ML20091C075

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Amend 145 to License DPR-46,changing TS to Reduce Low Reactor Water Level Scram Setpoint (Level 3) & Making Administrative Changes Re Editorial & Typos
ML20091C075
Person / Time
Site: Cooper Entergy icon.png
Issue date: 07/31/1991
From: Quay T
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20091C076 List:
References
NUDOCS 9108050256
Download: ML20091C075 (12)
v • d • e


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UNITED STATES j

  • / i NUCLEAR REGULATORY COMMISSION 5.k

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NEBRASKA PUBLIC SuWER DISTRICT DOCKET NO. 50-298 COOPER NbCLEAR STATION i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.145 l

License No. DPR-46

)

1.

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

A.

The application for amendment by Nebraska Public Power District (the licensee) dated April 25, 1991 as supplemented by the letter dated June 28, 1991, 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; 8.

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 activitie' 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 license 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 accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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9108050256 910731 PDR ADDCK 05000298 P

PDR

-2.

Accordingly, the license is amended by changes to the Technical Specifi-i-

cations as indicated in the attachment to this license amendment and l

Paragraph 2.C.(2) of f acility Operating License No. DPR-46 is hereby amended to read as follows:

2.

Technical Specifications l

The Technical Specifications contained in Appendix A, as revised through Amenoment No.145, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

l 3.

The license amendment is effective as of its date of issuance.

FOR THE 8" CLEAR REGULATORY COMMISSION l

f/wasfAd%L&u/w Theodore R. Quay, Director Project Directorate IV-1 l

Division of Reector Projttts 111, IV, and V Office of Nuclear Reactor Regulation t

Attachment:

Changes to the Technical l

Specifications Date of Issuance:

July 31, 1991 l

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W ATTACHMENT TO LICENSE AMENDMENT NO.145 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PACES i

8 8

10 10 l

20 20 26 28 50 50 526 52a 56 56 l

59 59 83 83 l

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l 1.

SAff.TX._Ilt1111 LIMITING SAFETY SYETEM SETTINGS 2.1.A.1 (Cont'd) d.

APPJi Rod B19ek Trin Setting The APRM rod block trip setting shall be:

S s 0.66 V 4 42% -.66 AW RB where:

S

- Rod block setting in g

percent of rated thermal power (2381 MWt)

W and 6V are defined in Specification 2.1.A.l.a.

In the event of operation with a mnximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP). the setting shall be modified as follows:

S

. 6 W + 42% - 0.66 Ak1 _ E RB MFLPD where.

FRP - fraction of rated thermnl power (2381 MWt)

MFLPD mnximum fraction of limiting power density where the limiting power density for each type of fuel bundle is specified in the Core Operating Limits Report.

The ratin of FRP to MFLPD shall he p

set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

2.

Egaetor Water Low Level Sc rarr uni Isolation Trip Settint (except MSI'D l

+4.5 in.

on v e v.c l lavol instruments.

Amendnent No. 16,32,39,#1,#6,80,94, 3

123,f42,145

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n 87.100,121,145 Amendment No.

4.

2.1 Bases

(Cont'd) e 2.

Efjgter

  • tater low level Scram and Isolation Trio Settinr texpect MSIY)

The setpoint for low reactor water level scram is established at 1.evel 3 to ensure that during normal power operation the i.Stom of the separator skirt is not uncovered (this protsets available reactor recirculation purop NPSH from carryunder).

This level has been used in transient and accident analyses dealing with coolant inventory decrease. The results reported in USAR sections XIV 5 and XIV 6 show that when scram is initiated at 1.evel 3, the fuel and process barrier are adequately

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protected because MCPR remains well above the MCPR fuel cladding integrity limit in all cases, and reae or coolant system pressure does not reach the safety valve settings.

scram setting is approximately 30 inches below the normal operating. g ; and is thus adequate to avoid spurious scrams.

J 3.

Turbine Stor Valve Clostgf Scram tric settine The turbine stop valve closure scram trip anticipates the pressure,

neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves.

  • 'ith a scram trip setting of s10 percent of valve closure from full open, the resultant increase in suruce heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity limit even during the worst case transient that assumes the turbine bypass is closed.

This scrar is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressure.

4.

TurMne control Valve Fast closure Scram Trin Settinc Tha turbine tontrol valve fast closure ltm anticipates the pressure, neutron flux, and heat flux increase that ould result from fast closure of the turbine control valves due to oad rejection exceeding the capability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the loss of turbine control oil pressure as sensed by pressure switches, This setting and the feet that control valve closure time is approximately twice as ~long as that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure.

