ML20056C706

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Proposed Tech Specs Eliminating MSIV Closure Function & Scram Function of Main Steam Line Radiation Monitor
ML20056C706
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 07/15/1993
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20056C700 List:
References
JPTS-90-008, JPTS-90-8, NUDOCS 9307220062
Download: ML20056C706 (43)
v • d • e


Text

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.i ATTACHMENT I to JPN-93-050 PROPOSED TECHNICAL SPECIFICATION CHANGES ELIMINATION OF THE MAIN STEAM ISOLATION VALVE CLOSURE FUNCTION  ;

AND SCRAM FUNCTION OF THE MAIN STEAM LINE RADIATION MONITOR JPTS-90-008  ;

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT '

Docket No. 50-333 DPR-59 t

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9307220062 930715 3 DR ADOCK 0500

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3.1 BASES (cont'd) subchannel. APRM's B, D and F are arranged similarly in A Reactor Mode Switch is provided which actuates or the other protection trip system. Each protection trip system bypasses the .arious scram functions appropriate to the has one more APRM than is necessary to meet the particular plant operating status. Reference paragraph minimum number required per channel. This allows the 7.2.3.7 FSAR.

bypassing of one APRM per protection trip system for maintenance, testing or calibration. Additional IRM channels The manual scram function is active in all modes, thus have also been provided to allow for bypassing of one such providing for a manual means of rapidly inserting control channel. The bases for the scram setting for the IRM, rods during all modes of reactor operation.

APRM, high reactor pressure, reactor low water level, main steam isolation valve (MSIV) closure, and generator load The APRM (high flux in startup_ or refuel) System provides rejection, turbine stop valve closure are discussed in protection against excessive power levels and short reactor Sections 2.1 and 2.2.- periods in the startup and intermediate power ranges.

Instrumentation for the drywell is provided to detect a loss of The IRM System provides protection against short reactor -

coolant accident and initiate the core standby cooling periods in these ranges.

equipment. A high drywell pressure scram is provided at the.

same setting as the Core and Containment Cooling Systems (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent retum to criticality. This instrumentation is a backup ,

to the reactor vessel water level instnJmentation.

Amendment No. l}d, gd, gd ..

33

JAFNPP 3.1 BASES (cont'd)

The Control Rod Drive Scram System is designed so that The IRM high flux and APRM .115% power scrams all of the water which is discharged from the reactor by a provide adequate coverage in the startup and intermediate scram can be accommodated in the discharge piping. range. Thus, the IRM and APRM systems are required to ,

Each scram discharge instrument volume accommodates be operable in the refuel and startup/ hot standby modes, in excess of 34 gallons of water and is the low point in The APRM s120% power and flow referenced scrams the piping. No credit was taken for this volume in the provide required protection in the power range (reference design of the discharge piping as concerns the amount of FSAR Section 7.5.7). The power range is covered only by water which must be accommodated during a scram. the APRMs. Thus, the IRM system is not required in the run mode.

During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to The high reactor pressure, high drywell pressure, reactor the piping from the reactor could not be accommodated, low water level and scram discharge volume high level which would result in slow scram times or partial control scrams are required for startup and run modes of plant-rod insertion. To preclude this occurrence, level detection operation. They are, therefore, required to be operational instruments have been provided in each instrument volume for these modes of reactor operation.

which alarm and scram the reactor when the volume of water reaches 34.5 gallons. As indicated above, there is The requirement to have the scram functions indicated in sufficient volume in the piping to accommodate the scram Table 3.1-1 operable in the refuel mode assures that without impairment of the scram times or amount of shifting to the refuel mode during reactor power operation insertion of the control rods. This function shuts the does not diminish the protection provided by the Reactor reactor down while sufficient volume remains to Protection System.

accommodate the discharged water and precludes the situation in which a scram would be required but not be Turbine stop valve closure occurs at 10 percent of valve able to perform its function adequately. closure. Below 217 psig turbine first stage pressure (30 percent of rated), the scram signal due to turbine stop A Source Range Monitor (SRM) System is also provided to valve closure is bypassed because the flux and pressure supply additional neutron level information during startup scrams are adequate to protect the reactor.

but has no scram functions (reference paragraph 7.5.4 FSAR).

t Amendment No. }$,1/4 34

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' JAFNPP ,

s TABLE 3.1-1 (coqt'dl ,

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT l-^ -

Minimum No. Modes in Which Function of Operable . Must be Operable Tota 1 Number of instrument instrument Channels-Channels per - Refuel Startup Run Provided by Design Action -

- Trip System (1)- Trip Function Trip Level Setting' '(6) for Both Trip Systems (1)

E 2 APRM Downscale a 2.5 indicated on 'X 6' instrument Channels ALor B scale (9) 2 High Reactor Pressure s 1045 psig X(8)' X -X '4 Instrument Channels- A 2 High Drywell Pressure - s 2.7 psig X(7) X(7) X. 4 Instrument Channels A ,

2 Reactor Low Water a 177 in. above TAF X X- X - 4 Instrument Channels -A. ,

Level 3 High Water Level in - s 34.5 gallons per -X(2) X X ' 8 Instrument Channels - A-Scram Discharge Volume . Instrument Volume 4 - Main Steam Line'- s.10% valve closure- X(5) . 8 Instrument Channels. A Isolation Valve Closure d

L Amendment No. If,[p,7)l,g,[1/.9/1[~

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JAFNPP - *.

TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENI NOTES OF TABLE 3.1-1-(cont'd)

C. High Flux IRM.

D. Scram Discharge Volume High Level when any control rod in a control cell containing 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.
9. The APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
10. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal complement.
11. See Section 2.1.A.1.
12. The APRM Flow Referenced Neutron Flux Scram setting shall be less than or equal to the limit specified in the Core Operating Limits Report.
13. The Average Power Range Monitor scram function is varied as a function of recirculation flow (W). The trip setting of this function must be maintained as specified in the Core Operating Limits Report.
14. The APRM flow biased high neutron flux signalis fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is place in the Run position.

l . . - -

Amendment No. p, qd, g4, {[/, gd,7/,7[b19 , if, if9, if2 43

e-JAFNPP TABLE 4.1-1 (Cont'd)

. REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT FUNCTIONAL TEST MINIMUM FUNCTIOEAL TEST FRECLUENCIES FOR SAFETY INSTRUMENT AND CONTROL CIRCUITS Instrument Channel Grouo (2)

Functional Test Minimum Frec.uency (3)

Main Steam Line Isolation Valve Closure A Trip Channel and Alarm Once/ month.(1)

Turbine Control Valve EHC Oil Pressure A Trip Channel and Alarm Once/ month.

Turbine First Stage Pressure Pemiissive B Trip Channel and Alarm (4) Once/ month.(1)(8).

Turbine Stop Valve Closure A Trip Channel and Alarm Once/ month.(1)

' NOTES FOR TABLE 4.1-1

1. Initially once every month until acceptable failure rate data are available; thereafter, a request may be made to the NRC to change the test frequency. The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design instrument operates in an environment similar to that of JAFNPP.
2. A description of the three groups is included in the Bases of this Specification.
3. Functional tests are not required on the part of the system that is not required to be operable or are tripped.

If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to retuming the system to an operable status.

4. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a simulated electrical signal into the instrument channels.

Amendment No. l}4,6/,1)$,JM 45

JAFNPP -

TABLE 4.1-2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION '

E l MINIMJJM CALIBBATION FRERtJEMCiES1QELBEA_CIOR N PROTECTION _IN_STRUMENT CHANNEL _S lastrument Channel Group (1) Calibration Minimum Freouency_(2)'

IRM High Flux C Comparison to APRM on Maximum frequency once/ week Controlled Shutdowns APRM High Flux Output Signal B Heat Balance . Daily Flow Bias Signal B Intemal Power and Flow Test Every refueling outage with Standard Pressure Source LPRM Signal B TIP System Traverse Every 1000 effective full power hours High Reactor Pressure B Standard Pressure Source Note (6)

High Drywell Pressure B Standard Pressure Source Note (6)

Reactor Low Water Level B Standard Pressure Source Note (6)

High Water Level in Scram A Water Column, Note (5) Once/ operating cycle, Note (5)

Discharge Instrument Volume High Water Level in Scram B Standard Pressure Source Every 3 months Discharge Instrument Volume Main Steam Une isolation A Note (4) Note (4)

I' Valve Closure Turbine First Stage Pressure B Standard Pressure Source Note (6)

Permissive Amendment No. 4k ,4)$, Q d ; 7[,8k 1/6,~

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TABLE -4.1-2 (Cont'd) .

. REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CAllBRATION IVilNIMuhLCALLBBATION FREQt1EliCIES FOR REACTOR PRQTECTION INSTRUMENT CHANNELS Instrument Channel Group (1) Calibration Minimum Freauency (2)

Turbine Control Valve Fast A Standard Pressure Source Once/ operating cycle :

Closure Oil Pressure Trip Turbine _Stop Valve Closure A Note (4) Note (4)

~ NOTES FOR TABLE 4.1-2

, 1. A description of three groups is included in the Bases of this Specification.-

21 Calibration test is not required on the part of the system that is not required to be operable, or is tripped, but is ' required prior to return to service.

l 3. Deleted

4. Actuation of these switclies by normal means will be performed during the refueling outages.

.5. Calibration shall be performed utilizing a water column or similar device to provide assurance that damage to a float or other portions of the -

l- float assembly will be detected.-

! 6. Sensor calibration once per operating cycle. Master / slave trip unit calibration once per 6 rnonths.

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Amendment No?48,6/,E}8,ik .

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JAFNPP 3.2 BASES (cont'd)

High radiation monitors in the area of t5 main steam lines The trip settings of approximately 300 percent of design flow have been provided to detect gross Rei failure as in the for this high flow or 40 F above maximum ambient for high control rad drop accident. A trip setting of 3 times normal temperature are such that uncovering the core is prevented and full-power background is established to close the main steam fission product release is within limits.

line drain valves, the recirculation loop sample valves, the mechanical vacuum pump isolation valves, and trip the pumps, The RCIC high flow and temperature instrumentation ure to limit fission product release. For changes in the Hydrogen arranged the same as that for the HPCI. The trip settings of Water Chemistry hydrogen injection rate, the trip setpoint may approximately 300 percent for high flow or 40 F above be adjusted based on a calculated value of the expected maximum ambient for temperature are based on the same radiation level. Hydrogen addition will result in an increase in criteria as the HPCI.

the N-16 carryover in the main steam.

The reactor water cleanup system high temperature Pressure instrumentation is provided to close the main steam instrumentation are arranged similar to that for the HPCI. The isolation valves in the run mode when the main steam line trip settings are such that uncovering the core is prevented and pressure drops below 825 psig. The reactor pressure vessel fission product release is within limits.

thermal transient due to an inadvertent opening of the turbine bypass valvas when not in the run mode is less severe than The instrumentation which initiates ECCS action is arranged in a the loss of feedwater analyzed in Section 14.5 of the FSAR, dual bus system. As for other vital instrumentation arranged in therefore, closure of the main steam isolation valves for this fashion, the specification preserves the effectiveness of the thermal transient protection when not in the run mode is not system even during periods when maintenance or testing is required. being performed. An exception to this is when logic functional testing is being performed.

The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping The control rod block functions are provided to prevent of this instrumentation results in actuation of HPCI isolation excessive control rod withdrawal so that MCPR does not valves. Tripping logic for the high flow is a 1 out of 2 logic. decrease to the Safety Limit. The trip Amendment No.1/,3/,3[,p,p,1/4,1[7 57

9 JAFNPP TABLE 3.2-1 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION

. Minimum No.

of Operable Total Number of instrument Instrument Channels Channels Per Provided by Design Trio System (1) Instrument Trio Level Settino for Both trio Systems Actions (2) 2 (6) Reactor Low Water Level a 177 in. above TAF 4 Instrument Channels A 1 Reactor High Pressure s 75 psig 2 Instrument Channels D (Shutdown Cooling isolation) 2 Reactor Low-Low-Low Water Level a 18 in. above the TAF 4 Instrument Channels A 2 (6) High Drywell Pressure s 2.7 psig 4 Instrument Channels A 2 High Radiation Main s 3 x Normal Rated 4 Instrument Channels E Steam Line Tunnel Full Power Background 2 Low Pressure Main Steam Line 2 825 psig (7) 4 Instrument Channels B 2 High Flow Main Steam Line s 140% of Rated Steam Flow 4 Instrument Channels B 2 Main Steam Line Leak s 40*F above max ambient 4 Instrument Channels B Detection High Temperature 4 Reactor Cleanup System Equipment s 40*F above max ambient 8 Instrument Channels C Area High Temperature 2 Low Condenser Vacuum a 8" Hg. Vac (7)(8) 4 Instrument Channels B Closes MSIV's Amendment No. 1/, p, f, f, p,1[3,1[9,1/2

-64

JAFNPP TABLE 3.2-1 (Cont'd)

INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION NOTES FOR TABLE 3.2-1

1. Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
2. From and after the time it is found that the first column cannot be met for one of the trip systems, that trip system shall be tripped or the appropriate action listed below shall be taken.

A. Initiate an orderly shutdown and have the reactor in cold shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Initiate an orderly load reduction and have main steam lines isolated within eight hours.

C. Isolate Reactor Water Cleanup System.

D. Isolate shutdown cooling.

E. - Isolate the main steam line drain valves, the recirculation loop sample valves, and the mechanical vacuum pumps, within eight hours.

3. Deleted
4. Deleted
5. Two required for each steam line.
6. These signals also start SBGTS and initiate secondary containment isolation.
7. Only required in run mode (intertocked with Mode Switch).
8. Bypassed when mode switch is not in run mode and turbine stop valves are closed.

I Amendment No. /,4/,6/, g6,1/2.1[9

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JAFNPP TABLE 4 2-1 MJNIMUM TEST AND CAllBRATION FREQUENCY FOR PCIS Instrument Channel (8) Instrument Functional Test Calibration Frequency. Instrument Check (4)

1) Reac+or High Pressure (1) Once/3 months None (Shutdown Cooling Permissive)
2) Reactor Low-Low-Low Water Level (1)(5) (15) Once/ day
3) Main Steam High Temp. (1)(5) (15) - Once/ day
4) Main Steam High Flow (1)(5) (15) Once/ day
5) Main Steam Low Pressure (1)(5) (15) Once/ day -
6) Reactor Water Cleanup High Temp. (1) Once/3 months None
7) Condenser Low Vacuum (1)(5) (15) Once/ day l 8) Main Steam Une High Radiation (1)(5) (11) Once/ day Logic System Functional Test (7) (9) Frequency
1) Main Steam Une Isolation valves Once/6 months Main Steam Une Drain Valves Reactor Water Sample Valves
2) RHR - Isolation Valve Control Once/6 months Shutdown Cooling Valves
3) Reactor Water Cleanup Isolation Once/6 months
4) Drywell Isolation Valves ' Once/6 months .

TIP Withdrawal Atmospheric Control Valves

5) - Standby Gas t System . Once/6 months Reactor Building isolation l NOTE- See notes o!!owing Table 4.2-5.

' Amendment No. jy/,9 6 , }66, Igt,1%

78 l

_ _ - _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ - - _ . . . .a

JAFNPP .

NOTES FOR TABLES 4.2-1 THROUGH 4.2-5

1. Initially once every month until acceptance failure rate data are 7. Simulated automatic actuation shall be performed once each available; thereafter, a request may be made to the NRC to operating cycle. Where possible, all logic system functional change the test frequency. The compilation of instrument tests will be performed using the test jacks.

failure rate data may include data obtained from other boiling water reactors for which the same design instruments. operate 8.' Reactor low water level, and high drywell pressure are not in a environment similar to that of JAFNPP. included on Table 4.2-1 since they are listed on Table 4.1-2.

2. Functional tests are not required when these . instruments are 9. The logic system functional tests shall include a calibration of not required to be operable or are tripped. Functional tests time delay relays and timers necessary for proper functioning of shall be performed within seven (7) days prior to each startup. the trip systems.-
3. Calibrations are not required when these instruments are not 10. At least one (1) Main Stack Dilution Fan is required to be in required to be operable or are tripped. Calibration tests shall operation in order to isokinetically sample the Main Stack.

be performed within seven (7) days prior to each startup or prior to a pre-planned shutdown. 11. Perform a calibration once per operating cycle using a radiation -

source. Perform an instrument channel alignment once every 3

4. Instrument checks are not required when these instruments months using the built-in current source.

are not required to be operable or are tripped.

