ML20011D704

Proposed Tech Specs Re Core Operating Limits Rept,Movable Control Assemblies,Shutdown Rod Insertion Limit,Control Rod Insertion Limit & Heat Flux Channel Factor

ML20011D704
Person / Time
Site:	Beaver Valley
Issue date:	12/14/1989
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20011D701	List: ML20011D700 ML20011D703 ML20011D704
References
NUDOCS 8912280294
Download: ML20011D704 (46)

Text

..

l ATTACHMENT A-1 Revise the Beaver Valley Unit No.

1 Technical Specifications as follows:

Remove Paces Insert Paces 1-8 1-8 3/4 1-18 3/4 1-18 3/4 1-19 3/4 1-19 3/4 1-23 3/4 1-23 3/4 1-23a_

3/4 1-23a 3/4 1-24 3/4 1-25 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-4 3/4 2-5 3/4 2-5

~l 3/4 2-6 3/4 2-6 3/4 2-6a 3/4 2-6a 3/4 2-7 3/4 2-8 3/4 2-8 B3/4 2 B3/4 2-1 i

B3/4 2-2

.B3/4'2-2 B3/4 2-4 B3/4 2-4 B3/4 2-5 B3/4'2-5 6-22 6-22 l

6-23 6-23 i

i 8912280294 891214 fDR ADOCK 05000334 F

PDC

Ah !

4 1.0 DEFINITIONS (Continued)

{

I' I

2)

Major changes in the design of radwaste treatment systems (liquid, gaseous and solid) that could significantly J

increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those l

previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);

3)

Changes in system design which may invalidate the accident analysis as described in the SER (e.g., changes in tank l

capacity that would alter the curies released); and I

4)

Changes in system design that could potentially result in a significant increase -in occupational exposure of operating i

personnel (e.g., use of temporary equipment without adequate

{

shielding provisions).

MEMBER (S) OF THE FUBLIC 1.36 MEMBERS OF THE PUBLIC shall include all persons-who are not occupationally associated with the. plant.

This category does not l

include employees of the

utility, its contractors or its vendors.

i Also excluded from this category are persons who enter the site to service equipment or-to make deliveries and persons who traverse i

portions of the site as the consequence of.a public highway, railway, I

or waterway located within the confines of the site boundary.

This 1

category does include persons who use portions of the site for i

recreational, occupational, or other purposes not associated with the l

plant.

1 CORE OPERATING LIMITS REPORT 1.37 The CORE OPERATING LIMITS REPORT-(COLR) is 'the u_nj' specific document that provides core operating limits for me current operating reload cycle.

These cycle specific core operating limits a

shall be determined for each reload cycle in accordance with Specification 6.9.1.14.

Plant operation within taese operating-l limits is addressed in individual specifications.

I l

l l

BEAVER VALLEY - UNIT 1 1-8 l

PROPOSED d

i v

e

=

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length shutdown and control rods shall be OPERABLE and positioned within i 12 steps (indicated position, as determined in accordance with Specification 3.1.3.2) of their group step counter demand position.

APPLICABILITY:

MODES 1* and 2*

ACTION:

a.

With one or more full length rods inoperable due to being immovable as a

result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With more than one full length rod misaligned from its group step counter demand position by more than i

12 steps (indicated position determined in accordance with Specification 3.1.3.2), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, c.

With one full length rod trippable but inoperable due to causes other than addressed by ACTION a above, or misaligned from its group step counter demand position by more than i 12 steps (indicated position determined in accordance with Specification 3.1.3.2), POWER OPERATION may continue provided i

that within one hour either:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable.and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits provided in-the CORE OPERATING LIMITS REPORT.

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification _3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

The THERMAL POWER level is reduced to less than or equal to 75%

of RATED THERMAL POWER within the hour

and, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high noutron flux trip setpoint is reduced to less than or equal to 85%

of RATED THERMAL P0WER.

BEAVER VALLEY-UNIT 1 3/4 1-18 PROPOSED

LIMITING CONDITION FOR OPERATION (Continu:d)

-+

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

j c) A power distribution map isobtagnedfromthemovable incore detectors and FQ(Z) and FA H are verified to be l

within their' limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, d) A reevaluation of each accident analysis of Table 3.1-1 is-performed within 5

days; this reevaluation shall confirm that the previously analyzed results of-these accidents renain valid for the duration of operation.

under these conditions.

d. With more than one rod tripable but inoperable due to causes

[

other than addressed by Action a above, POWER OPERATION may continue provided that:

1. Within one

hour, the remainder of the rods'in the bank (s) with the inoperable rods.are aligned to within i 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits provided in the CORE OPERATING LIMITS REPORT.

The THERMAL POWER -level shal'1 be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and 2.

The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

SURVEILLANCE REQUIREMENTS l

4.1.3.1.1 Each shutdown and control rod notifully inserted in the core shall be determined to be OPERABLE by movement of at.least 10 steps in any one direction at least'once per 31 days.

4.1.3.1.2 The position of each full length rod shall be determined i

to be within i

12 steps-of the associated group demand counter by verifying the individual rod position at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-

~

except during intervals when the Rod-Position-Deviation monitor is-inoperable, then verify the group position at least once per 4' hours.

l l

l

• See Special Test Exception 3.10.2 and 3.10.4 i

BEAVER VALLEY - UNIT 1 3/4 1-19 r

PROPOSED l

REACTIVITY CONTROL SYSTEM SHUTDOWN ROD INSERTION LIMIT l

LIMITING CONDITION FOR OPERATION j

1 3.1.3.5 All shutdown rods shall be within the insertion limits specified in the CORE OPERATING' LIMITS REPORT.

APPLICABILITY:

MODES 1* and 2*#

1 ACTION!

With a

maximum of one shutdown rod inserted beyond the insertion l

limit, except for surveillance testing pursuant to Specification 4.1.3.1.1, within one hour either:

a. Restore the rod to within the limit, or I

b. Declare the rod to be inoperable and apply Specification f

3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the l

insertion limit by use of the group demand counters, and verified by the analog rod position indicators **

a. Within 15 minutes prior to withdrawal of any rods in control banks A,

B, C or D during an approach to reactor criticality, and l

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

1 l

1 See Special Test Exception 3.10.2 and 3.10.4.

H

• • For power levels below 50% one hour thermal " soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps.

With Keff 21.0.

