ML20077R755

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Proposed Tech Spec Table 3.3.6-2 Re Control Rod Block Instrumentation Setpoints
ML20077R755
Person / Time
Site: Clinton Constellation icon.png
Issue date: 08/16/1991
From:
ILLINOIS POWER CO.
To:
Shared Package
ML20077R752 List:
References
NUDOCS 9108260039
Download: ML20077R755 (11)
v • d • e


Text

_ _ _ _ _ _

TABLE 3.3.6-2 -

f0NTROL RCD BLOCK INSTRUMENTATION SETPOINTS ~-

p TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 8 1. ROD PATTERN CONTROL SYSTEM

$ se a. Low Power Setpoint -f*)%-of-RATED-THERMAL-POWER-

b. -(*)%-o f-RATED-THERMAL-POWER- }

ou 8 2>t RWL High Power Setpoint -t")% of-RATED-THERMAL-POWER-

-(*)%mf-RATED-THERMAL-POWER f

@h nc n

2. APRM d ,
a. Flow Biased Neutron Flux  :

33 oc

~ - Upscale MCPIAC6 w;4h In g +  !

g 1) During two recirculation ewe

c.>

loop operation:

~

a) Flow Biased e 0.66W + 58%** with a maximum < 0.66W + 61%** with a maximum of of b) 111gh Flow Clamped i 108.0% of RATED THEPJiAL POWER 1 110.0% of RATED THERMAL POWER

$ 2) During single recirculation loop operation:

E a) Flow Biased < 0.66(W-AW) + 42%**

'

  • i 0.66(W-aW) + 45%** i b) High Flow Clamped Sot required OPERABLE Not required OPERABLE  !
b. Inoperative NA NA  !

! c. Downscale > 5% of RATED THERMAL POWER

- > 3% of RATED THERMAL POWER  :

d. Neutron Flux - Upscale -

i

, Startup $ 12% of RATED T11ERMAL POWER $ 14% of RATED THERMAL POWER g 3. SOURCE RANGE MONITORS

$ a. Detector not full in NA NA

[ b. Upscale 5 1 x 105 cps 5 1.6 x 105 cps o c. Inoperative NA NA

{

i

[ d. Downscale 1 3 cps 1 1.8 cps

- 4. INTERMEDIATE RANGE MONITORS E a. Detector not full in NA NA

- ~ , ,

g '; o "> .

b. Upscale $ 108/125 division of full scale < 110/125 division of full scalee g?g
c. Inoperative NA NA
d. Downscale w4gg  ;

3 5/125 division of full scale 3 3/125 division of full scale o, 3 g g r

~n M kl i  !

.. W '

Attachment 2

= I' to U 601872 4- h' LS.90 005 d Page 6 of 7 t

ltuart for page 3/4 3 66 Trin Setnoid Allowable Value i I

a. Low Power Setpoint 138.0 1 2.3 psig* 2 115.0 psig,$ 175.0 psig* f, i- b. RWL liigh Power s 361.6 psig* $ 400 psig*

! Setpoint 4

i

! s i-l 1 i

),

I L

l l

~_ __- _ __- - _. - _ , _ _ _ - _ _ _ _ . . , _ . . . _ . . _ _ _ _ . _ _ _ _ . . . _ _. b

A t t a c hinen t 2 to U 601872 [

1.S-90-005 '

., l'a g e 7 of 7  ;

TABLE 3.3.6-2 (Continued)  !

CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS i i

TABLE NOTATIONS

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. v_ - ,

The Average Power Range Monitor rod block function is varied as a function ,

of recirculation loop flow (W). The trip setting of this function must be -

maintained in accordance with note (a) of Table 2.2.1-1. '

i

  1. Instrument zero is 758' 5" ms1. '
    1. Instiument zero is 758' 4 1/2" ms1.

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q* These are. +va,nc em

%3c. ytessucc. va a.es ,

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CLINTON - UNIT 1 3/4 3-67 Amendment No.18

. ,,,, ALM ^QN g %010g' {

( () f M M 6 fC* r0 , 0 At.t achment 2 Ti%Cnaa.Eq.5 h JM(,hte '

llS.f;$""

Page 1 of 7 h<faC.ghu wyfy.f ,L g (

basky.rsuttd -

p j. 0 Qy Clinton Power Station (CPS) Technical Specification Table 3.3.6 2,

" Control Rod Block Instrumentstion Setpoints," currently identifies that the Trip Setpoints and A11swable Values (in terms of % of RATED TilERMAL POWER) associated with the Rod Pattern Control System (RPCS) Low Power Setpoint. (Item 1.a) and Rod Withdrawal Limit er (RWL) liigh Power Setpoint.

(Itcm 1.b) are to be determined during the startup test pro 6 ram.

Accordingly, the data necessary to establish these setpoints, was cbtained during the startup tert program, and the values determined iDr the Trip Setpoints and Allowable Values were subrnitted to the !GC for their review on February 5, 1988 (reference Illinois Power Company (IP) letter U 601118). Ilowever, prior to !;RC approval of IP's February $.

1988 amendment request, IP identified that a modification to the main

] turbino control system would be performed durint, the second refueling

\ outage. This modificationwoggconvert the high pregyup turbine gg. M'4 control valvisequenoing f rom [ lull arc steam admission g tf78dal-arc

___ s tenm _oslainlmp It was recognized that this modification could result g' in a change to the RPCS Low Power Sctpoint (LPSP) and liigh Powe Setpoint (PPSP). As a result, IP requested withdrawal of that portion of the February 5, 1988 amendment request until completion of the turbine control modification and the determination of new setpoints for the LPSP and !! PSP.

giht(

liodification of the main turbine control system was completed during the second refueling outage. Results of testinP/ conducted during the J.uluttqw.pt_ reactor startup were utilized to[ confirm the setpoints for m

'\ the RPCS LPSP and IIPSP1, determined during the startup test program (d--r -tff) )

W g [di determine the appropriate setpoints. This request involves propcse N changes to CPS Technical Specification Tcbic 3.3.6-2 to incorporate the revised RPCS L?SP and llPSP.

Description of Proposed Char &g i

l In accordance with 10CFR50.90, the following proposed changes to l Technical Specification 3/4.3.6, " Control Rod Block Instrumentation,"

l are being proposed:

1) The RPCS Low Power Setpoint (Table 3.3.6 2, Item 1.a) Trip

) Setpoint is being revised from "(*)% of RATED TilERMAL POWER" to "138.0 1 2.3 psig*" and Allowable Value is being revised from l

"(*)% of RATED TilERMAL POWER" to ";> 115.0 psig, g 17 g,*

,O l RWl kp

2) The RPOS ligh Power Setpoint (Tabic 3.3.6 2, Item 1.b)

Setpoint is being revised from "(*)% of RATED TilERMAL POWER" to "s 361.6 psig*" and Allowable Value is being revised from "(*)% of RATED TilERMAL POWER" to "g 400 psig*."

3) Footnote "*" is being revised to delete the statement that the g3 setpoints are to be determined during the startup test program and 4 0/

add a statement that these setpoints are the values of turbine V first stage pressure.

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

Attnehment >

. to U.60187?

LS.90 00's Page 2 of /

htEi1LLealle1 LIE _l'Intncil Chanceti As stated in the Bases f or this Technical Specification, the control rod block functions are provided consistent with the requirenients of the Technical Specifications in Section 3/4.1.4 " Control Rod Program controls," and Section 3/4.2 " Power Distribution Limits," to provido a backup to CPS administrat ive controln on cont rol rod inanipulations.

These controla ensure that fuel thermal limits and initial conditions of design basis accidents are preserved during control rod movements. Two of these control rod block funct ions. consist of the Rod Pattern Cont.rol System and the Rod Withdrawal Limiter. Each of these two functions are more fully described bebw.

