ML20086K149

Amend 151 to License DPR-46,changing TS to Allow Plant Operation in Plant Operating Domain on Power/Flow Map Above Rated Rod Line

ML20086K149
Person / Time
Site:	Cooper
Issue date:	11/29/1991
From:	Larkins J Office of Nuclear Reactor Regulation
To:
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ML20086K152	List: ML20086K149 ML20086K160
References
NUDOCS 9112120241
Download: ML20086K149 (28)

Text

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!X UNITED STATES 33,

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- 1 NUCLEAR REGULATORY COMMISSION k h.e ~ f WASHINGTON, D C. M%5 gv

....+

_ NEBRASKA PUBLIC POWER DISTRICT DOCKET NO.50-29R COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.151 License No. DPR-46 1.

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

A.

The application for amendment by Nebraska Public Power District (the licensee) dated July 29, 1991, as supplemented October 3, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 regulations of the Comission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without 'ndangering 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 Comission's regulit i a s and all applicable requirements have been satisfied.

9112120241 911129 DR ADOCK 050 0

0 f

2 I

2y AccordinglyIndicated in the attachnient to this license amendment andth i

cations as-I Paragraph 2.C.(2) of Facilit ammded to read as follows: y Operating License No. DPR 46 is hereby

-2.

Technical Specifications The Technical fpecifications contained in Appendix A, as revised

{

through Amendment No.151, are hereby incorporated in the license.

The licensee shall operate the facility ice accordance with the Technical Spec ? lcations.

s 3.

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

FOR THE NUCLEAR REGUI.ATORY COMMIS$f0N John T. tarkins, Director Project Directorate IV-1

~,

Division of Reactor Projects ill, IV, and V-l Office of Nuclear Reactor Regulation Attachment -

Changes to the Technical Specifications j

Date of Issuance: November 29. 1991 1

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w-ATTACHME,NT TO LICENSE AMENOMENT NO. 151 FACILITY OPE'1ATING LICENSE NO. OPR-46 DOCKET NO. 50 298 Replace the following sages of t.:e Appendix A Technicel Specifications with the enclosed paget. T1e revised pages are identified by Amendtnent number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 6

6 7

7 8

8 18 18 19 19 27 27 28 28 31 31 42 42 43 43 61 61 61a 62 6?

62a 62a 77 77 86 86 96 96 102 10?

104 104 210 210 211 211 212 212 214 214 214a 214a i

214b 214b E

s i

SAFFTY LIMITS

,_ LIMITING SAFETY SYSTFM SETT M S 1.1 BTL CIADD1HG.. INTECRITY 2.1 }TEL CIADDINE.1NIfCRITY 6poliesbility Applicability l

The Safety. Limits established to The Limiting safety System Settings i

preserve the fuel claoding integrity apply to trip settings of the in-apply to those variables which moni-strureents and devices which are tor the luel thermal behavior.

provided to prevent the fuel clad-ding integrity Safety Limits from Qbl1911YA being exceeded.

The' objective of the Safety Limits obtective 4

is to establish limits below which

~

the integrity of the fuel cladding The objective of the Limiting Safety

}

is preserved.

System Settings is to define the level of the process variables at j

-Action which automatic protective action is initiated to prevent the fuct clad-If a Safety Limit is exceeded, the ding integrity Safety Limits f rort reactor shall be in at least hot being exceeded, shutdown withir,2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, Specifications Specifications A.

Tric Settineg A.

Reactor Pressure >800 osia and Core Flow >10% of Rated The limiting safety system trip settings shall be as spe-The existence of a minimurn cified below:

critical power ratio (MCPR) less than 1.06 for two recir.

1.

Neutron Flux Trip Settinrs culation loop operation (1,07 7

for single loop operation) a.

APRM Flux Scram Trip Set t inc shall constitute violauton of (Run Mode) the fuel claddirs integrity safety.

(1) Flow Re ferenced Scram Trio Settine B.

' Core Thermal Power tirait;

~

(Reactor Pressure <800 osia When the Reactor Mode Selector and/or Core Flow-(1011 is in the RUN position, the i

APRM flow referenced-flux

?

When the reactor pressure is scram trip setting shall be:

(800 psia or core flow - is S50.58 W + 621 -.58 AW 1ess than 10% of rated, the core thermal per r shall not where:

r exceed 25% of rated thernal S - Setting in percent of power.

rated therrnal power (2}tI Nv')

C.

Power Transient U - Two loop recirculation To ensure that the Safety flow rate in percent of Limit established in Specifi-rated (rated loop cation 1.1. A and 1.1.8 is not recirculatioit flow rate exceeded, each required scram is that recirculation flow shall be initiated by its rate which provides_100%

expected rcram signal.

The core flow at 100% power) l Safety Lirait shall be assumed AW - Dif ference - between two-to be exceeded when scram is loop and single-loop ef-accomplished by a neans other fective drive flow at the than the expected scram sig*

saree core flow.

nal.

