ML17265A606

LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp

ML17265A606
Person / Time
Site:	Ginna
Issue date:	03/24/1999
From:	St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17265A605	List: ML17265A604 ML17265A606
References
REF-PT21-99 LER-99-001, LER-99-1, NUDOCS 9903310302
Download: ML17265A606 (8)

Text

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150.0104 EXPIRES 06/30/2001 (6.1998) Estimated burden per responsa lo comply with this mandatory Informa Reported 50 hrk IBSSOn learned are Incorporated Info 'equest licensing procass and fed back lo indusby. Forwanl cornrnants ragardi LICENSEE EVENT REPORT (LER) burden ossrnato lo Iha Records Management Branch (T< F33), U.s. Nudea Regularly CornrIEssion. WasNngton. DC 205550001. Bnd lo the Pape Reduction Project (31500104). Office of Management and Budget (See reverse for required number of Washington. DC 20503. If an lnforrnallon collection does nol display digits/characters for each block) currently vaEId OMB control number, lha NRC may not conduct or sponsor and a person is nol required lo respond lo, the hfonnatlon collecson.

FAcluTY NAME ill DOCKET NUMBER l2l PAGE (3)

R. E. Ginna Power Plant 05000244 1 OF 7 TITLE fel Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis EVENT DATE 5 LER NUMBER 6 REPORT DATE 7 OTHER FACILITIES NVOLVED 8 FACILITYNAME DOCKET NUMBER SEDVENTIAL REVISION MONTH DAY YEAR NUMBER NUMBER DAY YEAR 05000 FAC'ILrrY NAME DOCKET NUMBER 02 22 1999 1999 - 001 - 00 03 24 1999 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR E: Chock ona or moro 11 MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(aH2)(i) 50.73(a)(2)(viii)

POWER 20.2203 a 1 20.2203 a 3 I 50.73 a 2 ii 50.73 a 2 x LEVEL (10) 75 20.2203(a)(2)(i) 20 2203(aH3HII) 50.73(a) (2) (ill) ?3.71 20.2203 a 2 ii 20.2203 a 4 50.73 a 2 iv OTHER Part 21 20.2203(a)(2H(ill 50.36(c)(1) 50.73(BH2Hv) Specify in Abstract below 20.2203(a)(2Hiv) 50.36(c) (2) 50.73(a) (2) (vil) or ln NRC Form 366A LICENSEE CONTACT FOR THIS LER 12 NAME TELEPHONE NUMBER firielude Area Codel John St.Martin - Technical Assistant (716) 771-3641 REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX CAUSE SYSTEM COMPONENT MANVFACTVRER TO EPIX SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).

NO DATE (15) 06 18 1999 ABSTRACT (Limit to 1400 spaces, i.eapproximately 15 single spaced typewritten lines) (16)

On February 22, 1999, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a main steamline break inside containment with an assumed single failure of a main feedwater regulating valve. These errors had a nonconservative impact on calculated peak containment pressure.

The first-error involved the omission of a previously unaccounted for volume of high temperature feedwater. The second error resulted in an incorrect feedwater isolation time.

The consequences of these two modeling errors has the potential for containment to exceed its design basis value of 60 psig for a main steamline break with a concurrent failure of a main feedwater regulating valve.

The cause of the plant being in a condition that was outside the design basis of the plant was deficiencies in the Westinghouse main steamline break mass and energy release analysis.

Immediate corrective action was to place a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water/Containment fan cooler heat removal and pressure reduction capability, for Nodes 1 through 4 (when containment is required to be operable) to ensure adequate post-accident environmental conditions margin existed.

Co'rrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions, if needed, will be identified in a supplement to this LBR.

99033l0302 990324 PDR ADQCK 05000244 S PDR

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (9.1999)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 SEOUENTUIL REVISIOII y@R IIUMBER IIUMBER R. E. Ginna Nuclear Power Plant 05000244 1999 001 00 '2 OF 7 TEXT (If more space is required, use additional copies of NRC form 36Q) (17)

I. PRE" EVENT PLANT CONDITIONS:

In 1992, the NSSS vendor for Ginna Station (Westinghouse) analyzed a full spectrum of cases for the main steam line break (MSLB) and containment response for the R.E.

Ginna plant. The set of analyses was performed to support the program to reduce the boric acid storage tank (BAST) boron concentration to 2000 ppm. This analysis entailed over 120 cases, which varied break size and location, initial power level, single failure, offsite power availability, and different reactivity coefficients (rodded versus unrodded). In 1995, Westinghouse re-visited the steamline break mass and energy release event due to replacement steam generators (RSGs) and an increase to 18 month fuel cycles. Many of the previous cases could be eliminated due to:

1., An integral flow restrictor on the RSGs, eliminating any breaks larger than 1.4 square feet (the RSGs have flow restricting outlet nozzles which make any break size larger than the throat area of the restrictor not credible and thus, break flow areas greater than 1.4 square feet no longer need to be considered), and

2. Credit being taken for the main steam non-return check valve, which eliminated several types of breaks and the need to consider a main steam isolation valve (MSIV) single failure for containment overpressurization cases.