Pod PALLtrn C.entroLjiyftem (RPCSl An described in Updated Safety Analysis Report (USAR) Section 7.6.1.'

the purpose of the RPCS is to reduce the consequences of a postuinted control Rod Drop Accident (CRDA) by restricting the patterns of control rods that can be established to predeterruined sets below the LPSP. As discussed in USAR Section 15.4.9, the RPCS ensures that the fuel design peak enthalpy of 280 fcal /gm. is not exceeded during postulated RDA initiated from any p'lant operatis)g condition.

-CalMicMfc:Wo m The CRDA analyzed in USAR SectionMS.4.9 is the result of a postulated event in which a highest. reactivity worth control rod within the constraints of the RPCS drop- from the fully inserted or an intermediate position in t.he reactor corr in accordance with this CRDA analysis, the highest worth control roa asstaed to becomo deccupled from its drive mechanism. The drive mechanism is t. hen withdrawn; however, .;e decoupled control rod i: assu:acd to be stuck in place. At a later moment, the cont.rol rod is assumed to suddenly fall free and drop to t.he position of the control rod drive mechanism. This results in the removal of large negative reactivity from the core and results in a localized power excursion. The velocity of the dropped cont.rol rod (and hence, the rate of negative reactivity removal) is limited by the velocity limiter on the bottom of the control rod blade. The reactor is assumed to scram as a result of high neutron flux effected by Average Power Range Monitors (APRMs) (120% power neutron flux).

l The LPSP has been chosen to correspond to the reactor power level at l

which a limiting CRDA can no longer occur. In accordance with the CRDA l analysis, this power level has been determined to be 20% of RATED l

TilERMAL POWER. Therefore, no restrictions on the control rod pattern are required above 20% of RATED TIIERMAL POWER.

Papd WithdrawA.l Limiter (RWL)

As described in USAR Section 7.6.1.7, the# RWL limits continuous control rod vithdrawal to prevent excessive cha go in the hect flux rate in the event of a control rod rithdrawal error This event is assumed to occur as the result of an operator error in which a singic control rod or gang of control rods is withdrawn continuously until the RWL blocks further withdrawal. The analysis for this event demonstrates that this event is

Attachment 2

.r to U 601872 s LS 90 005  :

Page 3 of 7  ;

i of ininimal concern below 35% of RATED TilERMAL POWER, even without  :

cont.inuous control rod withdrawal- restraints. Above the LPSP, t.he RWL restricts continuous control rod withdrawal to four notches (2 feet) until the llPSP is reached (70% of RATED TilERMAL POWER). Above the llPSP,- [

continuous control rod withdrawal is limited to two notches (1 foot). .

In accordance with the RWE analysis, the llPSP was chosen to provide more restrictive restraints on continuous control rod withdrawal at reactor j power levels above 70% of RATED 'NIERMAL POWER.  !

3-Propor.ed Serpointg The reactor thermal power inputs to the RPCS and RWL are derived from main turbine first stage pressure instrumentation. The initial LPSP and IIPSP setpoints were established based upon a predicted relationship ,

between reactor chermal power and turbine first stage pressure (with t margin for calculation uncertainties). This predicted relationship was-  ;

obtained from stearn flow dSta determined at various reactor power IcVels i from several reactor heat balances. The turbine inanufacturer's measured f steam flow vs. turbine first stage pressure relationship was-also [

utilized. The combination of these two re12.tionships yielded the predicted relationship between reactor thermal-power and turbine first i stage pressure. Data obtained furing the startup test program confirmed }

the validity of the predicted relationship. j Following modification of the main turbine control system during the second refueling _ outage, turbine first stage pressure measurements were taken at numerous power levels during the subsequent reactor startup. .