12. (Deleted)
5. This instrumentation is exempt from the . functional test definition. The functional test will consist of injecting a 13. Calibration and instrument check surveillance for SRM and IRM simulated electrical signal into the measurement channel, instruments are as specified in Tables 4.1-1,4.1-2,4.2-3.
6. These instrument channels will be calibrated using simulated 14. Functional test is performed once each operating cycle.

electrical signals once every three months.

15. Sensor calibration once'per operating cycle. Master / stave trip unit calibration once per 6 months.

Amendment No. :)4, h,5//,89,181 84

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NOTES FOR TABLE 3.1Q-2 P

(a) Functional tests, calibrations and instrument checks need not be perfonned when these instruments are not required to be operable or are tripped.

(b) Instrument checks shall be performed at least once per day during these periods when the instruments are required to be operable.

(c) A source check shall be performed prior to each release.

(d) Uguid radwaste effluent line instrumentation surveillance requirements need not be performed wuen the instruments are not required as the result of the discharge path .

not being u'.i!!zod. .

(e) An instrument channel calibration shall be performed with known radioactive sources standardized on plant equipment which has been calibrated with NBS traceable standards.

(f) Simulated automatic actuation shall be performed once each operating cycle. Where possible, all logic system functional tests will be pedormed using the test Jacks.

(g) Refer to Appendix A for instrument channel functional test and instrument channel calibration requirements (Table 4.2-1). These requirements are performed as part of main steam high radiation monitor surveillances.

(h) The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.

(i) This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel.

These instrument channels will be calibrated using simulated electrical signals once every three months.

Amendment No. @

39 i

f' .

Attachment ll to JPN-93-050 SAFETY EVALUATION FOR-PROPOSED TECHNICAL SPECIFICATION CHANGES ELIMINATION OF THE MAIN STEAM ISOLATION VALVE CLOSURE FUNCTION AND SCRAM FUNCTION OF THE MAIN STEAM LINE RADIATION MONITOR I. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. Fitzpatrick Technical Specification will eliminate the scram and Main Steam Line Isolation Valve (MSIV) closure functions associated with the Main Steam Line Radiation Monitors (MSLRM). The specific changes are described below:

Page 33,Jases 3.1 Delete the following paragraph:

"High radiation levels in the main steam line tunnel, above levels due to the nitrogen and oxygen radioactivity, are an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds three times normal background.

The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector offgas monitors which cause an isolation of the main condenser offgas line. During the Hydrogen Addition Test, the normal background Main Steam Une Radiation Level is expected to increase by a factor of approximately 5 at the maximum hydrogen addition rate as indicated in note 16, Table 3.1-1. The scram setpoint will be reset to three times the projected background radiation level prior to performance of the test. The setpoint will be restored to normal following completion of the hydrogen addition test."

Move the start of the last paragraph to page 34.

Page 34. Bases 3.1 Revise to accommodate a redistribution of text from page 33.

Page 41a. Table 3.1-1 Delete the Main Steam Line High Radiation scram requirement.

Page 43. Notes of Table 3.1-1 Delete Note 16 and the footnote regarding the proposed hydrogen addition test.

Attachm:nt 11 to JPN-93-050 i

. .,- SAFET( EVALUATION l' L Page 2 of 13 Page 45. Table 4.1-1 Delete the Main Steam Une High Radiation functional test requirement.

Eage 46. Table 4.1-2 Delete the Main Steam Une High Radiation calibration requirement.

Page 47. Table 4.1-2 (cont'd)

Delete Note 3 regarding the calibration of the Main Steam Une High Radiation instrument channel.

Page 57. Bases 3.2 Replace the following paragraph:

"High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop accident. With the .

established setting of 3 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident. Reference Section 14.6.1.2 FSAR. During -

the Hydrogen Addition Test _the normal background Main Steam Une Radiation Level is expected to increase by approximately a factor of 5 at the peak hydrogen concentration as indicated in note 16, Tablo 3.1-1. 'With the hydrogen addition, the fission product release would still be well within the 10 CFR 100  ;

guidelines in the event of a control rod drop accident."

with: 4 "High radiation monitors in the area of the main steam lines have been provided to detect gross fuel failure as in the control rod drop accident. A trip setting of 3 times normal full-power background is established to close the main steam line drain valves, the recirculation loop sample valves, and the mechanical vacuum pump isolation valves, and trip the pumps, to limit fission product release. For changes in the Hydrogen Water Chemistry hydrogen injection rate, the trip setpoint may be adjusted based on a calculated value of the radiation level expected. Hydrogen addition will result in an increase in the N-16 carryover in the main steam."

Attachm:nt 11 to JPN-93-050 SAFETY EVALUATION-Page 3 of 13 Eage 64. Table 3.2-1 Delete the note "(9)" designation after trip level setting for the High Radiation Main i Steam Line Tunnel.

Replace action note "B" for the High Radiation Main Steam Une Tunnel with "E".

Page 65. Table 3.2-1 (cont'd)

Add Note 2.E.

" Isolate the main steam line drain valves, the recirculation loop sample valves,-

and the mechanical vacuum pumps, within eight hours."

Delete Note 9 that reads:

"The trip level setpoint will be maintained at < 3 times normal rated full power background. See note 16 to Table 3.1-1 for re-setting trip level setpoint just prior to and following the Hydrogen Addition Test."

Page 78. Table 4.2-1 Add the following testing requirements for the " Main Steam Une - High Radiation" Instrument Channel:

Instrument Functional Test: "(1)(5)"

Calibration Frequency: *(11)"

Instrument Check: "Once/ day" The proposed calibration frequency for the MSLRM is the same as currently specified in Table 4.1-2 of the Technical Specifications. ' The proposed monthly frequency for the -

instrument functional test is consistent with NUREG 0123, Rev. 3, " Standard Technical Specifications for General Electric Boiling Water Reactors," and with the other PCIS trip functions currently in Table 4.2-1.

j i

Attachment ll to JPN-93-050 SAFETY EVALUATION L

Page 4 of 13 Page 84. Notes for Table 4.2-1 through 4.2-5 Replace Note 8:

" Reactor low water level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2-1 since they are tested on Table 4.1-2." -

with:

" Reactor low water level and high dryweil pressure are not included on Table 4.2-1 since they so li< ted on Table 4.1-2."

Replace Note 11:

"Uses same instrumentation as Main Steam Une High Radiation. See Table 4.1 -2."

with:

" Perform a calibration once per operating cycle using a radiation source.

Perform an instrument channel alignment once every 3 months using the built-in current source."

Apoendix B (Radiological Effluent Technical Soecificationt Page 39. Notes for Iable 3.10-2 Replace the references to " Tables 4.1-1 and 4.1-2 respectively" in note (g) with " Table 4.2 1."

11. PURPOSE OF THE PROPOSED CHANGES The Main Steam Une Radiation Monitor (MSLRM) system consists of four redundant radiation detectors located extemal to the main steam lines outside of primary containment. The monitors are designed to detect a gross release of fission products indicative of fuel failure. The MSLRM currently provides readout, alarm, and trip functions upon detection of excessive radiation levels. A trip initiates a reactor scram, isolates the mechanical vacuum pumps, and initiates a Group i primary containment _ .

Isolation signal for the Main Steam isolation Valves (MSIV), main steam line drain valves, and recirculation loop sample valves.

Elimination of the MSLRM scram and MSIV closure functions was recognized as a generic improvement by the BWR Owners Group, who submitted a safety evaluation justifying removal of these functions to the NRC in May 1987 as Licensing Topical-

Attachment il to JPN-93-050

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SAFETY EVALUATION Page 5 of 13 Report NEDO-31400, titled " Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Une Isolation Valve Closure Function and Scram Function of the Main Steam Une Radiation Monitor" (Reference 1). The report was approved by the NRC in a Safety Evaluation Report sent to the Owners Group on May 15,1991 (Reference 2).

Elimination of these trip functions provides the following safety-related enhancements:

1. Reduction in Scram Frequency Elimination of the trip functions will reduce the exposure to spurious trips. This .

Is especially true when calibration of reactor scram instrumentation, including the main steam line high radiation monitors, is in progress since this activity involves inserting a half scram signal. As noted in NEDO-31400, eight scrams -

have been attributed to the MSLRM trip system since 1980. Failure of one of the MSLRMs was a contributing factor to an automatic reactor scram of the FitzPatrick plant on May 25,1993. This event is described in LER 93-013.

Removing the trips would represent a reduction in transient initiating events. As described in NEDO-31400, an evaluation using MSLRM initiated scram data for BWR plants, results in a 0.3% reduction in the generic core damage frequency probability.

