ML20080R844

Amend 94 to License NPF-30,revising TS Table 3.3-1 & 4.3-1 & Bases Pages B 2-8 & B 3/4 4-1.Changes Reflect Reanalysis of Boron Dilution Transient for Shutdown Modes to Address non-conservatisms in Previous Event Analysis

ML20080R844
Person / Time
Site:	Callaway
Issue date:	03/07/1995
From:	Wharton L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20080R847	List: ML20080R844
References
NUDOCS 9503100102
Download: ML20080R844 (11)

Text

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UNITED STATES -

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. NUCLEAR REGULATORY COMMISSION 3'

F WASHINGTON, D.C. 20S55-0001 n

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UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT l' DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE

' Amendment-No. 94 License No. NPF-30 1.

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

A.

The application for amendment filed by Union Electric Company-(UE, the licensee) dated August 4, 1994, and October 31, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regn1ations 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-t D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have i

been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-30 is hereoy amended to read as follows:

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I 9503t00102 950307 I DR ADOCK 0 3

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(2)

. Technical Soecifications and Environmental Protection Plan The Technical l Specifications' contained in Appendix A, as revised through Amendment No. 94, and the Environmental Protection Plan-contained in Appendix B, both of which are attached hereto,-are hereby incorporated into'the license. :UE shall operate.the facility i

in accordance with the Technical ~ Specifications 'and lthe i:#

Environmental. Protection Plan.

3.

This license amendment.is effective as of.its date of. issuance. The Technical Specifications are to be implemented within 30 days from the date of issuance.

'FOR THE NUCLEAR REGULATORY COMMISSION k

. Raynard Wharton, Project Manager Project Directorate III-3 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: March 7, 1995 I

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ATTACHMENT TO LICENSE AMENDMENT No.~94-p OPERATING LICENSE NO. NPF-30 i

DOCKET NO. 50-483 1

Revise Appendix A Technical Specifications by removing-the pages identified

-I below and insarting the enclosed pages. The revised pages are identified by the' captioned' amendment number and contain. vertical lines indicating the area-of change. The corresponding. overleaf pages, indicated by an asterisk, are J

also provided to maintain document completeness..

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EfMQYE INSERT

)

1 B 2-8 B 2-8 3/4 3-4*

3/4 3-4*

j 3/4 3-5 3/4 3 !

3/4 3-11*

3/4 3-11*

3/4 3-12 3/4.3-12

)

3/4 3-12a 3/4 3-12a B 3/4 4-1 B 3/4 4-1

'B 3/4 4-2*

B 3/4 4-2*

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4 LIMITING SAFETY SYSTEM SETTINGS BASES' REACTOR TRIP SYSTEM INTERLOCKS The Reactor Trip' System interlocks perform the following functions:

P-6 On increasing power, ~P-6. allows the manual block' of the~ Source Range

+

trip (i.e., prevents premature block of Source Range trip), provides.

a backup block for Source Range Neutron Flux Multiplication, and l

allows deenergization of the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored. '

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant. loop, reactor coolant pump bus i

undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed. trips are automatically. blocked.

P-8 On. increasing power, P-8 automatically enables Reactor ~ trips on low-flow in one or more reactor coolant loops. On decreasing power,-the P-8 automatically blocks the single loop Low Flow trip.

a P-9 On increasing power, P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Ra'ge trip and de-energizes the Source Range high voltage power. On decreasing power; the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.

Provides input to P-7.-

P-13 Provides input to P-7.

i CALLAWAY - UNIT 1 B 2-8 Amendment No. 94

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TABLE 3.3-1 (Continued) 1 REACTOR TRIP SYSTEM INSTRUNENTATION

~

E MINIMUM d

TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT-0F CHANNELS TO TRIP OPERABLE MODES ACTION g-5.j C

18.

Reactor Trip System Interlocks J-l y

a.

Intermediate Range Neutron Flux, P-6 2

1 2

2N-8 b.

Low Power Reactor Trips Block, P-7 l

P-10 Input 4

2 3

1-8 j-or P-13 Input 2

1 2

1 8

R c.

