Text

Proposed Tech Specs Change to TS 3/4.2.2 Modifies TS to Be IAW NRC Approved W Methodologies for Heat Flux Hot Channel factor-FQ(Z).Changes to TS Section 6.9.1.6 Are Adminstrative in Nature
	ML20199L080
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Site:	Millstone
Issue date:	01/18/1999
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20199L067	List: ML20199L063; ML20199L080;
References
	NUDOCS 9901270120
	Download: ML20199L080 (36)
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1-2yt&

I October 18,1995 POWER DISTRIBlTTION LIMITS 3/4.Y.2 HIAT FLUX HOT CHANNEL FACTOR -3 F (Z)

FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.2.2.1 Fo(Z) shall be limited by the following relationships:

Fo (Z) $ F*"" K(Z) for P > 0.5 P

p*aw Fo(Z) S K(Z) for P S 0.5 0.5 Fo" - the Fo limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR).

Where: P, HEMM POM RATED THEMAL POWER, and K(Z) -

the normalized eF(Z) as a function of core height specified in the COLR.

APPLICABILITY: MODE 1. -

ACTION:

f&I g With Fo(Z) exceeding limit: )

a. For RAD eration with Fo(Z) ou de the applicable limit specified in COLR:

l (1) Within 15 minutes ontrol the AFD to within n FD limits which are dete ed by reducing the applicab AFD limits by 1% AFD for h percent Fo(Z) exceeds i limits. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , r et the AFD alarm setpoin ' to these modified

{ limits, i

a k

f MSet6 h 9901270120 990118 PDR ADOCK 05000423 P PDR gLSTONE-t11T3 3/4 2-5 Amendment No. pp, 79,77,120

(

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

i Millstone Unit 3 Technical Specification Change Request No. 3-27 98 5

Insert A:

l With Fn(Z) exceeding its limit : -

a. For RAOC operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

(1) Reduce THERMAL POWER at least 1% for each 1% F n(Z) 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER

' OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% Fn(Z) exceeds the ,

limit, and i

(2) Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by item (1)., above; THERMAL POWER may then be increased provided Fn(Z) is demonstrated through incore mapping to be within its limits.

i

b. For RAOC operation with Specification 4.2.2.1.2.c not being satisfied, one of ;

the following actions shall be taken- l l

(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by at least 1% AFD for each percent i Fn(Z) exceeds its limits. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints I to these modified limits, or (2) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

Where it is necessary to calculate the percent that Fn (Z) exceeds the limits for item (1) above, it shall be calculated as the maximum percent over the core height (Z), consistent with Specification 4.2.2.1.2.f, that Fn(Z) exceeds its limit by the following expression:

J

., . , Millstone Unit 3 Technical Specification Change Request No. 3-27-98

,. j . !

Fy(Z)x W(Z) ,

p,, -1 x 100 for P > 0.5 i

0 x K(Z) !

__ P _ _

i i

i Fy(Z)x W(Z) !

-1 x 100 for P < 0.5 i F,,, :

x K(Z) !

__ 0.5 . ._ j i

i

c. h For base load operation with Specification 4.2.2.1.4.c not being satisfied, one ;

of the following actions shall be taken: '

' (1) Place the core in an equilibrium condition where the limit in 4.2.2.1.4.c .;

is satisfied, and remeasure F q "(Z), or I (2) Reduce THERMAL POWER at least 1% for each 1% F (Z) n 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent i 4

POWER OPERATION may proceed provided the Overpower AT Trip i Setpoints have been reduced at least 1% for each 1% Fn(Z) exceeds the j

' limit. The percent that F n(Z) exceeds its limit shall be calculated as the l

maximum percent over the core height (Z), consistent with ;

Specification 4.2.2.1.4.f, by the following expression: l Fy(Z) x W(Z),o

-1 x 100 for P > APLno F,,

-n x K(Z)

__. P .

k u ,

'. l u 1 1

, ,. ., - - - . . .-. , + - , ,- , .

3 -4 16 POWER DISTRIBtJTION LIMITS

LIMITING CONDITION FOR OPERATION (Continued)

(2) Reduce THERMAL POW limit within 1 at least 1% for each 1% Fo(Z) exceeds the Neutron Flux- inutes and similarly reduce the Power Range {

OPERATION gh Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER POWER OP y proceed for up to a total of 72 h urs; subsequent TION may proceed provided the erpower AT Trip Setpo s have been reduced at least 1 for each 1% Fo(Z) exc ds the limit, or

]

(3) erify that the requirements of pecification 4.2.2.1.3 for base load operation operation. are sa sfied and enter base load Where it is necessary to ca ulate the percent that Fo(Z) exceeds the limits for items (1) d (2) above, it shall be calculated as the maximum percent over he core height (Z) that Fo(Z) exceeds its limit by the followin expression:

~ '

F[( X W(Z)

~1 x 100 for y a r

gr x K(Z)

.5

, p enw-

/ Ff(Z) x w(g) p arr 1 X 100 for P < 0. 5

~

0.5 -

b. For base lo the COLR, operation outside the applicable imit specified in rform either of the following acti s:

(1) Place the core in an equilibrium co ition where the limit in 4.2.2.1.2.C is satisfied, and rem sure Fo(Z), or

_ J 1

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MILLSTONE - UNIT 3 3/4 2-6 one Amendment No. ff.120

POWER DISTRIBUTION L1 HITS 3-27$

Dec mber 29,1994

. LIMI. TING CONDITION FOR OPERATION (Continued)

(.2.) Reduce .