BEAVER VALLEY - UNIT 1 3/4 1-23 l

PROPOSED l

l l

l

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be within the insertion limits specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY:

Modes 1* and 2*#

ACTION:

With the control banks inserted beyond the insertion limits, except l

4 for surveillance testing pursuant to Specification 4.1.3.1.1, either:

a. Restore the control banks to'within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or i

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction or RATED THERMAL POWER which is allowed by the bank position insertion limits specified in the CORE OPERATING l

LIMITS REPORT, or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS

^

4.1.3.6 When the Rod Insertion Limit Monitor _is OPERABLE, the deviation between the position indicated by'the individual analog rod position instrument channel and the position indicated by the I

corresponding group demand -counter shall be checked ** manually for each rod at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When the Rod Insertion Limit Monitor is inoperable, the deviation between indicated positions shall be checked ** manually at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

See Special Test Exception 3.10.2 and 3.10.4 l

with Keff 2 1.0

• • For power levels below 50%, one hour thermal " soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps, l

l r

l-BEAVER VALLEY - UNIT 1 3/4 1-23a PROPOSED (next page is 3/4 2-1) l

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ULLY INSERTED)

FRACTION OF RATED THERMAL POWER Figure 3,11 Rod Group Insertion Limits Versus Thermal Power Three Loop Operation t

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l BEMER VALLEY - UMT 1 3/4 1-24 DELGTE I

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(FULLY INSERTED)

FR ACTION OF RATED THERMAL POWER FIGURE 3.12 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER TWO LOOP CPERATION' l

BUVER '/ ALLEY - UNIT 1 3/4 1 25 b&L ETE l

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I

O 3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE ( APD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY:

MODE 1 ABOVE 50% RATED THERMAL FUWER*

ACTION:

l A. With the indicated AXIAL FLUX DIFFERENCE outside of the target band and with THERMAL POWER:

1. Above 90% of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

2.

Between 50% and 90% of RATED THERMAL POWER:

i a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the target l

band for more than 1

hour-penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

2) The indicated AFD is within the target-band.

Otherwise, reduce THERMAL POWER to less than 50% of l

RATED THERMAL POWER within 30 minutes and' reduce the Power Range. Neutron Flux-High Trip Setpoints to 5 5S%

of RATED THERMAL POWER within the next-4 hours.

b) Surveillance testing of the Power Range Neutron Flux l

Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits.. A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be l

accumulated with the AFD outside of.the target band during this testing without penalty deviation.

• See Special Test Exception 3.10.2 BEAVER VALLEY - UNIT 1 3/4 2-1 PROPOSED i

}

~

l i

i POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

b. THERMAL POWER shall not be increased above 90%

of RATED THERMAL POWER unless the indicated AFD is within the target l l

band and ACTION a.2.e.) 1), above has been satisfied.

c. THERMAL POWER shall not be increased above 50%

of RATED I

THERMAL POWER unless the indicated AFD has not been outside of deviation l the target band for more than 1

hour penalty' cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1. At least once per 7

days when the AFD Monitor Alarm is OPERABLE, and

2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AXIAL FLUX _ DIFFERENCE for each OPERABLE-excore channel at'least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least-once per 30 minutes-thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

l 4.2.1.2

.The indicated-AFD shall be considered outside of its target l l

band when at least 2

of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside the: target band.

POWER OPERATION outside of the target band shall be accumulated on a time basis of:

l l

a. One minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels i

equal to or above 50% of RATED. THERMAL POWER, and-

b. One-half minute penalty deviation for each one minute-of POWER OPERATION outside of the target band at THERMAL POWER levels L

between 15% and 50% of RATED THERMAL POWER.

l BEAVER VALLEY - UNIT 1 3/4 2-2 PROPOSED

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t FLUX DIFFERENCE (Al) %

FIGURE 3.21 AXIAL FLUX 0lFFERENCE t.iMITS AS A FUNCTION OF RATED l

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i POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-Fg(Z)

T LIMITING CONDITION FOR OPERATION s

3.2.2 F (Z) shall be limited by the following relationships:

g F9(Z) 5 (CEQ) [K(Z)] for P >0.5 l

P t

F (Z) 5 [gEQ) [K(Z)] for P 50.5 g

0.5 where:

CFQ = The FQ limit at RATED THERMAL POWER provided in the CORE OPERATING LIMITS REPORT, K(Z) = The normalized FQ(Z) as a function of core height provided in the CORE OPERATING LIMITS REPORT, and P = THERMAL POWER RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

With Fq(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit within 15 minutes and similarly-reduSe the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up'to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the overpower a T Trip Setpoints have been reduced at least 1% for

~

each 1%

Fn(Z) exceeds the -limit.

The overpower A T Trip Setpoint Peduction shall be performed 'with the reactor subcritical.

b. Identify and correct the cause of the out of limit condition-prior to increasing THERMAL POWER: THERMAL POWER may then be ~

increased provided Fg(Z) is demonstrated through incore I

mapping to be within its limit.

BEAVER VALLEY - UNIT 1 3/4 2-5 PROPOSED

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The. provisions of Specification-4.0.4-are not applicable.

4.2.2.2 F

shall be evaluated to determine if F (Z) is within iYE limit by:

g

a. Using-the movable.

incore

. detectors to obtain a

power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Increasing the measured F

component of the power distribution map by 3% to accobnt for manufacturing tolerances x

and further increasing the value by 5%

to account for measurement uncertainties.

C

c. Comparing the F computed (Fxy ) obtained in b, above to:

xy

1. The F

limits for RATED THERMAL POWER (F

)

for the appropr$ ate measured core planes given in e and YYbelow, and-x

2. The relationship:

F xh = FTy (1+PFXY(1-P))

l P

L where Fxy is the limit'for fractional THERMAL POWER-operation expressed as a function of F P, PFXY is the Power Factor multiplier-for F provided in-the CORE xy OPERATING LIMITS REPORT, and P is the fraction of RATED THERMAL POWER at which Fxy was-measured.

d. Remeasuring F according to the following schedule:

xy

1. When F is greater than the F limit for the xy xy L

appropriate measured core plane but.less.than the F xy relationship, additional power distribution maps shall be taken and F compared to F and Fxy:

xy y

a) Either -within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATEp THERMAL POWER or greater,-the THERMAL POWER at'which F was last determined, or xy b) At least once per 31 EFPD, whichever occurs iirst.

l BEAVER VALLEY - UNIT 1 3/4 2-6 l

PROPOSED i

- O POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (continued)

F f is less than or equal to the F((P limit for

2. When the x

the appropriate measured core

plane, additional power distribution maps shall be taken and F

compared to xy RTP F

and F at least once per 31 EFPD.

l xy

e. The F

limit for Rated Thermal Power (F

) shall be xy xy provided for all core planes containing bank "D" control rods and all unrodded core planes in the CORE OPERATING LIMITS REPORT.

I

f. The F

limits of e,

above, are not applicable in~ the~

xy following core plane regions as measured from the bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2. Upper core region'from 85 to 100%, inclusive.