Amendment No. 48,P2,88,H,133,151

-6

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- ~. _ _ _ _.. _ _

4 SAFETY LIMITS-LIMIT 1110 SAFETY SYSTEM SEITINGS 1.1 (cont'd) 2.1.A.1 (cont'd)

D.

cold Shutdown AV - O for two recirculation loop operation.

i

)

Whenever the reactor is in the cold shutdown condition with irradiated (2) Eind APP.M Flux Scram Trio Sctti.D&

fuel in the reactor vessel, the water level shall not be less than The fixed APRM flux scram trip set.

i 18 in, above the top of the normal ting shall not be allowed to exceed active fuel zone (top of active fuel 1201 of rated thermal power.

is defined in Figure 2.1.1).

i b.

APRM Flux Scram Trip setting (Refuel or Startup and Hot Standby Mode)

When the Reactor Mode telector Switch is in the REFUEL or STARTUP positi n, the APRM scram shall be 9

set at less than or equal to 15% of rated power.

c.

IEH The IRM flux scram setting shall be i

5120/125 of scale.

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-t No. 32,29,46,20,88,94,151 7-133,142,-

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$AFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.A.1 (Cont'd) d.

APRM Rod Block Trio Settine The APRM rod block trip setting shall be:

S s 0.58 V + $01 -.58 AV g

with a maxinwn of s 108% or rated power.

where Sg - Rod block sotting in percent of rated thermal power (2381 MVt)

W and AV are defined in Specifica-tion 2.1.A.1.a.

l 2.

Reactor Water Low Level Scram and l

Igolation Trio 4,t11Mgxcept MSIV)_

h 412.5 in, on vessci level instru.

ments.

e Amendment No. 16,32,39,41,46,80,91, 8'

133,142,145, 151

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.1 Eastsj (Cont'd) 6 An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety 1.imit is reached. The APRM scram trip setting wa.,

deterrained by an analysis of margins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious scrans which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety 1.imit yet allows operating eargin that reduces the possibility of unnecessary scrams.

j l

b.

APRM Flux Scram Tilgjictrine (Refuel or Ste,rt & llot Standby. Mg4t.1 For operation in the startup mods while thc reactor is at low pressure, the APRM 4

scram setting of 15 percent of rated power provides adequate thernal margin betwee' the setpoint and the safety lirait, 25 percent of rated.

The margin is adequate to accomudate anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, l

tersperature coefficients are areall, and control rod patterns are constrained to be uniform by operating procedure backed up by the rod worth minir.!zer, and the rod sequences control system.

Worth of individual rods is very low in a uniform rod pattern.

Thus, of all possibic sources of reactivity input, uniform control rod withdrawai is the most probabic cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peake, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scran before the power could exceed the safety limit.

The 15-percent APRM scram remains active until the mode switch is placed in the RUN position. This change can occur when reactor pressure is greater than Speelfication 2.1.A.6.

i F

Amendment. No. Shf,142, 151

-18

.2

2.1 hpfa (Cont *d) c.

IPJi_ flux scram Trin Settine The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic thannels. The IRM la a 5 decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one half of a decade in size.

The IRM scrate trip setting of 120 divisions is active in each range of the IRM.

Por exatrple, if the instrument were m range 1, the scrats setting would be 120 divisions for tha: range; likewise, if the instrument were on range 5, che scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal.

For in. sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitatior, of withdrawing control roda, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well bef ore any Safety Limit is exceeded.

In order to ensure that the IRM provided adequate protection against the single rod withd aval error, a range of rod withdrawal accidents was analyzed This analysis included starting the accident at various power levels.

The roost severe case involves an initial condition in which the reactor is j'ist suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density.

Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is by passed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the MCPR fuel cladding integrity rafety limit.

Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

d.

APRM Rod Block Trin Setting Reactor poser level may be varied by moving coritrol rods or by varying the recirculation flow rate.

The APRM system provides a control rod block which is dependent on recirculation flow rate to limit rod withdrawal, thus protecting against a MCPR of less than the MCPR fuel cladding integrity safety limit.

The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore the worst case MCPR which could occur during steady state op" stion is at 108%

of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in core LPRM cystem.

Amendment No. 94,151 _

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_.__._m___.__._____._

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM L

ApplirJtbility:

Arnlicability:

Applies to the instrumentation and Applies to the surveillance of the

. associated devices which initiate a instrumentation and assor:iated de-reactor scram, vices which initiate reactor serem.

Objective,1 Objective:

To assure the operability of the To specify the type and frequency of reactor protection system, surveillance to be applied to the protection instrumentation.

Specification:

Specification:

The - setpoints, minimum number of A.

Instrumentation systems shall trip systems, and minimum number of W

fM 1

d ed instrument channels that must be calibrated as indicated in f

operable for each position of the Tables 4.4.1 and 4.1.2 resPec-reactor mode switch shall be as

• IV'I '

Y given in Table 3.1.1.

B.

Deleted.

C.

Deleted.

D.

When it is determined that a channel has failed in the un-safe condition, the other RPS channels that monitor the same variable shall be functionally tested immediately before the trip system containing the failure is tripped.

The trip system containing the unsafe failure. may be placed in the untripped condition during the period in which surveillance testing is being performed on the other RPS channels.

Amendment _ No. 16,32, M,#6,80,151

.. 27-

. -. - ~ - -

COOPER NUCLEAR STATION y

m TABLE 3.1.1

{

REACTOR PROTECTION SYSTEM.INSTRUMENTATIO* REQUIREMENTS 3

'[

. Applicability Conditions of Operable When Equipment

+

Minimum Number Action Required 6

teactor Protection.