L POWER at least 1% for each 1% Fo(Z) exceeds the L

limi it61n 15 minutes and similarly reduce the Power Range I

He ron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; WER OPERATION may proceed fo ilp to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION ma proceed provided the Overpower AT Trip setpoints have be reduced at least 1% for ach 1%

Fo(Z) exceeds the limit. e percent that Fo exceed ts limit shall be calculated the maximum percent o r the core height (Z) by the lowing expression:

Fo"(Z) x )n , ,

F"/xK(2) l

c. Ide ify and cor:ect the caus of the out-of-limit cond ion prior t increasing THERMAL POWE above the reduced limit equired by l TION a or b, above; THEP L POWER may then be inc ased provided Fo(Z) is demonstrated rough incore mapping to e within this limit. ,

, L ~ -

SURVEILLANCE REQUIREMENTS 4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.1.2 For RAOC operation, Fo(Z) shall be evaluated to' determine if Fe(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution I# map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

reas g the measured Fo omponent of the r dis'tributio map 3% to account manufacturing t s and f er creasing the value 5% to account for surement unce inties.

Verify the requi ents of Specificat 3.2.2.1 are isfied.

.Tsem !.3 MILLSTONE - UNIT 3 3/4 2-7 Amendment No. 79 79, 99 C210

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

l

. . Millstone Unit 3 Techn.ical Specification Change Request No. 3-27-98 i

e. ,., l Insert B:

b. Evaluate the computed heat flux hot channel fact r by Performmg.both of the following :

(1) Determine the computed heat flux hot channel factor, F n "(Z), by increasing the. measured F n (Z) component of the power distribution .

map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and

!' (2) Verify that F n"(Z) satisfies the requirements of Specification 3.2.2.1 .

for all core plane regions, i.e. 0 - 100%, inclusive, t

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

. - _. . _ = . . . _ . . . . _ .

3-4-%

October 18,1995 '

POWER DISTRIBUTION LIMITS

'h SURNILLAkCEREQUIREMENTS(Continued)

c. Satisfying the following relationship:

Ff(Z) s F*" x K(Z) for P > 0.5 P x W(Z)

Ff(Z) $ F** x K(Z) for P $ 0.5 W(Z) x 0.5 where Fs(Z) is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FE" is the Fo limit, K(Z) is the normalized Fo(Z) as a function of core height, P ,

is the relative THERMAL POWER, and W(Z) is the cycle-dependent function during normal accountsF ower thatoperation. distribution transients encountered for p$", K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.6.

g

d. Measuring 7"o(Z) according to the following schedule: '

(1) Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which Fo(Z) was last determined,* or (2) At least once per 31 Effective Full Power Days, whichever occurs first.

e. With the maximum value of Ff(Z)

K(Z) !;

over the core height (Z) increasing since the previous detemination y of Fs(Z), either of the following actions shall be taken:

C A ,

m (1) Fy( hall be in ased over t specified Specifi on 4 .2.1.2c by an propriate f or specified the CO , or

.D1Sett- C A During power escalation.at the beginning of each cycle, oower level may be increased until a power level for extended operation has been achieved and power distribution map outlined.

Mil l STnNF . flwf T 1 in 9 o a-a--* **- -a -a -a *

, , Millstone Unit 3 Technical Specification Change Request No. 3 27-98 s' .'.

Insert C:

(1) Increase Fn "(Z) by an appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification ,

4.2.2.1.2.c, or I

l 1,

December 29,1994 3D

, POWER DISTRIBUTION LINITS SURVEILLANCE REQUIRENENTS (Continued)

)

b. Durin APL""g base load operation, if the THERMAL POWER is decreased below then the conditions of 4.2.2.1.3.a shall be satisfied before reentering base load operation.

~

4.2.2.1.4 During base load operation Fe(Z) shall be evaluated to determine if Fo(2) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APL"".

pRT p 73 % y p --e lr b. Increas he measured F component of power distri ion map 3% to accou or manufactur tolerances reasing the va y 5% to accoun Verify the re ements of Speci r measuremen ation 3.2.2.1 e satisfiedy.

certainties. further )I

c. Satisfying the following relationship:

utcosso sime wwS W twnorkruns TkkRNM.S puzzy $ f*"o x K(Z) , y g my oacummy, P x W(Z) ,

dere: Ff(Z) is the measured Fo(Z . F"" o is the F limit,sthe [>

Op __ normalized Fo(Z) as a function o core heighth P is the relative THERMAL POWERdAW(Z) , is the cycle-dependent function that accounts for limited power distribution transients gg g encountered during base load operation. FS", K(Z), and r W(Z) are specified in the COLR as per Specification 6.9.1.6.

d. . Measuring F#(Z) in conjunction with target flux' difference determi-nation according to the following schedule:

(1) Prior to entering base load operation after satisfying Sec-tion 4.2.2.1.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been maintained above APL"" for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and (2) At least once per 31 Effective Full Power Days.

k ')

l NILLSTONE - UNIT 3 3/4 2-10 Amendment No. JJ., 77,99 0210 l

r

. . Millstone Unit 3 Technical Specification Change Request No. 3-27-98 t

.. ., j insert D:

b. Evaluate the computed heat flux hot channel factor by performing both of the following :

(1) Determine the computed heat flux hot channel factor, Fn"(Z), by increasing the measured Fn (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that F n"(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e. 0 - 100%, inclusive-.

