Text

Proposed Tech Specs for Increase in Nuclear Enthalpy Rise Hot Channel Factor
	ML20070A105
Person / Time
Site:	Farley
Issue date:	06/17/1994
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20070A104	List: ML20070A100; ML20070A105;
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	NUDOCS 9406280195
	Download: ML20070A105 (7)
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Enclosure 3 Technical Specification Pages Unit 1 Paae Action B 2-2 Replace 3/4 2-8 Replace B 3/4 2-4 Replace 1

l 9406280195 940617 PDR ADOCK 05000348 P PDR

. Changes marked with strikethroughs and bold, ItalicIzcd print.

Safety Limits L

Bases The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using an enthalpy hot channel factor, FN H, of 4v46-1. 70 for >

VANTAGE 5 fuel and an FNAH of 1.55 for LOPAR fuel and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an' increase in FN H at reduced power based on the expression:

i FN AH = 4r65 1.70 [1 + 0.3 (1 - P)] for VANTAGE 5 fuel and l FN AH = 1.55 [1 + 0.3 (1 - P)] for LOPAR fuel

[

where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for f the range of all control rods fully withdrawn to the maximum allowable control ,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 _ fl (delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the 1 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 ode requirements. l The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

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i FARLEY - UNIT 1 B 2-2 AMENDMENT NO.

,1

e

. Changes marked with strikethroughs and bold, ifr/icIzcd print.

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTnALPY HOT CHANNEL FACTOR -

LIMITING CONDITION FOR OPERATION 3.2.3 FN shall be limited by the following relationship:

FN ag s 4,45 1. 70 [1 + 0. 3 (1 - P)] for VANTAGE 5 fuel and l FN AH s 1.55 (1 + 0.3 (1 - P)] for LOPAR fuel where P= THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With F Nag exceeding its limit:

a. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore Psy to within the above limi ti and demonstrate through in-core mapping thatAV g is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of exceeding the limit, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER wi-ttin 2 heure and reduce the Power Range Neutron Flux - High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and l

b. Demonstrate through in-core mapping, if not previously perfo2med per a.1 above, that FN H is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit 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

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or j b, N above; subsequent POWER OPERATION may proceed provided that F AH is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding )

this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

1 l

FARLEY - UNIT 1 3/4 2-8 AMENDMENT NO.

1

. 1 1

. Changes marked with strikethroughs and bold, ItcIlcizcd print.

POWER DISTRIBUTION LIMITS BASES l FM AH will be maintained within its limits provided conditions a. though d.

above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. ,

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3%

allowance is appropriate for manufacturing tolerance. t When NF H is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for F AHU contains an 8% allowance for uncertainties. The 8% allowance is based on the following considerations:

a. Abnormal perturbations in the radial power shape, such as from rod misalignment, affect F Ng3 more directly than Fg,

b. Although rod movement has a direct influence upon limiting Fg to within its limit, such control is not readily available to limit FN H' "Ud

c. Errors in prediction for control power shape detected during startup physics tests can be compensated for in Fg by restricting axial flux ,

distribution This compensation for FN g is less readily available.

If Ysy exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore ,

Ysg to within its limits. This restoration may, for example, involve realigning any misa1Lgned rods or reducing power enough to bring Ysy vi+h4n 'its power dependent ilmit. When the Ysy limit is exceeded, the IniBR limit is not L Ikkely violated in steady state operation, because events that could i significantly perturb the Ysy value, e.g. , static control rod misalignment, are l considered in the safety analyses. However, the inTBR 14m4 t may be violated if a i DNB limiting event occurs while Ysy is above its 14m4 t. The increased allowed action ~ time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable time to restore Ysy to within its -?

limits without allowing the plant to remain in an unacceptable condition for an extended period of time.

