Text

Application for Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,revising Twelve RTS & ESFAS Allowable Values Contained in ITS Table 3.3.1-1 & Table 3.3.2-1
	ML20198N335
Person / Time
Site:	Byron, Braidwood
Issue date:	12/29/1998
From:	Krich R; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20198N339	List: ML20198N335; ML20198N347;
References
	NUDOCS 9901060100
	Download: ML20198N335 (14)
	v • d • e

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'% Commonwealth kliwn Company

.. . 1400 Opus Place M Downers Grove, IL 60515-5701 December 29,1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Application for Amendment to Appendix A, Technical Specifications, to Facility Operating Licenses Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Revision to Twelve Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) Allowable Values Pursuant to 10 CFR 50.90, Commonwealth Edison (Comed) Company is requesting changes to Appendix A, Technical Specifications, of Facility Operating License Nos.

NPF-37, NPF-66, NPF-72 and NPF-77. The proposed changes are to the Byron Station and Braidwood Station Improved Technical Specifications Table 3.3.1-1, " Reactor Trip System Instrumentation," and Table 3.3.2-1, " Engineered Safety Feature Actuation System Instrumentation," to revise twelve Allowable Values.

The improved Technical Specifications (ITS) Tables 3.3.1-1 and 3.3.2-1 provide the required Allowable Value (AV) for each applicable function for the Rea:: tor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) instrumentation, respectively. Based on the value of selected parameters for these functions, the RTS initiates a unit shutdown and the ESFAS initiates operation of necessary safety systems to protect the reactor core design limits at,d the Reactor Cooiant System pressure /

boundary, and to mitigate accidents. For the selected parameters, AVs are established. /

which provide a conservative margin with regard to instrument uncertair'.;es, to ensure that cafety limits are not violated during anticipated transients, and that the consequences of design basis accidents remain acceptable. The AVs are used as a basis for checking instrument channel functionality when performing instrument loop l calibrations. In the calculation of AVs, statistical allowances are provided to account for g/

calibration tolerances and instrument drift, which are assumed to occur between calibrations. These AV calculations are based on an NRC approved or endorsed i methodology that incorporates uncertainties applicable for each instrument channel.

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9901060100 981229 ?

PDR P ADOCK 05000454 pg A Unicom Company

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i l December 29,1998

) U.S. Nuclear Regulatory Commission l Page Two An evaluation was undertaken in order to address observed differences between AVs

and their respective design basis calculations for the RTS and ESFAS functions contained in ITS Tables 3.3.1-1 and 3.3.2-1. This evaluation found that certain of the !

current AVs were inconsistent with their supporting design basis calculation results. To !

resolve these differences, instrument loop uncertainty calculations were revised to incorporate the latest plant operating assumptions. For the RTS and ESFAS AVs, twelve current AVs need to be revised. The twelve AVs being proposed have been adjusted to reflect changes in plant variables, such as decreased instrument uncertainty due to the use of smaller Measurement and Test Equipment /M&TE) error terms than originally assumed that reflect more accurate station-specific M&TE, the update of bounding ex-core detector output signal values to account for existing core loading and burnup, the impact of implementing the Resistance Temperature Detector (RTD)

Bypass Elimination Modification, and use of additional conservatism applied to bound minor calculation revisions in the future.

This proposed amendment request is subdivided as follows.

i l 1. Attachment A gives a description and safyty analysis of the proposed changes.

2. Attachments B-1 and B-2 include the marked-up Improved Technical Specifications (ITS) Table 3.3.1-1, " Reactor Trip System Instrumentation," pages 3.3.1-14,15,16, .18 and 19, and Table 3.3.2-1, " Engineered Safety Feature Actuation System Instrumentation," pages 3.3.2-9,12 and 14, each with the requested changes indicated for Byron Station and Braidwood Station, respectively. Additionally attached for information are marked-up ITS Bases pages B 3.3.1-5 and 60, and pages B 3.3.2-4 and 57.

3. Attachment C describes Comed's evaluation performed in accordance with 10 CFR 50.91(a)(1), which provides information supporting a finding of no significant hazards consideration using the standards of 10 CFR 50.92(c).

4. Attachment D provides information supporting an Environmental Assessment.

This proposed changes have been reviewed and apprc .ad in accordance with the requirements of Comed's Quality Assurance Program.

