ML19351F315
| ML19351F315 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1981 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19351F309 | List: |
| References | |
| NUDOCS 8101120226 | |
| Download: ML19351F315 (16) | |
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O i ATTACHMENT I PROPOSED OPERATING LICENSE ADDITION RELATED TO ! CONTROL ROD DRIVE SCRAM DISCHARGE VOLUME CAPABILITY
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1 1 I ! l l l POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 JANUARY 6, 1981 l r 1< l l. 1" 81013 g 9p
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JAFNPP TABLE 3.1-1 (cont'd) REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT NOTES OF TABLE 3.1-1 (cont'd) C. High Flux IRM
& D. Scram Discharge' Instrument Volume High Level when any control rod in a control cell cont'aining fuel is not fully inserted E. APRM 15% Power Trip
- 7. Not required to be operable when primary containment integrity is not required.
- 8. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
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'Ihe APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
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, 10. An APRM will be considered operable if there are at least 2 LPRMinputs per level and at least 11 LPRM inputs of the normal complement.
- 11. See Section 2.1.A,1.
- 12. This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP) .
where: FRP - Fraction of rated thermal power (2436 MWt) MFLPD
- Maximum fraction of limiting power density where the limiting power density is 18.5 e KW/ft for 7x7 fuel and 13.4 MW/ft for 8x8, 8x8R and P8x8R fuel, The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used 6 l ; .-{
lb/hr) W - Icop; Recirculation flow in percent of rated (rated is 34.2 x 10 4 Sn
- Scram setting in percent of initial
- as a function of recirculation loop i
- 13. The Average Power Range Monitor scram function is varied (Figure 1.1-1)The trip setting o
' .. flow (W). i 43 j Amendment No. / J
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JAFNPP Table 4,1-1 l REACTOR PRCyrECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Minimum Frequency (3) Group (2) Functional Test Each refueling outage. A Place Mode Switch in Shutdown Mode Switch in Shutdown A Trip Channel and Alarm Every 3 months. Manual Scram Trip Channel and Alarm Every refueling outage or A RPS Channel Test Switch after channel maintenance. IRM C Trip Channel and Alarm (4) Once per week during re-High Flux fueling or startup and before each startup. C Trip Channel and Alarm (4) Once per week during re-Inoperative fueling or startup and before each startup. APRM Once/ week. High Flux B Trip output Relays (4) B Trip output Relays (4) Once/ week. Inoperative Once/ week. Downscale B Trip output Relays (4) B Calibrate Flow Bias Signal (4) Once/ month. (1) Flow Bias C Trip Output Relays (4) Once per week during refueling High Flux in Startup or Refuel or startup and before each startup.
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B Trip Channel and Alarm (4) Once/ month. (1) (Instrument check High Reactor Pressure once per day) A Trip Channel and Alarm Once/ month (1) High Drywell Pressure A Trip Channel add Alarm Once/ month (1)
Reactor Low Water Level (5) A Trip Channel and Alarm Once/ month and before each High Water Level in Scram startup (6) , (7) Discharge Instrument volume Trip Channel and Alarm (4) Once/ week. Main Steam Line High Radiation B 44 Amendment No. 42
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JAFNPP Table 4.1-1 (cont'd) REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL S TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR SArtif INSTRUMENT AND C Minimum Frequency (3) Group (2) Functional Test Once/ month. (1) A Trip Channel and Alarm Main Steam Line Isolation Valve Closure once/ month. A Trip Channel and Alarm Turbine Control Valve ENC 011 Pressure Every 3 months. (1) A Trip Channel and Alarm Turbine First Stage Pressure Permissive once/ month. (1) A Trip Channel and Alarm Turbine Stop Valve Closure Every 3 months. A Trip Channel and Alarm Reactor Pressure Permissive NOTES FOR TABLE 4.1-l_ st may be made l
- 1. Initially o. ace every month until acceptable failure rate data are available;thereafter,a tes in an environment requeT to the NRC to change the test frequency.
y obtained from other boiling water reactors for which the same design instrument opera . similar to that of JAFNPP. of this Specification.
