IR 05000387/2013004
| ML13318A960 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/14/2013 |
| From: | Fred Bower Reactor Projects Region 1 Branch 4 |
| To: | Rausch T Susquehanna |
| BOWER, FL | |
| References | |
| IR-13-004 | |
| Download: ML13318A960 (37) | |
Text
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION ber 14, 2013
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION - NRC INTEGRATED INSPECTION REPORT 05000387/2013004 AND 05000388/2013004
Dear Mr. Rausch:
On September 30, 2013 the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Susquehanna Steam Electric Station (SSES) Units 1 and 2. The enclosed integrated inspection report documents the inspection results, which were discussed on October 10, 2013, with Mr. Jeffrey Helsel, Plant Manager, and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents three NRC-identified and one self-revealing finding of very low safety significance (Green). All of these findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because they are entered into your corrective action program (CAP), the NRC is treating these findings as non-cited violations (NCVs) consistent with Section 2.3.2 of the NRCs Enforcement Policy. If you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Senior Resident Inspector at the SSES. In addition, if you disagree with the cross-cutting aspect of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Senior Resident Inspector at the SSES. In accordance with the Code of Federal Regulations (10 CFR) 2.390 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely, /RA/ Fred L. Bower, III, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket Nos. 50-387; 50-388 License Nos. NPF-14, NPF-22
Enclosures:
Inspection Report 05000387/2013004 and 05000388/2013004 w/Attachment: Supplemental Information
REGION I== Docket No: 50-387, 50-388 License No: NPF-14, NPF-22 Report No: 05000387/2013004 and 05000388/2013004 Licensee: PPL Susquehanna, LLC (PPL) Facility: Susquehanna Steam Electric Station, Units 1 and 2 Location: Berwick, Pennsylvania Dates: July 1, 2013 through September 30, 2013 Inspectors: J. Greives, Senior Resident Inspector P. Finney, Senior Resident Inspector, Salem A. Turilin, Resident Inspector S. Kennedy, Senior Resident Inspector, Calvert Cliffs K. Mangan, Senior Reactor Inspector J. Laughlin, Emergency Preparedness Inspector, NSIR E. Burket, Emergency Preparedness Inspector Approved By: Fred L. Bower, Chief Reactor Projects Branch 4 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS
IR 05000387/2013004 05000388/2013004 07/01/2013 - 09/30/2013; Susquehanna Steam
Electric Station, Units 1 and 2; Licensed Operator Requalification Program, Maintenance Risk Assessment and Emergent Work Control, Problem Identification and Resolution, Follow-up of Events and Notices of Enforcement Discretion.
The report covered a three-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. Inspectors identified four non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process(SDP), dated June 2, 2011. The cross-cutting aspects for the findings were determined using IMC 0310, Components Within The Cross-Cutting Areas, dated October 28, 2011. Findings for which the SDP does not apply may be Green, or be assigned a severity level after Nuclear Regulatory Commission (NRC) management review. All violations of NRC requirements are dispositioned in accordance with NRCs Enforcement Policy, dated June 7, 2012. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process (ROP), Revision 4, dated December 2006.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) because PPL did not adequately assess the risk of performing maintenance in accordance with station procedures. Specifically, PPL did not specify appropriate risk management actions (RMAs)while performing a standby liquid control (SLC) system flow surveillance in conjunction with having the E emergency diesel generator (EDG) unavailable. PPLs immediate corrective actions included entering the issue into their CAP as condition reports (CRs) 1721928 and 1781929, communicating the issue to applicable station personnel, and revising the risk assessment for use in future performance of the maintenance activities.
The performance deficiency is more than minor because it affected the Human Performance attribute of the Mitigating Systems cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
The item is similar to example 7.e. in NRC IMC 0612 Appendix E, Examples of Minor Issues. This example states, in part, that failure to perform an adequate risk assessment when required by 10 CFR 50.65 (a)(4) is not minor if the overall elevated plant risk would require, under plant procedures, RMAs or additional RMAs. In this case, the SLC flow surveillance was required to be screened as high operational risk due to the short duration limiting condition of operation (LCO) entry and medium or high operational risk due to changing risk to Yellow when performed in conjunction with the E EDG unavailability.