2. Maintain Availability of Condenser Heat Sink <

Elimination of this MSIV closure function will maintain availability of the main condenser as a heat sink to facilitate scram recovery.

3. Elimination of the Potential for Trips Due to Hydrogen Water Chemistry Hydrogen Water Chemistry injection rates influence nitrogen-16 leveis in the main steam lines, introducing the possibility that system transients could momentarily increase nitrogen-16 levels to the point that a high radiation trip could occur.
4. Increased Operator Control Over Radioactive Releases Elimination of this MSIV closure function allows the Steam Jet Air Ejectors to remain operable for a longer period of time. This permits the continued use of the Offgas Treatment System to process radioactivity during transients that may occur.

111. SMETY IMPLICATIONS OF THE PROPOSED CHANGES The bases for approval of the proposed changes is the NRC's Safety Evaluation Report approving the BWR Owners Group Ucense Topical Report NEDO-31400 (Reference 2). The SER concluded that the removal of the MSLRM trips that automatically shutdown the reactor and close the MSIVs is acceptable provided the -!

plant meets three conditions. These conditions are discussed later in this section.

Attachm:nt ll 'o JPN-93-050 SAFETY EVALUATION ,

Page 6 of 13  ;

As stated in NEDO 31400, and indicated in the FitzPatrick Final Safety Analysis Report, the automatic reactor shutdown on the MS RM trip is not given credit in the analysis of any design basis event for BWRs. The FSAR assumes that the MSIVs, on the MSLRM trip, close only in a control rod drop accident (CRDA). However, the Standard Review Plan (SRP) 15.4.9, Rev. 2. July 1981, recommends an assumption that the radioactive contents (noble and lodine) of the coolant resulting from the event are transferred to the condenser and turbine before the MSIVs close. ,

SRP 15.4.9 states that the plant site and dose mitigating engineered safety features are acceptable with respect to the radiological consequences of a postulated control rod drop accident if the calculated whole-body and thyroid doses at the exclusion area '

boundaries are well within the exposure guideline values in 10 CFR 100. "Well within"-

is defined in SRP 15.4.9 as 25% of the 10 CFR 100 exposure guideline values. This translates to guideline values of 75 Rem for thyroid doses and 6 Rem for whole-body doses.

NEDO-31400 confirmed that the radiological release consequence without the automatic MSIV trip is within the NRC's acceptance criteria as stated in SRP 15.4.9.

NEDO-31400 analyzes two scenarios for the CRDA as follows:

Scenario 1:

This is the FSAR bounding scenario that involves automatic MSIV closure. This scenario assumes that the fission product activity is airborne in the turbine and condenser following MSIV closure and leaks directly from the condenser to the atmosphere.

Scenario 2:

This scenario assumes that no automatic MSIV closure occurred and the fission products are transported to an augmented offgas system. The release of the activity to the environment would be from the normal offgas release point after holdup in the treatment system.

Calculations of post accident doses for the site boundary were performed for each of these scenarios to compare radiological consequences with the exposure guidance of SRP 15.4.9. The calculations, using the conservative assumptions of SRP 15.4.9, yield for both scenarios site boundary doses which are a small fraction of 10 CFR 100 and SRP 15.4.9 guidelines. Since the plant specific doses are a function of the plants dispersion coefficient (X/0) and offgas treatment system holdup times, NEDO-31400 provides curves for computing a plant specific value. Fitzpatrick utilizes an augmented offgas treatment system. The specific values of X/O and offgas holdup times for the FitzPatrick Nuclear Power Plant are shown on Table 1.

Attachment 11 to JPN-93-050 t.', -

SAFETY EVALUATION Page 7 of 13 Plant Soecific Analysis.

Site boundary doses were calculated for the James A. FitzPatrick Nuclear Power Plant (Reference 5) for the scenarios with and without MSIV closure (Scenarios 1 and 2, respectively). The site boundary doses are presented in Table 2. Two methods were used to calculate the plant specific site boundary doses. One method used in-house analysis capabilities, the other method used the curves in NEDO-31400. There is good agreement between the in-house calculations and the NEDO-31400 curves.

Plant specific values were used for the power level of the failed fuel rods, X/0, and offgas holdup times. The assumptions are bounded by NEDO-31400. The in-house calculation assumed a more conservative depletion factor for halogens in the turbine and condenser by partitioning and plateout than the NEDO-31400 value (50% instead -

of 90%). The doses presented in Table 2 have been corrected for the 90% depletion assumed in NEDO-31400. Under either scenario, the doses at the site boundary are well within the acceptance criteria of SRP 15.4.9 guidelines. The doses presented in Table 2 for the design basis CRDA were obtained from the FitzPatrick FSAR, Section -

l 14 6.1.2.

The plant specific doses calcu!ations were performed using power uprate initial conditions. These conditions are more conservative for the current licensed power level, and the calculations will not need to be revised to accommodate anticipated NRC approval of power uprate. Power uprate was requested in Technical Specification amendment request, JPTS 91-025 (Reference 6).

NRC Conditions The NRC in a May 15,1991 Safety Evaluation Report concluded that removal of the MSLRM trips that automatically shutdown the reactor and close the MSIVs is 3 acceptable provided the licensee references NEDO-31400 in support of their licensing ,

applications and meets the following conditions:  !

1. "The applicant demonstrates that the assumptions with regard to input values  !

(including power per assembly, X/O, and decay times) that are made in the .l generic analysis bound those for the plant."

NYPA Position ,

f-  !

The assumptions made in the generic analysis (NEDO-31400) bound those used in the plant specific analysis (Reference 5). Plant specific values used in the FitzPatrick analysis are shown on Table 1.

2. "The applicant includes sufficient evidence (implemented or proposed operating procedures, or equivalent commitments) to provide reasonable assurance that increased significant levels of radioactivity in the main steam lines will be contro!!ed expeditiously to limit both occupational doses and environmental releases."

4 L

1.

L __ _

Attachment il to JPN-93-050 SAFETY EVALUATION Page 8 of 13 NYPA Position Reasonable assurance is provided in the plant response to increased radiation I levels as detected by the offgas monitor (located between the Steam Jet Air Ejector (SJAE) and the Offgas treatment system). Abnormal Operating Procedure, AOP-03, "High Activity in Reactor Coolant or Offgas," currently controls the plant response. The offgas monitor is a more sensitive monitor than the MSLRM because the N-16 source, dominating the radiation source surveyed by the MSLRM, has decayed by the time the offgas monitor can be affected by any increased levels of activity. As required by the FitzPatrick Technical Specifications (Appendix B, Specification 3.5), a level of 500 millicuries /second, after a 15 minute delay, will close the SJAE isolation valve, and will prompt a power reduction to retum to acceptable levels within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or a shutdown within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the limit cannot be met. This graded response ensures that actions are taken to limit occupational doses and environmental releases. Prior to modification of the MSLRM trip, plant procedures will be reviewed, and revised as appropriate.

The FitzPatrick Operating Ucense permits bypassing of the Offgas Treatment System during plant startup. NEDO-31400 states that this condition is acceptable provided the offgas radiation monitors are being utilized to automatically isolate the offgas process line. The SJAE isolation feature, previously described, conforms with this NEDO-31400 criteria, and precludes a ,

direct release to the environment, in the event the offgas system is isolated, the offgas dose is equivalent to the FSAR design basis scenario (with MSIV isolation) since in this case the activity is assumed to be transferred to the main condenser, followed by a ground level release. The NRC conditions stipulated in their SER for the offgas radiation monitor will be implemented as describe below for condition 3.

3. "The applicant standardizes the MSLRM nnd offgas radiation monitor alarm setpoint at 1.5 times the nominal nitrogen-16 background dose rate at the monitor location, and commits to promptly' sample the reactor coolant to determine possible contamination levels in the plant reactor coolant and the need for additional corrective actions, if the MSLRM or offgas radiation monitors or both exceed their alarm setpoints."

NYPA Position Concurrent with the modification of the MSLRM trip, the alarm setpo!nts on the MSLRM and offgas radiation monitor will be adjusted to less than or equal to 1.5 times the normal full power N-16 background dose rate (accounting for the increased N 16 carryover due to hydrogen water chemistry). Prior to modification of the MSLRM trip, the plant procedures will be revised to require prompt sampling of the reactor coolant to determine the need for corrective actions, if the MSLRM or offgas radiation monitors, or both, exceed their alarm setpoints.

