ML17309A627

Forwards Response to Request for Addl Info Re Util 970929 Request for Amend to License DPR-18,revising TS on Main Steam Line Isolation Signal Set Point.Design Analysis DA EE-92-089-21,Rev 1 Encl

ML17309A627
Person / Time
Site:	Ginna
Issue date:	02/06/1998
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17265A155	List: ML17265A154 ML17309A627
References
NUDOCS 9802120041
Download: ML17309A627 (7)

Text

CATEGORY 1 REGULA Y INFORMATION DISTRIBUTIOA SYSTEM (RIDS)

ACCESSION NBR:9802120041 DOC.DATE: 98/02/06 NOTARIZED: YES DOCKET ¹ FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas &, Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION VISSING,G.S.

SUBJECT:

Forwards response to request for addi info re util 970929 request for amend to license DPR-18,revising TS on main steam line isolation signal set point. Design analysis DA EE-92-089-21,Rev 1 encl.

DTSTRTBUTTON CODE: A001D TITLE: OR COPTES RECETVED:LTR Submittal: General Distribution 3 ENCL i S1EE: 9Z, NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72) . 05000244 G

RECIPIENT COPIES " RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL 0

PD1-1 LA 1 1 PD1-1 PD 1 1 VISSING,G. 1 '

INTERNAL ILE CENTRE 01 1 1 NRR/DE/ECGB/A 1 1 EMCB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D

E NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 13 ENCL 12

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AND ROCHES7ER GAS AND ELEC7RkC CORPORA77ON ~ 89 FAS7 AtrrENUE, ROCHESTER, N. Y 1dbf9-0001 ARFA CODE716 54'6-27O0 ROBERT C. MECREDY Vice President Nudeor Operations February/1 998 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555

Subject:

Response to Request for Additional Information dated November 20, 1997 Rochester Gas & Electric Corporation R.E. Ginna Nuclear Power Plant Docket No. 50-244

Reference:

Letter from Guy S. Vissing, NRC, to R.C. Mecredy, RG&E, Request for Addi tional Infornration - Review of the Request for Aknendhnent Dated Septekkkber 29, 1997-Change to the Technical Specificatiokk related to the Main Steakn Line Isolation Sigkkal Set Poikkts (TAC No. M99702), dated November 20, 1997

Dear Mr. Vissing,

Enclosed please find a response to the subject request for additional information (RAI). Please contact us ifwe may be any further assistance.

Very trul yours, Dy Robert C. Mecredy Subscribe/ and sworn to before me on thisP'th day of February 1998.

b 30HM-SCOTT FlSH Notary Public in the State of Nee York Monroe Coun Notary. Public My Commission Expires 01FI5008 155 MDF858 Attachment IIIIIIIIIIIIIIIIIIIIIIIIIIIIII,IIIIIII

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.9802120041 ID d yern r20 1 7 The response to Questions 1 through 4 have been consolidated into one response.

Provide the bases for Fuzzction 4.d, "High Steanz Flow Coincident wi th Safety Injection and Coincident with T - Lo>v, "and Function 4.e, "High - High Steam Flow Coincident with Safety Injection" ofLCO Table 3.3.2-I. Provide a discussion of how the bases for these fiznctions will be met with the proposed Allowable Values and Trip Setpoints for Modes I, 2, azzd 3.

Provide a discussion of the applicable analyses and how these fzznctions are modeled.

Provide the setpoints used in the analysis.

Izz your submittal you stated, "...Function 4.e willnot provide closure of the MSIVs due to an inadvertent opening of azz atmospheric relief or safety valve. Consequezztly, only Fuzzction 4 d performs this firzzctiozz. " However, the setpoint for Function 4.d is higher than the capaci ty ofa single atmospheric relief valve. Please explain your statement.

Yozz I

stated, "Choosing 0% RTP equat'es to 0.66I'6 lbnzfhr and is also equal to two ARVs opening at I005 psig. However, should both ARVs open, the steazn generators would blow dow)zfz'ozzz each of the ARVs and, therefore, each steam line would ozzly be expected to expe'rience the effect of a single open ARV. Also, the proposed allowable value would require a steanz break to reszzltin aflow eqzzivalent to the capacity of two ARVs (ozz a steanz line) before the Allowable Value is'reached. Justify your selection of the allowable value for breaks resulting inflows ranging betweezz the proposed value and the sizes assuznedin UFSAR Section I5.I.6.

The steam line break analyses for Ginna Station are described in UFSAR Sections 6.2.1.2, 15.1.5, and 15.1.6. A discussion of the Chapter 15 analyses with respect to LCO Table 3.3.2-1 is provided below. Steam line breaks for the containment integrity analysis (UFSAR Section 6.2.1.2) occur upstream of the main steam isolation valves (MSIVs) such that the non-return check valves are credited with preventing blowdown from more than one steam generator. The non-return check valves are passive devices that are not assumed to fail; therefore, automatic isolation via the MSIVs is not assumed (see ITS bases page B 3.7-7).

A discussion of the Chapter 15 analyses with respect to LCO Table 3.3.2-1 is provided below.

