ML20134N838
| ML20134N838 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 02/18/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20134N841 | List: |
| References | |
| NUDOCS 9702250004 | |
| Download: ML20134N838 (10) | |
Text
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Osmmonwealth Fxthon Q)mpany d
. I400 Opm Place Downers Grote. IL 6051M701 k
. Febrnry 18,1997 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk
Subject:
Byron Nuclear Power Station, Units 1 & 2 Facility Operating Licenses NPF-37 & NPF-66 NRC Doc _ket No. 50-454 and 50-455 Braidwood Nuclear Power Station, Units 1 & 2 Facility Operating Licenses NPF-72 & NPF-77 NRC Docket No. 50-456 and 50-457
" Steam Generator Level Optimization" Pursuant to Title 10, Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90),
Commonwealth Edison Company (Comed) proposes to amend Appendix A, Technical Specifications, for Facility Operating Licenses NPF-37 and NPF-66 for Byron Nuclear Power Station, Units 1 & 2 (Byron), and Facility Operating Licenses NPF-72 and NPF-77 for Braidwood Nuclear Power Station, Units 1 and 2 (Braidwood).
Please note that although the proposed Technical Specifications amendment is applicable to Byron and Braidwood Unit 1 only, this license amendment request is being docketed to
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reflect Byron Units 1 and 2 and Braidwood Units 1 and 2 due to common Technical Specification pages being used for both units.
Comed proposes to revise Technical Specifications, (TS) for Byron and Braidwood to support steam generator replacement. Comed will be replacing the original Westinghouse D4 steam generators (OSGs) at Byron and Braidwood with Babcock & Wilcox International (BWI) steam generators. The installation of the BW1 steam generators necessitates an increase to the operating range (i.e., difference between the low-low and the high-high level trip setpoint) of the steam generators due to the decrease in narrow range span from 233 inches for the original Westinghouse steam generators to 180 inches for the BWI replacement steam generators. The increase in operating range will minimize
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the possibility ofina_dvertent plant trips following load changes and feedwater transients.
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A Unicom Company
U.S. Nuclear Regulatory Commission February 18,1997 Comed also proposes to eliminate outdated cycle specific notations for Byron and Braidwood Stations, since they are no longer valid.
This package consists of the following:
Attachment A Description and Safety Analysis of Proposed Changes to Appendix A Attachment B Proposed Changes to the Technical Specification Pages for Byron and Braidwood Stations Attachment C Evaluation of No Significant Hazards Attachment D Environmental Assessment Please note the affected Improved Technical Specifications (ITS) pages have been prepared and submitted showing the proposed changes for Byron and Braidwood.
The proposed changes in this license amendment have been reviewed and approved by both On-Site and Off-Site review in accordance with Comed procedures.
Comed is notifying the State ofIllinois of our application for this license amendment request by transmitting a copy of this letter and its attachment to the designated State Official.
The Byron Unit 1 Steam Generator Replacement Outage (SGRO) is scheduled during the eighth refuel outage (B1R08). The Braidwood Unit i SGRO is scheduled during the seventh refuel outage (AIR 07). Comed respectively requests the NRC Staff review and approve this license amendment request no later than November 3,1997, to support the current outage schedule for the lead station, Byron Unit 1.
l I affirm that this transmittal is true and correct to the best of my knowledge, information and belief.
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i U.S. Nuclear Regulatory Commission February 18,1997 Please address any comments or questions regarding this matter to Marcia.Lesniak, Nuclear Licensing Administrator at (630) 663-6484.