We are notifying the Stab c f Illinois of this application for amendment by transmitting a copy of this letter aH ^ Machments to the designated State Official. :

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\\oPSNW103\VOL4\LicSTAFralA\BYRBWD\av_ cover ltr. doc

December 29,1998 U.S. Nuclear Regulatory Commission Page Three i

Should you have any questions conceming this letter, please contact Mr. J. A. Bauer at (630) 663-7287.

Respectfully, R. M. K ch Vice President - Regulatory Services Attachments: Affidavit A - Description and Safety Analysis B Byron Station ITS/ITS Bases Marked-up Pages B Braidwood Station ITS/ITS Bases Marked-up Pages C - Significant Hazards Consideration D - Environmental Assessment ,

cc: Regional Administrator - USNRC, Region lli NRC Senior Resident inspector - Byron Station NRC Senior Resident inspector - Braidwood Station ;

Office of Nuclear Facility Safety -IDNS i 1

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

i STATE OF ILLINOIS )

COUNTY OF DUPAGE )

IN THE MATTER OF )

! Docket Numbers COMMONWEALTH EDISON (COMED) COMPANY )

BYRON STATION UNITS 1 anc' "; ) STN 50454 and STN 50455 1'

AND

$ BRAIDWOOD STATicN UNITS 1 and 2 ) STN 50456 and STN 50457

SUBJECT:

Application for Amendment to improved Technical Specifications; Revision to Twelve Reactor Trip System (RTS) and Engineered Safety Feature Actuation

System (ESFAS) Allowable Values i

AFFIDAVIT

I affirm that the content of this transmittal is true and correct to the best of my knowledge, information and belief.

R. M. Krich Vice President - Regulatory Services Subscribed and sworn to before me, a Notary Public in and for the State of Illinois, this 29 day of bEC EHbEA .1999 ll, OFFICIAL SEAi. I: '

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l ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CHANGES A.

SUMMARY

OF PROPOSED CHANGES Pursuant to 10 CFR 50.90, Commonwealth Edison (Comed) Company proposes to amend Appendix A, Technical Specifications, of Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77. The proposed amendment requests a change to the Byron Station and Braidwood Station Improved Technical Specifications (ITS) Table 3.3.1-1, " Reactor Trip System Instrumentation," and Table 3.3.2-1, " Engineered Safety Feature Actuation System Instrumentation," to revise twelve Allowable Values.

ITS Tables 3.3.1-1 and 3.3.2-1 provide the Allowable Value (AV) for each function for the Reactor Trip System (RTS) and Engineered Safety Feature Actuation System (ESFAS) instrumentation, respectively. An evaluation was undertaken in order to address observed differences between AVs and their respective design basis calculations for the RTS and ESFAS functions contained in ITS Tables 3.3.1-1 and 3.3.2-1 (Current Technical Specifications (CTS)

Tables 2.2-1 and 3.3-4). This evaluation found that certain of the current AVs were inconsistent with their supporting design basis calculation results. To resolve these differences, instrument loop uncertainty calculations were revised to incorporate the latest plant operating assumptions.

For the RTS and ESFAS AVs, twelve current AVs need to be revised. The twelve AVs being proposed have been adjusted to reflect changes in plant variables, such as decreased instrument uncertainty due to the use of smaller Measurement and Test Equipment (M&TE) error terms than originally assumed that reflect more accurate station-specific M&TE, the update of bounding ex-core detector output signal values to account for existing core loading and burnup, the impact of implementing the Resistance Temperature Detector (RTD) Bypass Elimination Modification, and use of additional conservatism applied to bound minor calculation revisions in the future.

The proposed changes are described in Section E, Tables 1 and 2, of this Attachment A. The marked-up ITS pages are provided in Attachments B-1 and B-2 for Byron Station and Braidwood Station, respectively. Additionally, the marked-up ITS Bases pages are provided for information as part of Attachments B-1 and B-2. In accordance with Byron Station and Braidwood Station ITS Section 5.5.14, " Technical Specifications (TS) Bases Control Program,"

changes to the ITS Bases resulting from the proposed AV changes will be evaluated under 10 CFR 50 59 and will be submitted to the NRC on a frequency consistent with 10 CFR 50.71(e).

