ML20108C404

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Proposed Tech Specs,Relocating Variable Low RCS Pressure Trip Setpoint & Associated Protective Limits to COLR
ML20108C404
Person / Time
Site: Arkansas Nuclear 
Issue date: 04/29/1996
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20108C395 List:
References
NUDOCS 9605060311
Download: ML20108C404 (27)
v • d • e


Text

_.

PROPOSED TECENICAL SPECIFICATION CHANGES 9605060311 960429 PDR ADCCK 050003i3 P

PDit

l LIST OF FIGURES Numbeh Title Eaga l

3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR COOLANT SYSTEM, NO2)(AL OPERATION COOLDOWN LIMITATIONS 20c 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H-O 33 i

i 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT i

INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c f

3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110be 4.4.2-1 NORMALIZED LIFTOFF FORCE - HOOP TENDONS 85b 4.4.2-2 NORMALIZED LIFTOFF FORCE - DOME TENDONS BSc l

4.4.2-3 NORMALIZED LIFTOFF FORCE - VERTICAL TENDONS 85d l

l 4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 110o2 4.18.3.a.3 l

5.1-1 MAXIMUM AREA BOUNDARY FOR RADIOACTIVE RELEASE CALCULATION l

(EXCLUSION AREA) lila j

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6.2-1 MANAGEMENT ORGANIZATION CHART 119 l

6.2-2 FUNCTIONAL ORGANIZATION FOR PLANT OPERATIONS 120 1

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l Amendment No. E,M,M,MS,4H, iv l

4 mom 9 l

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2.,

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Acolicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow when the reactor is critical.

Obioetive To maintain the integrity of the fuel cladding.

Specification 4

Themaximumlocalf3elpincenterlinetemperatureshallbe 2.1.1 s 5080 - (6.5 x 10-x Burnup, MWD /MTU)*F) for TACO 2 applications and $4642 - (5.8 x 10-3(x (Burnup, MWD /MTU)*F for TACO 3 applications.

Operation within this limit is ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip Setting Limits," as specified in the COLR.

2.1.2 The departure from nucleate boiling ratio shall be maintained greater than the limits of 1.3 for the BAW-2 correlation and 1.18 for the BWC correlation. Operation within this limit is ensured by compliance.with specification 2.1.3 and with the Axial Power Imbalance protective limits preserved by Table 2.3-1

" Reactor Protection System Trip Setting Limits," as specified in the COLR.

2.1.3 Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Variable Low RCS Pressure-Temperature Protective Limits as specified in the COLR.

Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. Thu upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB).

At this point there is a sharp reduction of the heat transfer coefficient, which could result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation. The BAW-2(1) and BWC(2) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The BAW-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30 (BAW-2) and 1.18 (BWC).

Amendment No, M,-lH,He 7

A QNBR of 1.30 (BAW-2) or 1.18 (BWC) corro: ponds to a 95 pere nt probability at a 95 percent confidence level that DNS will not occur; this is considered a conservative margin to DNB for all operating conditions.

The difference between the actual core outlet pressure and the indicated reactor coolant system pressure for the allowable RC pump combination has been considered in determining the Variable Low RCS Pressure-Temperature Protective Limits.

The Variable Low RCS Pressure-Temperature Protective Limits presented in the l

COLR represent the conditions at which the DNBR is greater than or equal to the i

minimum allowable DNBR for the limiting combination of thermal power and number of operating reactor coolant pumps which is based on the nuclear power peaking factors (3) as specified in the COLR with potential fuel densification effects.

The Axial Power Imbalance Protective Limits in the COLR are based on the l

more restrictive of two thermal limits and include the offects of potential fuel densification:

1.

The DNBR limit produced by the limiting combination of the radial peak, axial peak, and position of the axial peak.

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l 2.

The combination of radial and axial peak that prevents central' fuel molting at the hot spot as given in the COLR.

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Power peaking is not a directly observable quantity and therefore limits l

have been established on the basis of the reactor power imbalance produced by the power peaking.

The flow rates for the Variable Low RCS Pressure-Temperature Protective Limits specified in the COLR correspond to the expected minimum flow rates with.four pumps, three pumps, and one pump in each loop.

