05000336/LER-1996-031-01, :on 961003,discovered Potentially non-conservative Assumptions Used in Analysis Cold Result in Exceeding ASME Code Max Relief Valve Accumulation for Sg. Caused by Inadequate Design of Mssv.Plant Mods Initiated

:on 961003,discovered Potentially non-conservative Assumptions Used in Analysis Cold Result in Exceeding ASME Code Max Relief Valve Accumulation for Sg. Caused by Inadequate Design of Mssv.Plant Mods Initiated

ML20137N881
Person / Time
Site:	Millstone
Issue date:	04/03/1997
From:	Joshi R NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20137N861	List: 05000336/LER-1996-031-01, :on 961003,discovered Potentially non-conservative Assumptions Used in Analysis Cold Result in Exceeding ASME Code Max Relief Valve Accumulation for Sg. Caused by Inadequate Design of Mssv.Plant Mods Initiated
References
LER-96-031-01, LER-96-31-1, NUDOCS 9704090079
Download: ML20137N881 (3)

LER-1996-031, on 961003,discovered Potentially non-conservative Assumptions Used in Analysis Cold Result in Exceeding ASME Code Max Relief Valve Accumulation for Sg. Caused by Inadequate Design of Mssv.Plant Mods Initiated

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
3361996031R01 - NRC Website

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NRC FORM 366 U.s. NUCLEAR REGULATORY COMMisslON APPROYED BY OMB NO. 3160-0104 (4 95)

EXPlRES 04/30/98 l

LICENSEE EVENT REPORT (LER)

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M ANAGEMENT AND BWGG. W ASHINGTON. DC 20603.

MCIUTY NAME Of DOCKET NUMDER (2' PAGE (3)

Millstone Nuclear Power Station Unit 2 05000336 1OF3 TITLE 14) 1 Non-conservative Assumptions identified in Analysis for Peak Secondary System Pressure i

EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INv0LvED (8) sE U AL RE $ N MONTH DAY YEAR YEAR MONTH DAY YEAR NU R

10 03 96 96

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01 04 03 97

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l OPERATING THis REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE (9) 5 20.2201(b) 20.2203(aH2Hv) 50.73(a)(2)(i)

So.73(aH2)(viii) l POWER 20.2203(aH1) 20.2203(a)(3)(i)

X So.73(aH2Hii) 50.73(aH2)(x)

LEVEL (10) 000 20.2203(aH2)(i) 20.2203(a)(3)(ii) 50.73(aH2Hiii) 73.71

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Rf NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (include Area Codel R. G. Joshi, MP2 Nuclear Licensing (860) 440-2080 i

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THis REPORT (13)

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SYSTEM COMPONENT MANUFACTURER

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I SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR l-

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X NO DATE (15) j ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) l On October 3,1996, it was discovered that potentially non-conservative assumptions used in an analysis could result l

in exceeding the ASME Code maximum relief valve accumulation for the steam generators (SG) and main steam l

line piping during an analyzed design basis event. During a review of the analysis for the single main steam j

isolation valve (MSIV) closure event, it was discovered that potentially non-conservative assumptions were made in the modeling of the main steam line, main steam safety valves (MSSV), and SGs. A preliminary analysis of these i

conditions has been performed. The results indicate that the 110 percent of design rating of the SGs would be exceeded for the single MSIV closure and loss of load design basis events.

The cause of this event was an inadequate design of the MSSV inlet piping and the failure to adequately assess the l

l piping pressure losses to the MSSVs.

l As a result of this event, corrective actions will be implemented to ensure that the plant response to analyzed events will not result in exceeding the design requirements of the SG. These actions will include any necessary reanalysis and plant modifications. These actions will be completed prior to plant restart from the current Outage.

J 9704090079 970403 PDR ADOCK 05000336 S

PDR NRC FORM 366 (4-95)

NRC FORM,366A U.S. NUCLEAR REGULATORY COMMisslON (4-95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER i 6)

PAGE (3)

SEQUENTIAL REvlsioN A

NUMBER NUMBER 2OF3 Millstone Nuclear Power Station Unit 2 05000336 96 -- 031 -

01 TEXT fit more space is required, tise additional copie: of NRC Form 366A) (17) 1.

Description of Event

On October 3,1996, it was discovered that potentially non-conservative assumptions used in an analysis could result in exceeding the ASME Code maximum relief valve accumulation for the steam generators (SG) and main steam line piping during an analyzed design basis event. At the time of discovery of this event, the unit was in Mode 5 at 0 percent power.

