ML20096H269

Amend 165 to License DPR-35,revising TS to Convert Current percentage-based Scram Time Limits to notch-based Limits in TS 3.3.C & Updating Bases for notch-based Limits in TS 3/4.3.C

ML20096H269
Person / Time
Site:	Pilgrim
Issue date:	01/23/1996
From:	Marsh L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20096H272	List: ML20096H269 ML20096H282
References
NUDOCS 9601260326
Download: ML20096H269 (9)

Text

1. "49 S

g

-4 UNITED STATES g

j NUCLEAR REGULATORY COMMISSION o

2 WASHINGTON, D.C. 20eeH1001 i

49.....

BOSTON EDISON COMPANY DOCKET NO. 50-293 l

PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 165 License No. DPR-35 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Boston Edison Company (the licensee) dated dated July 14, 1995, as supplemented September 12 and December 8, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954,.as amended (the Act), and the Commission's rules and regulations; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in recordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment.

l 9601260326 960123 PDR ADOCK 05000293 P

PDR

. 3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMISSION h

g Ledyard B. Marsh, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 23, 1996 4

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1 ATTACHMENT TO LICENSE AMENDMENT N0.165 FACILITY OPERATING LICENSE NO. OPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 1

B2-2 B2-2 3/4.3-4 3/4.3-4 3/4.3-5 3/4.3-5 B3/4.3-6 B3/4.3-6 j

3/4.11-2 3/4.11-2 3/4.11-3 3/4.11-3 1

4 a

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- - - ~ -

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2.0 SAFIff LIMITS (Cont)

Fllft CIADDING Since the pressure drop in the bypass region is essentially all INTEGK1TY (2.1.1) elevation hand, the core prescare drop at low power and flows will j

(Cont) always be greater than 4.5 psi. Analyses show that with a bundle flew of 18 x 108 lbs/hs, bundle pressure drop is nearly l

independent of bundle power and has a value of 3.5 psi. Thus, the j

bundle flow with a 4.5 pst driving head wJl1 b* 5ceater than 28 a 108 lbs/hr. Full scale ATIAS test data taken at pressures free 14./ psia to 800 psia indicate that the fuel assembly static 1 l

power at this flow is approutmately 3.35 Wt.

With the design peaking tactors, this corresponds to a THERNAL FOWER of more than 50% of RATED THERNAL POWER. Thus, a THERMAL POWER limit of 25% of i

RATED THERNAL FUW58 for reactor pressure below 7s3 peig is i

conservative.

I The Safety Limit dCPR is determined using the Ceneral Electrie

'lhermal Analysis Basis, GETAE (1), which is a statistical model J

that condaines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Riectric critical Quality (X)

Boiling Length (L), GAIL, i

correlation.

The GEXL correlation is valid over the range of conditions used in the costs of the data used to develop the correlation. These j

conditions are:

1 l

Pressure:

800 to 1300 psia l

Mass Flux:

0.1 to 1.5 M1b/hr.ftt j

Inlet subcooling:

0 to 70 stu/lb Axial Profile:

1.5 chopped cosine 1.7 inlet peaked i

1.7 outlet peaked

)

R. Factor 0.95 to 1.20 l

Rod Array 9X9 GE 11 array l

i i

l MINIMuu caITICAL The fuel claddtng integrity safety Limit is set auch that no fuel j

POWER RATIC damage is caleviated to occur if the limit is not violated. Since I

(2.1.2) the parameters which ranult in fuel damage are not directly observable during reactor operation, the thermal and hydraulic eendittens resulting in a deparrure frns nue.lanta halling have been used to mark the beginning of the region where fuel damage i

seuld eseur. Although it is reeegnised thar a departure free nucleate boiling would not result in damage to SWR fuel rods, the d

erittant = =:r as (Cent) i j

j Amendment No. 15.42t-72 -195,-1293 133165 32-2 3

j j

1 LTMITING CONDITION FOR OPERATION SURVIILIANCE REQUIREMENT 3.3 REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)

B.

Control Rods (Cont)

B.

Control Rods (Cont)

4. Control rods shall not be

4. Prior to control rod withdrawal withdrawn for startup or for startup or during refueling unless at least two refueling, verify that at least source range channels have an two source range channels have i

observed count rate equal to or an observed count rate of at Breater than three counts per least three counts per second.

second.

5. The RBN shall be operable as required in Table 3.2.C-1, or control rod withdrawal shall be blocked.

C.

Scram Insertion Times C.

Scram Insertion Times

1. Average scram insertion time

1. Following each refueling for all operable control rods outage, or after a reactor 1

from de-energization of the shutdown that is greater than scram pilot valve solenoids to 120 days, each operable control drop out (DO) of Notches 04, rod shall be subjected to scrac 24, 34, and 44 shall be no time tests from the fully greater than:

withdrawn position.

If testing is not accomplished with the Notch Average Scram nuclear system pressure above Position Times (seconds) 950 psig, the measured scram insertion time shall be 44 DO 0.508 extrapolated to reactor 34 DO 1.252 pressures above 950 psig using 24 DO 2.016 previously determined 04 DO 3.578 correlations.

Testing of all operable control rods shall be l

completed prior to exceeding 40% rated thermal power.

Revision Amendment No. 15,-68,-124,-138 165 3/4.3-4 i

l LIMITING CONDITION FOR OPERATION SURVEILIANCE REQUIREMENT 4

3.3 REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)

C.

Scram Insertion Tf== (Cont)

C.

Scram Insertion Tf== (Cont)

2. Average scram insertion time

2. Within each 120 days of 1

for the three fastest operable operation, a minimum of lot of control rods in each group of the control rod drives, on a four control rods in all two-rotating basis, shall be scram by-two arrays from de-tested as in 4.3.C.l.