1 i

Attachm:nt il to JPN-93-050

'C r .

- SAFETY EVALUATION Page 9 of 13 Conclusion There are no adverse safety implications associated with removal of the MSLRM scram and MSIV closure function since the offsite radiation exposure levels without the " '

trips are comparable to those associated with. the trip function, and are well within 10 CFR 100 and SRP 15.4.9 guidelines.' The analysis has been performed in accordance.

with an NRC approved Ucensing Topical Report NEDO-31400 and conforms with conditions imposed by the NRC in a Safety Evaluation Report. The MSLRM isolation - ]

function is retained for the main steam drain valves, the recirculation loop sample - -

valves, and the mechanical vacuum pumps and associated isolation valves. The Technical Specifications for the offgas radiation monitor, and associated trip function, .

are not affected by this application.

t IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed amendment would _i not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or consequences of an accident i previously evaluated.

None of the FitzPatrick design basis events takes credit for reactor scram initiated from the MSLRM. Therefore, elimination of the scram trip signal will:

not increase the probability or consequences of accidents previously evaluated.

The Control Rod Drop Accident (CRDA) is the only design basis event which '

  • assumes that the reactor vessel isolation signal comes from the MSLRM. -The.

isolation trip will not prevent the CRDA from occurring, therefore its elimination . j will not increase the probability of the accident.-

The MSLRM isolation of the MSIVs was intended to mitigate the consequences  !

of a CRDA. The fission products transported to the main condenser before. .

MSIV closure, results in a ground level release due to condenser leakage.' 1' However, NEDO-31400, and the plant specific analysis using the NEDO-31400 assumptions and methodology, demonstrates that the isolation is actually of little benefit in this regards. Without MSIV isolation, the steam Jet air ejector remahs operational, and the fission products are processed through the . .j augmented offgas treatment system. The holdup time, charcoal adsorption, and elevated release, provided by the offgas treatment system, limits the offsite -

exposure levels. The analysis shows the offsite thyroid doses for the CRDA reduced to 'zero without MSIV closure. There is a small increase in the whole  ;

body doses without MSIV closure; however, the conservatively calculated j values are a small fraction of 10 CFR 100 and SRP 15.4.9 guidelines, l H

F Attachment ll to JPN-93-050 -

'L,. ..

SAFETY EVALUATION Page.10 of 13 Therefore, the elimination of the MSLRM isolation trip will not increase the L consequences of an accident previously evaluated.-

2. create the possibility of a new or different kind of accident from those previously evaluated.

The MSLRM scram and MSIV isolation, were originally intended to mitigate, not prevent an accident scenario. Other than the circuitry modifications required to accomplish the removal of the subject trips, no changes to the physical plant or .

to the manner in which the plant is operated are introduced by the requested ~

change. The change does not affect the remaining scram or vessel isolation functions. Therefore, no new or different kind of accident is created.

a 3. involve a significant reduction in the margin of safety.

The Licensing Topical Report NEDO-31400, as approved by the NRC, provides the results of a reliability assessment of the elimination of the MSLRM scram function on reactivity control failure frequency and core damage. frequency. The results of the analysis indicate a negligible increase, on a generic basis, in reactivity control failure frequency with the deletion of the MSLRM scram -

function (1.4 E-9 events / year). However, this increase in reactivity control failure frequency is offset by the reduction in the transient initiating events (inadvertent scrams). This reduction in transient initiating events represents a 0.3% reduction in the generic core damage frequency (Reference 1).

Safe operation of the plant is further enhanced by elimination of the ~

unnecessary scram and subsequent isolation of the reactor vessel. With -

implementation of these changes, the primary heat sink (main condenser) remains available, a large transient on the vessel and safety-related actuations are avoided, and the Offgas Treatment System remains available to control the '_

potential release pathway.

The existing MSLRM and offgas radiation monitoring instrumentation will remain l in service to provide information and alarms to plant operators. In the event either or both of these monitors alarm, the reactor coolant will be promptly sampled to determine activity levels and the need for additional corrective -!

actions. The offgas treatment system isolation trip function on high radiation l remains unaffected by this change. The MSLRM isolation functions, other than for the MSIV's, also remain unaffected by this change'-

For these reasons, the proposed changes will enhance the margin of safety.

V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not affect the ALARA or Fire Protection l Program at the FitzPatrick plant, nor will the changes impact the environment. i q

Q p e.  :-- .m ww

x Attachment il to JPN-93-050.

v , SAFETY EVALUATION - 1 Page 11 of 13 i

l The trip will be removed following NRC approval of the proposed Technical j Specification change.  ;

.VL C_QNCLUSION ,

The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they: ,

1. 'will not change the probability nor the consequences of an accident or.. _  :

malfunction of equipment important to safety as previously evaluated in the i

Safety Analysis Report; 2.

~

will not increase the' possibility of an accident or malfunction of a type different j

from any previously evaluated in the Safety Analysis Report; and ' l

3. will not reduce the margin of rafety as defined in the basis for any technical .

specification.

The changes therefore involve no signification hazards consideration, as defined in 10 '  !

CFR 50.92. -l Vil REFERENCES

1. Licensing Topical Report NEDO-31400, " Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam isolation. Valve Closure Function and Scram .  :

Function of the Main Steam line Radiation Monitor," May 1987.

2. NRC Safety Evaluation report, letter from Ashok C. Thadani, NRC to George J.

Beck, BWROG, " Acceptance for Referencing of Ucensing topical Report NEDO-31400," May 15,1991.

3. James A. FitzPatrick Nuclear Power Plant' Updated Final Safety Analysis Report. -; 1
4. USNRC Standard Review Plan 15.4.9, Rev. 2., July 1981, " Radiological Consequences of Control Rod Drop _ Accident (BWR).
5. New York Power Authority Calculation: JAF-CALC-RAD-00013, ."Radiologicai r Justification for Modification of the Main Steam Line Radiation MonitorTrip ,

Functions," March 13,1992.~

6. NYPA letter, R, E. Beedle to NRC, JPN-92-028, Proposed Changes to the -

Technical Specifications Regarding Power Uprate ( JPTS-91-025), dated June 5,1992.

l l

Ailachment il to JPN-93-050

, , SAFETY EVALUADON  !

Page 12 of 13 j 4

TABLE 1 PLANT SPECIFIC VALUES  ;

JAMES A. FITZPATRICK NUCLEAR POWER PLANT  ;

Offgas System Holduo Times Kryptons: 4.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />  ;

~

Xenons: 106.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Halogens: 100% removal (34,250 lbs. of charcoal in beds)

Site Boundan/ Discersion Coefficients

  • Time interval Conc. X/O Gamma X/O 8

(tn) (sec/m') (sec/m )

Stack Release 0-2" 5.24E-5 4.75E-5 Condenser Release 0-2 1.81 E-4 1.31 E-4 Concentration X/O is for thyroid exposure due to inhalation. Gamma X/O is for whole body extemal gamma exposure from a finite plume.

Fumigation conditions.

Other Plant Specific Values / Assumptions Reactor Power 2585.7 MWt Power Level of Failed Fuel Rods 0.106 MWt/ rod Radial Peaking Factor 1.5 No. of Failed Fuel Rods 850 i Main Condenser Leakage 1%/ day (Scenario 1 only) l Offgas System Flowrate 60 cfm

]

Fission Products Released 10% halogens  !

From Failed Rods 10% noble gases (except Kr-85),30% Kr-85

r '

i.: 4

~'

AttachmInt il to JPN 93-050.

'*e SAFETY EVALUATION-n .. ,

l Page 13 of 13- ,

~ TABLE 2 1 SITE BOUNDARY RADIATION EXPOSURES CONTROL ROD DROP ACCIDENT.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT THYROlD ' WHOLE BODY '.

MODEL (Rem) '(Rem)

SRP 15.4.9 Umit 75 6  :

Design Basis CRDR (FSAR) 3.94 0.242 6

NEDO-31400, Scenario 1* 0.323 (0.32) 0.013" (0.015).

(MSIV isolation) ,

-?

NEDO-31400, Scenarlo 2* O 1.457 (2.2) 3 (No MSIV isolation)  :

First value is from reference 5, adjusted for 90% depletion of halogens in the turtene and condenser, as assumed in NEDO-31400 Value in parenthesis' computed from curves in NEDO-31400. .