3. Grid plane-region 12%

of. core height (12.88 inches)

I measured from grid centerline.

4.

Core plane regions within 2% oficore height-(i2.88 inches)-

about the bank demand position-of the bank "D" control rods.

c 1

g. With F

exceeding Fxy',

the effects of F

on xy xy

/Fg (Z)-

shall be evaluated to determ,ine if F;

(Z) is within its g

limit.

4.2.2.3 When F

(Z) is measured pursuant to Specification g

4.10.2.2, an overall measured Fg (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing' tolerances and further increased-by 5% to account for measurement-uncertainty.

A BEAVER VALLEY - UNIT 1 3/4 2-6a PROPOSED

K(2) - NORMAL 12ED f (2)

Q AS A FUNCTION OF CORE HEIGHT N-LOOP EEAVER YALLEY - UNIT 1 x

y F.

iG.O. 3.0' 3,3 A

g f

_(10,S, o.M) a

0. 5 n

l er i c i

'w

i. 6 l u l @

N s

i

~~.

(12. 02431 5 X %'s n

~

x

\\.

\\

0.2 0

i 4

5 5

10 12

- le COP 5 HEIGHT (FT)

Tigure 3.2-2 A!4ENOME.';T NO. 58 SkAVER VALLEY - t/NI! 1 3/42-7 T>EL ETE

POWER DISTRIBUTION LIMITS N

NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F LIMITING CONDITION FOR OPERATION N

3.2.3 F

shall be limited by the following relationship:

N F 5 CFDH [1 + PFDH (1-P)]

where:

CFDH = F limit at RATED THERMAL POWER provide n the CORE OPERATING LIMITS REPORT, PFDH=ThePowerFactormultiplierforF$H provided in.the CORE OPERATING LIMITS REPORT, and P = THERMAL POWER RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

WithFMIexceedingitslimits:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2

hours 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.Demonstratethruin-core.mappingthatF$t is within'its limit l

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding.the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within:the'nextu2 hours, and

c. Identify and correct the cause of the out of' limit condition prior to increasing THERMAL gOWER, subsequent POWER OPERATION may proceed provided that FEH 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining'95% or greater RATED THERMAL POWER.

DEAVER VALLEY - UNIT 1 3/4 2-8 PROPOSED 1

\\

.o 3/4.2 POWER DISTRIBUTION LIMITS BASES

')

4 The specifications of this section provide assurance of fuel integrity during Condition I (Normal" Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in j

the core greater than or equal to the design DNBR limit during normal operation and in short term transients,-and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance j

that the initial conditions assumed for the LOCA analyses are met and 1

the ECCS acceptance criteria limit of 2200*F is not exceeded.

l The definitions of hot channel factors as used in these specifications are as follows:

Fg(Z)

Heat Flux Hot Channel

Factor, is defined as the maximum local heat flux on the surface of a

fuel rod at core elevation Z

divided by.the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

N F

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the j

g ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the Fo(Z) upper bound envelope times the normalized axial peaking factor is not l

exceeded during either normal operation, or.in the event of xenon

+

l redistribution following power changes.

I l

Target flux difference is determined at equilibriur xenon conditions.-

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be j

inserted near their normal position -for steady state. operation at I

high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at _ RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other i

THERMAL POWER levels are d

BEAVER VALLEY - UNIT 1 B 3/4 2-1 PROPOSED l

l

EQEfR DISTRIBUTION LIMITS

)

BASES i

i obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of q

the target flux difference value is necessary to reflect core burnup i

considerations.

Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the target band about the target flux l

difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviete outside of the target band at reduced THERMAL POWER Levels.

This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a substquent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited.

Accordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty i

deviation limit cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits specified in the CORE OPERATING LIMITS REPORT for THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50%

and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

Figure B

3/4 2-1 shows a

typical monthly target band near the beginning of core life.

l f

BEAVER VALLEY - UNIT 1 B 3/4 2-2 PROPOSED

l O

POWER DISTRIBUTION LIMITS BASES 3/4. 2. 2 AND 3/4. 2. 3 HEAT FLUX AND NUCTIAR ENTHALPY HOT CHANNEL FACTORS-F(Z)andPfH g

The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200'F.

Each of these hot channel factors are measurable but will normally only be determined periodically as specified in specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a. Control rods in a

single group move together with no individual rod insertion differing by more than i 12 steps from the group demand position,

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

c. The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.

N The relaxation in F

as a

function of THERMAL POWER allows 3H changes in tge radial power shape for all permissible rod insertion limits.

FAH will be maintained within its limits provided conditions a thru d above, are maintained.

When a

F measurement is

taken, both experimental error and g

manufacturing tolerance must be allowed for.

5% is the appropriate experimental error ellowance for a

full core map taken with the incore detector flux mapping system and 3%

is the appropriate allowance for manufacturing tolerance.

The specified limit of F$H contains an 8%

allowance for uncertainties which means ghat

normal, full

power, three loop operation will result in FE less than or equal to the design limitspecifiedintheCOREOPENATINGLIMITSREPORT.

BEAVFR VALLEY - UNIT 1 B 3/4 2-4 PROPOSED i

O POWER DISTRIBUTION LIMITS i

BASES Fuel rod bowing reduces the value of the DNB ratio.

Margin has been scintained between the DNBR value used in the safety analysee (1.33) i and the design limit (1.21) to offset the rod bow penalty and other

~

penalties which nay apply.

The radial peaking factor F

(Z) is measured periodically to provide assurance that the Nbt channel forRatedTherm81 Power (F[{[

factor, Fo (Z),

remg

)

within its limit.

The F

limit provided in the CORE OPERNhING LIMITS REPORT was determined om l expected power control maneuvers over the full range of burnup conditions in the core.

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

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.

The two-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow I

identification and correction of a dropped or misaligned rod.

In the event such action does not correct the

tilt, the margin for uncertainty on F

is reinstated by reducing the maximum allowed g

power by 3 percent for each percent of tilt in excess of 1.0.

4 e

DEAVER VALLEY - UNIT 1 B 3/4 2-5 PROPOSED

ADMINISTRATIVE CONTROLS O

The radioactive effluent release report to be submitted 60 days after January 1

of each year shall also include an assessment of radiation i

doses to the likely most exposed real individual from reactor t

releases for the previous calendar year to show conformance with 40 i

CFR

190, Environmental Radiation Protection Standards for Nuclear Power Operation.

Acceptable methods for calculating the doce contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1.

The SKYSHINE code (available from Radiation Shielding Information Center, ORNL) is acceptable for calculating the dose contribution from direct radiation due to N-16.

