ML17241A332

LER 99-004-00:on 990415,determined That as Found Cycle 10 Psv Setpoints Outside TS Limits.Root Cause Under Investigation.Psvs Replaced with pe-tested Valves During Cycle 11

ML17241A332
Person / Time
Site:	Saint Lucie
Issue date:	05/17/1999
From:	Frehafer K FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17241A331	List: ML17241A330 ML17241A332
References
LER-99-004, LER-99-4, NUDOCS 9905240096
Download: ML17241A332 (9)

Text

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2001 (6.1998)

Estimated burden per response to comply vrlth this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into the LICENSEE EVENT REPORT (LER) licensing process and fed back to industry. Fonvard comments regarding burden estimate to the Records Management Branch (TW F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555400t, and to the Papenvork (See reverse for required number of Reduction Project (3150.0104), Office of Management and Budget, digits/characters for each block) Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

FACILITYNAME (1) DOCKET NUMBER (2) PAGE (3)

St. Lucie Unit 2 05000389 Page 1 of 6 TITLE (4)

As Found Cycle 10 Pressurizer Safety Valve Setpoints Outside Technical Specification Limits EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8 SEQUENTIAL REVISION A IU NAM MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 04 15 1999 1999 - 004 - 00 05 17 1999 FACIUTY NAM DOCK NUM R OPERATING THIS REPORT IS SUBMIITED PURSUANT TO THE REQUIR EMENTS OF 10CFR ti: Check on e or more (11)

MODE (9) 20.2201(b) 20.2203(ax2)(v) X 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 050 20.2203 axl) 20.2203 a)(3)(t) 50.73(a)(2)(ii 50.73(a)(2)(x)

LEVEL (10) 20.2203(a)(2)(0 20.2203(ax3)(jj) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203 a)(4) 50.73(a) 2)(iv) 20.2203 a 2)(iii) 50.36(c 1 50.73(a)(2)(v) Specify In Abstract below or 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(ax2)(vii) in NRC Form 366A UCENSEE CONTACT FOR THIS LER (12)

TELEPHQNE NUMBER gncrude Area code)

Kenneth W. Frehafer, Licensing Engineer (561) 467 7748 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE RE ORTABLE TO EPIX CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX AB RV C710 YES SUPPLEMEN1'AL REPOR1'XPECTED 14 MONTH DAY YEAR EXPECTED YES SUBMISSION X NO 06 30 1999 (lf yes, complete EXPECTED SUBMISSION DATE). DATE (15)

ABSTRACT (Limit to 14(Uspaces, le., approximately 15single-spaced typewritten lines) (16)

On April 15, 1999, St. Lucie Unit 2 was in Mode 1 at approximately 50 percent reactor power. Wyle Labs informed FPL of unsatisfactory test results for the code pressurizer safety valves removed during the cycle 11 refueling outage. Wyle Labs was contracted to perform the offsite pressurizer safety valve testing and the testing was conducted within the required time restraints.

Technical Specification 3.4.2.1 requires the pressurizer safety valves to psia (+/-1 percent). The as-found settings of the removed St. Lucie Unit 2 lift at 2500 pressurizer safety valves were from 1.6 to 3.8 percent high, outside the Technical Specification tolerance limit of +/- 1 percent.

The root cause of the failed pressurizer safety valve testing is under investigation.

An LER supplement will be submitted once the root cause determinations are complete.

There is no present operability concern as the subject pressurizer safety valves were removed and replaced with pre-tested valves during the St. Lucie Unit 2 cycle 11 refueling outage. There was no affect on the health and safety of the public during past St. Lucie Unit 2 cycle 10 power operations because the limiting overpressure analyses remain bounded when actual St. Lucie Unit 2 cycle 10 operational parameters were considered.

9905240096 990517 PDR ADOCK 05000389 8 PDR NRC FORM 366 (6-igfrs)

NRC FORM 366A U.S. NUCLEAR REGUIATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET NUMBER (2 LER NUMBER (6) PAGE (3)

SEQUENTIAL REVISION NUMBER NUMBER St. Lucie Unit 2 05000389 Page 2 of 6 1999 - 004 - 00 TEXT (Ifmore space fs required. Use additional copies of NRC Form 366A) (17)

Description of the Event On April 15, 1999, St. Lucie Unit 2 was in Mode 1 at approximately 50 percent reactor power. Wyle Labs informed FPL of unsatisfactory test results for the code pressurizer safety valves (PSVs) [EIIS:AB:RV] "removed during the cycle 11 refueling outage.

'n accordance with the inservice testing (IST) program, pressure relief devices are tested per SPSI/ASME OM-1987, Part 1, "Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices." Section 1.3.3, "Test Frequency, Class 1 Pressure Relief Devices," of the code requires testing within 12 months of removal from service when the surveillance requirements are satisfied by installing a full complement of pre-tested valves. Wyle Labs was contracted to perform the testing and the testing was conducted within the required time restraints.