After eliminating cases that were no longer credible, cases for the 1995 analyses were selected from the 1992 BAST reduction program. Six of the most limiting cases were selected, including different break sizes, initial power levels, single failures, and offsite power assumptions. Another difference from the previous analyses is explained in the Ginna Station Updated Final Safety Analysis Report (UFSAR) Section 6. 2. 1. 2. 3:

"In the Reference 3 analyses (1992 BAST reduction program), the most limiting feedwater control system failure was assumed to be a failure in the auxiliary feedwater system control. This failure resulted in increased auxiliary feedwater flow rates for the entire duration of the transient. A failure of the isolation function of the main feedwater regulating valves (MFRVs) was not considered because a feedwater isolation signal also results in a main feedwater pump trip and in closure of the pump discharge valve (MFPDV). Thus, no single failure would have resulted in a failure to isolate main feedwater or a significant delay in the isolation of main feedwater to the affected steam generator. It has subsequently been realized that the condensate pumps would continue to pump through the main feedwater pumps at low steam generator pressures and that the pump discharge (MFPDV) closure time is approximately 80 seconds. Therefore, a failure of the isolation function of the main feedwater regulating valve (MFRV) could result in main feedwater flow via the condensate pumps for approximately 80 seconds to the affected steam generator. This new failure scenario is a more limiting failure than was assumed in the previous analyses due to the timing and magnitude of the increased flow."

NRC FORM 366A (6-1998)

1NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6.1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 6 NUPTIAL REVISION NUM88l NUMUSR R. E. Ginna Nuclear Power Plant 05000244 1999 PPg 00 3 OF 7 TEXT (If more space is required, use additional copies of NRC Form 36Q) (17)

The requirement to maintain operability of the main steam non-return check valves and the 80 second stroke time requirement of the MFPDV is addressed in the Ginna Station Improved Technical Specifications (ITS) Sections 3.7.2 and 3.7.3 respectively.

II'ESCRIPTION OF EVENT:

DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

0 February 22, 1999: Event date.

o February 22, 1999, 1540 EST: Discovery date and time.

o February 22, 1999, 1601 EST: NRC is notified of this condition per 10CFR50. 72 (b) (1) (ii) (B) ~

0 March 1, 1999, 2018 EST: Plant is shutdown in Mode 4 for the planned 1999 refueling outage.

B. EVENT:

On February 22, 1999, in activities unrelated to plant conditions, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV. (Refer to the letter RGE-99-207 from Mr. Steve M. Ira of the Westinghouse Electric Corporation to Mr. Peter Bamford, Rochester Gas and Electric Corporation (RG&E), dated February 22, 1999,

Subject:

Steamline Break Mass and Energy Release Analysis Nonconformance.) These errors had a nonconservative impact on calculated peak containment pressure.

The first error involves the volume of feedwater that exists between the specified isolation valve and the faulted steam generator. When the MFRV is assumed to fail to perform its isolation function, then the isolation function is performed by the MFPDV. Because of the relative locations of these valves to the steam generator, the volume of water is greater between the MFPDV and the steam generator than between the MFRV and the steam generator. This increased volume of high temperature feedwater was not accounted for in the analysis.

The second error was inadvertently isolating feedwater flow at 15 seconds instead of at 85 seconds (80 seconds valve stroke time and 5 seconds for signal delay) as was intended. This meant that feedwater flow should have continued for 70 seconds longer than was assumed in the analysis.

NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (9 I 999)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 SEQUEIITIAE REVISION BOMBER INMBER R. E. Ginna Nuclear Power Plant 05000244 1999 PPI 00 4 Of 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

The consequences of these two modeling errors has the potential for containment to exceed its design basis value of 60 psig for a MSLB with an assumed concurrent failure of a MFRV. This is due to adding higher energy fluid to containment for 70 seconds longer than was assumed in the 1995 MSLB analysis of record. There was also the potential for the containment to exceed its design basis value as documented in the 1992 analyses, due to discovery of a more limiting single failure.

At the time of discovery and until the planned shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature remained at approximately 35 degrees Fahrenheit. Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water System temperature, for Modes 1 through 4 (when containment is required to be operable). By maintaining this cooling medium for the containment fan coolers at such a low temperature, increased containment fan cooler heat removal ability was ensured and this, combined with a high cycle-specific excess shutdown margin, was sufficient to ensure containment pressure would remain less than 60 psig for this postulated scenario.

C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None E. METHOD OF DISCOVERY:

This event was disclosed to RG&E by the notification from its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV. (Refer to letter RGE-99-207 from Mr. Steve M. Ira of the Westinghouse Electric Corporation to Mr. Peter Bamford, Rochester Gas and Electric Corporation (RG&E), dated February 22, 1999,

Subject:

Steamline Break Mass and Energy Release Analysis Nonconformance.)

