ML20100J260

Proposed Tech Spec,Revising Table 3 3.1-1 & Sections 3.10.A.1 & 3.10.A.8 Re Refuel Interlocks & Control Rod Blocks

ML20100J260
Person / Time
Site:	FitzPatrick
Issue date:	12/06/1984
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
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ML20100J198	List: ML20100J194 ML20100J248 ML20100J260
References
NUDOCS 8412100267
Download: ML20100J260 (12)

Text

-. -

ATTACHMENT I

' Proposed Technical Specification Changes' Related to Refuel. Interlocks and Control Rod Blocks New York Power' Authority James A.

FitzPatrick Nuclear Power Plant Docket No. 50-333 dated December-6, 1984 8412100267 841206 PDR ADOCK 05000:133 P

PDR q

l

JAFNPP TABLE 3.

1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION-REQUIREMENT Minimum No.

Modes in-which Total of Oporable

' Trip Level Function Must be Number of Action Instrument Trip Function Setting Operable Instrument (1)

Channels Channels por. Trip Refuel Startup Run-Provided Cyctea (1)

(6) by Design (16) for Both l

Trip ~ Systems 1

Mode Switch in X

X X

1 Mode Switch-A Shutdown (4 Sections) 1 Manual Scram X

X 2 Instrument A

Channels 3

IRM High Flux 4 120/125 of full scale X

X 8 Instrument A

Channels 3

IRM Inoperative X

X 8 Instrument A

Channels 2

APRM Neutron Flux-fl5% Power X

X 6 Instrument A

Startup (15)

Channels-2 APRM Flow Referenced S &

X X

6 Instrument A or B Neutron Flux (12)(13)(14)(0.66W+54%)x Channels (Not to exceed 117%)

I FRP

~

L MFLPD 2

APRM Fixed'High Neutron 6120% Power X

6 Instrument A or B-Flux (14)'

Channels-2 APRM Inoperative (10)

X X

X 6 Instrument A or B Channels Amand=ent No. 1 7

41 m

1 JAFNPP TABLE 3.

1-1 REACTOR-PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.

Modes in Which Totals of Operable Trip Level Function Must be Number;of.

-Action Inctrument' Trip Function Setting Operable Instrument (1)

Channels Channels per Trip Refuel Startup Run.

Provided Cysten (1)'

(6) by Design (16).

for-Both-l Trip Systems

-2 APRM Downscale 27 2. 5 indicated

'X 6 Instrument. A or B on scale (9)

Channels

'2 High' Reactor 6 1045 psig x(8)

X X

4' Instrument. A Pressure Channels.

2 High Drywell 6.2.7.psig x(7) x(7).

X 4 Instrument A

Pressure Channels 2

Reactor Low Water it12.5 in.

X X

X 4 Instrument

'A Level indicat ed ' level-Channels

\\

(

177 in. above the top of active fuel)L 3

High Water Level 6: 34.5 gallons peri 'X(2)

X-X 8 Instrument-A in Scram Discharge Instrument Volume Channels Volume 2

Main Steam line 6 3x normal full X

X X

4 Instrument A

High Radiation power-background.

Channels 4

Main Steam line 6 10% valve X(3)(5)- X(3)(5)

X(5)

8. Instrument A

Isolation Valve closure Channels Closure

,Auendment No; 41a m

=

JAFNPP TABLE 3. 1-1 REACTOR PROTECTION SYSTEM-(SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.

Modes in Which Total of Operable Trip Level Function Must be Number of-Action Instrument Trip Function

. Setting Operable Instrument.

(1)

Channels Channels per Trip Refuel Startup Run Provided System (1).

(6) by Design (16) for Both Trip Syst'eus 2

Turbine Control 500 P 850=psig X(4) 4 Instrument A or C Valve Fast Closure

. Control oil pressure Channels between fast closure solenoid and disc dump valve A.endment No.

41b c

q

-i JAFNPP TABLE 3.

1-1'(cont'd)

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT-

'Minicum No.

Modes in Which Total of' Operable Trip Level

. Function Must be' Number of Action Inotrument Trip. Function Setting Operable Instrument (1)

Channels Channels.

por Trip Refuel Startup Run Provided Cyotsa (1).

(6)-(16):

by Design l

for Both Trip Systems 4

Turbine Stop.