The variable Low RCS Pressure-Temperature Protective Limit for four reactor coolant pumps operating is the most restrictive of all possible reactor coolant j

pump maximum thermal power combinations as specified in the COLR.

The Variable Low RCS Pressure-Temperature Protective Limits in the COLR represent the conditions at which the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.

If the actual pressure /temperatare point is below and to the right of the pressure / temperature line, the Variable Low RCS Pressure-Temperature Protective Limit is exceeded.,

l The local quality at the point of minimum DNBR is less than 22 percent (BAW-2)(1) i or 26 percent (BWC)(2).

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j Amendment No. M,H,99,444,448 8

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'Using a local quality limit of 22 perc nt (BAW-2) er 26 perc:nt (BWC) ct

/

the point of minimum DNBR as a basis for less than four reactor coolant pumps operating'of the Variable Low RCS Pressure-Temperature Protective Limits specified in the COLR.is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

The'DNBR as calculated by the BAW-2 or the BWC correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.

The maximum thermal power, as a function of reactor coolant pump operation is limited by the power. level trip produced by the flux-flow ratio (percent 1

l flow x flux-flow ratio), plus the appropriate calibration and instrumentation errors.

l For each combination of operating reactor coolant pumps of the Variable Low RCS l

-Pressure-Temperature Protective Limits specified in the COLR, a pressure-(

temperature point above and to the left of the curve would result in a DNBR greater than 1.30'(BAW-2)'or 1.18 (BWC) or a local quality at the point of I

minimum DNBR less than 22 percent (BAW-2) or 26 percent (BWC) for that particular reactor coolant pump combination.

The Variable Low RCS Pressure-Temperature Protective Limit for four reactor coolant pumps operating is the most restrictive because any pressure-temperature point above and to the left of this curve will be above and to the left of the other curves, l

t l

REFERENCES j.

(1)

Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized

[

Water, BAW-10000A, May, 1976.

(2)

BWC Correlation of critical Heat Flux, BAW-10143P-A, April, 1985.

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(3)

FSAR, Section 3.2.3.1.1.c.

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Amendment No. M,M,6,M,M,99,4H 9

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pump 3(o). Ths pump monitora also rostrict th3 pow 3r loval'for the number of pumps in operation.

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RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip is reached before the nuclear overpower trip setpoint. The trip l

setting 1Lmit'shown in Table 2.3-1 for high reactor coolant l

system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any l

design transient.(2)

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The low pressure (1800 psig) and variable low pressure (COLR) trip l

setpoint shown in Table 2.3-1 have been established to maintain the i

l DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.(2,3)

To account for the calibration and instrumentation errors, the accident analysis used the protective limit specified in the L9LR.

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D.

Coolant Outlet Temperature 1

The high reactor coolant outlet temperature trip setting limit l

(618F) shown in Table 2.3-1 has been established to prevent l

f excessive core coolant temperatures in the operating range.

Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 F.

l E.

Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig)-

l provides positive assurance that a reacaor trip will occur in l

the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of i

a low reactor coolant system pressure trip.

F.

Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for i

bypassing certain segments of the reactor protection system.

The reactor protection system segments which can be bypassed l

are shown in Table 2.3-1.

Two conditions are imposed when the l

bypass is used:

1.

A nuclear overpower trip set point of 55.0 percent of rated l

power is automatically imposed during reactor shutdown.

2.

A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

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Amendment 9,M,49,M,M4,4M 13 l

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Tchle 2.3-1 Reactor Protection System Trio Settina Limits One Reactor Coolant Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop ( )

-- l Operating (Nominal Operating (Nominal (Nominal Operating Shutdowyt Ooeratina Power - 100%)

Operatina Power. 75%)

Power. 49%)

Bypas's Nuclear power, % of 104.9 104.9 104.9 5.0(a) rcted, max Nuclear Power based on Protection System Maximum Protection System Maximum Protection System Maximum ~ Bypassed flow (b) and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for

% of rated, max Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump monitors, t of rated, max (c)

High RC system 2355 2355 2355 1720(a) pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig. min Variable low RC Specified in RCS Specified in RCS Specified in RCS Bypassed system pressure, Pressure-Temperature Pressure-Temperature Pressure-Temperature paig, min Protective Maximum Protective Maximum Protective Maximum Allowable Setpoints Allowable Setpoints Allowable Setpoints figure in COLR figure in COLR figure in COLE RC temp, F, max 618 618 618 618 High reactor building 4(18.7 psia) 4(18.7 psla) 4(18.7 psia) 4(18.7 pressure, psig, max psia)

(a) Automatically set when otner segments of the RPS (as specified) are bypassed.

