible to individuals with radiation levels such that an individual could receive in 1 hour a dose equivalent > 1000 arems but < 500 rads at one meter from sources of radioactivity shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Nuclear Shift Supervisor on duty and/or the radiation protection supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the 4 dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
| |
| 5.7.3 For individual areas accessible to individuals with radiation levels such that a major portion of the individual's body could receive in 1 hour a dose > 1000 mrems with measurement made at 30 centimeters from the source of radioactivity, but < 500 rads at one meter from sources of radioactivity that are located within large areas such as reactor containment, where no enclosure exists aO for purposes of locking.. and where no enclosure can be reasonably constructed around the individual area, that individual area shall be roped off and conspicuously posted, and a flashing light shall be activated as a warning device.
| |
| O FERMI m UNIT 2 5.0 24 Amendment No. 134
| |
| | |
| i Reactor' Core SLs B 2.1.1 v) B 2.0 SAFETY LIMITS (SLs)
| |
| B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences (A00s). i The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in {
| |
| Specification 2.1.1.2. MCPR greater than the specified !
| |
| limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
| |
| ; The fuel cladding is one of the physical barriers that i separate the radioactive materials from the environs. The 1 integrity of this cladding barrier is related to its (3 relative freedom from perforations or cracking. Although l
| |
| V some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this 1
| |
| source is incrementally cumulative and continuously ;
| |
| measurable. Fuel cladding perforations, however, can result ;
| |
| from thermal stresses, which occur from reactor operation significantly above design conditions.
| |
| While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold l beyond which still greater thermal stresses may cause gross,
| |
| ! rather than incremental, cladding deterioration. Therefore, i the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling l (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling.
| |
| Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp O
| |
| V FERMI UNIT 2 B 2.1.1 - 1 Amendment No. 134 I
| |
| h .- -
| |
| j
| |
| | |
| ~
| |
| l l
| |
| l Reactor Core SLs B 2.1.1 BASES ,
| |
| BACKGROUND (continued) reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place.
| |
| This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
| |
| APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal o)eration and A00s. The reactor core SLs are establis1ed to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.
| |
| The Reactor Protection System setpoints (LC0 3.3.1.1,
| |
| " Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER j level that would result in reaching the MCPR safety limit.
| |
| 2.1.1.1 Fuel Claddina Intearity General Electric Company (GE) critical power correlations are applicable for all critical power calculations at pressures a 785 psig and core flows a 10% of rated flow.
| |
| For operation at low pressures or low flows, another basis is used, as follows:
| |
| Since the pressure drop in the by) ass region is essentially all elevation head, t1e core pressure drop at low power and flows will always be > 4.5 psi.
| |
| Ref. 2) show that with a bundle flow of Analysey 28 x 10 l (b/hr. bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
| |
| Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 10' lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER > 50% RTP.
| |
| Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psig is conservative.
| |
| FERMI - UNIT 2 82.1.1-2 Amendment No. 134 L ,
| |
| | |
| j Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFE 17 ANALYSES (continued) 2.1.1.2 tgE The MCPR SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical )ower at which boiling transition is calculated to occur has >een adopted as a convenient limit. However, the uncertainties in monitoring i the core operating state and in the procedures used to l calculate the critical power result in an uncertainty in the j value of the critical power. Therefore, the MCPR SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.
| |
| The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The-probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations, Details of the fuel cladding integrity SL calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.
| |
| 2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the O
| |
| FERMI UNIT 2 B 2.1.1 - 3 Amendment No. 134
| |
| | |
| f.
| |
| Reactor Core SLs B 2.1.1
| |
| ( ' BASES i APPLICABLE SAFETY ANALYSES (continued) water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.
| |
| i SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
| |
| 1 I
| |
| APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
| |
| O. SAFETY LIMIT Exceeding a reactor core SL may cause fuel damage and create VIOLATIONS a potential for radioactive releases in excess of 10 CFR 100. " Reactor Site Criteria," limits (Ref. 3).
| |
| Therefore, it is required to insert all insertable control i rods and restore cogliance with the SLs within 2 hours.
| |
| The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.
| |
| REFERENCES 1. 10 CFR 50. Appendix A. GDC 10.
| |
| : 2. NEDE 24011 P A (latest approved revision).
| |
| : 3. 10 CFR 100.
| |
| I l
| |
| O FERMI UNIT 2 B 2.1.1 - 4 Amendment No. 134 l
| |
| l ,
| |
| | |
| p RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)
| |
| B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to l 10 CFR 50. Appendix A, GDC 14. " Reactor Coolant Pressure Boundary," and GDC 15. " Reactor Coolant System Design" (Ref.1), the reactor coolant pressure boundary (RCPB) shall be-designed with a low probability of failure and sufficient ,
| |
| margin to ensure that t1e design conditions are not exceeded during normal operation and anticipated operational l4 occurrences (A00s).
| |
| During normal operation and A00s RCS pressure is limited from exceeding the design pressure by more than 10t. in accordance with Section III of the ASME Code (Ref. 2). To (d~3 ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LC0 3.10.1, " Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS ,
| |
| components shall be pressure tested in accordance with the '
| |
| requirements of ASME Code, Section XI (Ref. 3). I Overpressurization of the RCS could result in a breach of l the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, " Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.
| |
| i O
| |
| FERMI - UNIT 2 B 2.1.2 - 1 Amendment No. 134 Y
| |
| | |
| RCS Pressure SL B 2.1.2 (v SASES l
| |
| APPLICABLE The RCS safety / relief valves and the Reactor Protection :
| |
| SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function i have settings established to ensure that the RCS pressure SL will not be exceeded.
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| The RCS pressure SL has been selected such that it is at a l pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel. i main steam piping, and reactor coolant system are designed I and built in accordance with a)plicable codes and standards, as detailed in Reference 5. T1e RCS pressure SL is selected I to be the lowest transient overpressure allowed by the )
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| applicable codes. 1 I
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| 1 l
| |
| SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code. Section III, is 110% of design I pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these allowances is the 110%
| |
| O of the suction piping design pressures: therefore. the SL on d maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.
| |
| APPLICABILITY SL 2.1.2 applies in all MODES.
| |
| SAFETY LIMIT Exceeding the RCS pressure SL may cause immediate RCS VIOLATIONS failure and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion Time ensures that the o)erators take prompt remedial action and also assures that tie probability of an accident occurring during this period is minimal.
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| (3 V
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| FERMI UNIT 2 B 2.1.2 - 2 Amendment No. 134 1
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| | |
| ! I i
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| RCS Pressure SL B 2.1.2 BASES REFERENCES 1. 10 CFR 50, Appendix A. GDC 14. and GDC 15. -
| |
| : 2. ASME. Boiler and Pressure Vessel Code. Section III. 1 Article NB 7000. <
| |
| : 3. ASME, Boiler and Pressure Vessel Code. Section XI.
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| Article IWB-5000.
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| : 4. 10 CFR 100.
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| : 5. UFSAR Section 3.2.
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| O 1
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| O FERMI UNIT 2 B 2.1.2 -3 Amendment No. 134
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| | |
| LC0 Applicability B 3.0
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| (]
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| V B 3.0 LIMITING CONDITION FOR OPERATION (LC0) APPLICABILITY BASES 4
| |
| LCOs LC0 3.0.1 through LC0 3.0.7 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
| |
| i LC0 3.0.1 LC0 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LC0 is required to be met (i.e., when the unit is in the MODES or other s)ecified conditions of the Applicability statement of eac1 Specification).
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| LC0 3.0.2 LC0 3.0.2 establishes that upon discovery of a failure to meet an LCO. the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within (Sj specified Completion Times when the requirements of an LC0 are not met. This Specification establishes that:
| |
| : a. Completion of the Required Actions within the specified l Completion Times constitutes compliance with a Specification; and
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| : b. Com)letion of the Required Actions is not required when an _C0 is met within the specified Completion Time, unless otherwise specified.
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| There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the D
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| (V FERMI UNIT 2 B 3.0-1 Amendment No. 134
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| | |
| LC0 Applicability B 3.0 O
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| G BASES LC0 3.0.2 (continued) remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time.
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| In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
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| Completing the Required Actions is not required when an LC0 is met or is no longer applicable, unless otherwise stated in the individual Specifications.
| |
| The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case.
| |
| An example of this is in LCO 3.4.10. "RCS Pressure and Temperature (P/T) Limits."
| |
| The Com)letion Times of the Required Actions are also applica)1e when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to.
| |
| performance of Surveillances, preventive maintenance, (d corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead. Doing so limits the time both subsystems / divisions of a safety function are inoperable and limits the time other conditions exist which may result in LC0 3.0.3 being entered.
| |
| Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
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| O FERMI UNIT 2 B 3.0-2 Amendment No. 134
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| | |
| LCO Applicability B 3.0 BASES LC0 3.0.2 0 mtinued)
| |
| When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case. the Completion Times of the associated Required Actions would apply from the )oint in time that the new Specification becomes applica)le and the ACTIONS Condition (s) are entered.
| |
| LC0 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
| |
| : a. An associated Required Action.and Completion Time is not met and no other Condition applies: or
| |
| : b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of O. Conditions are such that entering LC0 3.0.3 is warranted: in such cases, the ACTIONS specifically state a Condition corres)onding to such combinations and also that LC0 3.0.3 >e entered immediately.
| |
| Thi Specification delineates the time limits for placing the unit in a safe MODE or other s)ecified condition when operation cannot be maintained wit 11n the limits for safe operation as defined by the LC0 and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.
| |
| Upon entering LCO 3.0.3, I hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities O
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| FERMI-- UNIT 2 B 3.0-3 Amendment No. 134
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| | |
| LC0 Applicability l B 3.0 l BASES LC0 -3.0.3 (continued) of the unit. assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LC0 3.0.3 are consistent with the discussion of Section 1.3.
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| Completion Times.
| |
| A unit shutdown required 1n accordance with LC0 3.0.3 nay be
| |
| -terminated and LC0 3.0.3 exited if any of the following occurs:
| |
| : a. The LC0 is now met.
| |
| : b. A Condition exists for which the Required Actions have now been performed.
| |
| : c. ACTIONS exist that do not have expired Com)letion Times. These Completion Times are applica)le from the point in time that the Condition is initially entered p alid not from the time LC0 3.0.3 is exited.
| |
| V The time limits of LC0 3.0.3 allow 37 hours for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, the total allowable time to reach MODE 4, or other applicable MODE. is not reduced. For example, if MODE 2 is reached in 2 hours. then the time allowed for reaching MODE 3 is the next 11 hours, because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1. a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.
| |
| In MODES 1, 2, and 3. LC0 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LC0 3.0.3 do not apply in MODES 4 and 5 because the unit is'already in the most restrictive Condition required by LC0 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1. 2, or 3) because the O
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| FERMI UNIT 2 B 3.0-4 Amendment No. 134 l
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| | |
| LCO Applicability B 3.0 BASES LC0 3.0.3 (continued)
| |
| ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.
| |
| Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown. in accordance with LC0 3.0.3.
| |
| would not provide appropriate remedial measures for the associated condition of the unit. An exam)le of this is in LC0 3.7.7. " Spent Fuel Storage Pool Water _evel." LCO 3.7.7 has an Applicability of "During movement of irradiated fuel assemblies.in the spent fuel storage pool." Therefore. . this LCO can be applicable in any or all MODES. If the LC0 and the Required Actions of LC0 3.7.7 are not met while in MODE 1. 2. or 3. there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LC0 3.7.7 of " Suspend movement of irradiated fuel assemblies in the spent fuel storage pool" is the
| |
| ~ appropriate Required Action to complete in lieu of the actions of LC0 3.0.3. These exceptions are addressed in the individual Specifications. .
| |
| O 'Co 3.0.4 LC0 3.0.4 establ4shes ,imitat4 ens en chanoes 4n MooeS er other specified conditions in the A)plicability when an LC0 is not met. It precludes placing t1e unit in a MODE or other specified condition stated in that Applicability
| |
| .(e.g., Applicability desired to be entered) when the following exist:
| |
| : a. Unit conditions are such that the requirements of the LC0 would not be met in the Applicability desired to be entered; and
| |
| : b. Continued noncompliance with the LC0 requirements. if the A>plicability were entered, would result in the unit >eing required to exit the Applicability desired to be entered to comply with the Required Actions.
| |
| Compliance with Required Actions that permit continued I operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. . This is without regard to the status of the unit before or after the MODE l
| |
| . change. Therefore, in such cases entry into a MODE or 1 other specified condition in the Applicability may be made O
| |
| FERMI - UNIT 2 B 3.0-5 Amendment No. 134
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| | |
| i LC0 Applicability B 3.0 A BASES V
| |
| LC0 3.0.4 (continued) in accordance with the provisions of the Required Actions. '
| |
| The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good !
| |
| practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
| |
| 4 The provisions of LC0 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comp y with ACTIONS. In addition, the I provisions of LCO 3.0.4 s all not 3revent changes in MODES or other specified conditions in t1e Applicability that result from any unit shutdown.
| |
| Exceptions to LC0 3.0.4 are stated in the individual Specifications. The exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for i continued operation for an unlimited period of time. l Exceptions may apply to all the ACTIONS or to a specific '
| |
| Required Action of a Specification.
| |
| O L
| |
| LC0 3.0.4 is only applicable when entering MODE 3 from MODE 4. MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2.
| |
| Furthermore LC0 3.0.4 is applicable when entering any other specified condition in the Applicability only while 1 operating in MODE 1, 2, or 3. The requirements of LC0 3.0.4 ;
| |
| do not apply in MODES 4 and 5, or in other specified l conditions of the Applicability (unless in MODE 1, 2, or 3) ;
| |
| because the ACTIONS of individual specifications i sufficiently define the remedial measures to be taken. In i some cases (none currently in Fermi-2 Technical Specifications) these ACTIONS provide a Note that states "While this LC0 is not met, entry into a MODE or other specified condition in the Ap)licability is not permitted, unless required to comply witi ACTIONS." This Note is a requirement explicitly precluding entry into a MODE or other specified condition of the Applicability.
| |
| Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LC0 3.0.4 or where an exception to LC0 3.0.4 is stated, is not a violation of p
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| us FERMI - UNIT 2 B 3.0-6 Amendment No. 134
| |
| | |
| LC0 Applicability B 3.0 BASES LC0 3.0.4 (continued)
| |
| SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY ;
| |
| prior to declaring the associated equipment OPERABLE (or I variable within limits) and restoring compliance with the affected LCO.
| |
| LC0 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment l to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LC0 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of required testing to demonstrate:
| |
| ~a. The OPERABILITY of the equipment being r'eturned to service or
| |
| : b. The OPERABILITY of other equipment.
| |
| ) The administrative controls ensure the time the equipment is s returned to service in conflict with the requirements of the i ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY.
| |
| This Specificati0n does not provide time to perform any other preventive or corrective maintenance.
| |
| An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions and must be reopened to perform the required testing.
| |
| An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is ;
| |
| taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.
| |
| FERMI UNIT 2 B 3.0-7 Amendment No. 134
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| | |
| I LC0 Applicability.
| |
| B 3.0 BASES LC0 3.0.6 LC0 3.0.6 establishes an exception to LC0 3.0.2 for support systems that have an LC0 specified in the Technical ;
| |
| Specifications (13). This exception is provided because LC0 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LC0 be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the plant is maintained in a safe condition are sweified in the support system LC0's i Required Actions. T1ese Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.
| |
| , When a support system is inoperable' and there is an LCO specified for it in the TS the supported system (s) are required to be declared inomrable if determined to be inoperable as a result of tie support system inoperability.
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| l' However, it is not necessary to enter into the supported-l systems' Conditions and Required Actions unless directed to i do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported iO l
| |
| systems' LCOs' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the plant is maintained in a safe condition in the l
| |
| support system's Required Actions.
| |
| However, there are instances where a support system's :
| |
| Required Action may either direct a supported system to be !
| |
| declared inoperable or direct entry into Conditions and i Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's l Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.
| |
| Specification 5.5.11. " Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LC0 3.0.6 an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a O
| |
| FERMI UNIT 2 B 3.0-8 Amendment No. 134
| |
| | |
| LC0 Applicability B 3.0 BASES LC0 3.0.6 (continued) result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. . The SFDP implements the requirements of LCO.3.0.6.
| |
| Cross division checks to identify a loss of safety function for those su) port systems that sup) ort safety systems are
| |
| . required. T1e cross division checc verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.
| |
| If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
| |
| ; LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit.
| |
| These special tests and operations are necessary to demonstrate select unit performance characteristics -to perform special maintenance activities, and to perform special evolutions. Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to permit performances of these-special tests and operations, which otherwise could not be performed if required to comply with H
| |
| the requirements of these TS. Unless otherwise specified.
| |
| all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other l- specified condition not directly associated with or required I to be changed to perform the special test or operation will remain in effect. j The Applicability of a Special Operations LC0 represents a l condition not necessarily in compliance with the normal ;
| |
| l requirements of the TS. Compliance with Special Operations L LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LC0 or under the other applicable TS l requirements. If it is desired to )erform the special operation under the provisions of tie Special Operations LCO, the requirements of the Special Operations LC0 shall be followed. When a Special Operations LC0 requires another LC0 to be met, only the requirements of the LC0 statement
| |
| -are required to be met regardless of that LCO's FERMIL- UNIT 2 B 3.0-9 Amendment No. 134 L
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| | |
| LC0 Applicability i B 3.0 1 BASES LC0 3.0.7 (continued) {
| |
| Applicability (i.e., should the requirements of this other LCO not be met, the ACTIONS of the Special Operations LC0 apply, not the ACTIONS of the other LCO). However, there are instances where the Special Operations LC0 ACTIONS may direct the other LCOs* ACTIONS be met. The Surveillances of the other LC0 are not required to be met, unless specified in the Special 0)erations LC0. If conditions exist such that the A other LC0'pplica)ility s requirementsof (ACTIONS any otherand LC0SRs) is met, all the to are required be met concurrent with the requirements of the Special Operations LCO.
| |
| j l
| |
| l l
| |
| FERMI - UNIT 2 B 3.0 - 10 Amendment No. 134
| |
| | |
| y .
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| SR Applicability B 3.0 l
| |
| B 3.0 SURVEILLANCE REQUIREMENT'(SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
| |
| SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.
| |
| Systems and commnents are assumed to be OPERABLE when the '
| |
| associated SRs lave been met. Nothing in this Specification, however, is to be construed as implying that
| |
| , systems or components are OPERABLE when:
| |
| : a. The systems or components are known to be inoperable, although still meeting the SRs: or
| |
| : b. The requirements of the Surveillance (s) are known to be not met between required Surveillance performances.
| |
| Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LC0 are not applicable, unless otherwise specified. The SRs associated with a Special Operations LCO are only applicable when the Sacial Operations LC0 is used as an allowable exception to t1e requirements of a Specification.
| |
| Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this :
| |
| case, the unplanned event may be credited as fulfilling the l performance of the SR. This allowance includes those SRs - l whose performance is normally precluded in a given MODE or other specified condition.
| |
| i FERMI UNIT 2' B 3.0-11 Amendment No. 134 4
| |
| | |
| SR Applicability l 3.0 !
| |
| BASES SR 3.0.1 (continued)
| |
| Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. I Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
| |
| Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This '
| |
| includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not '
| |
| having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.
| |
| Some examples of this pro uss are:
| |
| : a. Control Rod Drive maintenance during refueling that requires scram testing at > 800 psi. However, if other appropriate testing is satisfactorily completed and the ,
| |
| scram time testing of SR 3.1.4.3 is satisfied, the i control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psi to perform other necessary testing,
| |
| : b. High pressure coolant injection (HPCI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing 1s satisfactorily completed, startup Jm proceed with HPCI considered OPERABLE. This allows o)eration to reach the specified pressure to complete tie necessary post maintenance testing.
| |
| I A
| |
| g .__
| |
| FERMI UNIT 2 B 3. 0 - 12 Amendment No. 134
| |
| | |
| SR Applicability 3.0 BASES SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."
| |
| interval.
| |
| SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant o)erating conditions that may not be suitable for conducting t1e Surveillance (e.g.,
| |
| transient conditions or other ongoing Surveillance or maintenance activities). '
| |
| The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the i SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in l the individual Specifications. An example of where SR 3.0.2 i does not apply is the Primary Containment Leak Rate Program, rm Specification 5.5.12.
| |
| O As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Com)1etion Time that requires performance on a "once per..." Jasis. The 25%
| |
| extension applies to each performance after the initial l performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%
| |
| extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
| |
| The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.
| |
| n v -
| |
| FERMI UNIT 2 B 3.0 - 13 Amendment No. 134
| |
| | |
| p SR Applicability j 3.0 t
| |
| BASES O SR 3.0.3' SR 3.0.3 establishes the flexibility to defer declaring I affected equipment inoperable or an affected variable i outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified
| |
| :requency, whichever is less, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
| |
| This delay period provides adequate time to complete l
| |
| Surveillances that have been missed. This delay period
| |
| . permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.
| |
| The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most
| |
| ; probable result of any particular Surveillance being l performed is the verification of conformance with the j ( requirements.
| |
| When a Surveillance with a Frequency based not on time '
| |
| intervals, but upon specified unit conditions or operational situations, is discovered not to have been performed when 4 l specified, SR 3.0.3 allows the full delay period of 24 hours l to perform the Surveillance.
| |
| SR 3.0.3 also provides a time limit for completion of Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.
| |
| : l. Failure to comply with specified Frequencies for SRs is
| |
| ; expected to be an infrequent occurrence. Use of the delay L period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.
| |
| If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the !
| |
| variable is considered outside the specified limits and the '
| |
| Completion Times of the Required Actions for the applicable LC0 Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay ,
| |
| O FERMI - UNIT 2 B 3.0 - 14 Amendment No. 134 L.m.. _
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| | |
| SR Applicability 3.0 BASES SR 3.0.3 (continued) period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LC0 Conditions begin immediately upon the failure of the Surveillance.
| |
| Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.
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| SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.
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| This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit.
| |
| The provisions of this Specification should not be f interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
| |
| However, in certain circumstances, failing to meet an SR l t
| |
| will not result in SR 3.0.4 restricting a MODE change or
| |
| ! other specified condition change. When a system, subsystem, i division, component device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed per SR 3.0.1. which states that I surveillances do not have to be performed on inoperable i equipment. .When equipment is inoperable, SR 3.0.4 does not l apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore. failing to perform the Surveillance (s) within the specified Frequency i does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the A)plicability.
| |
| However, since the LC0 is not met in t11s instance, LC0 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.
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| The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability O
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| i FERMI UNIT 2 B 3.0 - 15 Amendment No. 134
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| SR Applicability 3.0 BASES gs
| |
| \ SR 3.0.4 (continued) that are required to comply with ACTIONS. In addition, the provisions of LC0 3.0.4 shall not 3revent changes in MODES or other specified conditions in t1e Applicability that result from any unit shutdown. .
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| The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not
| |
| -necessary. The specific time frames and conditions necessary for meeting the SRs are s)ecified in the Frequency. in the Surveillance, or )oth. This allows performance of Surveillances when the prerequisite condition (s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LC0 prior to the performance or completion of a Surveillance. A Surveillance that could I not be performed until after entering the LC0 Applicability would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately. !
| |
| the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a 3 articular event, condition, or time has been reached. Furt1er discussion of the specific formats of SRs' annotation is found in Section 1.4. Frequency.
| |
| SR 3.0.4 is only applicable when entering MODE 3 from '
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| MODE 4, MODE 2 from MODE 3 or 4. or MODE 1 from MODE 2.
| |
| Furthermore. SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1. 2. or 3. Tne requirements of SR 3.0.4 do not apply in MODES 4 and 5. or in other specified conditions of the Applicability (unless in MODE 1. 2. or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be t sken.
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| FERMI UNIT 2 B 3.0 - 16 Amendment No. 134
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| r .
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| SDM B 3.1.1 O
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| v B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
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| BASES BACKGR00f0 SDM requirements are specified to ensure:
| |
| : a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events:
| |
| l b. . The reactivity transients associated with postulated
| |
| ! accident conditions are controllable within acceptable limits; and
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| : c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
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| .These requirements are satisfied by the control rods, as described in GDC 26 (Ref.1),' which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.
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| O APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2 SAFETY ANALYSES and 3) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control
| |
| ! rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LC0 3.1.6, " Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control rod removal error during refueling and fuel assembly insertion error during refueling accidents (Ref. 4). The analysis of these reactivity insertion events
| |
| ~ assumes the refueling interlocks are OPERABLE when the' ;
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| . reactor is in the refueling mode of operation. These '
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| interlocks prevent the withdrawal of more than one control rod from the core during refueling. (Special consideration
| |
| : and requirements for multi le control rod withdrawal during t
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| refueling are covered in S cial Operations LC0 3.10.6.
| |
| " Multiple Control Rod With rawal-Refueling.") The analysis l
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| . assumes this condition is acceptable since the core will be shut down with the highest worth control rod withdrawn, if adequate SDM is maintained. i O
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| FERMI - UNIT 2 B 3.1.1 - 1 Amendment No. 134 4
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| E SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)
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| Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity. Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage.
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| SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The specified SDM limit accounts for the uncertainty in the >
| |
| demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth centrol rod is determined by measurement. When SDM is demonstrated by calculations not associated with a test (e.g.. to confirm SDM during the fuel loading sequence),
| |
| additional margin is included to account for uncertainties in-the calculation. To ensure adequate SDM during the design process, a design margin is included to account for l uncertainties in the design calculations (Ref. 5). l APPLICABILITY In MODES 1 and 2. SDM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref. 2). In MODES 3 and 4 l SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. !
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| SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or a fuel assembly insertion error (Ref. 4).
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| O FERMI - UNIT 2 B 3.1.1 - 2 Amendment No. 134
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| i SDM B 3.1.1
| |
| . BASES ACTIONS U With SDM r ot within the limits of the LC0 in MODE 1 or 2. ,
| |
| SDM must be restored within 6 hours. Failure to meet the I specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours is acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, and the low probability of an event occurring during this interval.
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| M If the SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours. to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours is reasonable.-based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| i R
| |
| With SDM not within limits in MODE 3, the operator must i immediately initiate action to fully insert all insertable i control rods. Action must continue until all insertable i control rods are fully inserted. This action results in the l least reactive condition for the core. i D.1. D.2. D.3. and D.4 With SDM not within limits in MODE 4, the operator must i immediately initiate action to fully insert all insertable i control rods. Action must continue until all insertable control rods are fully inserted. This action results in the !
| |
| least reactive condition for the core. Action must also be initiated within 1 hour to provide means for control of ,
| |
| potential-radioactive releases. This includes ensuring secondary containment is OPERABLE: at least one Standby Gas
| |
| -Treatment (SGT) subsystem is OPERABLE: 'and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases.
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| O FERMI - UNIT 2 B 3.1.1 - 3 Amendment No. 134
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| SDM B 3.1.1 BASES ACTIONS (continued)
| |
| This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to >erform the surveillances needed to demonstrate the OPERABI_ITY of the components.
| |
| If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
| |
| E.1. E.2. E.3. E.4. and E.5 With SDM not within limits in MODE 5, the operator must incdiately suspend CORE ALTERATIONS that could reduce SDM (e.g., insertion of fuel in the core or the withdrawal of control rods). Suspension of these activities shall not preclude completion of movement of a component.to a safe condition. Inserting control rods or removing fuel from the i core will reduce the total reactivity and are therefore excluded from the suspended actions.
| |
| Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuC assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted.
| |
| Action must also be initiated within 1 hour to provide means for control of potential radioactive releases. This includes ensuring secondary containment is OPERABLE: at least one SGT subsystem is OPERABLE: and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases.
| |
| This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances G
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| V FERMI - UNIT 2 B 3.1.1 - 4 Amendment No. 134
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| SDM B 3.1.1 BASES ACTIONS (continued) as needed to demonstrate the OPERABILITY of the components.
| |
| If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case. SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE.
| |
| SURVEILLANCE SR 3.1.1.1 REQUIREENTS Adequate SDM must be maintained to ensure that the reactor can be made subcritical from any initial operating condition. Adequate SDM is demonstrated by testing before or during the first criticality after fuel movement within the reactor pressure vessel. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (B0C) test must also account for changes in core reactivity during the cycle.
| |
| Therefore, to obtain the SDM. the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity O during the operating cycle and the calculated B0C core reactivity. If the B0C is the most reactive point in the cycle, no correction to the BOC measured value is required (Ref. 6). For the SDM demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Ak/k) must be added to the SDM limit of 0.28% Ak/k to account for uncertainties in the calculation.
| |
| The SDM may be demonstrated during an in sequence control rod withdrawal or during local criticals.
| |
| The Frequency of 4 hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.
| |
| During MODE 5. adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example.
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| O FERMI - UNIT 2 B 3.1.1 - 5 Amendment No. 134 j i
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| SDM B 3.1.1 O BASES V !
| |
| SURVEILLANCE REQUIREMENTS (continued) bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence and would meet'the Frequency " prior to each in vessel fuel movement." These bounding analyses I include additional margins to the associated uncertainties as appropriate. Offload / reload sequences (i.e.. where all loaded fuel assemblies are consistent with an analyzed loading pattern) inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.
| |
| REFERENCES 1. 10 CFR 50, Appendix A. GDC 26.
| |
| : 2. UFSAR Section 15.4.9.
| |
| : 3. NEDE-24011-P-A 9 US. " General Electric Standard Application for Reactor Fuel." Su)plement for United O States Section S.2.2.3.1, Septem)er 1988.
| |
| b 4. UFSAR. Section 15.4.1.1.
| |
| : 5. UFSAR. Section 4.3.2.4.1.
| |
| : 6. NEDE-24011-P-A 9. " General Electric Standard 1 Application for Reactor Fuel." Section 3.2.4.1 Septeaber 1988.
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| l FERMI - UNIT 2 B 3.1.1 - 6 Amendment No. 134
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| i Reactivity Anomalies i L B 3.1.2 l I
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| B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES
| |
| , BACKGROUND In accordance with GDC 26, GDC 28. and GDC 29 (Ref. 1),
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| reactivity shall be controllable such that subcriticality is l
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| maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity anomaly is used as a measure of the predicted versus i measured core reactivity during power operation. The i continual confirmation of core reactivity is necessary to
| |
| ; ensure that the Design Basis Accident (DBA) and transient l
| |
| safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel
| |
| ! reactivity or control rod worth or operation at conditions i not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing i predicted versus measured core reactivity validates the ;
| |
| nuclear methods used in the safety analysis and supports the l le SDM demonstrations (LC0 3.1.1, " SHUTDOWN MARGIN (SDM)") in l
| |
| '( assuring the reactor can be brought safely to cold. i subcritical conditions.
| |
| When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by i.he negative reactivity of the control components, thermal feedback, 4 neutron leakage, and materials in the core that absorb l
| |
| neutrons, such as burnable absorbers, producing zero net reactivity.
| |
| l In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel l loaded in the previous cycles provide excess positive l,' reactivity beyond that required to sustain steady state operation at the beginning of cycle (B0C). When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, and whatever neutron poisons (mainly xenon and samarium) are present in the fuel.
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| /m FERMI - UNIT 2 B 3.1.2 - 1 Amendment No. 134 L
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| Reactivity Anomalies B 3.1.2 BASES BACKGROUND (continued)
| |
| The predicted core reactivity is calculated by a 3D core simulator code as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The core reactivity is determined from actual plant conditions and is then compared to the predicted value for the cycle exposure.
| |
| APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular. SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks.
| |
| Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity, o The comparison between measured and predicted initial core Q reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted reactivity for identical core conditions at BOC do not reasonably agree, then the assumptions used in the j reload cycle design analysis or the calculation models used i to predict reactivity may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at B0C, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured reactivity from the predicted reactivity that develop during fuel depletion may be an indication that the 3 assumptions of the DBA and transient analyses are no longer '
| |
| valid, or that an unexpected change in core conditions has occurred.
| |
| Reactivity anomalies satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
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| FERMI UNIT 2 83.1.2-2 Amendment No. 134
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| | |
| o s Reactivity Anomalies B 3.1.2 O "^ses LC0 The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the " Nuclear Design Methodology" are larger than expected. A limit on the difference between the monitored and the predicted reactivity of
| |
| * 11 Ak/k has been established based on engineering judgment. A > lt deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.
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| APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved. Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In MODE 2, control rods are typically being withdrawn during a startup. In MODES 3 and 4, all control rods are fully inserted and therefore the reactor is in the least reactive state, where monitoring core n reactivity is not necessary. In MODE 5. fuel loading V results in a continually changing core reactivity. SDM requirements (LC0 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and a SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement). The SDM test, required by LC0 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions: therefore, reactivity anomaly is not required during these conditions.
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| I l
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| l O ;
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| FERMI - UNIT 2 B 3.1. 2 - 3 Amendment No. 134
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| Reactivity Anomalies B 3.1.2 BASES ACTIONS L1 Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions.
| |
| The required Completion Time of 72 hours is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.
| |
| ill If the core reactivity cannot be restored to within the it Ak/k limit, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| 1 i
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| i O
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| FERMI - UNIT 2 B 3.1.2 - 4 Amendment No. 134 l l
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| Reactivity Anomalies ,
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| B 3.1.2 O
| |
| G BASES SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verifying the reactivity difference between the monitored and predicted reactivity is within the limits of the LC0 provides added assurance that plant operation is maintained within the assumptions of the DBA and transient analyses. A comparison of the monitored reactivity to the predicted reactivity at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a significant ,
| |
| amount. This may occur following a refueling in which new i fuel assemblies are loaded, fuel assemblies are shuffled I within the core, or fuel assemblies are removed and l reinserted as when control rods are re) laced or shuffled. I Also, core reactivity changes during t1e cycle. The 24 hour I interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicted reactivity can be made. I For the purposes of this SR, the reactor is assumed to be at '
| |
| equilibrium conditions when steady state operations (no control rod movement or core flow changes) at = 80% RTP have
| |
| ~ been obtained. The 1000 MWD /ST Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity. This comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results. Therefore, the comparison is only done when in MODE 1.
| |
| REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28 and GDC 29.
| |
| : 2. UFSAR, Chapter 15.
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| FERMI UNIT 2 B 3.1.2 - 5 Amendment No. 134
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| -12345X 1
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| l l
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| i
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| E .
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| Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS .
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| B 3.1.3 Control Rod OPERAB.TLITY BASES j l
| |
| 1 BACKGROUND' Control rods are components of the control rod drive (CRD)
| |
| System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational' occurrences, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to satisfy the requirements of GDC 26 GDC 27 GDC '28, and 29 (Ref.1).
| |
| The CRD' System consists of 185 locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive mechanism. The locking piston type CRDM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.
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| l This Specification, along with LC0 3.1.4, " Control Rod Scram i Times " and LC0 3.1.5, " Control Rod Scram Accumulators,"
| |
| ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the i assumptions used in the safety analyses of References 2 and 3.
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| O FERMI.- UNIT 2~ B3.1.3-1 Amendment No. 134
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| Control Rod OPERABILITY B 3.1.3 O BASES
| |
| - O APPLICABLE The analytical methods and assumptions used in the SAFETY ANALYSES evaluations involving control Nds are presented in References 2 and 3. The control rods provide the primary means for rapid reactivity control (reactor scram) for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.
| |
| The capability to insert the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated. Since the SDM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA) can possibly occur.
| |
| Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function.
| |
| 3 The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, " Reactor Core SLs." and LC0 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)"), the It cladding plastic strain fuel design limit (see Bases for LC0 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LC0 3.2.3,
| |
| " LINEAR HEAT GENERATION RATE (LHGR)"), and the fuel damage limit (see Bases for LC0 3.1.6. " Rod Pattern Control")
| |
| during reactivity insertion events.
| |
| The negative reactivity insertion (scram) provided by the CRD System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CRDA. The Bases for LC0 3.1.4.
| |
| LC0 3.1.5, and LC0 3.1.6 discuss in more detail how the SLs are protected by the CRD System.
| |
| Control rod OPERABILITY satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| O FERMI - UNIT 2 B3.1.3-2 Amendment No. 134 l
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| 1
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| | |
| Control Rod OPERABILITY l B 3.1.3 BASES LC0 The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times: therefore, an inoperable accumulator does
| |
| - not immediately require declaring a control rod inoperable.
| |
| Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.
| |
| -APPLICABILITY In MODES I and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 6re O located in LC0 3.9.5, " Control Rod OPERABILITY-Refueling."
| |
| ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod.
| |
| This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.
| |
| A.I. A.2. A.3. and A.4 A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. The Required Actions are modified by a Note, which allows the rod worth minimizer (RWM) to be bypassed if required to allow continued operation. LC0 3.3.2.1, " Control Rod Block Instrumentation," provides additional requirements when the O
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| FERMI UNIT 2. B3.1.3-3 Amendment No. 134 l .
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| Control Rod OPERABILITY B 3.1.3
| |
| ' BASES ACTIONS (continued)
| |
| RWM is bypassed to ensure compliance with the CRDA analysis.
| |
| With one withdrawn control rod stuck, the local scram reactivity rate and CRDA control rod worth assumptions may not be met if the stuck control rod separation criteria are i not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if: a) the stuck control rod occupies a location adjacent to two " slow" control rods:
| |
| b) the stuck control rod occupies a location adjacent to one
| |
| " slow" control rod, and the one " slow" control rod is also adjacent to another " slow" control rod; or c) if the stuck control rod occupies a location adjacent to one " slow" control rod when there is another pair of " slow" control rods adjacent to one another. The description of " slow" control rods is provided in LC0 3.1.4, " Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours. The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be l
| |
| shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CRDM. The control rod O can be isolated from scram and normal insert and withdraw pressure, yet still maintain cooling water to the CRD.
| |
| Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the lyi power setpoint (LPSP) of the RWM.
| |
| SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control rods.
| |
| Testing each withdrawn control rod ensures that a generic l problem does not exist. Thir. Completion Time allows for an '
| |
| exception to the normal " time zero" for beginning the allowed outage time " clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP :
| |
| of the RWM, since the notch insertions may not be compatible :
| |
| with the requirements of rod pattern control (LC0 3.1.6) and the RWM (LC0 3.3.2.1). The allowed Completion Time of 24 hours from discovery of Condition A concurrent with ,
| |
| THERMAL POWER greater than the LPSP of the RWM provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
| |
| O FERMI - UNIT 2 B3.1.3-4 Amendment No. 134
| |
| | |
| Cord.rol Rod OPERABILITY B 3.1.3 1
| |
| BASES ACTIONS (continued) 4 To allow continued operation with a withdrawn control rod which is stuck, an evaluation of adequate SDM is also required within 72 hours. Should a DBA or transient require a shutdown, to preserve the single failure criterion. an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be I evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.
| |
| The allowed Completion Time of 72 hours to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 4).
| |
| O u !
| |
| With two or more withdrawn control rods stuck, the plant must be brought to MODE 3 within 12 hours. The occurrence of more than one control rod stuck at a withdrawn Sosition increases the probability that the reactor cannot )e shut down if required. Insertion of all insertable control rods !
| |
| eliminates the possibility of an additional failure of a ;
| |
| control rod to insert. The allowed Completion Time of :
| |
| 12 hours is reasonable, based on operating experience, to !
| |
| reach MODE 3 from full power conditions in an orderly manner j and without challenging plant systems.
| |
| C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours and disarmed (electrically or hydraulically) within 4 hours. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.
| |
| The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods O
| |
| FERMI - UNIT 2 B 3.1.3 - 5 Amendment No. 134
| |
| | |
| 1 Control Rod OPERABILITY l B 3.1.3
| |
| (~N BASES V
| |
| ACTIONS (continued) can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.) is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LC0 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
| |
| The allowed Completion Times are reasonable, considering the small number of allowed ino)erable control rods, and provide time to insert and disarm t1e control rods in an orderly manner and without challenging plant systems.
| |
| D.1 and D.2 Control rods that are not in compliance with the prescribed withdrawal sequence may increase the potential reactivity worth of a dropped control rod during a CRDA. At s 101 RTP, action must be taken to restore compliance with p the prescribed withdrawal sequence or restore the control V rods to OPERABLE status.
| |
| 1 Control rod withdrawal sequences are normally established i consistent with the rules of the generic BPWS analysis.
| |
| Occasionally, operational limitations (e.g., power ,
| |
| suppression of failed fuel) may dictate the insertion of I control rods which do not meet the minimum cell separation criteria of the generic BPWS analysis. In such situations, 1 sufficient cycle specific analyses are performed to I demonstrate that the resulting control rod worths of the '
| |
| modified control rod withdrawal sequence are bounded by the rod worths allowed by rigorously following the rules of the generic BPWS analysis, thereby assuring that the 280 cal /gm fuel damage limit will not be violated during a CRDA.
| |
| The " prescribed withdrawal sequence" is defined as the combination of both the procedurally specified control rod movement sequence and any analytically allowed deviations from this sequence. Some prescribed withdrawal sequences (e.g., BPWS (Ref. 4)) have more flexibility in allowed deviations than other prescribed withdrawal sequences (e.g.,
| |
| a cycle specific sequence developed for power suppression of failed fuel may not allow any deviations).
| |
| n FERMI UNIT 2 B 3.1.3 - 6 Amendment No. 134
| |
| | |
| Control Rod OPERABILITY B 3.1.3 BASES ACTIONS (continued)
| |
| Condition D is modified by a Note indica) Clat the Condition is not applicable when > 10t Rb 51nce the prescribed withdrawal sequence is not required to be followed under these conditions, as described in the Bases for LC0 3.1.6. The allowed Completion Time of 4 hours is acceptable, considering the low probability of a CRDA occurring.
| |
| E.d If any Required Action and associated Completion Time of Condition A. C, or D are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of s ino)erable control rods could be indicative of a generic s pro)lem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 frcm full power in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by i moving control rods to a position with an OPERABLE I indicator, or by the use of other appropriate methods. The '
| |
| 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.
| |
| O ;
| |
| FERMI - UNIT 2 B 3.1.3 - 7 Amendment No. 134 I
| |
| i
| |
| | |
| l Control Rod OPERABILITY B 3.1.3
| |
| -BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
| |
| -The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RbH, since the notch insertions may not be compatible with the requirements of the prescribed withdrawal sequence (LCO 3.1.6) and the RWM (LC0 3.3.2.1).
| |
| The 7 day frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control rods are tested at a :
| |
| 31 day Frequency, based on the potential power reduction required to allow the control rod movement and considering the large testing sample of SR 3.1.3.2. Furthermore, the 31 day Frequency takes into account operating experience ;
| |
| related to changes in CRD performance. At any time, if a !
| |
| withdrawn control rod is immovable, a determination of that '
| |
| O control rod's ability to insert on a scram (OPERABILITY) must be made and appropriate action taken.
| |
| SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is s 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. l This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, i
| |
| and SR 3.1.4.4.~
| |
| The LOGIC SYSTEM FUNvTIONAL TEST in LC0 3.3.1.1, " Reactor Protection System (RPS)
| |
| Instrumentation," that overlaps this Surveillance and the functional testing of SDV vent and drain valves in LCO 3.1.8 " Scram Discharge Volume (SDV) Vent and Drain Valves," provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
| |
| O FERMI UNIT 2 B 3.1.3 - 8 Amendment No. 134
| |
| | |
| Control Rod OPERABILITY i B 3.1.3 )
| |
| i BASES .7 SURVEILLANCE REQUIREMENTS (continued)
| |
| SR_;4.1. 3.5 Cour, ling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended fanction when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position.
| |
| The verification is required to be performed any time a -
| |
| control rod is withdrawn to the " full out" position (notch !
| |
| position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the " full out" position during the performance of SR 3.1.3.2. This Frequency is l acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and I operating experience related to uncoupling events.
| |
| O aereaeaces 1. 1o cra so. anneedix 4. ooc 28. ooc 27. ooc 28.
| |
| and GDC 29.
| |
| : 2. UFSAR, Section 4.5.2.1.3.
| |
| ! 3. UFSAR, Chapter 15.
| |
| : 4. NED0-21231. " Banked Position Withdrawal Sequence "
| |
| Section 7.2, January 1977.
| |
| FERMI - UNIT 2 B 3.1.3 - 9 Amendment No. 134
| |
| | |
| g i
| |
| ! Control Rod Scram Times i B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Times BASES BACKGROUND . The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Ref. 1). The control rods are scrammed by positive means using hydraulic pressure exerted on the CRD piston.
| |
| When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action.
| |
| Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index i
| |
| tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston. As the drive moves upward and the accumulator pressure reduces below the reactor pressure, Oy a ball check valve opens. letting the reactor pressure com)lete the scram action. If the reactor pressure is low, suc1 as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure.
| |
| APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the control rod scram function are 3 resented in References 2 and 3. The Design Basis Accident OBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms t5 basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g. several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met, i
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| io U
| |
| FERMI - UNIT 2 B 3.1.4 - 1 Amendment No. 134
| |
| | |
| Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued)
| |
| The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, " Reactor Core SLs," and LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)")
| |
| and the it cladding plastic strain fuel design limit (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENEP.ATION RATE (APLHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to 3revent the actual MCPR from becoming less than the MCPR S . during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 4) and.
| |
| therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LC0 3.1.6. " Rod Pattern Control"). For the reactor vessel over)ressure protection analysis, the scram function, along wit 1 the safety / relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.
| |
| Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met ,
| |
| (Ref. 5). To account for single failures and " slow" l scramming control rods, the scram times specified in i Table 3.1.4 1 are faster than those assumed in the design basis analysis. The scram times have a margiri that allows up to approximately 7% of the control rods (e.g. ,
| |
| 101 x 71 - 13) to have scram times exceeding the specified I limits (i.e., " slow" control rods) assuming a single withdrawn stuck control rod (as allowed by LC0 3.1.3 ,
| |
| " Control N"' OPFRABILITY") and an additional control rod
| |
| , failing tc .a per the single failure criterion. The l scram times ree specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes
| |
| (" pickup") when the index tube ) asses a specific location and then opens (" dropout") as t1e index tube travels upward.
| |
| Verification of the specified scram times in Table 3.1.41 p
| |
| ,l --._
| |
| FERMI UNIT 2 B '3.1.4 - 2 Amendment No. 134
| |
| | |
| P Control Rod Scram Times i
| |
| B 3.1.4 l
| |
| 1 BASES l
| |
| LC0 (continued) is accomplished through measurement of the " dropout" times. i To ensure that local scram reactivity rates are maintained '
| |
| within acceptable limits, no more than two of the allowed
| |
| " slow" control rods may occupy adjacent (i.e,. face adjacent ,
| |
| j or diagonally adjacent) locations. j l Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the i table are considered " slow" and that control rods with scram times > 7 seconds are considered inoperable as required by l SR 3.1.3.4.
| |
| This LC0 applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LC0 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as
| |
| " slow" control rods.
| |
| APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant l
| |
| O conditions. These events are assumed to occur during startup and power operation: therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4. the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements j for control rod scram capability during these conditions. !
| |
| Scram requirements in MODE 5 are contained in LC0 3.9.5, !
| |
| * Control Rod OPERABILITY-Refueling." l ACTIONS A_J i
| |
| When the requirements of this LC0 are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analyses. Therefore, the plant must be brought to a MODE in which the LC0 does
| |
| , not a) ply. To achieve this !tatus, the plant must be ,
| |
| I broug1t to MODE 3 within 12 Iours. The allowed Completion l Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| (
| |
| FERMI UNIT 2 B 3.1.4 - 3 Amendment No. 134 l
| |
| | |
| Contr.ol Rod Scram Times B 3.1.4 BASES SURVEILLANCE ~ The four SRs of this LC0 are modified by a Note stating that REQUIREMENTS during a single control rod scram time surveillance the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated. (i.e., charging valve closed) the influence of the CRD pump head does not I
| |
| affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.
| |
| SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assuh.0d Control rod scram time. Measurement of the scram times with reactor steam dome pressure = 800 psig demonstrates acceptable scram times for the transients I analyzed in Reference 3.
| |
| Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing ;
| |
| effects of reactor steam dome pressure and stored ;
| |
| accumulator energy. Therefore, demonstration of adequate i scram times at reactor steam dome pressure = 800 psig !
| |
| O ensures that the measured scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following a shutdown
| |
| > 120 days or longer, control rods are required to be tested before exceeding 40t RTP following the shutdown. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY. the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System.
| |
| i l
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| l l
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| I O
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| FERMI UNIT 2 B 3.1.4 - 4 Amendment No. 134
| |
| | |
| Control Rod Scram Times B 3.1.4 BASES
| |
| -SURVEILLANCE REQUIREMENTS (continued)
| |
| . SR 3.1.4.2 Additional testing of a sample of control rods is required
| |
| -to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods (i.e., = 19 control rods tested).
| |
| The sample remains representative if no more than 20% of the control rods in the sam)le tested are determined to be
| |
| " slow" or inoperable. dith more than 20% of the sample declared to be " slow" or inoperable per the criteria in Table 3.1.41, additional control rods are tested until this 20% criterion (e.g., 20% of the entire sample size) is satisfied, or until the total number of " slow" and inoperable control rods (throughout the core, from all surveillances) exceeds the LC0 limit. For planned testing, the control rods selected for the sample should be different )
| |
| for each test and should be in addition to any scram time i testing required to satisfy SR 3.1.4.4 following work on control rods or the CRD System that could affect scram times. Data from scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. If O data is captured from a reactor scram, all rods are available for selection in the next required test sample.
| |
| The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LC0 3.1.3 and LC0 3.1.5, " Control Rod Scram Accumulators."
| |
| SR 3.1.4.3 When work that could affect the scram insertion time is
| |
| >erformed on a control rod or the CRD System, testing must j
| |
| )e done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The i required scram time testing must demonstrate the affected ;
| |
| control rod is still within acceptable limits. The limits i for reactor pressures < 800 psig are established and I maintained within approved plant procedures based on a high probability of meeting the acceptance criteria at reactor I
| |
| O FERMI - UNIT 2 B 3.1.4 - 5 Amendment No. 134
| |
| | |
| e Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) pressures = 800 psig. Limits for = 800 psig are found in Table 3.1.4 1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of Table 3.1.41, Note 2, the control rod can be declared OPERABLE and " slow."
| |
| Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification: replacement of a control rod: and maintenance or modification of a scram solenoid pilot valve, scram valve, aiston accumulator, isolation valve or check valve in t1e piping required for scram.
| |
| The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of o)erating conditions and the more frequent surveillances on otler aspects of control rod OPERABILITY, SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System and when fuel in a control cell is moved, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4 1 with the reactor steam dome pressure
| |
| = 800 psig. Where work has been performed on a CRD at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. For a control rod affected by work )erformed while shut down, however, a zero pressure and higi pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions.
| |
| Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria, The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILifY.
| |
| l O
| |
| FERMI - UNIT 2 B 3.1.4- 6 Amendment No. 134
| |
| | |
| l Control Rod Scram Times l B 3.1.4 BASES REFERENCES 1. 10 CFR 50. Appendix A GDC 10.
| |
| I 2. UFSAR, Section 4.5.2.2.3.
| |
| : 3. UFSAR, Chapter 15.
| |
| : 4. NEDE-24011 P-A 11. " General Electric Standard Application for Reactor Fuel," Section S.2.2.3.1.
| |
| I November 1995.
| |
| : 5. Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications " BWROG 8754. September
| |
| : 17. 1987.
| |
| O l
| |
| V l
| |
| l l
| |
| l l
| |
| l O
| |
| FERMI UNIT 2 B 3.1.4 - 7 Amendment No. 134
| |
| | |
| Control Rod Scram Accumulators B 3.1.5 O B 3.1 REACTIVITY CONTROL SYSTEMS O
| |
| B 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control 3 Rod Drive (CRD) System and are provided to ensure that the l control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure.
| |
| The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen, which provides the required l energy. The scram accumulators are necessary to scram the l control rods within the required insertion times of LC0 3.1.4. " Control Rod Scram Times."
| |
| APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the control rod scram function are ) resented in References 1 and 2. The Design Basis Accident (OBA) and transient analyses assume that all of the control rods scram at a l
| |
| O' specified insertion rate. OPERABILITY of each individual control rod scram accumulator, along with LC0 3.1.3.
| |
| l
| |
| " Control Rod OPERABILITY " and LC0 3.1.4. ensures that the scram reactivity assumed in the DBA and transient analyses can be met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated ;
| |
| control r d. ,
| |
| The screm function of the CRD System, and therefore the OPERABILITY of the accumulators, protects.the MCPR Safety l Limit (see Bases for SL 2.1.1 " Reactor Core SLs." and LCO 3.2.2. " MINIMUM CRITICAL POWER RATIO (MCPR)") and it cladding plastic strain fuel design limit (see Bases for LCO 3.2.1. " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)." and LC0 3.2.3 " LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LC0 3.1.4). In addition, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LC0 3.1.6. " Rod Pattern Control").
| |
| l Control rod scram accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| O FERMI UNIT'2 B 3.1.5 - 1 Amendment No. 134
| |
| | |
| Control Rod Scram Accumulators B 3.1.5 BASES l
| |
| LC0 The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability l exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.
| |
| APPLICABILITY In MODES 1 and 2, the scram function is required for- l mitigation of DBAs and transients, and therefore the scram '
| |
| accumulators must be OPERABLE to support the scram function.
| |
| l In MODES 3 and 4, control rods are not allowed to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod scram accumulator OPERABILITY I during these conditions. Requirements for scram l accumulators in MODE 5 are contained in LC0 3.9.5. " Control i Rod OPERABILITY-Refueling." !
| |
| ACTIONS The ACTIONS table is modified by a Note indicating that a
| |
| ^ separate Condition entry is allowed for each control rod C scram accumulator. This is acceptable since the Required Actions for each Condition provide appropriate compensatory i
| |
| actions for each affected accumulator. Complying with the Required Actions may allow for continued o)eration and subsequent affected accumulators governed )y subsequent !
| |
| ! Condition entry and application of associated Required l l Actions.
| |
| A.1 and A.2 l 1
| |
| With one control rod scram accumulator inoperable and the reactor steam dome pressure = 900 psig, the control rod may be declared " slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the '
| |
| required scram times in Table 3.1.4-1. Required Action A.1 is modified by a Note indicating that declaring the control rod " slow" only applies if the associated control scram time i was within the limits of Table 3.1.4-1 during the last scram time test. Otherwise, the control rod would already be considered " slow" and the further degradation of scram '
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| 1 O l
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| 1 b
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| FERMI UNIT 2 B 3.1.5 - 2 Amendment No. 134 1
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| | |
| l Control Rod Scram Accumulators B 3.1.5 i
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| ,O- BASES
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| ,L)
| |
| ACTIONS (continued) performance with an inoperable accumulatcr could result in excessive scram times, In this event, the associated control rod is declared inoperable (Required Action A.2) and LC0 3.1.3 is entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function, in accordance with ACTIONS of LC0 3.1.3.