The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter as outlined in Regulatory Guide 1.21.

In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be etaluated.

The assessment of radiation doses shall be performed in accordance with the ODCM.

The radioactive effluent release reports shall also include any licensee initiated changes to the ODCM made during the 6 month

period, i

CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a

reload cycle.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP-9272-P-A,

" WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",

July 1985 (Westinghouse Proprietary).

Methodology applied for the following Specifications:

3.1.3.5, Shutdown Rod Insertion Limits 3.1.3.6, Control Rod Insertion Limits l

3.2.1, Axial Flux Difference-Constant Axial Offset Control 3.2.2, Heat Flux Hot Channel Factor-FQ(Z) 3.2.3, Nuclear Enthalpy Rise Hot Channel Factor-FN delta H 2.

WCAP-9220-P-A, Rev.

1,

" WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION",

February 1982 (Westinghouse Proprietary).

Methodology applied for the following Specification:

3.2.2, Heat Flux Hot Channel Factor-FQ(Z) 3.

WCAP-8385,

" POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT",

September 1974 (Westinghouse Proprietary).

Methodology applied for the following Specification:

3.2.1, Axial Flux Difference-Constant Axial Offset Control BEAVER VALLEY - UNIT 1 6-22 PROPOSED

ADMINISTRATIVE CONTROLS

\\

4.

T.

M.

Anderson to K.

Kniel (Chief of Core Performance Branch,

Attachment:

Operation and Safety NRC)

January 31, 1980 Analysis Aspects of an Improved Load Follow Package. Methodology applied for the following Specification:

3.2.1, Axial Flux i

Difference-Constant Axial Offset Control 5.

NUREG-0800, Standard Review

Plan, U.

S.

Nuclear Regulatory Commission, Section 4.3, Nuclear

Design, July 1981.

Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset

{

Control (CAOC),

Rev.

2, July 1981.

Methodology applied for the following Specification:

3.2.1, Axial Flux Difference-Constant Axial Offset Control j

i The core operating limits shall be determined so that all. applicable J

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.

The CORE OPERATING LIMITS

REPORT, including any mid-cycle i

revisions or supplements

thereto, shall be provided upon issuance,

(

for each reload cycle, to the NRC Document Control Desk.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control desk, within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

(

I

)

BEAVER VALLEY - UNIT 1 6-22a l

l PROPOSED S

i ATTACHMENT A-2 Revise the Beaver Valley Unit No.

2 Technical Specifications as follows:

Remove Paces Insert Paaes 1-6 1-6 3/4 1-18 3/4 1-18 3/4 1-19 3/4 1-10 3/4 1-24 3/4 1-24 3/4 1-25 3/4 1-25 3/4 1-26 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-9 B3/4 2-1 B3/4 2-1 B3/4 2-2 B3/4 2-2 B3/4 2-4 B3/4 2-4 6-18 6-18 6-19 6-19

O Efl!Q110hl VENTING 3.34 YENilNG is the controlled process of discharging air or gas from a con-finement to maintain temperature, pressure, humidity, concentration or other i

operating conditions, in such a manner that replacement air or gas is not provided or required during VENilNG. Vent, used in system names, does not imply a VENTlWi process.

MAJOR CHANGES 1.35 KAJOR CHANGES to radioactive waste systems, as addressed in Para-graph 6.36.2, (liquid, gaseous and solid) shall include the following:

1)

Major changes in process equipment, components, structures, and ef fluent monitoring instrumentation from those described in the Final Safety Analysis Report (FSAR) or the Hazards Summary Report and evaluated in the staff's Safety Evaluation Report (SER)

(e.g., deletion of evaporators and installation of demineralizers; use of fluidized bed calciner/ incineration in place of cement solidification systems);

2)

Major changes in the design of radwaste treatment systems (liquid, gaseous, and solid) that could significantly increase the quantities or activity of effluents released or volumes of solid waste stored or shipped offsite from those previously considered in the FSAR and SER (e.g., use of asphalt system in place of cement);

3)

Changes in system design which may invalidate the accident analysis as described in the SER (e.g., changes in tank capacity that would alter the curies released); and 4)

Changes in system design that could potentially result in a significant increase in (.:cupational exposure of operating personnel (e.

temporary equipment without adequate shielding provisions). g., use of HEMBER(5) 0F THE PUBLIC 1.36 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant, lhis category does not include employees of the utility, its contractors, or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries and persons who traverse portions of the site as the consequence of a public highway, railway, or waterway located within the confines of the site boundary.

This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

CORE OPERATING LIMITS REPORT 1.1,7 The CORE OPERATING LIMITS REPORT (COL R) is the unit soect f ze coeutoent thet orovices cot'e ocetattrio 11 ra n t s f ot' the cu t'rere t 00et'at I ng t'e} OaC cycle. These cycle GDucafle core operatirio 12taxtn she11 tse det et ent ned for each reloeo cycle an accordance w2t*. Speci fi cat ion 6. 9.1.14 Plant ooeration within these coerat i ng l a roi t s as adcressed in 2ncavidual soecifacations.

BEAVER VALLEY - UNIT 2 1-6 fR0Po.sEb

REACTIVITY CONTROL SYSTEMS

?

3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONQ1110N FOR OPERATION 3.1.3.1 All full length shutdown and control rods shall be OPERABLE and with Specification 3.1.3.2) of their group step counter dema, APPLICABILITY:

MODES 1* and 2*

ACTION:

With one or more full length rods inoperable due to being immovable a.

as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With more than one full length rod misaligned from its group step counter demand position by more than i 12 steps (indicated position determined in accordance with Specification 3.1.3.2), be in HOT STANDBY w'ithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, With one full length rod trippable but inoperable due to causes other c.

than addressed by ACTION a above, or misaligned from its group step counter demand position by more than i 12 steps (indicated position determined in accordance with Specification 3.1.3.k), POWER OPERATION may continue provided that within one hour either:

1.

The rod is restored to OPERABLE status within the above alignment ti$eN k -h COAS $ M A n'U5 LH1M5 A&6&T 2.

The rod is declared inoperable and the remainder of the rods in the group with th of the inoperabigy,e inoperable rod are aligned to within i 12 steps sertion limits -of-Figurr 3.1=1rrod while maintaining the rod sequence and The THERMAL POWER level shall I

be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or I

1 3.

The rod is declared inoperable and the SHUTDOWN MARGIN require-I ment of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue'provided that:

a)

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the hour and, within l

the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip setpoint is i

I reduced to less than or equal to 85% of RATED THERMAL POWER.