0.0123 rem from noble gases, and 0.0006 rem from halogens. -

J 4

1 a

a. . . - .. . . , .. ,

ATTACHMENT 111 to JPN-93-050 MARKUP OF TECHNICAL SPECIFICATION PAGES FOR PROPOSED TECHNICAL SPECIFICATION CHANGES ELIMINATION OF THE MAIN STEAM ISOLATION VALVE CLOSURE FUNCTION AND SCRAM FUNCTION OF THE MAIN STEAfLLil1E RADIATION MONITOR JPTS-90-008 i

r New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

y,- ,

Attachment til to JPN-93-050 INSERTS FOR MMMUP TECHNICAL SPECIFICATION PAGES INSERT A High radiation monitors in the area of the main steam lines have been provided to detect gross fuel failure as in the control rod drop accident. A trip setting of 3 times normal full-power background is established to close the main steam line drain valves, the recirculation

. loop sample valves, and the mechanical vacuum pump isolation valves, and trip the pumps, to limit fission product release. For changes in the Hydrogen Water Chemistry hydrogen injection rate, the trip setpoint may be adjusted based on a calculated value of the expected radiation level. Hydrogen addition will result in an increase in the N-16 carryover in the main steam.

INSERT B isolate the main steam line drain valves, the recirculation loop sample valves, and the mechanical vacuum pump, within eight hours.

INSERT C Perform a calibration once per operating cycle using a radiation source. Perform an instrument channel alignment once every 3 months using the built-in current source.

n ,

JAFNPP 3.1 BASES (cont'd) ,

cubct.s nne l . 'ApRM's B. D and P are arran;ed similarly in the other (contamination. Discharge of excessiva protection trip system. Each protection amounts of radioactivity to the site trip system has one more APRM than is environs is prevented by the air ejector necessary to meet the minimum number offgas monitors which cause an isolation required per channel. This allows the of the main condenser offgas line.

bypasaing of one APRM per protection During the flydrogen Addition Test. the trip system for maintenance, testing or normal background Main Steam Line calibration. Additional IRM channels Radiation Level is expected to increase by a factor of approximately 5 at the have also been provided to allow for bypasaing of one such channel. The maximum hydrogen addition rate as bases for the scram setting for the IBM, indicated in note 16 Table 3.1-1. The APRM, high reactor pressure, reactor low scram setpoint will be reset to three water level, main steam isolation valve times the projected background radiation (MSIV) closure, and generator load re- level prior to performance of the test.

joction, turbine stop valve closure are The setpoint will be restored to normal i discussed in Sections 2.1 and 2.2. following completion of the hydrogen (additiontest. _,s/

Instrumentation for the drywell is provided to detect a loss of coolant A Reactor Mode Switch is provided which accidant and initiate the core standby actuates or bypasses the various scram cooling equipment. A high drywell functions appropriate to the particular pressure scram is provided at the "lant operating status. Reference emme aetting as the Core and Con- ' paragraph 7.2.3.7 FSAR.

tainment Cooling Systems (ECCS) initiation to minimize the energy which The manual scram function is active in cust be accommodated during a loss-of- all modes, thus providing for a manual coolant accident and to prevent return means of rapidly inserting control rods to criticality. This instrumentation is during all modes of reactor operation.

a bacxup to the reactor vessel water IcVel instrumentation. The APRM (high flux in startup or S refuel) System provides protection k filigh radiation levels in the main steam against excessive power levels and short line tunnel, above normal levels due reactor periods in the startup and to the nitrogen and oxygen radioactivity. intermediate power ranges.

A are al. indication of leaking fuel.

A scram The IRM System provides protection is initiated whenever such radiation level against short reactor periods in these oxceecs three times normal background. The ranges.

purpoce of this scram is to reduce the sourc, of such radiation to the extent neces ary to prevent excessive turbine The Control Rod Drive Scram System is designed so that all of the water which

  • Anandment No. J6 pf, 90 33 M o ve_ +o e- 3 4

c s

+

>}

-..j ,

JAFNPP L*

- 3.1 BASES (cont'd) P-ReIca+e.tof .

TYCM ff C, 33 '

is discharged from the_ reactor sc am can be The IBM high fluz and APRM 6.- 15% power scrams provido cccommodated la the. discharge piplag.- Bach scram adequate coverage'la the startup and latermediate

- Cischarge instrument volume accommodates la excess of range. - Thus, the IBM and APRM systems are required 34 gallons of water and is the low point in the

~

to be operable la the refuel.and startup/ hot standby piping. No credit was tekee for this volume la the modes. - The APPM(120% power and flow referenced design of the discharge piping as concerns the amount scrans provide. required protection la the power range cf water which must be accommodated during a scram. (reference FSAR Section 7.5.7). The; power range la l

covered only by the APRMs. Thus, the IBM system is During.moraal operation the discharge volume is not required la the rum mode.

empty; however, should it fill with water,-the water discharged to the piping from the reactor could not g be acccamodated, which would reem1t la slow scram The high reactor pressure,'high drywell pressure,

. reactor low water level had scram discharge. volume times or partial control fod;1asertion. To preclude high level scrans~-are required for startup and run this occurrence,-level detection lastrumments have. modes of pinat operation. They are, therefore, been provided la each'instrusment volume which alarm required to be operational for these modes of reactor-t and scram the reactor when the volume of water reaches operation.

34.5 gallons. As'ladicated above, there is aufficient

. volume la the piping to accosusodate the scram without The requirement to have the scrosa functions ladicated impairment of the acram~ times or amount of insertion la Table 3.1-1 operable in tae refuel modelassures of the control rods. This function shuts the reactor that shifting to the refuel inode durlag reactor pcser-down while sufficient volone ramalas to accommodate operation does.aot diminish the protection provided the discharged water:aad precludes the situation in by the Reactor Protectica System.

- which a scram would be required but act be able to perform its function adequately. Turbine stop valve closure occurs at 10 percent of valve closure. - Below 217 psig turbine first stage A Source Range Monitor-(SBM). System is also provioed 'preseure (30 percent of rated), the~ scram signal due

- to supply additional neutros level information durlag _to turbine stop valve closure is bypassed because the ctartup.but has ao scram functions (reference para- (Ima and pressure scrans are adequate to protect the

  • graph 7.5.4 FSAR). _

reactor.

n a

Amendment No. g . 134

-34 ,

4

-.b ~ v -- r M m

. - < .c_- - , . - - .wm ,....t. -- -.es *.em +-n- ww6-....r- y eeu--* ,n+m we +m-

  • i<w----w --tr== T i -- *.e., -*> - - - = ~ m ae- k -ww. im~.- .e e w y---* e. r e. r , s 4-<m.--+ . -- m m -w

JAFNPP .

TABLE 3.1-1 (cont'd)

BIACIQR_P10TECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Micimum No. Modes la Which Total Number of Operable Trip Level Function Must be of Instrument Icetrument Trip Function Setting 1 Operable Chamaels Action Ch2nnels ,

Provided by (1) per Trip Refuel Startup Run Design for Both Sy: tem (1) (6) Trly Systeme 2 APRM Downscale ),2.5 indicated on X 6 Instrument A or 5 scale (9) Channels 2 High Reactor [1045 psig X(8) X X 4 Instrument A Pressure Channels 2 High Drywell (2.7 psig X(7) X(7) X

  • 4 Instrument A Pressure Chamaels 2 Reactor Low Water 3177 in. above TAF X X X 4 Instrument A Level Chamaels 3 High Mater Level 634.5 gallons per X(2) X X 8 Instrument A in Scram Discharge Instrument Voltsee Chamaels Volume ,

2 Main Stease Line f3x normal full X X X 4 Instrussent A High Radiation power background (16) r%manals O

4 Main Stease Line f10% valve X(5) 8 Imatrument A laolation Valve closure - Chamaels Closure .

Amendsment No. 14, #3, g, 78. FT, 90, % , 122 41a

.[ . w 2

JAFNPP TABLE 3.1-1 (cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT - 0 NOTES OF TABLE 3.1 (cont'd) '

c. High Flux IRM.