Sincerely,
)q=: OFFICIAL SEAL:::======:mN JACQUEUNE T EVANSj ll moram rueuc. ct Air o, ruw g g, Q l
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John B. Hosmer w comssion exe, : uiis, u
< """"""""""==l Vice President Signed before me on this / 7 day of rua u
,1997 by Oael E Cre 9 Notary Public
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Attachment A:
Description and Safety Analysis of the Proposed Changes Attachment B-1:
Proposed Changes to Appendix A, Technical Specification, for the Byron Nuclear Power Plant, Units 1 & 2 Attachment B-la:
Proposed Changes to Appendix A, Improved Technical Specification, for the Byron Nuclear Power Plant, Units 1 & 2 Attachment B-2:
Proposed Changes to Appendix A, Technical Specification, for the Braidwood Nuclear Power Plant, Units 1 & 2 Attachment B-2a:
Proposed Changes to Appendix A, Improved Technical Specification, for the Braidwood Nuclear Power Plant, Units 1 & 2 4
Attachment C:
Evaluation of Significant Hazards Attachment D:
Environmental Assessment cc:
A. B. Beach, Regional Administrator - RIII G. F. Dick, Jr., Byron /Braidwood Project Manager - NRR S. D. Burgess, Senior Resident Inspector - Byron C. J. Phillips, Senior Resident Inspector - Braidwood Oflice of Nuclear Safety - IDNS kada.bybwd:sgrp:sgrievla doc 3
l ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 i
A.
DESCRIPTION OF THE PROPOSED CHANGE Commonwealth Edison (Comed) proposes to revise the steam generator level setpoints in the following Technical Specifications (TS) for Byron Nuclear Power Station (Byron) and Braidwood Nuclear Power Station (Braidwood):
TS 2.2.1, Table 2.2-1, Functional Unit 13.a," Reactor Trip System Instrumentation Trip Setpoint: Steam Generator Water Level-Low-Low" TS 3.3.2, Table 3.3-4, Functional Unit 5.b.1," Engineered Safety Features Actuation System Instrumentation Trip Setpeint: Steam Generator Water Level -
High-High" TS 3.3.2, Table 3.3-4, Functional Unit 6.c.1," Engineered Safety Features Actuation System Instrumentation Trip Setpoint: Steam Generator Water Level -
Low-Low Motor-Driven Pump and Diesel Driven Pump Start" Technical Specification Surveillance Requirement (TSSR) 4.4.1.2.2 and 4.4.1.3.2 defined steam generator water level for an operable steam generator during hot standby and hot shutdown, respectively.
TS 3.4.1.4.1.b, defined steam generator water level for operable RHR loop during cold shutdown -loops filled.
The installation of Babcock and Wilcox, International (BWI), replacement steam generators (RSGs) at the Byron Unit I and Braidwood Unit i Nuclear Power Stations necessitates an increase to the operating range (ie: ditTerence between the low-low and the high high level trip setpoint) of the steam generators due to the decrease in narrow range j
span from 233 inches for the original Westinghouse Model D4 steam generators (OSGs) to 180 inches for the BWI RSGs. The increase in operating range ( ie: difference between the low-low and the high-high level trip setpoints) will minimize the possibility of inadvertent plant trips following load changes and feedwater transients.
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Comed at:,o proposes to eliminate outdated cycle specific notations from page 2-5 for i
both P 4:dwood and Byron and pages 3/4 3-25 and 3/4 3-26 (Braidwood only) since they are ro longer valid.
The proposed change is discussed in detail in Section E of this attachment. The affected TS and Improved TS (ITS) pages are shown in Attachments B-1, B-la, B-2 and B-2a wi h the changes noted.
B.
DESCRIPTION OF TIIE CURRENT REQUIREMENT Th low-low level trip setpoint ensures a minimum operating level which provides ad quate secondary liquid mass to remove primary system sensible heat and core decay heat shortly after reactor trip. The low-low level setpoint also initiates auxiliary feedwater now for long-term cooling. The Reactor Trip Steam Generator (SG) Water Level low-low setpoint (TS 2.2.1, Table 2.2-1, functional unit 13.a), and Engineered Safety Features Actuation System low-low auxiliary feedwater (AFW) start setpoint (TS 3.3.2, Table 3.3-4, functional unit 6.c.1) are required to be set to the current value of 233% ornarrow range span (NRS) with an allowable value of 231% NRS.