No significant change in MCPR occurs.

Relevant transient ana.tyses are presented in Section XIV 5.1.1 of the USAR.

l Amendment No. N 145

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ct COOPER IWC1. EAR STATIOff I

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TABLE.3.1.1 L

REACTOR PROTECTIOil SYSTE!! IIISTRiHEtiTATIOff REQUIREMENTS 7o 4

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y Minimum ihmber Action Required j;

q An t icah i i i ty_c< null r f ans of Operable When Equipment 1

y Reactor Protection Mode Switch Position Trip Level Channels Per Operability is System Tristrunction Shutdown Stactuu refuel Run Settine Trip Systems (1) flot Assured (1)

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flode Switch in Shutdown X(7)

X X

X 1

A I

11anusi Scram X(7)

X X

X 1

A IRM (17)

X(7)

X X

(5) s 120/125 of in-3 A

liigh Flux dicated scale 3

Inoperative

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X (5) 3 A

APRM (17)

X s (0.66V+541-0.66aV)' T7F 2

C 4

liigh Flux (Flow biased)

(14)(19)(20)

HFLPD 4

m IIIgh Flux X(7)

X(9)

X(9)

(16) s 15I Rated Power 2

A i

Inoperative X(9';

.X(9)

X (13) 2 A

Downscale (12)

(12)

(12)

X(11) 2 2.5%

2 A

liigh Reactor Pressure X(9)

X(10)

X

$ 1045 psig 2

A 4

1 til51 - PS - 55 A, B, C, & D liigh Drywell Pressure X(9)(8) X(8)

X s 2 psig 2

A cr D PC-PS-12.A,B,C, & D f

Re,ctor I.ow Water Level X

X X

2 + 4.5 in Indi-7 A or D I

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!1BT-LIS-101 A.B.C. & D cated level I

.i Scram flischarge Instrument volume X

X(2)

X s 92 inches 3 (18)

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liigh u ter level CED-IS-2 31 A & h

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l CPI)-1S-214 A & B CED-lJ-231 C & D CRis-I.T-214 C & D

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COOPER NUCLEAR STATION

. c, TABLE 3.2.A (Page 1) a" PRIMARY CONTAINMENT AND REACTOR VESSEL 'ISOIATION INSTRI7tfEffTATION i

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Minimma Number Action Required m

of Operable

'When Coisporm. -

Instrument Components Per Operability 1.

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Instrisment I.D. No.

Settinr. Limit Trin System (1).Not Assured (2)

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11ain' Steam Line fligh R11P-RH-251, A.B.C,60

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A or B

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.'m Reactor Low *Jater Level NBI-LIS-101, A,B,C,6D $1. 214.5 in. Indicated Imvel 2(4)

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Reactor Low Low lev Vater NB1-LIS-57 A & B $1 2-145.5 in. Indicated Level 2

A or B level NBI-LIS-58 A & B #1 j.

Itain Stearn Line Leak MS-TS-121. A,B,C,6D

$ 200*F 2(6) 3

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Detection 122. 123, 124, 143, 144, j

145, 146, 147, 148. 149, F

'o 150 e

Main Steam Line fligh MS-dPIS-116 A,B,C,6D

$ 150% of Rated Steam 2(3)

B Flow 117, 118, 119 Flow j

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!!ain Stearn.Line Low MS-PS-134, A B.C,6D 2 825 psig 2(5)

B l'res sure fligh Drywell" Pressure PC-PS-12 A.B.C,6D s 2 psig 2(4)

A or B IIIgh Reactor Pressure RR-PS-128 A & B 5 75 psig 1

D Itaisi Condenser low ifs-PS-103, A.B.C,6D 2 7" ilg (7) 2 A or B Vacuum l

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Reactor Vater Cleantip RVCU-dP15-170 A & B s 200I of System Flow 1

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NOTES TOR TABLE 3.2.A (cont'd.)

Croup.

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Isolation Signals:

1.

Reactor Low Vater Level (24.5 inches) l 2.

High Dry Well Pressure (s 2 psig)

Isolations:

1.

PHR Shutoown Cooling System 2.-

Drywell floor and equipment drain sump discharge lines.

3.

TIP ball s'alves 4

Group 6 isolation relays Group 3 Isolation Signals:

1.