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| The allowed Completion Time of 8 hours is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures.
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| B.1. B.2.1. and B.2.2 With two or more control rod scram accumulators inoperable and reactor steam dome pressure a 900 psig, adequate pressure must be supplied to the charging water header. I With inadequate charging water pressure, all of the 1 accumulators could become inoperable, resulting in a potentially severe degradation of the scram performance.
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| O~ Therefore-, within 20 minutes from discovery of charging water header pressure < 940 psig concurrent with i
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| i Condition B, adequate charging water header pressure must be restored. The allowed Completion Time of 20 minutes is reasonable, to place a CRD pump into service to restore the charging header pressure, if required. This Completion Time is based on the ability of the reactor pressure alone to fully insert all control rods, The control rod may be declared " slow," since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 3.1.4-1. Required Action B.2.1 is modified by a Note indicating that declaring the control rod " slow" only ap) lies if the associated control scram time is within tie limits of Table 3.1.4-1 during the last scram time test. Otherwise, the control rod would already be considered " slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control rod is declared ino)erable
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| ; (Required Action B.2.2) and LC0 3.1.3 entered. T11s would l result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LC0 3.1.3.
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| I(t FERMI UNIT 2 B 3.1.5 -3 Amendment No. 134 L
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| | |
| Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS (continued)
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| The allowed Completion Time of 1 hour is reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.
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| C.1 and C.2 With one or more control rod scram accumulators inoperable and the reactor steam dome pressure < 900 psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure < 900 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during a depressurization event or at low reactor pressures. Therefore, immediately upon discovery of charging water header pressure < 940 psig, concurrent with Condition C. all control rods associated with inoperable accumulators must be verified to be fully inserted.
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| Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions. The O associated control rods must also be declared ino>erable V within 1 hour. The allowed Completion Time of 1 lour is reasonable for Required Action C.2 considering the low 4 probability of a DBA or transient occurring during the time !
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| that the accumulator is inoperable.
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| D.J. !
| |
| The reactor mode switch must be immediately placed in the I shutdown position if either Required Action and associated- 1 Completion Time associated with loss of the CRD charging pump (Required Actions B.1 and C.1) cannot be met. This ensures that all insertable control rods are inserted and ,
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| that the reactor is in a condition that does not require the 1 active function (i.e.. scram) of the' control rods. This Required Action is modified by a Note stating that the ;
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| action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.
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| A V
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| FERMI UNIT 2 B 3.1.5 - 4 Amendment No. 134
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| 1 i
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| Control Rod Scram Accumulators B 3.1.5 BASES SURVEILLANCE SR 3.1.5.1 I REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists ,
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| to provide sufficient scram force. The primary indicator of '
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| accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 )sig is established to assure a margin of accumulator OPERABI_ITY sufficient to scram the associated control rod (Ref.1).
| |
| Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant )
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| degradation in scram times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.
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| REFERENCES 1. UFSAR, Section 4.5.2.2.3.
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| : 2. UFSAR, Chapter 15.
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| O FERMI UNIT 2 B 3.1.5- 5 Amendment No. 134 t
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| | |
| c Rod Pattern Control B 3.1.6 i B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM)
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| (LC0 3.3.2.1, " Control Rod Block Instrumentation") so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods fully inserted to 10% RTP. The sequences limit the l potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).
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| This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1 and 2.
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| l APPLICABLE The analytical methods and assumptions used in evaluating i SAFETY ANALYSES the CRDA are summarized in References 1 and 2. CRDA p analyses assume that the reactor operator follows prescribed l
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| Q withdrawal sequences. . These sequences define the potential initial conditions for the CRDA analysis. The RWM l (LC0 3.3.2.1) provides backup to operator control of the l withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.
| |
| Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for U0 have been shown to be insignificant below fuel en,ergy depositions of 300 cal /gm
| |
| ! (Ref. 3), the fuel damage limit of 280 cal /gm provides a j
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| : l. margin of safety from significant core damage which would i l result in release of radioactivity (Refs. 4 and 5). Generic '
| |
| evaluations (Refs. I and 6) of a design basis CRDA (i.e., a l l CRDA resulting in a peak fuel energy deposition of l 280 cal /gm) have shown that if the peak fuel enthalpy remains below 280 cal /gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7) and the calculated offsite doses will be well within the required limits (Ref. 5).
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| ,y b
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| FERMI - UNIT 2 B 3.1.6 - 1 Amendment No. 134
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| 1 Rod Pattern Control B 3.1.6 1
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| fl BASES l V {
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| APPLICABLE SAFETY ANALYSES (continued) l Control rod patterns analyzed in Reference 1 follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10t RTP (Ref. 2). For the BPWS. the control !
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| rods are required to be moved in groups, with all control -
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| rods assigned to a specific group required to be within specified banked positions. The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant i operation. Generic analysis of the BPWS (Ref.1) has '
| |
| demonstrated that the 280 cal /gm fuel damage limit will not be violated during a CRDA while following the BPWS.
| |
| Control rod withdrawal sequences are normally established consistent with the rules of the generic BPWS analysis.
| |
| Occasionally operational limitations (e.g.. power suppression of failed fuel) may dictate the insertion of control rods which do not meet the minimum cell separation criteria of the generic BPWS analysis. In such situations, sufficient cycle s)ecific analyses are performed to demonstrate that tie resulting control rod worths of the A modified control rod withdrawal sequence are bounded by the V rod worths allowed by rigorously following the rules of the l generic BPWS analysis, thereby assuring that the 280 cal /gm i fuel damage limit will not be violated during a CRDA.
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| The " prescribed withdrawal sequence" is defined as the combination of both the procedurally specified control rod movement sequence and any analytically allowed deviations l from this sequence. Some prescribed withdrawal sequences 1 (e.g., BPWS (Ref. 8)) have more flexibility in allowed deviations than other prescribed withdrawal sequences (e.g..
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| a cycle specific sequence developed for power suppression of failed fuel may not allow any deviations).
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| Rod pattern control satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| FERMI - UNIT 2 B 3.1.6 - 2 Amendment No. 134
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| Rod Pattern Control J B 3.1.6
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| /7 BASES V
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| LC0 Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with CRDA assum rods.ptions. This LC0 only For inoperable applies control rodstorequired OPERABLEto becontrol inserted, separate requirements are specified in LC0 3.1.3, " Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the prescribed withdrawal sequence.
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| APPLICABILITY In MODES 1 and 2, when THERMAL POWER is s 10t RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5 since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the q reactor will remain subcritical with a single control rod V withdrawn.
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| ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours.
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| Noncompliance with the prescribed sequence may be the result of " double notching," drifting from a control rod drive 3 cooling water transient, leaking scram valves, or a power ;
| |
| reduction to s 10% RTP before establishing the correct i control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the rod pattern, or scram if warranted.
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| FERMI - UNIT 2 B 3.1.6 - 3 Amendment No. 134
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| Rod Pattern Control B 3.1.6 BASES ACTIONS (continued)
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| L Recuired Action A.1 is modified by a Note which allows the to be bypassed to allow the affected control rods to be l RWP returned to their correct position. LC0 3.3.2.1 requires.
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| verification of control rod movement by a qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. -A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2.
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| OPERABILITY of control rods is determined by compliance with LCO 3.1.3, " Control Rod OPERABILITY," LC0 3.1.4, " Control Rod Scram Times," and LC0 3.1.5, " Control Rod Scram Accumulators." The allowed Completion Time of 8 hours is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.
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| B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be O suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod l insertion to correct control rods withdrawn beyond their allowed position is allowed.since, in general, insertion of j control rods has less impact on control rod worth than '
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| ' withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be inserted to their correct position. ;
| |
| LC0 3.3.2.1 requires verification of control rod movement by a qualified member of. the technical staff.
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| -When nine or more OPERABLE control rods are not in compliance with the prescribed withdrawal sequence, the reactor mode switch must be placed in the shutdown oosition within 1 hour. With the mode switch in shutdown, t1e reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
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| O FERMI - UNIT 2 B 3.1.6 -4 Amendment No. 134
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| r Rod Pattern Control l B 3.1.6
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| . BASES SURVEILLANCE SR 3.1.6.1 REQUIREENTS The control rod pattern is verified to be in compliance with the prescribed withdrawal sequence at a 24 hour Frequency to !
| |
| ensure the assumptions of the CRDA analyses are met. The 24 hour Frequency was developed considering that the primary !
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| check on compliance with the prescribed withdrawal sequence is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at s 10% RTP.
| |
| REFERENCES 1. NEDE-24011 P-A 9 US, " General Electric Standard Application for Reactor Fuel, Supplement for United States," Section S.2.2.3.1, September 1988.
| |
| : 2. " Modifications to the Requirements for Control Rod Drop Accident Mitigating System," BWR Owners Group, July 1986.
| |
| : 3. NUREG 0979, Section 4.2.1.3.2, April 1983.
| |
| : 4. NUREG 0800 Section 15.4.9. Revision 2 July 1981.
| |
| : 5. 10 CFR 100.11.
| |
| : 6. NED0-21778 A, " Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water -
| |
| Reactors," December 1978.
| |
| : 7. ASME, Boiler and Pressure Vessel Code.
| |
| : 8. NED0 21231. " Banked Position Withdrawal Sequence,"
| |
| January 1977.
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| l l
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| O FERMI UNIT 2 B 3.1.6 - 5 Amendment No. 134 L_
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| l
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| [
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| SLC System i B 3.1.7 l
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| '(')
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| lv B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System l BASES l BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref.1) on anticipated transient without scram.
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| I The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The i borated solution is discharged near the bottom of the core I shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized {
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| n water is provided for testing purposes.
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| 1 U \
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| I APPLICABLE The SLC System is manually initiated from the main control l SAFETY ANALYSES room, as directed by the emergency operating 3rocedures, if l the operator believes the reactor cannot be slut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be
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| , inserted to accomplish shutdown and cooldown in the normal )
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| manner. The SLC System injects borated water into the I reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration equivalent to 720 ppm of natural boron, in the reactor coolant at 70*F. To allow for potential leakage and imperfect mixing in the reactor system, an additional amount t of boron equal to 25% of the amount cited above is added l (Ref. 2) . The volume versus concentration limits in l l Figure 3.1.7-1 and the temperature limit in SR 3.1.7.2 are l l calculated such that the required concentration is achieved )
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| accounting for dilution in the RPV with reactor water level '
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| at Level 8 and including the water volume in the residual O
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| V l FERMI - UNIT 2 B 3.1.7 - 1 Amendment No. 134 l
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| l
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| ! l L l
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| SLC System B 3.1.7 BASES APPLICABLE SAFETY ANALYSES (continued) heat removal shutdown cooling piping and in the 1 recirculation loop piping. This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected.
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| The SLC System satisfies the requirements of 10 CFR 50.36(c)(2)(ii) because operating experience and probabilistic risk assessments have shown the SLC System to i be important to public health and safety. Thus, it is '
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| retained in the Technical Specifications.
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| LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV. including the
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| /7 OPERABILITY of the pumps and valves. Two SLC subsystems are U required to be OPERABLE: each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.
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| APPLICABILITY In MODES 1 and 2. shutdown capability is required. In MODES 3 and 4 control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control l rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5.
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| only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LC0 3.1.1 " SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.
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| O V
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| FERMI - UNIT 2 B 3.1.7 - 2 Amendment No. 134 m
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| SLC System B 3.1.7 BASES ACTIONS M If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of )erforming the intended SLC System function and the low pro) ability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the plant.
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| IL1 If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.
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| M If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
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| 3 O
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| FERMI - UNIT-2 B 3.1.7 - 3 Amendment No. 134
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| SLC System B 3.1.7
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| [3 wJ BASES SURVEILLANCE SR '4.I.7.1. SR 3.1.7.2. and SR 3.1.7.3 REQUIREMENTS SR 3.1.7.1 through SR 3.1.7.3 are 24 hour Surveillances verifying certain characteristics of the SLC System (e.g..
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| the volume and temperature of the borated solution in the storage tank). thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution volume and temperature, including the temperature of the pump suction piping, are maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate oic in the storage tank or in the pump suction piping. The 24 hour Frequency is based on operating experience and has shown there are relatively slow variations in the measured parameters of volume and temperature.
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| SR 3.1.7.4 and SR 3.1.7.6 SR 3.1.7.4 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will p) y occur if required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on o)erating experience and has demonstrated the reliability of t1e explosive charge continuity.
| |
| SR 3.1.7.6 verifies that each manual valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manual valves in the SLC System flow path provides assurance that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This Surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment does not require any testing or valve manipulation: rather, it involves verification that those valves ca position.pable ThisofSR being doesmispositioned are inthat not a) ply to valves thecannot correctbe inadvertently misaligned, suc1 as check valves. The 31 day Frequency is based on engineering judgment and is consistent A
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| V FERMI - UNIT 2 B 3.1.7 - 4 Amendment No. 134 1
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| SLC System B 3.1.7 O
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| O BASES SURVEILLANCE REQUIREMENTS (continued) with the procedural controls governing valve operation that ensures correct valve positions.
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| SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank. SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.
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| SR 3.1.7.5 must also be performed anytime the temperature is restored to a 48'F to ensure that no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.
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| SR 3.1.7.7 Demonstrating that each SLC System pump develops a flow rate a 41.2 gpm at a discharge pressure = 1215 asig ensures that g pump performance has not degraded during t1e fuel cycle.
| |
| ( This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is in accordance with the Inservice Testing Program.
| |
| SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV.
| |
| including the firing of an explosive valve. The replacement charge for the ex)losive valve shall be from the same manufactured batc1 as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are O
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| FERMI UNIT 2 B 3.1.7 - 5 Amendment No. 134
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| SLC System B 3.1.7 BASES SURVEILLANCE REQUIREENTS (continued) testea every 36 months at alternating 18 month intervals.
| |
| The Surveillance may be performed in separate steps to prevent. injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency:
| |
| therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
| |
| Demonstrating that all piping between the boron solution storage tank and the explosive valve is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the test tank (this is followed by draining and flushing the piping with demineralized water).
| |
| The 18 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the piping.
| |
| This is especially true in light of the temperature verification of this piping required by SR 3.1.7.3.
| |
| However, if, in performing SR 3.1.7.3. it is determined that the temperature of this piping has fallen below the specified minimum. SR 3.1.7.9 must be performed once within 24 hours after the piping temperature is restored to = 48'F.
| |
| SR 3.1.7.10 Enriched sodium pentaborate solution is made by mixing-granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate to verify the actual B 10 enrichment must be performed >rior to addition l to the SLC tank in order to ensure that tw proper B-10 atom percentage.is being used. ;
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| [
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| i i
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| FERMI - UNIT 2 B 3.1.7 - 6 Amendment No. 134
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| SLC System B 3.1.7 O Bases REFERENCES 1. 10 CFR 50.62.
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| : 2. 'UFSAR, Section 4.5.2.4.
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| O J s
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| i G
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| FERMI . UNIT 2 B3.1.7-7 Amendment No. 134
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| SDV Vent and Drain Valves B 3.1.8 O B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves l
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| BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain ;
| |
| valves close to contain reactor water. The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into two instrument volumes. The two instrument volumes each receive approximately one half of )
| |
| .the control rod drive (CRD) discharges. The two instrument volumes are connected to a common drain line with two valves in series. The header is connected to a common vent line 1 with two valves in series. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1.
| |
| O APPLICABLE The Design Basis Accident and transient analyses assume all SAFETY ANALYSES of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:
| |
| : a. Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100 (Ref. 2); and
| |
| : b. Open'on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.
| |
| Isolation of the SDV can also be accomplished by manual ,
| |
| closure of the SDV valves. Additionally, the discharge of I reactor coolant to the SDV can be terminated by scram reset i or closure of the HCU manual isolation valves. For a l
| |
| . bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 (Ref. 2), and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation to ensure that the SDV has sufficient capacity to contain O
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| FERMI - UNIT 2 B 3.1.8- 1 Amendment No. 134
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| | |
| SDV Vent and Drain Valves B 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued) the reactor coolant discharge during a full core scram. To automatically ensure this capacity, a reactor scram (LC0 3.3.1.1, " Reactor Protection System (RPS)
| |
| Instrumentation") is initiated if the SDV water level in the instrument volume exceeds a specified setpoint. The setpoint is chosen so that all' control rods are inserted before the SDV has insufficient volume to accept a full scram.
| |
| SDV vent and drain valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV aiping. I Since the vent and drain lines are provided wit 1 two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, the valves are required to open on scram reset to ensure that a path is available for the SDV (Q,/ piping to drain freely at other times.
| |
| APPLICABILITY In MODES 1 and 2 scram may be required; therefore, the SDV vent and drain valves must be OPERABLE. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that only a single control rod can be withdrawn. Also, during MODE 5. only a single control rod can be withdrawn from a core cell containing fuel assemblies. Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram.
| |
| ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each SDV vent and drain line. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent v
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| FERMI UNIT 2 83.1.8-2 Amendment No. 134 l
| |
| | |
| c SDV Vent and Drain Valves )
| |
| 8 3.1.8 j
| |
| ( BASES ACTIONS (continued) inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions.
| |
| /L1 When one SDV vent or drain valve is inoperable in one or more lines, the valves must be restored to OPERABLE status within 7 days. The Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a scram occurring while the valve (s) are inoperable. The SDV is still isolable since the redundant valve in the affected line is OPERABLE. During these periods, the single failure criterion may not be preserved, and a higher risk exists to allow reactor water out of the primary system during a scram.
| |
| IL1 If both valves in a line are inoperable the line must be 4 isolated to contain the reactor coolant during a scram.
| |
| When a line is isolated, the potential for an inadvertent
| |
| ~N scram due to high SDV level is increased. Required (d ,
| |
| Action B.1 is modified by a Note that allows periodic draining and venting of the SDV when a line is isolated.
| |
| During these periods, the line may be unisolated under administrative control. This allows any accumulated water in the line to be drained, to preclude a reactor scram on SDV high level. This is acceptable since the administrative controls ensure the valve can be closed quickly, by a dedicated operator, if a scram occurs with the valve open.
| |
| The 8 hour Completion Time to isolate the line is based on the low probability of a scram occurring while the line is i not isolated ard unlikelihood of significant CRD seal leakage.
| |
| L1 If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is ceasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
| |
| O O
| |
| FERMI - UNIT 2 B 3.1. 8 - 3 Amendment No. 134
| |
| | |
| SDV Vent and Drain Valves B 3.1.8 C' BASES V) -
| |
| SURVEILLANCE SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when closed intermittently under administrative control for testing) to allow for drainage of the SDV piping. Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform the:r intended functions during ,
| |
| normal operation. This SR doas not require any testing or )
| |
| valve manipulation: rather, it involves verification that l the valves are in the correct position. j The31osyFrehuencyisbasedonengineeringjudgmontandis consistent wit the procedural controls governing valve operation, which ensure correct valve positions.
| |
| SR 3.1.8.2 SR 3.1.8.2 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The closure time of q 30 seconds after receipt of a scram signal is based on the Q bounding leakage case evaluated in the accident analysis.
| |
| Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 that overlaps this Surveillance and the scram time testing of control rods in LC0 3.1.3 to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un31anned transient if the Surveillance were performed with tie reactor at power.
| |
| 0)erating experience has shown these components usually pass t1e Surveillance when performed at the 18 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
| |
| REFERENCES 1. UFSAR, Section 4.5.2.2.2.3.
| |
| : 2. 10 CFR 100.
| |
| : 3. NUREG-0803. " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping." August 1981.
| |
| O FERMI - UNIT 2 B 3.1.8 - 4 Amendment No. 134
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| | |
| 1 APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
| |
| BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during anticipated operational occurrences (A00s) and that the peak cladding temperature '(PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.
| |
| 4 APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in References 1 and 2.
| |
| The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), anticipated operational transients, and normal operation that determine the APLHGR limits are presented in References 1. 2. 3. 4. 5. 6. and 7.
| |
| (~h V Fuel design evaluations are performed to demonstrate that the it limit on the fuel cladding plastic strain and other fuel design limits described in Reference 1 are not exceeded during A00s for operation with LHGRs up to the operating limit LHGR. APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the l fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the ;
| |
| limiting A00s (Refs. 5. 6. and 7). Flow dependent APLHGR limits are determined using the three dimensional BWR simulator code (Ref. 8) to analyze slow flow runout I transients. The flow dependent multiplier. MAPFACr.1s ;
| |
| dependent on the maximum core flow runout capability. The maximum runout flow is dependent on the existing setting of the recirculation scoop tube mechanical stop in the Recirculation Flow Control System.
| |
| Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power de>endent multipliers. MAPFAC,. are also generated. Due to t1e sensitivity of the transient response to initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve rs U.
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| FERMI UNIT 2 B 3.2.1 - 1 Amendment No. 134
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| | |
| APLHGR B 3.2.1 BASES APPLICABLE SAFETY ANALYSES (continued) fast closure scram trips are bypassed, both high and low core flow MAPFAC, limits are provided for operation at power levels between 25% RTP and the previously mentioned bypass power level. The exposure dependent APLHGR limits are reduced by MAPFAC, and MAPFACr at various operating conditions to ensure that all fuel design criteria ire met for normal operation and A00s. A complete discussion of the' analysis code is provided in Reference 9.
| |
| LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50. Appendix K. A complete discussion of the analysis code is provided in Reference 10.
| |
| The PCT following a postulated LOCA is a function of the average heat _ generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A O conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.
| |
| The APLHGR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The APLHGR limits specified in the COLR are the result of the fuel design, DBA. and transient analyses. The limit is determined by multiplying the smaller of the MAPFAC, and MAPFACr factors times the exposure dependent APLHGR limits.
| |
| APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 7) and operating experience have shown that as power is reduced, the margin to the required fuel design limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2. the intermediate range monitor scram function provides prompt scram initiation during any significant O
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| FERMI - UNIT 2 B 3.2.1 - 2 Amendment No. 134
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| | |
| APLHGR B 3.2.1 BASES APPLICABILITY (continued) transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels < 25% RTP. the reactor is operating with substantial margin to the fuel design limits: thus, this LC0 is not '.
| |
| required.
| |
| ACTIONS M If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA and transient analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the f27 rods. The 2 hour Com)letion Time.is sufficient va restore the APLHGR(s) to wit 11n its limits and is accep.able based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.
| |
| O If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.
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| FERMI - UNIT 2 B 3.2.1- 3 Amendment No. 134 i
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| | |
| APLHGR l B 3.2.1 l BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS >
| |
| APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is = 25% RTP and then every 24 hours thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after 4 THERMAL POWER = 25% RTP is achieved is acceptable given the large inherent margin to fuel design limits at low power levels.
| |
| REFERENCES 1. NED0 24011-P-A " General Electric Standard Application for Reactor Fuel" (latest approved version).
| |
| : 2. UFSAR, Chapter 4.
| |
| : 3. UFSAR, Chapter 6.
| |
| : 4. UFSAR, Chapter 15. I l
| |
| S. mE 56 0386, " Fermi 2 Single Loop Operation Analysis,"
| |
| April 1987, and NEDC-32313 P, "Enrico Fermi Energy Center Unit 2 Single Loop Operation," September 1994.
| |
| : 6. NEDC 31515, Rev.1, " Maximum Extended Load Line Limit and Feedwater Heater Out of-Service Analysis for Enrico Fermi Atomic Power Plant Unit 2 " August 1989.
| |
| : 7. NEDC-31843, " Maximum Extended Operating Domain Analysis for Detroit Edison Company Enrico Fermi Energy Center Unit 2." July 1990.
| |
| : 8. NED0 30130 A, " Steady State Nuclear Methods," May 1985.
| |
| : 9. NED0 24154, " Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors,"
| |
| October 1978.
| |
| : 10. NEDC 31928. " Fermi 2 SAFER /GESTR LOCA, Loss of-Coolant Accident Analysis," July 1991. Erratta and Addenda.
| |
| April 1992.
| |
| . O, FERMI UNIT 2 B 3.2.1-4 Amendment No. 134
| |
| | |
| MCPR B 3.2.2 i B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
| |
| BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2. The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (A00s). Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref.1),
| |
| the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
| |
| The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,
| |
| the bundle power level at the onset of transition boiling) r '
| |
| for a given set of plant parameters (e.g., reactor vessel 5
| |
| pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
| |
| l APPLICABLE The analytical methods and assumptions used in evaluating l SAFETY ANALYSES the A00s to establish the operating limit MCPR are ) resented in References 2, 3, 4, 5, 6, 7, and 8. To ensure t1at the ;
| |
| MCPR SL is not exceeded during any transient event that !
| |
| occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (ACPR).
| |
| When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.
| |
| q O
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| FERMI - UNIT 2 B 3.2.2 -1 Amendment No. 134
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| | |
| L MCPR B 3.2.2
| |
| . BASES-APPLICABLE SAFETY ANALYSES (continued)
| |
| The MCPR operating limits derived-from the transient analysis are dependent on the operating core flow and power state (MCPRr and MCPR,, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 6, 7, and 8). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 9) to analyze slow flow runout transients. The operating limit is dependent on the maximum recirculation scoop tube mechanical stop setting in the Recirculation Flow Control System.
| |
| Power dependent MCPR limits (MCPR,) are determined mainly by the one dimensional transient code (Ref.10). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPR operating limits are provided for operating between 25% R,TP and the previously mentioned bypass power level.
| |
| O Transients involving increase in pressure and power are sensitive to the size of the steam volume and the availability of this steam volume to accommodate the reactor steam production. Larger steam volumes and longer or earlier availability result in less-severe pressure transients. Thus operation of the turbine generator bypass valves and the availability of the moisture separator reheater have an effect on the transient results. For this reason the COLR contains MCPR limits for when the turbine bypass valves and/or moisture separator reheater are out-of-service (refer to LC0 3.7.6, "The Main Turbine Bypass System and Moisture Separator Reheater").
| |
| The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0' The MCPR operating limits specified in the COLR are the result of the design basis transient analysis. The operating limit MCPR is determined by the larger of the MCPRr and MCPR, limits.
| |
| O FERMI . UNIT 2 B 3.2.2-2 Amendment No. 134 l
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| | |
| MCPR B 3.2.2
| |
| ( BASES APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the moderator void ratio is small.
| |
| Surveillance of thermal limits below 25% RTP is unnecessary I due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.
| |
| Statistical analyses indicate that the nominal value of the initial MCPR expected at 25t RTP is > 3.5. Studies of the variation of limiting transia behavior have been performed over the range of power and ' .ow conditions. These studies encompass the range of key ac:ual plant parameter values {
| |
| important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR recuirements. and that margins increase as power is reducec to 25% RTP. This trend is ,
| |
| expected to continue to the 5% to 15% power range when entry I into MODE 2 occurs. When in MODE 2. the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels
| |
| < 25% RTP. the reactor is operating with substantial margin to the MCPR SL and this LC0 is not required. l
| |
| )
| |
| ACTIONS /L1 If any MCPR is outside the required limits an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low arobability of a transient occurring simultaneously with tie MCPR out of specification.
| |
| IL1 If the MCPR cannot be restored to within its required limits within the associated Completion Time the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status. THERMAL POWER must be reduced to < 25% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.
| |
| FERMI - UNIT 2 B 3.2.2-3 Amendment No. 134
| |
| | |
| MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is = 25% RTP and then every 24 hours thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. . The 24 hour Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 12 hour allowance after THERMAL. POWER = 25% RTP is achieved is acceptable given the large inherent margin to the MCPR safety limit at low power levels.
| |
| SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the value of r, which is a measure of the actual scram speed distribution compared with the assumed distribution. For r
| |
| > 0, the MCPR operating limit is then determined based on an Ox interpolation between the applicable limits for ODYN Option A (scram times of LC0 3.1.4," Control Rod Scram Times") and ODYN Option B (realistic scram times) analyses.
| |
| The parameter r and MCPR operating limit must be determined once within 72 hours after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle. The 72 hour Completion Time is acceptable due to the relatively minor changes in r expected during the fuel cycle.
| |
| REFERENCES 1. NUREG 0562. June 1979.
| |
| : 2. NED0 24011 P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
| |
| : 3. UFSAR, Chapter 4.
| |
| i
| |
| : 4. UFSAR, Chapter 6.
| |
| : 5. UFSAR, Chapter 15.
| |
| O FERMI UNIT 2 B 3.2.2-4 Amendment No. 134
| |
| | |
| MCPR B 3.2.2 I)
| |
| M BASES REFERENCES (continued)
| |
| : 6. 10E 56-0386, " Fermi 2 Single Loop Operation Analysis,"
| |
| April 1987, and NEDC-32313 P, "Enrico Fermi Energy Center Unit 2 Single Loop Operation," September 1994.
| |
| : 7. NEDC 31515. Rev.1 " Maximum Extended Load Line Limit and Feedwater Heater Out of Service Analysis for Enrico Fermi Atomic Power Plant Unit 2," August 1989.
| |
| : 8. NEDC 31843, " Maximum Extended Operating Domain Analysis for Detroit Edison Company Enrico Fermi Energy Center Unit 2," July 1990.
| |
| : 9. NED0-30130 A,' " Steady State Nuclear Methods " May 1985.
| |
| : 10. NED0 24154, " Qualification of the One-Dimensional Core '
| |
| Transient-Model for Boiling Water Reactors,"
| |
| October 1978. ,
| |
| 1 I
| |
| O FERMI - UNIT 2' B 3.2.2-5 Amendment No. 134
| |
| | |
| LHGR B 3.2.3
| |
| ( B 3.2 POWER DISTRIBUTION LIMITS l B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
| |
| BASES-BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (A00s).
| |
| Exceeding the fuel design limits could potentially result in fuel damage and rubsequent release of radioactive materials.
| |
| LHGR limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the anticipated operating conditions identified in Reference 1.
| |
| APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in References 2 and 3.
| |
| The fuel assembly is designed to ensure (in conjunction with
| |
| /7 the core nuclear and thermal hydraulic design, plant V equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, 50, and 100. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:
| |
| : a. Rupture of the fuel rod cladding caused by strain from the relative expansion of the U0 2 pellet: and
| |
| : b. Severe overheating of the fuel rod cladding caused by inadequate cooling.
| |
| A value of it plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).
| |
| 1 l
| |
| r~'s O
| |
| FERMI - UNIT 2 B 3.2.3 1 Amendment No. 134
| |
| | |
| rt LHGR B 3.2.3 BASES APPLICABLE SAFETY ANALYSES (continued)
| |
| Fuel design evaluations have been performed and demonstrate that the it fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limits specified in the COLR. The analysis also includes allowances for short term transient operation above the operating limit to account for A00s. plus an allowance for densification power spiking.
| |
| The LHGR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). 4
| |
| .LC0 The LHGR is a basic assumption in the fuel design analysis. I The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a It fuel cladding plastic strain. The operating limits to accomplish this objective are specified in the COLR.
| |
| 9 APPLICABILITY The LHGR limits are primarily derived from fuel design (V , analysis that are assumed to occur at high power level conditions. At core thermal power levels < 251 RTP. the reactor is operating with a substantial margin to the fuel design limits and, therefore, the _ Specification is only required when the reactor is operating at r 25% RTP.
| |
| ACTIONS &l If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to restore the LHGR(s) to within its required limits such that the plant is operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the LHGR(s) to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.
| |
| O FERMI -UNIT 2 B 3.2.3 2 Amendment No. 134
| |
| | |
| LHGR B 3.2.3 BASES ACTIONS (continued) 6.1 If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status, THERMAL POWER is reduced to < 25% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER TO < 25% RTP in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LHGR is required to be initially calculated within 12 hours after THERMAL POWER is a 25% RTP and then every 24 hours thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour Frequency is based on both engineering judgment and
| |
| (] recognition of the slow changes in power distribution during
| |
| 'v normal operation. The 12 hour allowance after THERMAL POWER >
| |
| = 25% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels.
| |
| REFERENCES 1. UFSAR, Chapter 15.
| |
| : 2. UFSAR, Chapter 4.
| |
| : 3. NED0 24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
| |
| 1 l
| |
| O !
| |
| FERMI UNIT 2 B 3.2.3 3 Amendment No. 134
| |
| | |
| r ,
| |
| RPS Instrumentation B 3.3.1.1 l
| |
| B 3.3 -INSTRUMENTATION
| |
| []
| |
| B.3.3.1.1 Reactor Protection System (RPS) Instrumentation BASES i
| |
| BACKGROUND The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits. to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). This can be accomplished either automatically or manually.
| |
| The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is !
| |
| achieved by specifying limiting safety system settings ]
| |
| (LSSS) in terms of parameters directly monitored by the RPS. 5 as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the-LCOs, establish the threshold for protective system I action to prevent exceeding acceptable limits, including l Safety Limits (SLs) during Design Basis Accidents (DBAs).
| |
| The RPS, as shown in the UFSAR. Figure 7.2-2 (Ref.1),
| |
| includes sensors, relays, bypass circuits, and switches that are necessary to cause initiation of a reactor scram.
| |
| Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve (TCV) fast closure, turbine stop valve (TSV) position drywell pressure, main steam line radiation, and scram discharge volume (SDV) water level as well as reactor mode switch in shutdown position and manual
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| . scram. signals. There are at least four redundant sensor s input signals from each of these parameters (with the exception of the reactor mode switch in shutdown scram signal). Most channels include electronic equipment (e.g.,
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| trip units) that compares measured input signals with i pre-established setpoints. When the setpoint is exceeded. ;
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| the channel output relay actuates, which then outputs an RPS :
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| trip signal to the trip logic. j O ;
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| FERMI -_ UNIT 2 B 3.3.1.1- 1 Amendment No. 134 L
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| RPS Instrumentation B 3.3.1.1 !
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| BASES BACKGROUND (continued)
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| The RPS is comprised of two independent trip systems (A and B) with two logic channels in each trip system (logic channels Al and A2, B1 and B2) as shown in Reference 1. The outputs of the logic channels in a trip system are combined in a one out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram. This logic arrangement is referred to as a one out-of two taken twice logic. The trip systems can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for 10 seconds after the full scram signal is received. This 10 second delay on reset ensures that the scram function will be completed.
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| Two scram pilot valves are located in the hydraulic control unit for each control rod drive (CRD). Each scram pilot valve is solenoid operated, with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD.
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| When either scram pilot valve sulenoid is energized, air j D pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de energized to cause a control rod to scram. The scram valves control the su )
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| and discharge paths for the CRD water during a scram. Onepply of the scram pilot valve solenoids for each CRD is controlled by trip system A, and the other solenoid is controlled by trip system B. Any trip of trip system A in conjunction with any trip in trip system B results in de energizing both solenoids, air bleeding off, scram valves opening, and control rod scram. 4 The backup scram valves, which energize on a scram signal to depressurize the scram air header, are also controlled by the RPS. Additionally, the RPS controls the SDV vent and drain valves such that when both trip systems trip, the SDV vent and drain valves close to isolate the SDV.
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| O FERMI UNIT 2 B 3.3.1.1 - 2 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE The actions of the RPS are assumed in the safety analyses of SAFETY ANALYSES, Reference 7. The RPS initiates a reactor scram LC0. and when monitored parameter values exceed the Allowable Values.
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| APPLICABILITY specified by the setpoint methodology and listed in Table 3.3.1.1-1 to preserve the integrity of the fuel cladding, the reactor coolant pressure boundary (RCPB), and the containment by minimizing the energy that must be absorbed following a LOCA.
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| RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Functions not specifically credited 'in the accident analysis are retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
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| The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1. Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
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| O (Ref. 11) . Certain channels must also respond within their assumed response times given by Reference 10.
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| Allowable Values where applicable are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS or between successive verifications of trip unit setpoints. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowtble Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required
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| ' Allowable Value.
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| Trip setpoints are those ) redetermined values of output at which an action should ta(e place. The setpoints are compared to the actual process carameter (e.g., reactor vessel water level), and when t1e measured output value of the process parameter exceeds the setpoint, the associated device (e.g. trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis or other design basis documents. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and O
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| FERMI - UNIT 2 B 3.3.1.1 - 3 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g. , drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. The methodology used to determine Fermi allowable values and trip setpoints is given by Reference 11.
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| The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV . valves, described in the Background section, are not addressed by this LCO.
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| The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in MODES 1, 2 and 5 to provide primary and diverse
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| .in tiation signals.
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| Certain RPS functions are required to be OPERABLE in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. Control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram. Provided all other control rods remain inserted, the RPS function is not required. In this condition, the required SDM (LC0 3.1.1) and refuel position one rod out interlock (LC0 3.9.2) ensure that no event requiring RPS will occur. No RPS function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LC0 3.3.2.1) does not allow any control rod to be withdrawn.
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| The specific Applicable Safety Analyses. LCO, and Applicability discussions are listed below on a Function by function basis.
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| O FERMI - UNIT 2 B 3.3.1.1 - 4 Amendment No. 134 1- - _ .
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| r RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES..LCO, and APPLICABILITY (continued)
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| Intermediate Ranoe Monitor (IRM) 1.a. Intermediate Ranoe Monitor Neutron Flux-Hioh The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the average power range monitors (APRMs). The IRMs are capalle of generating trip signals that can be used to prevent fuel damage resulting from abnormal operating transients in the intermediate power range. In this power range, the most
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| .significant source of reactivity change is due to control rod withdrawal. The IRM provides protection that is in addition to the roo worth minimizer (RWM).-which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref. 2). The IRM mitigation of the neutron flux excursion. provides To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref. 3) to evaluate the consequences of control rod v'
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| () withdrawal events during startup that are citigated only by
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| . the IRM. This analysis, which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against-local control rod withdrawal errors and results in peak fuel enthalpies below the 170 cal /gm fuel failure threshold criterion.
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| The IRMs are also capable of limiting other reactivity excursions during startup, such as the control rod drop
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| ' accident in the start-up MODE, although no credit is specifically assumed.
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| The IRM System is divided into two groups of IRM channels.
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| with four IRM channels inputting to each trip system. The analysis of Reference 3 assumes that one channel in each l trip system is bypassed. Therefore, six channels with three !
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| channels in each trip system are required for IRM OPERABILITY to ensure that no single instrument failure will '
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| preclude a scram from this Function on a valid signal. This trip is active in each of the 10 ranges of the IRM, which must be selected by the operator to maintain the neutron flux within the. monitored level of an IRM range.
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| O FERMI --UNIT 2 B 3.3.1.1 - 5 Amendment No. 134 s
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO,-and APPLICABILITY (continued)
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| The analysis of Reference 3 has adequate conservatism to permit an IRM Allowable Value of 122 divisions of a 125 division scale.
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| The Intermediate Range Monitor Neutron Flux-High Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists. In MODE 5, when a cell with fuel has its control rod withdrawn, the IRMs provide monitoring for and protection against unexpected reactivity excursions. In MODE 1, the RBM System-and the RWM provide protection against control rod withdrawal error events and the IRMs are not required.
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| Therefore, the IRM's are automatically bypassed when the
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| . mode switch is in the run position.
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| 1.b. Intermediate Ranoe Monitor-Inoo This trip signal provides assurance that a minimum number of IRMs are OPERABLE. Anytime an IRM mode switch is moved to any position other than " Operate." the detector voltage drops below a preset level, or when a module is not ) lugged C in, an inoperative trip signal will be received by t1e RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal.
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| This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
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| Six channels of Intermediate Range Monitor-Inop with three channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
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| Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.
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| This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux-High Function is required..
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| O FERMI - UNIT 2 B 3.3.1.1 - 6 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| Averaoe Power Ranoe Monitor The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP.
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| The APRM System is divided into 4 APRM channels and 4 2-out-of-4 voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The APRM System is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip"from any one unbypassed APRM will result in a " half trip in all four voter channels, but no trip inputs to either RPS trip system. A trip from any two O unby)assed APRM channels will result in a full-trip in each of t1e four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1, A2, B1, and B2). Three of the four APRM channels and all four of the voter channels are required to be 0"ERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a.
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| 2.b and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the l LPRMs are located, are required for each APRM channel.
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| 2.a. Averaoe Power Ranoe Monitor Neutron Flux-Uoscale (Setdown)
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| For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low Average Power Range Monitor Neutron Fluxpower levels, the Upscale (Setdown)
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| Function will provide a secondary scram to the Intermediate Range Monitor Neutron Flux-High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron O
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| FERMI UNIT 2 B 3.3.1.1 - 7 Amendment No. 134
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| r i
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| RPS Instrumentation I B 3.3.1.1 l BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| Flux-Upscale (Setdown) Function will provide the primary trip signal for a corewide increase in power.
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| The Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function is credited, along with the IRM Neutron Flux High Function, with initiating a reactor scram in the analysis of the continuous rod withdrawal during reactor startup event. This Function also indirectly ensures that before the. reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow.
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| Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER
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| < 25% RTP.
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| The Allowable Value is based on preventing significant I increases in power-when THERMAL POWER is < 25% RTP.
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| The Average Power Range Monitor Neutron Flux-Upscale j (Setdown) Function must be OPERABLE during MODE 2 when j control rods may be withdrawn since the potential for 1 A criticality exists. In MODE 1. the Average Power Range '
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| V Monitor Neutron Flux-Upscale Function provides protection against reactivity transients and the RWM and rod block j monitor protect against control rod withdrawal error events.
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| 2.b. Averaoe Power Ranoe Monitor Simulated Thermal Power-Ooscale The Average Power Range Monitor Simulated Thermal Power-U) scale Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant.
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| The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e.. at lower drive flows, the setpoint is reduced proportional to the reduction in power experienced as drive flow is reduced with a fixed control i rod pattern) but is clamped at an upper limit that is always i lower than the Average Power Range Monitor Neutron !
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| Flux-0) scale Function Allowable Value. The Average Power Range ionitor Simulated Thermal Power-Upscale Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring uG-FERMI J UNIT 2 B 3.3.1.1 - 8 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 p) i BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued) that the MCPR SL is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram.
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| For ra)id neutron flux increase events, the THERMAL POWER lags t1e neutron flux and the Average Power Range Monitor Neutron Flux-Upscale Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power-Upscale Function setpoint is exceeded.
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| The required trip setting for APRMs is dependent on whether the unit is in single recirculation loop operation or two-loop operation, as specified in Table 3.3.1.1-1 footnote (b). The setpoint variable 4W is defined as the difference in indicated drive flow (in t of rated drive flow that produces rated core flow) between two-loop and single-loop operation at the same core flow.
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| Each APRM channel uses one total drive flow signal representative of total core flow. The drive flow signal at rated drive flow is representative of rated core flow at C) b RTP. The total drive flow signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel 0PERABILITY requirements for this Function.
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| The Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power-Upscale Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of approximately 6 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.
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| The Average Power Range Monitor Simulated Thermal Power-Upscale Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5. Other IRM and/or APRM Functions provide protection for fuel cladding integrity.
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| ,m FERMI UNIT 2 B 3.3.1.1 - 9 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 2.c. Averaoe Power Ranoe Monitor Neutron Flux-Uoscale The Average Power Range Monitor Neutron Flux-Upscale Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux-Upscale Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety relief valves (SRVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Neutron Flux-Upscale Function to
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| , terminate the CRDA.
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| The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.
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| The Average Power Range Monitor Neutron Flux-Upscale Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux-4Jpscale Function is assumed in the CRDA analysis, which
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| -is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux-Upscale Function is not required in MODE 2.
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| 2.d. Averaoe Power Ranoe Monitor-Inoo This Function provides assurance that a minimum number of APRMs are OPERABLE. For any APRM channel, any time: 1) its mode switch is in any position other than "0PER": 2) there is a loss of-input power; 3) the automatic self test system detects a critical fault with the APRM channel; or 4) the firmware/ software watchdog timer has timed out, an Inop trip signal is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip
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| . system. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
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| O FERMI - UNIT 2 B 3.3.1.1 -10 Amendment No. 134 i
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| )
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| l RPS Instrumentation B 3.3.1.1 i BASES
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| -APPLICABLE SAFETY ANALYSES, LCO, and APPLICASILITY (continued)
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| There is no- Allowable Value for this Function.
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| This Function is required to be OPERABLE in the MODES where the APRM Functions are required.
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| 2.e. 2 out of 4 Voter The 2 out of-4 Voter Function provides the interface between the APRM Functions and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions.
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| Therefore, the 2 out of 4 Voter Function is required to be OPERABLE in MODES 1 and 2.
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| Both voter channels in each trip system (all four voter channels) are required to be OPERABLE. Each voter channel also includes self diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel to the associated trip system. The 2 out of 4 trip voter includes q separate outputs to RPS for the independently voted sets of V '
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| functions, each of which is redundant (four total outputs).
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| The 2-out of 4 Trip Voter Function is inoperable if any of its functionality is inoperable. Due to the independent voting of APRM trips and the redundancy of outputs, there may be conditions where the trip voter function is ino3erable, but trip capability for one or more of the other APRi functions through that Trip Voter is still maintained.
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| This may be considered when determining the condition of the other APRM functions resulting from partial inoperability of the trip voter function.
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| There is no Allowable Value for this Function.
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| G V
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| FERMI UNIT 2 B 3.3.1.1 - 11 Amendment No. 134
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| l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| : 3. Reactor Vessel Steam Dome Pressure-Hiah An increase in the RPV pressure during reactor operation l compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However the Reactor Vessel Steam Dome Pressure-High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. The overpressurization protection analysis of Reference 4 conservatively assumes scram on the Average Power Range Monitor Neutron Flux-Upscale signal, not the Reactor Vessel Steam Dome Pressure-High signal. Along with the SRVs, the reactor scram limits the peak RPV pressure to less than the ASME Section III Code limits.
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| High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The r Reactor Vessel Steam Dome Pressure-High Allowable Value is i chosen to provide a sufficient margin to the ASME Section III Code limits during the event.
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| Four channels of Reactor Vessel Steam Dome Pressure-High Function, with two channels in each trip system arranged in a one out of two logic, are required to be OPERABLE to ensure that no single instrument failure will 3reclude a scram from this Function on a valid signal. T1e Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
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| I
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| )
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| O kJ FERMI - UNIT 2 B 3.3.1.1 - 12 Amendment No. 134 i
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| )
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| : 4. Reactor Vessel Water Level-Low. Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level-Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| Reactor Vessel Water Level-Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
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| Four channels of Reactor Vessel Water Level-Low, Level 3 Function, with two channels in each trip system arranged in C] a one-out-of two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
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| The Reactor Vessel Water Level-Low, Level 3 Allowable Value is selected to ensure that the function will perform as predicted in the recirculation line break analysis. It also ensures that during normal operation the separator skirts are not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water-Low Low Low, Level 1 will not be required. The allowable value is referenced to the top of active fuel.
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| The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level-Low Low. Level 2 and Low Low Low. Level 1
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| . provide sufficient protection for level transients in all
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| 'other MODES.
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| /~'T V
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| FERMI - UNIT 2 B 3.3.1.1 - 13 Amendment No. 134
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| e RPS Instrumentation B 3.3.1.1
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| !m BASES i
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| APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| : 5. Main Steam Isolation Valve-Closure MSIV closure results in loss of the main turbine and the l condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve-Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient. However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor )
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| Neutron Flux-Upscale Function, along with the SRVs limits )
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| the peak RPV pressure to less than the ASME Code limits.
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| That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.
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| The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| V MSIV closure signals are initiated from Josition switches located on each of the eight MSIVs. Eac1 MSIV has one position switch that provides the originating sensor for two se)arate channels; one inputs to RPS trip system A while the otler inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve-Closure channels, each channel consisting of one position switch, which is shared with one other channel.
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| The logic for the Main Steam Isolation Valve-Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.
