ML20211F151

Forwards Response to 970729 RAI Re Steam Generator Level TS Amend

ML20211F151
Person / Time
Site:	Byron, Braidwood
Issue date:	09/22/1997
From:	Hosmer J COMMONWEALTH EDISON CO.
To:	NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9709300358
Download: ML20211F151 (8)

Text

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, , t owners (inm ,11. HM i % 4'01 September 22,1997 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk

Subject:

Byron and Braidwood Station Units 1 & 2 Response to Request for Additional Information Regarding Steam Generation Level Technical Specification Amendment '

NRC Docket Numbers: 50: 454. 455. 456 and 457

References:

1. J. Hosmer letter to the Nuclear Regulatory Commission dated February 18,1997, transmitting Technical Specification Amendment for the Steam Generator Level

2. G. Dick letter to I. Johnson dated July 29,1997, transmitt;ng Request for Additional Information Pertaining to Revirjon of Steam Generator Level Technical Specification for Byron and Braidwood Station Units 1 and 2 Reference i transmitted the Commonwealth Edison Company's (Comed) request to amend the technical specification regarding the revision of steam generator level setpoints. Subsequent to that submittal, the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI) via Reference 2. Attached is Comed's response to that RAl.

If you have any questions concerning this correspondenr.e, please contact this office.

Sincerely, AdL SPl- e w j John B. Hosmer j/

Engineering Vice President Attachment cc: Senior Resident inspector-Braidwood Braidwood/ Byron Project Manager-NRR Regional Administrator-Rill

• Office of Nuclear Safety-IDNS 9709300358 970922

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ATTACHMENT Question #1:

The submitial makes numerous references to a setpoint methodology that was approved in WCAP-12583. Provide information related to the staffs approval of this document or the setpoint methodology. Also, provide additional details on how the TS setpoint uncertainties, margins, and allowable values were determined using the approved methodology.

Response to Question #1:

Comed is unaware of any specific NRC staff approval of the Westinghouse Topical Report, WCAP-12583, " Westinghouse Setpoint Methodology for Protection Systems, May 1990." However, the methodology presented in WCAP-12583 has been the basis for previous Comed submittals to the NRC staff for instumentation setpoints. This includes the original Byron and Braidwood protection system setpoints which were based on the methodology outilnted in the " Westinghouse Setpoint Methodology for Protection Systems, December 1982."

By a letter dated March 25,1992, the NRC staff approved changas to the Byron Unit 1 and Braidwood Unit 1 Technical Specifications which revised the current Unit 1 Steam Generator Low-Low trip setpoint. These changes were consistent with WCAP-12583 and with a setpoint methodology update that accounted for additional uncertainties in steam generator level. Additionally, Comed performed a self-initiated reconciliation program in 1991 which was also consistent with the setpoint methodology presented in WCAP-12583. The setpoint reconciliation program incorporated plant specific error terms and assumptions, and represented plant as-built conditions and surveillance requirements. The staff found the allowable values and setpoints developed by this program to be acceptable and approved a Technical Specification amendment by letter dated April 13,1993. The NRC staff's SER for this change, and the previously mentioned Unit 1 Steam Generator Low-Low trip setpoint change, specifically referenced the methodology of WCAP-12583. Accordingly, Comed believes that the methodology of WCAP-12583 represents a setpoint methodology acceptable to the NRC staff and is, therefore,the basis of the currently proposed setpoint change request.

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The setpoints calculated (and submitted for approval) for the replacement steam generators include the Low-Low Steam Generator Water Level and the High-High Steam Generator Water Level. Both of these setpoints and associated allowable values have been developed in accordance with the staff approved methodoiogy discussed in the previous paragraphs. Details regarding the development of these setpoints are as follows:

Low-Low Steam Generator Level Setooint For the low-low water level setpoint, the channel statistical allowance (CSA) was determined by arithmetically combining the dependent errors and systematically combining the independent errors using the SRSS method. Blas terms, such as adverse environmental uncertainties, are combined arithmetically if applicable.

The combination of the terms yields a CSA of 15.04% span (with adverse environmental uncertainties) and 5.04% (without adverse environmental uncertainties).

The Safety Analysis Limit (SAL) assumed in the accident analysis for the Feedline Break (FLB) event is 0% span. The FLB is a high energy line break (HELB) transient; therefore, the setpoint calculation must include adverse environmental uncertainties. The setpoint (SP) was chosen to be 18% span.

