ML20052E731

Submits Justification for Proposed Tech Specs Changes Re Design Basis events.Marked-up Tech Specs Encl

ML20052E731
Person / Time
Site:	Calvert Cliffs
Issue date:	04/29/1982
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
NUDOCS 8205110391
Download: ML20052E731 (11)

Text

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BALTIM O RE GAS AND ELECTRIC CHARLES CENTER P.O. BOX 1475 BALT MORE, MARYLAND 21203 ARTHUR E*. LUNDVALL, JR.

April 29,1982 VicE PRESIDENT SuppLv

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Office of Nuclear Reactor Regulation dh(.

U. S. Nuclear Regulatory Commission of Washington, D. C. 20555 i'},.'$p p

ATTENTION: Mr. R. A. Clark, Chief k

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s Operating Reactors Branch #3

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Division of Licensing k/K Q.

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No.1, Docket No. 50-317 Amendment to Operating License DPR-53 Supplement 1 to Sixth Cycle License Application Ihoo/

REFERENCE (A):

A. E. Lundvall to R. A. Clark letter, dated February 17,

/

1982, Amendment to Operating License DPR-53 Sixth e//

Cycle License Application g

Gentlemen:

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6. Sc.AwGN /f In partial response to NRC requirements we are replacing existing pressured. b'25 transmitters in the Unit I containment with enviromnentally qualified Barton 4 g//

pressure transmitters during the Spring 1982 refueling outage. The uncertainties associated with those transmitters under all accident conditions have been underL kofj?

evaluation for some time and are now in the latter stages of confirmation. In late 7, gg//

1981, during the design of the Cycle 6 reload core we made some projections of what those uncertainties might be. The projections were considered in the Reload Safety Analysis for Cycle 6.

Ilowever, the projections were too imprecise to be factored into the reanalysis of each Design Basis Event and/or into the determination of LSSSs and LCOs for Cycle 6.

Technical Specification Changes and Justification We can now state the uncertainties more precisely. The Design Basis Events have been systematically reviewed for the effect that the uncertainties of the Barton transmitters scheduled for operability in Cycle 6 will have on the Safety Analyses. The Main Steam Line Break and the Steam Generator Tube Rupture events are adversely affected.

The effect can be best and most quickly accommodated by revising several RPS and ESFAS equipment setpoints. The attachment to this letter describes proposed changes to the Technical Specifications required to ensure the continued validity of the Safety Analyses.

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Office of Nuclear Reactor Regulation April 29, 1982 Page 2 Safety Analysis and Review The proposed changes to the Technical Specifications do not constitute an unreviewed safety question since the shift in RPS and ESFAS setpoints are in the conservative direction by an amount sufficient to at least offset the additional transmitter uncertainties associated with the most limiting Design Basis Events already analyzed in Reference (A).

The Plant Operations and Safety Review Committee (POSRC) and the Offsite Safety and Review Committee (OSSRC) have reviewed the proposed changes to the Technical Specifications and have concluded that they do not constitute an unreviewed safety question nor do they present an undue risk to the health and safety of the public.

Very truly yours, BALTIMORE GAS AND E ECTRIC COMPANY dh a

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' ale. K5hdy'all, Jr.

Vice President - Sup AEL/WJL/djw Attachment STATE OF MARYLAND, CITY OF BALTIMORE, TO WIT:

Arthur E. Lundvall, Jr., being duly sworn states that he is Vice President of l

the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he executed the foregoing Amendment for the purposes therein set forth; that the statements made in said Amendment are true and correct to the best of his knowledge, information, and belief; and that he was authorized to execute the Amendment on behalf of said Cornoration.

WITNESS My IIand and Notarial Seal this df day of 198L l

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Notary Pdblic My Commission Expires:

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Copies To:

J. A. Biddison, Esquire (w/o Encl)

G. F. Trowbridge, Esquire (w/o Encl)

D. II. Jaffe - NRC P. W. Kruse - CE

1 ATI'ACHMENT The following Technical Specification changes are keyed to Tables 9-1 and 9-2 of the enclosure to A. E. Lundvall, Jr. to R. A. Clark letter, dated February 17,1982,

" Amendment to Operating License DPR-53 Sixth Cycle License Application."

TABLE 9-1 Change Tech Spec #

Action Add the following Tech Spec changes:

25 Table 2.2.1, Item 6 Change Steam Generator - Low Trip Page 2-9 Setpoint and Allowable Value from 570 to 635 psia 26 Table 2.2.1, Note (2)

Change Steam Generator - Low Page 2-10 Trip Bypass from 685 to 710 psia 27 B.2.2.1, Page B2-5 Change Steam Generator Pressure -

Low Trip Setpoint from 570 to 635 psia. Change the pressure instrument loop uncertainty factor from i22 psi to i87 psi based on Main Steam Line Break Event 28 B.2.2.1, Page B2-6 Change the floor pressure for the TM/LP trip from 1750 to 1875 psia 29 Table 3.3-1, Note (b)

Same action as for Change 26 Page 3/4 3-4 30 Table 3.3-3, Note (a)