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| The Main Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.
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| Sixteen channels of the Main Steam Isolation Valve-Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate O
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| L.)
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| FERMI UNIT 2 B 3.3.1.1 -14 Amendment No. 134 1
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| J
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| r RPS Instrumentation B 3.3.1.1
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| /3 Q BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) high. a pressurization transient can occur if the MSIVs close. In MODE 2, the MSIV closure trip is automatically bypassed, and the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.
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| : 6. Main Steam Line-Hiah Radiation Main Steam Line-High Radiation Function ensures prompt reactor shutdown upon detection of high radiation in the vicinity of the main steam lines. High radiation in the vicinity of the main steam lines could indicate a gross fuel failure in the core. The scram is initiated to limit the fission product release from the fuel. This Function is not specifically credited in any accident analysis but is being retained for overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
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| Main Steam Line-High Radiation signals are initiated from four radiation monitors. Each monitor senses high gamma radiation in the vicinity of the main steam line. The Main
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| /_ Steam Line-High Radiation Allowable Value is selected high C) enough above background radiation levels to avoid spurious scrams, yet low enough to promptly detect a gross release of fission products from the fuel.
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| Four channels of Main Steam Line-High Radiation Function with two channels in each trip system. arranged in a one-out of two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid signal. This Function is required in MODES 1 and 2 where considerable energy exists such that steam is being produced at a rate which could release considerable fission products from the fuel.
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| The Allowable Value is based on the NRC guidelines of 3.6 times the full oower background radiation level with nominal full power lydrogen injection rate. This Allowable Value remains fixed at this nominal full-power basis even when operating at reduced power and/or reduced hydrogen injection rates.
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| I FERMI - UNIT 2 B 3.3.1.1 - 15 Amendment No. 134 !
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| l
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY (continued)
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| : 7. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure-High Function is assumed to scram the plant coincident with the Reactor Vessel Water Level-Low. Level 3 Function in the analysis of the LOCA inside primary containment. The reactor scram reduces the amount of energy to be absorbed and helps the ECCS ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.
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| Four channels of Drywell Pressure-High Function, with two channels in each trip system arranged in a one out of-two O logic, are required to be OPERABLE to ensure that no single
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| (,) instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS. resulting in the limiting transients and accidents.
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| 8a. 8b. Scram Discharae Volume Water Level-Hiah The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.
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| Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level-High Functions are an input to the RPS logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the UFSAR. However, they are retained to ensure the RPS remains OPERABLE.
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| SDV water level is measured by two diverse methods. The level is measured by four float type level switches and four level transmitters for a total of eight level signals. The outputs of these devices are arranged so that there is a FERMI UNIT 2 B 3.3.1.1 - 16 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) signal from a level switch and a level transmitter to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8.
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| The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram.
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| l Four channels of each ty)e of Scram Discharge Volume Water ;
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| Level-High Function, wit 1 two channels of each type in each !
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| trip system, arranged in a one out of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES 1 l and 2, and in MODE 5 with any control rod withdrawn from a '
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| core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.
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| : 9. Turbine Stoo Valve-Closure O
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| () Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves. The Turbine Stop Valve-Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 7. For this event. the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.
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| Turbine Stop Valve-Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A: the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve-Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve-Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER = 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first O
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| .V FERMI - UNIT 2 B 3.3.1.1 - 17 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) stage pressure of = 161.9 psig; therefore, to consider this function OPERABLE, the turbine bypass valves must remain shut at THERMAL POWER = 30% RTP.
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| The Turbine Stop Valve-Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.
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| Eight channels of Turbine Stop Valve-Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is = 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP since the Reactor Vessel Steam Dome Pressure-High and the Average Power Range Monitor Neutron Flux-Upscale Functions are adequate to maintain the necessary safety margins.
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| : 10. Turbine Control Valve Fast Closure
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| /7 V
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| Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure Function is the primary scram signal for the generator load rejection event analyzed in Reference 7. For this event, the reactor {
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| scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.
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| Turbine Control Valve Fast Closure signals are initiated by the de energization of.the solenoid dump valve at each 3 control valve. Redundant relay signals are provided to each l RPS logic channel such that fast closure of one control valve in each RPS trip system will initiate a scram. This Function must be enabled at THERMAL POWER = 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure of
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| = 161.9 psig; therefore, to corsider this Function OPERABLE, the turbine bypass valves must remain shut at THERMAL POWER
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| = 30% RTP.
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| O FERMI - UNIT 2 B 3.3.1.1 - 18 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES' APPLICABLE SAFETY ANALYSES, LCO,'and APPLICABILITY (continued)
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| There is no Allowable Value for the Turbine Control Valve Fast Closure Function since the channels are actuated solely on de energization of the solenoid dump-valve.
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| Four channels of Turbine Control Valve Fast Closure Function i with two channels in each trip system arranged in a !
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| one out of two logic are required to be OPERABLE to ensure i that no single instrument. failure will preclude a scram from this Function on a valid signal. This Function is required.
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| consistent with the analysis assumptions, whenever THERMAL POWER is = 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel Steam Dome Pressure-41ig' lnd the Average Power Range Monitor Neutron Flux-Upscuie Functions are adequate to maintain the necessary safety margins.
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| : 11. Reactor Mode Switch-Shutdown Position The Reactor Mode Switch-Shutdown Position Function provides signals, via the manual scram logic channels, to each.of the I four RPS logic channels, which are redundant to the O.
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| automatic protective instrumentation channels and provide I manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
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| The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.
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| There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.
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| Four channels of Reactor Mode Switch-Shutdown Position Function, with two channels in each trip system arranged in a one-out of-two logic, are available and required to be OPERABLE. The Reactor Mode Switch-Shutdown Position Function is required to be OPERABLE in MODES 1 and 2. and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
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| I O
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| FERMI.- UNIT 2 B 3.3.1.1 - 19 Amendment No. 134
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| I RPS Instrunentation '
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| E 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) l
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| : 12. Manual Scram )
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| The Manual Scram push button channels provide signals, via ,
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| the manual scram logic channels, to each of the four RPS j logic channels, which are redundant to the automatic q protective instrumentation channels and provide manual 4 reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. I There is one Manual Scram push button channel for each of .
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| the four RPS logic channels. In order to cause a scram it )
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| is necessary that at least one channel in each trip system be actuated.
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| There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.
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| Four channels of Manual Scram with two channels in each trip system arranged in a one out-of-two logic are available and
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| \ required to be OPERABLE in MODES 1 and 2. and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion i Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial '
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| entry into the Condition. However. the Required Actions for inoperable RPS instrumentation channels provide appropriate com)ensatory measures for separate inoperable channels. As suc1, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.
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| O FERMI UNIT 2 B 3.3.1.1- 20 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 I '
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| BASES i
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| ACTIONS-(continued)
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| A.1 and A.2 ,
| |
| Because of the diversity of sensors.available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (Refs. 9 and 13) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the 1 Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the f ino wrable channel cannot be restored to OPERABLE status wit 11n the allowable out of service time, the channel or the associated trip system must be placed in the tripped 1 condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip _(e.g.,
| |
| as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered O and its Required Action taken.
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| As noted, Required Action A.2.is not applicable for APRM Functions 2.a. 2.b. 2.c. and 2.d. Inoperability of one required APRM channel affects both trip systems: thus, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Conaition C, as well as entry into Condition A for each a channel. l 2
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| l O
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| FERMI UNIT 2 B 3.3.1.1 - 21 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued)'
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| B.1 and B.2 Condition B exists when, for any one or more Functions, at 1 least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip 1 capability for that function, but cannot accommodate a {
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| single failure in either trip system. l Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single )
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| failure in both trip systems (e.g., each trip system remains in a one out-of one arrangement for a ty)ical four channel Function). The reduced reliability of t11s logic arrangement was not evaluated in References 9 and 13 for the 1 12 hour Completion Time. Within the 6 hour allowance, the I associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.
| |
| Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9 and 13, which justified a 12 hour allowable out O, of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the ino rable channels in that trip system should be placed in tri (e.g., a trip system with two inoperable channels could e in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is-in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in).
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| If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip.
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| The 6 hour Com)1etion Time is judged acceptable based on the remaining capa)1lity to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.
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| O FERMI UNIT 2 B 3.3.1.1 - 22 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1
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| ( BASES ACTIONS (continued)
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| Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case 1 where placing the inoperable channel or associated trip l system in trip would result in a scram), Condition D must be I entered and its Required Action taken.
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| As noted, Condition B is not applicable for APRM Functions 2.a. 2.b. 2.c, and 2.d. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system, as are the APRM 2 out of-4 voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a. 2.b, 2.c, and 2.d. and these Functions are not associated with specific trip systems as
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| /~
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| are the APRM 2 out of 4 voter and other non APRM channels.
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| C] Condition B does not apply.
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| C.l Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function ;
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| result in the Function not maintaining RPS trip capability. l A Function is considered to be maintaining RPS trip '
| |
| capability when sufficient channels are OPERABLE or in trip l (or the associated trip system is in trip), such that both !
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| trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one out of two taken twice logic and the IRH and APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation Valve-Closure), this would require both tri) systems to have each channel associated with the MSIVs in t1ree main steam lines (not necessarily the same main steam lines for both trip systems) OPERABLE or in trip (or the associated trip system in trip). l (g
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| 3 FERMI - UNIT 2 B 3.3.1.1 - 23 Amendment No. 134
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| o RPS Instrumentation B 3.3.1.1 I BASES ACTIONS (continued)
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| For Function 8 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE.or in trip (or the associated trip system in trip).
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| The Completion Time is intended to allow the operator time to evaluate, and repair or place in trip any discovered ino)erabilities that result in a loss of RPS trip OPE MBILITY. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
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| El Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and M0DE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Com)letion Time has expired. Condition D will be entered for tlat channel and provides for transfer to the appropriate subsequent O Condition.
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| E.1. F.1. G.I. H.1. and H.2 If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time. the plant must be
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| ) laced in a MODE or other specified condition in which the
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| _C0 does not apply. Alternately, for Condition H, the MSLs may be isolated (Required Action H.1), and, if allowed (i.e., plant safety analysis and minimal steam flow in MODE 2 allows operation with the MSLs isolated), operation with the MSLs isolated may' continue. Isolating the MSLs conservatively accomplishes the safety function of the i inoperable channel. The allowed Completion Times are i reasonable, based on operating experience, to reach the :
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| specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)."
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| l FERMI - UNIT 2 B 3.3.1.1 - 24 Amendment No. 134
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| ~
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| l RPS Instrumentation l B 3.3.1.1
| |
| ! BASES
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| . V)
| |
| , ACTIONS (continued) ,
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| I i
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| l L1 1
| |
| If the channel (s) is not restored to OPERABLE status or :
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| placed in trip (or the associated trip system placed in .
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| trip) within the allowed Completion Time, the plant must be I
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| ) laced in a MODE or other specified condition in which the I C0 does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods I in core cells containing no fuel assemblies do not affect j the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable l control rods in core cells containing one or more fuel assemblies are fully inserted.
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| l SURVEILLANCE As noted at the beginning of the SRs. the SRs for each RPS l REQUIREMENTS instrumentation Function are located in the SRs column of ieble 3.3.1.1-1.
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| O The Surveillances are modified by a Note to indicate that
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| () '
| |
| when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to i 6 hours, provided the associated Function maintains RPS trip i capability. For the case of the APRM Functions 2.a. 2.b.
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| 2.c. and 2.d. RPS trip capability is maintained with any two OPERABLE APRMs remaining. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
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| This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to aerform channel l l . Surveillance. That analysis demonstrated tlat the 6 hour testing allowance does not significantly reduce the 4
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| l probability that the RPS will trip when necessary.
| |
| SR 3.3.1.1.1 and SR 3.3.1.1.2 1
| |
| Performance of the CHANNEL CHECK once every 12 hours or once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring
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| , O
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| , V
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| ' FERMI - UNIT 2 B 3.3.1.1 - 25 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) the same parameter should read approximately the same value.
| |
| Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
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| The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.1.1.3 (D To ensure that the APRMs are accurately indicating the true
| |
| 'O core average power, the APRMs are calibrated to the reactor i power calculated from a heat balance when = 25% RTP. The Frequency of once per 7 days is based on minor changes in ,
| |
| LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. i A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at = 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR, LHGR, and APLHGR). At
| |
| = 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met oer SR 3.0.2. In this event, the SR must be performed witlin 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
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| r~
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| (_)T FERMI UNIT 2 B 3.3.1.1- 26 Amendment No. 134
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| RPS Instrumentation B 3.3.1.1 n 1 BASES
| |
| (]
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required l contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL !
| |
| FUNCTIONAL TEST of a relay. This is acceptable because all I of the other required contacts of the relay are verified by I other Technical Specifications and non Technical l Specifications tests at least once per refueling interval with applicable extensions.
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| Any setpoint adjustment shall be consistent with the j assumptions of the current plant specific setpoint '
| |
| methodology.
| |
| As noted. SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1. since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without J O utilizing jumpers, lifted leads, or movable links. This d allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve ,
| |
| hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
| |
| A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).
| |
| SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required i channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non Technical Specifications tests at least once per refueling interval l with applicable extensions. In accordance with Reference 9.
| |
| the scram contactors must be tested as part of the Manual l
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| [
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| (
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| FERMI - UNIT 2 B 3.3.1.1 - 27 Amendment No. 134 L.
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| b RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREENTS (continued)
| |
| ' Scram Function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on References 9 and 10. (The Manual Scram Function's CMANNEL FUNCTIONAL TEST Frequency was credited in the Reference 9 analysis to extend many automatic scram Functions
| |
| * Frequencies.)
| |
| SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.
| |
| The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to fully withdrawing SRMs from the core since indication is being transitioned from the SRMs to the IRMs.
| |
| The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the i
| |
| s system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap i between IRMs and APRMs exists when sufficient IRMs and APRMs l concurrently have onscale readings such that the transition between MODE I and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs si:nilarly exists when, prior to fully withdrawing the SRMs from the core, IRMs are above the downscale rod block and show increasing flux on range 1 before SRMs have reached 1/2 decade below the upscale rod block.
| |
| As noted. SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs. maintaining overlap is not required (APRMs may be i reading downscale once in MODE 2). I If o/erlap for a group of channels is not demonstrated !
| |
| (e.g.. IRM/APRM overlap), the reason for the failure of the i Surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition ,
| |
| should be declared inoperable.
| |
| FERMI . UNIT 2 B 3.3.1.1 - 28 Amendment No. 134 L q
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| | |
| RPS Instrumentation B 3.3.1.1
| |
| ( BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs.
| |
| SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)
| |
| System. This establishes the relative local flux profile for approariate re3resentative input to the APRM System.
| |
| The 1000 4fD/T ("slort" ton) Frequency is based on operating experience with LPRM sensitivity changes.
| |
| SR 3.3.1.1.,3 and SR 3.3.1.1.13 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all Q of the other required contacts of the relay are verified by V other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9.
| |
| 1 The 18 month Frequency is based on the need to perform this J Surveillance under the conditions that apply during a plant outage and the potential for an un)lanned transient if the Surveillance were performed with tie reactor at power.
| |
| Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.
| |
| l l
| |
| FERMI - UNIT 2 B 3.3.1.1 - 29 Amendment No. 134
| |
| | |
| RPS Instrumentation B 3.3.1.1
| |
| ) BASES SURVEIll.ANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.1.10 This Surveillance provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.11. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology. but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions. the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| The Frequency of 92 days is based on the reliability analysis of Reference 9.
| |
| SR 3.3.1.1.11 and SR 3.3.1.1.14 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel Q
| |
| v responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive l calibrations consistent with the plant specific setpoint methodology.
| |
| SR 3.3.1.1.11 Note 1 states that neutron detectors are i excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift. and because of the difficulty ;
| |
| of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2 and the 1000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8).
| |
| SR 3.3.1.1.11 Note 2 is provided that requires the IRM SR to be performed within 12 hours of entering MODE 2 from MODE 1.
| |
| Testing of the MODE 2 IRM Function cannot be performed in MODE 1 without utilizing jumpers lifted leads or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
| |
| b)
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| FERMI - UNIT 2 B 3,3.1.1 - 30 Amendment No. 134
| |
| | |
| RPS Instrumentation B 3.3.1.1 TO V
| |
| BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| The Frequency of SR 3.3.1.1.11 is based upon a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.1.1.14 is based upon = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| SR. _;l . 3.1.1.12 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will Derform the intended function. For the APRM Functions thi's test supalements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including for Function 2.b only, the recirculation flow input function excluding the flow transmitter), the 2 out of 4 voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shrill be consistent with the assumptions of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.12 is based on the reliability analysis of A)
| |
| ( ,
| |
| Reference 13. (NOTE: The actual voting logic of the 2-out-of 4 voter channels is tested as part of SR 3.3.1.1.19.)
| |
| For Function 2.a. a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or l{fted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
| |
| SR 3.3.1.1.15 and SR 3.3.1.1.19 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific ;
| |
| channel. The functional testing of control rods (LC0 3.1.3), and SDV vent and drain valves (LC0 3.1.8).
| |
| overlaps this Surveillance to provide complete testing of the assumed safety function. For the 2-out-of-4 Voter j Function, the LSFT includes simulating APRM trip conditions :
| |
| at the APRM channel inputs to the 2-out-of 4 trip voter !
| |
| channel to check all combinations of two tripped inputs to the 2-out of 4 trip voter logic in the voter channels, r~ t k
| |
| FERMI UNIT 2 B 3.3.1.1 - 31 Amendment No. 134
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| | |
| RPS Instrumentation 1 B 3.3.1.1 BASES i
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| The 18 month Frequency of SR 3.3.1.1.15 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.
| |
| Additionally, the 24 month Frequency of SR 3.3.1.1.19 is )
| |
| based on Reference 13. I SR 3.3.1.1.16 ,
| |
| 1 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure Functions will not be inadvertently bypassed when THERMAL POWER is = 30* RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
| |
| Additionally, consideration is given to the fact that main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine
| |
| ( first stage pressure: where turbine first stage 161.9 psig conservatively correlates to 30% RTP) pressure
| |
| . the main of turbine bypass valves must remain closed at THERMAL POWER
| |
| = 30% RTP to ensure that the calibration remains valid.
| |
| If any bypass channel's setpoint is nonconservative (i.e., !
| |
| the Functions are bypassed at = 30% RTP. either due to open main turbine bypass valve (s)-or other reasons), then the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure Functions are considered inoperable.
| |
| Alternatively, the bypass channel can be placed in the i conservative condition (nonbypass). If placed in the i nonbypass condition, this SR is met and the channel is !
| |
| considered OPERABLE.
| |
| The Frequency of 18 months is based on engineering judgment, reliability of the components, and = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| O FERMI - UNIT 2 B 3.3.1.1-32 Amendment No. 134
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| | |
| L .
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| RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| -SR 3.3.1.1.17 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10. RPS RESPONSE TIME for the APRM 2-out-of-4 Voter Function includes the output relays of the voter and the associated RPS relays and contactors. (The digital portion of the APRM and 2 out-of 4 voter channels are excluded from the RPS RESPONSE TIME testing because self-testing and calibration I checks the time base of_ the digital electronics.)
| |
| Confirmation of the time base is adequate to assure required response times are met.
| |
| As noted. neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.
| |
| In addition. Note 2 states the response time of the sensors O for Functions 3 and 4 are excluded from RPS RESPONSE TIME testing. The sensors for these Functions are assumed to operate at the sensor's design response time. This allowance is supported by Reference 12, which determined that significant degradation of the sensor channel response time can be detected during )erformance of other Technical Specification SRs and that t1e sensor response time is a small part of the overall RPS RESPONSE TIME testing.
| |
| RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip
| |
| . system. in lieu of the 8 channels specified in Table 3.3.1.1 1 for the MSIV Closure Function. This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. The 18 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time
| |
| ' degradation, but not channel failure. are infrequent occurrences.
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| O FERMI - UNIT 2 B 3.3.1.1 -33 Amendment No. 134
| |
| | |
| cv RPS Instrumentation B 3.3.1.1 i BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.1.18 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and acctracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. For the APRM Simulated Thermal Power - Upscale Function, this SR also includes calibrating the associated recirculation loop flow channel. j SR 3.3.1.1.18 is modified by a Note that states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the i difficulty of simulating a meaningful signal. Changes in l neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR'3.3.1.1.3) J and the 1000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8).
| |
| The Frequency of SR 3.3.1.1.18 is based upon 24 month (V3 calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. l l
| |
| 1 REFERENCES 1. UFSAR, Figure 7.2-2. i
| |
| : 2. UFSAR, Section 15.4.1.2.
| |
| : 3. NEDO 23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
| |
| : 4. UFSAR, Section 5.2.2.3.
| |
| : 5. UFSAR, Section 15.4.9.
| |
| : 6. UFSAR, Section 6.3.3.
| |
| : 7. UFSAR, Chapter 15.
| |
| : 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.
| |
| . FERMI - UNIT 2 B 3.3.1.1 - 34 Amendment No. 134
| |
| | |
| RPS Instrumentation B 3.3.1.1 l
| |
| -BASES REFERENCES (continued)
| |
| : 9. NED0 30851-P-A , " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
| |
| March 1988.
| |
| : 10. UFSAR, Table 7.2-4.
| |
| : 11. NEDC-31336, " Class III, October 1986 General Electric Instrument Setpoint Methodology."
| |
| : 12. NED0 32291, " System Analyses for Elimination of Selected Response Time Testing Requirements," January 1994: and Fermi-2 SER for Amendment 111, dated April 18, 1997.
| |
| : 13. NEDC-32410P A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
| |
| Retrofit Plus Option III Stability Trip Function,"
| |
| October 1995, and Supplement 1. May 1996.
| |
| l O I
| |
| )
| |
| l l
| |
| O FERMI - UNIT 2 B 3.3.1.1 -35 Amendment No. 134
| |
| | |
| +
| |
| i SRM Instrumentation .
| |
| B 3.3.1.2 B.3.3 ' INSTRUMENTATION
| |
| .B 3.3.1.2 Source Range Monitor (SRM) Instrumentation-BASES BACKGROUND The SRMs provide the operator with information relative to the neutron flux level at very low flux levels in the core.
| |
| As such, the SRM indication is used by the operator to monitor the approach to criticality and determine when criticality is achieved. The SRMs are maintained fully ,
| |
| inserted until the count rate is greater than a minimum allowed count rate (a control rod block is set at this q condition). After SRM to intermediate range monitor (IRM) 1 overlap.is demonstrated (as required by SR 3.3.1.1.6), the SRMs are normally fully withdrawn from the core.
| |
| The SRM subsystem of the Neutron Monitoring System (NMS) consists of four channels. Each of the SRM channels can be bypassed, but only one at any given time, by the operation of a bypass switch. Each channel includes one detector that can be physically positioned in the core. Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning equipment, and O electronics associated with the various SRM functions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate. Each channel also includes indication, alarm, and control rod blocks. 'However, this i LC0 specifies OPERABILITY requirements'only for the I monitoring and indication functions of the SRMs.
| |
| During refueling, shutdown, and low power operations, the ;
| |
| primary indication of- neutron flux levels is provided by the SRMs_ or special movable detectors connected to the normal SRM circuits. The SRMs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of unexpected subcritical multiplication that could be indicative of an approach to criticality.
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| O FERMI - UNIT 2- B_ 3.3.1.2 - 1 Amendment No. 134
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| | |
| ! l SRM Instrumentation B 3.3.1.2
| |
| ! BASES APPLICABLE Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES during refueling and low power operation is provided by LC0 3.9.1. " Refueling Equipment Interlocks": LC0 3.1.1.
| |
| "SHU1DOWN MARGIN (SDM)"; LC0 3.3.1.1. " Reactor Protection System (RPS) Instrumentation": IRM Neutron Flux-High and Average Power Range Monitor (APRM) Neutron Flux-High.
| |
| Setdown Functions: and LC0 3.3.2.1, " Control Rod Block Instrumentation."
| |
| The SRMs have no safety function and are not assumed to function during any UFSAR design basis accident or transient analysis. However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical Specifications.
| |
| LC0 During startup in MODE 2. three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux O level and reactor period until the flux level is sufficient V to maintain the IRM on Range 3 or above. All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the core.
| |
| In MODES 3 and 4. with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the core.
| |
| In MODE 5. during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2 1. footnote (b).
| |
| requirement that the bundles being spiral reloaded or spiral ,
| |
| offloaded are all in a single fueled region containing at I least one OPERABLE SRM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence).
| |
| FERMI - UNIT 2 B 3.3.1.2 - 2 Amendment No. 134 l
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| l
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES LC0 (continued)
| |
| In nonspiral routine operations two SRMs are required to be OPERABLE to provide monitoring of reactivity changes ;
| |
| occurring in the reactor core. Because of the local nature of reactivity changes during refueling, adequate coverage is provided by requiring one SRM to be OPERABLE in the quadrant of the reactor core where CORE ALTERATIONS are being performed, and the other SRM to be OPERABLE in an adjacent quadrant containing' fuel. These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS. i Special movable detectors, according to footnote (c) of Table 3.3.1.2 1. may be used during CORE ALTERATIONS in place of the normal SRM nuclear detectors. These special detectors must be connected to the normal SRM circuits in the NMS such that the applicable neutron flux indication
| |
| , can be generated. These special detectors provide more flexibility in monitoring reactivity changes during fuel loading since they can be positioned anywhere within the core during refueling. They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for SRMs.
| |
| For an SRM channel to be considered OPERABLE. it must be providing neutron flux monitoring indication.
| |
| APPLICABILITY The SRMs are required to be OPERABLE in MODES 2, 3. 4. and 5 prior to the IRMs being on scale on Range 3 to provide for neutron monitoring. In MODE 1. the APRMs provide adequate {
| |
| monitoring of reactivity changes in the core: therefore, the SRMs are not required. In MODE 2, with IRMs on Range 3 or above. the IRMs provide adequate monitoring and the SRMs are not required.
| |
| 4 ACTIONS A.1 and B.1 In MODE 2, with the IRMs on Range 2 or below. SRMs provide I the means of monitoring core reactivity and criticality. I With any number of the required SRMs inoperable, the ability to monitor neutron flux is degraded. Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.
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| O FERMI UNIT 2 B 3.3.1.2 - 3 Amendment No. 134
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES ACTIONS (continued)
| |
| Provided at least one required SRM remains OPERABLE, Required Action A.1 allows 4 hours to restore the required SRMs to OPERABLE status. This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is sufficient time to take corrective actions to restore the required SRMs to OPERABLE status or to establish alternate IRM monitoring capability. During this time, control rod withdrawal and power increase is not precluded by this Required Action. Having the ability to monitor the core with at least one required SRM, proceeding to IRM Range 3 or greater (with overlap required by SR 3.3.1.1.6),
| |
| and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and 1 allowing continued operation. I With three required SRMs inoperable Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes. Required Action A.1 still applies and allows 4 hours to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair. rather than to immediately shut down, with no SRMs OPERABLE.
| |
| C.d In MODE 2, if the required number of SRMs is not restored to OPERABLE status within the allowed Completion Time, the reactor shall be ) laced in MODE 3. With all control rods fully inserted, tie core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
| |
| D.1 and D.2 With one or more required SRMs inoperable in MODE 3 or 4 the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available. Placing the reactor mode switch in the shutdown O .
| |
| FERMI - UNIT 2 B 3.3.1.2 - 4 Amendment No. 134
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES-ACTIONS (continued) position prevents subsequent control rod withdrawal by maintaining a control rod block. The allowed Completion Time of 1 hour is sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the SRM occurring during this interval.
| |
| E.1 and E.2 With one or more required SRMs inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded. - CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes. fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
| |
| Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
| |
| l SURVEILLANCE The SRs for each SRM Applicable MODE or other specified
| |
| . REQUIREMENTS conditions are found in the SRs column of Table 3.3.1.2-1.
| |
| SR 3.3.1.2.1 and SR 3.3.1.2.3 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It i
| |
| is based on the assumption that instrument channels i monitoring the same parameter should read approximately the i same value. Significant deviations between the instrument l channels could be an indication of excessive instrument drift in one of the channels or something even more serious.
| |
| A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| . O !
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| FERMI UNIT 2 ' B 3.3.1.2 - 5 Amendment No. 134
| |
| | |
| l SRM Instrumentation B 3.3.1.2
| |
| , BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| Agreement criteria are determined by the plant staff based L on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the
| |
| : instrument has drifted outside its limit.
| |
| l l The Frequency of once every 12 hours for SR 3.3.1.2.1 is l based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4. reactivity changes j i are not expected: therefore, the 12 hour Frecuency is '
| |
| relaxed to 24 hours for SR 3.3.1.2.3. The CFANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.1.2.2 To provide adequate coverage of potential reactivity changes
| |
| 'in the core, one SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed. and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to O be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact. OPERABLE.
| |
| In the event that only one SRM is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM.
| |
| The 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities that include steps to ensure that the SRMs required by the LCO are in the proper quadrant.
| |
| FERMI - UNIT 2 B 3.3.1.2 - 6 Amendment No. 134 l
| |
| | |
| i SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient irradiated fuel assemblies, to establish the minimum count rate.
| |
| l To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated l core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical.
| |
| A The Frequency is based upon channel redundancy and other V information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours.
| |
| SR 3.3.1.2.5 and SR 3.3.1.2.6 l Performance of a CHANNEL FUNCTIONAL TEST demonstrates the ;
| |
| associated channel will function properly. A successful i test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. SR 3.3.1.2.5 is required in MODE 5 and the 7 day Frequency ensures that the channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.
| |
| /
| |
| t
| |
| (
| |
| FERMI UNIT 2 B 3.3.1.2 - 7 Amendment No. 134
| |
| | |
| SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place in MODES 3 and 4, and since core
| |
| )
| |
| reactivity changes in MODE 2 are typically due to control >
| |
| rod movement, the Frequency has been extended from 7 days to 31 days. The 31 day Frequency is based on operating ex arience and on other Surveillances (such as CHANNEL CHICK) that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.
| |
| Verification of the signal-to noise ratio also ensures that I the detectors are inserted to an acceptable operating level. I In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector.
| |
| Any count rate obtained while the detectors are fully withdrawn is assumed to be " noise" only. The Note to SR 3.3.1.2.5 and Note 1 to' SR 3.3.1.2.6 modify this requirement to not require _ the signal-to noise ratio to be determined when the associated SRM count rate is = 3.0 cps.
| |
| This is acceptable since there is no limitation on signal-
| |
| .to noise ratio when the SRM is'= 3.0 cps.
| |
| The Note 2 to SR 3.3.1.2.6 allows the Surveillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the desire not to perform the Surveillance while at higher power levels.
| |
| Although the Surveillance could be performed while on IRM ,
| |
| Range 3, the plant would not be expected to maintain steady ;
| |
| state operation at this power-level. In this event, the l 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
| |
| 1 O
| |
| FERMI UNIT 2 B 3.3.1.2 - 8 Amendment No. 134
| |
| | |
| i SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.1.2.7 Performance of a CHANNEL CALIBRATION at a Frequency of 18 months verifies the performance of the SRM associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The. neutron detectors are excluded from the CHANNEL CALIBRATION because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful li fe.
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| Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. If not performed within the previous 18 months (plus 25% allowed by SR 3.0.2). the SR must be performed within 12 hours of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability-with the 18 month Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the A)plicability and the desire not to perform the Surveillance w111e at higher power levels. Although the Surveillance could be performed while on IRM Range 3. the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
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| REFERENCES None.
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| l l
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| l l
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| O FERMI - UNIT 2 B 3.3.1.2 - 9 Amendment No. 134 '
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| Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES.
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| BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod-blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.
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| The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds-a predetermined setpoint during control rod manipulations. It is assumed to function O to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the preset power level. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM ;
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| channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit.
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| The RBM channel signal is generated by averaging a set of "ncal power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four average power range monitor (APRM) channels supplies a reference signal for one of the !
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| RBK channels and a signal from another of the APRM channels supplies the reference signal to the second RBM channel.
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| This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the preset power level, the RBM is automatically by)assed. The RBM is also automatically bypassed if a peripleral control rod is selected (Ref.1). ;
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| l O
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| FERMI. UNIT 2 B 3.3.2.1 - 1 Amendment No. 134
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| l Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND (continued)
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| A rod block signal is also generated if an RBM downscale trip or an inoperable trip occurs, since this could indicate a malfunctioning RBM channel. The downscale trip will occur if the RBM channel signal decreases below the downscale trip set point after the RBM channel signal has been normalized.
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| The inoperable trip will occur if too few LPRM inputs are available, if a module is not plugged in, or the function switch is in any position other than " operate."
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| .The purpose of the RWM is to control rod patterns during startup and shutdown at low power conditions, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods 1 inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. -The prescribed control rod sequence is stored in the RWM,.which will initiate control rod withdrawal and insert i blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based on position indication for each control rod.
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| The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the low power setpoint at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits.
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| With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during N00E 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a I control rod block to all control rods.
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| APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, i LCO. and The RBM is designed to prevent violation of the MCPR !
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| APPLICABILITY. SL'and the cladding it plastic strain fuel design limit that i may result from a single control rod withdrawal error (RWE) )
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| event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A O
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| FERMI - UNIT 2 B 3.3.2.1 - 2 Amendment No. 134
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| F l Control Rod Block Instrumentation )
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| B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) statistical analysis of RWE events was performed to determine the RBM response for both channels for each event.
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| From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level, and are specified in the Core Operating Limits Report )
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| : (COLR). Based on the s {
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| limits are established,pecified Allowable Values, operating j The RBM Function satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| Two channels of the RBM are required to be OPERABLE, with )
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| their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block from this ,
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| Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.
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| Nominal trip set]oints are specified in the setpoint calculations. T1e nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values (n) between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
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| Trip setpoints are those predetermined values of output at which an action should tace place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured out)ut value of the process parameter exceeds the setpoint, t1e associated device (e.g.,
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| trip unit) changes state. The analytical limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g. , drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
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| The RBM is assumed to mitigate the consequences of an RWE event when operating = 30t RTP. Below this power level, the
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| } consequences of an RWE event will not exceed the MCPR SL 13 V
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| FERMI - UNIT 2 B 3.3.2.1 -3 Amendment No. 134 l
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| Control Rod Block Instrumentation B 3.3.2.1 I
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| BASES
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| {
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| APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) I and, therefore, the RBM is not required to be OPERABLE l (Ref. 3).
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| : 2. Rod Worth Minimizer The RWM enforces the prescribed withdrawal sequence to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used I in evaluating the CRDA are summarized in References 4, 5, 6, and 7. The prescribed withdrawal sequence requires that '
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| control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the prescribed withdrawal sequence are specified in LC0 3.1.6. " Rod Pattern Control."
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| The RWM Function satisfies Criterion 3 of 10 CFR i 50.36(c)(2)(ii).
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| Since the RWM is designed to act as a backup to operator 1 control of the rod sequences, only one channel of the RWM is I
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| /~ available and required to be OPERABLE (Ref. 7). Special
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| ( circumstances provided for in the Required Action of LC0 3.1.3 " Control Rod OPERABILITY," and LC0 3.1.6 may i necessitate bypassing the RUM to allow continued operation I with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the prescribed withdrawal sequence. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LC0 followed.
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| Compliance with the prescribed withdrawal sequence, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is s 10t RTP. When THERMAL POWER is > 10t RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 7).
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| In MODES 3 and 4, all control rods are required to be inserted into the core: therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subtritical.
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| O FERMI - UNIT 2 B 3.3.2.1 - 4 Amendment No. 134
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| Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| : 3. Reactor Mode Switch-Shutdown Position During MODES 3 and 4 and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch-Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
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| The Reactor Mode Switch-Shutdown Position Function 4 satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). l Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.
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| During shutdown conditions (MODE 3, 4, or 5), no positive 3 reactivity insertion events are analyzed because assumptions
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| '(V . are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod j withdrawal block is required to be OPERABLE. During MODE 5 1 with the reactor mode switch in the refueling position the !
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| refuel position one-rod out interlock (LC0 3.9.2) provides the required control rod withdrawal blocks.
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| ACTIONS A_J With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function: however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours is based on the low probability of an event occurring coincident with a failure in the remaining 0PERABLE channel.
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| O FERMI - UNIT 2 B 3.3.2.1 - 5 Amendment No. 134
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| Control Rod Block Instrumentation B 3.3.2.1
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| ~ BASES
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| ' ACTIONS-(continued)
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| IL1 If Required Action A.1 is not met and the associated Completion Time has expired, an RBM channel must be placed in tri) within 1 hour. If both RBM channels are inoperable, the RB1 is not capable of performing its intended function:
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| -thus, one channel must also be placed in trip. This
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| . initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
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| The 1 hour Completion Time is intended to allow the operator time to evaluate ~and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
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| C.1. C.2.1.1. C.2.1.2. and C.2.2 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod
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| 'O movement must be immediately suspended except by scram.
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| Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last 12 months.
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| Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double i check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff.
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| 'The RWM may be bypassed under these conditions to allow continued operations. In addition. Required Actions of LC0 3.1.3 and LC0 3.1.6 may require bypassing the RWM.
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| -during which time the RWM must be considered inoperable, and if during a reactor startup. would require Condition C be entered and its Required Actions taken.
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| O FERMI -UNIT 2 B 3.3.2.1 - 6. Amendment No. 134 I l
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| /
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| J
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| Control Rod Block Instrumentation B 3.3.2.1 l BASES ACTIONS (continued)
| |
| D_1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor .
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| Operator) or other qualified member of the technical staff. {
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| The RWM may be bypassed under these conditions to allow the I reactor shutdown to continue. )
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| )
| |
| E.1 and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are '
| |
| consistent with the normal rod block action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e.,
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| maintaining all control rods inserted), there is no
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| (] distinction between having one or two channels inoperable.
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| In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing I one or more fuel assemblies will ensure that the core is !
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| subcritical with adequate SDM ensured by LC0 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not ,
| |
| required to be inserted. Action must continue until all 1 insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
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| SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1 1.
| |
| The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely
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| 'for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the O
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| FERMI UNIT 2 B 3.3.2.1 - 7 Amendment No. 134
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| Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
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| This Note is based on the reliability analysis (Ref. 8) assumption of-the average time required to )erform channel Surveillance. That analysis demonstrated t1at the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.
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| SR 3.3.2.1.1 and SR 3.3.2.1.2 A CHANNEL FUNCTIONAL TEST ~is performed for the RWM to ensure that the entire system will perform the intended function.
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| A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is'an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non Technical Specifications tests at least once Jer -
| |
| refueling interval with applicable extensions. T1e O SR 3.3.2.1.1 CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a ,
| |
| selection error is indicated and a control rod block occurs, i Control rod withdrawal sequences are normally established consistent with the rules of the generic BPWS analysis.
| |
| Occasionally, operational limitations (e.g. power suppression of failed fuel) may dictate the insertion of
| |
| . control rods which do not meet the minimum cell separation criteria of the generic BPWS analysis. In such situations, sufficient cycle s)ecific analyses are performed to l demonstrate that t1e resulting control rod worths of the modified control rod withdrawal sequence are bounded by the I rod worths allowed by rigorously following the rules of the generic BPWS-analysis, thereby assuring that the 280 cal /gm ;
| |
| fuel damage limit will not be violated during a'CRDA.
| |
| The " prescribed withdrawal sequence" is defined as the combination of both the procedurally specified control rod movement sequence and any analytically allowed deviations
| |
| ~from this sequence. Some prescribed withdrawal sequences (e.g. BPWS) have more flexibility in allowed deviations than other prescribed withdrawal sequences (e.g., a cycle-A V
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| FERMI UNIT 2 B 3.3.2.1 - 8 Amendment No. 134
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| 1 . Control Rod Block Instrumentation B 3.3.2.1 BASES-SURVEILLANCE REQUIREMENTS (continued) specific sequence developed for power suppression of failed fuel may not allow any deviations).
| |
| SR 3.3.2.1.1 is performed during startup. As noted in the SRs, SR 3.3.2.1.1 is not required to be performed until I hour after any c3ntrol rod is withdrawn at s 10% RTP in
| |
| -MODE 2. The SR 3.3.2.1.2 CHANNEL FUNCTIONAL TEST is I performed by attempting to insert and withdraw a control rod not in compliance with the prescribed sequence and verifying a selection error is indicated and a control rod insert and withdraw block (respectively) occur. SR 3.3.2.1.2-is performed during a plant shutdown when transitioning to s 10% RTP. As noted, SR 3.3.2.1.2 is not required to be performed until 1-hour after THERMAL POWER is s 101 RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.1, and .
| |
| THERMAL POWER reduction to 5 10% RTP when in MODE 1 for SR 3.3.2.1.2. to perform the required Surveillance if the 1 92 day Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in.which to complete the SRs. The Frequencies are based on operating experience that shows the RWM usually passes the
| |
| , O Surveillance when performed at these Frequencies.
| |
| SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for each RBH channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical l Specifications and non Technical Specifications tests at least once extensions.per refueling interval with applicable Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 184 days is based on reliability analyses (Ref. 9).
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| k FERMI UNIT 2 B 3.3.2.1 - 9 Amendment No. 134
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| I-Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.2.1.4 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire t channel will perform the intended function. The CHANNEL
| |
| ! FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown t
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| Position Function is performed by attem) ting to withdraw any control rod with the reactor mode switc1 in the shutdown position and verifying a control rod block is present.
| |
| As noted in the SR. the Surveillance is not required to be performed until I hour after the reactor mode switch is in the shutdown position. since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads. or movable links. This allows entry into MODES 3 and 4 if the 18 month Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.
| |
| The 18 month Frequency is based on the need to perform this O Surveillance under the conditions that apply during a plant V. outage and the potential for an unalanned transient if the Surveillance were performed with t1e reactor at power.
| |
| 0)erating experience has shown these components usually pass tle Surveillance when performed at the 18 month Frequency.
| |
| SR 3.3.2.1.5 The power at which the RBM is automatically by)assed is based on the APRM signal's in)ut to each RBM clannel. Below the minimum power setpoint. t1e RBM is automatically bypassed. This power Allowable Value must be verified periodically to be less than 30t RTP. If this setpoint is nonconservative. then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be '
| |
| placed in the conservative condition (i.e.. enabling the RBM Function). If placed in this condition, the SR is met and j the RBM channel is not considered inoperable. The 24 month {
| |
| Frequency is based on the actual trip setpoint methodology utilized for these channels.
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| O FERMI UNIT 2 B 3.3.2.1 - 10 Amendment No. 134
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| | |
| Control Rod Block Instrumentation B 3.3.2.1 A
| |
| V BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.2.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a ,
| |
| meaningful signal. Neutron detectors are adequately surveilled in SR 3.3.1.1.1 and SR 3.3.1.1.7.
| |
| The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| SR 3.3.2.1.Z The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer.
| |
| This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. Control rod withdrawal sequences are normally established consistent with the rules of the generic BPWS analysis. Occasionally, operational limitations (e.g., power suppression of failed fuel) may dictate the insertion of control rods which do not meet the minimum cell separation criteria of the generic BPWS analysis. In such situations, sufficient cycle specific analyses are performed to demonstrate that the resulting control rod worths of the modified control rod withdrawal sequence are bounded by the rod worths allowed by rigorously following the rules of the generic BPWS analysis, thereby assuring that the 280 cal /gm fuel damage limit will i not be violated during a CRDA. I The " prescribed withdrawal sequence" is defined as the combination of both the procedurally specified control rod movement sequence and any analytically allowed deviations from this sequence. Some prescribed withdrawal sequences (e.g., BPWS) have more flexibility in allowed deviations than other prescribed withdrawal sequences (e.g., a cycle-FERMI UNIT 2 B 3.3.2.1- 11 Amendment No. 134 I
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| i
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| | |
| Control Rod Block Instrumentation B 3.3.2.1 BASES-SURVEILLANCE REQUIREMENTS (continued) specific sequence developed for power suppression of failed fuel may not allow any deviations).
| |
| The Surveillance is performed once )rior to declaring the RWM OPERABLE following loading of tie )rescribed withdrawal sequence into the RWM, since this is w1en rod sequence input errors are possible.
| |
| REFERENCES 1. UFSAR Section 7.6.2.13.5.
| |
| : 2. UFSAR, Section 7.6.1.20.
| |
| : 3. General Electric Energy, " Maximum Extended Operating Domain Analysis for Detroit Edison Company Enrico Fermi Energy Center Unit 2," NEDC - 31843P, July 1990.
| |
| : 4. NEDE-24011 P-A 10 US, " General Electric Standard Application for Reload Fuel," Supplement for United States, March 1991.
| |
| : 5. " Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.
| |
| : 6. NED0 21231, " Banked Position Withdrawal Sequence "
| |
| January 1977.
| |
| : 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011 P A." " General Electric Standard Application for Reactor Fuel, Revision 8.
| |
| Amendment 17," December 27, 1987.
| |
| : 8. NEDC-30851-P A. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
| |
| October 1988.
| |
| : 9. NEDC-32410P A, " Nuclear Measurement Analysis and
| |
| ' Control Power Range Neutron Monitor (NUMAC PRNM)
| |
| Retrofit Plus Option III Stability Trip Function "
| |
| October 1995, and Supplement 1. May 19%.
| |
| O FERMI UNIT 2 B 3.3.2.1 - 12 Amendment No. 134
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| | |
| p ,
| |
| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 8 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES-BACKGROUND The feedwater and main turbine high water level trip instrument: tion is designed.to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.
| |
| With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level. Level 8 reference point, causing the trip of the two feedwater pump turbines and the main turbine.
| |
| Reactor Vessel Water Level-High Level 8 signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). Four channels of Reactor Vessel Water Level-High. Level 8 instrumentation are provided as input to a one-out-of-two taken twice initiation logic that trips the two feedwater pump turbines and the main turbine. One channel isolates the non-safety related Standby Feedwater (SBFW) pumps as well. The SBFW isolation O portion of the channels is not subject to Technical Specifications since the SBFW system is not required for safety. The channels include electronic equipment (e.g.,
| |
| trip units) that compares measured input signals with pre established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a main feedwater, standby feedwater, and turbine trip signal to the trip logic.
| |
| A trip of the feedwater pump turbines limits further l increase in reactor vessel water level by limiting further '
| |
| addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.
| |
| O FERMI UNIT 2 B 3.3.2.2 - 1 Amendment No. 134 ;
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| )
| |
| | |
| l 1
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| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES APPLICABLE The feedwater and main turbine high water 'evel trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a
| |
| , turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref.1).
| |
| The Level 8 tri) indirectly initiates a reactor scram (above 30% RTP) from t1e main turbine trip and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR. Credit for this function is also taken in the analysis of the pressure regulator failure event (Ref.1), and the decrease in reactor coolant system flow rate events described in Reference 2.
| |
| Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| LC0 The LC0 requires four channels of the Reactor Vessel Water Level-High, Level 8 instrumentation to be OPERABLE to ensure that no single instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Level 8 signal. One out of two channels in each of the two O trip systems are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event.
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| The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
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| Trip setpoints are those aredetermined values of output at which an action should tace place. The setpoints are compared to the actual process )arameter (e.g., reactor vessel water level), and when t1e measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process i parameters obtained from the safety analysis. The Allowable !
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| Values are derived from the analytic limits corrected for calibration, process, and some of the instrument errors. A i
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| O FERMI - UNIT 2 B 3.3.2.2-2 Amendment No. 134
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| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES LC0 (continued) channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). - The trip setpoints derived in this manner provide adequate protection Decause instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
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| APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at = 25% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding it plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.1, " Average Planar Linear Heat Generation Rate (APLHGR)," and LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 25t RTP: therefore, these requirements are only necessary when operating at or above this power
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| ) level. 3 ACTIONS A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures 'for separate inoperable channels._ As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip. instrumentation channel. ,
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| O FERMI UNIT 2 83.3.2.2-3 Amendment No. 134
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| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 l
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| O 8^ses ACTIONS (continued)'
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| Ad With one or.more channels inoperable, and feedwater and main turbine high water level trip capability maintained, the remaining OPERABLE channels can provide the required trip signal. However, overall instrumentation reliability is reduced because a single failure in one of the remaining channels concurrent with feedwater controller failure, maximum demand event, may result in the instrumentation not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time with one or more channels inoperable. If the inoperable channel (s) cannot be restored to OPERABLE status within the Completion Time. the channel (s) must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel (s) in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel (s) in trip (e.g., as in the case where placing the inoperable channel (s) in trip A would result in a feedwater or main turbine trip).
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| Condition C must be entered and its Required Action taken.
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| The Completion Time of 7 days is based on the low probability of the event occurring coincident with a single failure in a remaining OPERABLE channel.
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| S_d With the feedwater and main turbine high water level trip instrumentation not capable of performing its design j function (feedwater and main turbine high water level trip '
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| capability is not maintained), continued operation is only permitted for a 2 hour period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the
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| .feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal. This requires at least one channel to each trip system be OPERABLE or in trip. If the required channels cannot be restored to OPERABLE status or placed in trip Condition C must be entered and its Required Action taken.
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| I FERMI UNIT 2 B 3.3.2.2 -4 Amendment No. 134 i
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| , I Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS (continued)
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| The 2 hour Completion Time is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip instrumentation occurring during this period. It is also consistent with the 2 hour Completion Time provided in LC0 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.
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| C1 With the required channels not restored to OPERABLE status or placed in trip, THERMAL POWER must be reduced to
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| < 25% RTP within 4 hours. As discussed in the Applicability section of the Bases, operation below 25% RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during analyzed events. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 25% RTP from full power conditions in an orderly manner and without Q
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| D ,
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| challenging plant systems.
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| SURVEILLANCE SR 3.3.2.2.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other ,
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| channels. It is based on the assumption that instrument j channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits.
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| O FERMI - UNIT 2 B 3.3.2.2 - 5 Amendment No. 134
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| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
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| (~]
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| v BASES SURVEILLANCE REQUIREMENTS (continued)
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| The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK .'
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| supplements less formal, but more frequent, checks of channel status during normal operational use of the displays ,
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| associated with the channels required by the LCO. l SR 3.3.2.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
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| r~.
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| Q The Frequency of 31 days is reasonable, based on operating experience and on other Surveillances that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. Furthermore.
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| 0)erating experience shows that failure of more than one clannel in a given 31 day period is a rare event.
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| SR 3.3.2.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
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| The Frequency is based upon the assumption of a = 18 month calibration interval in the determination of the magnitude' of equipment drift in the setpoint analysis.
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| FERMI - UNIT 2 B 3.3.2.2 -6 Amendment No. 134
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| p .
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| Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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| SR 3.3.2.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater and main turbine valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a valve is incapable of operating, the associated instrumentation would also be inoperable. The 18 month Frequency is based on the need to perform this Surveillance l under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.
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| REFERENCES 1. UFSAR. Section 15.1.2.
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| pO
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| : 2. UFSAR. Section 15.3.
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| I l
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| i
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| (
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| FERMI UNIT 2 B 3.3.2.2 - 7 Amendment No. 134 l
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| PAM Instrumentation B 3.3.3.1
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| .B 3.3 INSTRUMENTATION
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| /J B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The pririary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This '
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| information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events.