The test results showed that the actual turbine-first stage pressure is generally higher than the originally predicted value for a given reactor power level. The proposed setpoints are based on the more conservative r of the incasured +urbine first stage pressure or the originally predicted  ;

turbine _ first stage pressure (with margin for calculation i uncertainties). Determination of thes limits in this manner ensures j that the most conservative value is used in determining the LPSP and  ;

ilPSP instrument setpoints. Since the RPCS LPSP supports safety [

functions for both the RPCS and the RWL, a range has been specified for j both the Allowable Value and Trip Setpoint stc t that the LPSP will bc

)

reached between 20 and 35% of RATED M KHAL POWER. The RPCS LPSP lower i Allowable Value limit is based W an extrapolation of-the test results [

at 20% of RATED TilERMAL PO ERjand the upper Allowable Valne limit of the  !

RPCS LPSP was based on the predicted turbine first stage presrure (with

~

i margin for calculation uncertIinties) at 35% of RATED TilERMAL POWER. {

The upper Allowable Value limit of the RWL llPSP was based on the }

predicted turbine first stage pressure-(with margin for calculation j uncertainties) at-70% of RATED T11ERMAL POWER,  ;

f Based or, these Allowable Values, the Trip Setpoints for the LPSP and itPSP were established in accordance with the methodology of Regulatory ,

Cuide_1.105, ' Instrument Setpoints." Accordingly, each setpoint is j

- supported by a calculatic.n that includes instrument accuracy, i calibration uncertainties, and drift allowance during the 18 month  !

calibration _ interval. -Thus, these proposed setpoints will ensure that

}

P

_ . _ _ ___ _____._____.__-._a

  • l Att achinent 2 e to U 601872

',- LS90-00S page 4 of 7 plant operation remains within the assumptions of the CI:DA and RWE l analyses.

Basis for No Sir.nificant Hara.r.ds Consideration )

In accordance with 10CFR$0.92, a proposed change to the Operating l License (Technical Specifications) involves no significant hazards l consierations if operation of the facility in accordance with the propo' .1 change would not:- (1) involve a significant increase in the probability or consequencet of any accident previsusly evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a signific. ant reduction in a margin of safety. The proposed changes are evaluated against each of ,

these criteria below.-

1) The proposed changes are consistent with the Control Rod Drop Accident-(CRDA) analysis presented in Updated Safety Analysis Report (USAR) Section 15.4.9 and the Rod Withdrawal Error (RWE) 1 analysis-presented in USAR Section 15.4.2. Additionally, the proposed setpoints have been de. eloped based upon a-conservative relationship between turbine first stage pressure and reactor power level using a setpoint methodology which takes into account appropriate instrument uncertainties in accordance with Regulatory Guide 1.105. In addition. this relationship has been confirmed by rucasurements taken during plant startup from the second refueling '

outage. Further,.the proposed changes do not result in any change to plant equipment or operation. Therefore, the proposed changes do not result in a significant increase in the probability or consequences of any accident previously evaluated.

2) The proposed changes do not result in any changes to plant equipment or operation. As a result, no new failure modes are  :

introduced. . The proposed changes are clearly within the limits of plant operation as described in USAR Section 7.6.1.7. Therefore,

  • the proposed changes cannot create the possibility of a new or different kind of-accident from any accident previously evaluated.
3) The proposed changes are consistent with the CRDA and RWE analyses presented in the USAR and limit plant operation to those conditions assumed in the CRDA and RWE analyses. Additionally, the proposed setpoints have been developed based upon a conservative relationship between turbine first stage pressure and reactor power level using a setpoint methodology which takes into account appropriate instrument uncertainties in accordance with Regulatory Guide _1.105. In addition, this relationship has been confirmed by measurements taken during plant startup from the accond refueling outege. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

-Based upon the foregoing, Ip has concluded that these proposed changes do not involve a significant hazards consideration.