10 % valve X(4)(5) 8 Instrument..A or.C.

Valve Closure closure Channels.

NOTES OF TABLE 3.1-1

1. There shall-be two operable or tripped trip' systems for each function, except as specified-in 4.1.D.

From and after.the time that.the. minimum number of operable instrument channel for a trip system cannot be met,.that,affected trip system shall be placed in the~ safe (tripped) condition, or the approp,riate actions listed below.shall'be taken.

A.

Initiate insertion of operable rods and complete insertion of all operable rods within.four hours.

B.

Reduce power level to IRM range and place Mode Switch in the Startup Position within eight hours.

C.

Reduce power to less than 30 percent of rated.

2 Permissible to bypass, in Refuel and Shutdown; positions of the Reactor Mode Switch'.

3.

By passed when reactor pressure is 1005 psig..

4 Bypassed when turbine first stage pressure is less.than 217 psig or less'than 30 percent of rated.

5.

The design permits closure of any two. lines without a scram being initiated.

6. When the reactor is subcritical and the ~ reactor water temperature is less'than 212*F, only the following trip functions need to be operable:

A.

Mode Switch in Shutdown B.

Manual Scram 42 3

Amendment No.

e

-JAFNPP TA3LE 3.

1-1: (cont'd)

REACTOR PROTECTION SYSTEM-(SCRAM) INSTRUMENTATION REQUIREMENT NOT3S OF TABLE 3.1-1 (cont'd) 14 The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds.

The APRM fixed high' neutron flux singal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

15.

This Average Power Range Monitor scram function is fixed point ind is increased when'the reactor mode switch is place in.the Run position.

16.

When all rods are full-in and electrically disarmed, the reactor protection system _need.

not be operable.

Anond ent No.

43a t.

~.

w, JAFNPP-

.n

'.10 ~ cLIMITING CONDITIONS FOR 4.10'.' SURVEILLANCE REQUIREMENTS 3

OPERATION, 4.10- CORE ALTERATIONS

'3.10 CORE ALTERATIONS Applicability:

Applicability:

Applies to fuel handling and core re-Applies to the periodic testing of those r

activity limitations.

interlocks and instruments used during.

refueling.and core alterations.

f1 e

Objective Objectives To assure that core reactivity is within To verify the operability of instru-the capability of the control rods and.

mentation and interlocks used.during.

to prevent criticality during refueling, refueling and core alterations.

Specification:

Specification:

+n A.

Refueling Interlocks A.

Refueling Interlocks

1. The Reactor Mode Switch shall

1. Prior to any fuel, handling, with the.

be locked in the Refuel-head off the, reactor -vessel, the.

position during core altera-refueling interlocks-shall be tions and the refueling functionally; tested. They shall interlocks shall be operable also be tested at weekly. intervals except as specified in thereafter until no longer required Specifications 3.10.A.2, 3.10.A.8, and following.any repair work 3.10.D, and 3.10.E.

associated with the interlocks.

2. Fuel shall not be loaded into

2. Whenever the reactor is in the refue.1 mode the reactor core unless all and rod block interlocks are being bypassed control rods'are fully

.for core unloading, one licensed. operator-inserted except in accord-'

and a member of the reactor analyst-ance.with Specification 3.10.A.7.

. department shall verify'that the. fuel-from the cell has been removed before

3. The fuel grapple hoist load switch the corresponding control rod is withdrawn.

.shall be set atf650 lbs.-

Amendment'No. 5,

227

JAFNPP 3.10.(cont'd) control rod'after.the fuel assemblies-in the cell con-taining'(controlled by) that control' rod have been removed from the reactor core.

All other refueling interlocks shall be' operable.

7.

In the " refuel" mode, there are inter-locks which prevent the refueling bridge (ifLloaded) from moving toward the core unless'all control rods are fully inserted.

Those' interlocks may be bypassed during spiral loading except for those control cella.which contain fuel or that control cell which is being loaded.

Interlocks for all cells'containing fuel, or'for any cell about to be loaded, shall be operable.

O.

Refuel interlocks and rod blocks associated with one rod permissive need not be operable, if all rods are fully-inserted and electrically disarmed.

D.

Core Monitoring B.

Core Monitoring During core alterations two SRM's Prior to making alterations to shall be operable, one in the core the core the SRM's shall be quadrant where fuel or control rods

' functionally tested and checked are being moved and one in an

.for neutron' response.