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| The instruments that monitor these variables are designated as Type A. Category I, and non-Type A. Category I, in accordance with Regulatory Guide 1.97 (Ref. 1).
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| The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1.
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| O APPLICABLE- The PAM instrumer:tation LCO ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97. Type A variables so that the control room operating staff can:
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| * Perform the diagnosis specified in the Emergency Operating Procedures (EOPs). These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g.,
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| loss of coolant accident (LOCA)), and
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| * Take the specified, preplanned, manually controlled actions for v;hich no automatic control is provided, which are required for safety systems to accomplish their safety function.
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| The PAM instrumentation LC0 also ensures OPERABILITY of Category I, non Type A variables so that the control room operating staff can:
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| * Determine whether systems important to safety are performing their intended functions:
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| p\}
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| FERMI - UNIT 2 B 3.3.3.1 - 1 Amendment No. 134
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| PAM Instrumentation B 3.3.3.1 BASES APPLICABLE SAFETY ANALYSES (continued)
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| . Determine the potential for causing a gross breach of the barriers to radioactivity release:
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| . Determine whether a gross breach of a barrier has occurred: and
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| . Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.
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| The plant specific Regulatory Guide 1.97 Analysis (Refs. 2, 3, 4, 5, 6, and 7) documents the process that identified Type A and Category I, non Type A, variables.
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| Accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of 10 CFR 50.36(c)(2)(ii). Category I, non Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in minimizing the consequences of accidents.
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| Therefore, these Category I variables are important for reducing public risk.
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| O LC0 LC0 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident.
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| Furthermore, provision of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayd information.
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| The excepion to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information is the status of the primary containment penetrations. The LC0 requires one position indicator for each active PCIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive valve or via system boundary status. If a normally active PCIV is known to be closed and
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| ' deactivated, position indication is not needed to determine i O
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| FERMI - UNIT 2 B 3.3.3.1-2 Amendment No. 134
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| PM Instrumentation B 3.3.3.1 BASES LC0 (continued) status. Therefore, the position indication for valves in this state is not required to be OPERABLE.
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| The following list is a discussion of the specified instrument Functions listed in Table 3.3.3.11 in the accompanying LCO.
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| : 1. Reactor Vessel Pressure Reactor vessel pressure is a Type A. Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS). Two independent pressure transmitters with a range of 0 psig to 1500 psig monitor pressure. Wide range recorders are the primary indication used by the operator during an accident. Therefore. the PM Specification deals specifically with this portion of the i instrument channel.
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| : 2. 3. Reactor Vessel Water Level - Fuel Zone: Reactor Vessel Water Level - Wide Ranoe Reactor vessel water level is a Type A. Category I variable provided to su) port monitoring of core cooling and to verify operation of t1e ECCS. The wide range and fuel zone range water level channels provide the PM Reactor Vessel Water
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| -Level Function. The wide range water level channels measure from 220 inches above the top of active fuel to a point 10 inches above the top of active fuel. The fuel zone range water level channels measure from 50 inches above the top of active fuel to the bottom of active fuel. The two measurement systems provide overlapping ranges to give the operator water level information covering the area of interest during an accident. Wide range water level is ;
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| measured by two independent differential pressure transmitters. The output from these wide range channels is recorded on two independent recorders. Fuel zone range water level is measured by two independent differential pressure transmitters. The output from these fuel zone range channels is recorded on two independent recorders.
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| Each of these recorders also provides indication of uncompensated fuel zone range water level (indication of which is not required by this LC0 for PM OPERABILITY requirements) as well as com)ensated fuel zone range water level (i.e.. the required PA1 indication).
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| O FERMI - UNIT 2 B 3.3.3.1- 3 Amendment No. 134 i
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| PM Instrumentation B 3.3.3.1 BASES
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| (
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| LC0 (continued)
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| In addition to uncompensated and compensated water level indication, indication of reactor vessel pressure is provided on each recorder. This reactor pressure signal is used to derive the compensated fuel zone range water level indication, which provides a more accurate measurement during LOCA transients. This indication of reactor pressure is not required by this LC0 for PM OPERABILITY requirements, however, the signal input to the compensation circuit is required.
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| : 4. Sucoression Pool Water Level Suppression pool water level is a Type A. Category I variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function.
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| The wide range suppression pool water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS. The wide range water level indicators monitor the suppression pool water level from 144 inches below the
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| /~
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| b' normal water level to 56 inches above the normal water level. Two wide range suppression pool water level signals are transmitted from separate differential pressure transmitters and are continuously recorded on two recorders in the control room. These recorders are the primary indication used by the operator during an accident.
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| Therefore, the PM Specification deals specifically with this portion of the instrument channel.
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| : 5. Sucoression Pool Water Temoerature Suppression pool water temperature is a Type A, Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression pool water temperature instrumentation allows operators to detect trends in su)pression pool water temperature in sufficient time to ta(e action to prevent steam quenching vibrations in the-suppression pool.
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| Only two Category I thermocouple channels are needed for post accident monitoring of suppression pool water temperature (Refs. 3 and 4). The outputs for the PM sensors T50N404A and T50N405B are recorded on two O
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| O I FERMI - UNIT 2 B 3.3.3.1 -4 Amendment No. 134
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| FM Instrumentation B 3.3.3.1 BASES LC0 (continued) independent recorders in the control room (channel A is redundant to channel B). Both of these recorders must be OPERABLE to furnish two channels of PAM indication. These j recorders are the primary indication used by the operator I during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels.
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| : 6. Drywell Pressure Drywell pressure is a Type A. Category I variable provided to detect a breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two wide range drywell pressure signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals !
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| specifically with this portion of the instrument channel. !
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| : 7. 8. Primary Containment Hydrooen and 0xvoen Concentration Primary containment hydrogen and oxygen analyzers are Type C, Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach. This variable is also important in verifying the adequacy of mitigating actions.
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| I
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| : 9. Primary Containment Hioh Ranoe Radiation Monitor Primary containment area radiation (high range) is a Type E, Category I variable, and is provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke-site emergency plans. The instrumentation provided for this function consists of redundant sensors, microprocessors and indicators. A common 2-pen recorder in the control room continuously records signals from both channels. The redundant indicators in the relay room and the common recorder in the control room are the primary indication used by the operator during an accident.
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| Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
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| I l
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| O ;
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| 1 FERMI - UNIT 2 B 3.3.3.1 - 5 Amendment No. 134 i
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| PAM Instrumentation B 3.3.3.1 BASES LLO (continued)
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| : 10. Primary Containment Isolation Valve (PCIV) Position PCIV position is a Type B, Category I variable, and is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LC0 requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two total channels j of PCIV position indication for a penetration flow path with I two active valves. For containment penetrations with only 1 one control room indication, Note (b) requires a single i channel of valve position indication to be OPERABLE. This !
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| is sufficient to redundantly verify the isolation status of I each isolable penetration via indicated status of the active valve. as applicable, and prior knowledge of passive valve {
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| 1 or system boundary status. If a penetration flow Jath is !
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| isolated, position indication for the PCIV(s) in tie {
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| associated penetration flow path is not needed to determine i status. Therefore, the position indication for valves in an isolated penetration flow path is not required to be
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| ('3 OPERABLE. The PCIV po'.ition PAM instrumentation consists of U position switches, wiring, cabling, and control room indicating lamps for active PCIVs. Therefore, the PAM specification deals specifically with these instrument channels.
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| APPLICABILITY The PAM instrumentation LC0 is a)plicable in MODES 1 and 2.
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| These variables are related to t1e diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event l that would require PAM instrumentation is extremely low; therefore. PAM instrumentation is not required to be OPERABLE in these MODES.
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| ACTIONS Note 1 has been added to the ACTIONS to exclude the MODE !
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| change restriction of LC0 3.0.4. This exception allows I entry into the applicable MODE while relying on the ACTIONS l even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to A
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| U FERMI - UNIT 2 B 3.3.3.1 - 6 Amendment No. 134
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| T PM Instrumentation B 3.3.3.1
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| ( BASES ACTIONS (continued) diagnose an accident using alternative instruments and methods, and the low probability of an event requiring these instruments.
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| Note 2 has been provided to modify the ACTIONS related to PM instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered.
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| subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable PM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PM Function. {
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| L.1 When one or more Functions have one required channel that is O inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other non Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PM instrumentation during this interval.
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| U I If a channel has not beea restored to OPERABLE status in !
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| 30 days, this Required Action specifies initiation of action i in accordance with Specification 5.6.7, which requires a written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.
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| FERMI - UNIT 2 B 3.3.3.1 - 7 Amendment No. 134
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| 'PAM Instrumentation B 3.3.3.1 BASES ACTIONS (continued)
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| This action is appropriate in lieu of a shutdown requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.
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| L1 When one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PiJi instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an O accident occur. Condition C is modified by a Note that excludes hydrogen concentration or oxygen concentration channels. Condition D provides appropriate Required Actions for two inoperable hydrogen concentration or oxygen concentration channels.
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| L1 When two hydrogen concentration or oxygen concentration channels are inoperable, one hydrogen concentration or oxygen concentration channel must be restored to OPERABLE status within 72 hours. The 72 hour Completion Time is based on the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit: the length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit: and the availability of the hydrogen recombiners, the Primary Containment Nitrogen Inerting and Purge System, and the Post Accident Sampling System.
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| O FERMI' UNIT 2 B 3.3.3.1- 8 Amendment No. 134
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| E !
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| i PAM Instrumentation I B 3.3.3.1 i BASES ACTIONS (continued)
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| L1 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent.
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| Each time an inoperable channel has not met any Required Action of Condition C or D, as applicable, and the associated Completion Time has expired. Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition. 1 El For the majority of Functions in Table 3.3.3.1-1. if any Required Action and associated Completion Time of j Condition C or D are not met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this ,
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| status, the plant must be brought to at least MODE 3 within 12 hours. Tw allowed Completion Times are reasonable, based on operating experience, to reach the recuired plant conditions from full power conditions in an orcerly manner and without challenging plant systems.
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| u 3 Since alternate means of monitoring primary containment area l radiation have been developed sad tested. the Required Action is not to shut down the plant, but rather to follow
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| 'he directions of Specification 5.6.7. . These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.
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| O FERMI -'' UNIT 2 B 3.3.3.1 - 9 Amendment No. 134 !
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| i i
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| i
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| PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE- The following SRs apply to each PAM instrumentation Function REQUIREMENTS in Table 3.3.3.1-1, SR 3.3.3.1.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normelly a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels mon,1toring the same parameter should rcad approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift. in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure: thur, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determ,ined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indicatiori, and readability. If a 3' channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
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| The Frequency of 31 days is based upon plant operating experience, with regard to channel DPERABILITY and drift, which demonstrates that failure of niore than one channel of I a given Function in any 31 day interval is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operat.ional use of those displays associated with the required channels of this LCO.
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| l O
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| FERMI - UNIT 2 B 3.3.3.1 - 10 Amendment No. 134
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| 1 PAM Instrumentation B 3.3.3.1 f BASES SURVEILLANCE REQUIREMENTS (continued)
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| SR 3.3.3.1.2 and SR 3.3.3.1.3 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies the channel responds to measured parameter with the necessary range and accuracy.
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| The 18 month Frequency for all channels except the primary containment oxygen and hydrogen analyzers (per Note 1 to SR 3.3.3.1.3) is based on operating experience and consistency with the typical industry refueling cycles. The 92 day Frequency for the primary containment oxygen and hydrogen analyzers (per Note 1 to SR 3.3.3.1.2) is based upon vendor recommendations and instrument accuracy requirements.
| |
| SR 3.3.3.1.2 is modified by Note 2 stating that performance of the calibration of the oxygen and hydrogen monitors may be delayed until after exceeding 15? RTP (i.e., the power at which LC0 3.6.3.2 requires the primary containment to be inerted). This delay is allowed for up to 72 hours for one oxygen and one hydrogen monitor, and for 7 days for the e second oxygen and hydrogen monitor. These delays facilitate more accurate calibration methods, which can be employed with the primary containment inerted.
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| SR 3.3.3.1.3 is also modified by Note 2 stating that radiation detectors are excluded from calibration requirements.
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| REFERENCES 1. Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,"
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| Rev. 2, December 1980.
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| : 2. Detroit Edison Letter NRC 89 0148, " Additional Clarification to Fermi 2 Compliance to Regulatory Guide 1.97, Revision 2," dated June 19, 1989.
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| : 3. Detroit Edison Letter NRC 89 201, " Regulatory Guide !
| |
| 1.97 Revision 2 Design Review," dated September 12, 1989.
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| es A
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| FERMI - UNIT 2 B 3.3.3.1-11 Amendment No. 134 l
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| )
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| | |
| PAM Instrumentation B 3.3.3.1 BASES-REFERENCES (continued) )
| |
| : 4. NRC Letter " Emergency Response Capability Conformance to Regulatory Guide 1.97, Revision 2 (TAC No. 59620),"
| |
| dated May 2, 1990.
| |
| : 5. Detroit Edison Letter NRC 93-0105, " Fermi 2 Review of Neutron Monitoring System Against Criteria of NED0 31558A," dated September 28, 1993.
| |
| : 6. NRC Letter, " Regulatory Guide 1.97 Boiling Water i Reactor Neutron Flux Monitoring - Fermi 2 (TAC No. M5%20)," dated February 17, 1994.
| |
| : 7. NRC Letter " Regulatory Guide 1.97 - Boiling Water 1 Reactor Neutron Flux Monitoring Fermi 2 (MPA 17 l TAC No. M59620)," dated May 10, 1993. I O
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| j O
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| FERMI UNIT 2 B 3.3.3.1 -12 Amendment No. 134
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| l Remote Shutdown System B 3.3.3.2 l B 3.3 INSTRUMENTATION B 3.3.3.2 Remote Shutdown System i
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| BASES BACKGROUND The Remote Shutdown System provides the control room operator with sufficient instrumentation and controls to place and maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC) System, the safety / relief valves, and the Residual Heat Removal Shutdown Cooling System can be used to remove core decay heat and meet all safety requirements. The long term supply of water i for the RCIC and the ability to operate shutdown cooling I from outside the control room allow extended operation in MODE 3.
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| The design basis of this system does not include the loss of a offsite AC power. Other systems and procedures (e.g., the j n alternative shutdown system) accommodate that design basis i,j event combined with the coincident loss of control room habitability.
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| In the event that'the control room becomes inaccessible, the l operators can establish control at the remote shutdown panel l and place and maintain the plant in MODE 3. The design !
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| basis of this system includes the assumption that the ;
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| operator will initiate a manual scram to place the plant in I MODE 3, prior to establishing control from the Remote Shutdown Panel, where the plant can be maintained safely in MODE 3 for an extended period of time.
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| The OPERABILITY of the Remote Shutdown System control and instrumentation Functions ensures that there is sufficient information available on selected )lant parameters to place and maintain the plant in MODE 3 s1ould the control room become inaccessible.
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| Ov FERMI - UNIT 2 B 3.3.3.2 - 1 Amendment No. 134
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| Remote Shutdown System B 3.3.3.2 O BASES V
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| APPLICABLE The Remote Shutdown System is required to provide equipment SAFETY ANALYSES at appropriate locations outside the control room with a design capability to promptly shut down the reactor to MODE 3. including the necessary instrumentation and controls, to maintain the plant in a safe condition in MODE 3.
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| The criteria governing the design and the specific system requirements of the Remote Shutdown System are located in 10 CFR 50, Appendix A. GDC 19 (Ref. 1).
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| The Remote Shutdown System is considered an important contributor to reducing the risk of accidents: as such, it has been retained in the Technical Specifications (TS) as indicated in 10 CFR 50.36(c)(2)(ii).
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| LC0 The Remote Shutdown System LC0 provides the requirements for the OPERABILITY.of the instrumentation and controls necessary to place and maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table 3.3.3.2-1.
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| The Remote Shutdown System is OPERABLE if all instrument and control channels needed to support the remote shutdown function are OPERABLE.
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| The Remote Shutdown System instruments and control circuits covered by this LC0 do not need to be energized to be considered OPERABLE. This LC0 is intended to ensure that the instruments and control circuits will be OPERABLE if plant conditions require that the Remote Shutdown System be placed in operation.
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| APPLICABILITY The Remote Shutdown System LC0 is applicable in MODES 1 and 2. This is required so that the plant can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room.
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| A b
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| FERMI - UNIT 2 B 3.3.3.2 -2 Amendment No. 134
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| Remote Shutdown System B 3.3.3.2 BASES APPLICABILITY (continued)
| |
| This LC0 is not applicable in MODES 3, A, and 5. In these MODES, the plant is already subcritical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control becomes unavailable. Consequently, the TS do not require OPERABILITY in MODES 3, 4, and 5.
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| ACTIONS. A Note is included that excludes the MODE change restriction of LC0 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require a plant shutdown. This exception is acceptable due to the low probability of an event requiring this system.
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| Note 2 has been provided to modify the ACTIONS related to Remote Shutdown System Functions. Section 1.3, Completion Times, specifies that once a Condition has been entered.
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| subsequent divisions, subsystems, components, or variables p expressed in the Condition, discovered to be inoperable or d not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable Remote Shutdown System Functions provide appropriate compensatory measures for separate Functions.
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| As such, a Note has been provided that allows separate Condition entry for each inoperable Remote Shutdown System Function.
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| bal Condition A addresses the situation where one or more required Functions of the Remote Shutdown System is inoperable. This includes any Function listed in Table 3.3.3.2 1.
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| The Required Action is to restore the Function to OPERABLE status within 30 days. The Completion Time is based on omrating experience and the low probability of an event tlat would require evacuation of the control room.
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| O FERMI UNIT 2 B 3.3.3.2-3 Amendment No. 134
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| | |
| Remote Shutdown System B 3.3.3.2 BASES ACTIONS (continued) lL1 l If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought' to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience.. to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
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| SURVEILLANCE SR 3.3.3.2.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A ,
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| CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations O- '
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| between the instrument channels could be an indication of excessive instrument drift in one of the channels or ,
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| something even more serious. A CHANNEL CHECK will detect gross _ channel failure: thus, it is key to verifying the
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| -instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted l outside its limit. i The Frequency is based upon plant operating experience that demonstrates channel failure is rare.
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| l FERMI - UNIT 2 B 3.3.3.2-4 Amendment No. 134
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| Remote Shutdown System B 3.3.3.2 O
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| V BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.3.2.2 SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit pcrfarr.s the intended function. This verification is performed from the remote I shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. Operating c.voerience demonstrates .
| |
| that Remote Shutdown System control ch enels usually pass the Surveillance when performed at the 153 month Frequency. ]
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| j SR 3.3.3.2.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds !
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| to measured parameter values with the necessary range and j accuracy.
| |
| The 18 month Frequency is based upon operating experience ,
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| and consistency with the typical industry refueling cycle. j l
| |
| l REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.
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| I I
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| A V :
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| l FERMI - UNIT 2 B 3.3.3.2-5 Amendment No. 134
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| ATWS RPT Instrumentation B 3.3.4.1
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| ( 'B 3.3 INSTRUMENTATION B 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATVS RPT) Instrumentation BASES BACKGROUND The ATWS RPT System initiates an RPT, adding negative reactivity. following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
| |
| Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level-Low Low, Level 2 or Reactor Vessel Pressure-High setpoint is reached, the recirculation drive motor breakers trip. pump MG set generator field and The ATWS RPT System (Ref. 1) includes sensors, relays, bypass capability circuit breakers, and switches that are I necessary to cause initiation of an RPT. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre established setpoints. When the set)oint is exceeded, the channel j
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| ( output relay actuates, w11ch then outputs an ATWS RPT signal to the trip logic.
| |
| The ATWS r"T consists of two independent trip systems, each with t*c m:.anels of Reactor Vessel Pressure-High and two channel' ef Reactor Vessel Water Level-Low Low, Level 2 in each tr1, system. Each ATWS RPT trip system is a two out % /-two logic for each Function. Thus, either two Reactor Water Level-Low Low, Level 2 or two Reactor Vessel Pressure-High signals are needed to trip a trip system.
| |
| The outputs of the channels in a trip system are combined in j a logic so that either trip system will trip both recirculation pumps (by tripping the respective MG set generator field and drive motor breakers). Each breaker has two trip coils, which allows each trip system to trip both breakers. j The ATWS RPT logic delays the MG set generator field and drive motor breakers trip on low reactor water level for a time delay of 9 seconds. This time delay accounts for the difference in pump coastdown time compared to a direct trip of the MG set drive motor, which was assumed in the LOCA analysis.
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| b(*
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| FERMI UNIT 2 B 3.3.4.1 - 1 Amendment No. 134
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| | |
| ATWS RPT Instrumentation B 3.3.4.1 BASES BACKGROUND (continued)
| |
| There is one MG set generator field and drive motor breaker provided for each of the two recirculation pumps for a total of two breakers per pump. The output of each trip system is provided to both recirculation pump MG set generator field and drive motor breakers.
| |
| APPLICABLE The ATWS RPT is not credited in the safety analysis for SAFETY ANALYSES, accident or anticipated operational occurrences mitigation.
| |
| -LCO, and The ATWS RPT initiates an RPT to aid in preserving the APPLICABILITY integrity of the fuel cladding following events in which a scram does not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation is included as required by 10 CFR 50.36(c)(2)(ii).
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| The OPERABILITY of the ATWS RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a reguired number of OPERABLE channels in each trip system, with their setpoints p within the specified Allowable Value of SR 3.3.4.1.3. The A actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated recirculation pump MG set generator field and drive motor breakers. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.
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| Allowable Values are specified for each ATWS RPT Function specified in the LCO. Nominal' trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS, Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
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| Trip setpoints are those ) redetermined values of output at which an action should tace place. The setpoints are compared to the actual process )arameter (e.g., reactor vessel water level), and when t1e measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and some of the O
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| FERMI UNIT 2 B 3.3.4.1 - 2 Amendment No. 134
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| | |
| AWS RPT Instrumentation B 3.3.4.1 (9
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| V BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) ,
| |
| instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g.,
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| drift) . The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
| |
| The individual Fenctions are required to be OPERABLE in MODE 1 to protect against common mode failures of the Reactor Protection System by providing a diverse trip to mitigate the consequences of a postulated AWS event. The Reactor Vessel Pressure-High and Reactor Vessel Water Level-Low Low, Level 2 Functions are required to be OPERABLE in MODE 1 since the reactor is producing significant power and the recirculation system could be at high flow. During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event. In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is G low for a pressure increase or low water level, assuming an V ATWS event. Therefore, the ATWS RPT is not necessary.
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| MODES 3 and 4, the reactor is shut down with all control In rods inserted: thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical:
| |
| thus, an ATWS event is not significant. In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists.
| |
| The specific Applicable Safety Analyses and LC0 discussions are listed below on a Function by function basis.
| |
| : a. Reactor Vessel Water Level-Low Low. Level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the ATWS RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. ,
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| The reduction of core flow reduces the neutron flux and '
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| THERMAL POWER and, therefore, the rate of coolant boiloff.
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| FERMI UNIT 2 B 3.3.4.1 - 3 Amendment No. 134
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| | |
| I ATWS RPT Instrumentation B 3.3.4.1 O
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| (/
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| BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
| |
| Four chanrels of Reactor Vessel Water Level-Low Low, Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
| |
| : b. Reactor Vessel Pressure-Hioh Excessively high RPV pressure may rupture the RCPB. A1 increase in the RPV pressure during reactor operation p compresses the steam voids and results in a positive V reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Vessel Pressure-High function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV 3ressure to less than the ASME Section III Code Service
| |
| _evel C limits.
| |
| The Reactor Vessel Pressure-High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Vessel Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS RPT from this Function on a valid signal. The Reactor Vessel Pressure-High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code Service Level C allowable Reactor Coolant System pressure.
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| O V
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| FERMI UNIT 2 B 3.3.4.1 - 4 Amendment No. 134 l
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| u
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| | |
| ATWS RPT Instrumentation B 3.3.4.1 BASES ACTIONS A Note has been provided to modify the ACTIONS related to ATWS RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ATWS RPT instrumentation channels provide a>propriate compensatory measures for separate inoperable clannels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel.
| |
| A.1 and A.2 With one or more channels inoperable, but with ATWS RPT capability for each Function maintained (refer to Required Actions B.1 and C.1 Bases), the ATWS RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation
| |
| ( ' is reduced, such that a single failure in the remaining trip system could result in the inability of the AlVS RPT System to perform the intended function. Therefore, only a limited time is allewed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to 3rovide trip signals, the low probability of extensive num)ers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS RPT,14 days i provided to restore the inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in tri) with no further restrictions is not allowed if the inopera)le channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be i inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT), or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its
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| ' Required Actions taken.
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| !(
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| FERMI UNIT 2 B 3.3.4.1 - 5 Amendment No. 134
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| | |
| ATWS RPT Instrumentation B 3.3.4.1 i BASES ACTIONS (continued) fL1 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untriiped channels within the same Function result in the unction not maintaining ATWS RPT trip capability. A Function is considered to be maintaining ATWS RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires two channels of the Function in the same trip system to each be OPERABLE or in trip, and the recirculation pump HG set generator field and drive motor '
| |
| breakers to be OPERABLE or in trip.
| |
| The 72 tour Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of channels) and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining ATVS RPT tr,p capability.
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| O V C.l Reouired Action C.1 is intended to ensure that appropriate ;
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| Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not '
| |
| maintaining ATWS RPT trip capability. The description of a Function maintaining ATWS RPT trip capability is discussed in the Bases for Required Action B.1 above.
| |
| The 1 hour Completion Time is sufficient for the operator to i take corrective action and takes into account the likelihood l of an event requiring actuation of the ATWS RPT 1 instrumentation during this period. I n
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| V FERMI - UNIT 2 B 3.3.4.1 - 6 Amendment No. 134 i
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| J
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| ATWS RPT Instrumentation B 3.3.4.1 BASES ACTIONS (continued).
| |
| D.1 and D.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump may be removed from service since this performs the intended function of the instrumentation (Required Action D.1). The allowed Completion Time of 6 hours is reasonable, based on operr. ting experience, both 1 to reach MODE 2 from full power conditions and to remove a i
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| . recirculation pump from service in an orderly manner and without challenging plant systems.
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| l SURVEILLANCE The Surveillances are modified by a Note to indicate that i REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 2 hours provided the associated Function maintains
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| ( AlWS RPT trip capability. Upon completion of the Surveillance, or expiration of the 2 hour allowance, the channel must be returned to OPERABLE status or the a)plicable Condition entered and Required Actions taken. ;
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| T1e 2 hour testing allowance does not significantly reduce l the probability that the recirculation pumps will trip when '
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| necessary.
| |
| SR 3.3.4.1.1 i Performance of the CHANNEL CHECK once every 12 hours ensures '
| |
| that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| ,D G
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| FERMI UNIT 2 B 3.3.4.1 - 7 Amendment No. 134
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| ATWS RPT Instrumentation B 3.3.4.1 BASES SURVEllt.ANCE REQUIREMENTS (continued)
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| ^
| |
| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| The Frequency is based upon operating ex)erience that demonstrates channel failure is rare. T1e CHANNEL CHECK supplements less formal, but more frequent, checks of !
| |
| channels during normal operational use of the displays associated with the required channels of this LCO.
| |
| SR 3.3.4.1.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the ;
| |
| intended function. A successful test of the required i contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| Any setpoint adjustment shall be consistent with the assumptions of.the current plant specific setpoint methodology.
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| The Frequency of 31 days is reasonable, based on operating experience and on other Surveillances that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. Furthermore, o>erating experience shows' that the failure of more than one clannel in a given 31 day period is a rare event.
| |
| SR 3.3.4.1.3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts'between successive calibrations consistent with the plant specific setpoint methodology.
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| 1 O
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| FERMI UNIT 2 B 3.3.4.1 - 8 Amendment No. 134
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| l l
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| ATWS RPT Instrumentation B 3.3.4.1
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| ( BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| The Frequency is based upon the assumption of a = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| SR 3.3.4.1.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific 1 channel. The system functional test of the pump breakers is I included as part of this Surveillance and overlaps the LOGIC {
| |
| SYSTEM FUNCTIONAL TEST to provide complete testing of the ;
| |
| assumed safety function. Therefore, if a braaker is incapable of operating, the associated instrument channel (s) would be inoperable. ;
| |
| The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un)lanned transient if the Surveillance were performed with t1e reactor at power.
| |
| Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
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| O I i
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| REFERENCES 1. UFSAR, Figure 7.7-3, Reactor Recirculation System FCD.
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| C\
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| U FERMI UNIT 2 B 3.3.4.1 - 9 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1
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| 'B 3.3 INSTRlNENTATION B 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation BASES-BACKGROUND The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis 1
| |
| -accident or transient. j 1
| |
| For most anticipated operational occurrences and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored.
| |
| The ECCS instrumentation actuates core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection-(HPCI), Automatic Depressurization System (ADS),
| |
| and the emergency diesel generate s (EDGs). The equi) ment involved with each of these systems is described in t1e l Bases for LC0 3.5.1, "ECCS-Operating." '
| |
| Core Soray System The CS System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of
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| ~
| |
| Reactor Vessel Water Level-Low Low Low, Level 1 or Drywell Pressure-High. Each of these diverse variables is '
| |
| monitored by four redundant transmitters, which are, in turn, connected to four trip units, ' The outputs of the eight trip units are connected to relays whose contacts are arranged in a one out of two taken twice logic-(i.e., two 4 trip systems).for each Function.
| |
| Once an initiation signal is received by the CS control circuitry, the signal is sealed-in until manually reset.
| |
| Automatic initiation starts both CS pumps in both loops five seconds after initiation when normal AC >ower is available.
| |
| If normal AC power is not available, eac1 pump starts five seconds after standby power becomes available to that pump.
| |
| When RPV pressure decreases below the injection permissive setpoint. the pum) injection valves open allowing water to be sprayed over t1e core.
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| O FERMI'- UNITL2 B 3.3.5.1 - 1 Amendment No. 134
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| i, ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND (continued)
| |
| The CS test line isolation valve, which is also a primary containment isolation valve (PCIV), is closed on a CS initiation signal to allow full system flow assumed in the accident analyses and maintain primary containment isolated in the event CS is not operating.
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| The CS System also monitors the pressure in the reactor to ensure that, before the injection valves open, the reactor pressure has fallen to a value below the CS System's maximum design pressure. The variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out of two taken twice logic.
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| Low Pressure Coolant Iniection System The LPCI is an operating mode of the Residual Heat Removal (RHR) System, with two LPCI subsystems. The LPCI subsystems may be initiated by automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water D Level-Low Low Low, Level 1: Drywell Pressure-High: or both. Each of these diverse variables is monitored by four redundant transmitters, which, in turn, are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one out of two taken twice logic (i.e., two trip systems) for each Function. Once an initiation signal is received by the LPCI control circuitry, the signal is sealed in until manually reset.
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| Upon receipt of' an initiation signal, if normal power is I available, all four RHR pumps start with no delay, i Otherwise, they each start when standby )ower is available. l Valves are automatically positioned for _PCI injection, which occurs when RPV pressure falls below the injection permissive setpoint. This setpoint is selected to protect the RHR system from overpressure. A Reactor Vessel Water Level-Low Low, Level 2 signal initiates the loop selection logic, a divisionalized subsystem using redundant sensors that determines which if any, recirculation loop is broken.
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| This logic then aligns the pumps injection valves to the unbroken loop to ensure the system function of flooding the core to at least 2/3 core height will be accomplished. Once selected, the injection path is sealed in for five minutes, after which a timer allows the operator to throttle the O
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| FERMI UNIT 2 B 3.3.5.1 -2 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1 BASES BACKGROUW (continued) flow. Once identified, the broken recirculation loop is isolated for ten minutes. These times ensure the LPCI design flows will reflood the core to at least 2/3 core height.
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| The loop selection logic has four channels that are arranged
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| ~ in a one out-of two taken twice logic for each of the following Functions: RPV Pressure (which initiates the break detection logic on a decreasing setpoint): Reactor Vessel Water Level-Low Low, Level 2 (which provides a break detection permissive signal): and differential pressures between the recirculation loop risers, and between the recirculation pump suction and discharge (which are used to determine recirculation pump operation and which loop (s) are unbroken), j The recirculation pump discharge valve on the unbroken loop is closed by the-loop selection logic to direct LPCI flow to the reactor through the suction line without being diverted through the recirculation pump.
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| s The RIR test / suppression pool cooling line isolation valve, !
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| suppression pool spray isolation valves, and containment I spray isolation valves (which are also PCIVs) are also closed on a LPCI initiation signal to allow the' full system flow assumed in the accident analyses and maintain primary
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| . containment isolated in the event LPCI is not operating.
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| Hiah Pressure Coolant In.iection System The HPCI System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low, Level 2 or Drywell Pressure-High. Each of these variables is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of two taken tstice logic for each Function.
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| The HPCI test line isolation valve (which is also a PCIV) is closed upon receipt of a HPCI initiation signal to allow the full system flow assumed in the accident analysis and maintain primary containment isolated in the event HPCI is j not operating. '
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| l O
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| FERMI UNIT 2 B 3.3.5.1 -3 Amendment No. 134 i
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| ECCS Instrumentation B 3.3.5.1 p
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| y BASES BACKGROUND (continued) >
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| The HPCI System also monitors the water levels in the condensate storage tank (CST) and the suppression pool because these are the two sources of water for HPCI operation. Reactor grade water in the CST is the normal source. Upon receipt of a HPCI initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position) unless both suppression pool suction valves are open. If the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. Two channels of level transmitters and trip units are used to detect low water it. vel in the CST. Either channel can cause the suppression pool suction valves to open and the CST suction valve to close. Two channels of level transmitters and trip units monitor suppression pool water level. The suppression pool suction valves also automatically open and the CST suction valve closes if high water level is detected in the suppression pool. To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.
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| The HPCI provides makeup water to the reactor until the reactor vessel water level reaches the Reactor Vessel Water Level-High, Level 8 trip, at which time the HPCI turbine trips, which causes the turbine's stop valve and the injection valves to close. The logic is two out of-two to provide high reliability of the HPCI System. The HPCI System automatically restarts if a Reactor Vessel Water Level-Low Low, Level 2 signal is subsequently received.
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| Automatic Deoressurization System The ADS may be initiated by either automatic or manual means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Low Low, Level 1: Drywell Pressure-High or Drywell Pressure-High Bypass Timer:
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| confirmed Reactor Vessel Water Level-Low, Level 3: and CS or LPCI Pump Discharge Pressure-High are all present and the ADS Initiation Timer has timed out. There are two transmitters each for Reactor Vessel Water Level-Low Low Low, Level 1 and Drywell Pressure-High, and one transmitter for confirmed Reactor Vessel Water Level-Low, Level 3 in !
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| each of the two ADS trip systems. Each of these I transmitters connects to a trip unit, which then drives a relay whose contacts form the initiation logic.
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| j v
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| FERMI - UNIT 2 B 3.3.5.1 - 4 Amendment No. 134 !
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| ECCS Instrumentation B 3.3.5.1
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| / BASES BACKGROUND (continued)
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| Each ADS tri) system includes a time delay between satisfying tie initiation logic and the actuation of the ADS valves. The ADS Initiation Timer time delay setpoint chosen is long enough that the HPCI has sufficient operating time to recover to a level above Level 1. yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI fails to maintain that level. An alarm in the control room is annunciated when either of the timers is timing. Resetting the ADS initiation signals resets the ADS Initiation Timers.
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| The ADS also monitors the discharge pressures of the four LPCI pumps and the four CS pumps. Each ADS trip system includes a single pressure transmitter from each CS pump and two pressure transmitters from each LPCI pump in the associated Division (i.e.. Division 1 LPCI subsystems A and C. and CS pumps A and C input to ADS trip system A: and division 2 LPCI subsystems B and D. and CS pumps B and D input to ADS tri) system B). The signals are used as a permissive for A)S actuation, indicating that there is a source of core coolant available once the ADS has depressurized the vessel. Either one LPCI pump (two out-of-s~) two logic) or both core spray pumps in the same division is sufficient to permit automatic depressurization.
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| The ADS logic in each trip system is arranged in two :
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| strings. Each string has a contact from each of the '
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| following variables: Reactor Vessel Water Level-Low Low Low. Level 1: Drywell Pressure-High; and Drywell Pressure-High Bypass Timer. One of the two strings in each trip system must also have a confirmed Reactor Vessel Water Level-Low. Level 3. Each of these contacts in both logic strings must close, the ADS initiation timer must time out.
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| and a CS or LPCI pump discharge pressure signal must be present in both strings to initiate an ADS trip system.
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| Either the A or B trip system will cause all five ADS relief valves to open. Once the Drywell Pressure-High signal. the ADS Low Water Level Actuation Timer, or the ADS initiation signal is present. it is individually sealed in until manually reset. The reactor vessel water Level Low Low Low.
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| Level 1 does not seal in to give HPCI an opportunity to restore level before ADS initiates a blow down.
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| Manual inhibit switches (one for each ADS trip system) are provided in the control room for the ADS.
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| v FERMI UNIT 2 B 3.3.5.1 - 5 Amendment No. 134
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| 1 ECCS Instrumentation B 3.3.5.1 BASES-BACKGR0Vf0 (continued)
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| Emeroency Diesel Generators The EDGs may'be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or Drywell Pressure-High. The EDGs are also initiated upon loss of voltage signals. Each of these diverse variables is monitored by four redundant transmitters, which are, in turn. connected to four trip units. The outputs of. the four trip units are connected to relays whose contacts are connected to a one out of-two taken twice logic to initiate all four EDGs (ll.12,13, and 14). The EDGs receive their initiation signals from the CS System initiation logic. The EDGs can also be started manually from the control room and locally: from the associated EDG room. The EDG initiation signal is a sealed in signal and must be manually reset.
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| The EDG initiation logic is reset by resetting the associated ECCS initiation logic. Upon receipt of a loss of coolant accident (LOCA) initiation signal, each EDG is automatically started, is ready to load in approximately 10 seconds, and will run in standby conditions (rated voltage and speed, with the EDG output breaker open). The O EDGs will only energize their respective Engineered Safety Feature buses if a loss of offsite power occurs. (Refer to Bases for LC0 3.3.8.1.)
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| APPLICABLE The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES, analyses of References 2, 3, and 4. The ECCS is initiated LCO, and to preserve the integrity of the fuel cladding by limiting APPLICABILITY the mst LOCA peak cladding temperature to less than the 10 C R 50.46 limits.
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| ECCS instrumentation satisfies Criterion 3 of 10 CFR I 50.36(c)(2)(ii). Certain instrumentation Functions are !
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| retained for other reasons and are described below in the !
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| individual Functions discussion.
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| 4 The OPERABILITY of the ECCS instrumentation is dependent
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| - umn the OPERABILITY of the individual instrumentation clannel Functions specified in Table 3.3.5.11. Each Function must have a required number of OPERABLE channels.
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| . with their setpoints within the specified Allowable Values, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
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| O FERMI - UNIT 2 B 3.3.5.1 - 6 Amendment No. 134
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| r.
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| ECCS Instrumentation B 3.3.5.1 O
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| v BASES APPLICABLE SAFETY ANALYSES. LC0. and APPLICABILITY (continued)
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| Table 3.3.5.1-1 footnote (b) is added to show that certain ECCS instrumentation Functions are also required to be OPERABLE to perform EDG initiation.
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| Allowable Values are specified for each ECCS Function specified in the table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpcints do not exceed the Allowable Value between CHANNEL CALIBRATIONS or between successive verifications of the trip unit setpoints.
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| Operation with a trip setpoint less conservative than the nominal trip set)oint, but within its Allowable Value, is acceptable. A clannel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to
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| , the actual process parameter (e.g., reector vessel water level), and when the measured out)ut value of the process parameter exceeds the setpoint. t1e associated device (e.g.,
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| trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained q from the safety analysis. The Allowable Values are derived V from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip I setpoints are then determined, accounting for the remaining !
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| instrument errors (e.g., drift). The trip setpoints derived l in this manner provide adequate protection because instrumentation uncertainties, process effects. calibration !
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| tolerances, instrument drift, and severe environment errors '
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| (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. I In general, the individual Functions are required to be ;
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| OPERABLE in the MODES or other specified conditions that may require ECCS (or EDG) initiation to mitigate the consequences of a design basis transient or accident. To ensure reliable ECCS and EDG function, a combination of Functions is required to provide primary and secondary initiation signals.
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| The specific Applicable Safety Analyses. LCO, and Applicability discussions are listed below on a Function by Function basis.
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| O V
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| FERMI - UNIT 2 B 3.3.5.1 - 7 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| Core Sorav and Low Pressure Coolant In.iection Systems 1.a. 2.a. Reactor Vessel Water Level-Low Low Low. Level 1 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
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| The low pressure ECCS and associated EDGs are initiated at Level 1 to ensure that core spray and flooding functions are available to 3revent or minimize fuel damage. The Reactor Vessel Water _evel-Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in Reference 2. In addition, the Reactor Vessel Water Level-Low Low Low, Level 1 Function is directly assumed in the analysis of the recirculation line break (Ref. 1). The core cooling function of the ECCS, along with the scram action of the I Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| e Reactor Vessel Water Level-Low Low Low, Level 1 signals are I initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual '
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| water level (variable leg) in the vessel.
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| The Reactor Vessel Water Level-Low Low Low, Level 1 ;
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| Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.
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| Four channels of Reactor Vessel Water Level-Low Low Low.
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| Level 1 Function are only required to be OPERABLE when the ECCS or EDG(s) are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS and EDG initiation, Refer to LC0 3.5.1 and LC0 3.5.2, "ECCS-Shutdown," for Applicability Bases for the low pressure FCCS subsystems: LC0 3.8.1, "AC Sources-Operating": and LC0 3,8.2, "AC Sources-Shutdown," for Applicability Bases for the EDGs. l O
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| O FERMI UNIT 2 B 3.3.5.1 -8 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1
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| ( BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 1.b. 2.b. Drvwell Pressure-Hiah High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated EDGs are initiated upon receipt of the Drywell Pressure-High Function in order to minimize the
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| )ossibility of fuel damage. The Drywell Pressure-High r unction, along with the Reactor Water Level-Low Low Low.
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| Level 1 Function, is directly assumed in the analysis of the recirculation line break (Ref. 3). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.
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| The Drywell Pressure-High Function is required to be OPERABLE when the ECCS or EDG is required to be OPERABLE in 3 conjunction with times when the primary containment is (V required to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure-High Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude ECCS and EDG initiation. In MODES 4 and 5, the Drywell Pressure-High Function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure-High setpoint. Refer to LC0 3.5.1 for Applicability Bases for the low pressure ECCS subsystems and to LC0 3.8.1 for Applicability Bases for the EDGs.
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| 1.c. 2.c. Reactor Steam Dome Pressure-Low (In.iection Permissive)
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| Lew reactor steam dome pressure signals are used as permissives for the low pressure ECCS subsystems. This ent ures that, prior to opening the injection valves of the lou pressure ECCS subsystems, the reactor pressure has fellen to a value below these subsystems
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| * maximum design pressure. The Reactor Steam Dome Pressure-Low is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in Reference 2. In addition, the Reactor Steam ,
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| Dome Pressure-Low Function is directly assumed in the l i
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| O-i FERMI - UNIT 2 B 3.3.5.1 - 9 Amendment No. 134 i l
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| i j
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| 1 ECCS Instrumentation B 3.3.5.1 l
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| () BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) analysis of the recirculation line break (Ref.1). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| The Reactor Steam Dome Pressure-Low signals are initiated from four pressure transmitters that sense the reactor dome pressure.
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| The Allowable Value is low enough to prevent overpressuring the equi) ment in the low pressure ECCS, but high enough to ensure tlat the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.
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| Four channels of Reactor Steam Dome Pressure-Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation. Refer to LC0 3.5.1 and LC0 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.
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| 1.d. 2.h. Manual Initiation
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| ]
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| The Manual Initiation channel provides manual initiation capability by means of individual component controls. There is one manual initiation channel for each of the CS and LPCI subsystems (i.e., two for CS and two for LPCI).
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| The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the low pressure ECCS function as required by the NRC in the plant licensing basis.
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| There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the i position of the individual components. Each channel of the Manual Initiation Function is only required to be OPERABLE when the associated ECCS is required to be OPERABLE. Refer to LC0 3.5.1 and LC0 3.5.2 for Applicability Bases for the low pressure ECCS subsystems.
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| O FERMI - UNIT 2 B 3.3.5.1 - 10 Amendment No. 134
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| [
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| l-ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY' ANALYSES, LCO, and APPLICABILITY (continued) 2.d Reactor Vessel Water Level-Low Low Level 2 (Looo Selection Loaic)
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| LPCI Loop selection logic is initiated on decreasing RPV water level at level 2. This gives the logic time to detect the broken recirculation loop and select the unbroken recirculation loop for LPCI injection. The LPCI pumps are
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| ' initiated at level 1.
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| Reactor Vessel Water Level '.ow Low, Level 2 signals are initiated from four level transmitters that sense the l- difference between the pressure due to a constant column of l water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The transmitter !
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| l signals feed trip units whose outputs drive relays. Output contacts of the relays are configured in a one-out-of two q l
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| taken twice initiation logic. ,
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| j The same instrumentation and relay logic is used for HPCI I initiation (Function 3a). That system's design basis .
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| establishes the Allowable Value while accounting for O measurement uncertainties. LPCI loop selection initiation is not directly assumed by any safety or transient analysis, but is required to function to su) port the LPCI system. i l
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| j which is assumed to function in t1e accident analysis (Ref. 1).
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| Four channels are required to be OPERABLE whenever LPCI is required to be OPERABLE to ensure that no single instrument failure can preclude LPCI initiation.
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| 2.e. Reactor Steam Dome Pressure-Low (Break Detection Loaic)
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| This function is provided in the LPCI break detection logic.
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| If only one recirculation pump is running, the logic trips that pump in order to obtain a meaningful measurement of recirculation riser differential pressure (Function 2.f).
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| Reactor Steam Dome Pressure-Low inhibits the break i detection logic from acting ~on the value of riser 1 t differential pressure until reactor pressure has fallen below the set point due to the pump trip. This allows the logic to identify the broken recirculation loop. Although !
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| this function is not directly assumed by the safety '
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| analysis, it is required for the LPCI loop selection logic, L
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| FERMI - UNIT 2- B 3.3.5.1 - 11 Amendment No. 134 j
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| ECCS Instrumentation B 3.3.5.1 BASES-APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) and LPCI to be OPERABLE, and is therefore a supporting function for that assumed by the analysis of Reference 1.
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| Reactor Steam Dome Pressure-Low signals are initiated from four pressure transmitters that sense reactor steam dome pressure. The Allowable Value was selected, allowing for measurement uncertainties, to give adequate time, based on reactor pressure decrease following RPT, for an accurate riser differential pressure measurement to be made. The logic for this Function is one-out-of two taken twice.
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| Four channels of Reactor Steam Dome Pressure-Low are required to be OPERABLE when LPCI is required to be OPERABLE to ensure that no single instrument failure can preclude LPCIinjection.
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| 2.f. Riser Differential Pressure-Hiah (Break Detection)
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| The LPCI break detection logic determines which recirculation loop is broken by comparing the pressure of the two recirculation loops. The broken loop will indicate a lower pressure than the unbroken loop. The loop with the n)
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| ( '
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| higher pressure is then used for LPCI injection. If both pressures are the same, loop B is selected by default.
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| Riser Differential Pressure-High signals are initiated from four differential pressure transmitters that sense the difference between corresponding recirculation loop riser pipes. Logic is one out-of two taken twice.
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| The Riser Differential Pressure-High Allowable Value is selected, allowing for measurement uncertainties, based on the analytical limit of 1.0 psid between corresponding risers.
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| Four channels of Riser Differential Pressure-High are required to be OPERABLE to ensure that no single instrument failure prevents LPCI injection into the unbroken riser loop and support the LPCI function.
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| .) l
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| -FERMI - UNIT 2 B 3.3.5.1 - 12 Amendment No. 134
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| ]
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| 1 ECCS Instrumentation B 3.3.5.1 l l
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| I 3 BASES l (Q \
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| APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) 2.0. Recirculation Pumo Differential Pressure-Hiah (Break Detection)
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| The differential pressure between pump suction and discharge indicates whether the pump is running. This information is used by the LPCI break detection logic to determine whether the sensed riser differential pressure is meaningful. If both pumps are running or not running, the logic can use the ,
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| riser differential pressure (Function 2.f) value to 1 determine which loop is broken. If only one pump is running, the logic trips the running pump, waits until the .
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| reactor pressure decreases to a pre selected value (Function j 2.e), then identifies the broken loop based on riser {
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| differential pressure. . This function is necessary to j support LPCI injection, which was assumed in the safety analysis (Ref.1). ,
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| Recirculation Pump Differential Pressure-High signals are initiated from four differential pressure transmitters at each of the two recirculation pumps. The output relay signals are configured in one out of-two taken twice logic !
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| 'injection.
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| ""*''""*The''""'' "'''""*"' ' " "'' ""*"*"' "" '" '
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| (D s Allowable Value is selected allowing for measurement uncertainties to be low enough to distinguish between pump running and not running based on the pump head
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| {
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| curve. Four channels of Recirculation Pump Differential !
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| Pressure-High for each of the two recirculation pumps are required to be OPERABLE to support LPCI injection and preclude a single instrument failure.
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| HPCI System 3.a. Reactor Vessel Water Level-Low Low. Level 2 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water Level-Low Low, Level 2 is one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in Reference 3. Additionally, the Reactor Vessel Water Level-Low Low. Level 2 Function associated with HPCI is directly assumed in the analysis of the recirculation line break (Ref. 1). The core cooling function of the ECCS.
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| along with the scram action of the RPS, ensures that the FERMI - UNIT 2 B 3.3.5.1 - 13 Amendment No. 134 l 1
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| I ECCS Instrumentation B 3.3.5.1 l
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| BASES (V)
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| APPLICABLE SAFETY ANALYSES. LCO and APPLICABILITY (continued) fuel peak cladding temperature remains below the limits of 10 CFR 50.46. j Reactor Vessel Water Level-Low Low. Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of I water (reference leg) and the pressure due to the actual {
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| water level (variable leg) in the vessel. !
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| The Reactor Vessel Water Level-Low Low. Level 2 Allowable Value is high enough such that for complete loss of l feedwater flow. the Reactor Core Isolation Cooling (RCIC) !
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| System flow with HPCI assumed to fail will be sufficient to 1 avoid initiation of low pressure ECCS at Reactor Vessel )
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| Water Level-Low Low Low. Level 1. 1 l'
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| Four channels of Reactor Vessel Water Level-Low Low.
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| Level 2 Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single i instrument failure can preclude HPCI initiation. Refer to !
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| LC0 3.5.1 for HPCI Applicability Bases.
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| 3.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. The HPCI System is initiated upon receipt of the Drywell Pressure-High Function in order to minimize the possibility of fuel damage. The Drywell Pressure-High Function, along with the Reactor Water Level-Low Low. ,
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| Level 2 Function, is directly assumed in the analysis of the '
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| recirculation line break (Ref. 4). The core cooling function of the ECCS along with the scram action of the RPS. ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment.