Therefore, the total allowance (TA) can be calculated as l SAL - SPl = 18% span.

Since the CSA for the FLB event is 15.04%, the margin is calculated as TA -

CSA = +2.96% span. The most conservative allowable value (AV) is calculated using the methodology presented in Section 4.0 of WCAP-12583. The value of AV is determined by subtracting the smallest " trigger value," T, (which in this case is Ti = 1.95% span) from the setpoint yielding a value of AV = SP - Ti =

16.1% span (conservatively rounded).

The SAL assumed in the accident analysis for the Loss of Normal Feedwater (LONF) event is 10% span. The LONF is not a HELB transient; therefore, the setpoint calculation does not require the application of adverse environmental uncertainties. The setpoint (SP) was chosen to be 18% span. Therefore, the total allowance (TA) can be calculated as l SAL - SPl = 8% span. Since the CSA for the LONF event is 5.04%, the margin is calculated as TA - CSA = +2.96%

span. The most conservative allowable value (AV)is calculated using the methodology presented in Section 4.0 of WCAP-12583. The value of AV is determined by subtracting the smallest " trigger value," T, (which in tYs case is Ti = 1.95% span) from the setpoint yielding a value of AV = SP - Ti = 16.1%

span (conservatively rounded).

K nla bybwdsgrpni stpt

The original submittal dated February 18,1997, stated that the SAL for the FLB and LONF events was 0% span. A subsequent revision to the LONF analysis revised the LONF SAL to 10% span causing SP, AV, and the margin to be re-evaluated based on the new SAL. The results are discussed above. Therefore, the proposed low low water level setpoint remains at 18% (AV = 16.1%) and applies to both accident analysis transients for which the low-low water level setpoint is applicable.

Hiah Hioh Steam Generator Level Setoolnt For the high high water level setpoint, the channel statistical allowance (CSA) was determined by arithmetically combining the dependent errors and systematically combining the independent errors using SRSS methodology.

Blas terms, such as adverse environmental uncertainties, are combined arithmetically if applicable. The combination of the terms yields a CSA of 8.63%

span (without adverse environmental uncertaintles).

The SAL assumed in the accident analysis for the Feedwater Malfunction (FWM) event which results in an increase in feedwater flow is 100% span. The FWM is not a HELB transient; therefore, the setpoint calculation does not require the application of adverse environmental uncertainties. The setpoint (SP) was chosen to be 88% span. Therefore, the total allowance (TA) can be calculated as [ SAL - SPl = 12% span. Since the CSA for the FWM event is 8.63%, the margin is calculated as TA - CSA = +3.37% span. The most conservative allowable value (AV) is calculated using the methodology presented in Section 4.0 of WCAP-12583. The value of AV is determined by adding the smallest

" trigger value," T, (which in this case is T3 = 1.95% span) to the setpoint yielding a value of AV = SP + Ti = 89.9% span (conservatively rounded).

Question #2; The changes in the setpoint levels are due to the changes in the steam generator (SG) design and the resultant reduction in the narrow range span.

Provide a physical description of the location of the level taps on the OSG and RSG.

Response to Question #2:

The enclosed figure provides a schematic representation of the level tap locations on both the OSG and the RSG. Since the figure is a schematic, it is not to scale.

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Question #3:

Confirm that the new SG water level setpoints are consistent with the inputs usad in the reanalyses of all transients and accidents that are affected by the SG water level setpoints. Also, confirm that the results cf the reanalyses meet acceptance criteria with these new SG water level setpoints. For the limiting case for each setpoint change (i.e., the case most impacted by the change in high level setpoint), provide a summary of the results of the reanalyses including a comparison to the acceptance limits for that analysis.

Response to Question #3:

The transient analyses which credit SG level trips include feedwater nalfunction (FWM) resulting in increase in feedwater flow, loss tf normal feedwater (LONF) and feedline break (FLB). FWM credits the SG high-high level trip and LONF and FLB credit the SG tow-low level trip. All three transients have been analyzed using RELAP5 and the revised safety analysis SG level trip setpoints.

The Technical Specification trip setpoints are based on the analysis trip setpoints with the appropriate uncertainties applied.

The FWM event is an ANS Condition 11 event with the following acceptance criteria:

. Pressure in the reactor coolant and main steam systems shall not exceed 110 percent of the design value.

. Fuel cladding shall be maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit for the correlation used.

In addition, an analysis is performed to show margin to steam generator overfill.