OPERATOR ACTION:

After review of this Westinghouse correspondence, Reactor Engineering and Analysis personnel notified Operations supervision, who notified the Control Room operators. The NRC Resident was notified at this time. At approximately 1540 EST on February 22, 1999, plant staff determined that a non-emergency one hour notification, per 10CFR50.72 (b) (1) (ii) (B), should be made to the NRC Operations Center. The Shift Supervisor made this notification at approximately 1601 EST on February 22, 1999.

NRC FORM 366A (6.19981

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (9 I998I LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

'ACILITYNAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 SEBUERTNL REVISIOII YEAR RUM BER INCUMBER R. E. Ginna Nuclear Power Plant 05000244 1999 001 00 5 OF 7 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Operators maintained a restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature from the time of discovery until plant shutdown on March 1 to ensure adequate margin ex'isted.

SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT:

A. IMMEDIATE CAUSE:

The immediate cause of the plant being in a condition that was outside the design basis of the plant was deficiencies in the Westinghouse MSLB mass and energy release analysis.

ROOT CAUSE:

The underlying cause of the deficiencies in the Westinghouse analysis was the analysis did not account for a greater volume of high temperature feedwater and used an incorrect feedwater isolation time. The additional mass and energy release from the MSLB are penalties in the containment pressure calculation.

IV. ANALYSIS OF EVENT:

This event is reportable in accordance with 10 CFR 21 and in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (ii) (B), which requires a report of, "Any event or condition ... that resulted in the nuclear power plant being In a condition that was outside the design basis of the plant".

An assessment will be performed considering both the safety consequences and implications of. this event. A supplement to this LER will be submitted with the results of this assessment. The following are results and conclusions of a preliminary assessment:

There were no operational or safety consequences or implications attributed to the deficiencies in the Westinghouse MSLB mass and energy 'release analysis because:

NRC FORM $ 66A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION IB.IBB8)

LICENSEE EVENT REPORT (LER)

TEXT CONTMJATION FACILITYNAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 SEOUEMUIE REVISION VE4R NUMBER NUMBER R. E. Ginna Nuclear Power Plant 05000244 1999 001 00 6 OF 7 TEXT (Ifmore space is required, use additional copies of NRC Form 366A) (17) o 'he6.2.1.2.3.

MSLB containment analysis A revised analysis is presented in was Ginna Station UFSAR performed by offsetting the adverse effects Section of the additional mass and energy release through refinements in the existing analysis assumptions. These improvements do not alter the total quantity of steam released to containment, but provide benefits in lowering the rate of the release and increasing the rate of containment heat, removal. Applying these analysis improvements to the MFRV failure cases initiated from 70% power (approximate power level at time of discovery) yields peak containment pressures that are less than those calculated in 1995.

0 The current Westinghouse analysis shows that the peak containment pressure design basis value of 60 psig would not be exceeded provided that service water (SW) temperature (SW supplies the containment recirculation fans) is limited to less than 45 degrees Fahrenheit. Prior to entering Mode 4 for the planned 1999 refueling outage, the screenhouse bay temperature was being maintained at approximately 39 degrees Fahrenheit, while lake temperature was approximately 35 degrees Fahrenheit. Thus, plant operation remained within its design basis since the time of discovery.

o Any small increase in containment pressure beyond the 60 psig design basis value would not be expected to fail the containment. In 1996, the containment was successfully tested to 72 psig during the Structural Integrity Test for the steam generator replacement project.

o ~ A steam line break does not result in severe radiological consequences at Ginna Station; thus full containment integrity is not considered required for this event (even though containment isolation is assumed to occur).

V. CORRECTIVE ACTION:

ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL'TATUS:

o For the operation of Ginna Station between the time of discovery and plant shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature has remained at approximately 35 degrees Fahrenheit. Thus,- plant-operation remained within its design basis since the discovery date.

NRC FORH 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6 I 998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 1 DOCKET 2 LER NUMBER 6 PAGE 3 SEOUENTIAE REVISION y@R NUMBER NUMBER R. E. Ginna Nuclear Power Plant 05000244 1999 001 00 7 OF 7 TEXT (Ifmore space is required, use additional copies of NRC Form 366A) (17) o Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature for Modes 1 through 4 (when containment is required to be operable) to ensure adequate margin existed.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

o Additional analyses are being performed to support future cycle operation. Results of these analyses will be provided in a supplement to this LER. During the shutdown, RG&E is working closely with Westinghouse to resolve this issue.

o The reactivity feedback in the analysis will be examined more closely to determine whether there is a return to power after the postulated MSLB in future cycles of operation. If necessary, the Ginna Station Core Operating Limits Report (COLR) will be changed to reflect more conservative shutdown margin levels.

o Additional corrective actions, if needed, will be identified in a supplement to this LER ~

VI. ADDITIONAL INFORMATION:

A. FAILED COMPONENTS:

None PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Station could. be identified.

C. SPECIAL COMMENTS:

None