Mode Switch Position Trip Level Channels Per Operability is I

y System Trip Function Shutdown Startup. Refuel.

Run Setting Trip Systems (1) Not Assured (1) w

,@ - Mode Switch in Shutdown X(7)

X X

X 1

A k - Manual Scra:n X(7)

X X

X 1

A

?

IRM (17).

X(7)

X X

(5)

$ 120/125 of in-3 A

iligh Flux dicated scale U

Inoperative.

X X

(5) 3 A

I'

.APRM (17)

X s (0.58U+621-0.58AW) 2 C

High Flux (Flow biased)

(14)(19)

.l

.with a maximum of j

120I of rated power H1 h Flux (fixed)

X

$ 120I Rated Power 2

A 4

5 High Flux X(7)

X(9)

X(9)

(16)

$ 15% Rated Power 2

A Inoperative X(9)

X(9)

I (13) 2 A

f, Downscale (12)

'(12)

(12)

X(ll) 2 2.5%

2 A

High Reactor Pressure

.X(9)

X(10)

X

$ 104~,psig 2

A l

NBI-PS-55 A,B,C, A '

High Drywell Pressure X(9)(8) X(8)

X

$ 2 psig 2

A or D

[

PC-PS-12 A,B C. & D Reactor Low Water Level X

X X

2 + 4.5 in. Indi-2 A or D NBI-LIS-101 A B.C. & D cated level 1-Scram Discharge Instrument Volume X

X(2)

X

$ 92 inches 3 (18)

A High Water Level CRD-LS-231 A & B-j-

CRD-LS-234 A & B.

j

.CRD-LT-231 C & D CRD-LT-234 C & D j.

11. The APRM downscale trip function is only active when the Reactor Mode Selector Switch

}

is in RUN. When in RUN, thi.

metion is automatically bypassed when the companion IRM instrumentation is operable and not upscale.

12. The APRM downscale trip is automatically bypassed when the Reactot Mode Selector Switch is not in RUN.

13. An APRM will be considered inoperable if there are less than 21.PRM inputs per icvel or there is less than 11 operable 1.PRM detectors to an APRM.

14. W is the two loop rec heulation flow in percent of rated flow.

15. This note deleted.

16. The 151 APRM scram is bypassed when the Reactor Mode Selector Switch is in RUN.

l

17. The APRM and IRM instrument channels function in both the Reactor Protection System and Reactor Manual Control System (Control Rod Withdraw Block, Section 3.2.C.).

A f ailure of one channel will af fect both of these systeins.

18 The minimum number operable associated with the Scrain Discharge Instrument Volume are threc instruments per Scram Discharge Instrument Volume and three level devic.es per RPS channel.

19. AW is the difference between two loop and single loop effective drive flow and is used for single recirculation loop operation.

AW-0 for two recirculation loop operation.

Amendment No. 39,46,80 P2,88,94, 31 108, 151

o LIMITXNG CONDITf0NS FORflPERATION SURVEILIANCE REQMIREMENTS 3.1 BASES (Cont'd.)

4.1 BASES (Cont'd.)

there is proper overlap in the neu-Por the APRM system, drif t of elec-tron monitoring system functions and tronic apparatus is not the only thus, that adequate coveraSe is consideration in deterinining a cali-provided for all ranges of reactor bration frequency. Change in power operation, distribution and loss of chamber sensitivity dictate a calibration every seven days.

Calibration on this frequency assures plant opera-tion at or below thermal limits.

A comparison of Tables 4.1.1 and 4.1.2 indicatos that two instrument channels have not been included in the latter table. These are: mode switch in shutdown and manual scram.

All of the devices or sensors asso-ciated with these scram functions are simple on off switches and, hence, calibration during operation is not applicable.

B.

The sensitivity of 1.PRM detectors decreases with exposure to neutron flux at a slow and approximately cons ant rate. This is compensated for in the APRM system by calibrat-ing once a week using a heat balance data and by calibrating individual LTW s every six weeks of power operation above 20% of rated power.

i i

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l Amendment No. 16,32,33,46,94, 151

-42

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

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" INTENTIONALLY LEFT BIANK" i

J Amendment No.- ff,32.39,46.80,151 -

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COOPER NUCLEAR STATION

[h TABLE 3.2.C CONTROL ROD WITHDRAWAL BLOCK INSTRUMFliTATION

.m a

E Minimu:n Number Of Function-Trip Level Setting Operable Instrument I

gg Channels / Trip System (5) l b

([

APRM Upscal* (Flow Bias)

$ (0.58V + 50% - 0.58 AV) (2)(13) 2(1) with a maximum of 5108% of rated power es la APRM Upscale (Startup)

' 2" APRM Downscale (9) s 12%

2(1) 2 2.5%

2(1) 4 b 03.

j ; ;,

APRM Inoperative (10b) 2(1)

!: ca RBM Upscale (Power Referencad).