Both of these categories required additional RMAs in accordance with station procedures.
In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was associated with RMAs only and the incremental core damage probability was < 1E-6 and the incremental large early release probability was < 1E-7. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Control in that PPL failed to appropriately plan work activities by not incorporating risk insights. Specifically, PPL did not appropriately assess the risk of performing maintenance activities by including required risk manage actions as specified in station procedures. [H.3(a)]. (Section 1R13)
- Green.
The inspectors identified a Green NCV of TS 5.4.1, Procedures, because PPLs emergency operating procedure step for terminating injection sources during a rapid depressurization required for an anticipated transient without scram (ATWS) was inadequate to ensure that cold unborated water was not injected into the core. Specifically, PPLs emergency operating procedure (EOP) does not terminate injection from the high pressure coolant injection (HPCI) system during the transient and procedural guidance is insufficient to ensure that operators will maintain level in the prescribed ATWS band while injecting with HPCI. In addition to entering the issue into the CAP as CRs 1708885 and 1745775, PPLs immediate corrective actions included issuance of Operations Directive 13-02 which states that HPCI must be controlled, up to and including overriding injection, to ensure that reactor pressure vessel water level is maintained in the prescribed ATWS band during the duration of the rapid depressurization. Planned corrective actions include requiring termination of HPCI injection prior to initiation of a rapid depressurization (Action Request 1719605).
The performance deficiency is more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedure for terminating injection prior to rapidly depressurizing the reactor during an ATWS could have resulted in operators failing to control level in the prescribed EOP band, potentially resulting in cold unborated water being injected into the core. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding is related to the cross-cutting area of problem identification and resolution (PI&R), in that PPL did not identify a performance issue completely, accurately, and in a timely manner commensurate with the safety significance. Specifically, PPL failed to identify that guidance in EOP basis document was insufficient to ensure that operators maintained level in the EOP band. [P.1(a)]. (Section 1R11)
Cornerstone: Barrier Integrity
- Green.
The inspectors identified a green, self-revealing, non-cited NCV of 10 CFR 50 Appendix B, Criterion 5, Instructions, Procedures, and Drawings, which states, in part, that procedures shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. The inspectors determined that PPLs residual heat removal (RHR) shutdown cooling procedure failed to ensure that water properties (pressure and temperature) in the suction piping was controlled to ensure water hammer event would not happen when establishing a low pressure injection standby lineup. As a result, a water hammer occurred in the piping which caused the suction relief valve to fail open. PPLs immediate corrective actions included entering the issue into their CAP as CRs 1746612 and 1754913, replacing the relief valve, walking down the piping and associated supports and communicating to operations personnel to declare RHR inoperable when aligned to shutdown cooling (SDC) while reactor coolant temperature is above 200 degrees Fahrenheit.
This finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the water hammer event resulted in a stuck open relief valve on the RHR suction piping whose leak rate exceeded the assumed leakage from engineered safeguard systems in PPLs post-event control room dose calculations. Because conditions for RHR system operation had been established, the team assessed this finding in accordance with the NRC Inspection Manual Chapter (IMC)0609, Significance Determination Process, Appendix G, Shutdown Operations Significance Determination Process, using Attachment 1, Checklist 5. The finding did not require a quantitative assessment because none of the checksheet guidelines requiring a phase 2 analysis were affected. Therefore, the finding was determined to be of very low safety significance (Green). The finding had a cross cutting aspect in the problem identification and resolution area associated with operating experience because PPL did not implement and institutionalize operating experience through changes to station processes, procedures, equipment, and training programs. Specifically, PPLs review of IN 2010-11 did not ensure the transition of RHR from SDC to LPCI standby was completed successfully by incorporating adequate steps into the operating procedure. [P.2(b)]. (Section 4OA3)
Cornerstone: Emergency Preparedness
- Green.
The inspectors identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR 50.54(q) for failing to follow and maintain an emergency plan that meets the requirements of emergency planning standard 10 CFR 50.47(b)(4).