- 2. A description of the three groups is included in the Bases ble or are tripped.
Functio 6al tests are not required on the part of the system that is not required to be opera
- 3. h ll be performed prior to If tests are missed on parts not required to be operable or are tripped, then they s a returning the system to an operable status.
, This instrument channel I l j 4. 'Ihis instrumentation is exempted from the instrument channel test definition. fu I
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JAFNPP Table 4.1-1 (cont'd) . REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL TESTS MINIMUM FUNCTIONAL TEST FREQUENCIES FOR' SAFETY' INSTRUMENT AND CO NOTES FOR TABLE 4.1-1 (cont'd) I
- 5. The water level in the reactor vessel will be perturbed and the corresponding level indicator changes will be monitored. This perturbation test will be performed every month after completion of the functional test
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program.
- 6. Functional test of the instruments before each startup is required only if a scram has occurred since the last functional test or calibration.
- 7. The functional test shall be performed utilizing a water column or similar device to provide assurance that
- damage to a float or other portions of the float assembly will be detected.
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JAFNPP Table 4.1-2 REACTOR PRCyrECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CIIANNEL Minimum Frequency Once/ week Group (1) Calibration (4) Instrument Channel Comparison to APRM on Maximum frequency once/ week C IRM High Flux Controlled Shutdowns APRM High Flux Heat Balance Daily B Output Signal B Internal Power and Every refueling outage Flow Bias Signal Flow Test with Stan-dard Pressure Source B TIP System Traverse Every 1000 effective full LPRM Signal power hours Standard Pressure Once/ operating cycle B High Reactor Pressure Source Standard Pressure Every 3 months A High Drywell Pressure Source
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Pressure Standard Every 3 months A Reactor Low Water Level l Water Column, Note (6) Once/ operating cycle, Note (6) , A High Water level in Scram Dis- , charge Instrument Volume ) A Note (5) Note (5) 2 Main Steam Line Isolation Valve Closure Standard Current Every 3 months B
! Main Steam Line High Radiation Source (3)
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Standard Pressure Every 6 months A
. Turbine Plant Stage Pressure Source i Permissive .' Standard Pressure Once/ operating cycle A
A Turbine Control Valve Past Closure Source ' Oil Pressure Trip 46 Amendment No. 42, 4G
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JAFNPP Table 4.1-2 (cont'd) REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMEN Minimum Frequency (2) Group (1) Calibration (4) Instrument Channel Note (5) Note (5) A Turbine Stop Valve Closure Every 6 months A Standard Pressure Reactor Pressure Permissive Source NCyrES FOR TABLE 4.1-2
- 1. A description of three groups is included in the Bases of this Specification.
i d to be operable, or 2. Calibration test is not required on the part of the system that is not requ re is tripped, but is required prior to return to service. Calibration using a radiation source 3.
'Ihe current source provides an instrument channel alignment.
shall be made each refueling outage. be checked once per 4. Response time is not a part of the routine instrument channel test but will operating cycle. f ling outages.
- 5. Actuation of these switches by normal means will be performed during the re ue that Calibration shall be performed utilizing a water column or similar device to provide assurance damage to a float or other portions of the float assembly will be detected.