1

^See Special Test Exceptions 3.10.2 and 3.10.3 BEAVER VALLEY - UNIT 2

\\

3/4 1-18 ffDlb5G 1

i REACTIVITY CONTROL SYSTEMS LIM 111RG_CILNQlll0N FOR OPERATION (Continued) b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, c)

A power distribution map is obtained from the movable incore detectors and F (Z) and F are verified to be 9

H within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

d)

A reevaluation of each accident analysis of Table 3.1-1 is 1

performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions.

l d.

With more than one rod trippable but inoperable due to causes other than addressed by Action a above, POWER OPERATION may continue ande) s'n -lhe. (CLE 01512t*7t.16-l.lit T5 AENdi 1.

Within one hour, th remainder of the rods in the bank (s) with the inoperable rods are aligned to within 112 steps of the inoperable rods whi p maintaining the rod sequence and insertion limits-of figun 3.1-1.

The THERMAL POWER level l

shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and 2.

The ino hours. perable rods are restored to OPERABLE status within 72 SURVEILLANCE REOUIREMENTS 4.1.3.1.1 Each shutdown and control rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

4.1.3.1.2 i 12 steps of the associated group demand counter by verifyi rod position at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during intervals when the Rod i

Position Deviation monitor is inoperable, then verify the group position at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l

{

BEAVER VALLEY - UNIT 2 3/4 1-19 f(3/M C-b

O REACT 1v1TY CONTROL SYSTEM SHUT 00WN ROD INSERTION LIMIT LIC11N3._fCCll10N FOR OPEFA110N

. 3a r, s aiLh fl.a^lo' ',}s sps chi ~, ik 3.1.3.5 All shutdown rods shall be-fully withdreur. g gp g 7 gjp APPLICABILITY:

MODES 1* and 2**

4/r'1/75 dE/vA C ACTION

,ssa}s)kyel1he.sntN*e# l*.*" l1 With a maximum of one shutdown rod not ful'.y withdr:un, except for surveil-I lance testing pursuant to Specification (4.1.3.1.1), within one hour either:

p 4,; -/h /, L,'f

$5.5ly withdr - the to, or i

c.

b.

Declare the rod to be inoperable and apply Specification (3.1.3.1).

SLIRVEILLEE RE001P,EMENT5 4fk..;1he1Nte.');$ }.k<Y '

4.1.3.5 Each shutdo n rod shall be determii.ed to be fully withdr:wnt I

Within 15 minutes orior to withdrawal of any rods in control banks A, a.

B, C, or D during an approach to reactor criticality, and b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

• See Special Test Exception 3.10.2 and 3.10.3 AWith Keff > 1.0 BEAVER VALLEY - UNIT 2 3/4 1-24 AcoPoseb

O REACT!vlTY CONTROL SYSTEMS CONTROL ROD INSFRTION LIMITS LitilllELCOE21110R rOR_ OPE *LtJ10N w;$,k 4 I,k;h spee.;&cd4 A.

3.1.3.6 The control banks shall be-limited in-phm ical insertion 0; sh^.~

1. ^

Tig,ure 3.1 1.

COM Ot'EMTIMr 4ter171 REPpk7';

l APPLICABILITY:

MODES 1* and 2*#

ACTION:

With the control banks inserted beyond the e k ve insertion limits, except fcr l

surveillance testing pursuant to Specification 4.1.3.1.1, either:

Restore the control banks to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or a.

b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion (v; Won in-iS: b ; fi g i, or l

in 1,k,h specifrid sh & CC2E OfMAT/# L/n/U 'f%t'r c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Suh11U.Ri;E REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• See Special Test Exception 3.10.2 and 3.10.3

1. wi th Ke f f > 1. 0

[n c.x/ jtape U 7/V,2-/f I

BEAVER VALLEY - UNIT 2 3/4 1-25 P20PdSEb

o l

l (Full Wit hdra wn) 228- \\'

i i

l/

\\

'~

200 c

l f

i i

! X/

/

i 100-o

.s C

\\\\ li t

c Q

[

\\'N t

!/

l 50 j

s

/

4

/

i 4

i

/

s

/

0

.2 4

.6

.8 i'.0

/(ruiirin..ri.e>

s FR ACTION OF RATED THERMAL POWER

\\

FIGURE 3.1-1 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER THREE LOOP OPERATION 4

BEAVER VALLEY - UNIT 2 3/4 1-26 b6LETE

t 3/4.2 POWER DISTRIBUTICN LIMITS I

AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION l

3. 2.1 The indicated AXII.L FLUX DIFFERENCE (AFD) shall be maintained within-e-

" - rcent-targ

"-d

'449-d " ' - -- - - - """""'"'-----4"-----

h Sat 9efba d Sp'ecb e N 50E 56 5k'N755LINTN b5

?(5kE'$

l

~

APPLICABILITY:

M2E I above 50 Percent RATED THERMAL POWER

• ACTION:

1 With the indicated AX1AL FLUX DIFFERENCE outside of the : ? p r::r.t a.

target band 2:;t th: ' r;;t f1= differen;e and with THERMAL POWER:

1.

Above 90 percent of RATED THERKAL POWER, within 15 minutes:

a)

Either restore the indicated AFD to within the target band limits, or b)

Reduce THERMAL POWER to less than 90 percent of RATED THERMAL POWER.

2.

Between 50 percent and 90 percent of RATED THERMAL POWER:

a)

POWER OPERATION may continue provided:

1)

The indicated AFD has not been outside of the t--

? p:r::r.t target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

/~A ef N 2)

The indicated AFD it, within the lic$t; ;h;= ;r, figure 3.2-1.

OtheNise, reduce THERMAL POWER to less than 50 percent of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55 percent of RATED THERhAL POWER within the next 4~ hours.

b)

Surveillance testing of the Power Range Neutron Flux Chan-nels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintali,ed within the limits, cf Tigere 3.2-1.

A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation may be accumulated with +,he AFD outside of the target band during this testing without penalty deviation.

• See Special Test Exception 3.10.2 BEAVER VALLEY - UNIT 2 3/4 2-1 120 FvMD l

~~ ~ ~

l

\\

_ POWER DISTRIBUTION LIMITS i

LIMITING CONDIllON FOR OPERATION (Continued)

ACTION:

(Continued) b.

THERMAL POWER shall not be increased above 90 percent of RATED THERMAL POWER unless the indicated AFD is within the-

' ;;r :nt target band and ACTION a,2.a) 1), above has been satisfied, I

t THERMAL POWER shall not be increased above 50 percent of RATED c.

THERMAL POWER unless the indicated AFD has not been outsid the : 7 p;r;;nt target band for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

SURVEILLANCE REOUIREMENTS 4.2.1.1 its limits during POWER OPERATION above 15 perce

{

Monitoring the indicated AFD for each OPERABLE excore channel:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and i

2.