Power Range Neutron Flux, P-8 4

2 3

1 8

4 d.

Power Range Neutron-Flux, P-9 4

2 3

1 8

e.

Power Range Neutron Flux, P-10 4

2 3-1, 2 8

f.

Turbine Impulse Chamber-Pressure, P-13 2

1 2

1 8

19.. Reactor Trip Breakers 2

1 2

12 9,12 A3, 4*, 5*

10 2

1 2

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20. Automat'ic Trip and Interlock Logic 2

1

-2 1 2

-- 31 A3, 4*, 5*

10 2

1 2

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_ - _ - - _ - _ - _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ = _ _ - _ _ _

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TABLF3;3-l'(Continued)~

m TABLE NOTATIONS 3

Only-if the Reactor. Trip System breakers happen to'be in the closed position.and the. Control Rod Drive: System.is capable of rod withdrawal..

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The~ boron dilution flux multiplication signals 'may be blocked during reactor startup in accordance with approved procedures..

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l The. provisions of Specification 3.0.4 are not applicable.

H.

Below.the P-6 (Intermediate Range Neutron Flux-Interlock) Setpoirit..

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- Below the P-10,(Low Setpoint Power Range Neutron Flux Interlock).

a Setpoint.

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.(1)

The applicable MODES for these channels notedz in Table 3.3-3 are-more restrictive and, therefore, -applicable.

'i ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels-one less than the Minimum 1

Channels OPERABLE requirement, 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 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 />.

ACTION 2 -

With the number of OPERABLE channels one less than-the Total'-

Number of Channels,:STARTUP and/or POWER OPERATION may proceed provided.the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition-within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,-

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable. channel may be bypassed for-up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for i

surveillance testing of other_ channels per Specification ~

i 4.3.1.1, and Either,. THERMAL POWER is restricted to' less than or equal to?

c.

o 75% of RATED THERMAL POWER and the Power Range Neutron Flux j

Trip Setpoint is reduced to less than'or equal to 85% of-RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT' POWER.

'l TILT RATIO is monitored at least once:per 12. hours.per Specification 4.2.4.2.

ACTION 3 -

With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER le' vel:

a.

Below the P-6 (Intermediate Range Neutron. Flux interlock)

Setpoint, restore the inoperable channel-to OPERABLE status prior to increasing THERMAL'P0WER above the P-6 Setpoint; or

b. Above the P-6 (Intermediate Range Neutron Flux interlock)

Setpoint but below 10% of RATED THERMAL. POWER, restore the

. inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

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CALLAWAY - UNIT 1 3/4 3-5 Amendment No. A7,H,94 l

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. TABLE 4.3-1 (Continued)

E REACTOR TRIP SYSTEM INSTRtMENTATION SURVEILLANCE REQUIREMENTS.

5 i

TRIP e

AMALOG ACTUATING-MODES FOR 5

CdANNEL

- DEVICE-CHANNEL CHANNEL OPERATIONAL OPERATIONAL.

WHICH ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST-LOGIC TEST _IS REQUIRE 0.

18.

Reactor Trip System Interlocks (Continued) d.

Power Range Neutron Flux. P-10 N.A.

R(4)

R N.A.

N.A.

1.2 '

e.

Turbine Impulse Chamber Pressure. P-13 N.A.

R R

N.A.

M.A.-

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19. Reactor Trip Breaker N.A.

N.A.

N.A.

M(7.11)

N.A.

1.2.3*.4*.5*;

Yd

20. Automatic Trip and Inter 1ock logic M.A.

N.A.

N.A.

N.A..

M(7)

-1.2.3*.4* 5*

21.

Reactor Trip Bypass Breaker M.A.

M.A.

M.A.

M(17).R(18)

N.A.

1.2.3* 4* 5*'

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'o TABLE-4~.3-14 Continued)

TABLE NOTATIONS W

5, Jonly if the Reactor Trip ' System breakers happen to be closed and' th'e.

Control Rod Drive System is capable of rod withdrawal.