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| Four channels of the Drywell Pressure-High Function are required to be OPERABLE when HPCI is required to be OPERABLE
| |
| 'to ensure that no single instrument failure can preclude HPCI initiation. Refer to LC0 3.5.1 for the Applicability Bases for the HPCI System.
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| FERMI - UNIT 2 B 3.3.5.1 - 14 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 3.c. Reactor Vessel Water Level-Hich. Level 8 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to trip the HPCI turbine to prevent overflow into the main steam lines (MSLs). The Reactor Vessel Water Level-High.
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| Level 8 function is not assumed in the accident and transient analyses. It was retained since it is a potentially significant contributor to risk.
| |
| Reactor Vessel Water Level-High, Level 8 signals for HPCI are initiated from two level transmitters from the wide range water level measurement instrumentation. Both Level 8 signals are required in order to trip the HPCI turbine.
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| This ensures that no single instrument failure can preclude HPCI initiation. The Reactor Vessel Water. Level-High, Level 8 Allowable Value is chosen to prevent flow from the HPCI System from overflowing into the MSLs.
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| Two channels of Reactor Vessel Water Level-High, Level 8
| |
| ( Function are required to be OPERABLE only when HPCI is required to be OPERABLE. Refer to LC0 3.5.1 and LC0 3.5.2 for HPCI Applicability Bases.
| |
| 3.d. Condensate Storace Tank Level-Low Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
| |
| Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically own, and then the CST suction valve automatically closes.
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| T11s ensures that an adequate supply of makeup water is available to the HPCI pump. To prevent losing suction to the pump, the suction valves are interlocked so that the
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| . suppression pool suction valves must be open before the CST
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| , suction valve automatically closes. The Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool.
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| O
| |
| -FERMI UNIT 2 B 3.3.5.1 - 15 Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1 t BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Condensate Storage Tank Level-Low signals are initiated from two level transmitters. The logic is arranged such that either level transmitter can cause the suppression pool suction valves to open which, in turn, will cause the CST suction valve to close. The Condensate Storage Tank Level-Low Function Allowable Value is high enough to ensure adequate pump suction head while water is being taken from the CST.
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| Two channels of the Condensate Storage Tank Level-Low Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source.
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| Refer to LC0 3.5.1 for HPCI Applicability Bases.
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| 3.e. Suooression Pool Water Level-Hioh Excessively high suppression pool water could result in the loads on the suppression pool exceeding design values should there be a blowdown of the reactor vessel pressure through the safety / relief valves. Therefore, signals indicating high suppression pool water level are used to transfer the
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| 'j suction source of HPCI from the CST to the suppression pool to eliminate the possibility of HPCI continuing to provide additional water from a source outside containment. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
| |
| This Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool. ;
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| 1 Suppression Pool Water Level-High signals are initiated I from two level transmitters. The logic is arranged such !
| |
| that either transmitter can cause the suppression pool suction valves to open which, in turn, will cause the CST suction valve to close. The Allowable Value for the Suppression Pool Water Level-High Function is chosen to ensure that HPCI will be aligned for suction from the suppression pool before the water level reaches the point at which suppression pool design loads would be exceeded.
| |
| FERMI UNIT 2 B 3.3.5.1 - 16 Amendment No. 134
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| | |
| 1 ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) ;
| |
| 1 Two channels of Suppression Pool Water Level-High Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can oreclude HPCI swap to suppression pool source. Refer to
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| _C0 3.5.1 for HPCI Applicability Bases.
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| 3.f. Manual Initiation 4
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| The Manual Initiation channel provides manual initiation capability by means of individual component controls. There is one manual initiation channel for the HPCI System.
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| The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the HPCI function as required by the NRC in the plant licensing basis.
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| There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the :
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| position of individual controls. The Manual Initiation j Function is required to be OPERABLE only when the HPCI Oy System is required to be OPERABLE. Refer to LC0 3.5.1 for HPCI Applicability Bases.
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| Automatic Deo.tessurization System 4.a. 5.a. Reactor Vessel Water Level-Low Low Low. Level 1 Low RPV water level indicates that the capability to cool ;
| |
| the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore ADS receives one of the signals necessary for initiation from this Function. The Reactor Vessel Water Level-Low Low Low, !
| |
| Level 1 is one of the Functions assumed to be OPERABLE and l capable of initiating the ADS during the accident analyzed ;
| |
| in Reference 1. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| Reactor Vessel Water Level-Low Low Low, Level 1 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual '
| |
| water level (variable leg) in the vessel. Four channels of q(/
| |
| FERMI UNIT 2 B 3.3.5.1 - 17 Amendment No. 134
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| ECCS Instrumentation B 3.3.5.1
| |
| ( BASES APPLICABLE SAFETY ANALYSES LC0. and APPLICABILITY (continued)
| |
| Reactor Vessel Water Level-Low Low Low, Level 1 Function are required to be OPERA 6LE only when ADS is required to be OPERABLE to ensure that no swgle instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
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| The Reactor Vessel Water Level-Low Low Low, Level 1 Allowable Value is chosen to allow time for the low pressure core flooding systems to initiate and provide adequate cooling.
| |
| 4.b. 5.b. Drywell Pressure-Hiah High pressure in the drywell could indicate a break in the RCPB. Therefore, ADS receives one of the signals necessary for initiation from this Function in order to minimize the possibility of fuel damage. The Drywell Pressure-High is assumed to be OPERABLE and capable of initiating the ADS during the accidents analyzed in Reference 1. The core cooling function of the ECCS, along with the scram action of q
| |
| y ,
| |
| the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| Drywell Pressure-High signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.
| |
| Four channels of Drywell Pressure-High Function are only required to be OPERABLE when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels in)ut to ADS trip system B. Refer to LC0 3.5.1 for ADS Applica)ility Bases.
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| 4.c. 5.c. Automatic Deoressurization System Initiation Timer The purpose of the Automatic Depressurization System Initiation Timer is to delay depressurization of the reactor vessel to allow the HPCI System time to maintain reactor vessel water level. Since the rapid depressurization caused by ADS operation is one of the most severe transients on the reactor vessel, its occurrence should be limited. By delaying initiation of the ADS Function, the operator is i
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| O <
| |
| U j FERMI UNIT 2 B 3.3.5.1- 18 Amendment No. 134 I
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| | |
| ECCS Instrumentation B 3.3.5.1
| |
| / BASES APPLICABLE SAFETY ANALYSES, LC0, and APPLICABILITY (continued) given the chance to monitor the success or failure of the HPCI System to maintain water level, and then to decide whether or not to allow ADS to initiate, to delay initiation further by recycling the timer, or to inhibit initiation permanently. The Automatic Depressurization System Initiation Timer Function is assumed to be OPERABLE for the accident analyses of Reference 1 that require ECCS initiation and assume failure of the HPCI System.
| |
| There are two Automatic Depressurization System Initiation Timer relays, one in each of the two ADS trip systems. The i Allowable Value for the Automatic De)ressurization System l Initiation Timer is chosen so that t1ere is still time after I depressurization for the low pressure ECCS subsystems to provide adequate core cooling.
| |
| Two channels of the Automatic Depressurization System Initiation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel r inputs to ADS trip system B. Refer to LC0 3.5.1 for ADS
| |
| \ Applicability Bases.
| |
| 4.d. 5.d. Reactor Vessel Water Level-Low. Level 3 The Reactor Vessel Water Level-Low, Level 3 Function is used by the ADS only as a confirmatory low water level signal . ADS receives one of the signals necessary for initiation from Reactor Vessel Water Level-Low Low Low, Level 1 signals. In order to prevent spurious initiation of the ADS due to spurious Level 1 signals, a Level 3 signal must also be received before ADS initiation commences.
| |
| Reactor Vessel Water Level-Low, Level 3 signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Allowable Value for Reactor Vessel Water Level-Low Level 3 is selected at the RPS Level 3 scram Allowable Value for convenience. Refer to LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," for the Bases discussion of this Function.
| |
| O FERMI - UNIT 2 B 3.3.5.1 - 19 Amendment No. 134
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| | |
| ECCS Instrumentation
| |
| - B 3.3.5.1 BASES APPLICABLE' SAFETY ANALYSES. LCO, and APPLICABILITY (continued)
| |
| Two channels of Reactor Vessel Water Level-Low, Level 3 Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to one string of ADS trip system A, while the other channel inputs to one string of ADS trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases.
| |
| 4.e. 4. f. 5.e. 5. f. Core Sorav and Low Pressure Coolant In.iection Pumo Discharoe Pressure-Hioh The Pump Discharge Pressure-High signals from the CS and LPCI pumps are used as permissives for ADS 'nitiation, indicating that there is a source of low pn ssure cooling water available once the ADS has depressurind the vessel.
| |
| Pump Discharge Pressure-High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in Reference 1 with an assumed HPCI failure. For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the core cooling functions. This core A cooling function of the ECCS, along with the scram action of
| |
| () the RPS ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
| |
| Pump discharge pressure signals are initiated from twelve pressure transmitters, two on the discharge side of each of the four LPCI pumps and one on the discharge of four CS pumps. In order to generate an ADS >ermissive in one trip system, it is necessary that either _PCI pump or both CS pumps in the associated division indicate the high discharge pressure condition. The Pump Discharge Pressure-High Allowable Value is less than the pump discharge pressure when the pump is operating in a full flow mode and high
| |
| . enough to avoid any. condition that re:ults in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running. The actual operating point of this function is not assumed in any transient or accident analysis.
| |
| Twelve channels of Core Spray and Low Pressure Coolant Injection Pump Discharge Pressure-High Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two CS channels, one each associated with CS pumps A and C, and four LPCI channels, O
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| FERMI UNIT 2 B 3.3.5.1 - 20 Amendment No. 134
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| | |
| 1 ECCS Instrumentation B 3.3.5.1 BASES APPLICABLE SAFETY ANALYSES LCO. and APPLICABILITY (continued) two each associated with LPCI pumps A and C, are required for trip system A. Two CS channels, one each associated with CS pumps B and D, and four LPCI channels, two each associated with LPCI pumps B and D, are required for trip system B. Refer to LC0 3.5.1 for ADS Applicability Bases.
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| 4.a. 5.o. Drywell Pressure-Hioh Byoass One of the signals required for ADS initiation is Drywell Pressure-High. However, if the event requiring ADS initiation occurs outside the drywell (e.g., r.eain steam line break outside containment), a high drywell pressure signal may never be present. Therefore, the Drywell Pressure-High Bypass Timer is used to bypass the Drywell Pressure-High Function after a certain time period has elapsed. Operation of the Drywell Pressure High Bypass Timer Function is not assumed in any accident analysis. The instrumentation is retained in the TS because ADS is part of the primary success path for mitigation of a DBA.
| |
| There are four Drywell Pressure-High Bypass Timer relays.
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| two in each of the two ADS trip systems. The Allowable O Value for the Drywell Pressure-High Bypass Timer is chosen to ensure that there is still time after depressurization i
| |
| for the low pressure ECCS subsystems to provide adequate l core cooling. ]
| |
| Four channels of the Drywell Pressure-High Bypass Timer Function are only required to be OPERABLE when the ADS is required to-be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Refer to LC0 3.5.1 for ADS Applicability Bases.
| |
| 4.h. 5.h Manual Inhibit The Manual Inhibit switch channels allow the operator to inhibit ADS operation without repeatedly pressing the ADS timer-reset pushbuttons. There is one switch for each ADS trip system for a total of two. ;
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| The Manual Inhibit Function is not assumed in any accident l of transient analysis in the UFSAR. however, the operator is !
| |
| I required to manually inhibit the ADS function in an
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| -Anticipated Transient Without Scram (ATWS) situation. The l
| |
| Manual Inhibit Function is retained because the function was i determined to be risk significant. The estimated increase !
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| FERMI. UNIT 2 B 3.3.5.1 - 21 Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1 h BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) in the core damage frequency if the ADS inhibit function is always assumed to fail is more than a factor of four. Thus this function meets criterion 4 of 10 CFR 50.36(c)(2)(ii).
| |
| There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the switch. Two channels of the Manual Inhibit Function (one channel per trip system) are only required to be OPERABLE when the ADS is required to be OPERABLE. Refer to LC0 3.5.1 for the ADS Applicability Bases.
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| 4.1. 5.1. Manual Initiation The Manual Initiation channels provide manual initiation capability to each ADS valve.
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| The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the ADS functions as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of individual controls. The Manual Initiation
| |
| : unction is only required to be OPERABLE when the ADS is required to be OPERABLE. Refer to LCO 3.5.1 for ADS Applicability Bases.
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| ACTIONS A Note has been provided to modify the ACTIONS related to ECCS instrumentation channels. Section 1.3. Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Recuired Actions of the Condition continue to apply for each adcitional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel.
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| O
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| . FERMI - UNIT 2 B 3.3.5.1 - 22 Amendment No. 134 I
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| ECCS Instrumentation B 3.3.5.1 O
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| u BASES ACTIONS (continued)
| |
| L1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the table is Function dependent.
| |
| Each time a channel is discovered inoperable Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
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| B.1. B.2. and B.3 Required Actions B.1 and 8.2 are intended to ensure that appropriate actions are taken if multiple.. inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action B.1 features would be those that are initiated by Functions 1.a. 1.b. 2.a. 2.b, 2.d. and 2.g (e.g., low pressure ECCS). The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic initiation capability is lost if (a) two Function 1.a channels are inoperable and untripped in the f same trip system, (b) two Function 2.a channels are inoperable and untripped in the same trip system, (c) two Function 1.b channels are inoperable and untripped in the same system, (d) two Function 2.b channels are inoperable and untripped in the same trip system (e) two Function 2.d channels are inoperable and untripped in the same trip system, or (f) two Function 2.g channels are inoperable and untripped in the same trip system. For low pressure ECCS.
| |
| since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each ino wrable channel would only require the affected portion of t1e associated system of low pressure ECCS and EDGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and EDGs being concurrently declared inoperable.
| |
| For Required Action B.2, redundant automatic initiation capability is lost if two Function 3.a or two Function 3.b channels are inoperable and untripped in the same trip system. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required O
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| FERMI UNIT 2 B 3.3.5.1-23 Amendment No. 134 i
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| L
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| i ECCS Instrumentation B 3.3.5.1
| |
| ' BASES ACTIONS (continued) 1 I
| |
| Action B.3 is not appropriate and the feature (s) associated with the inoperable, untripped channels must be declared inoperable within 1 hour. As noted (Note 1 to Required Action B.1), Required Action B.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5. the specific .
| |
| initiation time of the low pressure ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of initiation capability for 24 hours (as allowed by Required Action B.3) is allowed during MODES 4 and 5. There is no similar Note provided for Required Action B.2 since HPCI instrumentation is not required in MODES 4 and 5: thus, a Note is not necessary.
| |
| Notes are also provided (Note 2 to Required Action B.1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable.
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| This ensures that the proper loss of initiation capability ;
| |
| check is performed.
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| The Completion Time is intended to allow the operator time A to evaluate and repair any discovered inoperabilities. This V Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock." i For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above. For Required Action B.2. the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function in the same trip system. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.3. Placing the inoperable channel in trip would conservatively compensate
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| {
| |
| FERMI UNIT 2 B 3.3.5.1-24 Amendment No. 134 )
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| | |
| ECCS Instrumentation B 3.3.5.1 h
| |
| .J BASES-
| |
| \
| |
| ACTIONS (continued) for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
| |
| Alternately, if it is not desired to place the channel in trip (e.g. as in the case where placing the inoperable channel in trip would result in an initiation). Condition G must be entered and its Required Action taken. 1 C.1 and C.2 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function result in redundant automatic initiation capability being lost for the feature (s). Required Action C.1 features would be those that are initiated by Functions 1.c, 2.c, 2.e. and 2.f (i.e., low pressure ECCS).
| |
| Redundant automatic initiation capability is lost if either l
| |
| -(a) two Function 1.c channels are inoperable in the same trip system, (b) two Function 2.c channels are inoperable in the same trip system, (c) two Function 2.e channels are inoperable in the same trip system, or (d) two or more Function 2.f channels are inoperable. In this situation (loss of redundant automatic initiation capability), the s 24 hour allowance of Required Action C.2 is not appropriate
| |
| ' and the feature (s) associated with the inoperable channels must be declared inoperable within 1 hour. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each ino>erable channel would only require the affected portion of t1e associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g.,
| |
| both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable.. For Functions 1.c, 2.c, 2.e. and 2.f. the affected portions are the associated low pressure ECCS pumps. As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5.
| |
| the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours (as allowed by Required Action C.2) is allowed during MODES 4 and 5.
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| FERMI UNIT 2 B 3.3.5.1 - 25 Amendment No. 134
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| | |
| ECCS Instrumentation i B 3.3.5.1 m
| |
| Q BASES ACTIONS (continued)
| |
| Note 2 states that Required Action C.1 is only applicable for Functions 1.c. 2.c. 2.e. and 2.f. Required Action C.1 is not applicable to Functions 1.d. 2.h. and 3.f (which also require entry into this Condition if a channel in these Functions is inoperable), since they are the Manual Initiation Functions and are not assumed in any accident or trar.sient analysis. Thus, a total loss of manual initiation capability for 24 hours (as allowed by Required Action C.2) is allowed. Required Action C.1 is also not a)plicable to Function 3.c (which also requires entry into t11s Condition if a channel in this Function 1s inoperable), since the loss of one channel results in a loss of the Function (two out of-two logic). This loss was considered during the development of Reference 4 and considered acceptable for the 24 hours allowed by Required Action C.2.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock."
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| For Required Action C.1. the Completion Time only begins upon discovery that the same feature in both subsystems
| |
| (
| |
| (N) (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time. Condition G must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events.
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| 1 O
| |
| FERMI - UNIT 2 B 3.3.5.1 - 26 Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1
| |
| -BASES ACTIONS (continued)
| |
| D.1. D.2.1. and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic component initiation capability for the HPCI System. Automatic component initiation capability is lost if two function 3.d channels or two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within 1 hour after discovery of loss of HPCI initiation capability. As noted.
| |
| Required Action D.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the Function is already performed.
| |
| The Completion Time is intended to allow the operator time to evaluate.and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock."
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| O For Required Action D.1. the Completion Time only begins V upon discovery that the HPCI System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while :
| |
| allowing time for restoration or tripping of channels. ]
| |
| 1 Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable !
| |
| channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per Required Action D.2.2.. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to 1
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| O V
| |
| FERMI' UNIT 2 B 3.3.5.1 - 27 . Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS (continued) continue. -If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the HPCI System piping remains filled with water. Alternately, if it is not desired to perform Required Actions D.2.1 and 0.2.2 (e.g.,
| |
| as in the case where shifting the suction source could drain down the HPCI suction piping), Condition G must be entered and its Required Action taken.
| |
| E.1 and E.2 Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one Function 4.a channel. and one Function 5.a channel are inoperable and untripped (b) one Function 4.b channel and one Function 5.b channel are )
| |
| inoperable and untripped, or .(c) one Function 4.d channel and one Function 5.d channel are inoperable and untripped.
| |
| In this situation (loss of automatic initiation capability).
| |
| O- '
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| the % hour or 8 day allowance, as applicable, of Required Action E.2 is not appropriate and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock."
| |
| For Required Action E.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the i paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or l tripping of channels.
| |
| Because of the diversity of sensors available to provide ;
| |
| initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable i channel to OPERABLE status if both HPCI and RCIC are OPERABLE. If either HPCI or RCIC is inoperable, the time is O
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| : FERMI UNIT 2 B 3.3.5.1 - 28 Amendment No. 134
| |
| | |
| p y ECCS Instrumentation B 3.3.5.1 G BASES b -ACTIONS (continued) shortened to % hours. If the status of HPCI or RCIC j
| |
| ~ changes such that the Completion Time changes from 8 days to l 96 hours, the % hours begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the " time zero" for beginning the 8 day " clock" begins upon discovery of the inoperable, untripped channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action E.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if i it is not desired to place the channel in trip (e.g., as in ;
| |
| the case where placing the inoperable channel in trip would l result in an initiation) Condition G must be entered and !
| |
| its Required Action taken. j F.1 and F.2 RequiredActionF.1isintendedtoensurethata$propriate actions are taken if multiple, inoperable channe s within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS. Automatic !
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| initiation capability is lost if either (a) one Function 4.c l channel and one Function 5.c channel are inoperable. (b) a i combination of Function 4.e. 4.f, 5.e. and 5.f chanr.els are inoperable such that channels associated with seven or more low pressure ECCS pumps are inoperable, or (c) one or more Function 4.g channels and one or more Function 5.9 channels
| |
| -are inoperable.
| |
| In this situation (loss of automatic initiation capability), i the 96 hour or 8 day allowance, as applicable, of Required l Action F.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability. The Note to Required Action F.1 states that Required Action F.1 is only ap.olicable for 4 Functions 4.c, 4.e. 4.f 4.g. 5.c. 5.e. 5.f. and 5.g.
| |
| Required Action F.1 is not applicable to Functions 4.h. 4.i.
| |
| 5.h. and 5.1 (which also require entry into this Condition if a channel in these Functions is inoperable), since they are the Manual Inhibit and Manual Initiation Functions and are not assumed in any accident or transient analyris.
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| FERMI UNIT-2 B 3.3.5.1 -29 Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS (continued)
| |
| Thus, a total loss of manual inhibit capability for % hours or 8 days (as allowed by Required Action F.2) is allowed.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock."
| |
| For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an 1 allowable out of service time of 8 days has been shown to be I acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status if t'oth HPCI and RCIC are OPERABLE (Required Action F.2). If either HPCI or RCIC is p inoperable, the time shortens to % hours. If the status of
| |
| .HPCI or RCIC changes such that the Com)1etion Time changes
| |
| 'd from 8 days to 96 hours, the % hours )egins upon discovery of HPCI or RCIC inoperability. However, the total time for .
| |
| an inoperable channel cannot exceed 8 days. If the status !
| |
| of HPCI or RCIC changes such that the Completion Time J changes from % hours to 8 days, the " time zero" for j beginning the 8 day " clock" begins upon discovery of the -
| |
| inoperable channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition G must be entered and its Required i Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.
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| L1 With any Required Action and associated Completion Time not met, the associated feature (s) may be incapable of performing the-intended function, and the supported feature (s) associated with inoperable untripped channels must be declared inoperable immediately.
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| O FERMI UNIT 2 B 3.3.5.1-30 Amendment No. 134
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| | |
| ECCS Instrumentation B 3.3.5.1
| |
| . BASES SURVEILLANCE As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.1-1.
| |
| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Function 3.c: and (b) for Functions other than 3.c and 3.f provided the associated Function or redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.
| |
| j SR 3.3.5.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter ;
| |
| indicated on one channel to a similar parameter on other -
| |
| channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between-the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited to 12 hours: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria it may be an indication that the
| |
| -instrument has drifted outside its limit.
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| l O
| |
| ' FERMI UNIT 2 B 3.3.5.1 - 31 Amendment No. 134 ;
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| J
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| | |
| ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| The Frequency is based upon operating ex)erience that demonstrates channel failure is rare. T1e CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.5.1.2 and SR 3.3.5.1.6 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of ;
| |
| the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
| |
| Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint i methodology.
| |
| The Frecucr,cy of 92 days for SR 3.3.5.1.2 is based on the 4 reliability analyses of Reference 4. The Frequency of l 18 months for SR 3.3.5.1.6 is based on engineering judgement I and the reliability of the components. !
| |
| l SR 3.3.5.1.3 1 i
| |
| This surveillance provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the ;
| |
| Allowable Value specified in Table 3.3.5.1-1. If the trip '
| |
| setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not i beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analyses. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology.
| |
| The Frequency of 92 days is based on the reliability analysis of Reference 4.
| |
| FERMI - UNIT 2 B 3.3.5.1-32 Amendment No. 134
| |
| | |
| [ j ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.5.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive i calibrations consistent with the plant specific setpoint i methodology. j I
| |
| The Frequency of SR 3.3.5.1.4 is based upon the assumption of a = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. l SR 3.3.5.1.5 l The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the ;
| |
| OPERABIF.ITY of the required initiation logic for a specific !
| |
| channel. The system functional testing performed in l LCO 3.5.1, LC0 3.5.2. LC0 3.8.1, and LC0 3.8.2 overlaps this l Surveillance to complete testing of the assumed safety 1 function.
| |
| O The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
| |
| Operating experience has shown that these components usually
| |
| ) ass the Surveillance when performed at the 18 month requency.
| |
| REFERENCES 1. UFSAR, Section 6.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. NEDC 31982 P, " SAFER /GESTR LOCA, Loss of Coolant Accident Analysis, including Errata and Addenda No.1,"
| |
| April 1992.
| |
| : 4. NEDC 30936 P A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2 " December 1988.
| |
| FERMI UNIT 2 B 3.3.5.1 - 33 Amendment No. 134
| |
| | |
| F RCIC System Instrumentation B 3.3.5.2 (D B 3.3 INSTRUMENTATION G
| |
| B 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation BASES BACKGROUND The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is unavailable, such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LC0 3.5.3, "RCIC System."
| |
| The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of reactor vessel Low Low water level. The variable is monitored by four transmitters that are connected to four trip units. The outputs of the trip units are connected to relays whose contacts are arranged in a one-out-of two taken twice logic arrangement. Once initiated, the RCIC logic p seals in and can be reset by the operator only when the g reactor vessel water level signals have cleared.
| |
| The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow and maintain primary containment isolated in the event RCIC is not operating.
| |
| The RCIC System also monitors the water level in the condensate storage tank (CST) since this is the normal source of water for RCIC operation. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position) unless the ) ump suction from the suppression pool valves is open. If t1e water level in the CST falls below a 1 preselected level, first the suppression pool suction valves automatically onen, and then the CST suction valve automatic 5ilycloses. Two channels of transmitter / trip unit in the HPCI system are used to detect low water level in the i CST. Either channel can cause the suppression pool suction l valves to open and the CST suction valve to close. To l prevent losing suction to the pump, the suction valves are i interlocked so that one suction path must be open before the i other automatically closes.
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| FERMI UNIT 2 B 3.3.5.2 - 1 Amendment No. 134
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| | |
| o RCIC System Instrumentation B 3.3.5.2 BASES BACKGROUND (continued).
| |
| c The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level L (Level 8) trip (two out-of two logic), at which time the RCIC steam supply, steam supply bypass, and cooling water supply valves close (the injection valve also closes due to the closure of the steam supply valves). The RCIC System restarts if vessel level again drops to the~ low level initiation point (Level 2).
| |
| APPLICABLE The function of the RCIC System to provide makeup coolant to SAFETY ANALYSES, the reactor is used to respond to transient events. The LCO,' and RCIC System is not an Engineered Safety Feature System and APPLICABILITY no credit is taken in the safety analyses for RCIC System operation. Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation, are included in the Technical Specifications as required by 10 CFR 50.36(c)(2)(ii).
| |
| Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
| |
| The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.2 1. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual tri) setpoint is not within its required Allowable Value. T1e actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
| |
| Allowable Values are specified for each RCIC System j instrumentation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations, i The nominal setpoints are selected to ensure that the I setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS or between successive verifications of trip unit.setpoints. Operation with a trip setpoint less conservative than the nominal tria setpoint, but within its Allowable Value, is acceptable. Iach Allowable Value specified accounts for instrument uncertainties appropriate :
| |
| .to the Function. These uncertainties are described in the setpoint methodology.
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| O FERMI - UNIT 2 B 3.3.5.2 - 2 Amendment No. 134
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| m .
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| RCIC System Instrumentation l B 3.3.5.2 ;
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| O v
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| BASES l 1
| |
| APPLICABLE SAFETY ANALYSES. LC0. and APPLICABILITY (continued) i The individual Functions are required to be OPERABLE in MODE 1. and in MODES 2 and 3 with reactor steam dome
| |
| )ressure > 150 psig since this is when RCIC is required to
| |
| >e OPERABLE.. (Refer to LC0 3.5.3 for Applicability Bases for the RCIC System.) !
| |
| The specific Applicable Safety Analyses. LCO. and Applicability discussions are listed below on a Function by i Function basis. 1
| |
| : 1. Reactor Vessel Water Level-Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that i normal feedwater flow is insufficient to maintain reactor i vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far.
| |
| fuel damage could result. Therefore, the RCIC System is !
| |
| initiated at Level 2 to assist in maintaining water level ;
| |
| above the top of the active fuel. l l
| |
| Reactor Vessel Water Level-Low Low Level 2 signals are p initiated from four level transmitters that sense the Q difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
| |
| The Reactor Vessel Water Level-Low Low. Level 2 Allowable Value is set high enough such that for com)lete loss of feedwater flow, the RCIC System flow with ligh pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1.
| |
| Four channels of Reactor Vessel Water Level-Low Low.
| |
| Level 2 Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.
| |
| Refer to LC0 3.5.3 for RCIC Applicability Bases.
| |
| 2 Reactor Vessel Water Level-Hich. Level 8 High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam supply, steam supply bypass. and cooling water supply valves to prevent overflow into the (3
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| V FERMI - UNIT 2 B 3.3.5.2 - 3 Amendment No. 134
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| | |
| r RCIC System Instrumentation B 3.3.5.2
| |
| (] BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) main steam lines (MSLs). (The injection valve also closes due to the closure of the steam supply valve.) '
| |
| Reactor Vessel Water Level-High, Level 8 signals for RCIC are initiated from two level transmitters from the wide range water level measurement instrumentation, which sense .
| |
| the difference between the pressure due to a constant column i of water (reference leg) and the pressure due'to the actual i water level (variable leg) in the vessel. '
| |
| The Reactor Vessel Water Level-High. Level 8 Allowable Value is high enough to preclude isolating the injection !
| |
| valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.
| |
| Two channels of Reactor Vessel Water Level-High. Level 8 i Function are available and are required to be OPERABLE when l RCIC is required to be OPERABLE to ensure that no single !
| |
| instrument failure can preclude RCIC initiation. Refer to i LCO 3.5.3 for RCIC Applicability Bases. l A
| |
| V 3. Condensate Storace Tank Level-Low Low level in the CST indicates the unavailability of an adequate su) ply of makeup water from this normal source.
| |
| Normally, t1e suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST.
| |
| However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve (consistency) automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
| |
| Two level transmitters and trip units in the HPCI system are used to detect low water level in the CST. The Condensate Storage Tank Level-Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST.
| |
| FERMI - UNIT 2 B 3.3.5.2 -4 Amendment No. 134 l l
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| I I
| |
| | |
| r ,
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| I i
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| RCIC System Instrumentation l B 3.3.5.2 ;
| |
| 1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Two channels of Condensate Storage Tank Level-Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLi to ensure that no single instrument failure can preclude RCIC swap to suppression pool source.
| |
| Refer to LC0 3.5.3 for RCIC Applicability Bases.
| |
| : 4. Manual Initiation l
| |
| )
| |
| The Manual Initiation channel provides manual initiation capability to individual valves. There is one manual initiation channel for the RCIC System.
| |
| The Manual Initiation Function is not assumed in any accident or transient analyses in the UFSAR. However, the Function is retained for overall redundancy and diversity of the RCIC function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function since the channel is mechanically actuated based solely on the position of the valve control. One channel of Manual Initiation is required to be OPERABLE when RCIC is required p)
| |
| ( to be OPERABLE.
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. Section 1.3.
| |
| Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.
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| O.
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| U a FERMI UNIT 2 B 3.3.5.2 - 5 Amendment No. 134 l
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| I
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| | |
| E RCIC System Instrumentation B 3.3.5.2 BASES ACTIONS (continued) eG1 Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is function dependent.
| |
| Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
| |
| B.1 and B.2 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC Systein. In this case, automatic initiation capability is lost if two Function 1 channels in the same trip. system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour after discovery of loss of RCIC initiation capability.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time.also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock." ,
| |
| For Required Action B.1, the Completion Time only begins ,
| |
| upon discovery that the RCIC System cannot be automatically !
| |
| initiated due to two inoperable, untripped Reactor Vessel Water Level-Low Low. Level 2 channels in the same trip system. The 1 hour. Completion Time from discovery of loss >
| |
| of initiation capability is acceptable because it minimizes .
| |
| risk while allowing time for restoration or tripping of !
| |
| channels.
| |
| Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref.1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable oct of service time, the channel must be placed in the l tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate O
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| FERMI - UNIT 2. B 3.3.5.2-6 Amendment No. 134
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| | |
| RCIC System Instrumentation B 3.3.5.2
| |
| (]
| |
| V BASES ACTIONS (continued) for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
| |
| Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.
| |
| L.1 A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref.1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level-High, Level 8 Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be p) 1 V
| |
| acceptable. This Condition also applies to the Manual Ir.itiation Function. Since this Function is not assumed in any accident or transient analysis, a total loss of manual initiation caoability (Required Action C.1) for 24 hours is allowed. The' Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events.
| |
| D.1. D.2.1. and D.2.2 l l
| |
| Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature (s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic initiation capability is lost if two Function 3 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed.
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| FERMI - UNIT 2 B 3.3.5.2 - 7 Amendment No. 134
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| F l i
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| RCIC System Instrumentation B 3.3.5.2 l
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| /7 BASES V
| |
| ACTIONS (continued) l The Completion Time is intended to allow the operator time I to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal
| |
| " time zero" for beginning the allowed outage time " clock."
| |
| For Required Action D.1. the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while l allowing time for restoration or tripping of channels.
| |
| Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable j out of service time of 24 hours has been shown to be i acceptable (Ref.1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel l cannot be restored to OPERABLE status within the allowable l out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 which performs p the intended function of the channel (shifting the suction
| |
| (-) source to the suppression pool). Alternatively. Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken.
| |
| L1 With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.
| |
| p k
| |
| FERMI - UNIT 2 B 3.3.5.2 - 8 Amendment No. 134
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| | |
| RCIC System Instrumentation B 3.3.5.2
| |
| (] BASES SURVEILLANCE As noted in the beginning of the SRs. the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.2 1.
| |
| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:
| |
| (a) for up to 6 hours for Function 2: and (b) for up to 6 hours for Functions 1 and 3 provided the associated Function maintains RCIC initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref.1) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.
| |
| SR 3.3.5.2.1 q Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A Q CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the i instrumentation continues to operate properly between each l CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based i on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| The Frequency is based upon operating experience that
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| ' demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LC0.
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| O FERMI UNIT 2 83.3.5.2-9 Amendment No. 134
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| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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| SR 3.3.5.2.2 and SR 3.3.5.2.6 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all l of the other required contacts of the relay are verified by J other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
| |
| Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint I methodology.
| |
| The Frequency of 92 days for SR 3.3.5.2.2 is based on the reliability analysis of Reference 1. The Frequency of 18 months for SR 3.3.5.2.6 is based on engineering judgement
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| , and the reliability of the components.
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| U SR 3.3.5.2.3 This surveillance provides a check of the actual trip !
| |
| setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.5.21. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
| |
| The Frequency of 92 days is based on the reliability analysis of o aference 1.
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| fN O
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| FERMI - UNIT 2 B 3.3.5.2 - 10 Amendment No. 134
| |
| | |
| RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.5.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive I calibrations consistent with the plant specific setpoint methodology.
| |
| The Frequency of SR 3.3.5.2.4 is based upon the assumption of a = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| SR 3.3.5.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in ,
| |
| LC0 3.5.3 overlaps this Surveillance to provide complete I testing of the safety function.
| |
| m U The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
| |
| Operating experience has shown that these components usually 3 ass the Surveillance when performed at the 18 month requency.
| |
| REFERENCES 1. Safety Evaluation Re) ort for Fermi Unit-2 Amendment No. 75, dated Septem)er 6. 1991.
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| (
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| V]
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| FERMI --UNIT 2 B 3.3.5.2 - 11 Amendment No. 134
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| | |
| Pricary Containment Isolation Instrumentation B 3.3.6.1 B 3.3 INSTRUMENTATION
| |
| 'B 3.3.6.1< Primary Containment Isolation Instrumentation l
| |
| BASES BACKGROUND The primary containment isolation instrumentation automatically initiates closure of appro)riate primary containment isolation valves (PCIVs). T1e function of the PCIVs in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the. analyses for a DBA.
| |
| The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Most channels include electronic L equipment (e.g., trip units) that compares measured input l signals with pre established setpoints. When the setpoint l is exceeded, the channel output relay actuates, which then ;
| |
| outputs a primary containment isolation signal to the I isolation logic. Functional diversity is provided by !
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| monitoring a wide range of independent parameters. The i
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| . input-parameters to the isolation logics are (a) reactor '
| |
| vessel water level, (b) area ambient and differential
| |
| ~
| |
| temperatures, (c) main steam line (MSL) flow and radiation.
| |
| (d). Standby Liquid Control (SLC) System initiation, (e) condenser pressure (f) main steam line pressure.
| |
| (g) high pressure coolant injection (HPCI) and reactor core !
| |
| isolation cooling (RCIC) steam line flow. (h) drywell l pressure (1) HPCI and RCIC steam line pressure, (j) HPCI and RCIC turbine exhaust diaphragm pressure, (k) reactor water cleanup (RWCU) differential flow, and (1) reactor steam dome pressure. Redundant sensor input signals from each parameter are typically provided for initiation of isolation. The only exceptions are SLC System initiation and RWCU differential flow. In addition, manual isolation of the valves is provided.
| |
| Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below.
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| ,b a
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| FERMI' UNIT 2 B 3.3.6.1- 1 Amendment No. 134
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| | |
| Prigary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND (continued)
| |
| : 1. Main Steam Line Isolation Most MSL Isolation Functions receive inputs from four channels. The outputs from these channels are combined in a one out of two taken twice logic to initiate isolation of all main steam isolation valves (MSIVs). The outputs from the same channels are arranged into two two out-of-two logic trip systems to isolate the two MSL drain valves at the containment boundary.
| |
| 'The exceptions to this arrangement are the Main Steam Line Flow-High Function and Area Temperature Functions. The Main Steam Line Flow-High Function uses 16 flow channels, four for each steam line. One channel from each steam line inputs to one of the four trip strings logic. Two trip strings logics make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string logic has four inputs (one per MSL), any one of which will trip the trip string logic. Either trip string logic can trip the trip system. This is effectively a one out of-eight taken twice logic arrangement to initiate n isolation of the MSIVs. Similarly, contacts from the same U four trip string logic output relays are connected into two two out of-two logic trip systems with each trip system isolating one of the two MSL drain valves.
| |
| The Main Steam Tunnel Temperature-High Function receives input from 16 channels. The logic is arranged similarly to the Main Steam Line Flow-High Function. The Turbine Building Area Temperature-High Function receives input from 8 channels. The inputs are arranged in a one out of two trip string logic, with two trip string logics per trip system. All MSIVs will close on one-out of two taken twice logic from the two trip systems. Similarly, contacts from the same four trip string logic output relays are arranged in two out-of-two logic that requires both trip systems to trip each of two MSL drain valves. Therefore, a Turbine Building Area Temperature-High condition sensed by at least one sensor in trip System A and at least one sensor in Trip System B will cause at least one of two MSL drain valves to close. At least four sensors input to four trip string logics must sense high temperature to close both MSL drain valves. {
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| l 10 i
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| I FERMI - UNIT 2 B 3.3.6.1- 2 Amendment No. 134
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| T Prirary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND (continued)
| |
| MSL Isolation Functions isolate the MSL and MSL drain isolation valves. The MSL Radiation-High Function also isolates the Reactor Water Sample System.
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| 1
| |
| : 2. Primary Containment Isolation l Primary Containment Isolation Functions receive inputs from four channels. The outputs from these channels are arranged into two two out of two logic trip systems. One trip system initiates isolation of all inboard primary containment isolation valves, while the other trip system initiates isolation of all outboard primary containment isolation ,
| |
| valves. Each logic closes one of the two valves on each 1 penetration, so that operation of either logic isolates the penetration.
| |
| 1 Primary Containment Isolation Drywell Pressure-High and Reactor Vessel Water Level-Low, Level 3 Functions isolate lines in the drywell sumps and traversing in core probe systems.
| |
| Primary Containment Isolation Drywell Pressure-High and (S '
| |
| ) Reactor Vessel Water Level-Low Low, Level 2 Functions isolate lines in the Reactor Water Samale Torus Water Management, Standby Gas Treatment, Com)ustible Gas Control, Nitrogen Inerting, and Primary Containment Monitoring Systems.
| |
| Primary Containment Isolation Drywell Pressure-High also affects isolation of lines in the RHR, CS, HPCI and RCIC systems. Primary Containment Isolation Reactor Vessel Water Level-Low Low, Level 2 Function also affects isolation of the Recirculation Pump Seal System and the Primary Containment Pneumatic Supply System.
| |
| : 3. and 4. Hiah Pressure Coolant In.iection System Isolation and Reactor Core Isolation Coolina System Isolation Most Functions that isolate HPCI and RCIC receive input from two channels, with each channel in one trip system using a' one out-of one logic. Each of the two trip systems in each isolation group is connected to one of the two valves on each associated penetration.
| |
| The exceptions are the HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High, Steam Supply Line Pressure-Low.
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| O (v .
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| FERMI UNIT 2 B 3.3.6.1 - 3 Amendment No. 134
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| Prtary Containment Isolation Instrumentation
| |
| ' B 3.3.6.1
| |
| (] BASES BACKGROUND (continued) and Drywell Pressure-High functions. These Functions receive inputs from four turbine exhaust diaphragm pressure and four steam sup)1y pressure channels for each system.
| |
| The outputs from tie turbine exhaust diaphragm pressure and steam supply pressure channels are each connected to two two out of two trip systems. Each trip system isolates one )
| |
| valve per associated penetration. 1 HPCI and RCIC Functions isolate the HPCI and RCIC isolation valves.
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| : 5. Reactor Water Cleanuo System Isolation The Reactor Vessel Water Level-Low Low, Level 2 Isolation Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into two two out of-two trip systems.
| |
| The Differential F1ow-High function is derived from three non redundant flow transmitters and a non redundant flow summer. The outaut of the summer is fed to two trip units, n the outputs of w11ch are channeled through relays into two l
| |
| () trip systems. One trip system isolates the inboard isolation valve, while the other trip system isolates the two outboard isolation valves.
| |
| SLC System Initiation Functions receive input from two channels, with each channel in one trip system using a one out of-one logic. Both channels are only input to the trip systems that isolates the outboard isolation valves.
| |
| The Area Temperature-High Function receives input from twelve temperature monitors, six to each trip system. The Area Ventilation Differential Temperature-High Function 1 receives input from four differential temperature monitors, two in each trip system. These are configured so that any one input will trip the associated trip system. One of the two trip systems is connected to the inboard valve and the other trip system is connected to the two outboard valves on each RWCU penetration.
| |
| RWCU Functions isolate the RWCU isolation valves.
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| (\
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| %-)
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| FERMI UNIT 2 B 3.3.6.1 -4 Amendment No. 134
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| | |
| a PrisIary Containment Isolation Instrumentation B 3.3.6.1 BASES BACKGROUND (continued)
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| : 6. Shutdown Coolina System Isolation The Reactor Vessel Water Level-Low, Level 3 Function receives input from four reactor vessel water level channels.. The outputs from the reactor vessel water level channels are connected to two two out of two trip systems.
| |
| The Reactor Vessel Pressure-High Function receives input from two channels, with each channel in one trip system-using a'one out of-one logic. Each of the two trip systems is connected to one of the two valves on each shutdown cooling penetration.
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| Shutdown Cooling System Isolation Functions isolate the ;
| |
| shutdown cooling isolation valves. J APPLICABLE The isolation signals generated by the primary containment 1 isolation instrumentation are implicitly assumed in the SAFETY ANALYSES, LCO, and safety analyses of References 1 and 2.to initiate closure APPLICABILITY of valves to limit offsite doses. Refer to LC0 3.6.1.3.
| |
| " Primary Containment Isolation Valves (PCIVs)," Applicable j O Safety Analyses Bases for more detail of the safety analyses.
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| 3 I
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| l Primary containment isolation instrumentation satisfies !
| |
| Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
| |
| The OPERABILITY of the primary containment instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.6.1 1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual tri) setpoint is not within its required Allowable Value. T1e actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
| |
| Each channel must also respond within its assumed response time, where appropriate.
| |
| Allowable Values are specified for each Primary Containment Isolation Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints j I
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| FERMI UNIT ?. B 3.3.6.1- 5 Amendment No. 134 J
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| Pricary Containment Isolation Instrumentation l L B 3.3.6.1 i
| |
| BASES l APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) i do not exceed the Allowable Value between CHANNEL i CALIBRATIONS or betw' e en successive verifications of trip unit setpoints. Operation with a trip setpoint less L conservative than the nominal trip setpoint, but within its l Allowable Value, is acceptable. Trip setpoints are those l
| |
| predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g..' reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from i E the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and l~
| |
| some of tne instrument errors. The trip setpoints are then i determined accounting for the remaining instrument errors l (e.g., drift). The trip set)oints derived in this manner ,
| |
| l provide adequate protection )ecause instrumentation I uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
| |
| Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g. minimum flow) also serve the dual function of automatic PCIVs. The signals that isolate these valves are '
| |
| also associated with the automatic initiation of the ECCS and RCIC. The instrumentation requirements and ACTIONS associated with these signals are addressed in LC0 3.3.5.1,
| |
| " Emergency Core Cooling Systems (ECCS) Instrumentation," and LC0 3.3.5.2. " Reactor Core Isolation Cooling (RCIC) System l
| |
| l Instrumentation." and are not included in this LCO. ;
| |
| In general, the individual Functions are required to be OPERABLE in MODES 1. 2. and 3 consistent with the Applicability for LC0 3.6.1.1, " Primary Containment."
| |
| Functions that have different Applicabilities are discussed i below in the individual Functions discussion. ;
| |
| The specific Applicable Safety Analyses, LCO. and !
| |
| Applicability discussions are listed below on a Function by Function basis.
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| l l
| |
| O FERMI . UNIT 2 B 3.3.6.1-6 Amendment No. 134 y
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| s
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| r i
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| Prirary Containment Isolation Instrumentation B 3.3.6.1 BASES l
| |
| APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Main Steam Line Isolation 1.a. Reactor Vessel Water Level-Low Low Low. Level 1 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
| |
| Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level-Low Low I (.ow, Level 1 Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals. ,
| |
| The Reactor Vessel Water Level-Low Low Low, Level 1 Function associated with isolation is assumed in the analysis of the recirculation line break (Ref.1). The isolation of the MSLs on Level 1 supports actions to ensure that offsite dose limits are not exceeded for a DBA.
| |
| Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) p and the pressure due to the actual water level (variable Q leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low Low, Level 1 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Reactor Vessel Water Level-Low Low Low, Level 1 Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LC0 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 100 limits.
| |
| This Function isolates the MSL and MSL drains isolation valves.
| |
| 1.b. Main Steam Line Pressure-Low Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low ,
| |
| reactor vessel water level condition and the RPV cooling i down more than 100*F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure-Low Function, although not credited in the analysis of the pressure regulator failure (Ref. 2), is a back up to the maximum steam flow limiter, which is credited by this analysis. For j j
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| FERMI - UNIT 2 B 3.3.6.1 - 7 Amendment No. 134 i
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| i Pricary Containment Isolation Instrumentation l B 3.3.6.1 BASES
| |
| . APPLICABLE SAFETY ANALYSES,'LCO, and APPLICABILITY (continued)
| |
| ]
| |
| this event, the closure of the MSIVs ensures that the RPV temperature change limit (100*F/hr) is not reached.
| |
| 'The MSL low pressure signals are initiated from four i transmitters that are connected to the MSL header The !
| |
| transmitters are arranged such that, even though physically '
| |
| separated from each other, each transmitter is able to detect low MSL pressure. Four channels of Main Steam Line Pressure-Low Function are.available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.
| |
| The Main Steam Line Pressure-Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref. 2).
| |
| This Function isolates the MSL nnd MSL drains isolation valves.
| |
| O. 1 c- a 4# ste '4"e e, -84 8 Main Steam Line Flow-High is provided to detect a break of the MSL and to initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow-High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 2). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits.
| |
| J The MSL flow signals are initiated from 16 transmitters that !
| |
| are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of Main Steam Line ,
| |
| Flow-High Function for each unisolated MSL (two channels per trip system) are available and are required to be O
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| FERMI UNIT 2 B 3.3.6.1- 8 Amendment No. 134 l I l
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| | |
| Prizary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL. l The Allowable Value is chosen to ensure that offsite dose i limits are not exceeded due to the break. i This Function isolates the MSL and MSL drains isolation
| |
| . valves.
| |
| 1.d. Condenser Pressure-Hiah j The Condenser Pressure-High Function is provided to prevent l
| |
| ' overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Pressure-High Function is assumed to be l OPERABLE and capable of initiating closure of the MSIVs.
| |
| The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser I pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby f- g preventing a potential radiation leakage path following an (j accident. This function is credited with closing the MSIV's by the analysis of the " Loss of Condenser Vacuum" event l (Ref. 2).
| |
| Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) to Table 3.3.6.11), the channels are not required to be OPERABLE in MODES 2 and 3 during reactor shutdown or for reactor startup when bypassed under administrative control, since the potential for condenser overpressurization is minimized. Keylocked switches are provided to manually bypass the channels when necessary for conducting startup and shutdown operations while condenser pressure is above the trip setpoint. This also allows limited period of time to unbypass the channels when the bypass is no longer needed and condenser pressure is below the trip setpoint.