The high-high level setpoint protects steam lines and the main turbine by ensuring they remain undamaged from the introduction oflow quality, two-phase flow from the steam generators into the steam lines. This is achieved with the Engineered Safety Features Actuation System high-high SG level trip (TS 3.3.2, Table 3.3-4, functional unit 5.b.1) which trips the main feedwater pumps and the turbine. The current high-high level trip setpoint is 5 81.4% NRS with an allowable value of 5 83.4% NRS.
Both TSSR 4.4.1.2.2, TSSR 4.4.1.3.2 and TS 3.4.1.4.1.b ensure a minimum SG operating l
level which provides adequate secondary liquid mass for core decay heat removal during hot standby and shutdown modes.
C.
BASES FOR TIIE CURRENT REQUIREMENT The current trip setpoints are based on acceptable results having been shown in the current limiting accident analyses of record using the Westinghouse Model D4 steam generators with the Technical Specification setpoints including measurement uncertainties.
The limiting accidents for the Reactor Trip Steam Generator Water Level low-low setpoint (TS 2.2.1, Table 2.2-1, functional unit 13.a), and Engineered Safety Features Actuation System low-low AFW start setpoint (TS 3.3.2, Table 3.3-4, functional unit 6.c.1) that support the low-!aw setpoint are the Loss of Normal Feedwater and Feedwater Line Break transients.
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The limiting accident for the Engineered Safety Feataes Actuation System high-high SG level trip (TS 3.3.2, Table 3.3-4, functional unit 5.b.1) is the Feedwater System Malfunction that results in an increase in feedwater to one or more steam generators.
TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS 3.4.1.4.1.b have no found basis for a specific minimum inventory (i e., level) to provide adequate decay heat removal. The requirement for a minimum inventory to remove decay heat is met with assurance that the tube bundle is completely covered. Assurance that the Unit I tubes remained covered is achieved with a steam generator water level within the span of the narrow range level indication. This can be assured by specifying a value equal to or greater than the low-low setpoint. (The current Unit 2 limit of 18% for shutdown conditions does assure the tubes are covered although 18% is below the current Unit 2 low-low setpoint. The Unit I approach is more anservative. Improved Technical Specifications will provide for consistency between the Units consistent with the approach utilized for Unit 1 in this submittal.)
D.
NEED FOR REVISION OF THE REQUIREMENT The installation of BWI RSGs at the Byron Unit I and Braidwood Unit I necessitates an increase in the operating range (ie: difference between the low-low and the high-high level trip setpoints) of the steam generators due to the decrease in narrow range span from 233 inches for the OSGs to 180 inches for the RSGs. The increase in operating range (ie:
difference between the low-low and the high-high level trip setpoints) whi minimize the possibility ofinadvertent plant trips following load changes and feedwr ter transients.
Decreasing the possibility ofinadvertent plant trips will decrease the probability of plant transients and thus will improve the overall safety of the plant.
The requirements for minimum inventory to remove decay heat during shutdown conditions are being modified to reflect the minimum steam generator water level that assures the water level remains within the span of the narrow range level indication with j
the RSGs.
An Improved Technical Specifications (ITS) submittal was made by Byron and Braidwood Stations on December 13,1996. Some of the affected current TS requirements have been relocated from the ITS. However, a discussion of affected ITS sections and markups of the ITS pages associated with the low-low and high-high level setpoints are provided in this package.
E.
DESCRIPTION OF THE REVISED REQUIREMENT All TS that explicitly state the present low-low and high-high SG level setpoints will be modified to reflect the results of safety analyses performed with the RSGs.
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m Revision to Current TS (Attachments B-1 and B-2)
The Reactor Trip Steam Generator Water Level low-low setpoint for the RSGs (TS 2.2.1, Table 2.2-1, functional unit 13.a) and Engineered Safety Features Actuation System low-low AFW start setpoint for the RSGs (TS 3.3.2, Table 3.3-4, functional unit 6.c.1) will be changed from 233.0% NRS to 218% NRS, with the allowable value changed from 231.0% NRS to 216.1% NRS. Notations are included to state the applicability of the revised value to be aner cycle 8 for Byron and aRer cycle 7 for Braidwood.