At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status, t

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereaf ter AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

the interval preceding each logging.of the indicated AXIAL The logged values i

1 l

i 4.2.1.2 band when at least 2 of 4 or 2 of 3 OPERABLE exco 1

i l

the AFD to be outside the target band.

target band shall be accumulated on a time basis of: POWER OPERATION outside of l

One-minute penalty deviation for each 1 minute of POWER OPERATION a.

50 percent of RATED THERMAL POWER, andoutside of the i

b.

One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels betwe 15% and 50% of RATED THERMAL POWER, l

i 1

BEAVER VALLEY - UNIT 2 3/4 2-2 l

l l

e Yl _

c,i.i 1-i i

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. tmACCEPTABLE M11.90) _M (11.f0)UNACCEPTART cP.t.a.AT!0N l -.:.j y-

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i

. h.-

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+

i.

=

_. ! 1 -

N 0 =.v-80 40

.40 2n In 0

10 to 3n 40 54 FLUX DIFFERENCE (41)%

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERwAL POWER 1

4/4 2-4 BEAVER VALLEY - UNIT 2 DELETE

l-POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR F (Z) n LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

9 CfQ F (Z) i (T (K(Z)) for P > 0.5 l

Q P-3BJ tra F (Z) 1

[K(Z))forP 1 0.5 g

g p, THERMAL POWER RATED THERMAL POWER

/

Zd %(Z) i H h br.ct hr. Obt: k;d ' ;; e$.g ; "e 3.3 u,,

~ ~ ~ ~ ~

/

gh r cer; h;ight heet hr..

/

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

q Reduce THERMAL POWER at least 1 percent for each 1 percent F (Z) a.

9 exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subse-quent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1 percent for each 1 percent Fn(Z) exceeds the limit.

The Overpower AT Trip Setpoint reduction stiall be performed with the reactor subtritical, b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER: THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.0

~

kve *, C F6l = Tie FQ tut of teAreb THaML M'M pnvded w

s'n A Cc2E OftUriNG 4 min KFPMG K(n) = TL. aces,al o %ecl FQ(t) as o %c& of cwe heGMpecJ.We<lo?ihe CMS Of'6dAmte diozy & fag (Q BEAVER VALLEY - UNIT 2 3/4 2-5 PRO Pos5=p

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS l

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F*Y shall be evaluated to determine if F (Z) is within its limit by:

9 a.

Using the movable incore detectors to obtain a pow'er distribution map at any THERMAL POWER greater than 5 percent of RATED THERMAL POWER.

b.

Increasing the measured F component of the power distribution map xy by 3 percent to account for manufacturing tolerances and further increasing the value by 5 percent to account for measurement uncertainties.

c.

Comparing the F computed (Ff)obtainedinb,aboveto:

xy x

1.

The F limits for RATED THERMAL POWER (FRTP) for the xy x

appropriate measured core planes given in e and f below, and

/FX Y h ike t'cw 2.

The relationship:

fu e f,,a m o /ty/,e',

Cet.6 odm,nc?iled n $

Eve F pp y y

L RTP p

=F

[1+

(1.p))

, NG pt'75 6*6 l

xy xy t

where F fis the limit for fractional THERMA x

POWER operation expressed as a function of F PhndPisthe l

J fraction of RATED THERMAL POWER at which F was measured.

xy l

d.

Remeasuring F according to the following scheoule:

xy 1.

When F is, greater than the FRTP limit for the appropriate x

x measured core plane but less than the F relationship, x

additional power distribution maps shall be taken and FfcomparedtoF RTP l

and F x

x xy a)

Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20 percent of RATED THERMAL POWER or greater, the THERMAL POWER at which F was last determined, or x

b)

At least once per 31 EFPD, whichever occurs first.

BEAVER VALLEY - UNIT 2 3/4 2-6 P&fn Eb

POWER DISTRIBUTION LIMITS iVRy1RijdC LHQUI REME NT $ ( Con t i n ue d ) -

C 2.

When the F is less than or equal to the FRTP limit for the xy xy appropriate measured cere plane, additional power distribution C

l maps shall be taken and F*Y compared to F*RTP and F*Y at least once per 31 EFPD.

Y e.

The F limit for Rated Thermal Power (F,P) shall be provided for T

xy all core planes containing bank "D" control rods and all unrodded core planes in-:

":di:1 P::kir[.- F:ctor Limit h crt e-0ff #64477#b / /

- Spcci fication 0. 0.1.14. //<t f.

The F limits of e, above, are not applicable in the following core xy plane regions as measured in percent of core height from the bottom of the fuel:

1.

Lower core region from 0 to 15 percent, inclusive.

2.

Upper core region from 05 to 100 percent inclusive.

3.

Grid plane regions of core height (2 2.88 inches) measured from grid centerline.

4.

Core plane regions within 12 percent of core height (12.88 inches) about the bant demand position of the bank "D" control rods.

C 9

With F exceeding F the effects of F on F (2) shall be x

x xy 9

evaluated to determine if F (Z) is within its limit.

n 4.2.2.3 Vhen F (Z) is measured pursuant to Specification 4.10.2.2, an overall 9

measured F (Z) shall be obtained from a power distribution map and increased o

by 3 percent to account for manuf acturing tolerances and further increased by 5 percent to account for measurement uncertainty.

BEAVER VALLEY - UNIT 2 3/4 2-7 Ofo/M wh

t j

i K(2) - NORMALIZED F (2) 9 AS A FUNCTION OF CORE HEIGHT 3-LOOP BEAVER VALLEY - UNIT 2 j

sN i

1.2 N

/

i

\\

8.0, 1.0 I \\

3*o

%,.e4 N

r, f

l 7

0.8 N

/

\\

/

N /

I E

12.0,.647 l

/

\\

/

\\

/

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0.2 -

/

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l

/

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o.0 12\\

0 2

4 e

a to CORE EGHT (FEET FROM BOTTOM)

FIGURE 3,2-2 BEAVER VALLEY'- UNIT 2 3/4 2-8 b6W7E

0 i

l POWER DISTRIBUTION LlHITS NUCLEAR ENTHALPY HOT CHANNEL FACTOR - FN LIMITING CONDITION FOR OPERATION N

3.2.3 FAH shall be limited by the following relationship:

cFD//

frD#

g F3g 1 %- (1 + -&+ (1-P)]

p, THERMAL POWER 2_

RATED THERMAL POWER APPLICABILITY:

MODE 1

/ ACTION:

i F[g exceeding its limit:

With Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within a.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to 155% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Demonstratethroughin-coremappingthatFhiswithinitslimit b.

