Text

Proposed Tech Specs Re Reactivity Control Sys
	ML20196F043
Person / Time
Site:	Wolf Creek
Issue date:	06/24/1988
From:	; WOLF CREEK NUCLEAR OPERATING CORP.
To:	;
Shared Package
ML20196F040	List: ML20196F036; ML20196F043;
References
	NUDOCS 8807050029
	Download: ML20196F043 (27)
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1 ATTACHMENT A SECTION III - MARKED-UP TECHNICAL SPECIFICATION PAGES l

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Attachment A Page 13 of'20 i REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUIDOWN N j._l,MITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 2968 gallons,

2) Between 7000 and 7700 ppm of boron,.and ,

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water'v lume of 55,416 gallons, i

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2) A minimum boron concentration of 4000 ppm, and

3) A minimum solution temperature of 37'F.

( APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE '

ALTERATIONS or positive reactivity changes. '

SURVEILLANCE REOUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and 3 3)

Verifying the Boric Acid Storage System solution temperature when it is the source of borated water, i

b. ,

At least once.per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the .'WST temperature when it is thethan less source 37 F.of borated water and the outside air tee.rerature i . .

WOLF CREEK - UNIT 1 3/4 1-11 1__ 3;- _. . _ _ . - - - - - -- --- -- ~ ~ ~ ~ ~ ' ~ " ~ ~ ' ~ ~

. Attachment A Page 14 of 20 REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - OPERATING N LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2, and 3 and one of the following borated water sources shall be OPERABLE as required by Specifica-tion 3.1.2.1 for MCSE 4:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 17,658 gallons,

2) Between 7000 and 7700 ppm of boron, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 394,000 gallons Aioo GLFoo

, 2) Between -eMO and 41M ppm of boron,

3) A minimua solution temperature of 37 F, and

4) A maximum solution temperature of 100 F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With the Boric Acio Storage System inoperaole and being used as one of the above requireo borated water sources in MODE 1, 2 or 3, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at. ;

least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTCOWNl MARGIN equivalent to at least 1%.ok/k at 200 F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With the RWST inoperable in MODE 1, 2, or.3, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY witnin l the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . I

c. t With no borated water source OPERABLE in MODE 4, restore one boratec water source within to OPERABLE the following status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD ShUTOC'nN 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

WOLF CREEK - UNIT 1 3/4 1-12

Attachment A Page 15 uf 20 REACTIVITY CONTROL SYSTEMS /

BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value equivalent to the most positive moderator density coef ficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core conditicn of all rods inserted (most positive MDC) to an all rocs withcrawn conditi 1 and, a conversion for the rate of change of moderator density with '

tempera ?e at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 ak/k/ F. The MTC value of

-0.2 x 10 4 Skik/ F represents a conservative value (with corrections for burnuo and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of

-4.1 x 10 4 ak/k/ F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the enc' of the fuel cycle are adequate to confirm that the MTC remain within its limits since this coefficient changes slowly due principally to the recuction in RCS boron concentration associated with fuel burnuo.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICAL'.TY This specification ensures that the reactor will not be mace critical with the Reactor Coolant System average temperature Icss than 551 F. This liraitation is required to ensure: (1) the moderator temperature coefficient is within it analy:ed temperature range, (2) the trip instrumentation is within its normal operating :*ange, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NOT 3/4.1.2 BORATION SYSTEMS The Boration Svstems ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function incluce: (1) torated water sources, (2) centrifugal charging ou os ,

(3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergenc'. .

power supply from OPERABLE diesel generators.

With the RCS average temperature equal to or greater than 350 F a minimum of t-o boron injection flow paths are required to ensure single functional caoscility in the event an assumeo failure renders one of the ficw oaths inoper-able. The buration capacility of either flow path is sufficient to oro. ice a SHUTCOWN MARGIN . om expected operating conditions of 1.3*.' 2k/k af.ee xenon cecay and cooldown to 200 F. T,1e maximum expected boration capabii'ty eceire-men , occurs at EOL i rom full power equilibrium xenon concitior s anc rece:res 17.552 gallons of 7000 p;m boratec water fro.t the boric acic storace .anss o-

, 7. ga'lons of ppm corated water frcm the RWST. A'.n e ::3 3.e 3;e temoerature less tnan 150 F. cnly one tcron injection flu., path is ecui-ec.

'w0L: CREEK - UNIT 1 0.900 5 3/4 1-2 S3,tEH j

Attachment A Page 16 of 20 s !

REACT!v!TY CONTCL SYSTEMS BASES BORti!ON SYSTEMS (Continued)

With the RCS temperature below 200 F, one Boration System is acceptable

-itacut single failure consideretion on the basis of the stable reactivity concition of the reactor and the additional restrictions prohib,iting CORE ALTERATIONS and positive rea:tivity changes in the event the single Boron Injection System becomes inocerable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps ex:ept the required OPERABLE pump to be inoperable in MODES 4 5, and 6 provide; assurance that a mass addition pr' essure transient can be relieved by the opera-tion of a single DORV or an RHR suction relief valve.