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P Attachment 2 to U 601872 LS 90 005 Page 1 of 7 hpf k rround Clinton Power Station (CPS) Technical Specification Table 3.3.6 2,

" Control Rod Illock Instrumentation Sotpoints," currently identifies that the Trip Setpoints and A110wabic Values (in terms of % of RATED THERMAL POWER) associated with the Rod Pattern Control Systern (RPCS) low Power Setpoint (Item 1.a) and Rod Withdrawal Limiter (RVL) liigh Power Setpoint (Item 1.b) are to F ictermined during the startup test _ program.

Accordingly, the necessary to establish these setpoints was obtained during the .cartup test program, and the values determined for the Trip Setpoints and Allowable Values ve.e subinitted to the NRC for their review on February 5,1988 (reference Illinois Power Company (IP) letter U-602118). llowever, prior to NRC approval of IP's February 5, 1988 amendrw a request, IP identified that a modification to the main turbine control system would be performed during the second refueling i.

outage. This modification would convert the high pressure turbine control valve control logic from the full arc steam admission mode to the partial arc stean admission mode. It was recognized that this modification could result in a change to the RPCS Low Power Setpoint (LPSP) and llir,h Power Setpoint (llPSP) . As a result., IP requested withdrawal of that portion <,f the February 5,1988 amendment request until completion of the turbine control snodification and the determination of new setpoints for the LPSP and itPSP.

gt CP6, a h.nMieh Hodification of t.he mai[ turbine control system was compic :e d during the second refueling outar,e. Results of testing conducted dur .r g the lh4lM)d,40 g _

subsequent reactor startup were utilized to either confirm It he setpoints opf J for the RPCS LPSP and llPSP which were determined during th$jsta x program)r to determine 4La-appsops4eee setpoints. This reque t Sherefm(c, involves proposed chang a to CPS Technical Specification Table SL?

to incorporate the rev sed HPCS LPSP and llPSP.

Description of Proposec mannes I

L In accordan n with 10CFR50.90, the following proposed changes to Technical Specification 3/4.3.6 " Control Rod Block Instrumentation,"

are being prononed MnominaA MP 'Seapo',nt Under 4hc Tr*p '$dpe'.nt Colornn br N

1) The f:y[RPCS Lowrevised Powerfrom Setpoint
  • (*a of(Table RATED3.3.6 2, item 1.a) fdp-i2 is being TilERMAL POWER" to l "138.0 1 2.3 psig*" an Allowable Value is being revised from

"(*)% of RATED THERHA '0WER" to "2 115.0 psig, s 175.0 psig* "

@ C 0 9.cc.indeA

2) - ThelRWL liigh Power Setpoint (Tabic 3.3.6 2. Item 1.b) THT Rgeint is being. revised from "(*)% of RATED TilERMAL POWER" to l "5 361.6 psig*" and, Allowable Value is being revised from "(*)% of RATED Ti!ERMAL POWER' -to "s 400 psig*."

d- +He OSsocicdeb

3) Footnote '*" is b'eing revised to delete the statement that the setpoints are to be determined during the startup test program and to add a statement that these setpoints are ahe- + ' ef- turbine first stage pressurey yg k g i

i

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

Attachment, 2 ,

8 t o U - (,0187 2

's LS.90 005  !

Page 2 of 7 [

Juti1Unlion for Proposed Clwnnes ,3 As stated in the Bases for thi[ Technical Specificatf or the control rod block functions are provided consistent with the requitements of the Technical Specifications in Section 3/4.1.4, " Control Rod Program '

controls," and Section 3/4.2, " Power Distribution Limit.s," to provide a '

m brPup to CPS administrative controls on control rod manipulations.

u a, s . g ensure that fuel thermal limits and initial conditions of i

_g g gtfg t Thc design basis accidents are preserved during control rod movements. -Tw.'Ttv:. p6nm>  !