I 1

1 I

3- H s l October 18.1995 PdNERDI'STRIBUTIONLINITS SURVE!LLANCE REQUIREMENTS (Continued)

e. With the maximum value of -

Ff(Z)

K(2) over the core height (2) increasing since the previous determination g of F#(Z), either of the following actions shall be taken:

E F ) shall be inc sed over that spe g ed in 4.2.2.1

[

(l an[) c by appropriate ctor specified in t V COLR, or (2) F7(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps . indicate that the maximum ]

value of Ff(Z) $W K(Z) over the core height (Z) is not increasing. l

f. The limits specified in 4.2.2.1.4.c and 4.2.2.1.4.e are not i.

applicable in the following core plane regions:

(1) Lower core region 0% to 15%, inclusive. l (2) Upper core region 85% to 100%, inclusive.

4.2.2.1.5 When Fo(Z) is measured for reaso6s other than meeting the require-ments of Specifications 4.2.2.1.2 or 4.2.2.1.4, an overall measured. Fe(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. '

\

MILLSTONE - UNIT 3 3/4 2-11 Amendment No. 77, 77, 77, 120 0328

Millstone Unit 3 Technical Specification Change Request No. 3-27 98 s' '.

Insert E:

(1) Increase Fn "(Z) by an appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.4.c, or

MMS ADMINISTRATIVE CONTROLS

. . October 18.1995 CORE OPERATING LTMITS REPORT (Cont.)

- 2. e '. i Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 3.

Control Rod Insertion Limits for Specification 3/4.1.3.6, 4.

Axial Flux Difference Limits, target band, and APL"" for Specifica-tions 3/4.2.1.1 and ?/4.2.1.2, _

l S.

Heat Flux Hot Channel Factor, K(z), W(z), APL"", and W(z)g for Specifications 3/4.2.2.1 and 3/4.2.2.2.

K G0T 6.

p% Nuclear Enthalpy3/4.2.3.

for Specification Rise Hot Channel Factor, Power Factor Multiplier 6.9.1.6.b 3~nfer F shall be those previously reviewed and approved by the NRC in

1. 1 WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATIO July 1985 (H Proprietary). (Methodology for Specifications 3.1.1.3--

Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit 3.1.3.6--Control Bank Insertion Limits, 3.2.1--Axial Flux Difference, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuclear Enthalpy Rise Hot Channel Factor.)

2.

WCAP-8385, " Power Distribution Control aboad Following Procedures - Topical Report,* September 1981 (H Proprietary). , [&

2 g.

T. M. Anderson January to K. Kniel 31, 1980 -

Attachment:

(Chief of Core Performance Branch, NRC) of an Improved Load Follow Package. Operation and Safety-Analysis Aspec

' i '.

NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Comissio Section 4.3, Nuclear Design, July 1981 Branch Technical Position Revision 2, July 1981.CPB 4.3-1, Westinghouse Constant Axial Off if5' g g g WCAP-10216-P-AhRELAXATIONOFCONSTANTAXIAL SURVEILLANCE TECHNICAL SPECIFICATION,"tober Rev.191, FFSELC NTROL FQ (M Proprietary).

(Methodology for Specifications 3.2.1--Axial Flux Difference [ Relaxed Axial Offset Control) and 3.2.2--Heat Flux Channel Factor (W(z) surveillance requirements for Fo Methodology).)

5 g.

WCAP-9561-P-A, ADD. 3, Rev. 1, "BART A-1:

A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TPANSIENTS--SPE l THIMBLE tary). MODELING M ECCS EVALUATION MODEL," July 1986 (E Prop (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.)

(, /. hhdendAm 1 \Sev.2-F-h M WCAP-10266-P-A, A v.* 2, "THE 1981 VERSION OF WESTINGHOUSE EVALU MODEL USING M

for Specific4 BASH CODE," March 1987 (M Proprietary). (Methodology ation 3.2.2--Heat Flux Hot Channel Factor.)

MILLSTONE casi - UNIT 3 6-20 i Amendment No. 7f, ;y, 79. JF, JJ,120

t i

1

. Millstone Unit 3 Technical Specification Change Request No. 3-27-98 i

s> . . >. <

1 1

Insert F. ,

1

7. Shutdown Margin Monitor minimum count rate for Specification 3/4.3.5.

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JULY 26, 1993

. , ADMINISTRATIVE CONTROLS 2-2 $

CORE OPERATING LIMITS REPORT (Cont.)

,# I 17 , WCAP-Il946, " Safety Evaluation Supporting a More Negative EOL 1 Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," September 1988 (M Proprie-tary).

8. # WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATI0'N MODEL USING THE NOTRUMP CODE," August 1985 (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

9. Hf. WCAP-10079-P-A, "NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL

{

NETWORK CODE," August 1985 (M Proprietary). (Methodology for l Specification 3.2.2 - Heat Flux Hot Channel Factor.) i 10 . M WCAP-12610, " VANTAGE + Fuel Assembly Report," June 1990 (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

Zel5EW G 1

-M t

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l

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MILLSTONE - UNIT 3 6-20a Amendmer. No.8i 0117

l l

l

. . Millstone Unit 3 Technical Specification Change Request No. 3-27-98 l

l 3- !

l l

Insert G: I I

l l 11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, " Safety Elvaluation for Topical Report, NUSCO-152, Addendum 4, ' Physics Methodology for PWR l Reload Design,' TAC No. M91815," July 18,1995.

i

12. Letter from E. J. Mroczka to the USNRC, " Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal- Boron Dilution Analysis," B13678, i December 4,1990. I i

13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, " Issuance of Amendment (TAC No. 77924)," March 11,1991.

14. Letter from M.11. Brothers to the USNRC, " Proposed Revision to Technical l

Specification, Shutdown Margin Requirements and Shutdown Margin Monitor Operability for Modes 3,4 and 5 (PTSCR 3-16-97)," B16447, May 9,1997.

l 15. Letter from J. W. Anderson (USNRC) to M. L. Bowling (NNECO), " Issuance i

of Amendment - Millstone Nuclear Power Station, Unit No. 3 l (TAC No. M98699)," October 21,1998.

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March 11, 1991 b Vz?-73 POWER DISTRIBUTION LIMITS w

BASES' -

(-

3/4.2.2 and'3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) .

design Margin is maintained between the safety analysis limit DNBR and the limit DNBR.

This margin is more than sufficient to offset any rod bow penalty and transition ' core penalty. The remaining margin is available for plant design flexibility.

l When an Fn measurement is taken, an allowance for both experimental error and manufacturlng 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 i

3% allowance is appropriate for manufacturing tolerance. hm y /

, _ ^

~

The hot chan M cycle and he factor F (z) is measured periodically and increased by a

-dependent pdwer f or appropriate to ett lo5d oper RAOC or base on, W(z) or W(z)BL, provide assurance that e limit on the hot channel actor, F (z) is met l tio 0 (z) accounts for the ects of normal ope -

i ransients a6d was d ermined from expected er control maneuver over full range of burn conditions in the cor restrictive operati W(z)BL accounts fo he more limits allowed by b load operation wh result in f) less severe tran ent values. The W(z) d W(z)BL actions d ribed above for 5 normal operat n are specified in the LR per Specification 6.9.1.6.

~

When RCS flow rate and FN g are measured, no additional allowances are necessary prior to comparison with the limits of the Limiting Condition for Operation.

loop flow for RCS total flow rate and 4% for FMeasurement errors of 2.4%

determination of the design DNBR value. AH have been allowed for in The measurement error for RCS total flow rate is based upon performing a precision indicators.heat balance and using the result to calibrate the RCS flow rate potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned at least once for 18 months. Any fouling which might bias the RCS flow rate measurement

greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate fouling. the

('

MILLSTONE - UNIT 3 B 3/4 2-4 0028 Amendment No. 12,60

l , , Millstone Unit 3 Technical Specification Change Request No. 3-27-98 Insert 11:

The heat flux hot channel factor, F n(Z), is measured periodically using the incore detector system. These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive Fn"(Z), a computed value of Fn (Z) . Ilowever, because this value represents a steady state ;

condition, it does not include the variations in the value of F n(Z) that are present during nonequilibrium situations.

To account for these possible variations, the steady state limit of Fn (Z) is adjusted by an elevation dependent factor appropriate to either RAOC or base load operation, W(z) or W(z)nt, that accounts for the calculated worst case transient conditions. The W(z) and W(z)nt factors described above for normal operation are specified in the COLR per Specification l 6.9.1.6. Core monitoring and control under nonsteady state conditions e.re accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state Fn(z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation nonequilibrium limits is performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c.

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Docket No. 50-423 l

.. . B17609 i

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Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Heat Flux Hot Channel Factor and Core Operating Limit Report (COLR) Modifications (TSCR 3-27-98)

Retyped Pages - )

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i January 1999 l 1

n -

Nucisar R:gul tory Commission

, , B17609\ Attachment 2\Page 1 RETYPE OF PROPOSED REVISION i, .

Refer to the attached retype of the proposed revision to the Technical Specifications.

The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

(

'u

POWER DISTRIBUTION LINITS

. ' 3/4.2.2 NEAT FLUX HOT CHANNEL FACTOR - Fe (Z)

FOUR LOOPS OPERATING LINITING CONDITION FOR OPERATION ;

3.2.2.1 Fo(Z) shall be limited by the following relationships:

parp f a(Z) s K(2) for P > 0.5 P

(RTP f a(Z) s K(2) for P s 0.5 0.5 FS" = the Fo limit at RATED THERMAL POWER (RTP) provided in the core operating limits report (COLR).

THERMAL POWER Where: P= RATED THERMAL POWER, and K(Z) - the normalized Fo(Z) as a function of core height specified in the COLR.

APPLICABILITY: MODE 1. ;

ACTION:

'With Fo(Z) exceeding its limit:

a. For RA0C operation with Specification 4.2.2.1.2.b not being satisfied or for base load operation with Specification 4.2.2.1.4.b not being satisfied:

(1) Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the i limit within 15 minutes and similarly reduce the Power Range l Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER :

OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent i POWER OPERATION may proceed provided the Overpower AT Trip l setpoints have been reduced at least 1% for each 1% fo(Z) exceeds the limit, and l NILLSTONE - UNIT 3 3/4 2-5 Amendment No. 77, pp. 77, U p, 0007 l

1 i

)

POWER DISTRIBUTION LINITS l

LINITING CONDITION FOR OPERATION (Continued) I l

i( , . ..