The boron capability reouired below 200*F is sufficient to provice a SHUTDOWN MARGIN of 1% Sk/k af ter xenon decay and cooldown from 200'F to 140 F. This condition requires either 2968 allons of 7000 o m, orated water from the boric acid storage tanks or t4 696 7 gallons of 4444- ppm boratec water from the RWST. 28 0 g q,o g i The contained water volume limits include allowance for water not av ilable because of discharge line location and other physical characteristics. INS?p[r j$)

The limits on contained water volume and boron concentration of the RWST-also ensure a pH value of between 8.5 and 11.0 for the solution recirculated witnin containment after a LOCA. This pH band minimize. the evolution of iodine andsystems metnanical minimizes the effect of chloride and caustic stetss corrosion on and components.

The OPERABILITY of one Boration System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES i

The specifications of this section ensure that: (1) acceptable power cistribution limits are maintained. (2) the minimum SHUTOOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated acci-cent analyses are limited. OPERABILITY of the control rod position indicctori is required to determine control rod positions and thereby ensure complience with the control rod alignment and insertion limits. ,

Verification that the {

Digital Rod Position Indicator agrees with the demanded pusition within : 12 '

steos at 24, 48, 120, and 22S steps withdrawn for the Control Bants anc 'S, 210 and 228 steos withdrawn for the Shutdown Banks provices assurances t at Lne Digital Roc Position Indicator is operating correctly over the full ran;e of incit "tion. Since the Digital Roc Position System coes not indicate the actual snutc wn roc cosition between 18 steps and 210 steps, only points in tne incicatec ranges are pickec for verification of agreement ith 1esandec position.

WOL~ OREEK - UNIT 1 2 3/4 1-3

. Attachment A Page 17 of 20 Insert A N

in the case of the boric acid tanks, all of the contained volume is considered usable. The required usable volume may be contained in either or both of the boric acid tanks.

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Attachment A Page 18 of 20 g

3/4.5 EMERGENCY CORE COOLING SYSTEMS .

3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a. The isolation valve open and power removed,

b. A contained borated water volume of between 6122 and 6594 gallons,

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c. A boron concentration of.betweenHF@@@ and ppm, and

d. A nitrogen cover pressure of between 585 and 665 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed L isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be 17 at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open.

Pressurizer pressure above 1000 psig.

WOLF CREEK - UNIT 1 3/4 5-1 Amendment No. Il

Attachment A~

Page 19 of 20 EMERGENCY CORE COOLING SYSTEMS ('

3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERASLE witn:

a.

A minimum contained borated water volume of 394,000 gallons,

b. 14o0 Asco A boron concentration of between 0064L and f4ffr ppm of toron,

c. A minimum solution temperature of 37 F. and d'

A maximum solution temperature of 100 F.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 nour within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .or be in at least HOT STANDBY C'.s within N the( next SURVEILLANCE REOUIREMENTS l

4.5.5 l l

The RWST shall be demonstrated OPERABLE: !

i a.

At least once per 7 days by:

1)

Verifying the containea borated water volume in the tank. ana 2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST terceratu e anen the outsice than 100 F. air temoerature is either less tran 37*: ar greate- .

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II WOLF CREE.< - UN!T 1 3/4 5-10

, Attachment A Page 20 of 20 EMERGENCY CORE COOLING SYSTEMS S BASES

_ECCS SUBSYSTEMS (Continued)

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety injac-tion pumos exceot the required OPERABLE charging pump to be inoperable in MODE and 5 and in MODE 6 with the reactor vessel head on provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV cr RHR suction relief valve.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures tnat at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

for thruttle valve position stops and flow balance testing provide assuranceSurveillanc that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper point injection flow resistance is necessary andto: pressure drop in the piping system to each (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptacle level in of total ECCS the ECCS-LOCA analyses. flow to all injection points ecual to or above that assumed of ECCSLOCA.

intersystem check valves ensures that a failure of one valve will not cause and accessible, The Surveillance Requirements to vent the ECCS pump casingt cose, discharge pipingcan i.e., be reached ensures a without personnel hazard or nign raciation or water nammer in ECCS piping. gainst inoperable pumps caused by gas binding 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of ne by the ECCS in tne event of a LOCA.ECCS ensures that a sufficient supp concentration ensure that: The limits on RWST minimum volume anc coron permit recirculation cooling flow to the core, and (2) the reactor will rem ontrol, emet volumes

ntF rodsall in: rt:c :ncept f;rsubcriticalthyinm::t the~2:ct colde condition i

cert following These ' motions a cc O'-

su m w 4 it our o Tgcc je.

't. the LOCA analyses.