In the current UFSAR evalreton for this event, the peak primary and secondary system pressures for this ewnt are bounded by those for the loss of external electrical load (LOEL) because the reactor trip occurs significantly later than the turbine trip for the LOEL event. The replacement of the steam generators does not change this evaluation.

A comparison analysis between the D4 steam generator currently installed in Unit 1 and the replacement steam generator (RSG) was performed to show that the minimum DNBR is bounded by the current UFSAR analysis for the D4 steam generators. The model D4 steam generator was modeled in a case using the current high-high SG level trip setpoint. The RSG was modeled in a separate case using the proposed high-high SG level trip setpoint. The RSG case K:nh bybwdagrpni stpt

showed a smaller increase in power and a smaller decrease in primary side pressdre. Therefore, the current UFSAR minimum DNBR analysis for this event bounds the RSG.

An analysis was performed for the RSG using the proposed high-high SG level trip setpoint and initial conditions that maximize the potential for steam generator overfill. The results show that the RSG does not overfill.

The LONF event is an ANS Condition ll event with the following acceptance critena:

• Pressure in the reactor coolant and main steam systems shall not exceed 110 percent of the design value, e Fuel cladding shall be maintained by ensuring that the minimum DNBR remains above the 95/95 DNBR limit for the correlation used.

• An ultimate heat sink for decay heat removal must be assured.

in the current UFSAR evaluation for this event, the peak primary and secondary system pressures for this event are bounded by those for the loss of external electrical load (LOEL) because the reactor trip occurs significantly later than the turbine trip for the LOEL event. The replacement of the steam generators does not change this evaluation.

In the current UFSAR evaluation for this event, the total loss of reactor coolant flow event bounds this event with respect to DNBR because forced reactor coolant flow is maintained for this event. -The replacement of the steam generators does not change this evaluation.

This event was analyzed using RELAPS to verify that long-term cooling is maintained and that the pressurizer does not become water-solid. The results show that the heat removal capacity of the auxillary feedwater system and the RSGs are sufficient to remove core decay heat following reactor trip and to prevent the pressurizer from becoming water-solid.

The FLB event is classified as an ANS Condition IV event with the following acceptance criteria:

. Pressure in the reactor coolant and main steam systems shall not exceed 110 percent of the design value, e An ultimate heat sink for decay heat removal must be assured.

e The core remains intact for effective cooling.

• Radiation doses do not exceed the guidelines of 10CFR100.

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ln the current UFSAR evaluation for this event, the peak primary and secondary systern pressures for this event are bounded by those for the loss of external electrical load (LOEL) because the reactor trip occurs significantly later than the turbine trip for the LOEL event. The replacement of the steam generators does not change this evaluation.

An analysis was performed to demonstrate that the core remains covered throughout the transient and that long term heat removal was available. The results show that the core subcooling margin is available throughout the transient anJ that no fluid is lost from the reactor coolant system for the duration of the transient. Therefore, the core remains intact for effective cooling.

In the current UFSAR evaluation for this event, the doses resulting from the worst-case FLB event are based on the inventory of one SG released to the containment. Therefore, the doses are bounded by those predicted for the steamline break event. The replacement of the steam generators does not change this evaluation.

All translents that credit SG level trip setpoints have been analyzed or evaluated and demonstrate that acceptance criteria are met.

In addition to the above evaluations, Comed performs a core reload safety evaluation in accordance with approved Westinghouse methodology for each

\

reload cycle, incorporation of the RSG design into this process requires that the three transients discussed above be re-evaluated using the proposed SG level setpoints in order to validate the reload core design. This re-evaluation is currently in progress.

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_ _ . . _ _ _ _ . _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ . ._ ._ --._ - ._ . _m . ._ _ _ _

RSG/OSG LevelTap Compenson 7,

RSG OSG *

(WR/NR Level Upper Tap 3 452.708 ( 3 ,

I i WR/NR Level Upper Tap 448.870 l .

r i i i s

' 180.000 in ,1 l l i

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I 232.440 in NR Levellowertap 437.708 l 3 l ,

i  ! >

2 a i 2 .I J

f NR Levellowertap 429.500 l t

! i 1' l i 600.250 in i 559.000 in i

i l i t

• s l WR Levellower tap 402.688 WR Levellowertap 402.287 Support Pads Support Pads 1

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i Elevation values are shown in units of feet.

RSG reference: BWI Drawing 7720E001 Rev. 5

) OSG reference: Survey Data for SG 1A 1

8/8/97

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