(2) Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by item (1) above; THERMAL POWER may then be increased provided Fo(Z) is demonstrated through incore mapping to be within its limits,

b. For RAOC operation with Specification 4.2.2.1.2.c not being satisfied, one of the following actions shall be taken:

(1) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by at least 1% AFD for each percent Fo(Z) exceeds its limits. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these modified limits, or (2) Verify that the requirements of Specification 4.2.2.1.3 for base load operation are satisfied and enter base load operation.

Where it is necessary to calculate the percent that Fo(Z) exceeds the limits for item (1) above, it shall be calculated as the maximum percent over the core height (Z), consistent with Specification 4.2.2.1.2.f, that Fo(Z) exceeds its limit by the ,

following expression: !

-1 x 100 for P > 0. 5

, x K(Z)

-1 x 100 for P s 0. 5 0.5 l

L l c. For base load operation with Specification 4.2.2.1.4.c not being l

satisfied, one of the following actions shall be taken:

(1) Place the core in an equilibrium condition where the limit in 4.2.2.1.4.c is satisfied, and remeasure Fo"(Z), or NILLSTONE - UNIT 3 3/4 2-6 Amendment No. pp, #9 0007

POWER DISTRIBUTION LINITS

I

' LIMITING C0fGITION FOR OPERATION (Continued)

(2) Reduce THERMAL POWER at least 1% for each 1% fo(Z) 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed provided the Oserpower AT Trip Setpoints have been reduced at least 1% for each 1% Fo(Z) exceeds its limit shall be calculated as the maximum percent over the core height (Z), consistent with Specification 4.2.2.1.4.f, by the following expression:

, -1 x 100 for P k APL

p x K(2) l SURVEILLANCE REQUIREMENTS l

4.2.2.1.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.1.2 For RAOC operation, Fo(Z) shall be evaluated to determine if Fo(Z) I is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,

b. Evaluate the computed heat flux hot channel factor by performing both of the following:

(1) Determine the computed heat flux hot channel Factor, Fo"(Z) by increasing the measured Fo(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that Fo"(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e. 0-100% inclusive.

MILLSTONE - UNIT 3 3/4 2-7 Amendment No. pp, pp 77 0007

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) e>- .

'c. ' ' Satisfying the following relationship:

l Ff(Z) S F*"" x K(Z) for P > 0.5 ?

P x W(Z) ,

Ff(Z) s F*"" x K(Z) for P s 0.5 V(2) x 0.5 l

where F#(Z) is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FS" is the Fo limit, K(Z) is the normalized Fo(Z) as a function of core height, P i is the relative THERMAL POWER, and W(Z) is the cycle-dependent function that accounts for power distribution transients encountered i during normal operation. FS", K(Z), and W(Z) are specified in the CORE OPERATING LIMITS REPORT as per Specification 6.9.1.6.

d. Measuring F"o(Z) according to the following schedule:

(1) Upon achieving equilibrium conditions after exceeding by 10% or !

more of RATED THERMAL POWER, the THERMAL POWER at which Fo(Z) ;

was last determined,* or ;

(2) At least once per 31 Effective full Power Days, whichever occurs first. 1 l

e. With the maximum value of Ff(Z)

K(2) i over the core height (Z) increasing since the previous determination of FE(Z), either of the following actions shall be taken:

(1) Increase Fo"(Z) by an appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.2.c, or l

During power escalation at the beginning of each cycle, power level may l be increased until a power level for extended operation has been achieved l and power distribution map outlined. i l MILLSTONE.- UNIT 3 3/4 2-8 Amendment No. pp, pp, #7, Up 0007

, POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

b. Durin APL" gthen basetheload operation,ofif the conditions THERMAL 4.2.2.1.3.a POWER shall is decreased be satisfied beforebelow reentering base load operation. .

4.2.2.1.4 During base load operation Fo(Z) shall be evaluated to determine if Fo(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APL" .

b. Evaluate the computed heat flux hot channel factor by performing both of the following: ,

(1) Determine the computed heat flux hot channel factor, Fo"(Z), by !

increasing the measured Fo"(Z) component of the power ,

distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties, and (2) Verify that Fo"(Z) satisfies the requirements of Specification 3.2.2.1 for all core plane regions, i.e., 0 - 100% inclusive.

c. Satisfying the following relationship:

Fy(Z) $ F*""' x K(Z) for P > APL"*

P x W(Z) ,

where: F"(Z) is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, FS'" !

is the Fo limit, K(Z) is the normalized Fo(Z) as a function of j core height, P is the relative THERMAL POWER, and W(Z)et is the '

cycle-dependent function that accounts for limited power distribution transients encountered during base load operation.

