ML20098G720

Proposed Tech Specs Allowing Main Steam Line High Radiation Monitor Trip Level Setpoint to Be Raised During Proposed Hydrogen Addition Test

ML20098G720
Person / Time
Site:	FitzPatrick
Issue date:	10/02/1984
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20098G715	List: ML20098G714 ML20098G716 ML20098G720
References
NUDOCS 8410050387
Download: ML20098G720 (10)

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-TABLE 3.1-1'(cont'd) -

REAC'IOR PROTECTION SYSTEM (SCRAM) :INSTRUMENTATIO'I REQUIREMENT '

Minimum No. Modes in Which  : Total of Operable Function Must Be Number of Instrument Operable Instrument

, Channels Trip Function Trip Level Channels Action per Trip Setting l Provided by (1)

System (1) Design for Refuel Startup Run 'Both Trip (6) Systems 2 APR>t Downscale Y 2.5' indicated on X 6 Instrument . A or B scale (9) Channels 2 High Reactor $ 1045 psig X(8) X X 4 Instrument A

. Pressure Channels 2 High Drywell b 2.7 psig X(7) X (7) 'X 4. Instrument A Pressure Channels I 2 Reactor Low Water D 12.5 in indicated X X X 4 Instrument 'A l Level level Channels l ( 1177 in, above the top of active fuel) 3 Iligh Water Level S34.5 gallons per X(2) X X 8 Instrument A in Scram Discharge Instrument Volume Chanrols Volume l 2 Main Steam Line I3x normal full X X X 4 Instrument A l

liigh Radiation power beckground (16) Channels 4 Main Steam Line I 10) valve closure X(3) (5) X(3) (5)X(5) 8 Instrument A Isolation Valve Channels Closure 8410050387 841002 hnendment No. Jd g, fe'I, J'I PER ADOCK 05000333 P PDR

JAFNPP Table 3.1-1 (Ocnt'd) .

REMNOR PIUfECTION SYSTfM (SCRAM) INS'IRLNENIATION RBQUIRDENT .

NOTES OF TABLE 3.1-1 ((bnt'd)

14. 'Ihe APR4 flow hinW high neutron flux signal is fed through a time constant circuit of approxunately 6 secnnds. The APIM fixed high neutron flux signal dvs not incorporate the time constant, but responds directly to instantaneous neutron flux.

15. 'Ihis Average Power Range bbnitor scram function is fixed point and is increased when the reactor node stdtch is placed in the Run position.

16. During the proposed IWen Mdition Test, the normal background radiation level will increase by approximately a factor cf 5 for peak hydrogen concentration. 11mrefore, prior to performance of the test, the Main Steam Line Padiation Fbnitor Trip Icvel Setpoint will be raised to ithree times the increased radiation levels. 'Ihe test will -

be conducted at power levels 5 80% of normal rated power. Duria3 controlled power reduction, the setpoint will be readjusted prior to going below 20% rated power. If-due to a recirculation pmp trip or otler unanticipated power reduction event, the reactor drops below 20% rated power without the setpoint change, control rod withdrawal will be prohioited until the necessary trip setpoint adjustment is made.

l Amendment No. 43a i

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.s-a l JAFtPP ~  :

[ . TABLE 3.2 .

INSTRGENTATION THAT INITIATES PRIMME CONTADMENT ISOIATION -

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I Miniman Nurnber of -

Total Number'of Instr u ent i Operable Instrunent Channels -- -

Channels Provided by Design

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Action i per Trip System (1) Instrunent Trip Ievel Setting for Both Trip Systems -(2) 2 (6) Reactor Iow Water 2 12.5 in Indicated- . 4 Inst. Channels A IcVel Ievel (b 177 in. above -

the top of active fuel) 1 Reactor High Pressure S 75 psig. 2 ' Inst. Gannels D:

(Shutdown Cooling t Isolation) 2 Reactor Iow-Iow 2 -38 in. . indicated 4 Inst. Channels. Ap Water Icvel level ( > 126.5 - in. ' above 5 the top of active fuel) 2 (6) High Drywell Pressure i 2.7 psig 4 Inst. Channels A

