ML20217N480

Proposed Tech Specs Bases Pages Re Safety Limits,Reactivity Control Systems & Afs

ML20217N480
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Site:	Farley
Issue date:	05/01/1998
From:	SOUTHERN NUCLEAR OPERATING CO.
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ML20217N471	List: ML20217N467 ML20217N480
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NUDOCS 9805050334
Download: ML20217N480 (10)

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Attachment 1 Revised Technical Specification Pages Unit 1 l

Eagg Instruction B 2-1 Replace B 2-2 Replace B 3/4 7-2 Replace t

Unit 2 l

Eagg Instruction B 2-1 Replace B 2-2 Replace B 3/4 1-2 Rephce

! B ?!4 1-3 Repla'ce B 3/4 1-4 Replace B 3/4 7-2 Replace 9905050334 990501 PDR ADOCK 05000348 p PM .

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2.1 SAFETY LIMITS dASES l

2.1.1 REACTOR CORE l

The restrictions of this safety Limit prevent overheating of the fuel and I possible cladding perforation which would result in the release of fission

products to the reactor coolant. Overheating of the fuel cladding is prevented l 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 from 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 correlations which have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB thermal design criterion is that the probability of DNB not l occurring on the most 13miting rod is at least 95 percent (at a 95 percent {

confidence level) for any condition I or II event. l In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

Additional DNBR margin is maintained by performing the safety analyses to a higher DNBR limit. This nargin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERMAL POWER, Reactor Coolant System pressure and average temperature which satisfy the following criteria: l

, a. The average enthalpy at the vessel exit is less than the enthalpy of l saturated liquid (far left line segment in each curve).

b. The minimum DNBR satisfies the DNB design criterion (all the other line i segments in each curve). Each curve reflects the most limiting result using either fuel with optimized fuel assembly fuel rod diameter or fuel with standard fuel assembly fuel rod diameter. The fuel with optimized rod diameter is analyzed using the WRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively. The fuel with standard rod diameter is analyzed using the WRB-1 correlation with design limit DNBR values of 1.25 and 1.24 for the typical and thimble cella, respectively.

c. The hot channel exit quality is not greater than the upper limit of the quality range (including the effect of uncertainties') of the DNB correlations. This is not a limiting criterion for this plant.

FARLEY - UNIT 1 B 2-1 Revised by NRC letter dated l l

l J

~

l j . .

l l SAFETY LIMITS l .

IsASES .

l The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using an enthalpy hot channel factor 17# yf which bounds the limit specified in the COLR, and a reference cosine with s. peak of 1.55 for axial power shape.

An allowance is included for an increase in 17,#gf at reduced power based on the expression:

N RTP P Fm = pai 1 + pm (1-P)

RTP where P is the fraction of RATED THERMAL POWER, and p g and F

pg, are specified in the COLR These limiting heat flux conditions are higher than those 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 ft (delta I) function of the overtenperature trip. wnen the axial power imbalance is not within tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety 10mits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release I of radionuclides contained in the reactor coolant from reaching the containment J

atmosphere.

The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code'for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

i.

FARLEY - UNIT 1 B 2-2 Revised by NRC letter dated

-~

PLANT SYSTEM 3 sASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The design of the auxiliary feed water system ensures that the Reactor Coolant System can be cooled down to less than 350*F (to commence cooldown with the residual heat removal system) from normal operating conditions in the event of any of the following incidents:

Loss of Normal Feedwater Loss of Off-site Power Feed Line Break Main Steam Line Break Accidental Depressurization of Steam Generators Steam Generator Tube Rupture (SGTR)

High Energy Line Break Small Break LOCA Normal Cooldown following a Reactor Trip Station Blackout Each motor driven auxiliary feedwater pump delivers a total of at least 285 gpm to all steam generators which are at a pressure of 1138 psia. The minimwm flow requirement for a motor driven pump is based on a high energy line break in the steam supply line to the steam driven auxiliary feedwater pump. In this scenario, only one motor driven auxiliary feedwater pump will be the source of auxiliary feedwater. For all other scenarios listed above, except Station Blackout, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.

