ML20086E178
| ML20086E178 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/18/1991 |
| From: | Schuster T COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20086E179 | List: |
| References | |
| TAC-M79082, TAC-M79083, TAC-M79084, TAC-M79085, NUDOCS 9111270079 | |
| Download: ML20086E178 (5) | |
Text
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Commonwealth Edison C5 1400 0 pus Place O. Downers Grove, luinois 60515 November 18, 1991 Dr. Thomas E. Hurley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Supolement to Application for Amendment to facility Operating Licenses NPF-37/66 & NPF 72/77 Appendix A Technical Specifications TAC)99082 A79083 andM79084h79085 NRC _ DochtLNon__50_ _454 /A 55_and_50- 456 L45 L...____
Reference:
a) October 26, 1990 letter from T. Schuster to Dr. T. Murley containing an application for Amendment to the Byron /Braidwood facility Operating Licenses b) April 23, 1991 letter (Revised 4/30/91) from T. Schuster to Dr. T. Hurley requesting a delay of review of the proposed amendnient of Reference (a).
Deae Dr. Murley:
In Reference (a), pursuant to 10 CFR 50.90, Commonwealth Edison Company (CECO) proposed to amend Appendix A, Technical Specifications of facility Operating Licenses NPF-37/66 and NPF-72/77 for Byron /Braidwood Stations. The proposed amendment was to revise a portion of Technical Specification Tables 2.2-1 and 3.3-4, Reactor Trip System Instrumentation Trip Setpoints and Engineered Safety Features Actuation System Instrumentation Trip Setpoints respectively, to reduce the setpoint for Low-Low Steam Generator Level Reactor Trip and Auxiliary Feedwater Initistion from 40.8% to 34.8% level for the Unit 1 Model D-4 Steam Generators. This setpoint reduction utilized existing excess margin present in the original 40.8% setpoint.
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Dr. 1.E. Murley November 18, 1991 By the letter of Reference (b) Ceco requested that reviews and processing of the proposed amendment of Reference (a) be delayed until a setpoint study being conducted by CECO was completed. The scope of the setpoint study involved all setpoints provided by Westinghouse for the Byron /Braidwood Stations Technical Specifications including Steani Generator Level setpoints specified for Reactor Trip and Auxiliary feedwater initiation. The Setpoint Etudy was a comprehensive review of the design assumptions madc by Westinghouse comparing current plant specific data to the original Westinghouse assumptions. The letter of Reference (b) also indicated that the previously quoted margin to trip values of the proposed amendment may change as a result of the study.
As a result of plant specific calculations performed for the Byron /Braldwood S/G Level instrumentation and related calibration and surveillance activities, a 10.3% of span value of excess margin has been determined to be available for the current S/G Low-2 Level Trip Setpoint. The previously identified excess margin was 7.8% of span. As a result of the change in excess margin new values for the S/G Low-2 Level Setpoint, Allowable Value, Total Allowance, Z and Sensor Error have been identified. As indicated in our original submittal the setpoint reductions for Reactor Trip and Auxiliary feedwater Initiation will allow operation of the Unit 1 Steam Generators over a greater range during operational transients. In response to a level transient, these changes will permit greater time for an operator's manual actions to take effect and will reduce the potential for unwarranted reactor trips and Engineered Safeguards feature actuation of Auxiliary feedwater.
The revised description and bases of the proposed changes are contained in Attachment A. The revised Technical Specification Pages are contained in Attachment B. The proposed changes have been reviewed and approved by both on-site and off-site review committees in accordance with Commonwealth Edison procedures and Technical Specifications. Commonwealth Edison has reviewed this proposed amendment in accordance with 10 CFR 50.92 (c) anc has determined that no significant hazards consideration exists. This evaluatlan is documented in Attachment C. An applicability review of the need for an Environmental Assessment has been performed and is included in Attachment D.
Attachment E briefly discusses the changes made from the previous application for amendment submittal and the rationale for the changes.
Commonwealth Edison is notifying the State of Illinois of this supplement to an application for amendment by transmitting a copy of this letter and its attachment to the designated State Official.
To the best of my knowledge and belief the statements contained hercin are true and correct. In some respects, these statements are not based on my personal knowledge but upon information received from other Commonwealth Edison and contractor employees. Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
ZNLD/1249/4
i Dr. T.E. Hurley November 18, 1991 Please direct any questions you may have concerning this matter to this office.