D. Scram Discharge Volums High Level when any control rod in a control cell containing fuel is not fully inserted.

E. --- APRM 15% Power Trip.

7. Not required to be operable when @ containment integrity is not required.  ;
8. Not required to be operable when the reactor preneure vessel head is not bolted to the vessel. ,
9. The APRM downecale trip is -R-TC; bypeased when the IRM instrumentation is operable and not high.
10. An APRM will be considered operable if there are et least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal complement.
11. See Section 2.1.A.1.-
-l 12. The APRM Flow Referenced Neutron Flux Scram setting eheil be less then or aquel to the limit specsNed in the Core Operating umits Report. 1
13. The Average Power Range Monitor scram function is verlod as a function of recirculamon How (W). The trip setting of this function rnust be maintained as specited in the Core Operating Limits Report. l
14. The APRM Sow blamed high neutron Dux signal is led through a time constant circuit of approximately 6 seconds. The APRM fixed high' '

r neutron aux signeVoes not incorporate the time' constant, but responds direcey to instantaneous neutron flux.' g

15. This Average Power Range Monitor scram funceton is Exod ooint and is increened when the reactor mode switch is place in the Run position.E

- 16. : *During the proposed Hydrogen Adcstion Test, the background rarnadan level will incrosse by ----wd,r^ r; a factor of 5 for peak hydrogen l ' ,

concentration. Therefore, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to performance of the test, the Main Steam Une Rarumainn Monitor Trip Level Setpoint will be releed to < three amos the ar*ta*M redleelon levels. Upon w,T-f ": ~. of the Hydrogen Addition Test, the setpoint win be readjusted to its C priorestungwithin24 hours.

.i- -

fTNs specification is in effect only during Operating Cycle 10.

O Amendment No. 46,9d,64, Of,60', M. M, las, 3 117, 159.-162 i ~43

____. l ____ __ _______._.__2..

.__._c_._._. _ . . . -, _. ..u. _ __ .. . _ _ . . _- ~ . _ . _ . . _ _ _ _ . _ _ _ _ _ _ _ _. -

s :-

~

JAFNPP ..-

Table 4.1-1 (cont *d)

< - RE Ad'IQR , fRQTECTI QN . SY STEILL SCR AM ) INSTRip(EMT_.IUNCTIQHAL _ TEST-ti1M1 HUM IVlfCI19NAldEST_lREQurycigs roR_SME_TY_1pKTRUtiENT ANILCOFTPQL CIRCUITS -

~ -- - - . --- _ . - -.... .. .,.

1stattwiegt ChpGael Gaug D }_. _Euagdq1mLTe.3L - . _ Mlaintwn.It tgunney_[3 ) ,

Main Steam Line High Radiation B Trip Channel and' Alarm (4) Once/ week.

Main Steam Line Isolation Valve' - A- Trip Channel and Alarm Once/ month. (1)

Closure

. Turbine Control. Valve EHC Oil A Trip Channel and Alarm- Once/ month.

- Pressure Tushine First: Stage Pressure 'B Trip Channel and Alarm -(4) Once/ month. (1)(8)- c Permissive '

Turbine Stop Valve Closure A

' kl-Trip Channel and Alarm Once/ month. (1)  :

.O

-Il MQTES E R_ TABLE 4.1-1 ('

1. Initially once'every montu until. acceptable' failure rate data are availables thereafter, a request may be made to '

the NRC to change the test frequency. --Thelcompliation of instrument failure rate data may include' data obtainnc*. ~

f rom other boiling water reactors _for which the same design instrument operates in an environment similar t.o ' tt:at of JAFMPP. --

2. A description of the three groups'is lacluded'la the Bases of.this Specification.
3. Functional tests are.not required on the part of the system that is'not required to be operable lor are tripped; If tests are missed on parts aot required to be operable or_are tripped, ~ . then they shall be performed prior to

- returning the' system to.an operable status.

. - 4. ' This lastrumentation is esempted.-from the: Instrument' channel test definition. - This instrument channel functional, test will consist.of injecting.a simulated electrical signal'into-the lastrument channels.

Amendment No. ' )[, [ . [ , 136 45 w

g--- --. _-- J.-'__ wree 'w -e %.+r -=?*JW 9 e P' 99'm m1- 4 * *rW'*-wt T9 '*t'- 't-Tea sW

  • WT t'sw -'%1:- 9 -wyr w e 6'e f m- #r we e dr-- 6 4 - u tm , e --**W- ' ou u fm m. A h

lh A'

+

l TABLE 4.1-2 ..

L REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CAUBRATION L i

i MINIMUM CAU8 RATION FREQUENCIES FOR REACTOR PROTECTION INSTRUM (;

t. ;.w. .; cr . ;

oroup 0) cwallon ti Minimum n=. =ey m - '

7I IRM High Flux C Comparison to APRM on .

i Maximum frequency once/ week Controlled Shutdowns APRM High Flux Output Signal B ,HeatBolence Deny

, Flow Bias Signal B Intemel Power and Flow Test Every refuenng outage with Standard Pressure Source LPRM Signal B TIP System Traverse High Reactor Pressure Every 1000 elleceve fue power hows B

Standard PressureSource Note M

- High Drywon Pressure B Standard Pressure Source Note @

Reactor low WHer Level B Standard Pressure Source Note M  ;

High WWer Levelin Scram A oischargeinstrument vosume Weer Column, Note M Once/operaung cycle, Note @

HighWeer Lovdin Scram B onschargeinstrument voeume Standard Pressure Source Every 3 mones t

4 vesveClosure (Main Steam Line;;;. Radiation B Standard Current Source @ Cw/3 monthe Turbine First Stage Pressure B .i PWmisolve Standard Presswo Source Note M ,

t y

..%6 No. g, p W. W. W. W. ; 1M }

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i JAFNPP , ,

TABLE 4.1-2(Cont'd)

REACTOR PROTECTM SYSTEM (SCRAM) RISTRUMENT CAUBRATM unauUm CausRSTON FREOUENCIES FOR REACTOR PROTECTION 9tSTRUMENT CHANNELS Insomment Channel Group (1) Cahbrat60n Minwnum Frequency (2) c Turbine ControlVWwe Fast . A Standard Pressure Source Once/ operating cycle oosure ON Pressure Trip

'l habine StopVisive Closure A 'Noe t (4) Note (4)

NOTES FOR TABt.E 4.12

1. AC /d-iof threegroupsisincludedintheBasesof thisSpecification  ;
2. Cambrosion test is not required on the part of the system that is not required to be operable, or k tripped, but is required prior to retum to sen4ce.
3. [The cament source provides en instrument channel eAgnment. CeRwellon using a rettelton source sher be performed each re
4. Actuellon ci these switmos ty normal means will be performed during the retualing outages.

I 5. CeRwedian shed be performed ullRzing a weser column or simAer device to provide aneurance amoembly wlR be M damage to a nost or other portions of the lloat -

8. Sensor costreelon once per operating cycle. Master / sieve trip unit ceRregion once per 6 months.

Ol s Znsed ' , elete Amendment No. 4, W, W, tjNI, 10

47

_______________________m _ _ , _ -- -.-m . ,r,e4 4

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JAFNPP TwseY 3.2 BASES (cont'd) '

b High rarsahon monitors in the mair; steam line tunnel have The trip 3caurgs of approximately 300 percent of design nowi -

been provided to detect gross fuel failure as in the control rod

- for this high Row or 40"F above maximum'amtnent for high drop mecariant. With the antahkehed settmg of 3 times normal temperature are such that uncovering the core is prevented -

_ background, and main steam line lentatinri valve closure, and Assion product release is within limits. +

nasion product rdesse is limited so that 10 CFR 100 guidehnes we not excaartari for this accident. Reference Section 14.6.1.2 W RCIW h W WWhtwh we '

FSAR.- During the Hydrogen Adc2 tion Test, the normal -arranged the same as that for the HPCI. The trip settings of -

W f background Main Steam Une Rac2ation Level is _ expected to increasa by approximat _ , a factor of 5 at the peak hydrogen -

N as N HPCI'.'

(# , Q concentration as indicated in note 16. Table 3.1-1. With the

~

hydrogen addition, the fiselon product release would stin be' The reactor water cleanup system high te perature

well within the to CFR 100 guidelines in the event of a control instrumentation are arranged similar to that for the HPCI.L The1_ _ ,
rod drop accident. -

trip settings are such that ur.cr.wirg the core is prevented and :

Assi n product release is withm limits.

' is e h m main mesm -

isolation valves in the run mode when the main steam line ' The instrumentation which irwtiates ECCS action is arranged in "

pressure drops below 825 peig. The raarear pressure vessel a dual bus system.; As for other vital instrumentation arranged thermal transeent due to an inadvertent opening of the turtune - in this fashion, the'specificahon preserves the effectiveness of bypass valves when not in the run mode is less severe than the the system even dunng periods when mantenance or. testing is ;

loss of feedwater analyzed in Sechon 14.5 of the FSAR, being performed. An excephon to this is when logic functional 2 therefore, closure of the main steam ientasiart valves for thermal testingis being performed.