g gg3 ord control rod block iunctionsl nonotee-o+ the Rod Pat torn Control I

p o) (ad e&-4.he+

System and the Rod Withdrawal Limite r. Each of these two functions are t

gcg more fully described below. [g g6 i Rod Pnttern Control System (RPCS1 gge offk As described in Updated Safety Analysis Report (USAR tion 7.6.1.7, the purpose of the RPCS is to reduce the consequene )a of a postulated Control Rod Drop Accident (CRDA) by restricting th patterns of control [I rods that can be established to predetermined sets below the LPSP. As discussed in USAR Section 1$.4.9, the RPCS ensures that the fuel design r peak enthalpy of 280 calories / gram is not exceeded during po ulated i CRDA initiated from any plant operating condition. A, i I

4bt, The CRDA analyz d in USAR Section 15.4.9 is the result of a postulated event in which highest reactivi orth control rod within the constraints of the RPCS drops from ic fully inserted or an intermediate '

I position in the renctor enen In accordance with this CRDA analysis, the highest worth control rodMs assumed to become decoupled from its I gU iri drive mechanism. The drive mechanism is then withdrawn; however, the t

O- og decoupled control rod is assumed to be stuck in place. At a later d I moment, the control rod is assumed to suddenly fall free and drop to tho

\p6f.M

'ggde(

g position of the control rod drive mechanism. This results in the

$0% removal of large negative reactivity from the core and results in a 3 localized power excursion. The velocity of the dropped control rod (and l

~

hence, the rate of negative reactivity removal) is limited by the l velocity limiter on the bottom of the control rod blade. Th gactor is assumed to scram as a result of high neutron flux #^Aby verage  ;

Power Range Monitors (APRMs) (120% powar neutron flux).

Les RMtbTHERNM.% '

The LPSP has been chosen to correspond to the reactor power level at which a limiting CRDA can no longer occur. In accordance with the CRDA analysis, this power icvel has been determined to be 20% of RATED THERMAL POWER. Therefore, no restrictions on the control rod pattern are required above 20% of RATED THERMAL POWER.

[

Rod Withdrawal Limiter (RWD l

i 1

' As described in USAR Section 7.6,1.7, the RVL limits continuous control rod withdrawal to prevent excessive change in the heat flux rate in the l

[

event of a control rod withdrawal error (RUE) . This event is assumed to j occur as the result of an operator error in which a single control rod "

or gang of control rods is withdrawn continuously until the RVL blocks further withdrawal. The analysis for this event demonstrates that this l

Attachment 2 to U 001872 LS.90-00$

/N JCWT' Page 3 of 7 Me6 event is of minimal concern below 35% of RATED TilEPJiAL POWER, even without continuous control rod withdrawal restraints, povotheLPSP, the RWL restricts continuous control rod withdrawal to four notches (2 feet) until the llPSP is rouhed (70% of RATED TilERMAL POWER). Above the llPSP, continuous control rod withdrawal is limited to two notches (1 foot). In accordance with the RWE analysis, the llPSP was chosen to provide rnore restrictive restrainto on continuous control rod withdrawal at reactor power 1cvels above 70% of RATED TilEPJtAL POWER.

Pronopd Setnointa gghg The reactor thermal power inputs to the RPCS and RVL arejerive from main turbine first stage pressure instrumentation. 4he /nitial LPSP and llPSP wrointe were established based upon a predicted relationship between reactor thermal power and turbine first stage pressure (with margin for calculation uncertainties). This predicted relationship was obtained from steam flow data determined at various reactor power levels from several reactor heat balances, The turbine manuf acturer's incasured steam flow vs. turbine first stage pressure relationship was also utilized. The combination of these two relationships yielded the predicted relationship between reactor cnermal power and turbine first stage pressure. Data obtained during theptartup test program confirined the validity of the predicted relationship. L ggg sod C Following modification c f the% main turbino control system during the second refueling outagg turbine first stage pressure measureroents were y taken at numerous power levels during the subsequent reactor startup 4/l The test results showed that the actual turbine first stage press is e*imelPj higher than the originally predicted value for e-g4*en, eactor OS power levels. The proposed setpoints are based on the snore conservative (cfvH obhW,