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JAFNPP TABLE 3.2-3 INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum No. Total Number of of Operable Instrument Channels Action Instrument Trip Level Setting Instrument Provided by Design Channels Per for Both Channels Trip System 6 Inst. Channels (1) 2 APRM Upscale (Flow Biased) S 1 (0.66W+42%)x FRP MFLPD 6 Inst. Channels (1) 2 APRM Upscale (Start-up i 12% Mode)
>2.5 indicated on scale 6 Inst. Channels (1) 2 APRM Downscale 2 Inst. Channels (1) 1 (6) Rod Block Monitor S 1 0.66W+K (8)
(Flow Biased)
>2.5 indicated on scale 2 Inst. Channels (1)
' 1 (6) Rod Block Monitor Downscale 8 Inst. Channels (1) 3 IRM Downscale (2) >2% of full scale 8 Inst. Channels (1) 3 IRM Detector not in (7) Start-up Position 8 Inst. Channels (1)
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3 IRM Upscale 186.4% of full scale
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4 Inst. Channels (1) SRM Detector not in (3)
! 2 (4) l Start-up Position 4 Inst. Channels (1) 2 (4)(5) SRM Upscale <10 5 counts / sec i 1 Inst. Channel (9) (10) l
. 1 Scram Discharge Instrument i 18 gallons j Volume High Water Level
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j. NOTES FOR TABLE 3.2-3
- 1. For the Start-up and Run positions of the Reactor N>de Selector Switch, there shall be two operable or l tripped trip systems for each function. *The SRM and IRM blocks need not be operable in run mode, and 8
Arndent No. 49 72
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JAFNPp TABLE 3.2-3 (Cont'd) . INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3 From and after the time it is the APRM and RBM rod blocks need not be operable in start-up mode. found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested inanediately and daily thereafter, if this condition lasts longer than seven days, the system shall be tripped. From and after the time it is found that the first column cannot be met for both trip systems, the systems shall be tripped.
- 2. IRM downscale is bypassed when it is on its lowest range.
- 3. This function is bypassed when the count rate is > 100 cps.
- 4. One of the four SRM inputs may be bypassed.
- 5. 'Ihis SRM Function is bypassed when the IRM range switches are on range 8 or above.
- 6. The trip is bypassed when the reactor power is < 30%.
- 7. This function is bypassed when the Mode Switch is placed in Run.
- 8. S = Rod Block Monitor Setting in percent of initial.
6 I lb/hr). W=Loopbecirculationflowinpercentofrated,(ratedlooprecirculationflowis34.2x10 K = Intercept values of 39%, 40%, 41% and 42% can be used with appropriate MCPR limits from Section 3.1.B. 0 and the reactor water temperature is less than 212 F, the control
- 9. When the reactor is suberitical rod block is required to be operable only if any control rod in a control cell containing fuel is not fully inserted.
- 10. When the control rod block function associated with scram discharge instrument volume high water level is not operable when required to be operable, the trip system shall be tripped.
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T JAFNPP . TABLE 4.2-3 MINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION Instrument Functional Test Calibration Instrument Check (9) Instrument Channel Once/3 months once/ day (1) (3)
- 1) APRM - Downscale Once/3 months once/ day
- 2) APRM - Upscale (1) (3)
(3) (2) (2)
- 3) IRM Upscale (2)
(3) (2) (2)
- 4) IRM - Downscale (2)
Once/3 months once/ day l RBM - Upscale (1) (3) 5) Once/3 months Once/ day l (1) (3)
- 6) RBM - Downscale (2) (2)
- 7) SRM ' - Upscale (2) (3)
(2) (3) (2)
- 8) SRM - Detector Not in Stattup Position (2) (3) (2)
- 9) IRM - Detector Not in Startup Position Scram Discharge Instrument Volume - High Once/ month (2 ) Once/ operating Cycle (2) N/A l
.10) water level Il Logic System Functional Test (4 ) (6) Frequency [
- 1) System Logic Check Once/6 months t
NOTE: See listing of notes following Table 4.2-6 for the notes referred to herein. l
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JAFNPP 4.3 (cont'd) 3.3 (cont'd)
- a. Each partially or fully
- a. Control rods which withdrawn operable control cannot be moved with rod shall be exercised one control rod drive notch at least once each pressure shall be week when operating above 30 considered inoperable, percent power. In the event If a partially or fully power operation is continuing withdresn control rod with three or more inoperable drive cannot be moved control rods, this test shall with drive or scram be performed at least once each pressure, the reactor day, when operating above 30 shall be brought to percent power.
