JSP-541-91, Application for Amend to License NPF-62,changing Tech Spec 3/4.3.6, Control Rod Block Instrumentation to Revise Nominal Trip Setpoint

Application for Amend to License NPF-62,changing Tech Spec 3/4.3.6, Control Rod Block Instrumentation to Revise Nominal Trip Setpoint

ML20077R750
Person / Time
Site:	Clinton
Issue date:	08/16/1991
From:	Jamila Perry ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20077R752	List: JSP-541-91, Application for Amend to License NPF-62,changing Tech Spec 3/4.3.6, Control Rod Block Instrumentation to Revise Nominal Trip Setpoint ML20077R755
References
JSP-0541-91, JSP-541-91, U-601872, NUDOCS 9108260038
Download: ML20077R750 (7)

Text

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CLntan,lL 61777 1ei 717 935 022f3 J. Stephen Perry va e Presconi ILLIN 915 NMR u-6 1872 L4 7-91 (08 - 16) LP DE.100a JSP-0541-91 August 16, 1991 10CFR50.90 Docket No. 50-461 Document control Desk Nuclear Regulatory commission

. Washington, D.C. 20555

Subject:

Clinton Power Station Proposed Amendment of Facility Ooeratina It tgense No. NPP-62 Dear Sirt Pursuant to 10CFR50.90, Illinois Power Company (IP) hereby applies for amendment of Facility Operating License No. HPF-62, Appendix A - Technical Specifications, for Clinton Power Station (CPS). This request consists of proposed changes to Technical Specification 3/4.3.6, " Control Rod Block Instrumentation." For each of these proposed Technical Specification changes, a description, the associated justification (including a Basis For No Significant Hazards consideration), and marked-up copies of pages from the current Technical Specifications are provided in Attachment 2. In addition, an affidavit supporting the facts set.forth in this letter and its attachments is provided in Attachment 1.

IP has reviewed the proposed changes against the criteria of 10CFR51.22 for categorical exclusion from environmental impact considerations. The proposed changes do not involve a significant hazards consideration, or significantly increase the amounts or change the-types of effluents that may be released offsite, nor do they significantly increase individual or cumulative occupational radiation exposures.

Based on the foregoing, IP concludes the proposed changes meet the _ criteria given in 10CFR51.22 (c) (9) for a categorical exclusion-from the requirement for en Environmental Impact Statement.

Sincerely yours, 3S.

(. Perrk[

Vice President

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t Attachments cc liRC Clinton Licensing Project Manager i NRC Resident Inspector, V-690 l NRC Region III, Regional Administrator Illinois Department of Nuclear Safety )

l

l Attachment 1 l to U-601872

, i STATE OF ILLINOIS COUNTY OF DEWITT J. Stephen Perry, being first duly sworn, deposes and says:

That he in Vice President of Illinois Power Company; that the application for amendment of Facility Operating License NPF-62 has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said application and the facts contained therein are true and correct.

DATE: This Ik day of August 1991.

Signed: \b N m-b)StOphenPerbh Gubscribed and sworn to before me this /[//s day of August 1991.

: :::::: ::: :: ( 'b ' v iM F

'0FilCIAL SEAL' Notary Public Linda S. French Notary PuSc, State of Imre My Commisson Enpires 9/1/92  ;

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- _ _ _ .~____ . = _ _ _ . _ . . . _ _ _ . _ _ . . _ _ . _ . _

. t Attachment 2  !

. to U 001872

. L9 90 005 t Page 1 of 7 [

Bachround Clinton Power Station (CPS) Technical Specification Table 3.3.6 2,  ;

" Control Rod Block Instrumentation Setpoints," currently identifies that  !

the Trip Setpoints and' Allowable Values (in terrs of % of RATED TilERNAL  ;

POWER) associated with the Rod Pattern Control System (RPCS) Low Power >

Setpoint (item 1.a) and Rod Withdrawal Limiter 'RVL) 111gh Power Setpoint f (Item 1.b) are to be determined during the startup test program.  !

