05000266/LER-2010-001, Regarding Engineered Safety Features Steam Line Pressure Dynamics Modules Discovered Outside of Technical Specification Values
| ML101230151 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/30/2010 |
| From: | Meyer L Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2010-0074 LER 10-001-00 | |
| Download: ML101230151 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2662010001R00 - NRC Website | |
text
April 30,201 0 NRC 201 0-0074 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Unit 1 Docket 50-266 Renewed License No. DPR-24 Licensee Event Re~ort 2661201 0-001 -00 Enaineered Safetv Features Steam Line Pressure Dvnamics Modules Discovered Outside of Technical Specification Values Enclosed is Licensee Event Report (LER) 2661201 0-001 -00 for Point Beach Nuclear Plant (PBNP), Unit 1. This LER documents engineered safety features steam line pressure dynamic modules that were discovered to be outside of Technical Specifications (TS) values. Pursuant to 10 CFR 50.73(a)(2)(i)(B), the event is reportable as a condition prohibited by TS.
This submittal contains no new or revised regulatory commitments.
If you have questions or require additional information, please contact Mr. James Costedio at 9201755-7427.
Very truly yours, NextEra Energy Point Beach, LLC Q:-.v.
Site rry Vice Meyer President Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241
LICENSEE EVENT REPORT (LER)
- 4. TITLE Engineered Safety Features Steam Line Pressure Dynamics Modules Discovered Outside of Technical Specification Values C] 20.2203(a)(3)(i) 50.73(a)(2)(i)(C)
[7 50.73(a)(2)(vii)
C] 20.2203(a)(3)(ii)
C] 50.73(a)(2)(ii)(A)
C] 50.73(a)(2)(viii)(A)
C] 20.2203(a)(4)
C] 50.73(a)(2)(ii)(B)
C] 50.73(a)(2)(viii)(B)
C] 50.36(c)(l )(i)(A)
C] 50.73(a)(2)(iii)
C] 50,73(a)(2)(ix)(A) 50.36(c)(I)(ii)(A)
C] 50.73(a)(2)(iv)(A)
C] 50.73(a)(2)(x)
C] 50.73(a)(2)(v)(A)
C] 50.73(a)(2)(v)(B) 50.73(a)(2)(v)(C)
C] 50.73(a)(2)(v)(D)
- 12. LICENSEE CONTACT FOR THlS LER NAME TELEPHONE NUMBER (Include Area Code)
Ena Agbedia, Licensing Engineer 92017557654 I I MANU-I REPORTABLE
CAUSE
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SYSTEM / COMPONENT I Fsz&R I
T o EPlx 111
CAUSE
I SYSTEM COMPONENT FACTURER TO EPlX C] YES (If yes, complete 15. EXPECTED SUBMISSION DATE)
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ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On March 3, 2010, procedure 1 ICP 04.001 E, Reactor Protection and Safeguards Analog Racks Steam Pressure Refueling Calibration, was performed to satisfy Point Beach Nuclear Plant (PBNP)
Technical Specification (TS) Table 3.3.2-1 Function le, Surveillance Requirement (SR) 3.3.2.8.
The testing determined that the as-found values for five out of six Unit 1 engineered safety features (ESF) steam line pressure (SLP) dynamic compensation module lead time constants were slightly below the TS required time constant limit of greater than or equal to 12.0 seconds (the TS requirement is >/=I 2).
The basis calculation for module drift addressed static settings only and did not address dynamic settings. As a result of not addressing the dynamic settings, the tolerances for as-left settings were too restrictive and did not sufficiently account for instrument drift. Corrective actions included performing an on-line calibration check and adjusting the modules to within the allowable as-left tolerances. The Unit 2 SLP module tolerances were confirmed to be within TS-required values.
The out of specification lead time constants would not have prevented the associated channels from being able to perform their required safety-related functions in the event of a main steam line break (MSLB) accident. Pursuant to 10 CFR 50.73(a)(Z)(i)(B), the event is reportable as a condition prohibited by TS.
NRC FORM 366 (9-2007)
Point Beach Nuclear Plant
Event Description
On March 3, 2010, during the Unit 1 Refueling 32 outage (MODE 6) performance of 1lCP 04.001E, Reactor Protection and Safeguards Analog Racks Steam Pressure Refueling Calibration, it was discovered that the as-found values for five of the six Unit 1 steam line pressure (SLP) ESF instrument channel dynamic compensation module lead time constants [JE] were outside required Technical Specification (TS) values.
Event Analysis
TS LC0 3.3.2, ESFAS Instrumentation, Table 3.3.2-1, Function I
.e requires that three channels are required per steam line to be operable to provide safety injection (SI) during MODES 1 and 2, and in MODE 3 when pressurizer pressure is greater than 1800 psig where a secondary side break or stuck open valve could result in rapid depressurization of the steam lines. The function is not required to be operable in MODES 4, 5 or 6.