Respectfully, ,
t-a t- -
T.K. Schuster Nuclear Licensing Administratcr
Enclosures:
Revised Application for Amendment (Att. A-D)
Description of Changes to Original Proposal (Att. E) cc: H. Kropp-Byron S. Dupont-Braidwood A. Hsia-NRR R. Pulsifer-NRR B. Clayton-RIII Office of Nuclear facility Safety-IONS StMo of f"/,; r . County of /)/ ' g wv-we^"
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AUACNENLA D E SC RI E110LA ND_ DAS E S_0LJ H L EROP05 E D_ CHANG ES LtESCRIPJION :
The proposed changes would revise Technical Specifications Tables 2.2-1, Item 13a and Table 3.3-4. Item 6c. The change to Table 2.2-1 Item 13a would reduce the values specified for Trip Setpoint and Allowable Value to 33.0% and 31.0% of Span respectively, for the Steam Generator Water Level Low-Low Reactor Trip for the Unit 1 Model D-4 Steam Generators. The change to Table 3.3-4, Item 6c would reduce the values specified for Trip Setpoint and Allowable Value to 33.0% and 31.0% of span respectively, for the Steam Generator Water Level Low-Low Start for the Unit 1 Motor and Diesel Driven Auxiliary feedwater Pumps. Not Applicab'e (N/A) entries have been made in both lables for Total Allowance, Z and tensor Error. The revised values are indicated on the marked-up Technical Specification pages included in i Attachment D. i BASES-_0LTHLER0EOSED.. CHANGES l
The steam generator water level instrumentation is a safety grade system I designed to actuate a reactor trip due to a loss of heat sink. The basic function of the reactor protection circuits associated with Low-Low Steara Generator Water Level is to preserve the Steam Generator heat sink for removal of long term residual heat. Should a complete loss of feedwater occur, the reactor would be tripped on a low-Low Steam Generator Water Level. In addition, an auto-start signal is provided at the same setpoint to two redundant auxiliary feedwater pumps to supply feedwater in order to maintain residual heat removal capability after the trip. The reactor trip acts prior ,
to Steem Generator tube uncovery. This reduces the required auxiliary feedwatei capacity, increases the time interval before the auxiliary feedwater
- pumps are required, and minimizes the thermal transient on the Steam Generator and Reactor Coolant System. Ihe auto-start of the auxiliary feedwater pumps at the sane setpoint as_the trip ensures a secondary heat sink is continually available after a trip coincident with a loss of normal feedwater.
The relctor trip function generated at the Low-Low Steam Generator Water Level trip setpoint is assumed to provide primary protection for the loss of Normal feedwater/ Loss of All Non-Emergency AC Power events, feedline Break event.
Loss of Load / Turbine Trip event and certain cases of the superheated Steam Line Break Mass and Energy Release Calculations outside containment. All of these analyses assume a "Shfety Analysis Limit" value for the reactor trip setpoint lower (more conservative) than the nominal Technical Specification trip setpoint value because it must account for all applicable errors and uncertainties associated with the Low-Low Steam Generator Level trip
- function. The " Safety Analysis Limit" value assumed for these accidents for
- the low-Low Steam Generator Water Level Reactor Trip / Auxiliary feedwater initiation was-13.7% of span and remains at that value for the proposed changes. The margin between the original Technical Specification setpoint for Trip / Auxiliary feedwater Initiation (40.8%) and the " Safety Analysis Limit" ZNLD/1249/6 n-,-.-,,, --v-,. ~ , - . .-,r-v.-
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(continued) l value of 13.7% contained an excess margin of 10.3% of span. This excess margin was in excess of that reautred to account for instrument error and uncertainties identified in the statistical setpoint study. This change will permit reduction of the Unit 1 Low-Low Steam Generator Water Level Reactor Trip / Auxiliary feedwater Initiation Setpoint from 40.8% to 33.0%, reducing the ercess margin from 10.3% of span to 2.5% of span. However, as stated previously, the margin of safety is unaffected since the new Reactor Trip / Aux 111ary feedwater Initiation setpoint is bounded by the original
" Safety Analyses Limit" value used in the applicable safety analyses.
t An evaluation of the impact these changes had on the LOCA and non-LOCA Safety Analyses has been performed, including Steam Generator Tube Rupture. He a conclude that all regulatory and design limits will continue to be met after impicmentation of the proposed change.
The changes were also evaluated for their indirect effect on the Anticipated Transient Hithout Scram (ATHS) Mitigation System Setpoints. This system has been installed at both Byron and Braidwood Stations. The Byron /Braldwood ATHS mitigation system automatically trips the main turbine and initiates Auxillary feedwater when level falls below a prescribed setpoint in 3/4 Stear Generators. The steam generator level setpoint for ATHS initiation is specified as 3% of narrow range span below the nominal Low-low Steam Generator level Reactor Trip / Auxiliary feedwater Initiation Setpoint. Therefore, when the Reactor Trip Setpoint is changed the ATHS Setpoint must be changed accordingly. The guidance provided by Hestinghouse HCAP-11436 "AMSAC Generic Design Package" for selection of the ATHS Setpoint places restrictions on setpoint selection. The first is that it occur below the reactor protection system low-Low Steam Generator level trip setpoint. The second is that the reduced ATHS setpoint be no lower relative to the reactor trip setpoint, than .
the environmental and reference leg heatup allowances, for Byron and Braidwood stations the additive values of these two allowances is 12.6% of span. The third restriction is that the setpoint be no lower than 5% level.
It can be seen that the first two restrictions are met by virtue of the way the ATHS setpoint is specified. The third restriction, a minimum value of 5%
level, is easily met by the new setpoint value of 301. Therefore, the ATHS mitigation system-initiation assumptions are still valid.
The setpoint reductions for Reactor Trip and Auxiliary feedwater Initiation w111-allow operation of-the Unit 1 Model 0-4 Steam Generators over a greater range during operational transients. In response to a level transient, these ,
changes will permit greater time for an operator's manual actions to take effect and will reduce the potential for unwarranted reactor trips and ESF actuation of Auxillary feedwater.
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