Thereafter, adjacent quadrant.

For an SRM.to.be the SRM's will be' checked. daily.for considered operable, the following response, except as specified-in conditions shall be satisfied:

Specification 3.10.B.3 and 4..

Anendment 230

. )n

-; mm JAFNPP 3.10 (cont'd) 4.10 (cont'd)

1. The-SRM shall be inserted to the normal operating level.. (Use of special movable, dunking type detectors during initial fuel loading and major core alterations in place of normal detectors is permissable as long as the detector is connected into normal '

SRM circuit).

2. The SRM shall have a minimum of 3 counts /sce with all rods fully inserted in the core except as noted in 3 and 4 below.

3. Prior to spiral unloading, the SRM's shall have an initial count rate of 3 3 CPS.

During spiral unloading, the count rate of the'SRM's may drop below 3 CPS.

4. During Spiral reload, SRM operability will be verified by using a portable external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until enough fuel is loaded to maintain 3 CPS.

Alternatively, two fuel assemblies will be loaded in different cells containing control blades around each SRM to obtain the required 3. CPS.

Until these two assemblies have been loaded in a given quadrant, it is not necessary for the SRM in that quadrant to indicate the minimum count rate of 3 CPS.

The loading of fuel near the SRM's does not violate the intent of the spiral re-loading pattern.

Amendment 59, 230a

4-ATTACHMENT II Safety Evaluation for Technical Specifications Related to Refuel Interlocks and Control Rod _ Blocks

-(JPTS-84-21) t i

New York Power Authority James A.

FitzPatrick Nuclear Power Plant Docket No. 50-333 1

dated 1

December 6, 1984' h

t1

~.

I..

Description of the Proposed Changes The proposed changes to the FitzPatrick Technical Specifications relate to refuel interlocks -and control rod blocks.

Specifically, the.following' changes are being proposed:

On pages 41, 41a, 41b and 42 in the column heading under

" Refuel," add "(16)."

On page 43a add the following note:

"16.

When all rods are full-in and electrically disarmed, the reactor protection system need not be operable.~"

On page 227, amend the end of Section 3.10.A.1 to read "3.10.A.2, 3.10.A.8, 3.10.D, and 3.10.E."

On page 230 add Section 3.10.A.8:

-" Refuel interlocks and rod blocks associated with one rod permissive need not be operable if all rods are fully inserted andtelectrically disarmed."

Section 3.10.B.1 is moved from page 230 to a new page 230a.

II. Purpo'se of the Proposed Changes The proposed changes are necessary to allow refueling while the Reactor Protection System is inoperable during installation of Analog Trip Transmitter System components.

III. Impact of the Proposed Changes The Reactor Protection System limits the uncontrolled release of radioactive material from the fuel and the Reactor Coolant Prensure Boundary by terminating excessive temperature and pressure increases through the initiation of an automatic scram.

Since the proposed changes apply only when all rods are full-in and electrically disarmed, the safety of the plant will not be affected.

The purpose of the control rod blocks and refueling interlocks, as described in Reference 1, Sections 7.6 and 7.7, is to prevent inadvertent criticality by restricting the movement of control rods.

Since all rods will be electrically disarmed, the proposed change will not affect the safety of the plant.

In addition, the nuclear characteristics of the core assure that the reactor would remain subcritical even if the highest worth control rod were fully withdrawn.

1.

.s.

'The commission has provided guidance concerning.the application of the standards for making a "no significant hazard considerations" determination by providing certain examples in the Federal Register (F.R.) Vol. 48, No. 67 dated April 6,

1984, page 14870.

The proposed changes match Commission example (vi):

"A change which either may

resultLin 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 in that'it is within all acceptable criteria.

However, the

. probability or consequences of a previously-analyzed accident are.not increased, and safety margins are not reduced.

Operation of the FitzPatrick plant in accordance with the proposed amendments, therefore, would nots (1) involvo a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or differant kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

IV.. Implementation of the Changes Implementation of the changes, as proposed, will not impact the fire protection program at FitzPatrick, nor will the changes impact the environment.

V.

Conclusion The incorporation of these changes involves no significant hazard considerations, as defined in 10 CFR 50.92.

VI. References 1.

James A.

FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).

2.

James A.

FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev.

1, July 1983.

2.