! 2 High Rmbntion Main 6 3 x Normal ~ Rated 4 Inst. Channels B Steam Line Tunnel Full Power Background (9) 2 Iow Pressure Main 1825 psig (7) 4 Inst. Channels '

B Steam Idne 2 High Flow Main Steam i 140%-of Rated Steam 4 Inst. Channels B' Idne Flow 2 Main Steam Iane Icak 6 40 F above max 4 Inst. Channels B Detection High anbient Tenperature 3 reactor Cleanup Sys- 6 e0'F above max 6 Inst. Channels C tem Equipnent Area ambient High 'Dmperature 2 Iow Condenser Vacuum D. 8" Hg. Vac (8) .4 Inst. Channels B Closes NSIV's Icendment No. )Ma )*f, fu JW 64

yc JAFNPP Table 3.2-1 (Cont'd)

INSTRUMENTATION TIRT INITIATES PRIMARY CINTAINMENT ISOIATION NOTES EUR TABLE 3.2-1

- 1. hibenever Primary Containment integrity is rcquired by Section 3.7, there shall be two operable or tripped trip systes for each function.

2.. Frtm and after the time it is found that the first column cannot be met for one of the trip systes, 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 mcin steam lines isolated within eight hours.

C. Isolate Reactor Water Cleanup System.

D. Isolate shutdown cooling.

3. Deleted

4. Deleted

5. 'IWo required for each steam line.

6. These signals also start SBGrS and initiate secondary containment isolation.

7. Only required i t run mode (interlocked with Mode Switch) .

-8. Bypassed when reactor pressure is less than 1005 psig and turbine stop valves are closed.

The trip' level setpoint will be maintained at <-3 times normal rated full

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power background. See note 16 to Table 3.1-1 for re-setting trip level setpoint just prior to the Hydrogen Addition Test, and re-setting of the Main Steam Line Padiation Ibnitor for pot.or levels below 20%.

Amenchnent No. Jl 3E', pf 65

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t-source of such radiation to the extent'

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3.1 BASES (cont'd) JAENPp necessary to prevent excessive turinne ,

contamination. Discharge of excessive amounts of radioactivity to tim site.

subchannel. AP M's B, D and F are arranged environs is prevental by the air ejector similarly in the other protection tr19 offgas nonitors thich cause an isolation system. Each protection trip system has of tie main condenser offgas line. During one more APM than is necessary to meet the the Hydrogen Mdition 'Ibst, the normal back-minimum nu s er required per channel. This ground Main Steam Line Radiation Ievel.is allows the bypassing of one APM per protec- expected to increase by a factor of approxi .

tion trip system for maintenance, testing mately 5 at the maximtun hy&% addition or calibration. Mditional IM channels rate as indicated in notr 16, Table 3.1-1.

have also been provided to allow for by- ':he scram setpoint will na reset to three passing of one such channel. The bases for times the projected background radiation the scram setting for the I W , AP M , high level prior to performance of the test.

reactor pressure, reactor low water level, h setpoint will be restored to normal main steam isolation valve (MSIV) closure, following ccmpletion of the hydrogen and generator load rejection, turbine stop addition test.

valve closure are discussed in Sections .

2.1 and 2.2. A Reactor Mode Switch is provided which actuates or bypasses the various scram Instrumentation (pressure switches) for the functions appropriate to the particul e plant drywell are provided to detect a loss of operating status. Reference paragraph 7.2.3.7 coolant accident and initiate the core FSAR.

standby cooling equipnent. A high drywell pressure scram is provided at the same The manual scram function is active in all setting as the Core and Containment Cooling rnodes, thus providing for a manual means of Systems (EOCS) initiation to nunimize the rapidly inserting control rode during all energy which nust be au.umadated during a modes of reactor operation.

loss-of-coolant accident and to prevent return to criticality. This instrumentation The APM (high flux in startup or refuel) is a backup to the reactor vessel water System provides protection against excessive level instrumentation. Power levels and short reactor periods in the startup and intormv h te power ranges.