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter exNeding the limit or reduce THERMAL POWER to less than 5 percent of RATED THERM L POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and l

l

< Identify and correct the cause of the out of limit condition prior c.

j to increasing THERMAL POWER, subsequent POWER OPERATION may proceed provided that Fh is demonstrated through in-core mapping to be within its limit at a nominal 50 percent of RATED THERMAL POWER prior to exceeding this THERMAL power, at a nominal 75 percent of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter attaining 95 percent or greater RATED THERMAL POWER, tdest? CFbH: 71e F 'lalf of Ursb rHootAt powse pmdec/

h y

sh -fh. C0AE OPEA'ArtM 4/!!!TS A'EPMT;

)

PFbH = Tje fwer Fcic.h ~~/hp/les k Ff pmWed a A. coes omans lieurs ieercen c,.]

BEAVEP, VALLEY - UNIT 2 3/4 2-9 NoPOSE])

i

\\

o 9

3/4.2 POWER DISTRIBUTION LIMITS 1

j BASES The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and II (Incidents of Moderate frequency) events by:

(a) maintaining the minimum DNBR in the core > 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed de-t sign criteria.

In addition, limiting the peak linear power density during Con-dition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of hot channel factors as used in these specifications are os follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat 9

flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F"h Nuclear Enthalpy Rise Hot Channel factor, is defined as the ratio of O

the integral of linear power along the rod with the highest integrated power to the average rod power.

3/4.2.1 AX1AL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F (2) upper bound 9

envelope ef 2.22 times the normalized axial peaking factor is not exceeded during l

either normal operation or in the event of xenon redistribution following power i

changes.

Target ~ flux differen:e is determined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL PDWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the-!-7%-target band about the target flux difference, l

during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

Thir otsiation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time BEAVER VALLEY - UNIT 2 B 3/4 2-1 f00fC. SED

POWER DISTRIBUTION LIMITS Sf' e. CihtY 151 Ytt gas

/?EMer 4..*

/

AXI AL FLUX DIFFERENCE (AFD) (Continued) i duration timit of the deviation is limited.

Accordin viation limit cumulative during the previous 24 hou ' gly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty de-is provided for operation outside of the target band but within the limits cf,ipre 1.2-1 whHe-d THER-l MAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERKAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

reflects this reduced significance.

The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

Durin levels between 50% and 90% and between 15% and 50%g operation at THERMAL POWER of RATED THERMAL POWER, the computer' outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

Figure B 3/4 2-1 shows a typical monthly target band near the beginning of core life.

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTO F G)and F q

g The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200'F.

Each of these hot channel factors art measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

Control rods in a single group move together with no individual rod a.

insertion dif fering by more than i 12 steps from the group. demand

position, b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

BEAVER VALLEY - UNIT 2 B 3/4 2-2 M CAGSfh

POWER DISTRIBUTION LIMITS BAJES 3/4.2.2 and 3/4 2.3 HEAT FLUX AND NUCLEAR ENTHALPY N

q AND F g (Continued)

~

~

The control rod insertion limits of Specifications 3.1,3.5 and 3.1.3 6 c.

are maintained.

d.

DIFFERENCE is maintained within the limits.The axial power d The relaxation in FN as a function of THERMAL POWER allows ch nges in g

the radial power shape for all permissible rod insertion limits.

F6H will be maintained within its limits provided conditions a thru d above, are maintained When an F tolerance must be allowed for.q measurement is taken, both experimental error 5% is the appropriate experimental error allow-ance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

The specified limit of FN g contains an 8% allowance for uncertainties which means that normal, full power, three loop operation will result in FNg i ;,3j;,gg, Fuel rod bowing reduces the value of the DNB ratio.

to offset this reduction in the generic margin.

Credit is available The generic design margins, worst case which occurs at a burnup of 24,000 WD/MTV). total e,

This margin includes the following:

leu ba*s cwefaalk g

/4fjt pe; Q 1.

Design Limit DNBR of 1.30 vs. 1.28 Grid Spacing (K ) of 0.046 vs. 0.059 4 f gg. pygpg.

2.

3 3.

Thermal Dif fusion Coef ficient of 0.038 vs. 0.059 g g g,g 4.

DNBR Multiplier of 0.865 vs. 0.88 5.

Pitch reduction The radial peaking factor fxy (Z) is measured periodically to provide assurance that the hot channel factor, Fg (Z), remains within its limit. The F*Y limit for Rated Thermal Power (Fg3p

&E 0/EAArti 4provided in the @4edhd-Peek 4ng N~

4ector-t-imi-t-Report-peMpeeH4eet4en 0.0.1.1? was determined f rom expected

• E 4/W / 73 2& fear power control maneuvers over the full range of burnup conditions in the core 3/4.2.4 QUADRANT POWER TIL1 RATIO bution satisfies the design values used in the power capab BEAVER VALLE.Y - UNIT 2 B 3/4 2-4 PlcPost%

\\

I

\\

AMIN 151RMlVLCONIROLS SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued The radioactive ef fluent release report to be submitted 50 days af ter January 1 of each year shall also include an assessment of radiation doses.to the likely most exposed real individual from reactor releases for the previors calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation.

Guide 1.109, Revision 1. dose contribution from liquid and gaseous effluent Information Center, (ORNL) is acceptable for calculating the dose cont from direct radiation due to N-16.

The radioactive ef fluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during addition, the un,each calendar quarter as outlined in Regulatory Guide 1.21.

air doses shall be evaluated. restricted area boundary maximum noble gas gamma air and beta In formed in accordance with ODCH. The assessment of radiation doses shall be per-initiated changes to the ODCM made during the 6 month period.Th M M PEAKlHG FACTOR LIMIT REPORT 6.9.1.14 The it for Pated Thermal Power (F RTP allcoreplanesTontain nk "0" control rods aM a)ll unroshall be prov*

at least 63 days prior to cyc e ' ' tal criticality, lanes limit would be submitted at some other h event that the dur.in ore life, it will be submitted 60 days prior to the date tt i

otherwise exempted by the com become effective unless n.

Any informa '

eded to support F will be by request from the not be included in this report.

C4 (

nee SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

ECCS Actuation, Specifications 3.5.2 and 3.5.3 a.

b.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

Inoperable Meteorological Monitoring Instrumentation, c.

Specification 3.3.3.4.

BEAVER VALLEY - UNIT 2 6-18 h.?ofC38)

_l

INSERT 1 CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT bef ore each reload cycle or any remaining part of a reload cycle. The encl yti c al methods used to determine the core operating l i mi ts shcIl be those previousl y reviewed and approved by the NRC in:

1.

WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (Westinghouse Proprietary).