The Engineered Safety Features Actuation System high-high SG level trip for the RSGs (TS 3.3.2, Table 3.3-4, functional unit 5.b.1) will be changed from 5 81.4% NRS to $
88% NRS, with the allowable value changed from 5 83.4% NRS to $ 89.9% NRS.
Notations are included to the state of applicability of the revised value to be aRer cycle 8 for Byron and aner cycle 7 for Braidwood.
TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS 3.4.1.4.1.b will be changed from 41.0% NRS to 18% NRS for the RSGs. Notations are included to state the applicability of the revised value to be aRer cycle 8 for Byron and aner cycle 7 for Braidwood.
Notes at the bottom of page 2-5 (TS 2.2.1, Table 2.2-1) with cycle specific reactor coolant loop flow values are deleted, since the earlier cycles have been completed. For Braidwood only, TS 2.2.1, Table 2.2-1, functional unit 13.b Trip Setpoint and Allowable Value columns are modified to delete reference to values for cycles already completed.
l Likewise, for Braidwood only, TS 3.3.2, Table 3.3-4, functional units ' N2 and 6.c.2 Trip Setpoints and allowable value columns are modified to delete referent a values for cycles already completed.
Revision to ITS (Attachments B-la and B-2a)
The Reactor Trip Steam Generator Water Level low-low allowable value for the RSGs (TS 3.3.1, Table 3.3.1-1, functional unit 14.a) and Engineered Safety Features Actuation System low-low AFW start allowable value for the RSGs (TS 3.3.2, Table 3.3.2-1, functional unit 6.b.1) will be changed from 231.0% to 216.1%.
The Engineered Safety Features Actuation System high-high SG level allowable value for the RSGs (TS 3.3.2, Table 3.3.2-1, functional unit 5.b.1) will be changed from 5 83.4%
to 5 89.9%.
The Improved Technical Specifications and Surveillance Requirements and Basis for SR 3.4.5.2, SR 3.4.6.2, and SR 3.4.7.2 will change the minimum steam generator water level from 33% to 18% for both Braidwood and Byron Unit I aner the RSGs have been installed.
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BASES FOR TIIE REVISED REQUIREMENT F.
The determination of the impact of the RSGs on the limiting accidents that suppo low-low setpoint were evaluated by Frarra;ome Technologies in 1996 for Byron Braidwood Unit 1. These accidents are the Loss of Normal Feedw Line Break transient. These accidents were analyzed with use of the fully-certifie RELAPS/ MOD 2-B&W computer code us.'g the methodology approved by the BAW-10169. All acceptance criteria were shown to be met with a low-low set 0% NRS demonstrating that the proposed low-low level setpoint change is acce The TS low-low setpoint and its associated allowable value were calculated 12583. In summary, the setpoint is the accident analysis setpoint (0% NRS) plus calculated uncertainties and additional margin. Uncertainties were calculated methodology approved in WCAP 12583 and determined to be approximat The setpoint was conservatively chosen as 18% NRS. The setpoint allow per the WCAP is 1.9%, which yields a TS allowable value of 16.1% NRS.
The determination of the impact of the RSGs on the limiting accident that su high-high setpoint was also evaluated by Framatome Technologies i
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Braidwood Unit 1. The limiting accident is the Feedwater System Malfunct I
in an increase in feedwater to one or more steam generators. This accide hdl with use of the fully-certified RELAP5/ MOD 2-B&W computer code using met o o approved by the NRC in BAW-10169.
All acceptance criteria were shown to be met with a high-high setpoint of d
addition to the above acceptance criteria, it was shown that the steam generato completely fill with liquid. This assures that the steam lines and turbine rema t
undamaged with no introduction oflow quality, two-phase flow from the steam g into the steam lines during the transient. With all acceptance criteria met, th high-high level setpoint change is demonstrated to be acceptable.