[

=4= the measured turbine first stage pressure or the originally predicted &ctn 6Whef turbine first stage pressure (with snargin for calculation .

uncertainties). Determination of these limits in this manner ensure /

that the most conservative value 44 used-in datermin4cc sthe ISSP and i

llPSP instrument setpoints. Since the RPCS ISSP supports safety #kQ)Mhc8g functions for both the RPCS and the RWL, a range has been specified for both the Allowable Value and Trip Setpoint such that the LPSP will be reached between 20 and 35% of RATED TilERMAL POWER. The RPCS LPSP lower Al n@ "h ilmit is based on an extrapolation of the test results i at 20% of RATED TilEP s '0VER; the upper Allowable Value litnit e&-4.he M cS 14GP was based on the redicted turbine first stage pressure (with yg.bg margin for calculation uncertainties) at 356 of RATED TilERMAL POWER.

{ The upper Allowable Value lirait of the RWL llPSP was based on the

> predicted turbine first stage pressure (with margin for calculation uncertainties) at 70% of RATED TilERHAL POWER.

&.O4C.~ OVLdbeh Based on thes lowable Values, the Trip Setpoints for the LPSP and f itPSP were' established in accordance with the trethodology of Regulatory Guide 1.105, " Instrument Setpoints." Accordingly, each setpoint is supported by a calculation that includes instrument accuracy, calibration uncertainties, and drift allowance during the 18. month calibration interval. Thus, these proposed setpoints will ensure that

p Attachment 2 to U 601872 LS 90 005 page 4 of 7 plant operation remains within the assumptions of the CRDA and RWE analyses.

Bonin for No Significant Hazards Consideratica i

In accordance with 10CFR50.92, a prenosed change to the Operating License (Technical Specifications) involves no significant hazards considerations if operation of the facility in accordance with the proposed change would not: (1) involve a significant increase in the probability or consequences of any accident previously evaluated, or (2) +

create the possibility of a new or different kind of accident from any accident previously ovaluated, or (3) involve a significant reduction in a margin of safety. The proposed changes are evaluated against each of -

-these criteria below.

1) The proposed changes are consistent with the Control Rod Drop Accident (CRDA) analysis presented in Updated Safety Analysis Report (USAR) Section 15.4.9 and the Rod Withdrawal Error (RWE) analysis presented in USAR Section 15.4.2. Additionally, the proposed setpoints have been developed based upon a conservative relationship between turbine first stage pressure and reactor power level using a setpoint methodology which takes into account c appropriate instrument uncertainties in accordance with Regulatory  ;

Guide 1.105. In addition, this relationship has been confirmed by measurements taken during plant startup irca the second refueling outage. Further, the proposed changes do not result in any change to plant. equipment or operation. Therefore, the proposed changes do not result in a significant increase in the probability or consequences of any accident previously evaluated.

2) The proposed changes do not result in any changes to plant equipment or operation. As a result, no new failure modes are introduced. The proposed changes are clearly within the limits of plant operation as described in USAR Section 7.6.1.7. Therefore, .

the proposed changes cannot create the possibility of a new or .

different kind of accident from any accident previously evaluated. l

'3) .The proposed changes are consistent with the CRDA and RWE analyses presented in the USAR and limit plant operation to those [

conditions assumed in the CRDA and RWE analyses. Additionally, i the proposed setpoints have been developed based upon a r conservative relationship between turbine first stage pressure and [

reactor power level usin6 a setpoint methodology which takes into L y - account appropriato instrument uncertainties in accordance with Regulatory Guide 1.105. In addition, this relationship has been confirmed by measurements taken during plant startup from the second refueling outage. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the foregoing, Ip has concluded that these proposed changes do not involve a significant hazards consideration.

i 5

b

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