(14) (16) 1(15) l Lowest Rated LTSP ITSP HTSP l

MCPR Limit (LPSP $ P < IPSP)

(IPSP $ P < HPSP)

(HPSP s P) 1 2 1.20 s 117.0/125 s 111.2/125 s 107.4/125 l

2 1.25 s 120.0/125 s 115.2/125 s 110.2/125 2 1.30 s 123.0/125 5 118.0/125 s 113.2/125

' S:

RPM Power Range:

(14)

Not Applicable (15)

Low ?ower Setpoint (LPSP) s 30% of rated power Intermediate Power Setpoint (IPSP) s 65% of rated power l

High Power Setpoint (HPSP) s 85% of rated power 1

RBM Downscale (9) 2 91/125 (16).

1(15) l RBM Ynoperative (10c) (15) 1(15)

IRM Upscale (8) s 108/125 of Full Scale 3(1) l 1

IRM Downscale (3)(8) 2 2.5/125 of Full Scale 3(1)

=

IRM Detector Not Full In (8) 3(1) i i

l'

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COOPER NUCLEAR STATION S

TAELE 3.2.C (page 2)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION E

g.

Minimum Number Of' Function Trip Level Setting Operable Instrument g.

Channels / Trip System (5)

IRM Inoperative (8)

(10a) 3(1) i SRM Upscale (8)-

's 1 x 10 Counts /Second 1(1)(6)

SRM Detector Not Full In (4)(8)

(2 100 cps) 1(1)(6)

SRM Inoperative-(8)

(10a) 1(1)(6)

Flow Bias Comparator 5 101 Difference In Recirc. Flows 1

Flow Blas Upscale /Inop.

5 110% Rectre. Flow 1

SRM Downscale (8)(7) 2 3 Counts /Second (11) 1(1)(6) h SDV Water Level High

.$ 46 inches 1(12)

O CRD-231E, 234E I

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NOTES FOR TABl.E 3.2.C 1.

For the startup and run positions of the Reactor Mode Selector Switch, the Control

- Rod Withdrawal Block Instrumentation t *1p systern shall be operable f or each function.

The SRM and IRH rod blocks need not be operable in the RUN mode, and the APRM (flow biased) rod blocks need not be operable in STARTUp mode.

The Control Rod Withdrawal Block Instrumentation trip systein is a one out of "n" trip systein, and as such requires that only one instrument channel specified in the function column must exceed the Trip 1.evel Setting to cause a rod block.

By utilizing the RPS bypass logic (see note 5 below and note 1 of Table 3.1.1) for the Control Rod Withdrawal Block Instrwnent tion, a suf ficient number of instrument channels will always be operable to provide redundant rod withdrawal block protection.

2. W is the two. loop recirculation flow rate in percent of rated.

Trip level setting is in percent of rated power (2281 MVt).

3.

IRM downscale is bypassed when it is on its lowtst range.

4.

This function is bypassed when the count is 2 100 cps and IRM above range 2.

5.

By design one instrument channel; i.e., one APRM or IRM per RPS trip system raay be bypassed.- For the APRM's and IRM's, the minimma inumber of channels specified is that sninianum number required in each RPS channel anc' does not refer to a minimum number required by the control tod block instrumentation trip function. By design only one of two RAM's or one of four SIW s may be bypassed. For the SRM's, the rnintinua number of channels specified is the minimum inumber required in each of the two circuit loops of the Control Rod Block Instrumentation Trip Systers.

6.

IRM channels A,E,C,0 all in range 8 or higher bypasses SRM channels A6C functions.

'IRM channels B,F,D,ll all in range 8 or higher bypasses SRM channels B&D functions.

7.

This function is bypassed when IRM is above range 2.

1 8.

This function is bypassed when the Reactor Mode Selector Switch is placed in RUN.

9.

This function is only active when the Reactor Mode Selector Switch is in RUN.

10. The inoperative trips are produced by the fo11owin6 functions:

a, SRM and IRM (1) Mode switch not in operate (2)- Power supply voltage low t

(3) Circuit boards not in circuit t

(4). l.oss of negative supply voltage L

e 1

1

- Amendment. NO.: 27,93,94,108,115, 62 1#2,.151 t

l l

NOTES POR TABLE 3.2.0 (Continued) b.

APRM i

(1) Mode switch not in operate (2) Less than 11 LPRM inputs (3) Circuit boards not in circuit c.

RBM (1) Mode switch not in operate (2) Circuit boards not in circuit (3) RAM fails to null (4) less than required number of LPRM inputs for rod selected

11. During spiral unloading / reloading, the SRM count rate will be below 3 eps for-some period of time.

See Specification 3.10.B.

12. With the number of L,PERABLE channels less than required by the Minimum Number t

of Operable Instrument Channels / Trip System requirements, place the inoperable channel in the tripped condition within one hour.

13.

AW is the difference between two. loop and single. loop effective drive flow and j

is used for single recirculation loop operation. AV-0 for two recirculation loop operation.

14 One set of power referenced RBM upscale trip settings (LTSP, ITSP, and itTSP) is applied based or, the lowest rated MCPR limit given in Specification 3.11.C.

The P2M power range setpoints control the enforcement of the appropriate upscale tripp over the proper core thermal power range as follows:

a.