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. f ;The specified 18 month fre uency may be waived for Cycle 1:provided the surveillance:is performed: rior to restart following the first= refueling a

Toutage or June 1,-1986, wh chever. occurs first. The' provisions of.

Specification 4.0.2 are reset from performance of this surveillance.

5 lBelow P-6. (Intermediate Range Neutron Flux? interlock)-Setpoint.

1. 1. 1. Below P-10:(Low Setpoint Power Range Neutron Flux interlock) Setpoints.

(1)

If not performed in previous 31 days.-

l Com arison of calorimetric to excore power indication above 15% of ' RATED.-

(2)-

.THEkMAL POWER. Adjust'excore channel gains consistent with calorimetric" power if absolute difference is greater than 2%. The arovisions of l

Specification 4.0.4'are not applicable for entry.into MODE'2 or.1;

~l (3)

Single point comparison of incore to excore' AXIAL FLUX DIFFERENCELabove.

15% of RATED THERMAL POWER. Recalibrate if the: absolute difference is -

J greater than or equal to 2%. The provisions of Specification 4.0.4 are 1

not applicable for entry -into MODE 2 or 1.

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(4)

Neutron detectors may. be excluded from CHANNEL CALIBRATION.

(5)

For Source Range detectors, integral bias curves are obtained, evaluated, j

and com)ared to manufacturer's data.

For Intermediate Range 1and Power Range ciannels, detector plateau curves shall be obtained, evaluated, and compared to manufacturer's data.

For the Intermediate Range and-Power Range Neutron Flux channels the provisions'of Specification 4.0.4 1

are not applicable for entry into' MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL-POWER. The provisions of Specification 4.0.4 are not-applicable for entry into MODE' 2 or 1.

Determination of the loop specific vessel 4T value should be made when performing the Incore/Excore quarterly recalibration,; under steady state conditions.

(7)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. The TRIP ACTUATING DEVICE OPERATIONAL' TEST shall independently verify the OPERABILITY of the Undervoltage and Shunt Trip Attachments of j

the Reactor Trip Breakers.

(8)

Deleted (9)-

Quarterly surveillance in MODES 3*,

4', and 5* shall also include verification that permissives P-6 and P-10 are in their required state-for existing plant conditions by observation.of the permissive annunciator window. Quarterly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an i

increase of 1.7 times the count rate within a 10-minute period..

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CALLAWAY - UNIT 1 3/4 3-12 Amendment No./l),28,8A,81,94

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TABLE 4.3-l~(Continued) c..

TABLE NOTATIONS (10)

Setpoint verification is not required.

(11).

Following maintenance or. adjustment of the: Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.

(12)'

At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Flux Multiplication test signal the normal CVCS discharge valves will close and the centrifugal charging pumps l

.~

suction valves from the RWST will open within. 30 seconds.

(13)

' Deleted (14)

Deleted (15)

The surveillance MODES specified.for 'these channels in Table 4.3-2 are more restrictive and therefore, applicable.

(16)

TheTRIPACTUATINGDEVjCEOPERATIONALTESTshal1~ independently verify the OPERABILITY of the Undervoltage and Shunt Trip circuits-for the Manual Reactor. Trip function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit.

(17)

Local manual shunt trip prior to placing breaker in service.

(18)

Automatic Undervoltage Trip.

Complete verification of OPERABILITY of the manual reactor trip switch circuitry shall be performed prior to startup from the first shutdown to Mode 3 occurring aften August 7, 1992.

CALLAWAY - UNIT 1 3/4 3-12a Amendment No.)),$$,7f, AJ,AA,7),94

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1 21.id ltEACTOR COOLANT SYSTEM j

BASES

[

3/4.4'1 REACTOR C0OLANT: LOOPS AND C00LANT CIRCULATION

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The plant is: designed to operate with ail: reactor coolant loops in operation and maintain DN8R above the: safety analysis DN8R limits during y

la11< normal operations:and anticipated transients.

In MODES'I and 2 with.