Once corrective action has been taken, e.g., realignment of misaligned rods or reduction of power, an incore flux map must be obtained and the measured .

value of Ysy verified not to exceed the allowed 1Luit. Twenty additional hours \

are provided to perfozm this task above the four hours allowed by Action Statement 3/4.2.3.a. The coupletion time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable because of the low probability of having a LarB limiting event within this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period

' and, in the event that power is reduced, an increase in EnrB margin is obtained at lower power levels. Additionally, operating experience has indicated that this coupletion time is sufficient to obtain the incore flux map, perfoza the l required calculations, and evaluate Ysy.

l FARLEY . UNIT 1 B 3/4 2-4 AMENDMENT NO.

i I. j i

)

. I

' Safety Limits Bases The curves of Figures 2.1-1 and 2.1-2 are based on the most '

limiting result using an enthalpy hot channel factor, FN H, of 1.70 for l VANTAGE 5 fuel and an FN AH of 1.55 for LOPAR fuel and a reference cosine with a ,

peak of 1.55 for axial power shape. An allowance is included for an increase in Fh AH at reduced power based on the expression:

FN ag = 1.70 (1 + 0.3 (1 - P)] for VANTAGE 5 fuel and l FN H = 1.55 [1 + 0.3 (1 - P)) for LOPAR fuel where P is the fraction of RATED THERMAL POWER.

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 f1 (delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature delta T trips will reduce the setpoints to provide protection consi tent 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 i 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 i design pressure, to demonstrate integrity prior to initial operation.

Y r

CARLEY - UNIT I B 2-2 AMENEMENT NO.

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR -

LIMITING CONDITION FOR OPERATION l

3.2.3 FNAH shall be limited by the following relationship N

F Ah s 1.70 [1 + 0.3 (1 - P)] for VANTAGE 5 fuel and l FN AH s 1.55 [1 + 0.3 (1 - P)) for LOPAR fuel where P= THERMAL POWER RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With F Ngg exceeding its limit:

a. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore FN AH to within the above limit; and demonstrate ,

through in-core mapping that FN is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of exceeding the limit, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoints to 5 55% of RATED THERMAL POWER within the next +

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and l l 1

b. Demonstrate through in-core mapping, if not previously performed per '

a.1 above, that FN gg is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceedi;., the limit 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

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that FN H is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED T!!ERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

1 FARLEY - UNIT 1 3/4 2-8 AMENDMENT NO.

l

1

)

' POWER DISTRIBUTION LIMITS l BASES FgN will be maintained within its limits provided conditions a. though d.

above are maintained. The relaxation of FN AH as a function of THERMAL POWER ,

allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3%

allowance is appropriate for manufacturing tolerance.

When F NAH is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection ,

~

system. The specified limit for FNAH contains an 8% allowance for uncertainties. The 8% allowance is based on the followfag considerations:

a. Abnormal perturbations in the radial power shape, such as from rod misalignment, affect FN AH more directly than Fg,

b. Although rod movement has a direct influence upon limiting Fg to within its limit, such control is not readily available to limit FNAH, and i

c. Errors in prediction for control power shape detected during startup !

physics tests can be compensated for in Fg by restricting axial flux distribution This compensation for F NAH is less readily available.

If FN exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore FN AH to within its limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring FN AH within.its power dependent limit. When the FN AH lindt is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the FN H value, e.g., static control rod misalignment, are considered in the safety analyses. However, the DNBR limit may be violated if a ;

DNB limiting event occurs while FN AH is above its limit. The increased allowed

~

action time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable time to restore NF H to within its limits without allowing the plant to remain in an unacceptable condition for an r extended period of time.

Once corrective action has been taken, e.g., realignment of misaligned rods or reduction of power, an incore flux map must be obtained and the measured value of FNAH verified not to exceed the allowed limit. Twenty additional hours are provided to perform this task above the four hours allowed by Action Statement 3/4.2.3.a. The completion time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable because of the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period and, in the event that power is reduced, an increase in DNB margin is obtained at lower power levels. Additionally, operating experience L-s indicated that this completion time is sufficient to obtain the incore flux map, perform the requf. red calculations, and evaluate FNAH*

FARLEY - UNIT 1 B 3/4 2-4 AMENDMENT NO.

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ML20070A105

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