The containec water volume limit includes an allowance for water not usable because of tank discharge line location or other pnysical characteristics .

also ensure a pH value of between 8.5 and 11.0 for the solu within containment after a LOCA. This pH band minimizes tne evolution of iodine anc mechanical minimizes systems anc components. the effect of chloride anc caustic stress corrosio t 1

l WOLF CREE < - uu : '. E 3/a 5-2 l

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ATTACHMENT B F Wr Mu PHer Revision / Reactor Coolant AH 8r System Temperature Measurement Ibcertainties/Special Test Exception 1

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l Attachment B to WM 88-Draft i Page 1 of 16 l June 24, 1988 l SECTION I - SAFETY EVALUATION ,

INTRODUCTION AND SUPMARY Historically, increasing the allowable F" with decreasing power has been permitted for all previously approved WesNnghousedesigns. The increase is ,

permitted by the DNB protection setpoints and allows for radial power ;

distribution changes with rod insertion to the insertion limit. The current Wolf Creek Generating Station, Unit 1 Technica1 Specifications N

(Section 3/4.2.3) require a 0.2 part power multiplier on'F AH. The NRC has in recent years approved a 0.3 multiplier for a number of operating plants such as Turkey Point, R. E. Ginna, -Trojan, Cook Unit 1, Zion, Indian Point Unit 3, Point Beach and Surry. Described in this submittal is a changetotheWogf Creek . Unit 1 Technical Spacifications reflecting a design change in the Fgg -

part power multiplier.

F" s 1.33 n.0 0.3(1-,n where P = fraction of rated thermal power The only change from the current F" Technical Specification .is the mgitiplier on the quantity (1-P) from d $ to 0.3. No change was made in the F limit at full power.

AH This change is requested for Wolf Creek Generating Station to gliow optimization of the core loading pattern by minimizing restrictions on F3g at low power. This change will also minimize the probability of making rod insertion limit changes in future reload cycles to satisfy peaking factor criteria at lower power.

The revision to the Technical Specification parameters involving instrument allowances for the Reactor Coolant System (RCS) Resistance Temperature Detectors (RTDs) are associated with the reactor trip setpoints for overtemperature delta - T (OTAT) and overpowerA T (OPAT), and reactor coolant flow setpoints and RCS flow measurement uncertainty.. These changes result from improvements in the calculation methodology through the use of a plant specific calibration technique.

The proposed Technical Specification change involving Special Test Exception Specification 3.10.4isintendedtocorrectaninconsistencyintge i Technical Specifications identified during the review of the proposed F AH '

modification.

A. F AH

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The proposed Wolf Creek Generating Station Technicgl Specification change which impacts DNBR evaluation is the valus of F * " "

AH following equation:

F I

I+ ~

l H

where: K (part power multiplier) has increased from 0.2 to 0.3

Attachment B to WM 88-Draft Page 2 of 16 June 24, 1988 FAH " me88ured radial Peaking factor with appropriate uncertainties P = fractional core power level at less than 1002 Rated Power or,

= 1.0 at greater than or equal to 100Z Rated Power The increase in the part power multiplier (K) from 0.2 to 0.3 has a direct impact on DNBR calculations. The core limits for Wolf Creek Unit 1 (reflected in Technical Specifications Figure 2.1.1) represent restrictions of average enthalpy at the vessel exit and minimum DNBR.

The average enthalpy at the vessel exit must be less than the enthalpy of saturated liquid to assure the proportionality between vessel T and core-power. The exit enthalpy restriction is more limiting than DNBR at lower power, and vessel exit limit lines are not impacted by the radial peaking factor as shown in the following relation:

h -hin + Q/G < h, out where:

h = average coolant enthalpy at vessel exit (BTU /lb,)

out h = vessel inlet lant enthalpy (BTU /lb ,)

in Q = total core power (BTU /hr)

G = total core coolant flow (1b,lhr)

Therefore, a change in radial peaking factor will not impact core limits at power levels restricted by vessel exit boiling limits.

At power levels greater or equal t 100Z g is not impacted by K (i.e.,

N 1-P=0) and the peak FAH used t RatedPowerF"khecorelimitsat generate these power levels is unchanged. Therefore, the core limits restricted ;

by DNBR at these power levels will not change. The core limits at power !

levels less than 100% rated power which are not restricted by vessel exit boiling limits will be impacted by the change in the part power multiplier.

2. Nuclear Desian Evaluation The proposed technical specification changes do not impact the other nuclear design bases used to evaluate the reload cores. The standard calculation methods described in WCAP-9272P-A, ' Westinghouse Reload Safety Evaluation Methodology," continue to apply. The 0.3 multiplier in conjunction with the Technical Specifications control rod insertion limits insures that peaking factor limits are not exceeded during anticipated power control maneuvers.

Attachment B to WM 88-Draft Page 3 of 16 June 24, 1988

3. Accident Evaluation The methods used for accident evaluations are described in WCAP-9272-P-A. These are the same as those applied to the Wolf Creek Generating Station Cycle 1 analysis. l To ensure adequate core protection, thg Reactor Core Safety Limits were reevaluated due to the increased F im . aey na a ns AH conclude that the reactor core safety limits are not changgd and the FSAR non-LOCA safety analysis remain applicable for the increased F ultiplier.