High radiation levels in the main steam line tunnel above that due to the normal The IM Systcm provides protection against nitrogen and oxygen radioactivity are an short reactor periods in these ranges.

irxlication of leaking fuel. A scram is initiated whenever such radiation level The Control Rod Drive Scram Systen is exceeds three times normal background. designed so that all of the water which The purpose of this scram is to reduce the Amendment No. )( 33

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3.2 BASES (cont'd) .

A High radiation monitors in the main steam line tunnel have been provided to detect The trip. settings of 4.300 l percentiof .de' sign gross fuel. failure as in the control rod flow f r this high flow 'of 43*F above maximum drop accident. With the established set % *" .f r.high temperature are such that..

of 3 times normal background, and main steam unc vering the core is prevented and fission lire isolation valve closure, fission product product release is within limits..

release is limited so that 10 CFR 100 gni h lines are not eacceadaa for this accident.. .%e RCIC high flow and tenperature instrumentation Reference Section 14.6.1.2 FSAR. During the .are arranged the same'as that.for the HPCI.':% e

% Mtim % h mM MW trip setting.of - 300 percent for high. flow and' Main Steam Line Badiation Ievel is expected 40*F:above maximum ambient' for temperature 'are -

to increase by approximately a factor of 5 at based on the same criteria as the HPCI.

l the peak hy&vgen cx)ncentration as indicated in note 16, Table 3.1-1. With the hy&ugen W e reactor. water cleanup ~ system high flow tem-nadition, the fission product release would perature instrumentation'are arranged.similar still be well within the 10 CFR 100 guidelines to that fer the HPCI. W e' trip settings are in the event of a control rod drop accident. such that. uncovering the core is prevented and l

I fission product release is within limits..

Pressure instrumentation'is provided to close the main steam isolation valves in the run mode  % e instrurantation which initiates ECCS action-when the main steam line pressure. drop below is arranged in.a dual bus system. :As for other l 825 psig. W e reactor pressure. vessel thermal

. vital instrumentation arranged in'this fashion, l

i transient due to an inadvertent opening of the specification preserves the effectiveness of the turbine bypass valves when not.in the run the system even during eriods when maintenance mode is less severe than the loss of feedwater or testing is being performed. An exception to analyzed in section 14.5 of the FSAR, therefore, this is when logic functional testing is being.

closure of the main steam isolation valves for performed, thermal transient protection when not in the ,

%e control rod block functions are provided to -

run mode is not required.

prevent excessive control rod withdrawal so that t

%e HPCI high flow and tenperature instru- MTR does not de-l mentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic.

Amendment No. p f[, ,

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ATTACHMENT-II Safety Evaluation of Proposed Change to Technical Specifications - Main Steam Line Radiation Monitor. Trip Level Setpoint Related to Hydrogen Addition Test

- + New York-Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 9

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Section I Description of the Change The proposed change to the Technical Specifications is shown in Attachment I to the Amendment Application. This change occurs in Table 3.1-1 pages 41a and 43a, Table 3.2-1 pages 64 and 65 and in the Bases on pages 33 and 57. The change on the above mentioned pages alters the main steam line radiation monitor trip level setting to take into account the higher background radiation level due to hydrogen addition to the primary coolant.

Section II Purpose of the Change The purpose of the change is to avoid spurious trips of the main steam line radiation monitor during the hydrogen addition test. During this test, the addition of hydrogen reduces the concentration of oxygen in the coolant water and increases the N-16 carryover in the steam. This results in a higher background radiation level seen by the main steam line radiation monitor, which would be above the existing trip setpoint.

Section III Impact of the Change The main steam line radition monitors have only a single design basis which is to initiate a reactor scram and isolate the main steam lines upon detecting high radiation, due to gross fission product release during a control rod drop accident (CRDA).

This is discussed in FSAR Sections 7.2.3.6, 7.3.4.8 and 7.12.

The results of a CRDA are more severe at power levels < 10% as stated in FSAR Section 14.6.1.2 and the hydrogen addition test will be conducted at power levels > 80% of rated power. If, due to a recirculation pump trip or any other unanticipated power reduction event, the reactor power decreases below 20% of rated power, control rod withdrawal will be prohibited until the necessary re-adjustment is made to the trip setpoint.