All RBM trips are automatically bypassed below the Low Power Setpoint (LPSP).

b.-

The upscale Low Trip Setpoint (LTSP) is applied at the LPSP and up to the Intermediate Power Setpoint (IPSP).

s c.

The upscale Intermediate Tril, Setpoint'(ITSP) is' applied at the IPSP up to the High Power Setpoint (llPSP),

d.

The upscale liigh Trip Setpoint (llTSP) is applied at and above tM HPSP.

t

15. The RPM is only required when core thermal power is h 30% of rated power and limiting control rod pattern (defined in Specification 3.3.B.5) exists.

t a

Requirements for operating with a limiting control rod pattern are specified in Specification 3.3.B.5.b.

16. -RBM trip level settings are relative to the initialization reference signal of 100/125 of full scale which takes piace upon control rod selection.

b L

JAmendment No. 61,7/7/,H 108,151 62a.

~

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TABLE 4.2.C" 1

g SURVEILIANCE REQUIREMENTS FOR ROD WITHDRAVAL BIDCK INSTRUMENTATION e

i..

S Functional P

Function' Test Freq.

Calibration Freq.,

Instrument Check y.

. g AFRM Upscale (Flow Bias)

(1)

(3)

Once/3 Months Orce/ Day AFRM Upscale (Startup Mode)

(1)

(3)

Once/3 Months Once/ Day.

AFRM Downscale (1)

(3)

Once/3 Months Once/ Day AFRM Inoperative (1)

(3)

N.A.

Once/ Day RBM Upscale (Power Referenced)

(1)

(3)

Once/6 Months Once/ Day RBM Fower Range (3)

Once/6 Months N.A.

RBM Downscale (1)

(3)

Once/6 Months once/ Day RBM Inoperative (1)

(3).

N.A..

Once/ Day a:

IRM Upscale (1).(2)

(3)

Once/3 Months Once/ Day IRM Downscale (1)

(2)

(3)

Once/3 Months Once/ Day IRM Detector Not Full In (2)' (Once/ operating cycle Once/Oper. Cycle (10)

Once/ Day IRM Inoperative (1)- (2)

(3)

N.A.

N.A.

SRM Upscale-(1)

(2)

(3) ence/3 Months Once/ Day SRM Downscale (1)

(2)

(3)

Once/3 Months Once/ Day Z

SRM Detector Not Full In (2)

(Once/ operating cycle Once/Oper. Cycle (10)

N.A.

SRM Inoperative (1)

(2)

(3)

N.A.

N.A.

t Flow Bias Comparator (1)

(8)

Once/Oper. Cycle N.A.

Flow Bias Upscale (1)' (8)

Once/3 Months N.A.

Rod Block Logic.

(9)

N.A.

N.A.

RSCS Bypass (1)

Once/3 Months N.A.

SDV High Water Level Quarterly Once/Oper. Cycle N.A.

i.

a i

t i

1 m

m

3.2 BASPS (cont'd.)

C.

Control Rod Blo.c.h Actuatten The conttd rod block functions are provided to provent excessive control rod withdrawal so that MCPR does not decrease to the safety limit CPR.

The trip logic for this function is 1 out of n:

e.g., any trip on one of six APRM's, eight IRM's, or four SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single feilure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This time period is only 31 of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is f'.ow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow.

The APRM provides gross core protection; i.e.,

limits the gross core power increase froin withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the safety limit CPR.

The RBM rod block function prov0es local protection of the core; 1.e.,

the prevention of the MCPR reaching the safety limit CPR in a local region of the core, for a single rod withdrawal errca f rem a limiting control rod pattern.

Additional details nre provided in Bases Section 3.3.B.$.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factut of 10 above the indicated level.

A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The downscale trips are set at 2.5 indicated on scale.

The fh

'sparator and scram discharge volt no high level components have only one logic cw..41 and are not required for safety.

The SDV high level rod block does.

provide adequato time to determine the cause of the level increase and take corrective action prior to automatic scram.

The refueling interlocks also operate one lob c channel, and are required for safaty i

only when the mode switch is in the refueling position.

D.

Radiation Monitorine Svitems Isolation and Initiation Functions 1.

Steam Jet Air Ejector Off-cas System Two air ejector off gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off gas line. Isolation is inittaced when both instrwnents reach their high trip point or one has an upscale trip and the other a

+

downscale trip.

There is a - fif teen minute delay accounted for by the 30 minute holdup time of the off. gas befc re it reaches the stack.

1 i

l L

l-Amendment Nc. 30,61,77,83,151 l l-

O 4.

l 1.1MITING CONDITION FOR OPEIMTION SURVEILIANCE REOp1REMENT _,

3.3.B.3 (cont'd)

A.3.B.3 b (cont'd) e.

If Specifications 3.3.B.3a 1)

The correctness of the Banked through d cannot be met, the Position Withdrawal Sequence reactor shall not be started, input to the RVM computer or if the reactor is in the run shall be verified.

or startup modes at Iras than 20% rated power, it shall be 2)

The RkH computer on line diag.

brought to a shutdown condition nostic test shall be success.

immediately.

tully performed.

f.