(b) Reactor coolant system flow, %

(c)The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.

(d) Operation with one Reactor Coolant Pump operating in each loop is limited to 24 hrs. with the reactor critical.

Amendment No. 4,Gt,44,49,5G,64,94,494,444,448 15

_.-~_._~m.

6.12.3 CORE OPERATING LIMITS REPORT l

6.12.3.1 The core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle for the following Specifications:

2.l' Safety Limits, Reactor Core - Axial Power Imbalance i

protective limits and variable Low RCS Pressure-Temperature Protective Limits 2.3.1

~ Reactor Protection System trip setting limits -

Protection System Maximum Allowable Setpoints for Axial Power Imbalance and Variable low RC system pressure l

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3.1.8.3 Minimum Shutdown Margin for Low Power Physics Testing 3.5.2.1 Allowable Shutdown Margin limit during Power Operation 3.5.2.2 Allowable Shutdown Margin limit during Power Operation j

with inoperable control rods l

3.5.2.4 Quadrant power Tilt limit 3.5.2.5 Control Rod and APSR position limits 3.5.2.6 Reactor Power Imbalance-limits 6.12.3.2 The analytical methods used to determine the core operating i

31mits addressed by the individual Technical Specification shall I

l be those previously reviewed and approved by the NRC in Babcock

& Wilcox Topical Report BAW-10179P-A, " Safety Criteria and E

Methodology for Acceptable Cycle Reload Analyses" (the approved i

i revision at the time the reload analyses are performed). The approved revision number shall be identified in the OORE OPERATING l

LIMITS REPORT.

')

6.12.3.3 The core operating limits shall be determined so that all i

applicable limits (e.g. fuel thermal-mechanical limits, core i

thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of i

the safety analysis are met, j

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6.12.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle i

revisions or supplements thereto, shall be provided upon issuance j

for each reload cycle to the NRC. Document Control Desk with j

copies:to the Regional Administrator and Resident Inspector.

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l Amendment No. M, 99, 88, H e, M 9, 142 (next page is 146)

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i MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY) i The pages would be revised as follows:

I Remove Insert iv iv 7

7 8

8 9

9 9a 9b 9c 13 13 14a 14b 15 15 142 142 l

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LIST OF FIGURES Number Title PaQe i

2.I 1

^^"" P"OT20"'!ON CAPOTY LIMITC 0 ;.

2.1 2 O^."," ""OTOO" ION CAT:TT I.!MITC Ob 2.1 0 99RilH-PRO 9999MN-6APIPPY-frIM I T C 0;

2.3 1

""O"00"'I'l: OYO"' M ;OXIM"". AI.LO'- A"L OCT"OINT 1h l

2.2 2

""OT"OTI'."" CTCT M "AXIM":- ALLONAOL" 0"T"^!%TC 10b l

3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION COOLDOWN LIMITATIONS 20c 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H-O 33 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110bc 4.4.2-1 NORMALIZED LIFTOFF FORCE - HOOP TENDONS 85b 4.4.2-2 NORMALIZED LIFTOFF FORCE - DOME TENDONS 85c 4.4.2-3 NORMALIZED LIFTOFF FORCE - VERTICAL TENDONS 85d 4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 110o2 4.18.3.a.3 5.1-1 MAXIMUM AREA BOUNDARY FOR RADIOACTIVE RELEASE CALCULATION (EXCLUSION AP.EA) lila 6.2-1 MANAGEMENT ORG ANIZATION CHART 119 6.2-2 FUNCTIONAL OAGANIZATION FOR PLANT OPERATIONS 120 I

t Amendment No. W,M,99,MG,444, iv

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i 2., $AFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE l

Apolicability l

Applies to reactor thermal power, reactor power imbalance, reactor coolant I

system pressure, coolant temperature, and coolant flow when the reactor is critical.

l Obiective 1

I To maintain the integrity of the fuel cladding.