Steam line pressure-low provides protection against the MSLB, feed line break and inadvertent opening of a steam generator (SG) relief or safety valve. The steam line pressure-low provides a signal for control of the main steam atmospheric steam dump valves. A failure of a steam line pressure channel will not create a control failure that would result in a low steam line pressure SI event.
The lead constant value is required to be greater than or equal to 12 seconds, and the lag value is required to be less than or equal to 2 seconds. The ESFAS steam line pressure instruments monitor main steam line pressure and actuate on a 2-out-of 3 (213) steam line pressure-low condition to provide protection against a MSLB, main feedwater line break, or an inadvertent opening of a SG relief or safety valve.
The results of an evaluation determined that the dynamic response for all six of the compensation modules lead time constants had drifted low after being set within 0.021 seconds of their ideal setting. None of the Unit 1 modules exceeded the required TS lag value.
All six channels were found to be within the static calibration as-found tolerances. Therefore, the module outputs were able to reach the allowable field trip setpoint of 530 psig when actual process pressure was 748 psig, providing an actual margin of 41 3 psig to the safety limit of 335 psig. The downstream bistables would have tripped to produce the steam line pressure low SI actuations that are required prior to reaching the TS-required value of 500 psig within 1.I 31 3 seconds, which would provide a minimum margin of 165 psig to the analyzed safety limit of 335 psig.
Safety Significance
A rupture of a steam pipe is assumed to include any accident which results in an uncontrolled steam release from a SG. The release can occur due to a break in a pipe line or due to a valve malfunction. The steam release results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. With a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin. If the most reactive control rod is assumed stuck in its fully withdrawn position, there is a possibility that the core will become critical and return to power even with the remaining
Point Beach Nuclear Plant control rods inserted. A return to power following a steam pipe rupture is a potential problem only because of the high hot channel factors which may exist when the most reactive rod is assumed stuck in its fully withdrawn position. Assuming the worst case combination of circumstances which could lead to power generation following a steam line break, the core is ultimately shut down by the boric acid in the SI system.
The SI system actuates on 213 pressurizer low pressure signals; or 213 low pressure signals in any steam line; or 213 high containment pressure signals.
Based on the worst case as-found out of tolerance data obtained during the calibration procedure, the variations in the settings of the leadllag functions in the signal would not have been consequential should an actual event have occurred because all six channels remained capable of providing the TS-required steam line pressure low SI dynamic response within 1.1313 seconds. The module output would reach 530 psig within this time when actual system pressure is 748 psig. This provides a margin of 41 3 psig to the analyzed safety limit of 335 psig. This is within the required time of 1.5 seconds of the evaluated accident.
All six channels were within TS-required static calibration as-found tolerances. Therefore, the module outputs were able to reach the trip setpoint of 530 psig and their downstream Bistables would have tripped to produce the steam line pressure low SI actuations that are required prior to reaching the TS-required value of 500 psig. This would provide a minimum margin of 165 psig to the analyzed safety limit of 335 psig.
Accordingly, the safety significance of this event is low.
Cause
While the modules had drifted low over the last operating cycle, Engineering determined that over time the instrument tolerances had been tightened based upon increases in the accuracy of the calibration equipment and methods. The basis calculation for the SLP compensation modules had addressed only the static response settings and not the dynamic settings. Therefore, the apparent cause of the event was that the basis document for the static settings did not permit meaningful trending of calibration and surveillance monitoring data.
Corrective Action
The following corrective actions were taken:
- 1. ESFAS SLP channel dynamic compensation modules lead and combined constants were adjusted to within the current allowable as-left tolerances.
- 2. The Unit 2 ESFAS SLP channel dynamic compensation modules calibration data was reviewed as part of the extent of condition review. It was confirmed that all six identical Unit 2 modules were within required TS values. Therefore, no adjustments were required. At the time of the review, Unit 2 was in MODE I, operating at 100% power.
Point Beach Nuclear Plant Additional corrective actions to be taken include:
- 1. A trending and monitoring plan for the compensation modules will be developed and implemented that includes dynamic calibration data.
- 2. An on-line calibration check of the six Unit 1 compensation modules will be performed to assure that the settings have remained within acceptable tolerances.
- 3. The data obtained from the trending and monitoring plan will be used to develop a basis document for the static settings that supports the dynamic calibration method, ideal values and tolerances for the compensation modules.
- 4. Affected calibration procedures will be revised to incorporate new basis values.
Corrective actions have been entered into the site's corrective action program.
Previous Occurrences
A review of recent LERs identified the following previous conditions that involved leadllag time constants for steam line pressure outside technical specification Values:
LER Number Title -
ESFAS Instrumentation, LeadILag Time Constants for Steam Line Pressure outside Technical Specification Values Failed Components Identified: None.
Additional Information
None.