FS'", K'( Z) , and W(Z)3t are specified in the COLR as per Specification 6.9.1.6.

d. Measuring F"(Z) in conjunction with target flux difference determi-nation according to the following schedule:

(1) Prior to entering base load operation after satisfying Sec-l tion 4.2.2.1.3 unless a full core flux map has been taken l in the previous 31 EFPD with the relative thermal power having been maintained above APL" for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and (2) At least once per 31 Effective Full Power Days. 1 L MILLSTONE - UNIT 3 3/4 2-10 Amendment No. pp, pp, 77, 0007 i

, , POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

e. With the maximum value of F;(Z) l K(2) over the core height (Z) increasing since the previous determination of F#(Z), either of the following actions shall be taken:

(1) Increase Fo"(Z) by appropriate factor specified in the COLR and verify that this value satisfies the relationship in Specification 4.2.2.1.4.c, or (2) Fy(Z) shall be measured at least once per 7 Effective Full Power Days until 2 successive maps indicate that the maximum value of Ff(Z) l K(2) over the core height (Z) is not increasing.

f. The limits specified in 4.2.2.1.4.c and 4.2.2.1.4.e are not applicable in the following core plane regions:

(1) Lower core region 0% to 15%, inclusive.

(2) Upper core region 85% to 100%, inclusive.

4.2.2.1.5 When Fo(Z) is measured for reasons other than meeting the require-ments of Specifications 4.2.2.1.2 or 4.2.2.1.4, an overall measured Fo(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.

l l

NILLSTONE - UNIT 3 3/4 2-11 Amendment No. JP, pp. 77, 179, oso?

ADMINISTRATIVE CONTROLS CORE OPERATING LINITS REPORT (Cont.)

f , 2. ,. . Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, 5

3. Control Rod Insertion Limits for Specification 3/4.1.3.6,

4. Axial Flux Difference Limits, target band, and APL" for Specifica-tions 3/4.2.1.1 and 3/4.2.1.2, l

5. Heat Flux Hot Channel Factor, K(z), W(z), APL" , and W(z)g for Specifications 3/4.2.2.1 and 3/4.2.2.2.

i

6. Nuclear Enthalp.y Rise Hot Channel Factor, Power Factor Multiplier l for Specificati ' 3/4.2.3. '

7. ShutdownMarginMonitorminimumcountrateforSpecification3/4.3.5.l 6.9.1.6.b 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 (W Proprietary). (Methodology for Specifications 3.1.1.3-- 1 Moderator Temperature Coefficient, 3.1.3.5--Shutdown Bank Insertion Limit. 3.1.3.6--Control Bank Insertion Limits, 3.2.1--Axial Flux Difference, 3.2.2--Heat Flux Hot Channel Factor, 3.2.3--Nuclear ,

Enthalpy Rise Hot Channel Factor.) l

2. T. M. Anderson to K. Kniel (Chief of Core Performance Branch, NRC), l )

January 31, 1980--

Attachment:

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

3. NUREG-800, Standard Review Plan, U.S. Nuclear Regulatory Commission, l Section 4.3, Nuclear Design, July 1981 Branch Technical Position

-CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC),

Revision 2, July 1981.

4. WCAP-10216-P-A-RIA, " RELAXATION OF CONSTANT AXIAL 0FFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," Rev. 1, February 1994 (W Proprietary). (Methodology for Specifications 3.2.1--Axial Flux Difference [ Relaxed Axial Offset Control] and 3.2.2--Heat Flux Hot Channel Factor [W(z) surveillance requirements for Fo Methodology].)

5. WCAP-9561-P-A, ADD. 3, Rev. 1, "BART A-1: A COMPUTER CODE FOR THE l BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS--SPECIAL REPORT:

, THIMBLE MODELING H ECCS EVALUATION MODEL," July 1986 (W Proprie- i tary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel i Factor.)

6. WCAP-10266-P-A, Addendum 1, Rev. 2-P-A, "THE 1981 VERSION OF THE l 1

WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE," March 1987 (W Proprietary). (Methodology for Specification 3.2.2--Heat Flux Hot Channel Factor.) j l l NILLSTONE - UNIT 3 6-20 Amendment No. 7J, 77, pp, pp, 77, i

( 8** 179,

l ADMINISTRATIVE CONTROLS 5'

CORE OPERATING LIMITS REPORT (Cont.)

7. WCAP-11946, " Safety Evaluation Supporting a More Negative EOL l Moderator Temperature Coefficient Technical Specification for the I Millstone Nuclear Power Station Unit 3," September 1988 (H Proprie-tary).

8. WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL.17 l

USING THE NOTRUMP CODE," August 1985 (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

9. WCAP-10079-P-A, "NOTRUMP - A N0DAL TRANSIENT SMALL BREAK AND GENERAL l NETWORK CODE," August 1985 (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

10. WCAP-12610, " VANTAGE + Fuel Assembly Report," June 1990 (W l Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

11. Letter from V. L. Rooney (USNRC) to J. F. Opeka, " Safety Evaluation for Topical Report, NUSCO-152, Addendum 4, ' Physics Methodology for PWR Reload Design,' TAC No. M91815," July 8, 1995.

l

12. Letter from E. J. Mroczka to the USNRC, " Proposed Changes to Technical Specifications, Cycle 4 Reload Submittal - Boron Dilution Analysis,"

B13678, December 4, 1990.