1one-reactor coolant loop not'in operation this specification requires _that

the plant be in at;1 east

HOT STAND 8Y.within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

.i In MODE 3, two reactor coolant loops provide sufficient heat removal

capability for removing decay heat even in the event of a bank withdrawal i

accident however single failure considerations require that three loops-

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be OPERAELE. A single reactor coolant loo) provides sufficient' heat removal if a' bank withdrawal accident can >e prevented, i.e.

by opening the Reactor Trip System. breakers.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single

}

reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations

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require that at least two' loops'(either RHR cr RCS)- be OPERABLE.

.In MODE 5 with reactor coolant loops not filled, a single RHR loop

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provides sufficient heat removal capability for removilg decay heat; but i

single failure ~ considerations,-and the unavailability.-of.the' steam

.i generators'as a heat removing component, require that at least two RHR t

loops be OPERABLE.

. The operation of one reactor coolant i

'provides adequate flow to ensure mixing, pump (RCP) in MODES 3, 4, and 5l 1

prevent stratification and' produce gradual ~ reactivity changes during boron concentration reductions in the Reactor Coolant System.- The reactivity change. rate associated with t

boron reduction will, therefore, be within the transient mitigation i

caphbility of the Boron Dilution Mitig; tion System (BDNS)...With no reactor coolant' loop in operation in either MODES 3, 4, or 5,. boron dilutions must be terminated and dilution sources isolated. The boron.

dilution analysis in these MODES takes credit for the mixing. volume-i associated with having at least one reactor coolant loop tin operation.

l The restrictions on starting a reactor coolant pump in MODES 4 and 5 are provided to arevent RCS pressure transients caused by energy additions.fromtieSecondaryCoolantSystem,-whichcouldexceedthelimits j

of Appendix G to 10 CFR Part 50. The RCS will be protected against n

overpressure innsients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature i

of each' steam generator is less than 50*F above each of the RCS cold leg i

temperatures.

j CALLAWAY - UNIT 1.

B 3/4 4-1 Amendment No. A5,94 i

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-REACTOR COOLANT SYSTEM BASES I

~

3/4.4.7 5AFETY VALVfg The pressurf ter Code safety valves operate to prevent the RC5 from being

' pressurized above its safety Lief t of 2735 psig.

Each safety valve is destensa to relieve 420,000 )bs per hour of saturated steam. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. -In the event that no safety valves are OPERABLE,.

an operating RNA loop, connected to the RC$ provides everpressure relief capability and will prevent RC$ everpressurIsation. In addition, the Over-pressure protection Systes provides a diverse means of protection against RCS 4

overpressurization at low temperatures.

During operation, all pressurfaer Code safety valves must be OPERA 8LE to prevent the RC$ from being pressurized above its safety Limit of R735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of-load assualng no Reactor trip i

and also assuming no operation of the power-operated relief velves or steaa 1

dump valves.

Demonstration of the tafety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI 1

of the ASME Soiler and Pressure Code.

3/4.4.3 PRE 550RIZER The 12-hour periodic surveillance is sufffcfent to ensure that the p. era-t meter is restored to within its Ifait following expected transient operation.-

i 1he maximum water volume also ensures that a steam bubble is formed and thus the RC5 is not a hydraulically solid systes. The requirement that a minimus number of pressurizer heaters be OPERA 8LE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to j

relieve RCS pressure and prevent a high pressurizer pressure reactor trip l

during all design transients up to and including the design ste load decrease t

with steam dump. Operation of the PORVs minimizes the undesira le opening of the spring-loaded pressurizer Code safety valves. Each PORY has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

The PORVs are equipped with automatis actuation circuitry and manual control capability. Because no credit for automatic operation is taken in the FSAR analyses for MODE 1, 2 and 3 transients where operation of the PORVs has a beneficial impact on the results of tie analysis, the PORVs are considered OPERABLE in either the manual or automatic mode. The automatic mode is the preferred configuration, as this provides pressure relieving capability without reliance on operation action.

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CALLAWAY - UNIT 1 B 3/4 4 2 Amen $nent No.83 i