Reactor Low Water Level (24.5 inches) 2.

Reactor Vater Cleanup System High Flow (5200 of system flow) 1 3.

Reactor Vater Cleanup System High Area Temperature ($ 200'F)

Isolations:

1.

Reactor Vater Cleanup System Croup 4-Isolation-Signals:

Provided by instruments on Table 3.2.B (HPCI)

Isolations:

Isolates the HPCI steam line C; up 5 Isolation Signals:

Provided by instruments on Table 3.2.B (RCIC)

Isolations:

Isolates the'RCIC steam line.

Group 6 Isolation Signals:

1.

Group 2 Isolation Signal 2.

Reactor Building H&V Exhaust Plenum High Radiation M 100 mr/hr'

. Amendment'No. 103.107,145 523

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.3 pg COOPER NUCIIAR. STATION j

g TABtX 3.2.B (Page 4) r j

IIPCI SYSTEff CIRCtIITRY REQUIREMENTS j

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Minimum Num!er of Action Required When

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Instrument Operable Components Component 0,erability 1

D Instrument

!.D. No; Settint-I.f rui t Per Trip System (1)

Is Not Assure <f l

a Reactor low low Vater NBI-LIS-72 A.B.C.

2-37 in. Indicate <* Level 2

A I.evel

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t Reactor liigh' Vater NBI-I.IS-101, B & D

$*58.5 In. Indicated level 2(2)

A l'

Level

  1. 2 liigh Drywell Press.

PC-PS-101 A,B,C, & D

$2 psig 2(2)

A IIPCI Turbine liigh HPCI-PS-97 A & B

$150 psig 1(2)

A Exhaust Pressure j

IIPCI Pump Iew IIPCI-PS 1 sl5" Ilg Vacuum 1(2)

A j

p Suction Press.

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i IIPCI. Purnp low IIPCI-FS-78 2400 gpm 1(2)

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Discharge Flow IIPCI low Steam IIPCI-PS-68, A B.C 2I00 psig 2(2)

A Supply Pressure

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IIPCI St eam Line

!!PCI-dPIS - 76 130 555210* 11,0 1

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Illgh AP IIPCI-dPIS-77

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2 IIPCI Steam I.ine IIPCI -TS - 101, A, B,C - 6 5200*F 2(4)

A i-Space 111 Temp.

D -102, 103, 104, s

11PCI-TS-125,126,12 7.128 i

RIIR-TS-150,151,152,153 154,155,156,157,158,159

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l Emerg. C<nnt. Storage IIPCI-IS-74 A & B 20* II 0 (10,000 gal.

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% *2 nt; Amendment No. 38,83,99,141,145 59

., _. _ _ _ _. _ _ _. _..... _.,. _ _, _ _. ~,. _. _. _ _ _.,. - _ _ _. -

I 3.2 EM.Il In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or ter:tinates operator 9trors before they result in serious consequences.

This set of specifications provides the limitin5 conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standbj gas treatment systems. The objectives of the specifications are (1) to assur<. the effectiveness of the protective instrumentation when required even dur.ng periods when portions of such systems are out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate performance.

Wen necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

Some of the settings on the instrumentation that initiate or control core and contaircent coolir g have tolerances explicitly stated where the high and low values are both crttical and may have a substantial effect on safety.

The set points of other ins. rumentation, where only the high or low end of the setting has a direct bearing on tafety, are chosen at a level away from the normal operr. ting range to prevent inadvartent actuation of the safety system involved and exposure to abnor=al situations.

A.

Prinr" Contalp..ent Isolatien Functions Actuation of primary containment valves is initiated by protective instrumentation shown in Tablo 3.2. A which senses the conditions for which isolation is required.

Such instrum.ntation must be available whenever primary contain: rent integrity is required.

The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

The low water level instrumentation, set to trip at 168.5 incbes (+4.5 inches) above the top of the active fuel, closes all isolation valves except those in Groups 1, 4, 5, and 7.

Details of valve grouping and required closing times are given in Specification 3.7.

For valves which isolate at this level this trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve closing time.

Required closing times are less than this.

The low low low reactor water level instrumentation is set to trip when the water level is 19 in..es

( 145.5 inches) above the top of the active fuel.

This trip l

closes Groups i and 7 Isolation Valves (Reference 1), rctivates the remainder of the CSCS subsystems, and starts the emergency diesel generators. These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished, Amendment No. /5,87,88,1N,145

-83

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