Accordingly, the data necessary to establish these setpoints was  !

obtained during the startup test program, and the values determined for  !

the Trip Setpoints and Allowable Values were submitted to the NRC for J their review on February 5, 1988 (reference Illinois Power Company (IP) i letter U-601118). licwever, prior to NRG approval of IP's rebruary 5, [

1998 amendment request, IP identified that a modification to the main  !

turisine control system would be performed during-the second refueling outage. This modification would convert the high pressure turbine  ;

control valve control logic from the full arc steam admission mode to i the partial-arc steam admission mode. It was recognized that this modification could result in a change to the RPCS Low Power Setpoint (LPSP) and 111gh Power Setpoint (llPSP). As a result, IP requested ,

withdrawal of that portion of the February 5, 1988 amendment requen- l until completion of the turbine control modification and the l determination of new setpoints for the LPSP and llPSP. l Modification of the main turbine control system was completed during the second refueling outage at CPS. Results of testing conducted during the subsequent reactor startup were utilized to either confirm that the setpoints for the RPCS LPSP and HPSP which were determined during the  ;

initial startup test program continue to be appropriate or to determine  :

new setpoints. This request therefore involves proposed changes to CPS [

Technical Specification Table 3.3.6 2 to incorporate the revised RPCS j LPSP and llPSP. i i

Descrintion of Proposed Changes  ;

In accordance with 10CFR50.90,-the following proposed changes to [

Technical Specification 3/4.3.6, " Control Rod Block In=trumentation,"  ;

are being proposed: l

-1) The nominal trip setpoint under the Trip Setpoint column for the RPCS Low Power Setpoint (Table 3.3.6 2, Item 1.a) is being revised from "(*)% of RATED YllERHAL POWER" to "138.0 1 2.3 psig*" and the ,

associated Allowable Value is being revised from "(*)% of RATED ,

TilERMAL POWER" to "2; 115.0 psig, s 175.0 psig*." g

2) The nominal trip setpoint under the Trip Setpoint column for the l RWL liigh Power Setpoint (Table 3.3.6-2, Item 1.b) is being revised ,

from "(*)% of RATED TilERMAL POWER" to "5 361,6 psig*" and the i associated Allowable Value is being revised from "(*)% of RATED  !

TilERMAL POWER" to "s 400 psig* . "  ;

'i

3) Footnote."*" is being revised to delete the statement that the  !

setpoints are to be determined during the startup test program and -l to add a statement that these setpoints are turbine first stage {

pressure values. .

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Attachment 2

.. to U 601872 LS-90 005 Page 2 of 7 Justification for Proposed Changes As stated in the Bases for Technical Specification 3/4.3.6, the control rod block functions are provided consistent with the requirements of the Technical Specifications in Section 3/4.1.4, " Control Rod Program Controls," and Section 3/4.2, " Power Distribution Limits," to provide a backup to CPS administrative controls on control rod manipulations. The administrative controls and control rod block functions ensure that fuel thermal limits and initial conditions of design basis accidents are preserved during control rod movements. The primary control rod block functions are the Rod Pattern Control System and the Rod Withdrawal Limiter. Each of these two functions are more fully described below.

Epilattern Control System (RPCJ1 As described in Updated Safety Analysis Report (USAR) Section 7.6.1.7, the purpose of the RPCS is to reduce the consequences of a postulated Control Rod Drop Accident (CRDA) by restricting the patterns of control rods that can be established to predetermined sets when operating below the LPSP. As discussed in USAR Section 15.4.9, the RPCS ensures that

.the fuel design peak enthalpy of 280 calories / gram is not exceeded during a postulated CRDA initiated from any plant operating condition.

The CRDA analyzed in USAR Section 15.4.9 is the result of a postulated event in which the highest reactivity worth control rod within the constrainte of the RPCS drops from the fully inserted or an intermediate position in the reactor core. In accordance with this CRDA analysis, the highest worth control rod, initially in a fully inserted or intermediate position, is assumed to become decoupled from its drive mechanism. The drive mechanism is then withdrawn; however, the decoupled control rod is assumed to be stuck in place. At a later moment, the control rod is assumed to suddenly fall free and drop to the position of the control rod drive mechanism. This results in the removal of large negative reactivity from the core and results in a l localized power excursion. The velocity of the dropped control rod (and hence, the rate of negative reactivity removal) is limited by the j velocity limiter on the bottom of the control rod blade. The reactor is

assumed to scram as a result of high neutron flux sensed by the Average Power Range Monitors (APRMs) (120% of RATED THERMAL POWER neutron flux).