The sequence restraints imposed 3)

Proper annunciation of the on the control rods-may be re-selection error of at least moved by the use of the indi.

one out of sequence control vidual rod position bypass rod in each fully inserted switches for scram testing only group shall be verified, those rods which are fully.

withdrawn in the 100% to 50%

4)

The rod block function of the rod density range.

RkH shall be verified by with.

drawing the first rod as an 4.

Control rods shall not be with, out.of sequence control rod no drawn for startup unless at more than to the block point.

t i

1 cast two scarce rance channels have an observed count rate c.

When required, the presence of 4

equal to or greater than three a second licensed operator or counts per second.

other qualified employee-to verify the following of the j

5.

OperatLE'L.Yli a_Lipiting correct rod program shall be

- Control Rod Partein (for Rod verified.

Pitfidraval irrer. RWE) 4 Prior to control rod withdraw.

a.

A Lluiting Control Rod Pattern al for startup, verify that at for RVE exists when either:

least two source range chan.

nels have an observed count

1) Core thermal power is 2 30%

rate of at least three counts and < 90% of rated power per second.

and the MCPR is less than 1.70. or 5.

OprJatLon with a 1.imiting E2ntrol Red Partern ( f or_.. Rod

2) Core thermal power is > 90%

Withdrawal Error. RWE) of rated power and the MCPR is less than 1.40.

During operation when a Limit-ing Control Rod Pattern for b.

During operation with a Limin RWE exists and only one RNi ing Control Rod Pattern for RWE channel is operable, an in-either:

strument functional test of' the RAM channel shall be per.

1)- Both rod block monitor formed prior to withdrawal of (RBM) channels thall be the control rn'(s).

A 1.imit-opernble, or ing Control xud Pattern for i

RWE is defined by Specifica.

2) Vith one BM channel inop.

tion-3.3.B.5.

erable, control, rod with-drawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.or 3)- With both RBM channels in-operable, hall control rod with-drawal s

be blocked until operability of at least ono channel is re-stored.

4 Amendment No. H7,151 96-

__ =

,J y.,.

-.m_

m.

. ~.,. _.

...__..m.

r--m

3.3 and 4.3 MSI21 (Cont *d) 5.

The RBM provides local protection of the core; i.e., the Trevention of boiling transition in a local region of the core, for a single rod withdrawal error from a Lissiting Control Rod Pattern a6 defined in Specification 3.3.P.S.

The trip point is referenced to a power signal provided by the APRMs.

A statistical analysis (Reference 5) of many ningle control rod withdrawal errors has been performed.

At the 95/95 level the results show that with the specified trip settings, rod wathdrawal is blocked at MCPRs which are greater than Safety Liisit 1.1, A, thus allowing adequate saargin. This analysis assumed operation at steady state operating limit MCPRs (Specification 3.11.C) prior to the postulated rod withdrawal error.

The RBM functions are required when core thermal power is (Specification 3.3.B.3) greater than 30% and a Limiting Control Rod Pattern exists.

When both RBM channels are operating either channel will assure required withdrawal blocks occur even assuming a single failure of one channel. Vhen a Limiting Control Rod Pattern exists, with one RBM channel inoperable for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, t6 sting of the RBM before withdrawal of control rods assures that improper control rod withdrawal will be blocked.

A Limiting _ Control Rod Pattern for rod withdrawal error (RWE) exists when

_ (a) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% $ P < 90%) anu the MCPR is less than 1.70, or (b) core thermal power is greater than or equal to 90% of rated (P 2 901) and MCPR is less than 1.40.

Under these conditions, corrplete withdrawal of a control rod could result in MPR violating the safety limits.

Therefore, RBH operation is required to blok control rod withdrawal. RBM setpoints have been selected such that the required cortrol rod bbeks shall occur even if one of the redundant RBM channels fails.

Shou'.d one RBM channel become inoperable during the use of such patterns, it is j % % that testing of the RBM system before withdrawal of such rods assures the remaining channel operability.

-C.

Agram Insertion Titres lhe control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel dange; i.e., to prevent the MCPR from becoming ISs than the safety limit. The liraiting power transient is defined in Reference 2.

n... lysis of this transient shows that the negative reactivity rates resulting from the scram provide the require orotection, and MCPR remains greater than the safety limit.

The surveillance requirement for scra.n testing of all the control rods after each refueling outag,e and 10% of the control rods at 16 week intervals is adequate for determining the operability of the control rod system yet is not so frequent as to cause excessive wear on the control rod system components.

The numerical values assigned to the predicted scrarn performance are based on the analysis of data frota other BWR's with control rod drives the same as those on Cooper Nuclear Station.

The occurrence of scram times within the limits,-but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod drives.

In the analytical treatment of the transients which are assumed to serain on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and start of motion of the control rods.

This is adequate and conservative when compared to the typical time delay cf about 210 milliseconds estimated from scram test results. Approximately the first 90 milliseconds of each of these time antervals result from the sensor and circuit delays; at this point, the pilot scram solenoid deenergized. Approximately 120 in1111 seconds later, Amendraent No'. 27,71,80,f00,133,1#7, 151 102

e-3.3 and 4.3 EASES:

(Cont'd)

G.

Scram Discharte Volume To ensure the Scram Discharge Volume (SDV) does not fill with water, the vent and drain valves shall be verified open at least once every 31 days. This is to preclude establishing a water inventory, which if sufficiently large, could result in slow scram times or only a partial control rod insertion.