' ~

. transient protection when not in the run mode is not required.

-The W rod M Wh we W to pm . 'i The HPCI high Row and temperature instrumentahon are , excessive control rod vnthdrawal so that MCPR does not '.

, provided to detect a break in the HPCI steam piping Tripping decrease to the Safety Umit.lThe trip '

of this instrumentatioit results in actuation of HPCI isolabon valves. Tripping logic for the high flow is a 1 out of 2 logic.

4 1

Amendment No. L4',3f, alf,4tf,90,13( 147 '

57 t .

m... m. . . ~ ., -w.. + .me.-+r- ,,,,w..-w-w e e.3.~ . . . . , -4 ..-

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x ,

g ] , s TABLE 3.2-1 INSTRUMENTAT!ON THAT INITIATES PRIMARY CONTAINMENT ISOL.ATION Minimum No.

of Operable Total Number of instrument instrument Channels Channels Per Provided by Design THp System (1) instrument Trip Level Setting for Both Trip Systems .. Action (2) 2M Reactor LowWester Level 1177 in.above TAF 4 inet. Channels A 1 Reactorligh Pressure 175 poig 2 Inst. Channels D (shutdownCooungisosassorg 2 Reactor Low-Low-Low Noter Level 118in.above TAF 4 Inst. Channels A 2M High DryweE Pressure $2.7 poig 4 Inst. Channels .A 2- High Pare =eart Main < 3 x Normal Reted m E 4 inst. Channels .

Steam UneTunnel Fue Power Background (9) aserk a E ,, J+

Q

).

2 Law Pressure Main Steera une 1825 poig (7) 4 inst. Channels B 2 54WiFlow Main Steam Une : < 140% of Resed Steam Flow 4 Inst. Channels B 2 Main Steam Uno Laek l -

< 407 above max ambient 4 inet. Channels B-rus.nervitegh Ternperature S '

4 Reactor Cleanup System Equipment

-(.

1407 above max ambient 8 inst Channels C

- AreaHighTemperature

(

.2- Low Condenser Vacuum 18".Hg. Vac (7)(8) 4 Inst. Channels B Caooes MSIV's -

g

) -

.e,

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~'

4 JAFNPP TABLE 3.21 (Cont'd) 4 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION .

NOTES FOR TABLE 3.2-1

1. Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
2. From and after the time it is found that the first column cannot be met for one of the trip sy' stems, that trip system shall be tripped or the appropriate action listed below shall be taken.

A. Initiate an orderly shutdown and have the reactor in cold shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Initiate an orderly load reduction and have main steam lines isolated within eight hours.

C. Isolate Reactor Water Cleanup System.

D. Isolate shutdown cooling.

E. %

3. Deleted Ig3erp 3

. 4. Deleted

5. Two required for each steam line.
6. These signals also start SBGTS and initiate secondary containment isolation.
7. Only required in run mode (interlocked with Mode Switch).
8. Bypassed when mode switch is '1ot in run mode and turbine stop valves are closed.

The trip leveisetposin wiii im maintained at < 3 times normal rated full power background. See note 16 to Table 3.1-1 for.re-setting trip level C9. setpoint just prior to and following the Hydrogen Addition Test.

j A

Amendment No. g36, frf, ptf, 122, 159 65

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JAFNPP .. .

TABLE 4.2-1 Mif8 MUM TEST AND CALIBRATION FREQUENCY FOR PCIS L- Instrument Channel (8) instrument Functional Test Calibration Frequency instrument Check (4)

, .1) _ Reactor'High Pressure . (1) Once/3 months - None y -(Shutdown Coolmg Permissivel *

. V1 2) Reactor Low-Low-Low Water Level (1)(5) (15) Once/ day. -

F 3r Mein Steam High Temp. (1)(5) (15) . Once/ day - ~

Mein Steam High Flow - (IH5) . (15) Once/ day "h ^j4)5)

Main Stoom Low Pressure (1H5) (15) .

Once/ day -

' 6) Reactor Water Cloenup High Temp. (1)- Once/3 months None:

7) . Condenser Low Vacuum  : (1HS) (15) Once/ day -

l sj ik h w L s e_fliqk R U d a -(1)(5)' 00 oue/ day.

Logic System Functional Test (7) (9)" Frequency

\

- 1) Mein Steam Line isoletion Valves - Once/6 months J Mein Steam Line Drain Valves A Reactor Water Sample Valves
2) RHR - Isolation Valve Control Once/6 months Shutdown Coolmevalves
3) Reactor Water Cleanup leoletion - Once/6 months
4) Drywell isoletion Valwes Once/6 months'

' TIP Withdrawal Atmospheric Control Valves

5) Standby Gas Treatment System Once/6 months- ,

Reactor Buddmg isoistion NOTE: See notes followmg Table 4.2-5.

~ Amendment No. , i16,'1 1; 1 2.190f

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NOTES FOR TABl_ES 42-1 THROUGH 42-5

7. mA sher be W once e lL 1. iruttally once overy month untN mmaptance failure rate data are

.available; theresher, a request may be made to the NRC to operaung @ h W. aR @ syWem W change the test frequency. The eue *:":-i of instrument tests d be W using h wW p fadure rate data may include data obtained from other boinng 8.- Reactor - low water level. ' hiah ! drywell pressure (snd : hiahl l water reactors for which the same design instruments operate , fradiahon main steam line tunn}elnot included on Table 42 -

in a environment similar to that of JAFNPP. y soce may are tested on Tacle 4.1-2.

2. Functional tests are not required when these instruments are 9. The logic system functional tests shen include a cabbrabon of.

not required to be operable or are tripped. Functional tests. time delay relays and timers necessary for proper functioning

( shen be performed wNhin seven M days prior to each start @, of thetripsystems.

3. Caltrations are not required when these instruments are not 10. At least one (1)' Main Stack DNution Fan is required to be in- . ,

required to be operable or are tripped. CaNbration tests shaN operahon in order to lookinetically sample the Main Stack.

be Pedonned wthin som M days W to each W or 0'

11. ( Uses same instrumentation as ~ Main : Steam .Une' : High i.

@ to a prW h '

Radiation. See Table 4.1-2. ~ ' ' i 4-

~

4. Instrument onecks are not required when these instruments are -

MM [

12.

l not required to be operable or are tripped.

13. Cahbrabon and check surveillanos for SRM and IRM - A -

This instrumentation . is : exempt : from . the functional test 4

5. in Tables 4.1-1,4.1-2,4.2 3*

Instruments are as

'i donnition. . The functional test ' wiR consist of injecting a ~

simulated electrical signal into the measurement channel. ~ 14. Functional test is' once each operating cycle.

6. These instrument channels wel be cabbrated using simulated 15. Sensor cahbrahon once operaung cycle. Master / slave trip electrical signals once every three months. unit cal @ ration once per6 J. %serk C l i

k Amendment No. 34, A,pr,pt, j 181 4 j

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NOTES FOR TA31.E 3.10-2 ,

(a) Tunctional tests, calibrations and instrument checks need not be per-formed when these instruments are not required to be operable or are tripped.

(b) Instrument checks shall be perferned at least once per day during these periods when the instruments are required to be cperable.

(c) A source check shall be performed prior to each release.

(d) Liquid radvaste affluent line instrumentation surveillance requirements need not be performed when the instruments are not required as the result of the discharge path not being utilized. .

(e) An instrument channel calibration shall be performed with known radioac .

tive ' sources standardized on plant equipment which has been calibraced with N35 traceable standards.

(f) Simulated automatic actuation shall be performed once each operating cycle. Where possible, all logic system functional tests vill be per-formed using the test jacks.

(g) Refer to Appendix A for instrument channel functicul test and hatrument channel calibration requirements (Tables 4.1-1 and 4.1-2 respectively).

These requirements are performed a part of main steam high radiation monitor surveillances. _ _

Dserb : I alk 4 2-i (b) The logic system functional tests sha e--s^tehirre cine delay, relays and timers necessary for proper functioning of the trip systems.

(1) This instrumentation is excepted from the functional test definition.

The functional test will consist of injecting a simulated electrical signal into the measurement channel. These instrument channels will be calibrated using simulated electrical signals once every three months.

I I I I

U j l

l A=end=ent No. 93 39 y l

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