Specification Themaximumlocalfgelpincenterlinetemperatureshallbe 2.1.1 l

5 5080 - (6.5 x 10~

x (Burnup, MWD /MTU)*F) for TACO 2 applications and $4642 - (5.8 x'10~

x (Burnup, MWD /MTU)*F for TACO 3 applications.

Operation within this limit is ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip Setting Limits," as specified in the 00LR.

2.1.2 The departure from nucleate boiling ratio shall be maintained greater than the limits of 1.3 for the BhW-2 correlation and i

1.18 for the BWC correlation.- Operation within this limit is ensured by compliance with Specification 2.1.3 and with the Axial Power Imbalance protective limite preserved by Table 2.3-1

" Reactor Protection System Trip Setting Limits," as specified in y

the COLR.

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2.1.3 Reactor. Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the safety Variable Low RCS Pressure-Temocrature Protective Limitg shown as specified in Tig;;; 2.1

+ the COLR.

l Bases j

To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is-accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant' temperature. The upper boundary l

of the nucleate boiling regime is termed departure from nucleate boiling l

(DNB). At this point there is a sharp reduction of the heat transfer' j

l coefficient, which could result in high cladding temperatures and the

]

l possibility of cladding failure. Although DNB is not an observable j

i j

parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation. The BAW-2(1) and BWC(2) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux i

distributions. The BAW-2 correlation applies to Mark-B fuel and the BWC l

correlation applies to Mark-BZ-fuel.

The local DNB ratio (DNBR), defined 1

1 as the ratio of the heat flux that would cause DNB at a particular core l

location to the actual heat flux, is' indicative of the margin to DNB.

The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30 (BhW-2) and 1.18 (BWC).

Amendment No. 94,H 4,446 7

___._m.__

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

_ _ ~ _.

A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corrocponds to a 95 percant probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all' operating conditions.

.r The difference between the actual core outlet pressure and the indicated

-reactor coolant system pressure for the allowable RC pump combination has been considered in determining the ;;;; p.;t;; tic; ;;f;ty Variable Low RCS Pressure-Temperature Protective 4 Limits.

The oweve Variable Low RCS Pressure-Temperature Protective Limits presented in Tig;;; 0.1 1 ghg gggg represente the conditions at which the DNBR is greater than or equal to the minimum allowable DNBR for the limiting combination of thermal power and number of operating reactor coolant pumps.

Thi; ;;. 2 3 is based on the deHow&ng nuclear power peaking factors (3) as specified in the COLR with potential fuel densification effectsta a

i 0.00, TjlH=171iIN=101-7

=

The Axial Power Imbalance Protective Limits in the COLR are based on the more restrictive of two thermal limits and include'the effects o? potential fuel densification:

l 1.

TheDNBRlimitproducedby;nu;1;;;p;a;;p;_hingf_;t;;efTk

= 0.02 cr the limitina combination of the radial peak, axial peakt th;t yi:ld; n; 1;;; then th; :::

and position of the axial peaks limit.

2.

The combination of radial and axial peak that prevents central fuel molting at the hot spot as given in.the COLR.

Power peaking is not a directly observable quantity and therefore' limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The flow rates for ;;...; 1, 0, _al 2 ef rig;;; 0.1 0 the variable Low RCS Pressure-Temperature Protective Limits specified in the COLR correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each 1

loop, ;;;;;;;i;;1y.

l The ;;.._ Of Tig;;; 2.1 1 Variable Low RCS Pressure-Temocrature Protective Limit for four reactor coolant cumos operatino is the most restrictive of all possible reactor coolant pump maximum thermal power combinations as specified shown in Tig;;; 2.1-4ggfagLB. The ;;.J;; Of Tig;;; 2.1 0 Variable Low RCS Pressure-Tomoerature Protective Limits in the COLR represent the conditions at which the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.

If the actual' pressure / temperature point is the ;;fety variable Low below and to the right of the pressure / temperature linea RCS Pressure-Temocrature Protective Limit is exceeded. The local quality at the point of minimum DNBR is less than 22 percent (BAW-2)(1) or 26 percent (BWC)(2).