the Cold shutdown con-dition within 24 hours b. The scram discharge volume drain and shall not be started and vent valves shall be verified unless (1) investigation open at least once per 31 days has demonstrated that the (these valves may be closed inter-cause of the failure is mittently for testing under admin-not a failed control rod istrative control). drive mechanism collet l housing, and (2) adequate c. A second licensed operator
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shutdown margin has been shall verify the conformance demonstrated as required to Specification 3.3.A.2.d by Specification 4.3.A. before a rod may be bypassed in the Rod Sequence Control If investigation demonstrates System, that the cause of control l rod failure is a cracked d. Once per week check status collet housing, or if this of pressure and level alarms
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possibility cannot be ruled for each accumulator. out, the reactor shall not be started until the affected ll control rod drive has been lg ' replaced or repaired. 89 Amendment No. /
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- e. When it is initially determined that l
- b. The control rod e. control rod is incapable of normal directional control insertion, an attempt to fully insert valves for inoperable the control rod shall be made. If control rods shall be the control rod cannot be fully disa: aed electrically. inserted
- c. Control rods with shutdown margin test shall be made scram times greater to demonstrate under this condition than those permitted that the core can be made suberitical by for any reactivity condition during
' Specification 3.3.C.3 the remainder of the operating cycle are inoperable, but with the analytically determined, if they can be highest worth control rod capable of inserted wiv.h control withdrawal, fully withdrawn, and all rod drive pressure other control rods capable of inser-they need not be tion fully inserted. If Specification , disarmed 3.3.A.1 and 4.3.A.1 are met, reactor electrically. startup may proceed.
- d. Control rods with a failed " Full-in" or
" Full-out" position switch may be bypassed in the Rod Sequence Control System and considered operable if the actual rod position
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must be moved in
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i amenament no. A 8'"
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JAFNPP . 4.3 (cont'd) 3.3 (cont'd)
- 2. At 8-week intervals,15 percent of
- 2. The average of the scram insertion the operable control rod drives shall times for the three fastest operable be scram timed above 950 psig. When-control rods of all groups of four ever such scram time measurements are control rods in a two-by-two. array made, an evaluation shall be made to shall be no greater thans provide reasonable assurance that Average Scram proper control rod drive performance Control Rod is being maintained.
Notch Position Insertion Time Observed (Sec)
- 3. All control rods shall be determined 0.361 operable once each operating cycle 46 by demonstrating the scram discharge 38 0.977 volume drain and vent valves operable 24 2.112 when the scram test initiated by 04 3.764 placing the mode switch in the SHUTDOWN position is performed as required by Table 4.1-1 and by verifying that the drain and vent valves:
- a. Close within 80 seconds after receipt of a signal for control rods to scram , and
- b. Open when the scram signal is reset or the scram discharge instrument volume trip is bypassed.
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Amendment No. AF
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ATTACHMENT II
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SAFETY EVALUATION RELATED TO SCRAM DISCHARGE VOLUME 4 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT
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' DOCKET NO. 50-333
'JAl$ARY 6, 198r
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Section I - Description of Modification The modification provides surveillance requirements for SDV vent and drain valves and LCO/ surveillance requirements for the RPS and control rod block SDIV limit switches, in accordance with the NRC letter dated July 7, 1980 to all BWR Licensees. Section II - Purpose of the Modification The purpose of the modification is to ensure that the SD reactor operation. Section III - Impact of the Change These modifications will not alter the conclusion reached in the FSAR and SER accident analysis. Section IV - Implementation of the Modification The modification as proposed will not impact the Fire Protection Program at JAF. Section V - Conclusion The incorporation of these modifications: a) will not increase the probability nor the consequences of an accident as previously evaluated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different typeand than c) any evaluated previously indoes the Safety Analysis Repor not constitute an for any Technical Specification, and d) unreviewed safety question. Section VI - References (a) JAF FSAR (b) JAF SER .
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