The vent and drain valves shut on a scram signal thus providing a contained volume (SDV) capable of receiving-the full volume of water discharged by the control rod drives at any reactor vessel pressure, rollowing a scram the SDV is discharged into the reactor building drain system.

REFERENCES 1.

"Ceneral Electric Standard Application for Reactor Fuel," NEDE 24011 P A-(1stest approved revision).

2.

" Supplemental Reload Licensing Submittal for Cooper Nuclear Station," (applicable

-reload document).

3.

General Electric Service Information Letter No. 380, Revision 1, dated February 10, 1984.

4.

General Electric Service Information Letter No. 316. Reduced Notch Vorth Procedure, November, 1979, 5.

" Extended Load Line Limit and ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14, 'NEDC 31892P, Revision 1, May 1991.

Amendment No. 177,133,151 104

LIMITING CONDITIONS FOR OPERATION SI'RVElldNiCE RffRillMUIS_

3.11 1TEL RODS 4.11 ITEL RODS Applicability Applj.cabilit y The Litelting Conditions for Opera.

The Surveillance Requirements apply tion associated with the fuel rods to the parameters which monitor the apply to those parameters whien fuel rod operating conditions.

tronitor the fuel rod operating con-ditions.

Objective Objective The Objective of the Surveillance Requirernents is to specify the type The Objective of the IJtniting Condi-and frequency of surveillance t o be tions for Operation is to assure the applied to the fuel rods.

performance of the fuel rods.

Specificatiom Specificatio n A.

Ave ra gdlanur_Line a r. lie a t A.

Averagdlansr Lincar_llrM Genention Kate ( APIRGE1 Ceneration Rate (APuiGR)

The APulGR for each type of During steady state power operation, fuel as a function of average with both recitculation loops oper-planar exposure shall be oe-i l

ating, the APUICR for each type of termined daily dur'.ng reactor fuel as a function of average planar operation at 2251 rated ther-exposure shall not exceed the limit-taal power, ing value specified in the Core Operating Limits Report for two recirculation loop operation.

For single loop operation these values are reduced for each fuel type as specified in the Core Operating y

Limits P.eport.

If at any tirne dur-ing steady state operation it is determined by normal surveillance that the limiting value for APUlGR is being exceeded action shall be initiated within 15 minutes to re-store operation to within the pre-scribed limits.

If the APUlGR is not returned to within the pre-scribed limits within two (2) hours, reduce reactor power to s 25% of rated power within the next four (4) hours. Surveillance and correspond-ing action shall continue until the prescribed limits are again being

~

met.

I I

Amendment No. 55,H 106,133,142,151

-210-l

M LIMITING CONDITIONS FOR OPERATIO!j SURVEILIANCE REOUIREMENTS 3.11. fi Heat Generation Rate 4.11.B Linear llent Generation Rate wm During steady state power operation.

The 1)lGR as a function of core the linear heat generation rate height shall be checked daily during (1JIGR) of any rod in any fuel assem.

reactor operation at ;t 251 rated bly at any axial location shall not thermal power.

exceed the maxistus allowable illGR as specified in the Core Operating Lignits Report.

If at any time during steady state operation it is determined by normal surveillance that the limiting value for IJ1CR is being exceeded, action shall then be initiated to restore operation to within the prescribed limits.

Surveillance and corres-ponding action shall continue until the prescribed limits are again being met.

i i

?

f i

i Amendment No. 77,0 3,f n,151 211 m,._..,..-,_,_,_,

L1HITING CONDITAONS FOR OPEPATION SURVEILIANCE _ RPOUIRP.MENTS I

C.

Minitrum Critical Power Ratio (MCPR)

C.

liinimum Critical Power Ratio (MCPR)

During steady state power operation the HCPR for each type - of fuel at MCPR shall be determined daily dur-ing reactor power operation at > 25%

rated power and flow shall not be rated thermal power and following I

lower than the limiting value speci-fled in the Core Operating 1.imits any change in power level or distri-Reports for two recirculation loop bution that could cause operation on operation.

If, at any time during the operating limit MCPR.

steady state operation it is deter.

mined by normal surveillance that the limiting value for MCPR is being exceeded, action shall then be ini-tiated within 15 minutes to restore operation to within the prescribed limits. If the steady state MCPR is not returned to within the pre-scribed limits within t.wo (2) bours, reduce reactor power to s 25% of rated power within the next four (4)

-hours, Surveillance and correspond-ing action shall continue until the prescribed limits are again being mot.

g L.

For one recirculation loop operation

- l the MCPR limits are 0.01 higher than the comparable two loop values.

L t

Amendment No. FJ,H,733,1#2,151

-212-e

.,,.,,-,.e m

,,---<>e-.a,~

m..

---.w-

,mv-,w,

~

w cc.-+.

eu

'I 4

3.11 BASES A.

AverargJlanar Linear fleat Generation Rate (ApulGR)

This specification assures that the peak cladding temperature following the postulated design basis loss of coolant accident will not exceed the limit specified in 10CFR50.46.

The peak cladding temperature following a postulated loss of coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.

Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than i 20*F relative to the peak temperature for

'7 a typical fuel design, the limit on the average linear heat generation tatn is sufficient to assure that calculated teperatures are within the 10CFR50.46 limit.

The limiting value for APulGR fct each fuel type is specified in the Core Operating Limits Report.

The flow dependent correction factor MAPFACr specified in the Core Operating Limits Report is applied to the MAP 141tR limits at rated conditions to assure that (1) the 10CFR50.46 limit would not be exceeded during a lhCA initiated f rom less than rated core flow conditions and (2) the fuel thermal mechanical design criteria would be met during abnormal operating transients initiated from less than rated core flow conditions (Reference 11).

The power dependent-correction factor MAPFAC, specified in the Core Operating Litaits Report is applied to the MAPulGR limits at rated conditions to assure that the. fuel theriaal mechanical design criteria would be net during abnormal operating transients initiated from'less than rated core power conditions (Reference 11).

The ApulGR values are reduced for single loop operation per Reference 10.

B.

Linear Heat Cencration Rate (UlGR)

This specification assures that the linear heat generation rate in any rod is less than.the design linear heat generation if fuel pellet densification is postulated. The UlGR as a function of core height shall be checked daily during reactor operation at 2 25% power to determine if fuel burnup, or control rod movement has caused changes in power distribution.

For UlGR to be a limiting value below=25% rated thermal power, the MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible-control rod pattern.

Pellet densification power spiking in GE fuel has been accounted for in the safety analysis presented in References 1 and 2; thus no ad,)ustment to the U1GR limit for densification effects is required.

Amendment No. 62,80,9#,133,1;2, 151-

-214-r w,

,--n.

n e,~

.e r n. - - - -. --

n

---~-

p.

i o

i 3.11 Bases:

(Cont'd)

C.

Minimwn Critical Power Ratio (MCPR)

Tho. required operating limit MCPR's at steady state operating conditions as specified in Specification 3.11C are derived from the established fuel cladding integrity Safety Limit and an analysis of abnormal operational transients (Reference 2 and 11). For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the more liniting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).

The models used in the transient analyses are discussed in Reference 1.

i Flow dependent and power dependent MCPR litsits (MCPRr and MCRPp) are used to define the required Operating Liteit MCPR (OLCPR) such that the above Safety Limit MCPR requirement is met for all power / flow conditions. MCRP, provides the

- thermal margin required to protect the fuel from transients resulting from inadvartent core flow increases. MCRPp protects the fuel from the other liiniting i

abnormal operating transients, including localized events such as a rod withdrawal error.

J Direct scram on Turbine Stop Valve Closure or Turbine control Valvo fast closure provides the fastest response to an abnorinal operating transient such as load rejection, turbine trip, or feedwater controller failure. These direct scrams are bypassed at low power (Peyp...), to reduce the frequency of serains during power ascension.

For operation at or above P,,,, (30% of_ rated power), the 37 required OlftCPR is the larger of MCPRr or MCRP, at the existing core power / flow state; where MCPRr and MCRPp are determined in the Core Operating Lignits Report by multiplying the scram time dependent MCPR limit for rated power and flow MCPR(100) by the K, factor. Below 30% of rated power, when the direct scrams are bypassed, a slightly more severe transient response results. To compensate for the anore severe transient response,- two power dependent MCPR limits are established, one for high flow (>50% of. rated) conditions and one for low flow (550% of rated) conditions.

These limits are specified in the Core Operating Limits Report.

Further information on the MCPR operatinj; liinits for off rated-conditiv i is presented in Reference 11.

References for Bases 3.11

'1.

"Ceneral Electric Standard Application for Reactor Fuel," NEDE-24011 P. A.

(The approved revision at the time the reload analyses are performed.)

The approved revision number shall be identif'ied in the-Core operating Limits Report.

2.

" Supplemental Reload Licensing Submittal for Cooper Nuclear Station," (applicable reload document).

3 8.

Deleted 9.

Letter (with' attachment), R. H. Buckholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Madel," September 5, 1980.

10.

" Cooper Nuclear Station Singleiloop Operation," NEDO 24258, 11.

" Extended Imad Limit. and ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14," NEDC 31892P,' Revision 1, May 1991.

Amendment No.. 62,80,94,142,151

-214a.

I i

o f

4.11 lagn:

f A&B.

Average and Local LHCR l

The IJtGR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

L C.

Minimum Critical Power Ratio _(MCPR) - (Survelll g e_.Et.quirement)

'i At core thermal power levels less than or equal to 251, the reactor will be operating at less than or equal to minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting HCPR value is-in excess of requirements by a considerable margin, With this low void L

content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

During initial start up testing of the plant, a MCPR evaluation was made at 251 thermal power level w!th minimum recirculation pump speed.

The MCPR margin wAs thus demonstrated such that subsequent MCPR evaluation below this power level was shown to be unnecessary.

The daily requirement for calculating MCPk above 251 rated thermal power is sufficient since power distribution shifts are very slow when - there have not been significant power or control rod changes.

The requirement for calculating MCPR when an operating limit MCPR is l

approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that-could place operation at a thermal limit.

- Amendment No. 62,80,7#,I#2,151 214b.

.