Amendment No. M,W,M,MG,MG 8

l l

Using a local quclity limit of 22 percsnt (BAW-2) or 26 percsnt (BWC) ct the point of minimum DNBR as a basis for ;;rve; 2 ;..d 2 less than four reactor coolant Dumos oDeratinQ of Tig;;; 2.1 3 the Variable Low RCS Pressure-Temperature Protective Limits specified in the COLR is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

The DNBR as calculated by the BAW-2 or the BWC correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.

The maximum thermal power, as a function of reactor coolant pump operation is limited by the power level trip produced by the flux-flow ratio (percent flow x flux-flow ratio), plus the appropriate calibration and instrumentation errors.

For each eweve combination of operatino reactor coolant Dumos of Tigur; 2.1 2_ghg Variable Low RCS Pressure-Temperature Protective Limits soecified in the COLR, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (BAW-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22 percent (BAW-2) or 26 percent (BWC) for that particuler reactor coolant pump ;itu; tion combination. Curv; 1 cf Tigur; 2.1 2 The Variable Low RCS Pressure-Temocrature Protective Limit for four reactor coolant oumos I

operatino is the most restrictive because any pressure-temperature point above I

and to the left of this curve will be above and to the left of the other curves.

l l

REFERENCES l

(1)

Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May, 1976.

(2)

BWC Correlation of Critical Heat Flux, BAW-10143P-A, April, 1985.

1 (3)

FSAR, Section 3.2.3.1.1.c.

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J Amendment No. M,M,M,M,M,M,4H 9

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Amendment No. 44,4+3 9e I

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REACTOR OUTLET TEhPERATURE,

  • F CURVE GPM POJER PUMPS OPERATING (TYPE OF LIMIT) 11*4%

FOUR PUMPS (DNBR LIMIT) 1 374,880 (100%)

2 280,035 (74.7%)

90.8%

THREE PUMPS (QUALITY LIMIT) 3 184,441 (49.2%)

63.7%

ONE PUMP IN EACH LOOP (QUALITY LIMIT)

  • 106.5% OF DESIGN FLOW i

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l Amendment No. M,M,4-1-3 Ge l

pump 2(c).

Th3 pump monitore alco r0ctrict ths powar Icyc1 for the number of pumps in operation.

C.

RCS Pressure i

During a startup accident from low power or a slow rod withdrawal from high power, the system high prinssure trip is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Figur: Table 2.3-1 for high reactor coolant l

system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for ar.y design transient.(2)

The low pressure (1800 psig) and variable low pressure (12.00T,g; 0700; (COLR) trip setpoint shownr in rigure Table 2.3-1 have l been established to maintain the DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.(2,3)

To account for the calibration and instrumentation errors, the accident analysis used the ;;faty protective lOmit of rigur; 2.1 1 specified in the COLR.

D.

Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618F) shown in rigur: Table 2.3-1 has been established to prevent l

excessive core coolant temperatures in the operating range.

Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of 620 F.

E.

Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even f.n the absence of a low reactor coolant system pressure trip.

F.

Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system.

l The reactor protection system segments which can be bypassed are shown in Table 2.3-1.

Two conditions are.Lmposed when the bypass is used:

1 l.

A nuclear overpower trip set point of 55.0 percent of rated i

power is automatically imposed during reactor shutdown.

i l

2.

A high reactor coolant system pressure trip setpoint of l

1720 psig is automatically imposed.

Amendment G,G1,49,M,4G4,4 M 13 l

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I e

2500 l

P=2355 PSIG T=618 'F l

2355

\\

2300 ACCEPTABLE Q-OPERATION N

Dm 2100 c.,

bsJ 1900 8

P=(13.89 Tout-6766)PSIG UNACCEPTABLE OPERATION P=1800 PSIG

< 1700 1500 560 580 600 620 640 660 REACTOR OUTLET TEhPERATURE, F

M ALLOWABLB-6B9PORFP Pigur; ::. 2-+

Amendment No. M,49,H,404,44-3 44e l

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

-.. -... - -.. -... - ~.... -...... -...