Met hodol ogy applied for the f ollowing Specifications 3.1.3.5, Shutdown Rod Insertion Limits 3.1.3.6, Control Rod Insertion Limits 3.2.1, Axial Flux Olfierence-Constant Axial Offset Control 3.2.2, Heat Flux Hot Channel Fact or-FQ ( Z )

3.2.3, Nuclear Enthalpy Rise Hot Channel Factor-FN delta H 2.

WCAP-9220-P-A,Rev.

1,

" WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION", February 1982 (Westinghouse Proprietary).

M2thodology applied f or the f ollowing Specification:

3.2.2, Heat Flux Hot Channel Fact or-FQ ( 2 )

3.

WCAP-83BS, " POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPORT", September 1974 (Westinghouse Proprietary).

Methodology applied f or the f ollowing Specif i cation 3.2.1, Axial Flux Di f f erente-Constant Axi al Offset Control 4.

T.

M.

Anderson to K.

Kniel(Chief of Core Perf ormance Branch,NRC)

January 31, 1980 --

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

Methodology applied for the f ollowing Speci f ication:

3.2.1, Axial Flux Di f f erence-Const ant Ax i al Offset Control 5.

NUREG-OBOO, Standard Review Plan, U.

S.

Nuclear Regulatory Commi ssi on, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.2, July 1981.

Me thodol ogy applied for the f ollowing Specification 3.2.1, Axial Flux Difference-Constant Axial Of f set Control The core operating limits shall be determined so that all applicable limits (e.g.,

f uel thermal-mechanical l i mi t s, core thermal-hydraulic limits, ECCS limi ts, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. 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.

e ATTACHME*fT B Safety Analysis Beaver Valley Power Station Proposed Technical Specification Change BV-1 Change No. 167 BV-2 Chanae No. 29 Description of amendment request:

The proposed amendment would revise applicable specifications by replacing the cycle specific parameter limits with reference to the Core operating Limits Report (COLR) which contains the values of those limits.

This change reflects the guidance provided by the NRC in Generic Letter 88-16 concerning the relocation of cycle specific technical specification limits to the COLR.

A definition of the Core operating Limits Report (COLR) has been added to the Definition section of the technical specifications describing this as a

unit specific document providing these limits for the current operating reload cycle.

The definition also notes that the values of these cycle specific parameter limits are to be determined in accordance with Specification 6.9.1.14.

Specification 6.9.1.14 has been revised to require the core operating limits to be determined and provided in the COLR for each reload cycle in accordance with the referenced NRC approved methodology-and that the core operating limits are consistent with the applicable safety analysis limits.

In

addition, this report and any mid cycle revisions must be provided to the NRC upon issuance.

The individual specifications have been reviewed to reference the limits specified in the COLR in lieu of listing tne limits in each specification.

The following specifications have been revised to reference the COLR:

3.1.3.1 Movable Control Assemblies (currently references Figures 3.1-1 and 3.1-2 which are being deleted from Specification 3.1.3.6) 3.1.3.5 Shutdown Rod Insertion Limit 3.1.3.6 Control Rod Insertion Limit (Figures 3.1-1 and 3.1-2 are being deleted) and an editorial change for BV-1 to correct the reference from Special Test Exception 3.10.3 to 3.10.4.

3.2.1 Axial Flux Dif ference (AFD) (Figure 3.2-1 is being deleted) 3.2.2 Heat Flux Hot Channel Factor-FQ(Z) (this includes the F limits currently provided in the Radial Peaking Factor Limik Report) (Figure 3.2-2 is being deleted) 3.2.3 Nuclear Exthalpy Hot Channel Factor-FN delta H Bases 3/4.2.1 Axial Flux Difference (AFD) is being revised to remove the specific value of the FQ(Z) upper bound envelope, to remove the specific 17% AFD value, and to replace reference to Figure 3.2-1 with reference to the COLR.

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a ATTACHMENT B (Continued) e Bases 3/4.2.2 and 3/4.2.3 Heat Flux and Nuclear Enthalpy Hot Channel Factors-FQ(Z) and FN delta H is being revised to replace reference to the Radial Peaking Factor Limit Report with reference to the COLR.

These changes will eliminate the need for future technical specification changes resulting from changes in reload cycle limits.

The cycle specific parameter limits will continue to be determined in accordance with the NRC approved methodology and will be consistent with the safety analyses limits.

The plant will continue to be operated within the analyzed limits and these limits w211 be provided to the NRC in the COLR to allow trending the values of the limits without the need for prior NRC

approval, therefore, appropriate measures exist to control the values of these limits.

Since there is no change in the methodology used for determining these limits and the limits are used in the same manner as the existing limits are used these changes are determined to be administrative in nature and do not affect the UFSAR or reduce the safety of the plant.

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ATTACHMENT C O

No Significant Hazard Evaluation Beaver Valley Power Station Proposed Technical Specification Change BV-1 Change No. 267 BV-2 Chanae No. 29 Basis for proposed no signi41 cant hazards consideration determination:

The Commission has provided standards for determining whether a

significant hazards consideration exists in accordance with 10CFR50.92(c).

A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a

significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

The proposed changes do not involve a

significant hazard consideration because:

1.

The proposed changes reflect the guidance provided in Generic Letter 88-16 for requests to remove the values of cycle specific parameter limits from the technical specifications.

The establishment of these limits in accordance with an NRC approved methodology and the incorporation of these limits into the COLR will ensure that proper steps have been taken to establish the values of these limite.

In addition, submittal of the COLR will allow the NRC to continue to trend the values of these limits without the need for prior staff approval of these limits and without introduction of an unreviewed safety question.

The revised specifications with the removal of the values of cycle specific parameter limits and the addition of the referenced report for those limits does not create the possibility of a new or different kind of accident from those previously evaluated.

2.

These changes are administrative in nature because the values of cycle specific parameter limits will continue to be determined in accordance with an NRC approved methodology consistent with the applicable limits of the safety analysis.

Consequently, the proposed change on the removal of the values of cycle specific limits does not involve a significant increase in the probability or consequences of an accident previously evaluated.

3.

The proposed amendment alters neither the requirement that the plant be operated within the limits for cycle specific parameters nor the required remedial actions that must be taken when those limits are not met.

With the removal of the values of these limits from the technical specifications, they have been incorporated into the COLR that is submitted to the Commission.

Hence, appropriate measures exist to control the values of these limits.

These changes do not alter the methods used to establish those

limits, therefore, these changes do not involve a

significant reduction in the margin of safety.

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e ATTACHMENT C (Continu;d) l e.

t Therefore, based on the above considerations, implementation of i

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the proposed changes will not-involve a significant hazard.

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