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| 1 U
| |
| FERMI - UNIT 2 B 3.3.6.1 - 9 Amendment No. 134
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| | |
| l- Prizary Containment Isolation Instrumentation B 3.3.6.1 BASES.
| |
| l APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILITY (continued)
| |
| This function is provided primarily as a backup to the closure of the turbine stop valves on Condenser Pressure-High. It is provided because of the potential consequence of exceeding off site doses, and because the turbine stop ,
| |
| valves and associated control system does.not meet l Protection System standards.
| |
| This Function isolates the MSL and MSL drains isolation valves.
| |
| l 1.e. and 1.a. Area Temoerature-Hiah Area temperature is provided to detect a leak in the RCPB l and provides diversity to the high flow instrumentation. If
| |
| ' the leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for these i instruments is not taken in any transient or accident i
| |
| , analysis in the UFSAR, since bounding analyses are performed
| |
| -for large breaks, such as MSLBs.
| |
| Area temperature signals are initiated from resistance temperature detectors (RTD) located in the area being O monitored. Sixteen channels of Main Steam Tunnel Temperature-High Function and eight channels of Turbine Building Area Temperature-High Function are available.
| |
| , Each of these Functions consist of two trip strings per trip system (for a total of 4 trip strings). For the Main Steam Tunnel Temperature-High Function, each trip string has inputs from four channels. Two channels per trip string of each Function are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation '
| |
| function.
| |
| The ambient temperature morinoring Allowable Value is chosen to detect a feedwater line break inside the steam tunnel.
| |
| l These Functions isolate the MSL and MSL drains isolation valves.
| |
| 1.f. Main Steam Line Radiation-Hiah l
| |
| High MSL radiation indicates there is a major fission product release due to a fuel cladding failure, and could provide an active role in mitigating release due to a control rod drop accident (Ref. 2). While MSIV closure initiated by Main Steam Line Radiation-High is not required FERMI . UNIT 2 8 3.3.6.1 - 10 Amendment No. 134 l'
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| l
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| Primary Containment Isolation Instrumentation B 3.3.6.1 O BASES G
| |
| APPLICABLE SAFETY ANALYSES LCO. and APPLICABILITY (continued) to ensure compliance with those guidelines of 10 CFR 100.
| |
| (Ref. 9) it is retained to maintain the overall diversity of parameters that cause an MSIV closure.
| |
| Main Steam Line Radiation-High signals are initiated from steam tunnel monitors that sense the presence of excessive radiation levels, indicative of a fuel cladding failure.
| |
| Four channels are available and required to be OPERABLE to l ensure that no single instrument failure can preclude the '
| |
| isolation function.
| |
| 4 The Allowable Value is based on the NRC guidelines of 3.6 l times the full power background radiation level with nominal i full power hydrogen injection rate. This allowable value )
| |
| remains fixed at this nominal full power basis even when operating at reduced power and/or reduced, or eliminated, hydrogen injection rates.
| |
| This Function shares common instrumentation with the RPS.
| |
| This function isolates the MSIVs. the MSL drains, and the Reactor Water Sample System.
| |
| 1.h. Manual Initiation The Manual Initiation channels provide manual isolation ca) ability. There is no specific UFSAR safety analysis that ta(es credit for this Function. It is retained for the overall redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the '
| |
| position of the valve control.
| |
| One channel of Manual Initiation Function per valve is available and required to be OPERABLE in MODES 1. 2. and 3.
| |
| since these are the MODES in which the MSL isolation automatic Functions are required to be OPERABLE.
| |
| A FERMI UNIT 2 B 3.3.6.1 - 11 Amendment No. 134
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| | |
| m Primary Containment Isolation Instrumentation B 3.3.6.1 BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Primary Containment Isolation 2.a. Reactor Vessel Water Level-Low. Level 3 Low RPV water level indicates that the ca) ability to cool the fuel may be threatened. The valves w1ose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded.
| |
| The Reactor Vessel Water Level-Low. Level 3 Function associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA.
| |
| Reactor Vessel Water Level-Low, Level 3 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of l-Reactor Vessel Water Level-Low, Level 3 Function are available and are required to be OPERABLE to ensure that no O single instrument failure can preclude the isolation function.
| |
| The Reactor Vessel Water Level-Low. Level 3 Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LC0 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.
| |
| This Function shares common instrumentation with the RPS.
| |
| This Function isolates the drywell sumps and TIP isolation valves.
| |
| 2.b. Reactor Vessel Water Level-Low Low. Level 2 Low RPV water level indicates that the ca) ability to cool the fuel may be threatened. The valves w1ose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 2 supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded.
| |
| The Reactor Vessel Water Level-Low Low, Level 2 Function associated with isolation is implicitly assumed to be isolated post LOCA.
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| O FERMI . UNIT.2 B 3.3.6.1 -12 Amendment No. 134
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| | |
| 1 Prinary Containment Isolation Instrumentation !
| |
| B 3.3.6.1 i
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| BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) :
| |
| Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value~was chosen to be the same as the ECCS Level 2 Initiation Allowable Value (LC0 3.3.5.1), since isolation of these valves is not critical to orderly plant shutdown.
| |
| This Function-isolates Reactor Water Sample System. TWMS, Drywell and Suppression Pool Ventilation System, Nitrogen {
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| Inerting System Recirculation Pump Seal System, Primary /
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| ' Containment Pneumatic Supply System, and PCMS isolation valves.
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| 2 '. c . Drywell Pressure-Hiah High drywell pressure can indicate a break in the RCPB i inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell
| |
| . pressure supports actions to ensure that offsite dose limits 4 of 10.CFR 100 are not exceeded. The Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident J analysis as these leakage paths are assumed to be isolated post LOCA.
| |
| High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High per Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
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| This Function shares common instrumentation with the RPS.
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| The Allowable Value was selected to be the same as the RPS i Drywell Pressure-High Allowable Value (LC0 3.3.1.1), since i this.may be indicative of a LOCA inside primary containment. l 4
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| V FERMI' . UNIT 2 B 3.3.6.1 - 13 Amendment No. 134
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| | |
| Pricary Containment Isolation Instrumentation B 3.3.6.1 BASES
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| -APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued)
| |
| This Function isolates certain RHR. CS, HPCI and RCIC isolation valves, as well as groups of drywell sumps, TIP, Reactor Water Sample System,11dMS, Drywell and Suppression
| |
| . Pool' Ventilation System, Nitrogen Inerting System, Recirculation Pump Seal System. Primary Containment ;
| |
| . Pneumatic Supply System, and PCMS isolation valves. I 2.d. Manual Initiation The' Manual . Initiation channels provide manual isolation capability. There is no specific UFSAR safety analysis that takes credit for this Function. It is retained for overall redundancy'and diversity of the isolation function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function since the i channels are mechanically actuated based solely on the position of the valve control.
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| One channel of the Manual Initiation Function per valve is available and required to be OPERABLE in MODES 1, 2, and 3.
| |
| since these are the MODES in which the Primary Containment
| |
| ' Isolation automatic Functions are required to be OPERABLE.
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| Hioh Pressure Coolant Injection and Reactor Core Isolation Coolina Systems Isolation 3.a. 4.a. HPCI and RCIC Steam Line Flow-Hioh Steam Line Flow-High Functions are provided i.c detect a break of the RCIC or HPCI steam lines and initiate closure ;
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| of the steam line isolation valves of the appropriate '
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| system. If the steam is allowed to continue flowing out of !
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| the break, the reactor will depressurize and the core can 1
| |
| , uncover. Therefore, the isolations are initiated on high ,
| |
| flow to The isolation '
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| action, prevent or minimize along with the scramcore damage.
| |
| function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any UFSAR accident analyses since the '
| |
| bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments l prevent the RCIC or HPCI steam line breaks from becoming !
| |
| bounding.
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| FERMI . UNIT 2 . B 3.3.6.1 - 14 Amendment No. 134
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| | |
| Pritary Containment Isolation Instrumentation B 3.3.6.1
| |
| "^''s O APPLICABLE SAFETY ANALYSES.' LCO, and APPLICABILITY (continued)
| |
| The HPCI and RCIC Steam Line Flow-High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines flow elements.
| |
| Two channels of both HPCI and RCIC Steam Line Flow-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. A time delay relay is used in the logic to delay isolation long enough to prevent spurious isolation resulting from start up pressure transients.
| |
| The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains the MSLB event as the bounding event.
| |
| These Functions isolate the HPCI and RCIC isolation valves, as appropriate.
| |
| 3.b. 4.b. HPCI and RCIC Steam Suoolv Line Pressure-Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue p operation of the associated system's turbine. These Q isolations are for equipment protection and are'not assumed in any transient or accident analysis in the UFSAR.
| |
| However, they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations (Ref. 3).
| |
| The HPCI.and RCIC Steam Supply Line Pressure-Low signals are initiated from transmitters-(four for HPCI and four for RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line i Pressure-Low Functions are available and are required to be !
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| OPERABLE to ensure that no single instrument failure can '
| |
| preclude the isolation function.
| |
| The Allowable Values are selected to be high enough to prevent damage to the system's turbine.
| |
| These Functions isolate the HPCI and RCIC system isolation valves, as appropriate.
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| O
| |
| . FERMI - UNIT 2 B 3.3.6.1 - 15 Amendment No. 134
| |
| | |
| Pritary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Additionally, the HPCI and RCIC steam supply line pressure-low signals are combined with ECCS drywell pressure-high signals to isolate the HPCI and RCIC turbine exhaust line vacuum breaker.
| |
| 3.c. 4.c. HPCI and RCIC Turbine Exhaust Diaohraam Pressure-Hiah High turbine exhaust diaphragm pressure indicates that the pressure may be too high to continue operation of the associated system's turbine. That is, one of two exhaust diaphragms has ruptured and pressure is reaching turbine casing pressure limits. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the UFSAR. These instruments are included in the TS because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations (Ref. 3).
| |
| The HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High signals are initiated from transmitters (four for HPCI and four for RCIC) that are connected to the area between the g~
| |
| rupture diaphragms on each system's turbine exhaust line.
| |
| Four channels of both HPCI and RCIC Turbine Exhaust Diaphragm Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Allowable Values are high enough to prevent damage to the system's turbine.
| |
| These Functions isolate the HPCI and RCIC system isolation valves, as appropriate.
| |
| 3.d. 4.d. HPCI and RCIC Eauioment Room Temoerature-Hioh Area temperatures are provided to detect a leak from the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation offsite dose limits may be reached.
| |
| These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
| |
| FERMI UNIT 2' B 3.3.6.1 - 16 Amendment No. 134
| |
| | |
| l
| |
| [ Primary Containment Isolation Instrumentation l B 3.3.6.1 l
| |
| BASES APPLICABLE SAFETY ANALYSES, LC0. and APPLICABILITY (continued)
| |
| HPCI and RCIC Equipment Room Temperature-High signals are I
| |
| initiated from thermocouples that are appropriately located to protect the system that is being monitored. Two instruments monitor each area. Two channels for each HPCI and RCIC Equipment Room Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Allowable Values are set low enough to detect a leak equivalent to 25 gpm.
| |
| These Functions isolate the HPCI and RCIC system isolation 1- valves, as appropriate.
| |
| 3.e. 4.e. Drwell Pressure-Hiah High drywell pressure can indicate a break in the RCPB. The HPCI and RCIC isolation of the turbine exhaust is provided to prevent communication with the drywell when high drywell pressure exists. A potential leakage path exists via the turbine exhaust. The isolation is delayed until the system O. becomes unavailable for injection (i.e., low steam line i pressure). The isolation of the HPCI and RCIC turbine exhaust by Drywell Pressure-High is indirectly assumed in the UFSAR accirier;t analysis because the turbine exhaust l leakage path is not assumed to contribute to offsite doses. j High drywell pressure signals are initiated from pressure !
| |
| transmitters that sense the pressure in the drywell. Two i channels of both HPCI and RCIC Drywell Pressure-High j Functions are available and are required to be OPERABLE to )
| |
| ensure that no single instrument failure can preclude the j isolation function. ,
| |
| The Allowable Value was selected to be the same as the ECCS I Drywell Pressure-High Allowable Value (LC0 3.3.5.1), since this is indicative of a LOCA inside' primary containment.
| |
| This Function is combined with the HPCI and RCIC steam supply line pressure-low signals to isolate the HPCI and ,
| |
| RCIC turbine exhaust line vacuum breaker.
| |
| l, l
| |
| l FERHI - UNIT 2 8 3.3.6.1 - 17 Amendment No. 134
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1
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| ( BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 3.f. 4.f. Manual Initiation The Manual Initiation channels provide manual isolation ca) ability. There is no specific UFSAR safety analysis that ta(es credit for these Functions. They are retained for overall redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for these Functions, since the channels are mechanically actuated based solely on the position of the push buttons.
| |
| One channel of Manual Initiation Function per valve is 1 available and required to be OPERABLE in MODES 1. 2. and 3 since these are the MODES in which the HPCI and RCIC systems
| |
| * Isolation automatic Functions are required to be OPERABLE.
| |
| Reactor Water Cleanuo System Isolation 5.a. Differential Flow-Hiah The high differential flow signal is rovided to detect a break in the RWCU System. This will detect leaks in the RWCU System when area or differential temperature would not provide detection (i.e., a cold leg break). Should the reactor coolant continue to flow out of the break, offsite dose limits may be exceeded. Therefore. isolation of the RWCU System is initiated when high differential flow is sensed to prevent exceeding offsite doses. A time delay is provided to prevent spurious trips during most RWCU operational transients. This Function is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large break; such as MSLBs.
| |
| The high differential flow signals are initiated from i transmitters that are connected to the RWCU punp outlet and RWCU system discharge to condenser and feedwater. The l outputs of the transmitters are compared ( ui a common !
| |
| summer) and the resulting output is sent to two high flow '
| |
| trip units. If the difference between the inlet and outlet flow is too large, each trip unit generates an isolation signal . Inoperability of the non redundant circuitry causes the channels in both trip systems to be inoperable. The remainder of the circuit is redundant and can be considered on a per trip system basis.
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| I FERMI - UNIT 2 B 3.3.6.1 - 18 Amendment No. 134
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1
| |
| -BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| The Differential Flow-High Allowable Value ensures that a break of the RWCU piping is detected.
| |
| This Function isolates the RWCU isolation valves.
| |
| 5.b. 5.c. Area and Area Ventilation Differential Temperature- Hich RWCU area and area ventilation differential temperatures are provided to detect a leak from the RWCU System. The isolation occurs even when very small leaks have occurred and is diverse to the high differential flow instrumentation for the hot portions of the RWCU System. If the small leak continues without isolation, offsite dose limits may be reached. Credit for these instruments is not taken in any transient or accident analysis in the UFSAR since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
| |
| l Area and area ventilation differential temperature signals are initiated from temperature elements that are located in the area or room that is being monitored. Twelve O thermocouples provide input to the Area Temperature-High Function (two per area). Two channels per area are required to be OPERABLE to ensure that no single instrument failure can preclude the . isolation function.
| |
| Eight thermocouples provide input to the Area Ventilation Differential Temperature-High Function. The output of these thermocouples is used to determine the differential temperature in four rooms containing RWCU piping and equipment. Each channel consists of a differential temperature instrument that receives inputs from
| |
| .thermocouples that are located in the inlet and outlet of
| |
| -the room cooling system and for a total of four available channels (one per room). i The Area and Area Ventilation Differential Temperature-High Allowable Values are set low enough to detect a leak l equivalent to 25 gpm.
| |
| These Functions isolate the RWCU isolation valves, as appropriate.
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| I FERMI - UNIT 2- B 3.3.6.1 - 19 Amendment No. 134
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILITY (continued) i 5.d. SLC System Initiation The isolation of the RWCU System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the RWCU System (Ref. 4). SLC System initiation signals are initiated from the two SLC pump start signals.
| |
| There is no Allowable Value associated with this Function since the channels are mechanically actuated based solely on
| |
| 'the position of the SLC System initiation switch.
| |
| Two channels (one from each pump) of the SLC System Initiation Function are available and are required to be OPERABLE only in MODES 1 and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LC0 3.1.7).
| |
| As noted (footnote (b) to Table 3.3.6.1 1), this Function is only required to close one of the RWCU isolation valves since the signals only provide input into one of the two O, trip systems.
| |
| 5.e. Reactor Vessel Water Level-Low Low. Level 2 Low RPV water level indicates that the capability to cool !
| |
| the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate i the potential sources of a break. The isolation of the RWCU '
| |
| l System on Level 2 supports actions to ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Vessel Water Level-Low Low, Level 2 Function associated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting).
| |
| Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from four level transmitters that sense the l difference between the pressure due to a constant column of !
| |
| water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE to ensure that no O
| |
| FERMI - UNIT 2 B 3.3.6.1- 20 Amendment No. 134
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| | |
| m Primary Containment Isolation Instrumentation B 3.3.6.1
| |
| .( BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) single instrument failure can preclude the isolation function.
| |
| The Reactor Vessel Water Level-Low Low. Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LC0 3.3.5.1),
| |
| since the capability to cool the fuel may be threatened.
| |
| This Function isolates the RWCU isolation valves.
| |
| 5.f. Manual Initiation The Manual' Initiation channels provide manual isolation capability. There is no specific UFSAR safety analysis that takes credit for this Function. It is retained for overall redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on the position of the valve control.
| |
| One channel of the Manual Initiation Function per valve is
| |
| ,. available and required to be OPERABLE in MODES 1, 2, and 3 since these are the MODES in which the RWCU System Isolation automatic Functions are required to be OPERABLE.
| |
| Shutdown Coolina System Isolation l
| |
| 6.a. Reactor Steam Dome Pressure-Hiah The Reactor Steam Dome Pressure-High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System. This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or-transient analysis in the UFSAR.
| |
| 'The Reactor Steam Dome Pressure-High signals are initiated from two transmitters that are connected to different taps on the RPV. Two channels of Reactor Steam Dome Pressure-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1. 2, and 3 since these O
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| FERMI - UNIT 2 B 3.3.6.1- 21 Amendment No. 134 k
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 I
| |
| BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) are the only MODES in which the reactor can be pressurized:
| |
| thus, equipment protection is needed. The Allowable Value was. chosen to be low enough to protect the system equipment from overpressurization.
| |
| This Function shares common instrumentation with the RPS.
| |
| This Function isolates the RHR shutdown cooling valves, as appropriate.
| |
| 6.b. Reactor Vessel Water Level-Low. Level 3 Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the
| |
| >otential sources of a break. The Reactor Vessel Water
| |
| _evel-Low Level 3 Function associated with RHR Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on Level 3 supports O actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.
| |
| Reactor Vessel Water Level-Low. Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level-Low. Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. As noted (footnote (c)
| |
| ' to Table 3.3.6.1-1). only two channels of the Reactor Vessel Water Level-Low. Level 3 Function are required to be OPERABLE in MODES 4 and 5 (and must input into the same trip system) provided the RHR Shutdown Cooling System integrity is maintained. System integrity is maintained provided the piping is intact and no maintenance is being performed that has the potential for draining the reactor vessel through the system.
| |
| FERMI - UNIT 2 B 3.3.6.1-22 Amendment No. 134
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| | |
| l Primary Containment Isolation Instrumentation B 3.3.6.1
| |
| / BASES APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILITY (continued)
| |
| The Reactor Vessel Water Level-Low, Level 3 Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level-Low, Level 3 Allowable Value (LC0 3.3.1.1), since the capability to cool the fuel may be threatened.
| |
| The Reactor Vessel Water Level-Low, Level 3 Function is only required to be OPERABLE in MODES 3, 4, and 5 to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2 another isolation (i.e., Reactor Steam Dome Pressure-High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.
| |
| This Function isolates the RHR shutdown cooling isolation valves, as appropriate.
| |
| 6.c. Manual Initiation The Manual Initiation channels provide manual isolation ca) ability. There is no specific UFSAR safety analysis that q taces credit for this Function. It is retained for overall V redundancy and diversity of the isolation function as required by the NRC in the plant licensing basis.
| |
| There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on the position of the push buttons.
| |
| One channel of the Manual Initiation Function per valve is available and required to be OPERABLE in MODES 1, 2, and 3 since these are the MODES in which the containment isolation automatic Functions are required to be OPERABLE.
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels.
| |
| Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
| |
| Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the q
| |
| b FERMI - UNIT 2 B 3.3.6.1 - 23 Amendment No. 134
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| | |
| )
| |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS (continued) l Condition. However, the Required Actions for inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable primary containment isolation instrumentation channel.
| |
| L.1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 1.f. 2.a. 2.c. and 6.b and 24 hours for Functions other than Functions 1.f. 2.a. 2.c. and 6.b has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the O inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an :
| |
| isolation), Condition C must be entered and its Required i Action taken. !
| |
| I iL.1 l Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic isolation capability being lost for the associated penetration flow path (s). The MSL Isolation Functions are considered to be maintaining isolation capability when ,
| |
| sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given Function on a valid signal. The other isolation functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the O
| |
| FERMI UNIT 2 B 3.3.6.1 -24 Amendment No. 134
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 l J
| |
| ( BASES ACTIONS (continued) two PCIVs in the associated penetration flow path can i receive an isolation signal from the given Function. For Functions 1.a.1.b 1.d. and 1.f. this would require both trip systems to have one channel 0PERABLE or in trip. For ,
| |
| Function 1.c. this would require both trip systems to have '
| |
| one channel, associated with each MSL, OPERABLE or in trip.
| |
| For Functions 1.e and 1.g each Function consists of channels that monitor several locations within a given area (e.g., different locations within the main steam tunnel area). Therefore, this would require both trip systems to have one channel per location OPERABLE or in trip. For Functions 2.a. 2.b. 2.c. 3.b. 3.c. 4.b. 4.c. 5.e. and 6.b. 1 this would require one trip system to have two channels, each OPERABLE or in trip. For Functions 3.a. 3.d. 4.a. 4.d, 5.a. 5.d. and 6.a. this would require one trip system to have one channel OPERABLE or in trip. For Functions 5.b and 5.c. each Function consists of channels that monitor :
| |
| several different locations. Therefore, this would require one channel per location to be OPERABLE or in trip (the channels are not required to be in the same trip system).
| |
| The Condition does not include the Manual Initiation
| |
| /^
| |
| Functions (Functions 1.ii. 2.d. 3.f. 4.f. 5.f. and 6.c),
| |
| ( since they are not assumed in any accident or transient analysis. Thus, a total loss of manual initiation capability for 24 hours (as allowed by Required Action A.1) is allowed.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| L1 Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.11. The applicable Condition specified in Table 3.3.6.1-1 is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A or B and the associated Com)letion Time has expired, Condition C will be entered for t1at channel and provides for transfer to the appropriate subsequent Condition.
| |
| /O V
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| FERMI - UNIT 2 B 3.3.6.1- 25 Amendment No. 134
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| l Primary Containment Isolation Instrumentation B 3.3.6.1 BASES
| |
| ' ACTIONS (continued).
| |
| D.1. D.2.1. and D.2.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time. the plant must be'placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and D.2.2). Alternately.
| |
| -the associated MSLs may be isolated (Required Action D.1),
| |
| and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL ,
| |
| isolated may continue. Isolating the affected MSL i accomplishes the safety function of the inoperable channel.
| |
| The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| L1 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must O- .
| |
| be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by placing the plant ,
| |
| in at least MODE 2 within 6 hours. 1 The allowed Completion Time of 6 hours is reasonable, based :
| |
| >- on . operating experience, to reach MODE 2 from full power !
| |
| conditions in an orderly manner and without challenging ,
| |
| plant systems.
| |
| 1 El If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration flow path (s) accomplishes the safety function of the inoperable channels.
| |
| For the RWCU Area and Area Ventilation Differential
| |
| -Temarature-High Functions, the affected penetration flow pat 1(s) may be considered isolated by isolating only that 4 portion of the system in the associated room monitored by the inoperable channel. That is, if the RWCU pump room A area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing O
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| FERMI UNIT 2 B 3.3.6.1-26 Amendment No. 134
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1
| |
| - BASES ACTIONS (continued) the B RWCU pum). For the RWCU Differential Flow-High ,
| |
| Function, if t1e flow element / transmitter monitoring RWCU l flow to radwaste and condensate is the only portion of the channel inoperable, then the affected penetration flow path (s) may be considered isolated by isolating the RWCU return to radwaste and condensate.
| |
| Alternately, if it is not desired to isolate the affected !
| |
| penetration flow path (s) (e.g., as in the case where ;
| |
| isolating the penetration flow path (s) could result in a l reactor scram), Condition H must be entered and its Required l Actions taken. -
| |
| The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path (s).
| |
| M If the channel is not restored to OPERABLE status or placed m in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path (s) is isolated. Isolating the affected penetration flow path (s) accomplishes the safety function of the inoperable channels.
| |
| The 24 hour Completion Time is acceptable due to the fact that these Functions (Manual Initiation) are not assumed in any accident or transient analysis in the UFSAR.
| |
| Alternately, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where isolating the penetration flow path (s) could result in a reactor scram), Condition H must be entered and its Required Actions taken.
| |
| H.1 and H.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or any Required Action of Condition F is not met and the associated Completion Time has expired, the plant must be ) laced in a' MODE or other specified condition in which the _C0 does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full O
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| v FERMI UNIT 2 B 3.3.6.1- 27 Amendment No. 134
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1
| |
| ,oO BASES ACTIONS (continued) power conditions in an orderly manner and without challenging plant systems.
| |
| I.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated SLC subsystem (s) is declared inoperable or the RWCU System is isolated. Since this Function is required to' ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated SLC subsystems inoperable or isolating the RWCU System.
| |
| The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System.
| |
| 'J.1 and J.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, iO these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is l restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.
| |
| SURVEILLANCE As noted at the beginning of the SRs the SRs for each REQUIREMENTS Primary Containment Isolation instrumentation Function-are found in the SRs column of Table 3.3.6.11.
| |
| The Surveillances'are modified by a note to indicate that when a channel is placed in an inopercble status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed. U)on completion of the Surveillance, or expiration of t1e allowance, the channel must be returned to OPERABLE status or the ap)licable Condition entered and Required Actions taken. T1is Note is based on the reliability analysis O
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| FERMI - UNIT 2 B 3.3.6.1 - 28 Amendment No. 134
| |
| | |
| Primary. Containment Isolation Instrumentation B 3.3.6.1 I
| |
| BASES
| |
| -SURVEILLANCE REQUIREMENTS (continued)
| |
| (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path (s) when necessary.
| |
| SR 3.3.6.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| O Aoceement criter4e ere determ4aea bx the nient steff besed on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside.the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
| |
| SR 3.3.6.1.2 and SR 3.3.6.1.6 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non Technical O
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| FERMI UNIT 2' B 3.3.6.1- 29 Amendment No. 134
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| | |
| 1 Primary Containment Isolation Instrumentation B 3.3.6.1 ;
| |
| BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| Specifications tests at least once per refueling interval with applicable extensions.
| |
| Any setpoint adjustment s' n all be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The 92 day Frequency of SR 3.3.6.1.2 is based on the reliability analysis described in References 5 and 6. The 18 month Frequency of SR 3.3.6.1.6 is based on engineering judgment and the reliability of the individual valve control components.
| |
| SR 3.3.6.1'.3 This surveillance provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under ;
| |
| these conditions, the setpoint must be readjusted to be i equal to or more conservative than that accounted for in the appropriate setpoint methodology. 1 The Frequency of 92 days is based on the reliability analysis of References 5 and 6.
| |
| SR 3.3.6.1.4 A CHANNEL CALIBRATION is a complete check of the instrument >
| |
| loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary
| |
| -range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| Tne Frequency of SR 3.3.6.1.4 is based on the assumption of a = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| O FERMI UNIT 2 B 3.3.6.1 - 30 Amendment No. 134
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| | |
| i Primary Containment Isolation Instrumentation B 3.3.6.1 l l
| |
| BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.6.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of. the required isolation logic for a specific channel. The system functional testing performed on PCIVs ;
| |
| in LC0 3.6.1.3 overlaas this Surveillance to provide !
| |
| complete testing of tie. assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. ,
| |
| 0)erating experience has shown these components usually pass j t1e Surveillance when performed at the 18 month frequency. '
| |
| SR 3.3.6.1.7 This SR ensures that the individual channel response times !
| |
| are less than or equal to the maximum values assumed in the accident analysis. The instrument response times must be i added to the PCIV closure times to obtain the ISOLATION SYSTEM RESPONSE TIME.
| |
| ISOLATION SYSTEM RESPONSE TIME acce tance criteria for the
| |
| -instrumentation portion are included in Reference 7, while the acceptance criteria for the PCIV closure times are included in Reference 8. This test may be performed in one measurement, or in overlapping segments, with verification that all components are tested.
| |
| A Note to the Surveillance states that the radiation detectors may be excluded from ISOLATION SYSTEM RESPONSE i TIME testing. This Note is necessary because of the difficulty of generating an appropriate detector input signal and because the principles of detector operation virtually ensure an instantaneous response time. Response times for. radiation detector channels shall be measured from detector output or the input of the first electronic component in the channel.
| |
| In addition. Note 2 states the response time of the sensors are excluded from the ISOLATION SYSTEM RESPONSE TIME testing. The sensors for the tested Functions are assumed to operate at the sensor's design response time. This allowance is supported by Reference 10 which determined that significant degradation of the sensor channel response time can be detected during performance of other Technical i 1
| |
| O FERMI - UNIT 2 B 3.3.6.1-31 Amendment No. 134 l 1
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| l i
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| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 I
| |
| O (V BASES SURVEILLANCE REQUIREMENTS (continued) !
| |
| Specification SRs and that the sensor response time is a small part of the overall response time testing. }
| |
| ISOLATION SYSTEM RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. The 18 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.
| |
| REFERENCES 1. UFSAR. Section 6.3.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. NE00-31466, " Technical Specification Screening Criteria Application and Risk Assessment "
| |
| November 1987.
| |
| : 4. UFSAR, Section 4.5.2.4.
| |
| : 5. NEDC-31677P A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"
| |
| July 1990.
| |
| : 6. NEDC-30851P-A Supplement 2 " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.
| |
| : 7. UFSAR, Section 7.3.
| |
| : 8. UFSAR, Section 6.2.
| |
| : 9. NED0 31400, " Safety Evaluation for Eliminating the BWR MSIV Closure Function and Scram Function of the MSL Radiation Monitor," Licensing Topical Plant Report for BWROG.
| |
| : 10. NED0 32291, " System Analysis for Elimination of Selected Response Time Testing Requirements," January 1994: and Fermi 2 SER for Amendment 111, dated April 18, 1997.
| |
| O V
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| FERMI UNIT 2 B 3.3.6.1 - 32 Amendment No. 134
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| | |
| y Secondary Containment Isolation Instrumentation B 3.3.6.2 B 3.3 ' INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs) trips the reactor building HVAC, and starts the Standby Gas Treatment (SGT)
| |
| System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. 1 and 2). Secondary containment isolation and establishment of vacuum with the SGT System within the assumed time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits.
| |
| The isolation instrianentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include s electronic equipment (e.g., trip units) that compares measured input signals with pre established setpoints. When the set >oint is exceeded, the channel output relay actuates, which tien outputs a secondary containment isolation signal !
| |
| to the isolation logic. Functional diversity is provided by l monitoring a wide range of independent parameters. The i input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, and (3) fuel pool ventilation exhaust high radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. In addition, manual initiation of the logic is provided.
| |
| The outputs of the logic channels in a trip system are arranged into two one out of-two trip system logics for the fuel pool ventilation exhaust high radiation; and are arranged into two two out-of-two trip system logics for the reactor vessel water level low and drywell pressure high trip system logics. One trip system initiates isolation of one automatic isolation valve (damp'er) in the penetration and starts one SGT subsystem while the other trip system O
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| , . FERMI - UNIT 2- B 3.3.6.2-1 Amendment No. 134
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| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 ,
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| ()
| |
| v BASES l BACKGROUND (continued) initiates isolation of the other automatic isolation valve in the penetration and starts the other SGT subsystem. Each logic closes one of the two valves on each penetration and starts one SGT subsystem, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of fission products.
| |
| APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the ,
| |
| LCO, and safety analyses of References 1 and 2 to initiate closure APPLICABILITY of valves and start the SGT System to limit offsite doses.
| |
| Refer to LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs) " and LCO 3.6.4.3, " Standby Gas Treatment (SGT) System." Applicable Safety Analyses Bases for more detail of the safety analyses.
| |
| The secondary containment isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). Certain instrumentation Functions are retained for other reasons and y are described below in the individual Functions discussion. 1 The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. A channel is inoperable if its actual trip i setpoint is not within its required Allowable Value.
| |
| Allowable Values are specified for each automatic Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the set)oints do not exceed the Allowable Value between CHANNEL CA IBRATIONS or between successive verifications of trip unit setpoints.
| |
| Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable.
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| 1 O
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| O l
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| l FERMI - UNIT 2 B 3.3.6.2 -2 Amendment No. 134
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| | |
| [L r l Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) l
| |
| ! Trip setpoints are those ) redetermined values of output at I which an action should ta(e place. The setpoints are
| |
| ~
| |
| compared to.the actual process )arameter (e.g., reactor vessel water level), and when t1e measured output value of the process parameter exceeds the setpoint, the associated device (e.g... trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.
| |
| The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, !
| |
| calibration tolerances, instrument drift, and severe environment errors-(for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
| |
| In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required.
| |
| The s Applibecific Applicable Safety ability discussions Analyses.
| |
| are listed below onLCO, and by a Function ,
| |
| Function basis. i 1
| |
| : 1. Reactor Vessel Water Level-Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
| |
| An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the '
| |
| potential of an offsite dose release. The Reactor Vessel Water Level-Low Low, Level 2 Function is one of the U Functions assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low Low, i Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
| |
| Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from level transmitters that sense the difference
| |
| -between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of O
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| FERMI UNIT 2 B 3.3.6.2 -3 Amendment No. 134
| |
| | |
| l
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| . \
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| Secondary Containment Isolation Instrumentation I B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Reactor Vessel Water Level-Low Low. Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. Instrument channels are derived from the RPS powered instruments to be consistent with the nuclear steam supply shut-off system design bases, which includes isolation on loss of power, or a deenergize-to-trip logic.
| |
| The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same at the High Pressure Coolant Injection / Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level-Low Low. Level 2 Allowable Value (LC0 3.3.5.1 and LC0 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened.
| |
| The Reactor Vessel Water Level-Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2 and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In
| |
| -MODES 4 and 5 the probability and consequences of these G events are low due to the RCS pressure and temperature V limitations of these MODES: thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (0PDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs.
| |
| : 2. Drywell Pressure-Hiah High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the mtential of an offsite dose release. The isolation on hig1 drywell pressure supmrts actions to ensure that any offsite releases are wit 11n the limits calculated in the safety analysis. i However, the Drywell Pressure-High Function associated with l 1 solation is not assumed in any UFSAR accident or transient analyses. It is retained for the overall redundancy and l
| |
| , FERMI UNIT 2 B 3.3.6.2-4 Amendment No. 134
| |
| | |
| Secondary Containment Isolation Instrumentation 8 3.3.6.2 L
| |
| . BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis. Instrument channels are derived from the RPS powered instruments to be consistent with the nuclear steam supply shut off system design bases, which includes isolation on loss of power, or a deenergize to trip logic.
| |
| High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation
| |
| . function.
| |
| The. Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LC0 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).
| |
| The Drywell Pressure-High Function is recuired to be OPERABLE in MODES 1, 2,.and 3 where consicerable energy
| |
| 'p exists in the RCS: thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
| |
| : 3. Fuel Pool Ventilation Exhaust Radiation-Hioh High fuel pool ventilation exhaust radiation is an indication of possible gross failure of the fuel cladding.
| |
| The release may have originated from the refueling floor due to.a fuel handling accident. When Fuel Pool Ventilation Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 3).
| |
| The Fuel Pool Ventilation Exhaust Radiation-High signals are initiated from radiation detectors that are located on.
| |
| .the' ventilation exhaust piping coming from the refueling floor zone. The signal from each detector is input to an i
| |
| O ~
| |
| 1 FERMI UNIT 2 B 3.3.6.2-5 Amendment No. 134 l i
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| e
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| _J
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| Secondary Containment Isolation Instrumentation B 3.3.6.2
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| /9 V
| |
| BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) individual monitor whose trip outputs are assigned to an isolation channel. Four channels of Fuel Pool Ventilation Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
| |
| The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.
| |
| The Fuel Pool Ventilation Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2. and 3 where considerable energy exists: thus. there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5. the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES: thus. this Function is not required. In addition, the Function is also required to be OPERABLE during CORE ALTERATIONS. OPDRVs. and movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or fm dropped fuel assemblies) must be provided to ensure that
| |
| \._) offsite dose limits are not exceeded.
| |
| : 4. Manual Initiation The Manual Initiation push button channels introduce signals into the secondary containment isolation and SGTS initiation logic that are redundant to the automatic protective instrumentation channels and provide manual isolation ca) ability. There is no specific UFSAR safety analysis that ta(es credit for this Function. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.
| |
| There are two push buttons for the logic. one manual initiation push button per trip system. There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on the position of the push buttons. I O l FERMI UNIT 2 B 3.3.6.2 -6 Amendment No. 134 l
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| | |
| Secondary Containment Isolation Instrumentation n B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| Two channels of Manual Initiation Function are available and are required to be OPERABLE in MODES 1, 2, and 3, and during CORE ALTERATIONS, OPDRVs, and movement of irradiated fuel I assemblies in the secondary containment. These are the I MODES and other specified conditions in which the Secondary Containment Isolation automatic Functions are required to be OPERABLE.
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.
| |
| Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition.
| |
| Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for ino wrable O
| |
| V secondary containment isolation instrumentation clannels provide appropriate compensatory measures for separate j
| |
| inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.
| |
| Al 4 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 4 and 5) to permit restoration of any inoperable channel to OPERABLE !
| |
| status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases).
| |
| If the ino wrable channel cannot be restored to OPERABLE status wit 11n the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore ;
| |
| O FERMI UNIT 2 B 3.3.6.2-7 Amendment No. 134
| |
| | |
| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES _
| |
| ACTIONS (continued).
| |
| capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation). Condition C must be entered and its Required
| |
| . Actions taken.
| |
| IL1 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated penetration flow path (s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip.
| |
| such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIVs in the associated penetration flow path and one SGT subsystem can be initiated on an isolation signal i from the given Function. For Functions 1 and 2, with two 1
| |
| A(/ two-out-of two logic trip systems, this would require one trip system to have two channels, each OPERABLE or in trip.
| |
| For Function 3 with two one out-of-two logic trip systems, this would require one trip system to have one channel 0PERABLE or in trip. The Condition does not include the Manual Initiation Function (Function 4), since it is not assumed in any accident or transient analysis. Thus, a total loss of manual initiation capability for 24 hours (as allowed by. Required Action A.1) is allowed.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because. it minimizes risk while allowing time for restoration or tripping of channels.
| |
| O FERMI --UNIT 2 B 3.3.6.2-8 Amendment No. 134
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| | |
| n Secondary Containment Isolation Instrumentation B 3.3.6.2
| |
| _ BASES ACTIONS (continued)
| |
| C.1.1. C.1.2. C.2.1. and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the secondary containment (closing the ventilation su) ply and exhaust automatic isolation dampers) and starting t1e associated SGT subsystem (Required
| |
| < Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue.
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| - Alternately, declaring the associated SCIVs or SGT subsystem (s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.2 and LC0 3.6.4.3) provide appropriate actions for the inoperable components.
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| One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily O, challenging plant systems.
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| SURVEILLANCE- As noted at the beginning of the-SRs. the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
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| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
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| 'This Note is based on the reliability analysis (Refs. 4 and 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary.
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| O FERMI JUNIT 2 B 3.3.6.2 -9 Amendment No. 134
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| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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| SR 3.3.6.2.1-Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A iTANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other
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| . channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties.
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| including indication and readability. If a channel is outside the criteria, it may be an indication that the
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| ' instrument has drifted outside its limit.
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| The Frequency is based on operating experience that
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| - demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal. but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.
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| SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
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| FERMI - UNIT 2 B 3.3.6.2 - 10 Amendment No. 134 l
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| l
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| Secondary Containment Isolation Instrumentation B 3.3.6.2 O 8^ses SURVEILLANCE REQUIREMENTS (continued)
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| Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
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| The Frequency of 92 days is based on the reliability analysis of References 4 and 5.
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| SR 3.3.6.2.3 This surveillance provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.6.21. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
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| The Frequency of 92 days is based on the reliability O analysis of References 4 and 5.
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| SR 3.3s6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
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| The Frequency of SR 3.3.6.2.4 is based on the assumption of a a: 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
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| FERMI ! UNIT'2 B 3.3.6.2 -11 Amendment No. 134
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| Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES
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| - SURVEILLANCE REQUIREENTS (continued)
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| SR 3.3.6.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LC0 3.6.4.2 and LC0 3.6.4.3. I respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
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| The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un)1anned transient if the Surveillance were performed with t1e reactor at power.
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| Operating experience has shown that these components usually pass the Surveillance.when performed at the 18 month Frequency. j i
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| REFERENCES 1. UFSAR. Section 6.3.
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| : 2. UFSAR, Chapter 15.
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| : 3. UFSAR. Section 15.7.4.
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| : 4. NEDC 31677P-A. " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation."
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| July 1990.
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| : 5. NEDC 30851P A Supplement 2. " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation." March 1989.
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| FERMI - UNIT 2 B 3.3.6.2 - 12 Amendment No. 134 g
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| LLS Instrumentation B 3.3.6.3
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| ~ B- 3.3 - INSTRUENTATION B 3.3.6.3 Low Low Set (LLS) Instrumentation BASES BACKGROUND The LLS relief mode functions to mitigate containment loads caused by re menings of an SRV by reducing the frequency of subsequent SR1 actuations following the initial SRV opening.
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| The steam discharge from the SRVs cause high frequency containment loads as well as thrust loads on the SRV discharge piping. The LLS allows time for the water leg that forms in the SRV discharge piping following SRV closure (from discharge piping residual steam condensation) to clear. Eliminating the water leg reduces the loading from the. subsequent SRV actuations to acceptable levels.
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| In addition, since subsequent SRV actuations will all involve the LLS SRVs, the LLS logic acts to reduce the number of challenges to the SRVs (by eliminating isolation cycling of the SRVs during some transients) and serves to dampen reactor pressure surges. Reducing the number of SRV challenges acts to reduce the probability of a stuck open relief valve event.
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| Upon initiation, the LLS logic will assign preset opening and closing setpoints to two preselected SRVs. These setpoints are selected such that the LLS SRVs will stay open longer: thus, releasing more steam (energy) to the suppression pool, and hence more energy (and time) will be required for repressurization and subsequent SRV openings.
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| The LLS logic increases the time between (or prevents) subsequent actuations to allow the high water leg created from the initial SRV opening to return to (or fall below) its normal water level; thus, reducing thrust loads from subsequent actuations to within their design limits. In addition, the LLS is oesigned to limit SpV subsequent actuations to one valve, so torus loads will also be reduced.
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| The LLS instrumentation logic is arranged in two divisions with logic channels A and C in one division and logic channels B and D in the other division (Ref. 1). Each LLS division controls one LLS valve. The LLS division will not actuate its associated LLS valve at its LLS setpoints until the arming portion of the associated LLS logic is satisfied.
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| Arming occurs when any one of the 15 SRVs opens as indicated by a signal from its SRV tailpipe pressure switches (one for FERMI ' UNIT 2 ' B 3.3.6.3 - 1 Amendment No. 134
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| _A .
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| LLS Instrumentation B 3.3.6.3
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| ( BASES BACKGROUND (continued) arming each division) coincident with a high reactor pressure signal.. Each division receives Tailpipe Pressure Switch arming signals from tailpipe pressure switches on each of the 15 SRVs. Each LLS division receives the Reactor Steam Dome Pressure High arming signal from one reactor pressure transmitter and trip unit assigned to that division. These arming signals (Tailpipe Pressure Switch and Reactor Steam Dome Pressure High) seal in until manually reset.
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| After arming, opening of each LLS valve is by a two out of two logic from two reactor pressure transmitters and two trip units set to trip at the required LLS opening setpoint. The LLS valve recloses when reactor pressure has decreased to the reclose setpoint of one of the two trip units used to open the valve (one out of-two reset logic).
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| This logic arrangement prevents single instrument failures from precluding the LLS SRV function. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre established setpoints. When the setpoint is exceeded, the channel output relay actuates, a which then outputs a LLS initiation signal to the initiation logic.
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| APPLICABLE The LLS instrumentation and logic function ensures that the SAFETY ANALYSES containment loads remain within the primary containment design basis (Ref. 2).
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| The LLS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| LC0 The LC0 requires OPERABILITY of sufficient LLS instrumentation channels to ensure successfully accomplishing the LLS function assuming any single instrumentation channel failure within the LLS logic.
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| Therefore, the OPERABILITY of the LLS instrumentation is dependent on the OPERABILITY of the instrumentation channel Function specified in Table 3.3.6.3-1. Each Function must (3
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| V -
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| FERMI UNIT 2- B 3.3.6.3 - 2 Amendment No. 134
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| l LLS Instrumentation B 3.3.6.3 BASES LC0 (continued) have a required number of OPERABLE channels, with their setpoints within the specified Allowable Value. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
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| Allowable Values are specified for each LLS actuation Function in Table 3.3.6.3-1. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure the setpoints do not exceed the Allowable Value between CHANNEL CALIBR4TIONS. Operation with a trip s(tpoint less conservative than the nominal trip setpoint. but within its Allowable Value, is acceptable.
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| Trip setpoints are those ) redetermined values of output at which an action should ta(e place. The setpoints are compared to the actual process )arameter (e.g., reactor vessel water level), and when t1e measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable A Values are derived from the analytic limits, corrected for C1 .
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| calibration, process, and some of the instrument errors.
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| The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties process effects, calibration tolerances, inst % ment drift, and severe environment errors (for cham-1s that must function in harsh environments as defined by lu CFR 50.49) are accounted for.
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| The Tailpipe Pressure Switch Allowable Value is based on ensuring that a pressure switch will actuate when the SRV is opened at = 150 psig reactor pressure, but will not actuate in response to a leaking SRV. That is, the pressure switch I is initiated only when an SRV has opened.
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| The React 7r Steam Dome Pressure-High was chosen to be the same as the Reactor Protection System (RPS) Reactor Steam Dome Pressure Allowable Value (LC0 3.3.1.1) because it would be expected that LLS would be needed for pressurization events. Providing LLS after a scram has been initiated would prevent false initiations of LLS at 100% power. The LLS valve open and close Allowable Values are based on the safety analysis performed for Reference 2.
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| FERMI - UNIT 2 B 3.3.6.3 -3 Amendment No. 134 ;
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| I l
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| LLS Instrumentation B 3.3.6.3 O BASES' V
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| APPLICABILITY The LLS instrumentation is recuired to be OPERABLE in MODES 1, 2. and 3 since consicerable energy is in the nuclear system and the SRVs may be needed to provide pressure relief. If the SRVs are needed, then the LLS function is required to ensure that the primary containment design basis is maintained. In MODES 4 and 5. the reactor pressure is low enough that the overpressure limit cannot be approached by assumed operational transients or accidents.
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| Thus, LLS instrumentation and associated pressure relief is not required.
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| ACTIONS A_d The failure of any reactor steam dome pressure instrument channel to provide the arming, SRV opening and closing pressure setpoints for an individual LLS valve does not affect the ability of the other LLS SRV to perform its LLS function. A LLS valve is OPERABLE if the associated logic has one Function 1 channel, two Function 2 channels, and at least three Function 3 channels OPERABLE. Therefore.
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| N 14 days is provided to restore the inoperable channel (s) to (V OPERABLE status (Required Action A.1). If the inoperable channel (s) cannot be restored to OPERABLE status within the allowable out of service time, Condition C must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action could result in an instrumented LLS valve actuation. The 14 day Completion Time is considered a>propriate because of the redundancy in the design (two L.5 valves are provided and any one LLS valve can perform the LLS function) and the very low probability of multiple LLS instrumentation channel failures, which renders the remaining LLS SRV inoperable, occurring together with an event requiring the LLS function during the 14 day Completion Time.
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| B.1. B.2. and B.3 Although the LLS circuitry is designed so that operation of a single tailpi>e pressure switch will result in arming both LLS logics: eac1 tailpipe pressure switch provides an input to both LLS logics. Since any overpressure event will normally open at least five SRVs and actuate their associated pressure switch inputs the LLS logic and instrumentation remains capable of performing its safety function even with several SRV tailpipe pressure switch FERMI - UNIT 2 B 3.3.6.3 - 4 Amendment No. 134
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| LLS Instrumentation B 3.3.6.3
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| ( ) BASES ACTIONS (continued) instrument channels inoperable. Therefore, it is acceptable for plant operation to continue provided that within 24 hours, per Required Action B.1 and B.2, verification and/or restoration is made to ensure at least: a) one tailpiae pressure switch in each Division OPERABLE on one OPERABLE SRV in the lowest SRV setpoint group; and b) at least 11 OPERABLE SRVs have at least one OPERABLE tailpipe pressure switch. Therefore, it is acceptable for plant operation to continue even with only one tailpipe pressure switch OPERABLE on each SRV. However, this is only acceptable provided each LLS valve is OPERABLE. (Refer to Required Action A.1 and C.1 Bases).
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| Required Action B.3 requires restoration of both tailpipe pressure switches on a 11 OPERABLE SRVs, including 4 SRVs out of the 5 lowest relief setpoint OPERABLE SRVs, to OPERABLE status, prior to entering MODE 2 or 3 from MODE 4.
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| This will ensure that sufficient switches are OPERABLE at the beginning of a reactor startup (this is because the switches are not accessible during plant operation). The Required Actions do not allow placing the channel in trip A since this action could result in an inadvertent LLS valve
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| () actuation. As noted, LC0 3.0.4 is not applicable, thus allowing entry into MODE 1 or 2 from MODE 2 or 3 with inoperable channels. This allowance is needed since the channels only have to be repaired prior to entering MODE 2 or MODE 3 from MODE 4.
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| A Note has been provided in the Condition to modify the Required Actions and Completion Times conventions related to LLS Function 3 channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable LLS Function 3 channels provide appropriate compensatory measures for separate inoperable Condition entry for each SRV with inoperable tailpipe pressure switches.
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| O FERMI UNIT 2 B 3.3.6.3 - 5 Amendment No. 134
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| LLS Instrumentation l B 3.3.6.3 l
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| : BASES ACTIONS (continued) ful If any Required Action and associated Completion Time of Conditions A or B are not met, or two LLS valves are inoperable due to inoperable channels, the LLS valves may be incapable of performing their intended function. Therefore, the plant must be placed in a MODE or other s)ecified condition in which the LC0 does not apply. T11s is done by placing the plant in at least MODE 3 within 12 hours and in H0DE 4 within 36 hours.
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| SURVEILLANCE As noted at the beginning of the SRs, the SRs for ea(.h LLS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.6.3-1. j SR 3.3.6.3.1 and SR 3.3.6.3.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the p intended function. A successful test of the required l 1 'j contact (s) of a channel relay may be performed by the )
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| verification of the change of state of a single contact of l the relay. This clarifies what is an acceptable CHANNEL j FUNCTIONAL TEST of a relay. This is acceptable because all l of the other required contacts of the relay are verified by i other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
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| The Frequency of 31 days is reasonable, based on o;,erating experience and on other indications that ensure pro)er functioning between CHANNEL FUNCTIONAL TESTS. Furtlermore, 0)erating experience shows that failure of more than one clannel in a given 31 day period is a rare event. l A portion of the SRV tailpipe pressure switch instrument channels are located inside the primary containment. The allowance for SR 3.3.6.3.2 to only perform the CHANNEL FlWCTIONAL TEST for portions of the channel outside of the primary containment is based on the location of these instruments and ALARA considerations and the requirefnent for p
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| L)
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| FERMI UNIT 2 B 3.3.6.3 - 6 Amendment No. 134
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| LLS Instrumentation B 3.3.6.3 1
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| BASES SURVEILLANCE REQUIREMENTS (continued) a complete CHANNEL CALIBRATION (SR 3.3.6.3.3) and LSFT (SR 3.3,6.3.4) every 18 months.
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| SR 3.3.6.3.3 CHANNEL CALIBRATION is a complete check of the instrument loop and sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
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| The Frequency of once every 18 months for SR 3.3.6.3.3 is based on the assumption of a a: 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
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| SR 3.3.6.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the n OPERABILITY of the required actuation logic for a specified I channel. The system functional testing performed in .
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| LC0 3.6.1.6. " Low Low Set (LLS) Safety / Relief Valves I (SRVs)." for SRVs overlaps this test to provide complete testing of the assumed safet/ function.
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| The Frequency of once every 18 months for SR 3.3.6.3.4 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the b,
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| potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
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| REFERENCES 1. UFSAR. Figure 7.3-13. j
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| : 2. UFSAR. Section 5.2.2. ,
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| i
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| )
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| J}v FERMI - UNIT 2 B 3.3.6.3 - 7 Amendment No. 134
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| CREF System Instrumentation B 3.3.7.1
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| [ B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation BASES BACKGROUND The CREF System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two CREF subsystems are each capable of fulfilling the stated safety function. The instrumentation.and controls for the CREF System automatically initiate action to pressurize recirculate and filter the main control room (MCR) air to minimize the consequences of radioactive material in the control room environment.