AH No reanalysis is requireo for non-LOCA accident evegts, since non-LOCA accident analysis are not impacted by the proposed F ultiplier change.

AH The current large and small break LOCA analysis of record remain N

applicable for the F increase at partial powers as long as the product of N

power times F is gYess than full power value. The LOCA analyses are performed ag M Z of rated power and use the Technical Specification upper limit for F3y. Thus Nthe LOCA analysis bounds all partial power conditions when the product of F times power decreases from the full power value with AH decreasing power. Since the Technical Specificatlog change for the partial power multiplieg does not increase the full power F and the product of the partialN p wer F timespowerislessthanthefullhHower limit, the partial AH multiP er increase has no impact on LOCA analyses.

power F

4. Conclusion Based on the above evaluation of the proposed F part power multiplier change, itisconcludedthatthechangedoesnotinvo$veanunreviewedsafety question.

B. RCS RTD SENSOR MEASUREMENT UNCERTAINTY-RELATED CHANGES

1. Evaluation The use of plant specific calculation inputs to establish RTD cglibration and drift allowances, as well as the previously discussed F part power multiplier, result in proposed revisions to the followkHng Technical Specifications:

a. OTAT reactor trip (Specification Table 2.2-1)

b. OPAT reactor trip (Specification Table 2.2-1)

c. RCS flow - low flow reactor trip (Specification Table 2.2-1)

d. The calculated RCS flow uncertainty (Specification 3.2.3 and Bases)

These changes are based on improvements in calculation techniques and the use of plant specific inputs. The basic setpoint methodology previously established by the NSSS vendor remains unchanged.

4 Attachment B to WM 88-Draft Page 4 of 16 June 24, 1988 The OT6T setpoint on Table 2.2-1 was revised because the parameter 'Z' is affected by the revised part power multiplier: and the 'S' parameter is changed to eliminate the two-entry format and to reflect a more accurate calculation of the parameter value. The changes to Note 1 (iii) and Note 2 of Table 2.2-1 are based on the part power multiplier revision previously evaluated. The OP4T change involves Note 4 of Table 2.2-1 and results from employing more accurate analysis through use of computer generated output.

For the RCS flow-low flow reactor trip the total allowance (TA), sensor error

'Z', trip setpoint, and allowable value are changed as a result of the change in the flow measurement uncertainty. The flow measurement uncertainty changes result from improved inputs to the calculation methodology. In the past, the NSSS vendor has performed these calculations using generic assumptions. The calibration procedure used at Wolf Creek is different than that assumed by the NSSS vendor. Tt ts difference in methodology allowed an evaluation of the setr oint to obtain more margin. A computer program was also used to improve the numerics of the calculations. The sensor error 'S' of 0.6 for the sensor drift reflects the value currently documented in the current, NSSS developed Wolf Creek setpoint methodology report. No other changes are made to this setpoint. The flow measurement uncertainty change also is reflected in a proposed revision to Figure 3.2-3 and in the Technical Specification Bases.

The RCS loop design flow value remains unchanged.

The proposed changes to the Technical Specifications involve reactor trip setpoints and calculated RCS flow uncertainty; therefore, the changes would not increase the probability of occurrence of accidents previously evaluated.

The consequences of previously evaluated accidents were investigated through an evaluation of the impact of the proposed changes on the applicable Wolf Creek safety analysis. Based on this evaluation, the safety analysis demonstrate continued compliance with required safety limits.

The possibility of an accident of a different type than previously evaluated will not be increased. The proposed changes simply affect the threshold values of the "as measured' sensor and rack errors which define reportability, the point at which the reactor will trip given a previously evaluated design base accident, and the measured radial peaking factor at power levels below 100%. All of these proposed changes protect the reactor core from exceeding safety limits given an accident or operational transients which has previously been evaluated.

The probability of a malfunction of equipment important to safety as previously evaluated will not be increased. The proposed changes will not increase the frequency at which equipment important to safety is actuated to protect the reactor core nor do the proposed changes alter the characteristic manner (rod drop time, valve movement, etc.) in which equipment responds once required to actuate. Therefore, probability of malfunctions is not increased.

The consequences of a malfunction vf equipment important to safety previously evaluated have not been increased. The impact of the proposed changes on the safety analyses has been evaluated. The results of the evaluation indicate that safety limits continue to be met.

The possibility of equipment malfunction of a type other than previously evaluated is not created. The proposed changes affect setpoints in euch a

Attachment B to WM 88-Draft Page 5 of 16 June 24, 1988 manner as to insure that safety limits are met. The changes are not such as to cause equipment malfunctions of any kind.

The proposed changes impact the analysis assumptions and inputs used in the safety analyses and the peaking factor limits employed in the core design.

The affected safety analyses and the peaking factor limits have been evaluated and or renalyzed, and it has been determined that all applicable safety criteria are met. Therefore, the margin of safety as defined in the bases has not been reduced.