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 level has been determined to be 20% of RATED 1HERMAL POWER. Therefore, no restrictions on the control rod pattern are required above 20% of RATED THERMAL POWER.

Rod Withdrawal Limiter (RWL)

As described in USAR Section 7.6.1,7, the RWL limits continuous control rod withdrawal to prevent excessive change in the heat flux rate in the event of a control rod withdrawal error (RWE). This event is assumed to occur as the result of an operator error in which a single 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 of minimal concern below 35% of RATED THERMAL POWER, even without continuous control rod withdrawal restraints. At power levels

h t

Attachment 2 )

p to U-601872 [

LS 90 005 .

Page 3 of 7

• above the LPSP, the RWL restricts continuous control rod withdrawal to l four notches (2 feet) until the llPSP is reached (70% of RATED TilERMAL POWER). Above the llPSP, concinuous control rod withdrawal is limited to  !

two notches (1 foot), in accordance with the RWE analysis, the llPSP was j chosen to provide more restrictive restraints on continuous control rod withdrawal at reactor power levels above 70n of RATED TilERMAL POWER. i f

Pronosed Setnoints }

t The reactor thermal power inputs to the RPCS and RWL are derived from main turbine first stage pressure instrumentation. Initial values for i the LPSP and llPSP vere 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 1cvels  ;

from st.veral reactor heat balances. The turbine manufacturer's measured  :

steam flow vs. turbine first stage pressure relationship was also  !

utilized. The combination of these two relationships yielded the [

predicted relationship between reactor thermal power and turbine first stage pressure. Data obtained during the initial startup test program  !

confirmed the validity of the predicted relationship.

Following modification of the main turbine control systetr during m  !

second refueling outage at CPS, turbine first stage pressure }'

. measurements were taken at numerous power levels during the subsequent reactor startup. The test results sh.;wed that the actual turbine first  !

stage pressure is slightly higher than the originally predicted value p for all reactor power levels. The proposed setpoints are based on the more connervative result obtained from either the measured turbine first  !

stage pressure or the originally predicted turbine first stage pressure  ;

(with margin far calculation uncertainties). Determination of these  ;

limits in this manner ensured that the most conservative value was l established for the LPSP and llPSP instrument setpoints. Since the RPCS }

LPSP supports safety functions for both the RPCS and the RWL, a range  !

has been specified for both the Allowabic Value and Trip Setpoint such  !

that the LPSP will be reached between 20 and 35% of RATED TilERMAL POWER. l The RPCS LPSP lower Allowable Value limit is based on an extrapolation [

of the test results at 20% of RATED TilERMAL POWER; the upper Allowable l Value limit was based on the originally predicted turbine first stage j pressure (with margin for calculation uncertainties) at 35% of RATED J t TilERMAL POWER. The upper Allowable Value limit of the RWL llPSP was l based on the originally predicted turbine first stage pressure (with [

margin for calculation uncertainties) at 70% of RATED TilERMAL POWER. l Based on the above-described Allowable Values, the Trip Setpoints for l the LPSP and llPSP vere established in accordance with the methodology of ,

Regulatory Guide 1.105, ' Instrument Setpoints." Accordingly, each  !

setpoint is supported by a calculation that includes instrument [

accuracy, calibration uncertaintics, and drift allowance during the i 18 month calibration interval. Thus,_these proposed setpoints will l ensure that plant operation remains within the assumptions of the CRDA  ;

and RWE analyses.

r 1

Attachment 2 to U.601872

, 1S.90 005 page 4 of 7  !

Ensis for No Significant Hazards Conalderation e In accordance with 10CFR50.92, a proposed change to the Operating 1.icense (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 evaluat. or (3) involve a significant reduction in a margin of safety. The prt,2 sed changes are evaluated against each of these criteria below, i

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-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 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 staSe 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 second refueling outage. Therefore, the proposed cuanges 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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