Change SIAS Pressurizer Pressure -

Page 3/4 3-15 Low Bypass from 1700 to 1800 psia 31 Table 3.3-3, Note (c)

Same action as for Change 26 Page 3/4 3-15 32 Table 3.3-4, Item 1.c Change SIAS Pressurizer Pressure -

Page 3/4 3-17 Low Trip Setpoint and Allowable Value from 1578 to 1725 psia 33 Table 3.3-4, Item 4.b Change Main Steam Line Isolation Page 3/4 3-17 (SGIS) Trip Setpoint and Allowable Value from 570 to 635 psia

2 TABLE 9-2 Change Tech Spec #

Explanation 25 Table 2.2.1, Item 6 The Steam Generator Pressure - Low Page 2-9 Trip Setpoint and its associated instrument loop uncertainties have been changed to reflect the uncertainty associated with Barton pressure transmitters during the Main Steam Line Break Event 26 Table 2.2.1, Note (2)

The Steam Generator - Low Trip Page 2-10 Bypass has been changed to reflect the change in the Trip Setpoint 27 B.2.2.1, Page B2-5 Same explanation as for Change 25 28 B.2.2.1, Page B2-6 The floor pressure for the TM/LP Trip Setpoint has been changed to reflect the uncertainty associated with the Barton pressure transmitters during a LOCA 29 Table 3.3-1, Note (b)

Same explanation as for Change 2S Page 3/4 3-4 30 Table 3.3-3, Note (a)

The SIAS Pressurizer Pressure -

Page 3/4 3-15 Low Bypass has been changed to reflect the change in the actuation setpoint 31 Table 3.3-3, Note (c)

Same explanation as for Change 26 Page 3/4 3-15 32 Table 3.3-4, Item 1.c The SIAS Pressurizer Pressure - Low Page 3/4 3-17 Trip Setpoint has been changed to reflect the uncertainty associated with the l.

Barton transmitters during a LOCA 33 Table 3.3-4, item 4.6 The SGIS Setpoint has been Page 3/4 3-17 changed to reflect the uncertainty 7

associated with Barton pressure

-transmitters during the Main Steam Line Break Event l

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n TABLE 2.2-1 (Cont'd)

?M REACTOR PROTECTIVE INSTR MENTATION TRIP SETPOINT LIMITS Si FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES f

4-Pressurizer Pressure - High 1.2400 psia 1 2400 psia E

5.

Containment Pressure - High 1 4 psig 1 4 psig 4

635 G35 6.

Steam Generator Pressure - Low (2)

> J70' psia

• c

> J7tf psia l

7.

Steam Generator Water Level - Low

> 10 inches below top

> 10 inches below top of feed ring.

of feed ring.

8.

Axial flux offset (3)

Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit lines not exceed the limit lines of Figurc 2.2-1.

of Figure 2.2-1.

9.

Thermal Margin / Low Pressure (1) a.

Four Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to Operating not exceed'the limit lines not exceed the limit lines of Figures 2.2-2 and 2.2-3.

of Figures 2.2-2 and 2.2-3.

b.

Steam Generator Pressure

< 135 psid

< 135 psid Difference - High (1)

~

F 10.

Loss of Turbine -- Hydraulic

> 1100 psig

> 1100 psig Fluid Pressure - Low (3) 3 11.

Rate of Change of Power - High (4) 1 2.6 decades per minute 1 2.6 decades per minute O

TABLE NOTATION

-4 (1) Trip may be bypassed be}ow 10 % of RATED THERMAL POWE.'; bypass shall be' automatically removed when THERMAL POWER is > 10~ % of RATED THERMAL POWER.

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TABLE 2.2-1 (Con t_' d_)

9 TABLE NOTATIONS _ (Cont'd)

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7/o 7/O M

(2)

Trip may be manually bypassed belowMpsia; bypass shall be automatically removed at or above A88f psia.

I

~ 3 (3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when g

THERitAL POWER is > 15% of RATED THERMAL POWER.

i (4) Trip may be bypassed below 10-4% and above 12% of RATED THERMAL POWER.

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LIMITING SAFETY SYSTEM SETTINGS I

BASES b

y operation of the reactor at reduced power if one or two,re pumps are taken out of service.

The low-flow trip setp'oints and Allowable Values for the various reactor coolant pump combinations have been derived in consideration of, instrument errors nd ^ ponse times of equipment involved to maintain the DNBR above 11195 under normal operation l

I and expected transients.

For reactor operation only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two-or three-pump d

operation prevents the minimum value of DNBR from going be position.

Changing these trip setpoints during two and three,pu 1'93 during J

normal operaticnal transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-High 9

The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam'line safety valves, provides reactor coolant Q

system protection against overpressurization in the event of loss of load without reactor trip.

This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its

+

concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

l Containment pressure-High

~The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.

The setpoint for this trip is identical to the safety injection setpoint.

Steam Generator Pressure-Low I

The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam genera and s c,35 subsequent cooldown of the reactor coolant.

ThesettingoSMjpsia l

is sufficiently below the full-load operating point of 850 sia so as not to interfere with nonnal operation, but still high enough to provide the required protection in the event of excessively p h_st_eam l

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flow.