Specifically, the licensee failed to take timely corrective actions to restore a degraded room flooded alarm in accordance with station procedures. The alarm was out-of-service from December 21, 2012 until September 23, 2013 without adequate compensatory measures in place. PPLs immediate corrective actions included entering the issue into their CAP as CR 1745962, changing the priority of the work order (WO) and listing it as a priority item on their Daily Leadership Alignment Package. PPL replaced the detector on September 23, 2013.
The performance deficiency is more than minor because it was associated with the facilities and equipment attribute of the Emergency Preparedness cornerstone and affected the objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.
Specifically, the performance deficiency would have resulted in untimely declaration of an Alert OA5 and Notice of Unusual Event (NOUE) OU5. In accordance with NRC IMC 0609, Appendix B, Emergency Preparedness SDP, the inspectors determined that this finding is of very low safety significance (Green) because it did not result in the loss or degradation of a risk significant planning standard. Specifically, one Alert and one NOUE EAL initiating condition would have been rendered ineffective such that a flooding event would have been declared in a degraded manner. The finding is related to the cross-cutting area of PI&R, CAP, in that PPL did not take appropriate corrective actions to address safety issues in a timely manner. Specifically, when the detector failed on December 21, 2012, adequate compensatory measures were not specified and the WO was not scheduled for completion for 12 months. [P.1(d)]. (Section 4OA2)
REPORT DETAILS
Summary of Plant Status
Unit 1 began the inspection period at 100 percent power. On July 10, 2013, operators reduced power to approximately 96 percent to mitigate the unexpected closure of a main turbine control valve (CV). Following discussion with the main turbine vendor, operators conducted a further unplanned power reduction to 75 percent on July 16, 2013 due to maintain margin to high pressure turbine blade vibration limits. On July 20, 2013, operators reduced power to 15 percent and removed the main turbine from service to perform planned repairs on the main turbine CV. Operators returned the unit to 100 percent on July 24, 2013. On August 27, 2013, the A reactor feed pump turbine automatically tripped which resulted in a reactor recirculation pump runback and an automatic unplanned power reduction to 65 percent. Operators returned the unit to 100 percent power the following day. On September 12, 2013, operators reduced power to 65 percent for a planned rod sequence exchange, returning to 100 percent the same day. On September 17, 2013, operators reduced power to 85% to allow isolation of a main condenser water box for maintenance. Power was restored to 100 percent on September 19, 2013 and remained at or near 100 percent power for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent power. On July 5, 2013, an automatic reactor recirculation rundown occurred due to a condensate system perturbation, resulting in a power reduction to 96 percent. Operators restored power to 100 percent the same day. On July 6, 2013, operators reduced power to approximately 65 percent at the request of the grid operator.
Operators returned the unit to 100 percent on July 8, 2013. On July 27, 2013, operators again reduced power to 65 percent at the request of the grid operator. Operators restored power to 100 percent on July 28, 2013. On September 13, 2013, operators commenced a reactor shutdown to perform repairs to the low pressure turbines. On September 14, 2013, operators manually scrammed the reactor during the shutdown when an unexpected reactor water level transient occurred while transitioning the A reactor feed pump from flow control mode to discharge pressure mode. On September 15, 2013, operators received a flood alarm associated with the B RHR room and declared an Unusual Event. Operators terminated the event by isolating appropriate portions of the RHR system. Following the completion of the maintenance activities, operators commenced a reactor startup on September 22, 2013.
Operators returned the unit to 100 percent power on September 27, 2013, and remained at or near 100 percent power for the remainder of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
==1R04 Equipment Alignment
.1 Partial System Walkdowns
a.
==
Inspection Scope The inspectors performed partial walkdowns of the following systems: Unit 2, Division II RHR Common, B EDG during A EDG unavailability Unit 2, Division II RHR in response to water hammer transient The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the Updated Final Safety Analysis Report (UFSAR), TSs, WOs, condition reports (CRs), and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether PPL staff had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization.
b. Findings
No findings were identified. ==1R05 Fire Protection