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t b

h 1

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f f

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L P

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...............m.....,

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Amendment 6,M,M,43,M,64,M,M, 44b l

M 4,4 M,+94

--,+,-r g

Table 2.3-4 Reactor Protection Syst c Jrio Settino Limits One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Operating (Nominal Operating (Nominal Loop (*II Shutdo,wn l Operatino Power - 100%)

Operatino Power. 75%)

(Nominal Operating Bvoass Power. 49%)

Nuclear power, % of 104.9 104.9 104.9 5.0(a) rated, max Nuclear Power based on Protection System Maximum Protection System Maximum Protection System Maximum Bypassed flow (b) and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for

% of rated, max Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump monitors, % of rated, max (c)

High RC system 2355 2355 2355 1720(a) pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig. min Variable low RC 12.00 Te--6%6N SD*C1fI'd l'. 00 "'e-6%6-(4)- Specified 12.00 T

-M66 N Bypassed system pressure, in RCS P-T Prot. Max. All.

in RCS P-T Prot. Max. All.

Specified in RCS P-T Prot.

psig, min Setooints floure in COLR Setooints floure in COLR Max. All. Setooints fioure in COLR RC temp, F, max 618 618 618 618 High reactor building 4(18.7 psia) 1(18.7 psia) 4(18.7 psia) 4(18.7 pressure, psig, max psia)

(a) Automatically set when other segments of the RPS (as specified) are bypassed.

(b) Reactor coolant system flow, %

(c)The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.

N is in dcs.cc; ich.cnhcit (?).

e (e,d,,) Operation with one Reactor Coolant Pump operating in each loop is limited to 24 hrs with the reactor critical.

Amendment No. E,94,43,49,se,M,94,-104,44-3,146 15

6.12.3 CORE OPERATING LIMITS REPORT 6.12.3.1 The core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle for the following Specifications:

2.1 Safety Limits, Re'wtor Core - Axial Power Imbalance protective limits and variable Low RCS Pressure-Temperature Protective Limits 2.3.1 Reactor Protection System trip setting limits -

Protection System Maximum Allowable Setpoints for Axial Power Imbalance and variable low RC system pressure j

3.1.8.3 Minimum Shutdown Margin for Low Power Physics Testing 3.5.2.1 Allowable Shutdown Margin limit during Power Operation 3.5.2.2 Allowable Shutdown Margin limit during Power Operation with inoperable control rods 3.5.2.4 Quadrant power Tilt limit 3.5.2.5 Control Rod and APSR position limits 3.5.2.6 Reactor Power Imbalance limits 6.12.3.2 The analytical methods used to determine the core operating limits addressed by the individual Technical Specification shall be those previously reviewed and approved by the NRC in Babcock

& Wilcox Topical Report BAW-lOl79P-A, " Safety Criteria and Methodology for Acceptable Cycle Reload Analyses" (the approved revision at the time the reload analyses are performed). The approved revision number shall be identified in the CORE OPERATING LIMITS REPORT.

6.12.3.3 The core operating limits shall be determined so that all appticable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.12.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

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Amendment No. M, 99, GB, 4 %, M9, 142 (next page is 146)

%9,4%

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^*was-m 4

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I ANO-1 CYCLE 13 COLR i

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1 ENTERGY OPERATIONS ARKANSAS NUCLEAR ONE - UNIT ONE i

CYCLE 13 CORE OPERATING LIMITS REPORT l

1.0 CORE OPERATING LIMITS l

This Core Operating Limits Report for ANO-1 Cycle 13 has been prepared m j

accordance with the requirements of Technical Specification 6.12.3. The core operating limits have been developed using the methodology provided in the references.

The following cycle - specific core operating limits are included in this report:

i 1)

Regulating control rod position setpoints, 2)

Reactor power imbalance setpoints, 1

3)

LOCA limited maximum allowable linear heat rate limits, l

l 4)

Axial power imbalance protective limits, l

l 5)

Protection system maximmn allowable setpoints for axial power imbalance, l

6)

KW/ft limit for axial power imbalance protective limits, l

7)

Minimum shutdown margin, i

8)

Axial power shaping rod insertion limits, l

9)

Quadrant power tilt limits, i

10)

Variable Low RCS Pressure-Temperature (P-T) Protective Limits, j

11)

RCS Pressure-Temperature (P-T) Protective Maximum Allowable Setpoints, 12)

Design Nuclear Power Peaking Factors.

94R-1009-03 Revision 2 i

COLR page1of23

4 9

Table Of Contents l

Eilaf l

Fig.1-A Rod Position Setpoints for Four-Pump Operation From 0 to 250 i 10 EFPD

- ANO-1 Cycle 13..