During a review of the analysis for the single main steam isolation valve (MSIV) closure event, it was discovered that potentially non-conservative assumptions were made in the modeling of the main steam line, main steam i

safety valves (MSSV), and SGs. It has been determined that the piping pressure losses between the SG and the MSSV inlets were not fully evaluated with respect to the MSSV performance and not fully addressed within plant safety analyses involving MSSV actuation.

Additionally in reviewing this event, it was identified that the relieving capacity of the MSSVs was significantly reduced due to the design of the connective piping between the main steam line and the MSSVs. The 6 inch connective piping is XXS grade piping with an inside diameter of less than 5 inches. The original analysis also did not identify peak steam generator pressure at the limiting location, resulting in a pressure slightly lower than the actual peak steam generator pressure. However, this error does not significantly affect the analysis.

A preliminary analysis of these conditions has been performed. The results indicate that the 110 percent of design rating of the SGs would be exceeded for the single MSIV closure and loss of load design basis events.

II. Cause of Event

The cause of this event was an inadequate design of the MSSV inlet piping and the failure to adequately assess the piping pressure losses to the MSSVs. The MSSVs were not capable of relieving their rated capacity at an acceptable pressure due to the smaller diameter installed connective piping. Additionally, subsequent analysis failed to account for the effects of the inadequate design in the analysis of peak pressures for the design basis events.

I The original design basis for Unit 2 did not require analysis for the single MSIV closure event. Analysis of this event was included at a later time and was originally performed by Westinghouse Electric Corporation. The identified potential non-conservative assumptions are believed to have existed in the original analysis also. The current analysis was performed by Siemens Power Corporation.

111. Analysis of Event The Single MSIV Closure Event is the limiting event for secondary side pressure. The closure of a single MSIV during operation will decrease the heat removal by the secondary system. Upon cessation of steam flow to the turbine, the pressure in the affected steam generator willincrease above the opening setpoint of the MSSVs.

l The peak analyzed SG dome pressure for this event is 1096 psia.

Due to the non-conservative assumptions identified in the analysis, the peak secondary side pressure would be greater than previously calculated. Based on the preliminary analysis, the peak secondary side pressure would exceed the design pressure rating of the SGs and main steam line piping. This would also be true for the Loss of Load Event. Therefore, this event is considered to be safety significant.

The effect of this condition on a small break loss of coolant accident (LOCA) was evaluated since the SG pressure is controlled by the MSSVs during the initial phases of the LOCA. The effect of the inlet losses on SG

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l NRC FORM 360A (4-95)

- NRC FORM 366A U.S. NUCLEAR REGULATORY Commission (4 95)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

SEQUENTIAL REvlslON YEAR NUMBER NUMBER 3OF3 Millstone Nuclear Power Station Unit 2 05000336 96 - 031 -

01 TEXT (if more space is required, use additional copies of NRC Form 366A) (17) pressure is not significant if all the MSSV banks have not opened. Since the net loss of capacity caused by the intet losses is less then the capacity of 1 MSSV, the maximum increase in SG pressure would equal the maximum difference in MSSV opening setpoint, or 10 psi. This magnitude change in SG pressure was evaluated I

by Siemens Power Corporation to be insignificant to peak fuel cladding temperature.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(B), any event or condition that resulted in the condition of the nuclear power plant, including the principal safety barriers, being seriously degraded, or that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant. This event i

was reported in accordance with 10 CFR 50.72(b)(1)(ii) on October 3,1996.

IV. Corrective Action

As a result of this event, corrective actions will be implemented to ensure that the plant response to analyzed j

events will not result in exceeding the design requirements of the SG. These actions will include any necessary i

reanalysis and plant modifications. These actions will be completed prior to plant restart from the current outage.

V.

Additional Information

None

Similar Events

LER 91-010: On October 18,1991, a reportability determination was made concerning a reanalysis of the main steam line break event inside the containment. The reanalysis confirmed that the assumptions made for the existing (1979) main steam line break (MSLB) analysis were non-conservative with respect to power level, break size, and single active failure. Using more restrictive assumptions, design limits for containment pressure and I

temperature could be exceeded. A multi-disciplinary task force was established to investigate containment response to postulated MSLBs. Plant modifications required to ensure an acceptable containment pressure response for a main steam line break inside the containment were installed and tested.

Manufacturer Data None Ells Codes SG - Steam Generator RV-Main Steam Safety Valve SB - Main Steam System ISV-Main Steam Isolation Valve WRC FORPA 366A (4 95)