B. DESCRIPTION OF THE CURRENT REQUIREMENTS ITS Tables 3.3.1-1 and 3.3.2-1 provide the required AV for each applicable function for RTS and ESFAS instrumentation. Based on the value of selected parameters for these functions, the RTS initiates a unit shutdown and the ESFAS initiates operation of necessary safety systems to protect the reactor core design limits and the Reactor Coolant System pressure boundary, and to mitigate accidents. For the selected parameters, AVs are established, which provide a conservative margin with regard to instrument uncertainties, to ensure that safety Page 1 of 10 Attachment A - Description and Safety Analysis

4 limits are not violated during anticipated transients, and that the consequences of design basis accidents remain acceptable. The AVs are used as a basis for checking instrument channel functionality when performing instrument channel calibrations and channel functional tests.

During channel calibrations and channel operational tests, the respective instrumentation channel is rendered inoperable and the appropriate TS Limiting Condition For Operation (LCO)

Action requirement is followed, as necessary. In all cases, the trip setpoint must be restored prior to returning the affected channel to service. Any instrument channel equipment that can not be calibrated to within its AV is repaired or replaced before returning the affected channel to service.

The AVs result from associated instrument channel uncertainty calculations which are the design basis in the determination of AVs. The AVs derived in the calculations are used as the basis for determining instrument channel operability. The AV is the limit that a channel trip ,

point may have when periodically tested, beyond which appropriate action must be taken. If the !

measured "As Found" trip setpoint is within the AV, then the associated instrumentation function ;

is considered to be operable. All channel setpoints have a calibration tolerance within the AV I which is the minimum allowed "As Left" setpoint tolerance. In the calculation of the AVs, i statistical allowances are provided to account for calibration tolerances and instrument drift which are assumed to occur between calibrations. The AV calculations are based on NRC approved or endorsed methodology that incorporates uncertainties applicable for each instrument channel. Section E, Tables 1 and 2, of this Attachment A provide the current AVs.

C. BASES FOR THE CURRENT REQUIREMENTS ,

The AVs are derived from the analyticallimits contained in the safety analyses. AVs provide a conservative margin with regards to instrument uncertainties to ensure that safety limits are not vio!ated during anticipated transients, and that the consequences of design basis accidents will be acceptable providing the unit is operated from within the LCOs at the onset of the event and the required equipment functions as designed. The operability of the RTS and ESFAS instrumentation ensures that the required reactor trip or engineered safety feature actuation occurs when the parameter being monitored reaches its setpoint.

To accommodate the instrument drift assumed to occur between operational tests and the ac;uracy to which setpoints can be measured and calibrated, AVs for the RTS and ESFAS setpoints have been specified in ITS Tables 3.3.1-1 and 3.3.2-1. An RTS trip or ESFAS actuation signal that may initiate within the associated AV is acceptable because an allowance has been made in the affected instrument uncertainty calculation to accommodate such deviation. The methodology used to derive the AVs is based upon combining the uncertainties in the instrumentation channels which are periodically tested. Inherent in the determination of the AVs are the magnitudes of these instrumentation channel uncertainties. Rack instrumentation utilized in these channels is expected to be capable of operating within the allowances of these instrument uncertainty magnitudes. Drift in excess of the AV exhibits the behavior that the instrument channel no longer meets the allowance provided in the uncertainty calculation. If the measured value of the associated bistable exceeds the AV without tripping, then that RTS or ESFAS function is considered inoperable.

The AVs are based on a methodology which incorporates uncertainties applicable for each instrument channel. A detailed description of the Westinghouse methodology utilized to Page 2 of 10 Attachment A - Description and Safety Analysis

-5 s,

calculate the RTS and ESFAS AVs, except for Turbine Trip Functions (Table 3.3.1-1 Function 15a, " Turbine Trip, Emergency Trip Header Pressure"), is presented in WCAP-12523, " Bases Document For Westinghouse Setpoint Methodology For Protection Systems, Commonwealth Edison Company, Zion / Byron / Braidwood Units," dated October 1990 (Reference 1). For i Function 15a, " Turbine Trip, Emergency Trip Header Pressure," a detailed description of the l Sargent & Lund/Engineers (S&L) methodology utilized to calculate the AV is presented in the '

S&L Mechanical Department Standard MES-3.6, " Calculations For Safety-Related Instrument Setpoint Margins," dated September 8,1986 (Reference 2). The key references in this S&L i standard that form the bas!s for the basic formula used in this turbine trip AV calculation are the Inst ument Society of America (ISA) Standard, ISA-S67.04-1988, "Setpoints For Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants" (Reference 3), and Regulatory Guide 1.105, " Instrument Setpoints for Safety-Related Systems," Revision 2, dated February 1986 (Reference 4), which endorses the ISA standard.