The steam driven auxiliary feedwater pump delivers a total of at least 350 gpm j to all steam generators which are at a pressure of 1138 psia. The minimum requirement for the steam driven pump is based on a Station Blackout event. In this scenario, {

the steam driven auxiliary feedwater pump will be the only source of auxiliary feedwater. For all other scenarios listed above, except high energy line break in the steam supply line to the steam driven pump, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secont y system specific activity ensure that the resultant off-site radiation dose wi.1 be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.

EARLEY-UNIT 1 B 3/4 7-2 Revised by NRC letter dated l

f'

)

1 2.1 SAFETY LIMITS I

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 from nucleate boil'.ng (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 correlations which have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions, The local DNB heat flux ratio, DNBR, defined as the ratio of the heet flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB thermal design criterion is that the probability of DNB not occurring on the most limiting rod is at least 95 percent (at a 95 percent confidence level) for any condition I or II event.

In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. As described in the FSAR, the effects of these uncertainties ha'e been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

Additional DNBR margin is maintained by performing the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERMAL POWER, Reactor Coolant System pressure and average temperature which satisfy the following criteria:

l a. The average enthalpy at the vessel exit is less than the enthalpy of saturated liquid (far left line segment in each curve).

I

b. The minimum DNBR satisfies the DNB design criterion (all the other line segments in each curve). Each curve reflects the most limiting result l

( using either fuel with optindzed fuel assembly fuel rod diameter or fuel j with standard fuel assembly fuel rod diameter. The fuel with optimized rod diameter is analyzed using the WRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively. The fuel with standard rod diameter is analyzed using the 1 WRB-1 correlation with design limit DNBR values of 1.25 and 1.24 for the typical and thimble cells, respectively.

c. The hot channel exit quality is not greater than the upper limit of the quality range (including the effect of uncertainties) of the DNB correlations. This is not a limiting criterion for this plant. I EARLEY - UNIT 2 B 2-1 Revised by NRC letter dated l I 1

I SAFETY LIMITS BASES The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting resultusinganenthalpyhotchannelfactorf((fwhichboundsthelimitspecified in the COLR, and a reference cosine with a peak of 1.55 for axial power shape.

Anallowanceisincludedforanincreaseinf[hg at reduced power based on the expression:

1 N RTP F Fm = pm 1 + pm (1- P)'

RTP where P is the fraction of RATED THERMAL POWER, and p g and F

p33 are specified in the COLR These limiting heat flux conditions are higher than those 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 ft (delta I) function of the Overtemperature trip. when the axial power imbalance is not within tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

, The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

l l

FARLEY - UNIT 2 B 2-2 Revised by NRC letter dated l

l REACTIVITY CONTROL SYSTEMS BASE 5 MODERATOR TEMPERATURE COEFFICIENT (Continued)

Once the equilibrium boron concentration falls below 100 ppm, MTC measurements may be suspended provided the measured MTC value at an equilibrium boron concentration 5 100 ppm is less negative than the 100 ppm MTC surveillance limit specified in the COLR. The difference between this value and the limiting EOL MTC value conservatively bounds the maximum change j in MTC between the 100 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) and the licensed end-of-cycle, including the effects )

of boron concentration reduction, fuel depletion, and end-of-cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel l burnup.

1 l

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY '

This specification ensures that the reactor will not be made critical with ,

the Reactor Coolant System average temperature less than 541*F. This I limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed tenperature range, 2) the protective instrumentation is within its nornal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTmn temperature.

3/4.1.2 BORATION SYSTEMS

?

The boron injection system ensures that negative reactivity control is l available during each mode of facility operation. The components required to  !

perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200*F, a minimum of two boron  !

, injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The i

, bore. tion capability of eitner flow path is sufficient to provide the required l SHUTDOWN MARGIN from expected operating conditions after xenon decay and cooldown to l 200*F. The maximan expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 11,336 gallons of 7000 ppm borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppm borated water t from the refueling water storage tank.

Incorporates Amerd. 120 & 127 EARLEY-UNIT 2 b 3/4 1-2 Revised by NRC Lecter dated

REACTIVITY' CONTROL SYSTEMS BASES

• BORATION SYSTEMS (Continued)

With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 180*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve. Two charging pumps may be capable of injecting into the PCS for a short time to allow the pumps to be swapped.

This allows seal in]ection flow to be continually maintained, thus, l . minimizing the potential for RCP number.one seal damage by reducing pressure transients on the seal and by preventing RCS water from entering the seal.

