05000400/LER-2013-002, Regarding Main Steam Safety Valve Setpoint Drift
| ML13353A628 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/19/2013 |
| From: | Kapopoulos E Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-13-129 LER 13-002-00 | |
| Download: ML13353A628 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 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)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) |
| 4002013002R00 - NRC Website | |
text
Ernest J. Kapopoulos, Jr.
Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill NC 27562-9300 919.362.2502 December 19, 2013 10 CFR 50.73 Serial: HNP-13-129 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 Subject: Licensee Event Report 2013-002-00 Main Steam Safety Valve Setpoint Drift Ladies and Gentlemen:
Duke Energy Progress, Inc. submits the enclosed Licensee Event Report 2013-002-00 in accordance with 10 CFR 50.73 for the Shearon Harris Nuclear Power Plant, Unit 1, which describes a condition where two main steam safety valve setpoints were found outside the technical specification tolerance.
This document contains no regulatory commitments. Please refer any questions regarding this submittal to Dave Corlett at (919) 362-3137.
Sincerely, Ernest J. Kapopoulos, Jr.
Enclosure: LER 2013-002-00 cc:
Mr. J. D. Austin, NRC Sr. Resident Inspector, Harris Nuclear Plant Mr. A. Hon, NRC Project Manager, Harris Nuclear Plant Mr. V. M. McCree, NRC Regional Administrator, Region II
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NRC FORM 366 (10-2010)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments regarding burden estimate to the FOIA/Privacy Section (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects.resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means used to impose 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.
LICENSEE EVENT REPORT (LER)
(See reverse for required number of digits/characters for each block)
- 1. FACILITY NAME Shearon Harris Nuclear Power Plant, Unit 1
- 2. DOCKET NUMBER 05000400
- 3. PAGE 1 of 4
- 4. TITLE Main Steam Safety Valve Setpoint Drift
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.
MONTH DAY YEAR FACILITY NAME None DOCKET NUMBER 10 23 2013 2013 - 002 - 00 12 19 2013 FACILITY NAME None DOCKET NUMBER
- 9. OPERATING MODE 1
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201(b) 20.2201(d) 20.2203(a)(1) 20.2203(a)(2)(i) 20.2203(a)(2)(ii) 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(2)(v) 20.2203(a)(2)(vi) 20.2203(a)(3)(i) 20.2203(a)(3)(ii) 20.2203(a)(4) 50.36(c)(1)(i)(A) 50.36(c)(1)(ii)(A) 50.36(c)(2) 50.46(a)(3)(ii) 50.73(a)(2)(i)(A) 50.73(a)(2)(i)(B) 50.73(a)(2)(i)(C) 50.73(a)(2)(ii)(A) 50.73(a)(2)(ii)(B) 50.73(a)(2)(iii) 50.73(a)(2)(iv)(A) 50.73(a)(2)(v)(A) 50.73(a)(2)(v)(B) 50.73(a)(2)(v)(C) 50.73(a)(2)(v)(D) 50.73(a)(2)(vii) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(ix)(A) 50.73(a)(2)(x) 73.71(a)(4) 73.71(a)(5)
OTHER Specify in Abstract below or Causal Factors The cause was determined to be setpoint drift incompatible with analysis specified criteria. As described in Electric Power Research Institute report TR-105872S and Nuclear Regulatory Commission Information Notice 2006-24, setpoint drift is a common phenomenon and the timing or degree of the drift cannot be predicted or determined with a high level of confidence. The common causal factors are an aggregate of aging, binding, bonding, and corrosion. Each of these alone are failure causes in their classic definition. The factors above preclude accurate determination or prediction of when and how much setpoint drift can occur on the MSSVs. Therefore, the MSSV setpoint out of tolerance conditions at Harris are principally driven by the close tolerance between technical specification requirements and the ability of the valve to perform within the required pressure band.
Corrective Actions
Upon discovery of the out of tolerance conditions, the two setpoints were adjusted to within technical specification tolerances which restored compliance with the technical specifications.
Because the setpoint drift does not have a highly reliable resolution path, avoiding the condition of MSSVs not meeting acceptance criteria can be accomplished by revision of the safety analysis and changing the Technical Specification to accommodate the observed drift. The corrective action to preclude recurrence will be implementation of a revised safety analysis that accommodates increased setpoint drift and supports revised technical specification setpoints.
Safety Analysis
The MSSVs are used to satisfy American Society of Mechanical Engineers (ASME) Code requirements for overpressure protection and are designed to prevent the system pressure from exceeding 110% of the design operating pressure. The valves are also credited in mitigating the effects of postulated accidents (e.g., loss of external electrical load and loss of normal feedwater).
The consequences of exceeding the ASME pressure limit could include damage to system components, leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.
The applicable limits are 110% of the main steam design pressure for American Nuclear Society (ANS) condition I and II events and 120% of the main steam system design pressure for ANS condition III and IV events. The limits are documented in the Harris Final Safety Analysis Report (FSAR), Chapter 15, for the respective events. As documented in the Safety Analysis Report for Cycle 18, the most limiting transients are turbine trip for ANS condition II events, and main feedline break for ANS condition III and IV events.
The safety analyses are predominantly concerned with main steam system overpressure. As a consequence, the out of tolerance reading for 1MS-44 (slightly below 1% tolerance) does not produce a result which is outside of the safety analysis.
The event with the smallest amount of overpressure margin is turbine trip at 18.9 pounds per square inch. The lowest margin is for a case with Temperatureaverage at 588.8° Fahrenheit. The event is analyzed crediting all the MSSVs at the nominal setting with the technical specification tolerance of
+1.0% added. All five stages of the MSSVs open for the limiting event. Therefore, all the safety valves would have to exhibit measured setpoints higher than the technical specification tolerance of
+1% to invalidate the analysis of record. Therefore, the safety analysis remains valid, and the impact on safety is very minor.
Additional Information
No previous Harris licensee event reports describing main steam safety valve setpoints outside technical specification tolerances within the last ten years were identified. Although isolated cases of MSSV setpoints were found outside of technical specification tolerances in the last three years, the isolated discrepancies did not meet reportability criteria. Multiple cases of MSSV setpoints outside technical specification tolerance did occur in 2001 and 2006, but a root cause analysis was not performed for those conditions because the failures were within the tolerance of the ASME code.
Therefore, corrective actions for the 2001 and 2006 events were not expected to prevent these conditions.
As stated previously, Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
This report contains no regulatory commitments.