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| In the event of a loss of coolant accident (LOCA) signal (Reactor Vessel Water Level-Low Low, Level 2 or Drywell Pressure-High), Fuel Pool Ventilation Exhaust Radiation-High, or Control Center Normal Makeup Air Radiation-High signal, the CREF System is automatically started in the recirculation mode. The air is then O recirculated through the charcoal filter, and sufficient outside air is drawn in through the north or south emergency intake to maintain the MCR slightly pressurized with respect to the auxiliary building.
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| The CREF System instrumentation has two trip systems, either of which can initiate both CREF subsystems (Ref.1). Each !
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| trip system receives input from each of the Functions listed above. The Functions are arranged as follows for each trip system. The Reactor Vessel Water Level-Low Low, Level 2 and Drywell Pressure-High are each arranged in a two out of-two logic. The Fuel Pool Ventilation Exhaust Radiation-High is arranged in a one-out-of-two logic. The Control Center Normal Makeup Air Radiation-High is arranged i in a one out-of-one logic. The channels include electronic l equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then ;
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| outputs a CREF System initiation signal to the initiation logic.
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| O
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| ,V FERMI - UNIT 2 B 3.3.7.1 - 1 Amendment No. 134
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| o CREF System Instrumentation B 3.3.7.1
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| ( BASES APPLICABLE The ability of the CREF System to maintain the habitability SAFETY ANALYSES, of the MCR is explicitly assumed for certain accidents as LCO, and discussed in the UFSAR safety analyses (Refs. E, 3, and 4).
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| APPLICABILITY CREF System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed the limits set by GDC 19 of 10 CFR 50, Appendix A.
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| CREF System instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
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| The OPERABILITY of the CREF System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.7.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual tri) setpoint is not within its required Allowable Value. T1e actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
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| Allowable Values are specified for each CREF System Function specified in the Table. Nominal trip setpoints are s .
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| specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS or between successive verifications of trip unit setpoints. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. For Functions 1 and 2, the analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. These Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The associated trip setpoints are then determined accounting for the remaining instrument errors (e.g. , drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for
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| (
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| FERMI UNIT 2 B 3.3.7.1 - 2 Amendment No. 134 u l
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| CREF System Instrumentation B 3.3.7.1 1
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| /]
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| V BASES-APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY (continued) channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
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| For Functions 3 and 4, the instrument trip setpoint is set ,
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| as a function of the background radiation level, with the alarm setpoint at approximately three times the background level. Calculations have determined that setpoints as high as a thousand times the typical background level are adequate to warn of airborne radionuclide concentrations ,
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| which might result in the limiting 5 rem dose to the control room operator provided by General Design Criteria 19 of Appendix A to 10 CFR 50. The allowable value was selected by allowing a margin to the setpoint based on engineering judgment.
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| The specific Applicable Safety Analyses. LCO, and Applicability discussions are listed below on a Function by Function basis.
| |
| : 1. Reactor Vessel Water Level-Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that sy the capability of cooling the fuel may be threatened. A low reactor vessel water level could indicate a LOCA and will
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| ~
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| automatically initiate the CREF System, since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personr.el.
| |
| Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available (two channels per trip system) and are required to be OPERABLE to ensure that no single instrument failure can preclude CREF System initiation. The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LC0 3.3.5.1, "ECCS Instrumentation").
| |
| The Reactor Vessel Water Level-Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3, and during {
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| operations with a potential for draining the reactor vessel (0PDRVs) to ensure that the control room personnel are A(>
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| FERMI - UNIT 2 B 3.3.7.1 -3 Amendment No. 134 1
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| ]
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| CREF System Instrumentation l B 3.3.7.1 i 1
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| 4 BASES {
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| (v') APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) j protected during a LOCA. In MODES 4 and 5 at times other i than OPDRVs, the probability of a vessel draindown event resulting in a release of radioactive material into the environment is minimal. In addition adequate protection is
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| )erformed by the Control Room Air Inlet Radiation-High unction. Therefore, this Function is not required in other 1 MODES and specified conditions.
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| : 2. Drywell Pressure-Hioh High pressure in the drywell could indicate a break in the reactor coolant pressure boundary. A high drywell pressure signal could indicate a LOCA and will automatically initiate the CREF System. since this could be a precursor to a potential radiation release and subsequent radiation exposure to control room personnel.
| |
| Drywell Pressure-High signals are initiated from four pressure transmitters that sense drywell pressure. Four channels of Drywell Pressure-High Function are available (two channels per trip system) and are required to be n OPERABLE to ensure that no single instrument failure can preclude CREF System initiation. The Drywell Pressure-High
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| !J Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Allowable Value (LC0 3.3.5.1). .
| |
| The Drywell Pressure-High Furction is required to be OPERABLE in MODES 1, 2, and 3 to ensure that control room personnel are protected in the event of a LOCA. In MODES 4 and 5, the Drywell Pressure-High Function is not required since there is insufficient energy i6 the reactor to ,
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| pressurize the drywell to the Drywell Pressure-High I setpoint. !
| |
| : 3. Fuel Pool Ventilation Exhaust Radiation-Hiah High radiation in the fuel pool ventilation exhaust could be i the result of a fuel handling accident. A fuel pool ventilation exhaust high radiation signal will automatically initiate the CREF System, since this radiation release could result in radiation exposure to control room personnel.
| |
| The fuel pool ventilation exhaust radiation equipment consists of four monitors and channels located in the refueling floor area (two channels on the east fuel pool ventilation exhaust, and two channels on the west fuel pool O
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| V FERMI - UNIT 2 B 3.3.7.1 -4 Amendment No. 134
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| o .
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| CREF System Instrumentation ;
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| B 3.3.7.1 i
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| ) BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) ventilation exhaust). Four channels of Fuel Pool Ventilation Exhaust Radiation-High Function are available (two channels per trip system) and are reouired to be OPERABLE to ensure that no single instrument failure can preclude CREF System initiation. The Allowable Value was I selected to ensure that the Function will promptly detect high activity that could threaten exposure to control room personnel.
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| The Fuel Pool Ventilation Exhaust Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and operations with a potential for draining the reactor vessel (0PDRVs), to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event.
| |
| During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE A~LTERATIONS), the probability of a LOCA or fuel damage is low; thus, the Function is not required. ;
| |
| : 4. Control Center Normal Makeuo Air Radiation-Hioh The control center normal makeup air radiation monitors :
| |
| measure radiation levels before filtration in the inlet I ducting of the MCR. A high radiation level may pose a ,
| |
| threat to MCR personnel; thus, automatically initiating the l CREF System. l The Control Center Normal Makeup Air Radiation-High Function consists of two independent monitors. Two channels of Control Center Normal Makeup Air Radiation-High are available and are required to be OPERABLE to ensure that no single instrument failure can preclude CREF System i initiation. Tne Allowable Value was selected to ensure l protection of the control room personnel. 1 The Control Center Normal Makeup Air Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS, OPDRVs. and movement of irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA or ibei damage is low; thus, the Function is not required.
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| FERMI - UNIT 2 B 3.3.7.1 - 5 Amendment No. 134
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| | |
| CREF ystem Instrumentation B 3.3.7.1 BASES ACTIONS A Note has been provided to modify the ACTIONS related to l CREF System instrumentation channels. Section 1.3, l
| |
| Completion Times, specifies that once a Condition has been l entered, subsequent divisions, subsystems, components, or
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| , variables expressed in the Condition, discovered to be .
| |
| l . inoperable or not within limits, will not result in separate !
| |
| entry into the Condition. Section 1.3 also specifies that (
| |
| Required Actions of the Condition continue to apply for each additional failure, with Completion Times ~ based on initial entry into the Condition. However, the Required Actions for inoperable CREF System instrumentation channels provide I appropriate compensatory measures for separate inoperable l channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREF System j instrumentation channel. 1 y b Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.7.1-1. The applicable Condition specified in the Table is Function dependent.
| |
| Each time a channel is discovered inoperable. Condition A is entered for that channel and provides for. transfer to the O5 appropriate subsequent Condition.
| |
| B.1 and B.2 Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREF System design, an allowable out of service time of 24 hours has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status.
| |
| However, this out of service time is only acceptable i provided the associated Function is still maintaining CREF 1 System initiation capability. A Function is considered to be maintaining CREF System initiation capability when sufficient channels are OPERABLE or in trip such that one l trip system will generate an initiation signal from the given Function on a valid signal. For Functions 1 and 2, this would require one trip system to have two channels i OPERABLE or in trip. For Function 3, this would require one I trip system to have one channel out of two OPERABLE or in i tri ). If the CREF System initiation capability is lost, the 24 wur allowance of Required Action B.2 is not appropriate.
| |
| l If the Function is not maintaining CREF System initiation capability, the CREF System initiation capability must be O ~
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| FERMI UNIT 2 B 3.3.7.1 -6 Amendment No. 134
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES
| |
| -ACTIONS (continued) declared inoperable within 1 hour of discovery of the loss of CREF System initiation capability in both trip systems.
| |
| The 1 hour Completion Time (B.1) is acceptable because it minimizes risk while allowing tinie for restoring or tripping of channels.
| |
| If the inomrable channel cannot be restored to OPERABLE status wit 11n the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition D must be entered and its Required Action taken.
| |
| C.1 and C.2 Because of the diversity of sensors available to provide initiation signals and the redundancy of the CREF System design, an allowable out of service time of 6 hours is provided to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function is still maintaining CREF System initiation capability. A Function is considered to be maintaining CREF System initiation capability when sufficient channels are OPERABLE or in trip such-that one trip system will generate an initiation signal from the given Function on a valid signal. For Function 4, this would require one trip system to have one channel OPERABLE or in trip. In this situation (loss of CREF System initiation capability), the 6 hour allowance of Required Actiori C.2 is not appropriate. If the Function is not maintaining CREF System initiation capability, the CREF System initiation capability must be declared inoperable within 1 hour of discovery of the loss of CREF System initiation capability in both trip systems.
| |
| The 1 hour Completion Time (C.1) is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.
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| O FERMI UNIT 2 B 3.3.7.1- 7 Amendment No. 134
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| | |
| CREF System Instrumentation '
| |
| B 3.3.7.1 J BASES ACTIONS (continued)
| |
| If the ino mrable channel cannot be restored to OPERABLE status wit 11n the allowable out of service time per Required Action C.2, Condition D must be entered and its Required Action taken.
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| The 6 hour Completion Time is based on the consideration j that this Function provides the primary signal to start the l CREF System: thus, ensuring that the design basis of the i CREF System is met.
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| D.1 and D.2 With any Required Action and associated Completion Time not met, the CREF System must be placed in the emergency recirculation mode of operation per Required Action D.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREF System in operation must provide for automatically ,
| |
| re initiating the System upon restoration of power following 1 a loss of power to the System. Alternately, if it is not desired to start the System, the CREF subsystem associated i with inoperable, untripped channel (s) must be declared
| |
| ("' ,
| |
| inoperable immediately. ,
| |
| SURVEILLANCE As noted at the beginning of the SRs the SRs for each CREF REQUIREMENTS System instrumentation Function are located in the SRs column of Table 3.3.7.1 1.
| |
| The Surveillances are modified by a Note to indicate that for functions 1, 2, and 3, when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the 1 associated Function maintains CREF System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average ,
| |
| time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the CREF System will initiate when necessary, fm O ._
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| FERMI - UNIT 2 B 3.3.7.1 - 8 Amendment No. 134
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.7.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or i i
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| something even more serious. A CHANNEL CHECK will detect gross channel failure: thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
| |
| Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
| |
| The Frequency is based u n operating ex mrience that demonstrates channel fai re is rare. T1e CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LC0.
| |
| SR 3.3.7.1.2 and SR 3.3.7.1.3 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all i of the other required contacts of the relay are verified by other Technical Specifications and non Technical Specifications tests at least once per refueling interval with' applicable extensions.
| |
| Any'setpoint adjustment shall be consistent with the assumptions of-the current plant specific setpoint methodology.
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| p b
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| FERMI 1 UNIT 2 B 3.3.7.1-9 Amendment No. 134
| |
| | |
| CREF System Instrumentation B 3.3.7.1 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| The Frequency of 31 days is reasonable, based on operating experience and on other Surveillances that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. Furthermore.
| |
| o)erating experience shows that failure of more than one c1annel in a given 31 day period is a rare event.
| |
| The Frequency of 92 days is based on the reliability analyses of Reference 5.
| |
| SR 3.3.7.1.4 This surveillance provides a check of the actual trip setpoints. Any setpoint adjustment shall be consistent with the assumptions of the current 31 ant specific setpoint methodology. The channel must )e declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.7.11. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology. but-is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under
| |
| ^ these conditions,.the setpoint must be readjusted to be equal to or more conservative than the setting accounted for in the appropriate setpoint methodology.
| |
| The Frequency of 92 days is based on the reliability analyses of Reference 5.
| |
| SR 3.3.7.1.5 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
| |
| The Frequency is based upon the assur.ption of a = 18 month calibration interval in the determination of the magnitude of. equipment drift in the setpoint analysis.
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| . FERMI UNIT 2 B 3.3.7.1 -10 Amendment No. 134 m
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| | |
| CREF System Instrumentation B 3.3.7.1 BASES
| |
| )
| |
| SURVEILLANCE REQUIREMENTS (continued)
| |
| SR 3.3.7.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.3, " Control Room Emergency Filtration (CREF)
| |
| System," overlaps this Surveillance to provide complete testing of the assumed safety function.
| |
| The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
| |
| 0)erating experience has shown these components usually pass t1e Surveillance when performed at the 18 month Frequency.
| |
| REFERENCES 1. UFSAR, Figure 9.4.2.
| |
| : 2. UFSAR, Section 9.4.1.
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| A
| |
| () 3. UFSAR. Section 6.4.1.
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| : 4. UFSAR, Chapter 15.
| |
| : 5. Safety Evaluation Re) ort for Fermi Unit-2 Amendment No. 75, dated Septem)er 6, 1991.
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| (O V
| |
| FERMI - UNIT 2 B 3.3.7.1- 11 Amendment No. 134
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| LOP Instrumentation .
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| B 3.3.8.1 O
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| y B 3.3 INSTRUMENTATION B 3.3.8.1 Loss of Power (LOP) Instrumentation BASES
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| =
| |
| BACKGROUND Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability _ of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV emergency buses.
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| Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite <
| |
| emergency diesel generator (EDG) power sources.
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| Each 4.16 kV emergency bus has its own independent LOP l instrumentation and associated trip logic. The voltage for l each bus is monitored at two levels, which can be considered as two different undervoltage Functions: Loss of Voltage and 4.16 kV Emergency Bus Undervoltage Degraded Voltage. j Bus undervoltage instrumentation for the loss of voltage A)
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| ( function monitors the Class 1E emergency bus for a level of voltage that is insufficient to o>erate the required ESF
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| ]'
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| equipment caused from a loss of t1e preferred off site power source. Bus undervoltage instrumentation.for the degraded voltage function monitors the Class 1E bus for a level of voltage from the preferred off-site power source that is insufficient to operate the required ESF equipment, but is not low enough to cause the loss of voltage function to operate, thereby ensuring adequate voltage for ESF o>eration and protecting the ESF equipment from damage caused )y low voltage operation. Both loss of voltage functions within a bus cause identical actions Class IE bus isolation, load shedding to prevent overloading of the associated EDG.
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| transfers, and automatic starting of the associated EDG and load sequencer.
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| Each Function is monitored by four undervoltage relays that compare measured input signals with pre established setpoints in each bus. When the setpoint of the undervoltage relay is exceeded continuously for 2 seconds, the channel output relay actuates. The outputs of the undervoltage relays are arranged in a one-out of-two taken-twice logic (i.e., two trip systems) for each Function in l each bus. The input signal to the undervoltage relays is at q
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| G 1 FERMI - UNIT 2 B 3.3.8.1 - 1 Amendment No. 134
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| | |
| l LOP Instrumentation B 3.3.8.1 BASES BACKGROUND (continued) the 120 Volt level and is derived from two independent step-down bus potential transformers (PT). Each PT is electrically connected to the associated Class 1E bus when the off-site power source is supplying the bus. Each bus uses one potential transformer connected to the line side of the Class 1E bus feed breaker providing input to one of the ,
| |
| redundant wired pair of undervoltage relays, and a second j potential transformer connected to the load side of the bus feed breaker providing input to the other redundant wired pair of undervoltage relays. A coincident trip in each redundant pair of undervoltage relay channels then causes a LOP trip signal to the trip logic. The degraded voltage function for each Class 1E bus has an additional time delay relay that is summed with time delay action provided in the undervoltage relays. The additional time delay is to l prevent actuation of the trip logic for a pre established time limit, unique for the associated divisional off-site power source.
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| Both LOP Functions are automatically bypassed whenever the associated EDG is supplying its respective ESF bus. This (3 ensures that the voltage dips encountered during load
| |
| () sequencing on the EDG will not interact with the load shedding feature.
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| APPLICABLE The LOP instrumentation is required for Engineered Safety i SAFETY ANALYSES. Features to function in any accident with a loss of offsite '
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| LCO, and power. The required channels of LOP instrumentation ensure APPLICABILITY that the ECCS and other assumed systems powered from the EDGs. provide plant protection in the event of any of the i Reference 2. 3 and 4 analyzed accidents in which a loss of I offsite power is assumed. The initiation of the EDGs on loss of offsite power, and subsequent initiation of the ECCS ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
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| Accident analyses credit the loading of the EDG based on the loss of offsite power during a loss of coolant accident.
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| The diesel starting and loading times have been included in r
| |
| the delay time associated with each safety system component j requiring EDG supplied power following a loss of offsite i
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| power.
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| FERMI - UNIT 2 B 3.3.8.1 -2 Arrondment No.134 L
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| p LOP Instrumentation B 3.3.8.1
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| ;i( BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
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| The LOP instrumentation satisfies Criterion 3 of 10 CFR l
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| 50.36(c)(2)(ii).
| |
| The OPERABILITY of the LOP instrumentation is dependent upon 3 the OPERABILITY of the individual instrumentation channel j Functions specified in Table 3.3.8.11. Each Function must 1 have a required number of OPERABLE channels )er 4.16 kV i
| |
| emergency bus, with their setpoints within t1e specified l Allowable Values. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. )
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| The actual setpoint is calibrated consistent with applicable j setpoint methodology assumptions.
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| l The Allowable Values are specified for each Function in the Table. Nominal trip setpoints are specified in the setpoint l
| |
| calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between l CHANNEL CALIBRATIONS. Operation with a trip setpoint less l conservative than the nominal trip setpoint, but within the Allowable Value, is acceptable. Trip setpoints are those
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| ; predetermined values of output at which an action should
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| ' (~'% take place. The setpoints are compared to the actual V process parameter (e.g., degraded voltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the i analytic limits, corrected for calibration, process, and i some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g. , drift). The trip set)oints derived in this manner provide adequate protection >ecause instrumentation uncertainties, process effects, calibration tolerances, i instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
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| : The specific Applicable Safety Analyses LC0, and i Applicability discussions are listed below on a Function by Function basis.
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| O V
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| FERMI --UNIT 2 B 3.3.8.1 - 3 Amendment No. 134 a
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| c.
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| LOP Instrumentation B 3.3.8.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
| |
| : 1. 4.16 kV Emeroency Bus Undervoltaae (Loss of Voltaae)
| |
| Loss of voltage on a 4.16 kV emergency bus indicates that offsite power may be com)letely lost to the respective emergency bus and is una)le to supply sufficient power for proper operation of the applicable equipment. Therefore, the power supply to the bus is transferred from offsite power to EDG power when the voltage on the bus drops below the Loss of Voltage Function Allowable Values (loss of voltage with a short time-delay). This ensures that adequate power will be available to the required equipment.
| |
| The Bus Undervoltage Allowable Values are low enough to
| |
| -prevent inadvertent power supply transfer, but high enough to ensure that power is available to the required equipment.
| |
| The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that power is available to the required equipment.
| |
| Four channels of 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) Function )er associated emergency bus are only
| |
| ' O required to be OPEMBLE when the associated EDG is required to be OPERABLE to ensure that no single instrument failure can preclude the EDG function. Refer to LC0 3.8.1, "AC Sources-Operating," and 3.8.2, "AC Sources-Shutdown," for Applicability Bases for the EDGs.
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| : 2. 4.16 kV Emeroency Bus Undervoltaae (Deoraded Voltaae)
| |
| A reduced voltage condition on a 4.16 kV emergency bus
| |
| -indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.
| |
| Therefore, power supply to the bus is transferred from offsite power to onsite EDG power when the voltage on the l bus drops below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay). This ensures that adequate power will be available to the required equipment.
| |
| The Bus Undervoltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough O
| |
| FERMI m UNIT 2 B 3.3.8.1 - 4 Amendment No. 134 t
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| L
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| H .
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| LOP Instrumentation B 3.3.8.1 BASES
| |
| , APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) to provide time for the offsite )ower supply to recover to
| |
| . normal voltages, but short enoug1 to ensure that sufficient power 1s available to the required equipment.
| |
| ~
| |
| Four channels of 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) Function per associated bus are only required to be OPERABLE when the associated EDG is required to be OPERABLE to ensure that no single instrument failure can preclude the EDG function. Refer to LC0 3.8.1 and LC0 3.8.2 for Applicability Bases for the EDGs.
| |
| ' ACTIONS' A Note has been provided to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion
| |
| . Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each
| |
| ' f/ 3 . . additional failure, with Completion Times based on initial
| |
| 'C-entry.into the Condition. However, the Required Actions for inoperable LOP instrumentation channels provide appropriate com>ensatory measures for separate inoperable channels. As suc1, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.
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| Al i
| |
| With one or more channels of a Function ino)erable, the Function may not be capable of performing t1e intended function (if LOP trip capability is lost, Condition B is also required to be entered). Therefore. 72 hours are allowed to restore the inoperable channel to 0PERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time.
| |
| Condition B must be entered and its Required Action taken.
| |
| The Completion Time is intended to allow the operator time.
| |
| to evaluate and repair any discovered inoperabilities. The 72 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration of channels.
| |
| O FERMI.- UNIT 2 B 3.3.8.1- 5 Amendment No. 134
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| | |
| p J
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| LOP Instrumentation I B 3.3.8.1 BASES ACTIONS (continued) ill If Required Action A.1 and associated Completion Time is not met, or the associated Function is not capable of performing the intended function, the associated EDG(s) is declared inoperable immediately. This requires entry into applicable i Conditions and Required Actions of LC0 3.8.1 and LC0 3.8.2. '
| |
| which provide appropriate actions for the inoperable EDG(s).
| |
| ' SURVEILLANCE As noted at the beginning of the SRs, the SRs for each LOP REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.8.1-1.
| |
| SR 3.3.8.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required f channel to ensure that the entire channel will perform the intended function. A successful test of the required contact (s) of a channel relay may be performed by the
| |
| : p. verification of the change of state of a single contact of ;
| |
| 'Q the' relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by i other Technical Specifications and non Technical
| |
| . Specifications tests at least once per refueling interval with applicable extensions. Any setpoint adjustment shall
| |
| 'be consistent with the assumptions of the current plant specific setpoint methodology.
| |
| The Frequency of 31 days is based on operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event.
| |
| SR 3.3.8.1.2 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test ~ verifies the channel ;
| |
| responds to the measured parameter within the necessary !
| |
| range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for in:trument drifts between successive i calibrations consistent with the plant specific setpoint methodology. This SR also ensures the sum of the degraded voltage time delay and the longest time delay of the four
| |
| ' FERMI UNIT 2 B 3.3.8.1 - 6 Amendment No. 134
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| L
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| , LOP Instrumentation B 3.3.8.1
| |
| ( BASES SURVEILLANCE REQUIREMENTS (continued) associated bus undervoltage relays remains consistent with the plant specific setpoint methodology.
| |
| Any setpoint adjustment shall be consistent with the assurnptions of the current plant specific setpoint methodology.
| |
| The Frequency is based upon the assumption of a = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
| |
| SR 3.3.8.1.3 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specific channel. The system functional testing performed in LC0 3.8.1 and LC0 3.8.2 overlaps this Surveillance to provide complete testing of the assumed safety functions.
| |
| The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an un31anned transient if the O Surveillance were performed with t1e reactor at power.
| |
| Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.
| |
| REFERENCES 1. UFSAR, Figure 8.3 8.
| |
| : 2. UFSAR, Section 3.6.
| |
| : 3. UFSAR, Section 6.3.
| |
| : 4. UFSAR, Chapter 15.
| |
| FERMI UNIT 2 B 3.3.8.1 - 7 Amendment No. 134
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| 'l i
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| J
| |
| | |
| l RPS Electric Power Monitoring B 3.3.8.2
| |
| /V 3 B 3.3 INSTRUMENTATION l
| |
| B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring BASES BACKGROUND RPS Electric Power Monitoring System is provided to isolate the RPS bus from the motor generator (MG) set or the alternate power supply in the event of overvoltage, '
| |
| undervoltage, or underfrequency. This system protects the loads connected to the RPS bus against unacceptable voltage and frequency conditions (Ref.1) and forms an important part of the primary success path of the essential safety circuits. Some of the essential equipment powered from the '
| |
| RPS buses includes the RPS logic, scram solenoids, and various valve isolation logic.
| |
| RPS electric power monitoring assembly will detect any abnormal high or low voltage or low frequency condition in the outputs of the MG set or the alternate power supply and will de energize its respective RPS bus, thereby causing all safety functions normally powered by this bus to -
| |
| de energize. l In the event of failure of an RPS Electric Power Monitoring System (e.g., both inseries electric power monitoring assemblies), the RPS loads may experience significant effects from the unregulated power supply. Deviation from the nominal conditions can potentially cause damage to the scram solenoids and other Class IE devices.
| |
| In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram action.
| |
| In the event of an overvoltage condition, the RPS logic relays and scram solenoids, as well as the main steam isolation valve (MSIV) solenoids, may experience a voltage higher than their design voltage. If the overvoltage condition persists for an extended time period, it may cause equipment degradation and the loss of plant safety function.
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| Two redundant Class 1E circuit breakers are connected in series between each RPS bus and its MG set, and between each RPS bus and its alternate power supply. Each of these circuit breakers has an associated independent set of Class 1E overvoltage, undervoltage, and underfrequency FERMI - UNIT 2 B 3.3.8.2 - 1 Amendment No. 134
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| | |
| RPS Electric Power Monitoring B 3.3.8.2 BASES BACKGROUND (continued) sensing logic. Together, a circuit breaker and its sensing logic constitute an electric power monitoring assembly. If the output of the associated power supply exceeds predetermined limits of overvoltage, undervoltage, or underfrequency, a trip coil driven by this logic circuitry opens the circuit breaker, which removes the associated power supply from service.
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| l l
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| APPLICABLE The RPS electric power monitoring is necessary to meet the l SAFETY ANALYSES assumptions of the safety analyses by ensuring that the equipment powered from the RFS buses can perform its intended function. RPS electric power monitoring provides protection to the RPS and other systems that receive power from the RPS buses, by acting to disconnect the RPS from the power su) ply under specified conditions that could damage I the RPS aus powered equipment.
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| RPS electric power monitoring satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). ,
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| t L j LC0 The OPERABILITY of each RPS electric power monitoring !
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| assembly is dependent on the OPERABILITY of the overvoltage.
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| undervoltage, and underfrequency logic, as well as the OPERABILITY of the associated circuit breaker. Two electric power monitoring assemblies are required to be OPERABLE for each inservice power supply. This provides redundant protection against any abnormal voltage or frequency conditions to ensure that no single RPS electric power monitoring assembly failure can preclude the function of RPS bus powered components. Each inservice electric power monitoring assembly's trip logic setpoints are required to be within the specified Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
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| I Allowable Values are specified for each RPS electric power monitoring assembly tr1p logic (refer to SR 3.3.8.2.2).
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| Nominal trip set)oints are specified in the setpoint calculations. T1e nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its C
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| s FERMI - UNIT 2 B 3.3.8.2 -2 Amendment No. 134
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| RPS Electric Power Monitoring B 3.3.8.2 BASES LC0 (continued)
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| Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Tri ) setpoints are those predetermined values of output at w11ch an action should take place. The setpoints are compared to the actual process parameter (e.g., overvoltage), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are deriwd from the limiting values of the process parameters obtained from the safety analysis. The Allowable i Values are derived from the analytic limits, corrected for j calibration, process, and some of the instrument errors.
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| The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
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| The Allowable Values for the instrument settings are based r on the RPS providing = 57 Hz,120 V i 10% (to all equipment), and 115 V i 10 V (to scram and MSIV solenoids).
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| The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels. The settings are calculated based on the loads on the buses and RPS MG set or alternate power supply being 120 VAC and 60 Hz.
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| 1 APPLICABILITY The o>eration of the RPS electric power monitoring assem) lies is essential to disconnect the RPS bus powered components from the MG set or alternate power supply during abnormal voltage or frequency conditions. Since the degradation of a nonclass 1E source supplying power to the RPS bus can occur as a result of any random single failure, l the OPERABILITY of the RPS electric >ower monitoring l assemblies is required when the RPS )us powered components are required to be OPERABLE. This results in the RPS Electric Power Honitoring System OPERABILITY being required in MODES 1, 2, and 3: and in MODES 4 and 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies or with both residual heat removal (RHR) shutdown cooling suction isolation valves open.
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| D v
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| FERMI UNIT 2 83.3.8.2-3 Amendment No. 134
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| n . 1 RPS Electric Power Monitoring B 3.3.8.2 4 BASES ACTIONS &J.
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| 1 If one RPS electric power monitoring assembly for an inservice power supply (MG set or alternate) is inoperable, or one RPS electric power monitoring assembly on each inservice power supply is inoperable, the OPERABLE assembly will still provide protection to the RPS bus powered components under degraded voltage or frequency conditions.
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| However, the reliability and redundancy of the RPS Electric Power Monitoring System is reduced, and only a limited time (72 hours) is' allowed to restore the inoperable assembly to OPERABLE status. If the inoperable assembly cannot be restored to OPERABLE status, the associated power supply (s) must be removed from service (Required Action A.1). This places the RPS bus in a safe condition. The alternate power supply with OPERABLE powering monitoring assemblies may then be used to power the RPS bus.
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| 1 The 72 hour Completion Time takes into account the remaining i OPERABLE electric power monitoring assembly and the low probability of an event requiring RPS electric power monitoring protection occurring during this period. It n allows time for plant operations personnel to take j
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| \ corrective actions or to place the plant in the required 1 condition in an orderly manner and without challenging plant systems.
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| Alternately, if it is not desired to remove the power supply from service (e.g., as in the case where removing the power supply (s) from service would result in a scram or isolation), Condition C or D, as applicable, must be entered and its Required Actions taken.
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| (L1 If both zwer monitoring assemblies for an inservice power supply ((i set or alternate) are inoperable or both power monitoring assemblies in each inservice power supply are inoperable, the system protective function is lost. In this condition,1 hour is allowed to restore one assembly to OPERABLE status for each inservice power supply. If one inoperable assembly for each inservice power supply cannot be restored to OPERABLE status, the associated power ;
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| supply (s) must be removed from service within 1 hour '
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| (Required Action B.1). The alternate power supply with !
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| OPERABLE assemblies may then be used to power one RPS bus.
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| The 1 hour Completion Time is sufficient for the plant O
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| FERMI' UNIT 2. B 3.3.8.2 -4 Amendment No. 134 1
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| i-RPS Electric Power Monitoring B 3.3.8.2 BASES ACTIONS (continued) operations >ersonnel to take corrective actions 6nd is acceptable )ecause it minimizes aisk while allowing time for restoration or removal from se;vice of the electric power monitoring assemblies.
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| Alternately, if it is not desired to remove the power
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| :;upply(s) trom service (e.g., as in the case where removing the power supply (s) from service would result in a scram or isolation).. Condition C or D, as applicable, must be entered and its Required Actions taken.
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| C.1 and C.2 If any Required Actior and associated Completion Time of Condition A or B are not met in MODE 1, 2 or 3. a plant shutdown must be perforned. This places the plant in a condition where minimal equipment, powered through the 1 inoperable RPS electric power monitoring assembly (s), is '
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| required and ensures that the safety function of the RPS (e.g., scram of control rods) is not required. The plant shutdown is accomplished by placing the plant in MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed
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| \ Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without )
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| challenging plant systems. j D.I. D.2.1. and D M If any Required Action and associated Completion Time of I Condition A or B are not met in MODE 4 or 5, or with any I control rod withdrawn from a core cell containing one or '
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| more fuel assemblin or with both RHR shutdown cooling valves open, the operator must immediately initiate action to fully insert all insertable control rods in core cells contair>ing one or more fuel assemblies. Required Action D.1 results in the least reactive condition for the reactor core and ensures that the safety function of the RPS (e.g., scram of control rods) is not required.
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| In addition, action must be imediately initiated to either restore one electric power monitoring assembly to OPERABLE status for the inservice power source supplying the required instrumentation powered from the RPS bus (Required Action D.2.1) or to isolate the RHR Shutdown Cooling System (Required Action D.2.2). Required Action D.2.1 is provided O
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| FERMI - UNIT 2 B 3.3.8.2 -5 Amendment No. 134
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| RPS Electric Power Monitoring B 3.3.8.2 0
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| 'k s _
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| ACTIONS (continued) because the RHR Shutdown Cooling System aay be needed to provku core cooling. All actions must continue until the t.pplicable Required Actions are completed.
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| SURVEILLANCE SR 3.3.8.24 REQUIREMENTS A CHANNE'l FUNCTIONAL TEST is performed on each overvoltage, undervoltage, and underfrequency channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint ethodology.
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| As noted in the Surveillance, the CHANNEL i'9NrTIGi4AL TEST is only required to be performed while the plant is in a condition in which t1e loss of the RPS bus will not jeopardize steady ststa power operation (the design of the system is such that the power source must be re'noved from service to conduct the Surveillance). The 24 hours is intended to indicate an outage of sufficient t'eration to allow for scheduling and proper performnce of tre
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| ( Surveillance.
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| The 184 day Frequency and the Note in ta Surveillance are based on guidance provided in Generic Letter 9109 (Ref. 2).
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| SR 3.3.8.L 2 CHANNEL CALIBRATION is a complete check of the instruneC loop .^,nd the sensor. This test verifies that the channel respono; to the measured parameter within the necessary range Lnd accuracy. CHANNEL CALIBRATION leaves the channel adjustad to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
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| The Frequency is based on the assumption of a = 18 month Cdlibrat fon interval in the determination of the magnitude of equipnent drift in the setpoint analysis.
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| 77 Q) -- . -
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| FERMI . UNIT 2 B 3.3.8.2-6 Amendment No. 134
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| RPS Electric Power Monitoring '
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| B 3.3.8.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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| SR 3.3.8.2.3 Performance of. a system functional test demonstrates that, with a required system actuation (simulated or actual) signal, the logic of the system will automatically trip open the associated power monitoring assembly. Only one signal per power monitoring assembly is required to be tested. l This Surveillance overlaps with the CHANNEL CALIBRATION to i provide complete testing of the safety function. The system !
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| functional test of the Class 1E circuit breakers is included j as part of this test to provide complete testing of the ,
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| safety function. If the breakers are incapable of operating, the associated electric power monitoring assembly would be inoperable.
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| The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Su?veillance were performed with the reactor at power.
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| Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month
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| /") Frequency.
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| xJ REFERENCES 1. UFSAR, Section 7.2.1.1.2.
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| : 2. NRC Generic Letter 91-09, " Modification of Surveillance Interval for the Electrical Protective Assemblies in Power Supplies for the Reactor Protection System."
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| l l
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| FERMI - UNIT 2 B 3.3.8.2- 7 Amendment No. 134 4
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| r_
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| Recirculation Loops Operating B 3.4.1 B 3.4 REACTOR COOLANT' SYSTEM (RCS)
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| - B 3.4.1. Recirculation Loops Operating BASES BACKGROUND .The Reactor Coolant Recirculation System is designed to
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| )rovide a forced coolant flow through the core to remove wat from the fuel. The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation. The forced flow, therefore, 6? lows operation at significantly higher power than would otherwise be possible. The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator. The Reactor Coolant Recirculation System consists of two recirculation pump loo)s external to the reactor vessel. These loo)s provide tw piping path for the driving flow of water to tw reactor vessel jet pumps. Each external loop contains one variable speed motor driven recirculation pump, a motor generator (MG) set to control pump speed and associated piping, valves, and instrumentation. The recirculation loops are part of the reactor coolant pressure boundary and, with the exception of the MG sets, are located inside the drywell structure. The jet pumps are reactor vessel internals.
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| The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold, fra which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. -This flow enters the jet pump at suction inlets and 1s accelerated by the driving flow. The drive flow and suction flow are mixed in the jet ) ump throat section. The total flow then passes through tw jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core. The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat j is transferred to the coolant. As it rises, the coolant ;
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| i FERMI UNIT.2 B 3.4.1 - 1 Amendment No. 134 l
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| )
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| Recirculation Loops Operating B 3.4.1 O B=s BACKGROUND (continued) begins to boil, creating . steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative i reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control system allows omratcrs to increase recirculation flow and sweep some of t1e volds from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of pover {
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| generation without having to move control rods and disturb I desirable flux patterns. I i
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| Each recirculation loop is manually started from the controi l room. The MG set provides regulation of individual l recirculation loop drive flows. The flow in each loop n '
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| manually controlled within limits established by the recirculation speed control system. I GDC 12 of 10 CFR 50 Appendix A (Ref. 4) states that the reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations l which can result in exceeding specified fuel design limits are not possible or can be reliably detected and suppressed.
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| BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed, along with other system parameters (xenon, subcooling, power distribution, etc.) the thermal-hydraulic / reactor kinetic feedback mechanism can be enhanced such that perturbations may result in sustained limit cycle or divergent oscillations in power and flow.
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| Two major modes of oscillations have been observed in BWRs.
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| The first mode is the fundamental or core-wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillt: tion in which one half of the core oscillates 180 degrees cut of phase with the other half. Studies have indicated that adequate margin to the Safety Limit MCPR may not exist ;
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| during oscillations.
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| 4 FERMI UNIT 2 B 3.4.1-2 Amendment No. 134 u.
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| Recirculation Loops Operating B 3.4.1 O Bases APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref.1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accfdant. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered.
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| The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (stice the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins (O
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| ._) during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR.
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| A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysit has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling provided the APLHGR requirements are modified accordingly (Ref. 3).
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| The transient analyses of Chapter 15 of the UFSAR have also been performed for- single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Prctection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor FERMI UNIT 2 B 3.4.1-3 Amendment No. 134
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| l Recirculation Loops Operating B 3.4.1 BASES APPLICABLE SAFETY ANALYSIS (continued) core flow. The APLHGR and MCPR setpoints for single loop o)eration are specified in the COLR. The APRM Simulated T1ermal Power - Upscale setpoint is in LC0 3.3.1.1. " Reactor Protection System (RPS) Instrumentation."
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| Thermal-hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal-hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference 4 are satisfied.
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| Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
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| 1 LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in G SR 3.4.1.2 not met, the recirculation loop with the lower b flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR limits (LC0 3.2.1 " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"). MCPR limits (LC0 3.2.2
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| " MINIMUM CRITICAL POWER RATIO (MCPR)"). APRM Simulated Thermal Power-Upscale setpoint (LC0 3.3.1.1) and limitation on THERMAL POWER may be applied to allow continued operation consistent with the assumptions of the safety analysis.
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| Operations that exhibit core thermal-hydraulic instability are not permitted. Additionally, in order to avoid potential power oscillations due to thermal-hydraulic instability, operation at certain combinations of power and flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" regions and are defined by Bases Figure B 3.4.11.
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| A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to establish the applicable limitations in accordance with the single loop analysis. The 4 hour period is sufficient to make the adjustment given the relatively small change required. This transition only results in applying the new single loop allowable nlues to APRM OPERABILITY. Any ARPM b
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| %J FERMI - UNIT 2 B 3.4.1 - 4 Amendment No. 134
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| Recirculation Loops Operating B 3.4.1 BASES LC0 (continued) non compliance with the required allowable value after this 4 hour allowance, results in ACTIONS of LC0 3.3.1.1 being entered: no ACTION of LC0 3.4.1 would apply. Similarly, any operation with APLHGR or MCPR out of limits results in the ACTIONS of LC0 3.2.1 or LC0 3.2.2 being entered; no ACTION I of LC0 3.4.1 would apply.
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| APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the roactor core and the limiting design basis transients and accidents are assumed to occur.
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| In MODES 3, 4, and 5, the consequences of an accider.t are reduced and the coastdown characteristics of the recirculation loops are not important. In addition, sufficient power to create power oscillations that threaten fuel design limits does not exist.
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| ' ACTIONS M With the requirements for matched recirculation loop flow not met, the recirculation loops must be restored to o)eration with matched flows within 2 hours or the loop with t1e lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop at a significantly lower flow than the other loop, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the loop to operating status.
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| Alternatively, if the single loop requirements of the LC0 are applied to RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence.
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| The 2 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
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| (^)
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| %)
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| FERMI - UNIT 2 B 3.4.1 - 5 Amendment No. 134
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| Recirculation Loops Operating B 3.4.1 1 BASES l ACTIONS (continued)
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| This Required Action does not require td.; ping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases ..here large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet i pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re establish forward flow or by tripping the pump.
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| IL1 When operating in the " Exit" region (refer to Figure B 3.4.11), the potential for thermal-hydraulic instabilities is increased and sufficient margin may not be available for operator response to sup)ress potential power oscillations. Therefore, action must )e initiated immediately to restore operation outside of the " Exit" )
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| region. Control rod insertion and/or core flow increases are designated as the means to accomplish this objective.
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| r Required Action B.1 is modified by a Note that precludes j t, core flow increases by restart of an idle recirculation loop, or by resetting a recirculation flow limiter. Core flow increases by these means would not support timely completion of the action to restore operation outside the
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| " Exit" Region.
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| Cal With no recirculation loops in operation in MODE 2, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from MODE 2 conditions in an orderly manner and without challenging plant systems.
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| FERMI - UNIT 2 B 3.4.1 - 6 Amendment No. 134
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| ~
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| r Recirculation Loops Operating B 3.4.1 gi j BASES ACTIONS (continued) 0.1 If operating with no recirculation pumps in operation in MODE 1 or operating in the " Scram" region (refer to Bases Figure B 3.4.1-1), or if core thermal-hydraulic instability is detected, then unacceptable power oscillations may result. Therefore, the reactor mode switch must be immediately placed in the shutdown position to terminate the potential for unacceptable power oscillations.
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| Thermal hydraulic instability is evidenced by a sustained 1 increase in APRM or LPRM peak to peak noise level reaching 2 !
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| or more times its initial level and occurring with a characteristic period of less than 3 seconds.
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| 1 If entry into this condition is an unavoidable and well known consequence of an event, early initiation of the Required Action is appropriate. Also it is recognized that !
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| during certain abnormal conditions, it may become {
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| operationally necessary to enter the " Scram" or " Exit" :
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| region for the purpose of: 1) protecting plant equipment.
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| O which if it were to fail could impact plant safety, or
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| \ 2) protecting a safety or fuel operating limit. In these cases, the appropriate actions for the region entered would be performed as required.
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| These requirements are consistent with References 5 and 6.
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| SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR provides frequent periodic monitoring for core thermal-hydraulic instability by monitoring APRM and LPRM signals for a sustained increase in APRM or LPRM peak to peak noise level reaching 2 or more times its initial level and occurring with a characteristic >eriod of less than 3 seconds. The 1 hour Frequency is )ased on the small potential for core thermal-hydraulic oscillations to occur outside the " Scram" or " Exit" regions. Therefore, frequent monitoring of the APRM and LPRM signals is appropriate when operating in the " Stability Awareness" region.
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| (}
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| %.J 1
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| FERMI - UNIT 2 B 3.4.1- 7 Amendment No. 134 j
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| Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
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| This SR is modified by a Note that states performance is only required when operating in the " Stability Awareness" region (refer to Bases Figure B 3.4.1-1) (i.e., in the power-to-flow region that is near regions of higher probability for core thermal-hydraulic instabilities). This is acceptable because outside the " Stability Awareness" region, power and flow conditions are such that sufficient margin exists to the potential for core thermal-hydraulic instability to allow routine core monitoring. Any 4 unanticipated entry into the " Stability Awareness" region (
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| would require immediate verification of core stability since the Surveillance would not be current.
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| 1' SR 3.4.1.2 This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i .e. ,
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| < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch p can therefore be allowed when core flow is < 70% of rated d core flow. The recirculation loop jet pump flow as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
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| The mismatch is measured in terms of )ercent of rated core flow. If the flow mismatch exceeds tie specified limits, the loop with the lower flow is considered "not in operation". The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.
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| p FERMI UNIT 2 B 3.4.1 - 8 Amendment No. 134
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| Recirculation Loops Operating B 3.4.1 BASES REFERENCES 1. UFSAR, Section 6.3.3.
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| : 2. NEDE 23785 P-A, " SAFER /GESTR Models for the Evaluation of the Loss of Coolant Accident," Revision 1, October 1984.
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| : 3. M)E-56 0386 " Fermi 2 Single Loop Operation Analysis."
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| Rev. 1. April 1987, and NEDC-32313-P, "Enrico Fermi Energy Center Unit 2 Single Loop Operation," September i 1994.
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| : 4. 10 CFR 50, Appendix A GDC 12.
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| i
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| : 5. NRC Generic Letter 94-02, "Long Term Solutions and U) grade of Interiin Operating Recommendations for T1ermal Hydraulic Instabilities in Boiling Water Reactors," July 1994.
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| : 6. BWROG Letter 94078, "BWR Owners' Group Guidelines for Interim Corrective Action " June 1994.
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| ^s (G
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| I I
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| 1 I
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| O l FERMI - UNIT 2 B 3.4.1 - 9 Amendment No. 134 l
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| L Jet Pumps B 3.4.2 6 3.4 REACTOR COOLANT SYSTEM (RCS)-
| |
| ~
| |
| B 3.4.2 Jet Pumps BASES-l BACKGROUf0 The Reactor Coolant Recirculation System is described in the Background section of the Bases for LC0 3.4.1,
| |
| " Recirculation Loops Operating," which discusses the o)erating characteristics of the system and how these claracteristics affect the Design Basis Accident (DBA) analyses.
| |
| The jet pumps are part of the Reactor Coolant Recirculation System and are designed to 3rovide forced circulation through the core to remove leat from the fuel. The jet pumps are located in the annular region between the core shroud and the vessel inner wall. Because the jet pump suction elevation is at two thirds core height, the vessel can be reflooded and coolant level maintained at two thirds core height even with the complete break of the recirculation loop suction elevation. pipe that is located below the jet pump O
| |
| V Each reactor coolant recirculation loop contains ten jet pumps. Recirculated coolant ) asses down the annulus between the reactor vessel wall and t1e core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure-flow into an external manifold from which individual recirculation inlet lines are' routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet .
| |
| pump at suction inlets and is. accelerated by the drive flow.
| |
| The drive flow and suction flow are mixed in the jet ) ump throat section. The total flow then passes through t1e Jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.
| |
| O FERMI - UNIT 2 B 3.4.2-1 Amendment No. 134
| |
| | |
| Jet Pumps B 3.4.2 BASES.
| |
| APPLICABLE Jet pump OPERABILITY is an explicit assumption in the design SAFETY ANALYSES basis loss of coolant accident (LOCA) analysis evaluated in Reference 1.
| |
| The capability of reflooding the core to two-thirds core j height is de>endent upon the structural integrity of the jet i pumps. If t1e structural system, including the beam holding i
| |
| a jet pump in place, fails, jet pump displacement and performance degradation could occur, resulting in an increased flow area through the jet pump and a lower core flooding elevation. This could adversely affect the water level in the core during the reflood phase of a LOCA as well as the assumed blowdown flow during a LOCA.
| |
| Jet pumps satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 1 LC0 The structural failure of any of the jet pumps could cause l significant degradation in the ability of the jet pumps to I allow reflooding to two-thirds core height during a LOCA.
| |
| OPERABILITY of all jet pumps is required to ensure that operation of the Reactor Coolant Recirculation System will O be consistent with the assumptions used in the licensing basis analysis (Ref.1).
| |
| APPLICABILITY In MODES l'and 2, the jet pumps are required to be OPERABLE since there is a large amount of energy in the reactor core and since the limiting DBAs are assumed to occur in these MODES. This is consistent with the requirements for o)eration of the Reactor Coolant Recirculation System
| |
| (.00 3.4.1).
| |
| In MODES 3. 4, and 5, the Reactor Coolant Recirculation System is not required to be in oporation, and when not in operation,- sufficient flow is not available to evaluate jet pump OPERABILITY.