Particles in the RCS water may cause wear on the seal surfaces and loss of l seal injection pressure may cause the seal not to fully reseat when pressure is reapplied. Low temperature overpressure protection is most critical during shutdown when the RCS is water solid. Mass input transients can I cause a very rapid increase in RCS pressure allowing little time for operator action to mitigate the event. For these reasons, more than one pump should be made capable of injecting into the RCS only when the RCS is in'a non water solid condition and when both RHR relief valves are OPERABLE l or.the RCS is vented via en opening of at least 5.7 square inches. A 5.7 l square inch opening is equivalent to the throat size area of two RHR relief ,

l valves. I The boron capability required below 200*F is sufficient to provide a  ;

SHUTDOWN MARGIN as specified in the COLR after xenon decay and cooldown from .

?OO*F to 140*F. This condition requires either 2,000 gallons of 7000 ppm

]

borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppm i borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical l characteristics.

l l

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 ano 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of

.- iodine and minimizes the effect of chloride and caustic stress corrosion on

mechanical systems and components.

Incorporates Amend. 118, 120, & 127 FARLEY-UNIT 2 B 3/4 1-3 Revised by NRC Letter dated

- REACTIVITY CONTROL SYSTEMS

. ggsES'

' 3/4.1.3- MOVABLE CONTROL ASSEMBLIES The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

The specifications of thisLaection ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod micalignment on associated accident-analyses. OPERABILITY of the control rod position indicators is required to

- determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

For purposes of determining compliance with Technical Specification 3.1.3.1, any inoperability of full length control rod (s), due to being immovable, invokes ACTION statement "a".

The intent of Technical Specification 3.1.3.l' ACTION statement "a" is to ensure that before leaving ACTION statement "a" and utilizing ACTION statement "c" that the rod urgent failure alana is illuminated or that an obvious electrical problem is j' detected in the rod control system by minimal electrical troubleshooting techniques.

Expeditious action will be taken to deterndne if rod immovability is due to an

- electrical problem in the rod control system.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria'are met. Misalignment of a rodLrequires measurement of l peaking factors.or a restriction in THERMAL POWER; either of these restrictions y provide assurance of fuel rod integrity during continued operation. In addition,

! those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg greater than or equal to 541*F and with all reactor coolant pumps operating ensures that the measured drop L times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

1 Incorporates Amend. 127 FARLEY-UNIT 2 B 3/4 1-4 Revised by NRC Letter dated

PLANT SYSTEMS BASES '

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The design'of the auxiliary feed water system ensures that the Reactor Coolant System can be cooled down to less than 350*F (to commence cooldown with the residual heat removal system) from normal operating conditions in the event of any of the following incidents:

Loss of Normal Feedwater Loss of Off-site Power Feed Line Break Main Steam Line Break l Accidental Depressurization of Steam Generators I Steam Generator Tube Rupture (SGTR)

High Energy Line Break Small Break LOCA Normal Cooldown following a Reactor Trip Station Blackout Each motor driven auxiliary feedwater pump delivers a total of at least 285 gpm to all steam generators which are at a pressure of 1138 psia. The minimum flow requirement for a motor driven pump is based on a high energy line break in the steam supply line to the steam driven auxiliary feedwater pump. In this scenario, only one motor driven auxiliary feedwater pump will be the source of auxiliary feedwater. For all other scenarios listed above, except Station Blackout, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.

The steam driven auxiliary feedwater pump delivers a total of at least 350 gpm to all steam generators which are at a pressure of 1138 psia. The minimum requirement for the steam driven purup is based on a Station Blackout event. In this scenario, the steam driven auxiliary feedwater pump will be the only source of auxiliary feedwater. For all other scenarios listed above, except high energy line break in the steam supply line to the steam driven pump, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total loss of off-site power. The contained water volume lindt includes an allowance for water not usable because of tank discharge line location or other physical characteristics. l I

3/4.7.1.4 ACTIVITY  !

The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coinciderit. l.0 GEM primary to secondary tube leak in the steam generator of the affected steam line. 'These values are consistent with the assumptions used in the accident analyses.

i EARLEY-UNIT 2 B 3/4 7-2 Revised by NRC letter dated j l