2. Conclusion Based on the above evaluation of RCS RTD sensor measurement uncertainty changes on Technical Specifications, it is concluded that the proposed changes do not involve an unreviewed safety question.

C. Special Test Exception Change

1. Evaluation Technical Specification 3.4.1.1 requires a verification that all RCS loops are in operation and circulating coolant. The applicability statement for Specification 3.4.1.1 invokes Special Test Exception 3.10.4 which allows suspending the limitations of Specification 3.4.1.1 during STARTUP and physics testing under certain conditions.

Specification 3.2.3 also establishes requirements for assuring RCS flow.

However, Specification 3.2.3 does not invoke the Special Test Exception of 3.10.4. This is apparently an oversight in the existing Technical Specifications, and proposed changes to eliminate the inconsistency have been prepared. The proposed changes would add a footnote to Specification 3.2.3 invoking the Special Test Exception and would modify Specification 3.10.4 to specifically refer to Specification 3.2.3.

Because the proposed changes are administrative in nature, no safety concern would be created by implementing them.

2. Conclusion 1

The proposed Technical Specification modifications involving implementing l Special Test Exception 3.10.4 are administrative in nature, eliminate an l inconsistency in the Technical Specifications and would not affect the safety I design bases of the plant.

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o 34 Attachment B to WM 88-Draft Page 6 of 16 June 24, 1988 1

4 SECTION II - ADDRESSING THE STANDARDS IN 10 CFR 50.92 This amendment request revises Wolf Crgek Generating Station, Unit No. 1,-

Technical Specifications to revise the FAH Part power multiplier from 0.2 to 0.3, to modify the OT4LT, OPAT, RCS flow-low flow trip, and RCS flow requirements based on RCS RTD uncertainty changes, and to eliminate an inconsistency in invoking a Special Test Exception.

The NRC has given guidance concerning the applicability of 10 CFR 50.92 by providing examples (50 FR 7751) for comparison. The proposed Technical Specification modifications to eliminate an inconsistent reference to Special Test Exception 3.10.4 are purely administrative and are encompassed by example (i) of actions not likely to involve a significant hazards consideration. The other proposed changes are discussed in more detail in the following paragraphs.

Standard 1 - Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes affect reactor core design parameters associated with accident mitigation or operational transients. These parameters are determined by design to protect the reactor core from exceeding safety limits. These changes do not affect-initiators of an event that would change the probability of occurrence of an accident previously evaluated. The consequences of previously evaluated accidents have not been increased. The impact of these changes on the safety analyses has been evaluated. The evaluation results indicate that the safety analyses continue to meet required safety limits.

Standard 2 - Create the Possibility of a New or Different Kind of Accident from Any Accident Previously Evaluated The proposed changes affect reactor core design psrameters associated with mitigation of postulated accidents and transients. No design changes to plant structures, systems or equipment or modifications to plant operation are being proposed. There are no new failure modes or mechanisms associated with the proposed revisions. The changes reflect revised parameters based on evaluations of existing analyses and or reanalysis, as necessary. The proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Standard 3 - Involve Significant Reduction in a Margin of Safety The proposed changes do not involve a significant reduction in a margin of safety. The changes do not affect any Technical Specification margin of safety. The changes proposed impact the analyses assumptions and inputs used in the safety analyses, the core power limits used in the reactor design, and reactor trip setpoints. The affected safety caalyses and core power limits have been evaluated and/or reanalyzed as necessary and it has been determined

_ .-_. _ _ __ _ . _ _ _ _ _ - _ _- ___ ~ _ - _ _ _ _ _ _

{

Attachment B to WM 88-Draft Page 7 of 16 June 24, 1988 that all applicable safety criteria are met. Therefore, the margin of safety as defined in the bases has not been reduced. Affected Technical Specifications have been revised to reflect the proposed changes.

Based on the above discussions and those presented its Attachment I , it has been determined that the requested Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations; or create the possibility of a new or different kind of accident 'over previous evaluations; or involve a significant reduction in a margin of safety.

Therefore, the requested changes do not involve a significant hazards consideration.

l

9 S

6 ATTACHMENT B SECTION III - MARKED-UP TECHNICAL SPECIFICATION PAGES i

i I

l

Attachment 8 Page 8 of 16 l

POWER DISTRIBUTION LIMITS l 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR g LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be. maintained within the region of allowable operation shown on Figure 3.2L3 for four loop operation.

Where:

N F

a' AH ,

R = 1.49 f.1.0 + (1.0 - P)]

b* P = THERMAL POWER , and RATED THERMAL POWER

c. F g = Measured values of F g obtained by using the movaole incore detectors to-obtain a power distribution mao. The measurea .

valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes measurement uncertainties of . ' for flow and 4P.

for incore measurement of F g. M ,

APPLICABILITY: MODE 1.