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| 4 O
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| FERMI UNIT 2 B 3.4.2-2 Amendment No. 134
| |
| | |
| Jet Pumps B 3.4.2 BASES ACTIONS dul An inoperable jet pump can increase the blowdown area and reduce the capability of reflooding during a design basis LOCA. If one or more of the jet pum)s are inoperable, the plant must be brought to a MODE in W11ch the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. The Completion Time of 12 hours is reasonable based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This SR is designed to detect significant degradation in jet pump performance that precedes jet pump failure (Ref. 2).
| |
| This SR is required to be performed only when the loop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation (this includes performing this SR when the loop is operating but may be declared "not in o>eration" in accordance with O' the ACTIONS of LC0 3.4.1). T1e jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. Jet pump plugging is also of concern since it adds flow resistance to the recirculation loop. Significant degradation is indicated if the specified criteria confirm unacceptable deviations from established patterns or relationships. The allowable deviations from the established patterns have been developed based on the variations experienced at plants during normal operation and with jet pump assembly failures (Refs. 2 and 3). Each recirculation loop must satisfy two of the performance criteria provided. Since refueling activities (fuel assembly replacement or shuffle, as well as any modifications to fuel sup) ort orifice size or core plate bypass flow) can affect t1e relationship between core flow, jet pump differential pressures, recirculation pum) speed, and recirculation loop drive flow, these relations 11ps may need to be re established each cycle. Similarly, initial entry into extended single loop operation may also require establishment of these relationships. During the initial weeks of operation under such conditions, while base lining new " established patterns", engineering judgement of the daily surveillance results is used to detect significant abnormalities which could indicate a jet pump failure.
| |
| FERMI - UNIT 2 B 3.4.2 -3 Amendment No. 134 l 1
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| | |
| Jet Pimps B 3.4.2
| |
| : BASES SURVEILLANCE REQUIREMENTS (continued) l The recirculation pump speed operating characteristics (loop drive flow versus pump speed and loop drive flow versus total core flow) are determined by the flow resistance from the loop suction through the jet pump nozzles. A change in the relationship indicates a plug, flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle. For criterion a., the loop drive flow versus pum) speed relationship must be verified. For criterion b., t1e l loop drive flow versus total core flow relationship must be verified.
| |
| Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal ft.w is due to the drive i flow manifold, which does not distribute flow equally to all risers. The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An apareciable change in this relationship is an indication tlat increased (or reduced) resistance has occurred in one of the jet pumps. This may be indicated by an increase in the relative (3 flow for a jet pump that has experienced beam cracks, failed U beam inletriser crack, or jet pump assembly crack.
| |
| The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 2). Normal flow ranges and established jet ) ump flow and differential pressure patterns are esta)lished by plotting historical data as discussed in Reference 2.
| |
| The 24 hour Frequency has been shown by operating experience to be timely for detecting jet pump degradation and is consistent with the Surveillance Frequency for recirculation '
| |
| loop OPERABILITY verification.
| |
| This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation.
| |
| y/
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| FERMI - UNIT 2 83.4.2-4 Amendment No. 134
| |
| | |
| Jet Pumps B 3.4.2 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| Note 2 allows this SR not to be performed when THERMAL POWER is s 25% of RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data.
| |
| REFERENCES 1. UFSAR, Section 6.3.
| |
| : 2. GE Service Information Letter No. 330, June 9, 1990.
| |
| : 3. NUREG/CR-3052, November 1984.
| |
| O o
| |
| L/
| |
| FERMI - UNIT 2 B 3.4.2-5 Amendment No. 134 l
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| I
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| | |
| x SRVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| B 3.4.3 Safety Relief Valves (SRVs)
| |
| BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system. the size and number of SRVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB). :
| |
| The SRVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The SRVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve ,
| |
| opens when steam pressure at the valve inlet overcomes the I spring force holding the pilot valve closed. Opening the ,
| |
| pilot valve allows a pressure differential to develop across !
| |
| the main valve piston and opens the main valve. This satisfies the Code requirement.
| |
| Each SRV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.
| |
| The SRVs that provide the relief mode are the low low set (LLS) valves and the Automatic Depressurization System (ADS) ,
| |
| valves. The LLS requirements are specified in LCO 3.6.1.6 j
| |
| " Low Low Set (LLS) Valves." and the ADS requirements are i specified in LC0 3.5.1. "ECCS-Operating." l APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient. Evaluations have determined that the most severe transient is the closure of I all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e.. failure of the direct '
| |
| scram associated with MSIV position) (Ref. 1). For the purpose of the analyses.11 SRVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
| |
| (
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| L)J FERMI UNIT 2 B 3.4.3 -1 Amendment No. 134
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| | |
| i SRVs !
| |
| B 3.4.3 l BASES APPLICABLE SAFETY ANALYSES (continued)
| |
| From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses ad11tional events that are expected to actuate the SRVs.
| |
| SRVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The safety function of 11 SRVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs.1 )
| |
| and 2). The requirements of this LC0 are applicable only to the capability of the SRVs to mechanically open to relieve excess pressure when the lift setpoint-is exceeded (safety function).
| |
| The SRV setpoints, and 3% allowance for setpoint drift, are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does Oi not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the UFSAR are also based on these setpoints.
| |
| Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.
| |
| -APPLICABILITY In MODES 1, 2, and 3, 11 SRVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The SRVs may be required to provide )ressure relief to discharge energy from the core until suc1 time '
| |
| that the Residual Heat Removal (RHR) System is capable of dissipating the core heat. ,
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| D U
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| FERMI - UNIT 2- B 3.4.3 -2 Amendment No. 134
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| | |
| SRVs B 3.4.3 BASES APPLICABILITY (continued)
| |
| In MODE 4. decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5. the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The SRV function is not needed during these conditions.
| |
| ACTIONS A.1 and A.2 With less than the minimum number of required SRVs OPERABLE. ;
| |
| a transient may result in the violation of the ASME Code !
| |
| limit on reactor pressure. If the safety function of any !
| |
| required SRVs cannot be maintained, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without O '
| |
| lallenging plant systems.
| |
| 1 SURVEILLANCE 'a 3.4.3.1 REQUIREMENTS mis Surveillance requires that the required SRVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the SRV safe lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating tem aratures and pressures. The SRV setpoint is i 3t for ,
| |
| OPEMBILITY. however, the valves are reset to i lt during i the Surveillance to allow for drift. l 1
| |
| The SR gives set pressures for all 15 SRVs installed. i However, since only 11 SRVs are required, the SR is met if j 11 SRVs are set properly.
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| O _
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| FERMI UNIT 2- B 3.4.3 -3 Amendment No. 134
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| ~
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| i SRVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| The Frequency is required by the Inservice Testing Program and is consistent with the fact that Surveillance must be performed during shutdown conditions.
| |
| SR 3.4.3.2 A manual actuation of each required SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.
| |
| Adequate reactor steam dome pressure must be available to l 4
| |
| perform this test to avoid damaging the valve. Also, I adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the SRVs divert steam flow upon opening.
| |
| Sufficient time is therefore allowed after the required i pressure and flow are achieved to perform this test.
| |
| Adequate pressure at which this test is to be performed is a 850 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by fi V turbine bypass valves open at least 20%. Plant startup is I allowed prior to wrforming this test because valve OPERABILITY and t1e setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed I until 12 hours after reactor steam pressure and flow are i adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE.
| |
| t The 18 month Frequency was developed based on the SRV t=sts i required by the ASME Boiler and Pressure Vessel Code.
| |
| Section XI (Ref. 3). Operating experience has shown that these components usually. pass the Surveillance when wrformed at the 18 month Frequency. Therefore, the requency was concluded to be acceptable from a reliability standpoint.
| |
| O FERMI UNIT 2 B 3.4.3 -4 Amendment No. 134
| |
| | |
| SRVs B 3.4.3-BASES REFERENCES 1. UFSAR, Section 5.2.2.3.5.
| |
| : 2. UFSAR, Chapter 15.
| |
| : 3. ASME, Boiler and Pressure Vessel Code, Section XI.
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| O I
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| O FERMI UNIT 2 B 3.4.3 -5 Amendment No. 134 L
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| =
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| RCS Operational LEAKAGE B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| B 3.4.4 RCS Operational LEAKAGE BASES.
| |
| BACKGROUND The RCS includes systems and components that contain or transport the coolant to or from the reactor core. The pressure containing components of the RCS and the portions of connecting systems out to and including the isolation valves define the reactor coolant pressure boundary (RCPB).
| |
| The joints of the RCPB components are welded or bolted.
| |
| During plant life. the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE through either normal operational wear or mechanical deterioration.
| |
| Limits on RCS operational LEAKAGE are required to ensure appropriate action is taken before the integrity of the RCPB i is impaired. This LC0 specifies the types and limits of )
| |
| LEAKAGE. This protects the RCS pressure boundary described l in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50. 1 Appendix A (Refs 1. 2. and 3). i The safety significance of RCS LEAKAGE from the RCPB varies I s widely depending on the source, rate, and duration. I Therefore, detection of LEAKAGE in the primary containment. )
| |
| is necessary. Methods for quickly separating the identified I LEAKAGE from the unidentified LEAKAGE are necessary to provide the operators quantitative information to permit them to take corrective action should a leak occur that is detrimental to the safety of the facility or the public.
| |
| A limited amount of leakage inside primary containment is expected from auxiliary systems that cannot be made 100%
| |
| leaktight. Leakage from these systems should be detected and isolated from the primary containment atmosphere, if possible, so as not to mask RCS operational LEAKAGE detection.
| |
| This LC0 deals with protection of the RCPB from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LC0 include the possibility of a loss of coolant accident.
| |
| FERMI ~ UNIT 2 B 3.4.4-1 Amendment No. 134
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| RCS Operational LEAKAGE B 3.4.4 l
| |
| BASES I APPLICABLE The allowable RCS operational LEAKAGE limits are based on SAFETY ANALYSES the 3redicted and experimentally observed behavior of pipe crac(s. The normally expected background LEAKAGE due to equipment design and the detection capability of the instrumentation for determining system LEAKAGE were also considered. The evidence from experiments suggests that, for LEAKAGE even greater than the specified unidentified LEAKAGE limits, the )robability is small that the imperfection or cract associated with such LEAKAGE would grow rapidly.
| |
| The unidentified LEAKAGE flow limit allows time for corrective action before the RCPB could be significantly compromised. The 5 gpm limit is much less than the expected flow from a critical crack in the primary system piping.
| |
| Crack behavior from experimental programs (Refs. 2 and 3) shows that leakage rates of hundreds of gallons per minute will precede crack instability (Ref. 4).
| |
| The low limit on increase in unidentified LEAKAGE assumes a failure mechanism of intergranular stress corrosion cracking (IGSCC) that produces tight cracks. This flow increase r3 limit is capable of providing an early warning of such
| |
| 'Q) deterioration.
| |
| No applicable safety analysis assumes the total LEAKAGE limit. The total LEAKAGE limit considers RCS inventory makeup capability and drywell floor sump capacity.
| |
| RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii),
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| e m
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| ]
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| FERMI UNIT 2 B 3.4.4 -2 Amendment No. 134
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| L RCS Operational LEAKAGE B 3.4.4 O
| |
| J BASES LCO RCS operational LEAKAGE shall be limited to:
| |
| : a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed. being indicative of material degradation. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAK!GE.
| |
| Violation of this LC0 could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
| |
| : b. Unidentified LEAKE The 5 g)m of unidentified LEAKAGE is allowed as a reasonaale minimum detectable amount that the containment air monitoring. drywell sump level monitoring, and primary containment sump flow monitoring equipment can detect within a reasonable time period. Violation of this LC0 could result in continued degradation of the RCPB.
| |
| : c. Total LEAKAGE The total LEAKAGE limit is based on a reasonable minimum detectable amount. The limit also accounts for LEAKAGE from known sources (identified LEAKAGE).
| |
| Violation of this LCO indicates an unexpected amount of LEAKAGE and. therefore. could indicate new or additional degradation in an RCPB component or system.
| |
| : d. Unidentified LEAKAGE Increase An unidentified LEAKAGE increase of > 2 gpm within the previous 24 hour period indicates a potential flaw in the RCPB and must be quickly evaluated to determine the source and extent of the LEAKAGE. The increase is measured relative to the steady state value: temporary changes in LEAKAGE rate as a result of transient conditions (e.g., startup) are not considered. As such, the 2 gpm increase limit is only applicable in MODE 1 when operating pressures and temperatures are established. Violation of this LC0 could result in continued degradation of the RCPB.
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| O -
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| FERMI: UNIT 2 B 3.4.4-3 Amendment No. 134 L
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| | |
| RCS Operational LEAKAGE B 3.4.4 l
| |
| BASES' APPLICABILITY' In MODES 1, 2, and 3, the RCS operational LEAKAGE LC0 aplies, because tN potential for RCPB LEAKAGE is greatest w1en the reactor is pressurized.
| |
| In MODES 4 and 5. RCS operational LEAKAGE limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are rcduced.
| |
| ACTIONS AJ With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unid!ntified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE: however, the total LEAKAGE limit would remain unchanged.
| |
| B.1 and B.2 An unidentified LEAKAGE increase of > 2 gpm within a 24 hour period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required Completion Time by evaluating service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE. This type piping is very susceptible to IGSCC. For an unidentified LEAKAGE increase greater than required limits (in accordance with LCA 3.0.2),
| |
| an alternative to this evaluation is to reduce the LEAKAGE increase to within limits (i.e., reducing the LEAKAGE rate such that the current rate is less than the "2 gpm increase in the previous 24 hours" limit: either by isolating the source or other possible methods).
| |
| The 4 hour Completion Time is reasonable to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety.
| |
| O -
| |
| l FERMI'' UNIT 2 B 3.4.4-4 AmenNt No.134 l r ---- -
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| J
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| | |
| i RCS Operational LEAKAGE B 3.4.4
| |
| ( BASES ACTIONS (continued)
| |
| C.1 and C.2 If any Require.d Action and associated Completion Time of Cor. .' tion A or B is not met or if pressure boundary LEAKAGE j exists, thei plant must be brought to a MODE in which the LC0 does not aph'y. To achieve this status, the plant must be brought to 400E 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable.
| |
| based on operatine experience, to reach the required plant conditions from f all power conditions in an orderly manner and without challenging plant safety systems.
| |
| SURVEILLANCE SR 3.4.4.1 l i
| |
| REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE (e.g., Primary Containment Atmospheric Gaseous Radioactivity, RPV head flange leak detection. and sump monitoring systems).
| |
| Leakage detection instrumentation is discussed in more
| |
| -(O_/ detail in the Bases for LC0 3.4.6, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates: however, any method may be used to quantify LEAKAGE within the guidelines of Reference 5. In conjunction with alarms and other administrative controls, an 8 hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and for tracking required trends (Ref. 6).
| |
| REFERENCES 1. 10 CFR 50, Appendix A. GDC 30.
| |
| : 2. GEAP 5620. April 1968.
| |
| : 3. NUREG 76/067, October 1975.
| |
| : 4. UFSAR, Section 5.2.7.4.3.3.
| |
| : 5. Regulatory Guide 1.45.
| |
| : 6. Generic Letter 88 01. Supplement 1.
| |
| / %
| |
| l)
| |
| FERMI - UNIT 2 B 3.4.4-5 Amendment No. 134 L
| |
| | |
| .,J P.W 1eakage B 3.4.5 B 3.4 REACTORP.00LANT M5 FEM (RCS)
| |
| B 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage i
| |
| i BASES
| |
| - j The function of RCS PIVs is to separate the high pressure
| |
| )
| |
| BACKGROUl0 l RCS from an attached low pressure system. This protects the i RCS pressure boundary described in 10 CFR 50.2, i 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50. Appendix A (Refs. 1. 2, and 3). PIVs are designed to meet the testing requirements of Reference 4. During their lives, these ,
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| valves can produce varying amounts of reactor coolant j leakage through either normal operational wear or mechanical '
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| deterioration.
| |
| The RCS PIV LC0 allows RCS high pressure operation when l leakage.through these valves exists in amounts that do not )
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| compromise safety. The PIV leakage limit applies to each individual valve. Leakage through these valves is not included in any allowable LEAKAGE specified in LCO 3.4.4.
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| "RCS Operational LEAKAGE."
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| l p Although this specification provides a limit on allowable d PIV leakage rate its main purpose is to prevent overpressure-failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs'between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components.
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| Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed event that could degrade the ability for low pressure injection.
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| A study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can l substantially reduce intersystem LOCA probability. j i
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| I FERMI f UNIT-2 B 3.4.5-1 Amendment No. 134 i
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| E
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| RCS PIV Leakage B 3.4.5 l l
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| h BASES BACKGROUND (continued)
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| PIVs are provided to isolate the RCS from the following connected systems:
| |
| : a. Residual Heat Removal (RHR) System:
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| : b. Core Spray System:
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| : c. High Pressure Coolant Injection System; and
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| {
| |
| : d. Reactor Core Isolation Cooling System.
| |
| The PIVs are listed in Reference 6.
| |
| APPLICABl.E Reference 5 evaluated various PIV configurations, leakage B FETY ANALYSES testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
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| This Specification provides for monitoring the condition of the reactor coolant pressure boundary (RCPB) to detect PIV degradation that has the potential to cause a LOCA outside of containment. RCS PIV leakage satisfies Criterion 2 of 10 CFR'50.36(c)(2)(ii).
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| l LC0 RCS PIV leakage is leakage into closed systems connected to l the RCS. Isolation valve leakage is usually on the order of i drops per minute. Leakage that increases significantly !
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| suggests that something is o)erationally wrong and !
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| corrective action must be tacen. Violation of this LC0 could result in continued degradation of a PIV. which could i lead to overpressurization of a low pressure system and the i loss of the integrity of a fission product barrier.
| |
| The LC0 PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 4) with the exception of the LPCI loop A and loop B outboard injection isolation valves E1150F015A and B (which are allowed a limit of 0.4 gpm through seat and 5 ml/ min external), and LPCI loop A and loop B inboard injection isolation testable check valves E1100F050A and B, (which are allowed a limit G
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| V FERMI - UNIT 2 B 3.4.5-2 Amendment No. 134
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| | |
| RCS PIV Leakage B 3.4.5 BASES LC0 (continued) of 10 gpm) (Ref. 8). These leakage acceptance criteria, along with an external leakage acceptance criteria are l s)ecified for the outboard LPCI injection isolation valves (E1150F015A and B) to assure adequate water seci is maintained inboard of these valves; as such, th? associated primary containment penetrations is classified as a hydrostatically tested penetration in accordance with 10 CFR 50, Appendix J.
| |
| Reference 7 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential). The observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the square root of. the pressure differential.
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| APPLICABILITY In MODES 1, 2, and 3, this LC0 applies because the PIV leakage potential is greatest when the RCS is pressurized.
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| In MODE 3, valves in the RHR shutdown cooling flow 3ath are O not required to meet the. requirements of this LC0 w1en in, or during transition to or from, the RHR shutdown cooling mode of operation.
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| In MODES 4 and 5 leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.
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| ACTIONS The ACTIONS are modi 11ed by two Notes. Note 1 has been !
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| provided to modify tie ACTIONS related to RCS PIV flow !
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| paths. Section 1.3, Completion Times, specifies once a !
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| Condition has been entered, subsequent divisions, '
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| subsystems, components, or variables expressed in the ,
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| Condition discovered to be inoperable or not within limits !
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| will not result in separate entry into the Condition. !
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| Section 1.3 also specifies Required Actions of the Condition i continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.
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| However, the Required Actions for the Condition of RCS PIV O
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| FERMI --UNIT 2 B 3.4.5-3 Amendment No. 134
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| | |
| RCS PIV Leakage B 3.4.5 BASES ACTIONS (continued) leakage limits exceeded provide appropriate com)ensatory i measures for separate affected RCS PIV flow patis. As such, a Note has been provided that allows separate Condition ;
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| entry for each affected RCS PIV flow path. Note 2 requires i an evaluation of affected systems if a PIV is inoperable.
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| The leakage may have affected system OPERABILITY, or isolation of a leaking flow path with an alternate valve may have degraded the sbility of the interconnected system to perform its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are tNen, if necessary, for the affected systems.
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| 86.1 If leakage from one or mm .''.C; eIVs is not within lirit, the flow path must be isolated by at least one closed manual, deactivated automatic, or check valve within )
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| 4 hours.
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| Required Action A.1 is modified by a Note stating that i A. check valves used for isolation must meet the same leakage 1 V requirements as the PIVs (i.e., meet SR 3.4.5.1).
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| Furthermore, the leakage must have been verified at the last i
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| i refueling outage or after the last time the value was disturbed (i.e., maintenance activities that could affect the leak tightness of the valve), whichever is more recent.
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| Four hours provides time to reduce leakage in excess of the !
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| allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours allows time for these actions and restricts the time of operation with leaking valves.
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| B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and MODE 4 within 36 hours. This action may reduce the leakage and also reduces the potential
| |
| , for a LOCA outside the containment. The Completion Times i are reasonable, based on operating experience. to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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| (O
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| .)
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| FERMI UNIT 2 B 3.4.5-4 Amendment No. 134 L
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| | |
| RCS PIV Leakaoe B 3.4.5
| |
| . BASES SURVEILLANCE SR 3.4.5.1 REQUIREMENTS Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve. A reduced leakage acceptance criteria and an external leakage acceptance criteria are specified for the LPCI injection isolation valves, E1150F015 i A and B, to assure adequate waisr 4 maintained inboard of these valves such that the associated primary containment penetration can be classified as a water tested penetration under Appendix J to 10 CFR 50. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs ;
| |
| are not individually leakage tested, one valve may have I failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost. !
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| The Frequency required by the Inservice Testing Program is within the ASME Code, Section XI, Frequency requirement and is based on the need to perform this Surveillance during an ,
| |
| outage and the potential for an unplanned transient if the l
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| /_}
| |
| C Surveillance were performed with the reactor at power. I I
| |
| This SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3.
| |
| Entry into MODE 3 is permitted for leakage testing at high differential 3ressures with stable conditions not possible in the lower 10 DES.
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| J FERMI -UNIT 2 B 3.4.5-5 Amendment No. 134
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| | |
| RCS PIV Leakage l B 3.4.5 BASES
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| )
| |
| REFERENCES 1. 10 CFR 50.2.
| |
| 2, 10 CFR 50.55a(c).
| |
| : 3. 10 CFR 50, Appendix A. GDC 55.
| |
| : 4. ASME, Boiler and Pressure Vessel Code, Section XI.
| |
| : 5. NUREG 0677, May 1980.
| |
| : 6. UFSAR, Table 5.215.
| |
| : 7. NEDC-31339, November 1986.
| |
| : 8. Amendment No. 98 to the Fermi 2 Operating License.
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| )
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| l l
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| l i
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| 1 l
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| t I
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| l l
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| s FERMI UNIT 2 B 3.4,5-6 Amendment No. 134
| |
| | |
| RCS Leakage Detection Instrumentation l B 3.4.6 3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| (G B 3.4.6 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of 10 CFR 50 Appendix A (Ref.1), requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.
| |
| Limits on LEAKAGE from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref. 2).
| |
| Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of leakage rates. The Bases for LC0 3.4.4, "RCS Operational LEAKAGE." discuss the limits on RCS LEAKAGE rates.
| |
| Systems for separating the LEAKAGE of an identified source
| |
| [] from an unidentified Source are necessary to provide prompt v and quantitative information to the operators to permit hem to take immediate corrective action.
| |
| Reference 2 requires three separate unidentified leakage detection systems in the design. Fermi 2 meets this requirement with a drywell floor drain sump flow monitoring system, a supplementary drywell floor drain level monitor, and an airborne gaseous radioactivity monitor. The priinary means of quantifying unidentified LEAKAGE in the drywell is the primary containment sump flow monitoring system.
| |
| The drywell floor drain sump flow monitoring system monitors the LEAKAGE collected in the drywell floor drain sump. This unidentified LEAKAGE consists of LEAKAGE from control rod drives, valve flanges or packings, floor drains, the closed cooling water systems, and drywell air cooling unit condensate drains, and any LEAKAGE not collected in the drywell equipment drain sump. The drywell floor drain sump has transmitters that supply level indications in the main control room.
| |
| The drywell floor drain sump flow monitoring system uses four basic leak detection methods to monitor the drywell floor drain sump. As the water in the sump is pumped out.
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| -p d
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| FERMI'- UNIT 2 B 3.4.6-1 Amendment No. 134 i
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| | |
| i RCS Leakage Detection Instrumentation B 3.4.6 O Bases BACKGROUND (continued) the flow is metered by a flow integrator. Level switches are used to set fill time and pump out time periods using adjustable reset timing devices. If the nominal pumping out )
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| or filling time for the sump is exceeded, an alarm 1s j generated in the control room. In addition, if both pumps l automatically start to handle the flow into the sump, an l alarm is generated. j The supplementary drywell floor drain level monitor provides )
| |
| a continuous analog level measurement of the drywell floor drain level. This sump level monitor provides a rate of-change measurement and alarm. The monitor has the sensitivity to detect a 1 gpm leak integrated over a 1 hour period (Ref. 3).
| |
| The primary containment atmosphere gaseous radioactivity monitoring system continuously monitors the primary containment atmosphere for airborne gaseous radioactivity.
| |
| A sudden increase of radioactivity, which may be attributed to RCPB steam or reactor water LEAKAGE, is annunciated in the control room. The primary containment atmosphere gaseous radioactivity monitoring system is not capable of O* quantifying LEAKAGE rates: however, it is included to meet diversity requirements.
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| 4 APPLICABLE A threat of significant compromise to the RCPB exists if the SAFETY ANALYSES barrier contains a crack that is large enough to propagate rapidly. LEAKAGE rate limits are set low enough to detect I the LEAKAGE emitted from a single crack in the RCPB (Refs. 4 i and 5). Each of the leakage detection systems inside the drywell is designed with the capability of detecting LEAKAGE i less than the established LEAKAGE rate limits and providing appropriate alarm of excess LEAKAGE in the control room.
| |
| A control room alarm allows the o mrators to evaluate the significance of the indicated LEA (AGE and, if necessary, shut down the reactor for further investigation and corrective action. The allowed LEAKAGE rates are well below the rates predicted for critical crack sizes (Ref. 6).
| |
| Therefore, these actions provide adequate response before a significant break in the RCPB can occur.
| |
| RCS leakage detection instrumentation satisfies Criterion 1 of.10 CFR 50.36(c)(2)(ii).
| |
| FERMI - UNIT 2 B 3.4.6-2 Amendment No. 134
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| 'd.
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| | |
| RCS Leakage Detection Instrumentation l B 3.4.6 !
| |
| l BASES LC0 The drywell floor drain sump flow monitoring system is required to quantify the unidentified LEAKAGE from the RCS.
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| The other monitoring systems provide early alarms to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage detection I systems inoperable, monitoring for LEAKAGE in the RCPB is degraded.
| |
| APPLICABILITY In MODES 1. 2, and 3, leakage detection systems are required to be OPERABLE to support LC0 3.4.4. This Applicability is consistent with that for LC0 3.4.4.
| |
| ACTIONS The ACTIONS are modif1ed by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when one or two of the primary containment sump flow monitoring system, primary containment atmosphere gaseous radioactivity monitoring system, and the
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| ('N drywell floor drain sump level monitoring system are
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| (.,/ inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.
| |
| L1 With the drywell floor drain sump flow monitoring system ino>erable, the plant has lost one means to quantify lea (age. However, the primary containment atmosphere gaseous radioactivity monitoring system and the drywell floor drain sum) level monitoring system will provide indication of c1anges in leakage.
| |
| With the drywell floor drain sump flow monitoring system inoperable, but with RCS unidentified and total LEAKAGE being determined every 12 hours (SR 3.4.4.1), operation may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available.
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| G O
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| FERMI - UNIT 2 B 3.4.6-3 Amendment No. 134
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| | |
| RCS Leakage Detection Instrumentation B 3.4.6 BASES ACTIONS (continued) 1L1-With the primary containment atmosphere gaseous radioactivity monitoring system inoperable, grab samples of the primary containment atmosphere must be taken and analyzed to provide periodic leakage information. Provided
| |
| .a sample is obtained and analyzed every 24 hours the plant l may continue operation since at least one other form of drywell leakage detection (i.e. drywell floor drain sump level monitoring system) is available.
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| The 24 hour interval provides periodic information that is adequate to detect LEAKAGE.
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| L.1 With the drywell floor drain sump level monitoring system inoperable, SR 3.4.6.1 must be performed every 8 hours to provide periodic information of activity in the primary containment at a more frequent interval than the routine Frequency of SR 3.4.6.1. The 8 hour interval provides Q
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| (/
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| periodic information that is adequate to detect LEAKAGE and recognizes that other forms of leakage detection are available. However, this Required Action is modified by a Note that allows this action to be not applicable if the primary containment atmosphere gaseous radioactivity monitoring system is inoperable. Consistent with SR 3.0.1.
| |
| Surveillances are not required to be performed on inoperable equipment.
| |
| D.1 and D.2 With both the primary containment atmosphere gaseous radioactivity monitoring system and the drywell floor drain sump level monitoring system inoperable, the only means of detecting LEAKAGE is the drywell floor drain sump flow monitoring system. This condition does not provide the required diverse means of leakage detection. The Required Action is to restore either of the inoperable monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures'that the plant will not be operated in a degraded configuration for a lengthy time period.
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| O FERMI UNIT 2 B 3.4.6-4 Amendment No. 134
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| | |
| RCS Leakage Detection Instrumentation B 3.4.6 l
| |
| BASES
| |
| -ACTIONS (continued) i E.1 and E.2 i l
| |
| If any Required Action of Condition A, B, C, or D cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least I MODE 3 within 12 hours and MODE 4 within 36 hours. The 1 allowed Completion Times are reasonable, based on operating )
| |
| experient.e. to perform the actions in an orderly manner and without challenging plant systems.
| |
| f.d.
| |
| With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate '
| |
| plant shutdown in accordance with LC0 3.0.3 is required.
| |
| SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the sj required primary containment atmosphere gaseous radioactivity monitoring system. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions.
| |
| SR 3.4.6.2 l This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The ,
| |
| test ensures that the monitors can perform their function in I the desired manner. The Frequency of 31 days considers i instrument reliability, and operating experience has shown it proper for detecting degradation.
| |
| SR 3.4.6.3 '
| |
| This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability. Operating experience has proven this Frequency is acceptable.
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| FERMI - UNIT 2 B 3.4.6-5 Amendment No. 134
| |
| | |
| RCS Leakage Detection Instrumentation B 3.4.6 l BASES REFERENCES 1. 10 CFR 50, Appendix A. GDC 30.
| |
| : 2. Regulatory Guide 1.45, May 1973.
| |
| : 3. UFSAR Section 5.2.7.1.3.
| |
| : 4. GEAP 5620, April 1968.
| |
| : 5. NUREG-75/067, October 1975.
| |
| : 6. UFSAR, Section 5.2.7.4.3.3.
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| ]
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| 1 i
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| O FERMI UNIT 2 B 3.4.6-6 Amendment No. 134
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| | |
| t RCS Specific Activity B 3.4.7
| |
| ( B 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| B 3.4.7 RCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment.
| |
| Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CFR 100 (Ref. 1).
| |
| This LC0 contains iodine specific activity limits. The n iodine isotopic activities per gram of reactor coolant are C/ ex)ressed in terms of a DOSE EQUIVALENT I-131. DOSE EQJIVALENT I 131 is calculated using the thyroid dose conversion factors listed in Table III of TID-14844, AEC.
| |
| 1962, " Calculation of Distance Factors for Power and Test Reactor Sites." The allowable levels are intended to limit the 2 hour radiation dose to an individual at the site boundary to a small fraction of the 10 CFR 100 limit.
| |
| APPLICABLE Analytical methods and assumptions involving radioactive SAFETY ANALYSES material in the primary coolant are presented in the UFSAR (Ref. 2). The specific activity in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely.
| |
| While other DBAs assume an initial coolant activity, and can result in offsite doses, the MSLB release forms the basis for determining the worst case offsite doses (Ref. 2). The limits on the specific activity of the primary coolant
| |
| (
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| ~.)
| |
| FERMI - UNIT 2 B 3.4.7- 1 Amendment No. 134
| |
| | |
| RCS Specific Activity B 3.4.7 O =^sts APPLICABLE SAFETY ANALYSES (continued) ensure that the 2 hour thyroid and whole body doses at the site boundary, resulting from an MSLB outside containment during steady state operation, will not exceed a small fraction of the dose guidelines of 10 CFR 100.
| |
| The limits on specific activity are values from a parametric evaluation of typical site locations. These limits are conservative because the evaluation considered more restrictive parameters than for a specific site, such as the location of the site boundary and the meteorological conditions of the site.
| |
| RCS s cific activity satisfies Criterion 2 of 10 CFR 50.36 c)(2)(ii).
| |
| LC0 The specific iodine activity is limited to s 0.2 Ci/gm DOSE EQUIVALENT I-131. This limit ensures the source term assumed in the safety analysis for the MSLB is not exceeded, so any release of radioactivity to the environment during an p
| |
| %J-MSLB is less than a small . fraction of the 10 CFR 100 limits.
| |
| APPLICABILITY. In MODE 1, and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are applicable since there is an escape path for release of radioactive material from the primary coolant to the environment in the event of an MSLB outside of primary containment.
| |
| In MODES 2 and 3 with the main steam lines isolated, such limits do not apply since an escape path does not exist. In MODES 4 and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced.
| |
| ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LC0 DOSE EQUIVALENT I-131 limit, but is s 4.0 C1/gm, samples must be analyzed for DOSE EQUIVALENT I-131 at least once every 4 hours. In addition, the specific activity must be restored to the LC0 limit within 48 hours. The Completion
| |
| :O FERMI UNIT 2 B 3.4.7 -2 Amendment No. 134 i-
| |
| | |
| 1 RCS Specific Activity B 3.4.7 3 ' BASES ACTIONS (continued)
| |
| Time of once every 4 hours is based on the time needed to take and analyze a sample. The 48 hour-Completion Time to restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems.
| |
| A Note to the Required Actions of Condition A excludes the MODE change restriction of LC0 3.0.4. This exception allows entry into the applicable MODE (S) while relying on the ACTIONS even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.
| |
| B.1. B.2.1. B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT I-131 cannot be restored to 5 0.2 pCi/gm within 48 hours, or if at any time it is > 4.0 p
| |
| Q pCi/gm. it must be determined at least once every 4 hours and all the main steam lines must be isolated within 12 hours. Isolating the main steam lines precludes the possibility of releasing radioactive material to the environment in an amount that is more than a small fraction of the requirements of 10 CFR 100 during a postulated MSLB accident.
| |
| l Alternatively, the plant can be placed in MODE 3 within 12 hours and in MODE 4 within 36 hours. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads). ;
| |
| In MODE 4, the requirements of the LCO are no longer ;
| |
| applicable.
| |
| The Completion Time of once every 4 hours is the time needed to take and analyze a sample. The 12 hour Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing '
| |
| the unit in MODES 3 and 4 are reasonable, based on operating experience. to achieve the required plant conditions from full power conditions in an orderly manner and without
| |
| ' challenging plant systems.
| |
| O -
| |
| FERMI - UNIT 2 B 3.4.7 -3 Amendment No. 1.34
| |
| | |
| ~
| |
| l RCS Specific Activity i B 3.4.7 j l BASES SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level.
| |
| This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less.
| |
| REFERENCES 1. 10 CFR 100.11.
| |
| : 2. UFSAR, Section 15.6.4.
| |
| O l
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| l i
| |
| V FERMI UNIT 2 B 3.4.7-4 Amendment No. 134-
| |
| | |
| r RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 B~3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown 1 BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to reduce the temperature of the reactor coolant to s 200*F. This decay heat removal is in preparation for wrforming refueling or maintenance operations, or for (eeping the reactor in the Hot Shutdown condition.
| |
| The two redundant, manually controlled shutdown cooling loops of the RHR System provide decay heat removal. Each loop consists of two subsystems: each subsystem having a separate motor driven pump, but sharing a common heat exchanger, and associated piping and valves. Both loops also have a common suction from the same recirculation loop.
| |
| Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the recirculation loops. The RHR heat exchangers transfer heat to the RHR Service Water System (LC0 3.7.1, " Residual O Heat Removal Service Water (R ESW) System").
| |
| APPLICABLE Decay heat removal by operation of the RHR System in the SAFETY ANALYSES shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. Although the Rm shutdown cooling subsystem does not meet a specific criterion of 10 CFR 50.36(c)(2)(ii), it was identified in the NRC Policy Statement as a significant contributor to risk reduction. Therefore, the RHR Shutdown Cooling System is retained as a Technical Specification.
| |
| LC0 Two RHR shutdown cooling subsystems are' required to be )
| |
| OPERABLE, and when no recirculation pump is in operation, i one shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves. The two subsystems have a common suction source and are allowed to have a common heat exchanger and O
| |
| FERMI - UNIT 2 B 3.4.8-1 Amendment No. 134 l
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 (3
| |
| v BASES LC0 (continued) common discharge piping. Thus, to meet the LCO. both pumps in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. Each shutdown cooling i' subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two i subsystems are required to be OPERABLE to provide i redundancy. Operation of one subsystem can maintain or l reduce the reactor coolant temperature as required. {
| |
| However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
| |
| I Note 1 permits both RHR shutdown cooling subsystems to be j shut down for a period of 2 hours in an 8 hour period. Note !
| |
| 2 allows one RHR shutdown cooling subsystem to be inoperable 1 for up to 2 hours for the performance of Surveillance tests. !
| |
| These tests may be on the affected RHR System or on some
| |
| /~N other plant system or component that necessitates placing V the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.
| |
| l APPLICABILITY In MODE 3 with reactor steam dome pressure below the RHR cut in )ermissive 3ressure (i.e., the actual pressure at which t1e interloc( resets) the RHR System may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant temperature. Otherwise, a recirculation pump is required to be in operation.
| |
| 1 In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing p
| |
| FERMI - UNIT 2 B 3.4.8-2 Amendment No. 134
| |
| | |
| I RHR Shutdown Cooling System-Hot Shutdown !
| |
| B 3.4.8
| |
| () BASES APPLICABILITY (continued) the steam in the main conden.ser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation. j The requirements for decay heat removal in MODE 4 are f discussed in LC0 3.4.9, " Residual Heat Removal (RHR) l Shutdown Cooling System-Cold Shutdown"; and for MODE 5 are j discussed in LC0 3.9.7. " Residual Heat Removal (RHR)-High l
| |
| Water Level": and LC0 3.9.8, " Residual Heat Removal (RHR)-Low Water Level."
| |
| ACTIONS A Note to the ACTIONS excludes the MODE change restriction l of LC0 3.0.4. This exception allows entry into the i a'plicable
| |
| > MODE (S) while relying on the ACTIONS even though j t1e ACTIONS may eventually require plant shutdown. This 1 exception is acceptable due to the redundancy of the l OPERABLE subsystems, the low pressure at which the plant is Q
| |
| V operating, the low probability of an event occurring during operation in this condition. and the availability of alternate methods of deccy heat removal capability.
| |
| A second Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3.
| |
| Completion Times, specifies once a Condition has been ,
| |
| entered, subsequent divisions, subsystems, components or i variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
| |
| A L)
| |
| FERMA - UNIT 2 B 3.4.8-3 Amendment No. 134
| |
| | |
| l l'
| |
| RHR Shutdown Cooling System-Hot Shutdown B 3.4.8 BASES ACTIONS (continued)
| |
| A.1. A.2. and A.3 .
| |
| With one of the two required RHR shutdown cooling subsystem )
| |
| inoperable for decay heat removal, except as permitted by LCO Note 2, the ino)erable subsystem must be restored to OPERABLE status wit 1out delay. In this condition, the remaining 0PERABLE subsystem can provide the necessary decay (
| |
| heat removal. The overall reliability is reduced, however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability.
| |
| Therefore, an alternate method of decay heat removal must be provided.
| |
| With both RHR shutdown cooling subsystems inoperable, an ,
| |
| alternate method of decay heat removal must be 3rovided in f addition to that provided for the initial RHR slutdown k cooling subsystem inoperability. This re-establishes backup I decay heat removal capabilities, similar to the requirements l of the LCO. The 1 hour Completion Time is based on the !
| |
| decay heat removal function and the probability of a loss of the available decay heat removal capabilities.
| |
| The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alterrate method capability. Alternate methods that can be used int.bde (but are not limited to) the Spent Fuel Pool Cooling System and the Reactor Water Cleanup System.
| |
| However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where l MODE 4 is entered.
| |
| 1 I
| |
| D 0 .
| |
| FERMI - UNIT 2 . B 3.4.8 - 4 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown B 3.4.8
| |
| _l BASES ACTIONS (continued)
| |
| B.1. B.2. and B.3 With no RHR shutdown cooling subsystem and no recirculation !
| |
| pump in operation, except as permitted by LC0 Note 1. !
| |
| reactor coolant circulation by the RHR shutdown cooling i subsystem or recirculation pump must be restored without delay.
| |
| Until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. This will provide assurance of continued temperature monitoring capability.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation (7 pump). the reactor coolant temperature and pressure must be U periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| SURVEILLANCE SR 3.4.8.1 i REQUIREMENTS {
| |
| This Surveillance verifies that one RHR shutdown cooling l subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room.
| |
| This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after. clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation. 1 The Note takes exception to the requirements of the bl O
| |
| FERMI - UNIT 2 B 3.4.8-5 Amendment No. 134 i
| |
| | |
| RHR Shutdown Cooling System-Hot Shutdown B 3.4.8
| |
| ) BASES SURVEILLANCE REQUIREMENTS (continued)
| |
| Surveillance being met (i.e., forced coolant circulation is not required for this initial 4 hour period). which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability.
| |
| REFERENCES None.
| |
| l 1
| |
| N FERMI UNIT 2 B 3.4.8-6 Amendment No. 134
| |
| | |
| 7 RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 O
| |
| v B 3.4 REACTOR COOLANT SYSTEM (RCS)
| |
| B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to maintain the temperature of the reactor coolant s 200*F. This decay heat removal is in preparation for wrforming refueling or maintenance operations, or for (eeping the reactor in the Cold Shutdown condition.
| |
| The two redundant, manually controlled shutdown cooling loops of the RHR System provide decay heat removal. Each loop consists of two subsystems: each subsystem having a separate motor driven pump, but sharing a common heat exchanger, and associated piping and valves. Both loops also have a common suction from the same recirculation loop.
| |
| Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor via the recirculation loops. The RHR heat exchangers transfer heat to the RHR Service Water (RHRSW) System.
| |
| APPLICABLE Decay heat removal by operation of the RHR S.* stem in the SAFETY ANALYSES shutdown cooling mode is not required for m'.tigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. Although the RHR Shutdown Cooling System does not meet a specific criterion of 10 CFR 50.36(c)(2)(ii), it was identified in l the NRC Policy Statement as a significant contributor to i risk reduct on. Therefore, the RHR Shutdown Cooling System is retained as a Technical Specification.
| |
| LC0 Two RHR shutdown cooling subsystems are required to be OPERABLE and when no recirculation pump is in operation, one RHR shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO both pumps 3
| |
| (G FERMI - UNIT 2 83.4.9-1 Amendment No. 134
| |
| | |
| F RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 BASES
| |
| {)
| |
| LC0 (continued) in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. In MODE 4 the RliR cross tie valve (E1150 F010) may be opened to allow pumps in one loo) to discharge through the opposite recirculation loop to ma(e a complete subsystem. Additionally, each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two ,
| |
| subsystems are required to be OPERABLE to provide l redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.
| |
| However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly .
| |
| continuous operation is required. l Note 1 permits both RHR shutdown cooling subsystems to be shut down for a period of 2 hours in an 8 hour period.
| |
| Note 2 allows one RHR shutdown cooling subsystem to be r] inoperable for up to 2 hours for the performance of V '
| |
| Surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of l redundancy. l APPLICABILITY In MODE 4, the RHR Shutdown Cooling System may be operated in the shutdown cooling mode tu remove decay heat to maintain coolant temperature below 200*F. Otherwise, a recirculation pump is required to be in operation. However, when decay losses to ambient are sufficient to maintain !
| |
| reactor coolant temperature steady at the existing i temperature the requirements for the RHR Shutdown Cooling l System are not necessary to assure continued safe operation. ,
| |
| In MODES 1 and 2. and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above A.
| |
| FERMI - UNIT 2 B 3.4.9-2 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 BASES APPLICABILITY (continued) this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure. the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.
| |
| The requirements for decay heat removal in MODE 3 below the cut in permissive pressure are discussed in LC0 3.4.8,
| |
| " Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown"; and for MODE 5 are discussed in LC0 3.9.7.
| |
| " Residual Heat Removal (RHR)-High Water Level": and LC0 3.9.8 " Residual Heat Removal (RHR)-Low Water Level."
| |
| , ACTIONS' A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies once a Condition has been entered, O subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional i failure, with Completion Times based on initial entry into i the Condition. However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
| |
| bal With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by LC0 Note 2 the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced.
| |
| Therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This
| |
| 'J FERMI UNIT 2 B 3.4.9-3 Amendment No. 134
| |
| | |
| RHR Shutdown Cooling System-Cold Shutdown B 3.4.9 l
| |
| l n
| |
| BASES (v)
| |
| ACTIONS (continued) e re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method (s) must be reconfirmed every 24 hours thereafter. This will provide assurance of continued heat removal capability.
| |
| The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay l heat removal by ambient losses can be considered as, or !
| |
| contributing to, the alternate method capability. Alternate !
| |
| methods that can be used include (but are not limited to) i the Spent Fuel Pool Cooling System and the Reactor Water Cleanup System.
| |
| B.1. B.2 and B.3 With no RHR shutdown cooling subsystem and no recirculation j
| |
| p) t pump in operation, except as permitted by LC0 Note 1 reactor coolant circulation by the RHR shutdown cooling subsystem or recirculation must be restored without delay.
| |
| ~
| |
| Until RHR or recirculation pump operation is re established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrance involving a loss of coolant circulation. This will provide assurance of continued temperature monitoring capability.
| |
| During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System or recirculation pump).
| |
| the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate.
| |
| p.
| |
| V FERMI UNIT 2 B 3.4.9-4 Amendment No. 134
| |
| | |
| 7 l
| |
| RHR Shutdown Cooling System-Cold Shutdown !
| |
| B 3.4.9 j BASES SURVEILLANCE SR 3.4.9.1 l REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RIE subsystem in the control room.
| |
| REFERENCES None.
| |
| l O
| |
| i
| |
| ~ %.
| |
| (O FERMI - UNIT 2- B 3.4.9-5 Amendment No. 134 i
| |
| | |
| RCS P/T Limits B 3.4.10 ,
| |
| O 8 3.4 acacroa coo'a T svsTeM <acs)
| |
| -B'3.4.10. RCS Pressure and Temperature-(P/T) Limits
| |
| -BASES BACKGROUlO . All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) omrations, power transients, and reactor trips. This _C0 limits the pressure and temperature '
| |
| changes during RCS heatup and cooldown, within the design.
| |
| assumptions and the stress limits for cyclic operation.
| |
| Figure 3.4.101 contains P/T limit curves for hydrostatic or leak testing (Curve A): for heatup by non nuclear means, cooldown following a nuclear shutdown and low power physics tests-(Curve B); and for operations with a critical core other than low power physics tests (Curve C). Other related P/T limits are provided in the SRs.
| |
| Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation 1 is within the allowable region..
| |
| The LC0 establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the !
| |
| component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel. ;
| |
| 10 CFR 50 ' Appendix G (Ref.1). requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III.
| |
| Appendix G (Ref. 2).
| |
| O FERMI - UNIT 2 B 3.4.10 - 1 Amendment No. 134
| |
| | |
| RCS P/T Limits i B 3.4.10 l l
| |
| n d BASES BACKGROUND (continued) l The actual shift in the RT, of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance I with ASTN E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted. '
| |
| l as necessary, based on the evaluation findings and the recommendations of Reference 5.
| |
| l The P/T limit curves are composite curves established by j superimposing limits derived from stress analyses of those l portions of the reactor vessel and head that are the uost restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions, j l
| |
| The criticality limits include the Reference 1 requirement )
| |
| that they be at least 40*F above the heatup curve or the -
| |
| cooldown curve and not lower than the minimum permissible C temperature for the inservice leakage and hydrostatic
| |
| 's testing.
| |
| The consequence of violating the LC0 limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
| |
| C
| |
| (
| |
| FERMI UNIT 2 B 3.4.10 - 2 Amendment No. 134
| |
| | |
| RCS P/T Limits B 3.4.10
| |
| (~3 c) BASES APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA. there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
| |
| RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
| |
| LC0 The elements of this LC0 are:
| |
| : a. RCS pressure, tenperature, and heatup or cooldown rate are within limits during RCS heatup, cooldown, and inservice leak and hydrostatic testing:
| |
| ! b. The tem)erature difference between the reactor vessel bottom lead coolant and the reactor pressure vessel (RPV) steam space coolant is within limit during recirculation pump startup, and during increases in i THERMAL POWER or loop flow while operating at low l THERMAL POWER or loop flow:
| |
| : c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel meets limit during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow with an idle recirculation loop;
| |
| : d. RCS pressure and temperature are within criticality i limits, prior to achieving criticality and '
| |
| : e. The reactor vessel flange and the head flange temperatures are within limits when the reactor vessel head bolting studs are under tension.
| |
| These limits define allowable operating regious and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
| |
| LJ FERMI UNIT 2 B 3.4.10 - 3 Amendment No. 134
| |
| | |
| RCS P/T Limits B 3.4.10 BASES LC0 (continued)
| |
| The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as in)uts for calculating the heatup, cooldown, and inservice lea (age and hydrostatic testing P/T limit curves. Thus, the LC0 for the rate of change of temperature restricts stresses caused by ,
| |
| thermal gradients and also ensures the validity of the P/T l limit curves.
| |
| Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:
| |
| : a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;
| |
| : b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick l vessel walls to become more pronounced): and l
| |
| : c. The existences sizes, and orientations of flaws in the O vessel material.
| |
| I APPLICABILITY The potential for vitlating a P/T limit exists at all times.
| |
| For example, P/T limit violations could result from ambient temperature conditior s that result in the reactor vessel metal temperature be og less than the minimum allowed temperature for bolt m. Therefore, this LC0 is applicable even when fuel is not loaded in the core.
| |
| ACTIONS 6 1 and A.2 Operation outside the P/T limits while in MODES 1, 2 and 3 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
| |
| The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
| |
| A b
| |
| FERMI - UNIT 2 B 3.4.10 -4 Amendment No. 134
| |
| | |
| E .
| |
| RCS P/T Limits B 3.4.10 m)
| |
| (
| |
| v BASES ACTIONS (continued)
| |
| Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The !
| |
| evaluation must verify the RCPB integrity remains acceptable j and must be completed if continued operation is desired. l Several methods may be used, including comparison with pre analyzed transients in the stress analyses, new ,
| |
| analyses, or inspection of the components. !
| |
| ASME Code, Section XI, Appendix E (Ref. 6). may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
| |
| The 72 hour Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.
| |
| Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered.
| |
| The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
| |
| Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
| |
| B.1 and B.2 l If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T l region for an extended period of increased stress, or a l sufficiently severe event caused entry into an unacceptable i region. Either possibility indicates a need for more l careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.
| |
| Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| O FERMI UNIT 2 B 3.4.10 - 5 Amendment No. 134
| |
| | |
| i RCS P/T Limits B 3.4.10 BASES ACTIONS (continued)
| |
| C.1 and C J Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
| |
| Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200*F. Several methods may be used, including comparison with pre analyzed transients, new analyses, or inspection of the components.
| |
| ASME Code, Section XI A)per. dix E (Ref. 6), may be used to support the evaluation: lowever, its use is restricted to evaluation of the beltline.
| |
| Condition C is modified by a Note requiring Required Action I A.2 be completed whenever the Condition is entered. The L Note emphasizes the need to perform the evaluation of the i effects of the excursion outside the allowable limits. l Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
| |
| SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that operation is within the limits of Figure 3.4.10 1 is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.
| |
| Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
| |
| Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
| |
| O V
| |
| FERMI UNIT 2 B 3.4.10 - 6 Amendment No. 134
| |
| | |
| m RCS P/T Limits B 3.4.10 BASES' (O)
| |
| SURVEILLANCE REQUIREMENTS (continued)
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| This SR has been modified with a Note that requires this Surveillance to be performed as applicable only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
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| SR 3.4.10.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified with'' the appropriate limits before withdrawing control rods that will make the reactor critical.
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| Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticaiity provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
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| SR 3.4.10.3. SR 3.4.10.4. SR 3.4.10.5. and SR 3.4.10.6 Differential temperatures within the applicable limits O ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
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| Limiting differential temperatures within the applicable limits, during a THERMAL POWER increase or recirculation flow increase in single loop operation, while THERMAL POWER s 30% RTP or operating loop flow s 50% of rated loop flow, ensure that thermal stresses will not exceed design allowances.
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| Performing the Surveillance within 15 minutes before starting the idle recirculation pump THERMAL POWER increase during single loop operation, or recirculation flow increase during single loop operation, provides adequate assurance j that the limits will not be exceeded between the time of the .
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| Surveillance and the time of the idle pump start, power increase, or flow increase.
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| An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 and
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