This setting was used with an uncertainty factor o g fst 1

4n-the-eccidens-andyses,-. wh reh war b a re d on the l

>ti a u n deam ke dred L t veort CALVERT CLIFFS - UNIT 1 B 2-5 Amendment No. 37,48 i

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LIMITING SAFETY SYSTEM SETTINGS I

BASES Steam Generator Water Level i

The Steam Generator Water Level-Lcw trip provides core protection by preventing operation with the steam generator water level below the i

minimum volume. required for adequate heat removal capacity and assures that the pressure of the reactor coolant system will not exceed its Safety Limit. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of-trip to provide a margin of more than 13 minutes before auxiliary feedwater is requjred.

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' Axial Flux Offset 7

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The axial flux offset trip is'provided to ensure that excessive axial peaking will not cause fuel damage.

The, axial flux offset is determined from the axially split excore,4e.ty)stors. The trip setpoints i

ensure that neither a DNBR of less thanQd95 nor a peak linear heat rate which corresponds to the temperature for fUe'l centerline melting will I

exist as a consequence of axial power maldistributions. These trip set-

~J points were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship.

Thermal Marcin/ Low Pressure y/N The Thermal Margin /Loy Prat re trip is provided to prevent operation when the DNBR is less tha Qh195J/g The trip is initiate ver the reactor coolant system pressure signal drops below eithe

, sia or a computed value as described below, whichever is higherrThe computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, and the number of reactor coolant pumps operating. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation pennitted for continuous operation are assumed in the genera-tion of this trip function.

In addition, CEA group sequencing in accor-dance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.

Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

CALVERT CLIFFS - UNIT 1 B 2-6 Amendment No. 33, W,48 l

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TABLE 3.3-1 (Continued)

TABLE NOTATION i

, ith the protective system trip breakers in the closed position and Wthe CEA drive system capable of CEA withdrawal.

1. The provisions of Specification 3.0.4 are not applicable.,

Trip may be bypassed below 10-4be automatically removed l

of RATED THERMAL POW (a) of RATED THERMAL POWER.

7to (b) Trip may be manually bypassed below.G85 psia; bypass shall.be~

automatically removed at or above 585' psia.

7/O b

(c) Trip may be b'ypassed'below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL POWER.

(d) Trip may be bypassed below 10-4% and above 12% of RATED THERMAL

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l POWER.

1 (e) Trip may. be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.

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There shall be at least two decades of overlap between the Wide (f)

Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Monitoring Channels.

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ACTION STATEMENTS With the number of channels OPERABLE one less than

~

ACTION 1 required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within j

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in either the bypassed a.

or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall j

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then'be either restored to OPERABLE status or placed in the tripped condition.

i b5 CALVERT CLIFFS - UNIT 1 3/4 3-4 Amendment No. 48

TABLE 3.3-3 (Continued) l

'b TABLE NOTATION If00 (a) Trip function ma bypassed in this MODE when pressurizer pressure is < J70 psia; bypass shal be automatically removed l

when pressurizer pressure is > 17 sia.

_ jgg (c) Trip function may be bypassed in this MODE below48S pII~a~;, bypass d

shall be automatically removed at or above The provisions of Specification 3.0.4 are not applicable.

ACTION STATEMENTS With the number of OPERABLE channels one less than the ACTION 6 Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With the number of OPERABLE channels one less than the ACTION 7 Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a a.

The inoperable channel is placed in either the bypassed Mi or tripped. condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

b.

Within one hour, all functional units receiving an input from the inoperable channel are also placeo in the same condition (either bypassed or tripped, as applicable) as that required by a. above for the inoperable channel.

c.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> while performing tests and maintenance on that channel provided the other inoperable channel is placed in the tripped condition.

CALVERT CLIFFS - UNIT 1 3/4 3-15 Amendment No.48

TABLE 3.3-4 99 GG ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES 99

--e s pp ALLOWABLE qq FUNCTIONAL UNIT TRIP SETPOINT VALUES 33 1.

SAFETY INJECTION (SIAS)

, i a.

Manual (Trip Buttons)

Not Applicable Not Applicable cg 5.-

b.

Containment Pressure - liigh

~< 4.75 psig

< 4.75 psig

/72.f

- /'72.7 N"

c.

Pressurizer Pressure - Low

>J&7tIpsia 1 1&7tfpsia 2.

CONTAINMENT SPRAY (CSAS) a.

Manual (Trip Buttons)

Not Applicable Not Applicable b.

Containment Pressure -- High

< 4.75 psig 1 4.75 psig 3.

CONTAINMENT ISOLATION (CIS) #

y a.

Manual CIS (Trip Buttons)

Not Applicable Not Applicable b.

Containment Pressure - liigh 1 4.75 psig

< 4.75 psig 4.

MAIN STEAM LINE ISOLATION a.

Manual (MSIV Hand Switches and Feed llead Isolation py Hand Switches)

Not A licable Not Apglicable gg 636 1 r7 psia

>,570' psia

/

&R b.

Steam Generator Pressure - Low 2=??

1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 1.a and 1.c).

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