Technical Specification 3.4.2.1 requires the PSVs to percent). The as-found settings of the Unit 2 PSVs were outside the Technical lift at 2500 psia (+/-1 Specification tolerance limit of +/- 1 percent. As shown below, the deviation ranged from 1.6 to 3.8 percent high for the three valves.

Set Acceptable As-Found Valve Serial Number Pressure Range Set Result Pressure V1200 N84217-00-0005 2500 psia 2475-2525 psia 2539.7 psia 1. 6% High V1201 N84217-00-0008 2500 psia 2475-2525 psia 2593.7 psia 3.8% High V1202 N84217-00-0007 2500 psia 2475-2525 psia 2567.7 psia 2.7% High No present operability concern exists,'s the PSVs were all removed and replaced with pre-tested valves'uring the St. Lucie Unit 2 cycle 11 (SL2-11) refueling outage under work orders (WO) 98001961, 98001960, and 98001959.

Cause of the Event The root cause determination is not complete, but preliminary investigation indicates that the PSVs may be susceptible to setpoint drift.

Additionally, the ANSI/ASME OM-1987, Part 1, code requires that a'ause determination be performed and corrective actions implemented for any valve exceeding its nameplate setpressure by 3 percent or greater. Only one valve, V1201 (S/N N84217-00-0008),

exceeded this 3 percent threshold and is the first new block body valve to exceed this criteria. Preliminary investigation into this failure indicates that have been caused by an inadequate manufacturing process.

it may An LER supplement will be submitted when FPL completes the root cause determinations.

Ana1ysis of the Event FPL reviewed NUREG-1022, Revision 1, "Event Reporting Guidelines 10 CFR 50.72 and 50.73, " and determined that this event is reportable under 10 CFR 50 .73(a) (2) (i) (B) as "any operation or condition prohibited by the plant's Technical Specifications."

Although discrepancies found in Technical Specification surveillance tests should be assumed to occur at the time of the test, the existence of multiple sequential test NRC FORM 366A (6.)996)

I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

NUMBER (2 YEAR SEQUENTIAL REVISION NUMBER NUMBER St. Lucie Unit 2 05000389 Page 3 of 6 1999 - 004 - 00 TEXT (Ifmore spaceis required, use additional copies of NRC Farm 366A) (17)

Ana1ysis of the Event (cont'd) failures involving safety valves may be an indication that the discrepancies arose over a period of time. Therefore, the condition may have existed during plant operation.

Analysis of Safety Significance As described in the UFSAR, Section 5.4.13.2, the reactor coolant system (RCS) is protected against overpressure by protective and control devices such as the pressurizer spray system, the power operated relief valves, and the high-pressure reactor trip. ln addition to these features, three ASME Code PSVs ensure that RCS piping and components are protected from overpressure in accordance with ASME code requirements. No present operability concern exists, as the PSVs were all removed and replaced with pre-tested valves during the cycle 11 refueling outage.

An assessment of the accident analyses was performed to determine deviations could have led to the violation of overpressurization limits during the if the setpoint operation of cycle 10. The function of pressurizer safety valves in the safety analyses is to mitigate the consequences of overpressurization events by limiting peak pressure below the acceptance limits. The limiting overpressurization events are in the category of "Decrease in Heat Removal by the Secondary System." The limiting events in this category affected by deviations in PSV setpoints are the feedline break and loss of condenser vacuum analyses.

Feedline Break A revised feedline break analysis has recently been performed for St. Lucie Unit 2 as part of the reload process improvement (RPI) to be implemented for cycle 12. The RPI is described in FPL Letter L-98-308, "St. Lucie Unit 2, Docket No. 50-389, Proposed License Amendment, Cycle 12 Reload Process Zmprovement." The feedline break analysis of letter L-98-308 bounds the operation of cycle 10, the current cycle and anticipated future cycles. The revised feedline break analysis, which used a conservative PSV setpoint of 2575 psia, showed acceptable results with respect to the overpressurization criteria for primary and secondary systems. Since the average as-found setpressure of the PSVs to be evaluated is 2568 psia (i.e., <2575 psia), the results of the assessment bounds cycle 10 operation with these as-found setpoints (a lower PSV setpoint would open the valves earlier helping in the mitigation of the overpressurization event).

Loss of Condenser Vacuum This is a limiting pressurization event for St. Lucie Unit 2. Similar to the feedline break analysis, a revised loss of condenser vacuum analysis has recently been performed for St. Lucie Unit 2 as part of the RPZ to be implemented for cycle

12. This analysis bounds the operation of cycle 10, the current cycle, and anticipated future cycles. The revised loss of condenser vacuum analysis, which used a PSV setpoint of 2550 psia, showed acceptable results with respect to the overpressurization criteria for primary and secondary systems. Since the average as-found setpressure of the PSVs to be evaluated here is 2568 psia ()2550 psia), the inputs and assumptions of the analysis were evaluated to determine the impact of as-found setpressures that could have existed during cycle 10 operation (a higher PSV setpoint would open the valves later adversely impacting the mitigation of the overpressurization event).

NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME 0) DOCKET LER NUMBER (6) PAGE (3)

NUMBER 2 YEAR SEQUENTIAL REVISION NUMBER NUMBER St. Lucie Unit 2 05000389 Page 4 of 6 1999 - 004 - 00 TEXT (Ifmore space is required. use additional copies of NRC Form 3rMA) (17)

Analysis of Safety Significance (cont'd)

A review of the RPI analysis determined that several key parameters are modeled conservatively as compared to the actual values applicable for cycle 10 operation.

The values used in the RPI analysis and the corresponding cycle 10 values are listed below. Although a higher PSV setpoint would make the consequences worse, other parameters such as a higher initial pressurizer pressure, a lower analysis high pressure trip setpoint, a higher initial primary system temperature, and a less positive (or a negative) moderator temperature coefficient (MTC) would make the event less severe as described below.

~ Initial Pressurizer Pressure & High Pressure Trip Setpoint:

Increasing the initial pressure (>30 psi) and reducing the trip pressure (20 psi) will'result in an earlier reactor trip, substantially reducing the heat input into the RCS. An initial pressure of 2220 psia is consistent with the cycle 10 normal operating pressure of 2250 psia minus an uncertainty of 30 psi. Also, the high pressurizer pressure trip value of 2400 psia for this analysis is acceptable for cycle 10 which accounts for an uncertainty of 30 psi on the Technical Specification trip setpoint of 2370 psia. The reduction in heat input into the RCS due to an earlier xeactor trip (based on these input changes) is expected to offset any adverse effects of a later opening of PSVs by 18 psi.

~ Initial RCS Temperature:

Increasing initial RCS temperature will result in increased heat removal to the secondary system providing some beneficial impact on the calculated peak RCS pressure.

~ Moderator Temperature Coefficient:

A less positive or a negative MTC would result in lower reactor power leading to a slower pressure increase up to the time of reactor trip. This reactor power at the trip time is important in determining the final peak pressures. An analysis value of -2 pcm/ F bounds cycle 10 operation which was predicted to have a MTC of

<-5 pcm/ F at all times during cycle 10 at hot full power.

The impact of the above parameters was evaluated using RETRAN computer code. RETRAN is maintained by EPRI and is used extensively in the nuclear industry for non-LOCA safety analyses. RETRAN calculations were first done with the key parameter values the same as in the RPI analysis. RETRAN calculations were then repeated with the values changed .to the cycle 10 values, as shown below.

NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

NUMBER 2 YEAR SEQUENTIAL REVISION NUMBER NUMBER St. Lucie Unit 2 05000389 Page 5 of 6 1999 - 004 - 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366A) (17)

Analysis of Safety Significance (cont'd)

Parameter RPI Value Cycle 10 Value Initial Pressurizer Pressure, psia 2180 2220 Initial Core Inlet Temperature, F 535 545 High Pressurizer Pressure Trip Setpoint, psia 2420 2400 Pressurizer Safety Valve Opening 2550 2568 Pressure, psia Moderator Temperature Coefficient (MTC), pcm/ F +3.0 -2.0 Initial Core Thermal Power, MIOth 2754 2754 It was found that the impact of the higher PSV setpoints is more than offset by the parameter changes reflecting cycle 10 operation. The peak pressures for this event during cycle 10 operation thus would have remained bounded by the UFSAR analysis values.

RETRAN calculations showed that the peak RCS pressure is reduced by approximately 10 to 30 psi using cycle 10 specific conditions with as-found PSVs setpoints. The input changes addressed above have an insignificant impact on the peak secondary pressure.

Conclusion As discussed above, the limiting overpressure events were bounded once the actual St.

Lucie Unit 2 cycle 10 operational parameters were considered in the analyses.

Therefore, FPL concludes that the as-found PSV setpoints did not adversely affect the health and safety of the public during past cycle 10 operation.

NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (l ) DOCKET LER NUMBER (6) PAGE (3)

NUMBER 2 SEQUENTIAL REVISION NUMBER NUMBER St. Lucie Unit 2 05000389 Page 6 of 6 1999 - 004 - 00 TEXT (Ifmore spaceis required. use additional copies of NRC Form 366A) (17)

Corrective Actions

1. All three St. Lucie Unit 2 PSVs were replaced with,pre-tested valves during the cycle 98001959.

ll refueling outage (SL2-11) via work orders (WO) 98001961, 98001960, and

2. FPL will supplement this LER to include the root cause and corrective actions once the root cause determinations are complete.

Additional Information Failed Com onents Identified Component: pressurizer safety valve Manufacturer: Crosby Model: HB-86-BP, forged block body design, size 3K6, assembly N84217 Similar Events None NRC FORM 366A (6-1998)

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