I D. NEED FOR REVISION OF THE REQUIREMENT Twelve current AVs associated with the RTS and ESFAS instrumentation functions are inconsistent with respect to the latest results of the Byron Station and Braidwood Station supporting design basis calculations. The revised supporting instrument loop uncertainty calculations utilize improved assumptions to reflect changes in plant variables, and form the new basis for the revised AVs. These revised AVs also reflect the use of additional conservatism applied to bound future minor calculation revisions. The major reasons for the proposed more restrictive changes to the twelve AVs are provided in Attachment A, Tables 3 and 4, respectively.

E. DESCRIPTION OF THE PROPOSED CHANGES Twelve AVs are being revised for certain RTS and ESFAS instrumentation functions where the latest revision of supporting design calculations resul;in AVs that are more restrictive than the current AVs. These proposed changes are listed in Attachment A, Tables 1 and 2, respectively.

l F. SAFETY ANALYSIS OF THE PROPOSED CHANGES The results of updated design calculations provide the basis for the revision to the twelve AVs.

These calculations were performed in accordance with the Westinghouse methodology described in WCAP-12583, " Westinghouse Setpoint Methodology for Protection Systems, Byron / Braidwood Stations," dated May 1990 (Reference 5), for all revised AVs except the

, revised AV for Table 3.3.1-1 Function 15a, " Turbine Trip, Ernergency Trip Header Pressure."

1 The NRC review and approval of this methodology is documented in the Safety Evaluation for Byron Station TS Amendment 53 and Braidwood Station TS Amendment 42, dated April 13, 1993. The Comed methodology used in calculating the proposed AV change for Table 3.3.1-1 Function 15a, " Turbine Trip, Emergency Trip Header Pressure," is described in the Comed Nuclear Engineering Standard NES-EIC-20.04, " Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy," Revision 0, dated October 14,1997 (Reference 6). This Comed methodology is consistent with the instrument Society of America (ISA) Standard, ISA-S67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Page 3 of 10 Attachment A - Description

and Safety Analysis

~

l l

l Plants" (Reference 7). This ISA standard is endorsed by Regulatory Guide 1.105, " Instrument l

Setpoints for Safety-Related Systems," Revision 2, dated February 1986 (Reference 4). The l NRC review and approval of this Comed methodology, as previously used to develop setpoint l changes for LaSalle County Station Unit 1, is documented in the Safety Evaluation for LaSalle I County Station TS Amendment 129, dated July 6,1998.

No changes have been made that affect the ability of the RTS or ESFAS instrumentation to perform its intended design function. The proposed changes continue to ensure that all of the design considerations are addressed. The changes to the AVs do not degrade or prevent any actions from taking place in response to an accident. The proposed changes to the twelve AVs for RTS and ESFAS instrumentation will not affect the trip setpoints at which a reactor trip or engineered safety feature actuation is initiated because the trip setpoints are not being changed and the instrumentation will continue to actuate in the same manner. No changes are being made to the analytical limits contained in the safety analyses. The only changes being made are to the AVs used as the basis for determining instrument channel operability, which have been derived from the analytical limits. The supporting design calculations that provide the basis for the proposed changes have been prepared, reviewed and approved in accordance with Comed's Quality Assurance Program. In all cases, the twelve proposed AVs provide limits closer to the trip setpoints for the respective instrumentation functions than the current AVs.

The difference between the trip setpoints and the revised AVs has been reduced, which increases the margin with respect to the analyzed safety limits.

Physical differences do exist in the measurement and test equipment used at each station.

However, the proposed AVs have been maintained consistent at both stations by selecting the bounding combination of error terms. We are in the process of administratively implementing these revised AVs in their respective surveillance procedures at both Byron and Braidwood Stations. This will ensure that all safety analysis assumptions remain valid during the interim period while this proposed amendment is being reviewed.