ACTION: '

With the combination of RCS total flow rate and R outside the region of acceotable operation shown on Figure 3.2-3:

I a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the combination of RCS total flow rate and R to within the above limits, or

2. Reduce THERMAL POWER to less than 50*. of RATED THERMAL POWER and reduce the Power Range Neutron Flux - Hign Trip Setooint to less than or eoual to 55*. of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through incore flux maoping and RCS total flow rate comoarison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5'.' of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and i

l

Sea Spec.w Tc.sr &creries SPt.c.iFicAmoN 3.lo.4

'a0LF CREEX - UNIT 1 3/4 2-S

_ , _ _ __ ... ___. . ,-~ _ . _ , _ . _ . . . _ . , _ _ . _ , . . . _ , . - . _ - . _ . _ . . _ . _ . _ . . _ _ _ . _

Attachment B Page 9 of 16

'\

2.5%

MEASUREMENT UNCERTAINTIES O 4-e% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF FbH N ARE INCLUDED IN THIS FIGURE 48 2.mEii _j. :..L ;q.2. . . .

..s.. . _ i a a L.g.:.g J..a j ; ,. .. ; . . j . _

.4551]9.:;}".? : silt:l . i:i: ~ ._

ilMii '".E.!:i!.ii.4.

,:x;.

. : F : .: -.: :f.:.

a.:l: :.l. t- u I: pl. .- .:::gn:u(: ;:

. g.: II U l a.

..._._~:._..

1

.._.... ...;. ..u..__.__...-..-.;.m..._.

.=::=_ = _ u=- -

_.= -- p= u -_....

.=

- =.,:.3-

_ :c=1==:_L z ____ ; _

._ .: u - .:._ "u_:._; __

=-

ACCEPTABLE g:#"#" n:_v : == UNACCEPTABLE =-

g - --s: _

'* ~#

OPERATION C-0 5_:5 OPERATION REGION E l5IfiO 155 REGION 5 44 . __ --

.__ __ n. . :._in. =. =. . . . t. . =. . . =.:.-n. v .n...n._==.=_....._,._._._..._.=....--h::---

O w

s-dig-N:-di'iih!"i:~" =i5 i:Ni"i'EElii:hiiihf'i:H'i"'i-"-

- :ii_:iH. ._H.E.i:.i.:h!.=__d.d g .. . . .. . _. n.j . :i..i..:.-

. i. l <al.i. .E._!i.ij i.i ._:iih:jd. .n..i_n.i.i j .". . :: :n. : j . -

g .n .a.: u = . u . . c { a : . . . . . j : . :!O.

2" ..;' *. ...:. 'j . . . . l n.r.- u=;;: ::: . . : . . : n a u: ..:.[a:. .

D C '-" -

3 g .". ...T_ .p_.T _. . .i=_ _ij. . -j :. .:.jii. .a j +.. n d

.. . H.i.n.'.fji..;n.i.:.i..H...:.

.. . . _f.i_=.l n.- . :l -

(' A$-25ii55iE~~iidiji:ff1 'idGiiiT ~i:1.=:UnEf:Ii:In"2i-~Eiii-"-"

J

-""-it_:i:E_ _:4_.:i_li. i:j.::.l.: 4.: . . _:i!=_i.:j _- .. ." . __a_ . . - _"- . in n ..j .

4 .. .

$ g i -

'#ii"55!!iM1 ! 5iiin Tif i'li" illk}ssidiijdi:: ii.hii.ii@.I i

.E

-H -

eli. ili;401F).4 .d..if.ii}i.i...i 0l :

get.\ ,.-,if fi.,E l ~ ",t_h ,

v .

,(1. =;=.: ...:. ..a.

Aw / -- -- - - - -v - "

"r- ' -

a 'T;%i[:3iRM:ld.i- =.T-i- -iihi-ji;-lii -j-i:- f: -

,~ ~- - t: n. = u- ni:.

<.. ::. : :n r.:.:: =f: :'- -.:.

38 - - - - - - ' - - ' - - - - ' ' ' '-'------"-------*---1.:

E

--- E.i-_-l.'.;iii:1-drd.id.:1 .

h.

.:;. .l.i_-;i

i_.h...Fji..: :.:ii

. .:_h.i.jii i .idi._:i.. ..n._i.:.i.. l".

. - r.= :n. n..:n.: :.r .r.a.. ::a..;:l .

J uu:: nn :

=:= r= =r_:2:a::f : n: n nu-l.a.a..a . _ ".nr_ . . . =::: . - . :}n:Il=_ :'rt.=l:

.. -_3

=n =j._= _:,.1:n;s =4 =r:n . I_=:_:=_-i:. . .,=_

.ln, n -G =l= = "-

36 E. _. _ ".__:= _--i=hpgM)-fyi;=;jisi:M.MG.i(Mj=EjijHid-l-"i-

.....u..............

== =31.: -- -. =-. ..=._.

.;-n-

. ._ _._ ---t e n .: ....::--

__ . . .=_.d

. . . . _ . . n.