UFSAR Section 15.1.5 provides the assumptions and results of steam line breaks equivalent to: (1) a full severance of a main steam line, and (2) a steam release through one main steam safety valve (MSSV). Although only hot zero power conditions are presented in the UFSAR, other power levels (e.g., 30%, 70%, 100%) have been evaluated to demonstrate that hot zero power is most limiting with respect to DNB. The LOFTRAN code is used for steam line break analyses. Main steam isolation is required for both UFSAR Section 15.1.5 cases with the MSIVs assumed to close within 5 seconds of receiving a close signal. This 5 second closing time is addressed by ITS surveillance requirement SR 3.7.2.1. The generation of the closing signal is described below:

For the large steam line break scenario, credit is taken for main steam isolation on high-high steam flow coincident with SI (i.e., LCO Table 3.3.2-1, Function 4.e).

Specifically, the MSIVs are assumed to receive a closure signal 2 seconds after the SI parameter for low steam line pressure is reached (358 psig per LCO Table 3.3.2-1, Function l.e). This is due to the fact that the high-high steam flow input of 3.7E6 ibm/hr is reached very rapidly (<< 1 second) with SI on low steam line pressure SI occurs ~

at approximately 1.5 seconds (see UFSAR Table 15.1-6). The fact that the is complete.

the high-high steam flow input is also verified after the LOFTRAN run The additional 2 second delay addresses signal delay time after the SI

'ccurring parameter has been met.

For the steam release equivalent to a MSSV lift, credit is taken for main steam isolation on high steam flow coincident with SI and low T(i.e., LCO Table 3.3.2-1, Function 4.d). Similar to the large steam line break, the MSIVs are assumed to receive a closure signal 2 seconds after the SI parameter is reached. The high steam flow input of 0.66E6 ibm/hr (i.e., proposed Allowable Value) is reached very rapidly since the flowrate through a MSSV is 0.82E6 ibm/hr per UFSAR Table 10.1-1 (also see UFSAR Figure 15.1-34). The low T, of 543'F is reached approximately 40-50 seconds after the break per UFSAR Figure 15.1-33. The SI occurs after 100 seconds (see UFSAR Table 15.1-6) with an additional 2 second time delay assumed. The fact that the SI occurs affer the high steam flow and low T~ input is also verified after the LOFTRAN run is complete.

UFSAR Section 15.1.6 describes the assumptions and results of the combined atmospheric relief valve (ARV) and main feedwater regulating valve (MFRV) failures. These combined failures are addressed due to postulated instrument failures with respect to the advanced digital feedwater control system (ADFCS). Several cases were examined (see Table 15.1-7) with the worst being the coincident failure of both ARVs and both MFRVs going full open.

Each ARV provides flow equivalent to 0.329E6 ibm/hr per UFSAR Table 10.1-1. As stated in UFSAR Section 15.1.6.1.2, manual operator action to initiate SI, feedwater isolation, and main steam isolation is assumed to occur at 600 seconds for hot zero power cases (i.e., ug automatic main steam isolation is assumed). For full power cases, a reactor trip terminates the event either automatically or manually at 600 seconds (i.e., ~n automatic main steam isolation is assumed). Given these assumptions, the full steam line breaks at hot zero power remain bounding.

In summary, Ginna Station has analyzed and documented in the UFSAR steam line breaks down to that equivalent to a MSSV steam release (0.82E6 lbm/hr) assuming automatic steam line isolation. In addition, steam line breaks up to the flow equivalent to two ARVs (or 0.66E6 Ibm/hr) have been analyzed assuming no steam line isolation until 600 seconds.

However, this steam flow is divided between two steam generators. This is not considered to be significant since the overall cooldown effect on the RCS would be the same. Also, the UFSAR analysis assumes that both MFW regulating valves failed full open to maximize the cooldown effect. This assumption is not valid for a steam break equivalent to 0.66E6 lbm/hr on one steam generator since ADFCS could not cause a break size this large. RGB has performed a LOFTRAN run with a 0.66E6 lbm/hr break in one steam generator with no steam line isolation until 600 seconds with MFW operating normally. For this case, the coincident ARV and MFW regulating valve failure remains bounding.

Therefore, RGkE has selected 0.66E6 lbm/hr as the Allowable Value for the LCO Table 3.3.2-1, Function 4.d since there is no existing UFSAR analysis demonstrating that manual steam line isolation is acceptable above this flowrate. The basis for the Trip Setpoint of 0.4E6 lbm/hr is addressed in response to Question 5 below.

Provide theinslnunent uncertainty calculation that was performed to confirtn the Allowable Value willnot be exceeded. Discuss the analytical value usedin this calculation.

A copy of Design Analysis EE-92-089-21, Revision 1 is attached. Section 10.0 of this analysis presents the actual setpoint evaluation based on the instrument loop uncertainties calculated in earlier sections of the analysis. As shown on the bottom of page 48, a setpoint of 0.4E6 Ibm/hr did not meet the current Allowable Value of 0.55E6 lbm/hr when all uncertainties were accounted for. Attachment 1 (page 52) shows that with a revised uncertainty analysis for this instrument loop based on Ginna historical data, and using the existing setpoint of 0.4E6 Ibm/hr, the new Allowable Value of 0.66E6 lb/hr would be met.