13. Letter from D. H. Jaffe (USNRC) to E. J. Mroczka, " Issuance of Amendment (TAC No. 77924)," March 11, 1991.

14. Letter from M. H. Brothers to the USNRC, " Proposed Revision to Technical Specification, Shutdown Margin Requirements and Shutdown Margin Monitor Operability for Modes 3, 4, and 5 (PTSCR 3-16-97),

B16447, May 9, 1497.

i

15. Letter from J. W. Anderson (USNRC) to M. L. Bowling (NNECO), " Issuance of Amendment - Millstone Nuclear Power Station, Unit No. 3 (TAC No.

M98699)," October 21, 1998.

l l

MILLSTONE - UNIT 3 6-20a Amendment No. $J. I me

I i

i POWER DISTRIBUTION LIMITS l

'macrt '

3/4.!?.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND I NUCL EAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Margin is maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is more than sufficient to offset any rod bow penalty and transition core penalty. The remaining margin is available for plant design flexibility.

When an Fa 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.

l The heat flux hot channel factor, Fo(Z), is measured periodically using the incore detector system. These measurements are generally taken with the core at or near steady state conditions. Using the measured three dimensional power distributions, it is possible to derive Fo"(Z), a computed value of Fo(Z).

However, because this value represents a steady state condition, it does not include the variations in the value of Fo(Z) that are present during nonequilibrium situations. i l \

To account for these possible variations, the steady state limit of Fo(Z) is adjusted by an elevation dependent factor appropriate to either RA0C or base load operation, W(Z) or W(Z)st, that accounts for the calculated worst case transient conditions. The W(Z) and W(Z)st, factors described above for normal l operation are specified in the COLR per Specification 6.9.1.6. Core monitoring l l and control under nonsteady state conditions are accomplished by operating the l core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion. Evaluation of the steady state Fo(Z) limit is performed in Specification 4.2.2.1.2.b and 4.2.2.1.4.b while evaluation i nonequilibrium limits are performed in Specification 4.2.2.1.2.c and 4.2.2.1.4.c. j When RCS flow rate and F"m are measured, no additional allowances are

! necessary prior to comparison with the limits of the Limiting Condition for !

Operation. Measurement errors of 2.4% for four loop flow and 2.8% for three l N

loop flow for RCS total flow rate and 4% for F m have been allowed for in determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi will be added if venturis are not inspected and cleaned '

at least once for 18 months. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

MILLSTONE - UNIT 3 B 3/4 2-4 Amendment No. U , pp, 0609

Docket No. 50-423

, .. B17609

!?. .-

I l

l Attachment 3 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Heat Flux Hot Channel Factor and Core Operating Limit Report (COLR) Modifications (TSCR 3-27-98) .

I Background and Safety Summary l

l l- .

t. '

['

1 January 1999 i

t ;.

l.-

Nucinar R:gulatory Commission

, , B17609\ Attachment 3\Page 1 Backaround This Technical Specification change is comprised of several unrelated issues which have been grouped together for convenience since they modify similar sections of the TS. The change to TS 3/4.2.2 " Heat Flux Hot Channel Factor-Fo(Z)" is technical in nature, the remaining TS changes are administrative.

With respect to TS 3/4.2.2, a prior TS change addressed removing action requirements which were integrated into the SURVEILLANCE REQUIREMENTS for this TS. In this regard the proposed Technical Specification change was approved by the NRC and issued in License Amendment 99 [2]. NNECO has subsequently identified that the changes incorporated into License Amendment 99 actually altered the intent of the Heat Flux Hot Channel Factor Technical Specification. As such, this proposed change modifies the existing Heat Flux Hot Channel Factor, TS 3/4.2.2, to be in accordance with NRC approved Westinghouse methodologies for Fo(Z) surveillance [3).

NNECO has also identified several administrative changes to Technical Specifications 6.9.1.6.a and 6.9.1.6.b that are not related to the Fo(Z) surveillance changes nor are these changes related to one and other. These TS changes are presented in the following.

The first change involves identifying an additional core operating limit in TS 6.9.1.6.a.

Currently, this TS does not administratively reflect each item that the Technical Specifications reference in the COLR. Specifically, Shutdown Margin Monitor minimum count rates required in TS 3.3.5 " Shutdown Margin Monitor" are listed the COLR, but not identified as a core operating limit in TS 6.9.1.6.a. This TS change will address this issue.

The second administrative change involves updating references to analytical methods used to determine core operating limits in TS 6.9.1.6.b. This change will bring conformity between the references in the Technical Specifications and the cycle specific COLR.

The third administrative change adds a reference in TS 6.9.1.6.b which relates to the NRC Safety Evaluation Report (SER) on Northeast Utility (NU) Topical Report NUSCO-152, Addendum 4. NUSCO-152 describes NU's ability to perform PWR physics calculations for MP3. This SER reference is being added to acknowledge that the NRC has approved NU to perform MP3 reload calculations.

[2] Letter from U. S. Nuclear Regulatory Commission to J. F. Opeka (NNECO)

" Issuance of Amendment (TAC No. M90035)," dated December 29,1994.

[3] WCAP-10216-P-A-R1A, " Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification," February 1994 (W Proprietary).

Nuclerr Regul:. tory Commission

-, B17609\ Attachment 3\Page 2 Th,e. fourth and final administrative change incorporates into TS 6.9.1.6.b references to the shutdown margin analysis methods reviewed and approved by the NRC. Adding these references satisfies a NNECO License Condition which was included as part of License Amendment 164 [1]. These four references provide documentation of the methodology used in developing the Shutdown Margin Requirements and Shutdown Margin Monitor Operability Requirements.