_ ..=.n . .. . =... _. 2- -n = : n . = n :. A..: i

==_:=.3...

2 :- . r.= _.:j=:::=n.ia

= n n::aIa nc. .-: m:= n =n- ja:: =urnl: na n.: - -j=:: - - = n p=. alm.j =..:- .:

--===== _

.=_-1. 1 :.-l=d,::n

- - n ::::.d:===:=r:

-: +2:=- ..un= : m

-- - _=. =_ =. :.1.:r....i:

._....._........=:_ = : :=u

=. =. . . In

: n - :. . . . . . . - . . - .

34 . . . _ _e. 22. :. =. i. n y n. r. -. - :. . _a+t-. _. .i n.

..=. . . . l. . .=.

_ . . ..o...

_ . . . .__ l.... . . .

..... 1: --

0.90 0.95 1.00 1.05 1.10 R = Fj/1.49 [1.0 1.0.P)]

FIGURE 3.2 3 RCS TOTAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION WOLF CREEK . UNIT 1 3/429

~

Attachment B Page 10 of 16 2.1 SAFETY LIMITS k BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset.of departure frem nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation (R-GRID). The W-3 DNS correlation (R-GRID) has been developed to predict the ONB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNS.

The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature fc which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an 'enthalpy hot channel factor, F g, of 1.55 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included f& an increase in FN dH at reduced power based on the expression:

n o. 3 FAH = 1.55 [1+ + + (1-P)]

Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than thcese calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (al) function of the Overtemperature trip. Wnen the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature with core Safety aT trips Limits. will recuce the Setpoints to provide protection consistent 4

VOLF CREEK - UNIT 1 8 2-1

Attachment B Page 11 of 16 >

s .

j D0 ER O!STRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY DISE HM CHANNEL FACTOR (Continued)

The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor,q F (z), remains within its limit. The F E xy limit for RATED THERMAL POWER (F ) as provideo in the Radial Peaking Factor Limit Report per Specification 6.9.1.9 sas determined from expected power control manuevers over the full range of burnup conditions in the core.

When RCS flow rate and F H are measured, no additional allowances are necessary C to comoarison with the limits of Figure 3.2-3.

aY Measurement errors of +4r for RCS total flow rate and 4% for FN g have been allowed for in cetermination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate f.

indicators. Potential fouling of the feedwater venture which might not be cetected could bias the result from the precision heat balance in a non-conservative manner. Therefore, an inspection is performed of the feedwater venture each refueling outage.

7

The 12-hour periodic surveillance of indicated RCS flow is sufficient to cetect only flow degradation wnich could lead to operation outside the acceptacle region of operation shown on Figure 3.2-3. This surveillance also provides acequate monitoring to detect any core crud buildup.

3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Racial power distribution measurements are made during STARTUP testing and periodically during power cperation.

The limit of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x y plane power tilts. A limit of tne1.02 was selected indicated to provide an allowance for the uncertainty associated with power tilt.

The 2-hour time allowance for operation with a tilt condition greater inan 1.02 but less than 1.09 is provided to allow identification and corree-tion of a cropped or misaligned control rod. In the event sucn ACTION coes rot correct the tilt, the margin for uncertainty on F is reinstatec oy recucing i

tne maxiev. allo-ec power by 3% for each cercent of tilt in exces s of 1.

'I a M CREEi. Ut:7 ; E 3/4 2-5 .

i

} =

TABLE 2.2-1

/

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g SENSOR o", R g"

m TOTAL x FUNCTIONAL UNIT ERROR _g ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE Ng

1. Manual Reactor Trip N.A. oo c N.A. N.A. N.A. N.A. '"

2. Power Range, Neutron Flux $"

i *> a. High Setpoint 7. 5 4.56 0 $109% of RTP* $112.3% of RTP*

b. Low Setpoint 8.3 4.56 0 $25% of RTP* 128.3% of RTP*

3. Power Range, Neutron Flux, 2.4 0.5 0 $4% of RTP* with High Positive Rate 16.3% of RTP* with i

a time constant a time constant 22 seconds 12 seconds

4. Power Range, Neutron Flux, 2.4 0.5 0 High Negative Rate 14% of RTP* with <6.3% of RTP* with

, a time constant a time constant 12 seconds 32 seconds 5- Intermediate Range, 17.0 8.41 0 525% of RTP* $35.3% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 1105 cps 11.6 x 105 cps

7. Overtemperature AT 3.4o a.m 7.2 N- -1. 56- See Note 1 See Note 2

^ 3. >

8. Overp' -er AT 5.5 1.43 0.15 See Note 3 See Note 4

9. Pressurizer Pressure-Low 3.7 0.71 2.49 11875 psig 11066 psig

10. Pressurizer Pressure-High 7. 5 0.71 2.49 $2385 psig $2400 psig

?+ 11. Pressurizer Water Level-High 8.0 2.18 1.96 592% of instrument z $93.9% of instrument span span