The TS high-high setpoint and its associated allowab i
minus calculated uncertainties and additional margin. Uncertainties were calcu l 9%
the methodology approved in WCAP 12583 and determined to be approximate y NRS. The setpoint was conservatively chosen as 88% NRS. The setpoint allowanc determined per the WCAP is 1.9% which yields a TS allowable value of 89.9%
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F.
BASES FOR TIIE REVISED REQUIREMENT The determination of the impact of the RSGs on the limiting accidents that support the low-low setpoint were evaluated by Framatome Technologies in 1996 for Byron and Braidwood Unit 1. These accidents are the Loss of Normal Feedwater and Feedwater Line Break transient. These accidents were analyzed with use of the fully-certified RELAP5/ MOD 2-B&W computer code using the methodology approved by the NRC in BAW-10169. All acceptance criteria were shown to be met with a low-low setpoint of 0% NRS demonstrating that the proposed low-low level setpoint change is acceptable.
The TS low-low setpoint and its associated allowable value were calculated per WCAP 12583. In summary, the setpoint is the accident analysis setpoint (0% NRS) plus calculated uncertainties and additional margin. Uncertainties were calculated using the methodology approved in WCAP 12583 and determined to be approximately 15% NRS.
The setpoint was conservatively chosen as 18% NRS. The setpoint allowance determined per the WCAP is 1.9%, which yields a TS allowable value of 16.1% NRS.
The determination of the impact of the RSGs on the limiting accident that supports the high-high setpoint was also evaluated by Framatome Technologies in 1996 for Byron and Braidwood Unit 1. The limiting accident is the Feedwater System Malfunction that results in an increase in feedwater to one or more steam generators. This accident was analyzed with use of the fully-certified RELAP5/ MOD 2-B&W computer code using methodology approved by the NRC in BAW-10169.
Al! acceptance criteria were shown to be met with a high-high setpoint of 100% NRS. In addition to the above acceptance criteria, it was shown that the steam generators do not completely fill with liquid. This assures that the steam lines and turbine remain undamaged with no introduction of k.v quality, two-phase flow from the steam generators into the steam lines during the transient. With all acceptance criteria met, the proposed
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high-high level setpoint change is demonstrated to be acceptable.
The TS high-high setpoint and its associated allowable value were calculated also per WCAP 12583. In summary, the setpoint is the accident analysis setpoint ( 100% NRS) minus calculated uncertainties and additional margin. Uncertainties were calculated using the methodology approved in WCAP 12583 and determined to be approximately 9%
NRS. The setpoint was conservatively chosen as 88% NRS. The setpoint allowance determined per the WCAP is 1.9% which yields a TS allowable value of 89.9% NRS.
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TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS 3.4.1.4.1.b assure a minimum inventory (i.e.,
level) to provide decay heat removal. The requirement for a minimum inventory to remove decay heat is met with assurance that the tube bundle is completely covered.
Assurance that the tubes remained covered is achieved with a steam generator water level within the span of the narrow range level indication, therefore, this level will correspond to 0% NRS plus uncertainties. Uncertainties were calculated using the methodology approved in WCAP 12583, and determined to be 18% NRS with additional margin for the TS value.
G.
IMPACT OF TIIE PROPOSED CHANGE This change will impact the setpoints for the the Reactor Trip System, and the Engineered Safety Features Actuation System. This change will also impact the minimum operating water level required during shutdown modes.
H.
SCHEDULE REQUIREMENTS The Byron Unit 1 Steam Generator Replacement Outage (SGRO)is scheduled during the eighth refuel outage (BIRO8). The Braidwood Unit i SGRO is scheduled during the seventh refuel outage (AIRO7). Approval of this change is requested by November 3, 1997, to support the current outage schedule for the lead station which is Byron Unit 1.
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ATTACHMENT B-1 MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A 4
TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66 1
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BYRON STATION UNITS 1 & 2 REVISED PAGES:
2-5 3/4 3-25 3/4 3-26 3/44-2 3/4 4-4 3/44-5 I
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- i i
1
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