.4 Fig.1-B Rod Position Setpoints for Four-Pump Operation From 250 i 10 EFPD to EOC l

- ANO 1 Cycle 13....

.5 l

Fig. 2-A Rod Position Setpoints for Three-Pump Operation From 0 to 250 i 10 EFPD l

- ANO-1 Cycle 13.

.6 Fig. 2-B Rod Position Setpoints for Three-Pump Operation From 250 i 10 EFPD to EOC

- ANO-1 Cycle 13.

.7 l

Fig. 3-A Rod Position Setpoints for Two-Pump Operation From 0 to 250 i 10 EFPD

- ANO-1 Cycle 13.

.8 l

Fig. 3-B Rod Position Setpoints for Two-Pump Operation From 250 i 10 EFPD to EOC

- ANO-1 Cycle 13.

..9

)

Fig. 4 Operational Power Imbalance Setpoints for Four-Pump Operation From 0 EFPD to EOC - ANO-1 Cycle 13..

. 10 i

Fig. 5 Operational Power Imbalance Setpoints for Three-Pump Operation From 0 EFPD to EOC - ANO-1 Cycle 13..

. 11 Fig. 6 Operational Power Imbalance Setpoints for Two-Pump Operation From 0 EFPD to EOC - ANO-1 Cycle 13...

.12 Fig. 7-A Mk-B8 LOCA Linear Heat Rate Limits..

.13 Fig. 7-B Mk-B8ZL and Mk-B9ZL LOCA Linear Heat Rate Limits..

.14 Fig. 8 Axial Power Imbalance Protective Limits..

.15 j

Fig. 9 Protection System Maximum Allowable Setpoints for Axial Power Imbalance.. 16 Fig.10 Variable Low RCS Pressure-Temperature (P-T) Protective Limits.

.. 17 Fig.I1 RCS Pressure-Temperature (P-T) Protective Maxunum Allowable Setpoints...

18 KW/FT Limit for Axial Power Inibalance Protective Limits..

19 Minimum Shutdown Margin..

. 20 Axial Power Shaping Rods (APSR) Insertion Limits..

.. 21 ANO-1 Cycle 13 Tilt Setpoints (Quadrant Power Tilt Limits)..

. 22

[

Design Nuclear Power Peaking Factors..

. 23

)

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94R 1009-03 Revision 2 COLR page 3 of 23

Figurs is refstred to by Technical Specification 2.1.3 1

i Figure 10 Variable Low RCS Pressure-Temperature (P-T) Protective Limits 2400 1 PUMP EACH LOOP 2200 3

'N 4-PUMP 7 a

io u) in y 2000 3-PUMP l a.

6 1

(23ABDO 1800 O

l 1600 580 600 620 640 660 1

REACTOR OUTLET TEMPERATURE,

  • F PUMPS OPERATING (TYPE OF LIMIT)

GPM POWER l

FOUR PUMPS (DNBR LIMIT) 374,880 (100%)

112%

THREE PUMPS (QUALITY LIMIT) 280,035 (74.7%)

90.8%

ONE PUMP IN EACH LOOP (QUALITY LIMIT) 184,441 (49.2%)

63.7%

  • 106.5% OF DESIGN FLOW i

94R 1009-03 Revision 2 COLR page 17 of 23 l

~

i Figuro 10 rGfGrrcd to by Technical specification Table 2.3-1 1

Figure 11 RCS Pressure-Temperature Protective Maximum Allowable Setpoints 2500 P=2355 PSIG T=618 'F 2355 2300 ACCEPTABLE Ca OPERATION k

Dv2 2100 J 1900 8

g P=(13.89 Tout-6766)PSIG UNACCEPTABLE OPER.TTION j

P=1800 PSIG

< 1700 1500 560 580 600 620 640 660 REACTOR OUFLET TEhPERATURE, F

4 94R-1009-03 Resision 2 COLR page 18 of 23

o Limit is referred to by Technical Specificaton 2.1 Bases Design Nuclear Power Peaking Factors Maximum Radial (F%)

1.71 Maximum Axial (fz) 1.65 Maximum Total (f )

2.83 q

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94R 1009-03 Revision 2 COLR page 23 of 23

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