' " *RTP = RATED THERMAL POWER

- Loop design flow = 95,700 gpn.

m o

n -

e TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP _SETPOINTS g EN

"' SENSOR E TUNCTIONAL UNIT TOTAL ERROR %k -7 ALLOWANCE (TA) Z (S)

[

z

12. Reactor Coolant Flow-Low 1- e-19 4-B TRIP SETPOINT

-190Wr of. loop ALLOWABLE VALUE N

-4 3.o 2.oc. o. s. i flow **

10^.1% of loop ,

de ~ n flow ** '

o

}",

"" h S9.9 % h

13. Steam Generator Water 23.5 21.18 2.51 8.1 *,

Level Low-Low > .5 6f narrow > 2. of narrow range instrument range instrument span span 14 Undervoltage - Reactor 7. 5 1.3 0 Coolant Pumps 110578 Volts A.C. 110355 Volts A.C.

15. Underfrequency - Reactor 2.3 0 0 Coolant Pumps 157.2 Hz 157.1 Hz j, 16. Turbine Trip

a. Low Fluid Oil Pressure N.A. N.A. N.A. 1590.00 psig 1534.20 psig

b. Turbine Stop Valve N.A. N.A. N.A. 11% open Closure 11% open

17. Safety injection Input N.A. N.A. N.A.

from ESF N.A. N.A.

a 9

3 E

o r

TABLE 2.2-1 (Continued) b 55 TABLE NOTATIONS (Continued) o? ??

NOTE 1

(Continued)

El YS }{

x T' < 31 5I

, 588.5'T (Nominal T,y9 at RATED THERMAL POWER); a, ,g g K3 = 0.000671; g; a)

7

,, P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure);

5 = Laplace transform operator, s 2; and f (AI) is a function of the indicated difference between top and bottom detectors of the -'

power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

q>

(i) for q - g between b -27% and + 7%, f3 (AI) = 0, where q and gb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q +q b iS total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt 9bexceeds -27%, the AT Trip Setpoint shall be automatically reduced by 1.57% of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of q qbexceeds +7%, the AT Trip Setpoint shall be automatically reduced by 1r06% of its value at RATED THERMAL POWER.

0.957?

3I NOTE 2: Th channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than

{ of AT span.

J j

eo 2.9 %

l

.O

-, r

w TABLE 2.2-1 (Continued) b TABLE NOTATIONS (Continued)

G o>

g NOTE 3: (Continued) m x r Ks =

0.00128/*F for T > T" and K -: 0 for T $ T"; mg

., ?.

g I = Average temperature. *F; _m m

[ T" =

Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, 5 588.5'F);

5 = Laplace transform operator, s 1; and f 2(AI) = 0 for all AI.

NOTE 4: T channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than ey 4:-04 of AT span.

l t; 'I. l %

it

-. 's

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

._. _ _ -- - ~

4.'"- Attachment B Page 16 of 16 SPECIAL TEST EXCEPTIONS N

(

3/4.10.1 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION t

3.10.4 The limitation t following requirements may be suspencec:

3.2..l e~o '

a.

Specificationd3.4.1.1 - During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:

1) The THERMAL POWER does not exceed the P-7 Interlock Setcoint, and

2) The Reactor Trip Setooints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER. .

t e

b.

Specification 3.4.1.2 - During the performance of hot cod crop time measurements in MODE 3 provided at least three reactor ccolant loops as listed in Specification 3.4.1.2 are OPERABLE. i APPLICABILITY:

During cperation below the P-7 Interlock Setpoint or performance of not roa crop time measurements. :

ACTION:

a.

With the THERMAL POWER greater than tne P-7 Interlock Setocint curing the performance of startup and PHYSICS TESTS, immediately cpes tne Reactor trip breakers, f

b.

With less than the above required reactor coolant loops CPERAELE during performarice of hot rod drop time measurements, immediataly place two reactor coolant loops in operation.

l SURVEILLANCE REOUIREMENTS 4

4.10.4.1 The THERMAL POWER shall be determined to de less than :-7 :nte-icc<

Setpoint at least once per hour during startup and PHYSICS TESTS. i 4.10.4.2 Each Intermediate and Pcwer Range channel, and P-T 1.;terle:i 153'.)

be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Ori:- t:

initiating startup and PHYSICS TESTS.

! 4.10.4.3 At least the above recuirec reacter ecolant 1:c:s s'a;. ::

CPERAELE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to initiation cf tne hot rod cc:: t: + cr.:-- e:

en u-ments arc at least orce :er a hours curirg the tet rea cet: t i

.e :n .:t-ty verif ing / c:rrect urear.er alignmenti cJ inci sta : wer a.-ai:n- :

4 ,

i

'Z' ? C:.Eis - t:::T 1 2.e 4 10-;

___ _ _ . _ _ _ _ _ - - - - - - - - - - - - ~ ~ ' ~- ~ ~ - ~ ~ ~ ~~ ~

ML20196F043

Text

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