The licensing basis for the CRDA states that the maximum control rod worth is established by assuming the worst single inadvertent operator error (Reference d). Assuming this operator error, References c and d_ establish the maximum control rod worth above 20% of rated power. Parametric studies utilizing the conservative GE excursion model (Reference b) indicate that the maximum peak fuel enthalpy for a dropped control rod of maximum worth is less than 120 calories per gram. Consequently, the conservatively calculated peak fuel enthalpy for a CRDA above 20% of rated power will have significant margin to the fuel cladding failure threshold of 170 calories per gram.

The proposed change to the Technical Specifications will not alter the conclusions reached in the FSAR and SER accident analyses.

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i' -Th@ Authority concidorc that thic proposed chcnga can be classified as not likely to involve a significant hazards consideration since this proposed change (as per 10 CFR 50.92 (c)(1), (2)_and (3)]:

1. noes not involve a significant increase in the probability or consequences of an accident previously evaluated, because the proposed test would be conducted at power levels ?> 80% of rated power. Above 20% of rated' power, there-is a significant margin between the calculated peak fuel enthalpy and the fuel cladding failure threshold enthalpy. For power levels below 10% of rated power for which the CRDA results become more severe, the trip setpoint will be readjusted to the original setting, as discussed above;

2. does not create the possibility of a new or different

- kind of ' accident from any accident- previously evaluated, because the only function of-these monitors is to detect gross fission product release in the event of a CRDA.- Below 20% of rated power, the monitors would be at their original setting. Above 20% of rated power, there will be a significant margin to the fuel cladding failure threshold; and

3. does not involve a significant reduction in margin of safety because the monitor setpoints will only be changed above 20% of rated power, and a significant margin of safety will still exist.

For these reasons the proposed change is similar to the example (vi) included in Federal Register, Vol. 48 No. 67 dated April 6, 1983, Page 14870. This Commission example is "(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan". The proposed change is similar to this example because the results of the change are within all

.accaptable criteria.

Section IV Implementation of the Change The change as proposed will not impact:

1. Radiation /ALARA Considerations Normal radiation and ALARA practices and procedures will be in effect during the course of the test. Appropriate approved access controls will be implemented for areas subject to the higher radiation levels that result from the test. Dose rate surveys will be conducted and radiation levels will be monitored in order to comply with ALARA requirements.

2. Fire Protection

. Hydrogen monitors.will;be installedJto. detect hydrogen

~ leaks.- Sufficient oxygen will be injected upstream of;the

,recombiner-unit-to assure efficient' hydrogen recombination. In:the unlikely event of any difficulty, the: test could.be terminated immediately.

r/. 3. Environment.

There.will;be no significant impact'on the environment.

Section v 1 Conclusion The. incorporation of.this change; a) will not increase the probability or-the consequences of an accident or malfunction

of: equipment important-to safety as evaluated previously in the

' Safety : Analysis Report; b) will not increase the possibility of an accident'or! malfunction of antype'other than-that evaluated

.previously;in the Safety Analysis Report; c) will not reduce

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the margin'of safety as defined in the Bases for any Technical:

. Specification; d).does not constitute an unreviewed safety.

question and; e) involves;no significant hazards consideration, as defined in.10-CFR 50.92.

s section VI References-a) JAF-FSAR Sections 7.2.3.6, 7.3.4.8,'7.12 and 14.6.1.2 sb) -R.C. Stirn, et al. Rod Drop Analysis for Large Boiling.

Water Reactors,. General Electric Company, March, 1972 (NEDO-10527) c) .R.C.'Stirn,

et al.' Rod Drop Accident Analysis for Large-Boiling Water Reactors Addendum No. 1 Multiple Enrichment Cores with Axial-Gadolinium,-General Electric Company, July,11972-(NEDO-10527, Supplement 1) d)' R.C. Stirn,.et'al. Rod Drop Accident Analysis for-Large

' Boiling Water Reactors Addendum No. 2 Exposed Cores,

-: General Electric Company,-January, 1973 (NEDO-10527, Supplement 2)

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