Safety Summary The changes to TS 3/4.2.2 promotes conformity with the NRC approved Westinghouse ;

methodology [3]. Selected SURVEILLANCE REQUIREMENTS are reworded for clarity. l The associated BASES are enhanced to explain the various SURVEILLANCE i REQUIREMENTS. These changes will ensure that the plant is operated within the i design basis accident assumptions. An unreviewed safety question (USO) does not exist because the changes to Specification 3/4.2.2 are in accordance with NRC approved Westinghouse methods, the changes to Specification 6.9.1.6.a and 6.9.1.6.b are administrative only, and no physical plant changes are being made.

l l

I

Docket No. 50-423

<- . B17909

. ll ...

Attachment 4 Millstone Nuclear Power Station, Unit No. 3 Heat Flux Hot Channel Factor and ,

Core Operating Limit Report (COLR) Modifications (TSCR 3-27-98)

Significant Hazards Consideration and Environmental Considerations January 1999

I Nuclear Regulatory Commission i

, B17609\ Attachment 4\Page 1 NNECO has reviewed the proposed revision in accordance with 10 CFR50.92 and has co,nglude,d.that the revision does not involve any Significant Hazards Considerations (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed Technical Specification revision does not involve an SHC !

because the revision would not:

1. . Involve a sigraficant increase in the probability or consequences of an accident previously evaluated. l To determino any potential impact, the proposed changes to the TS are grouped into the following two categories.

i a) Changes to Technical Specification 3/4.2.2 " Heat Flux Hot Channel Factor-Fa(Z)" !

b) Changes that are not related to the Heat Flux Hot Channel Factor TS, and are ,

administrative in nature. These include defining a new core operating limit and !

deleting, re-numbering, updating and adding references to analytical methods used to determine core operating limits in TS 6.9.1.6 " Core Operating Limit Report (COLR).

4 With respect to item 1.a changes related to the Heat Flux Hot Channel Factor, i Fo(Z), impact the initial conditions assumed in the accidents analyzed for MP3. l These initial conditions are power distributions which are consistent with reactor ;

operat!on as defined in the TS. The proposed changes to the Heat Flux Hot Channel Factor TS ensure that proper actions are taken to maintain peaking factors within the limits assumed in the MP3 accident analysis. The proposed changes are consistent with the NRC approved Westinghouse methodology for Fa(Z) surveillance. Changes to the SURVEILLANCE and ACTION statements will not :

change the probability of occurrence of any analyzed accidents. Furthermore, the consequences of analyzed accidents will not change since the power distribution l assumptions will not be challenged by reactor operation allowed by the Technical l Specifications.

With respect to item 1.b the administrative changes to the Technical Specifications do not affect existing or proposed Limiting Conditioas for Operation (LCO) or SURVEILLANCE REQUIREMENTS. Therefore, there is no impact on the design basis accidents.

Thus it is concluded that the proposed revision does not involve a significant increase in the probability or consequences of an accident previously evaluated.

l l

Nuciser Regulatory Commission

, . .B17609\ Attachment 4\Page 2

2. ,greate the possibility of a new or different kind of accident from any accident 4previollsly evaluated, a) Proposed changes to the Heat Flux Hot Channel Factor, TS 3/4.2.2 ensure that proper actions are taken to maintain peaking factors within the limits assumed in the MP3 accident analysis. The proposed changes are consistent with the NRC approved Westinghouse methodology for Fo(Z) surveillance. Maintaining safety analysis assumptions on power distributions cannot be an initiating event for any design basis accidents and will not create the possibility of a different type of i accident. Therefore the changes associated with the Heat Flux Hot Channel '

Factor limiting condition for operation do not represent a new unanalyzed accident.

i b) Since the administrative changes do not affect plant operation, the potential for an unanalyzed accident is not created. No new failure modes are introduced.

Thus, this proposed revision does not create the possibility of a new or different kind I of accident from any previously evaluated.

3. Involve a significant reduction on the margin of safety.

I a) The proposed changes ensure that Fa(Z), will remain within the safety analysis l assumptions. The LCO limits and SURVEILLANCE REQUIREMENTS are not altered. Therefore, the impact on the consequences on the protective boundaries is unchanged. Meeting the intent of the NRC approved Westinghouse methodology for Fo(Z), SURVEILLANCE ensures that power distributions assumed in the accident analysis will not be cha!!enged by reactor operations allowed by the Technical Specifications. Therefore, verification of no change in the margin of safety is encompassed by meeting the power distribution limits assumed in analyzed accidents.

b) Since the proposed changes do not affect the consequences of any accident previously analyzed, there is no reduction in the margin of safety.

Thus it is concluded that the proposed revision does not involve a significant reduction in the margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve a Significant Hazard Consideration.

Nucle r R::gulatory Commission

,, , B17609%ttachment 4\Page 3 ,

Environmental Considerations F. '

t NNECO has reviewed the proposed license amendment against the criteria of l 10CFR51.22 for environmental considerations. The proposed revision does not involve a Significant Hazards Consideration, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO :

concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) -

for categorical exclusion from the requirements for environmental review. '

\

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ML20199L080

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