G. IMPACT ON PREVIOUS SUBMITTALS This proposed amendment request does not impact any other licensing submittals currently with the NRC, with the exception of the proposed conversion to the Improved Technical Specifications. On December 13,1996, Comed submitted a request for conversion from the Byron Station and Braidwood Station Current Technical Specifications (CTS) to improved Technical Specifications (ITS)(Reference 8). The impact of the proposed AV changes on the ITS conversion is the same as the impact on the current AVs in the CTS. The affected ITS pages are provided in Attachments B-1 and B-2.

H. SCHEDULE REQUIREMENTS There are no specific schedule requirements for this submittal, however due to the nature of the change involved, a less than twelve-month approval cycle is requested. In the interim, the revised AVs are in the process of being administratively imposed at both stations.

Page 4 of 10 Attachment A - Description and Safety Analysis

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l. REFERENCES 1; WCAP-12523, " Bases Document For Westinghouse Setpoint Methodology For Protection Systems, Commonwealth Edison Company, Zion / Byron / Braidwood Units,"

dated October 1990, i

2. S&L Mechanical Department Standard MES-3.6, " Calculations For Safety-Related Instrument Setpoint Margins," dated September 8,1986. 1

3. Instrument Society of Amedca (ISA) Standard, ISA-S67.04-1988, "Setpoints For Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants." .]

4. Regulatory Guide 1.105, " Instrument Setpoints for Safety-Related Systems," Revision 2, dated February 1986.

5. WCAP-12583, " Westinghouse Setpoint Methodology For Protection Systems, Byron / !

Braidwood Stations," dated May 1990.

6. Comed Nuclear Engineering Standard NES-EIC-20.04, " Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy," Revision 0, dated October 14,1997. I 7.' Instrument Society of America (ISA) Standard, ISA-S67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants," approved September 30,1994. i

8. Comed letter from G. Stanley and K. Graesser to the USNRC, " Conversion to improved Standard Technical Specifications," dated December 13,1996.

t Page 5 of 10 Attachment A - Description and Safety Analysis

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i TABLE 1 I REVISED RTS ALLOWABLE VALUES CTS TABLE 2.2-1 ITS TABLE 3.3.1-1 CTS ITS Function Allowable Value

# From To 2.a 2.a Power Range Neutron Flux High s 111.36% RTP s 110.8% RTP i 1

2.b 2.b Power Range Neutron Flux Low s 27.36% RTP s 27.0% RTP "

3. 3.a Power Range Neutron Flux Rate High Positive Rate s 6.3% RTP s 6.2% RTP with time with time constant constant 2 2 sec 2 2 sec

4. 3.b Power Range Neutron Flux Rate High Negative Rate s 6.3% RTP s 6.2% RTP with time constant with time constant i 2 2 sec 2 2 sec !

5. 4. Intermediate Range Neutron Flux s 31.5% RTP s 30.0% RTP l

7. 6. Overtemperature AT 1.33% of AT Span 1.04% of AT Span

8. 7. Overpower AT 3.65% of AT Span 3.60% of AT Span

9. 8.a Pressurizer Pressure I Low 21869 psig 21875 psig 16.a 15.a Turbine Trip Emergency Trip Header Pressure 2 815 psig 2 910 psig I

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Page 6 of 10 Attachment A - Description and Safety Analysis

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- i TABLE 2 i

REVISED ESFAS ALLOWABLE VALUES i CTS TABLE 3.3-4 ITS TABLE 3.3.2-1 CTS ITS Function Allowable Value

# From To 1.d 1.d Safety injection I Pressurizer Pressure - Low 21813 psig 21817 psig 5.b.2 5.b.2 Turbine Trip and Feedwater isolation Steam Generator Water Level High-High (P-14)

Unit 2 s 82.8% NR Span s 81.5% NR Span 9.c 8.c ESFAS Interlocks l T., - Low-Low, P-12 2 546.9 F 2 548.0 F l

l A

Page 7 of 10 Attachment A - Description and Safety Analysis

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s TABLE 3 REVISED RTS ALLOWABLE VALUES CTS TABLE 2.2-1 ITS TABLE 3.3.1-1 :

CTS ITS Function Reason for Proposed Chance to AV 2.a 2.a Power Range Neutron Flux Smaller Measurement and Test Equipment Hi0h (M&TE) error terms that reflect the use of more accurate station-specific M&TE.