An energization of the scram evaluation shall be completed i

pilot valve solenoids to every 120 days of operation to dropout (DO) of Notches 04, provide reasonable assurance 24, 34 and 44 shall be no that proper performance is greater than:

being maintained.

1 Notch Average Scram Position Time (Seconds)

I l

44 DO 0.538 34 DO 1.327 24 DO 2.137 04 DO 3.793 l

ntr d c umu ators

3. The maximum scram insertion j

time for any operable control once a shift, check the status of rod from de-energization of the pressure and level alarms for the scram pilot valve each accumulator.

solenoids to dropout of Notch 04 shall not exceed 7.00 seconds.

D.

Control Rod Ace"=alators At all reactor operating pressures, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:

}

1. Inoperable accumulator.

2. Directional control valve electrically disarmed while in a non-fully inserted position.

3. Scram insertion time greater than the maximum permissible insertion time.

If a control rod with an inoperable accumulator is inserted " full-in" and its i

directional control valves are electrically disarmed, it shall l

not be considered to have an inoperable accumulator.

65 -124 -146 165 Amendment No.

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3/4.3 5 I

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l A&EER:

i 3/4.3 REACTIVITY CONTROL (Cont)

.C.

Scram Insertion Times

~

The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming i

less than the Safety Limit MCPR. Analysis of the limiting power transient shows that the negative reactivity rates resulting froe the scram with the

{

average response of all the drives as given in the above Specification, j

provide the required protection, and MCPR remains greater than the Safety i

Limit MCPR.

l i

The scram times for all control rods will be determined at the time of each

{

refueling outage. A representative sample of control rods will be scram tested during each cycle as a periodic check against deterioration of the l

l control rod performance.

a l

The limits on scram insertion time presented in Technical Specification 3.3.C include an allowance for the uncertainty in the location of the position indicator probes as well as an allowance for the uncertainty in the t

j control rod positions themselves when dropout of the reed switches occur.

?

D.

Control Rod Acen="1 storm j

Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean i

core.

The worst case in a nine-rod withdrawal sequence resulted in a k.gg i

<l.0 - other repeating rod sequences with more rods withdrawn resulted in k.gg

>1.0.

At reactor pressures in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet required scram insertion times due to the action of reactor pressure.

In addition, they may be 1

normally inserted using the control-rod-drive hydraulic system. Procedural control will assure that control rods with inoperable accumulators will be I

spaced in a one-in-nine array rather than grouped together.

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1 Revision Amendment No. 15,-42,-13s 165 B3/4.3 6 1

4

i LI)fITING CONDITIONS FOR OPERATION SURVEILIANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont) i B.

Linear Heat Generation Rate (MGR_)

B.

Linear Hest Generetion Rate (WGR)

During reactor power operation, The MGR as a function of core the MGR shall not exceed the height shall be checked daily limits specified in the CORE during reactor operation at 2 256 OPERATING LIMITS REPORT.

rated thermal power.

l l

If at any time during operation it is determined by normal surveillance that the limiting value for MGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the GGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall continue until C.

Minimum Critical Power Ratio reactor operation is within the (MCPR) prescribed limits.

1. MCPR shall be determined' daily C.

Minimum Critical Power Ratio during reactor power operation (MCPR) at > 254 rated thermal power

" "I ""I

""E*

1. During power operation MCPR shall be a the MCPR operating power level or distribution limit specified in the core that would cause operation with a limiting control rod pattern Operating Limits Report.

If at d "

any time during operation it is determined by normal Specification 3.3.5.5.

surveillance that the limiting value for MCPR is being

2. The value of r in Specification exceeded, action shall be 3.11.C.2. shall be equal to 1.0 unless determined from the initiated within 15 minutes t restore operation to within the result of surveillance testing of Specification 4.3.C as prescribed limits.

If the steady state MCPR is not returned to within the a) r is defined as prescribed limits within two (2) hours, the reactor shall be fave - rg brought to the Cold Shutdown 7_

j condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding 1.252 -rB i

action shall continue until

]

reactor operation is within the prescribed limits.

i Revision Amendment No. 15,-27,-39,-42,-54,-195,-133 165 3/4.11-2 l

LIMITING CONDITIONS FOR OPERATION bURVEILLANCE REQUIREMENTS l

t 3.11 REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)

C.

Minimum Critical Power Ration MCPR C.

Minimum Critical Power Ration MCPR (Cont'd)

(Cont'd)

2. The operating limit MCPR values

b. The average scram time to as a function of the r are dropout of Notch 34 is given in Table 3.3.1 of the determined as follows:

Core Operating Limits Report where r is given by n

specification 4.11.C.2.

I Ni ri rave "

i"1 n

I Ni i=1 Where: an n - number of surveillance tests performed to date in the cycle.

Ni-number of active control rods measured in the 12 surveillance test.

ri - average scram time to dropout of Notch 34 of all rods measured in the i* surveillance test.

I

c. The adjusted analysis mean scram time (rg) is 1

calculated as follows:

r, 1

1

!N

'B - p + 1.65.

l a

l n

l l

I ut l l i=1 l

L.

J Where:

mean of the distribution for y

average scram insertion time to i

dropout of Notch 34, 0.937 sec.

N1-total number of active control rods at BOC during the first i

surveillance test.

standard deviation of the a -

distribution for average scram insertion time to the dropout of Notch 34, 0.021 seconds.

Revision Amendment No. 24,-42-54r-59 -133,-133 165 3/4.11 3 i

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