IR 05000387/2008004
| ML083190088 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/13/2008 |
| From: | Paul Krohn Reactor Projects Region 1 Branch 4 |
| To: | Spence W PPL Corp |
| KROHN P, RI/DRP/PB4/610-337-5120 | |
| References | |
| IR-08-004 | |
| Download: ML083190088 (32) | |
Text
November 13, 2008
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION - NRC INTEGRATED INSPECTION REPORT 05000387/2008004 AND 05000388/2008004
Dear Mr. Spence:
On September 30, 2008, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Susquehanna Steam Electric Station Units 1 and 2. The enclosed integrated inspection report presents the inspection results, which were discussed on October 16, 2008, with Mr. Ronald Smith 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 two NRC-identified findings of very low safety significance (Green).
Both findings were determined to involve violations of NRC requirements. Additionally, a licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs), consistent with Section VI.A.1 of the NRC 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 Resident Inspector at the Susquehanna Steam Electric Station.
In accordance with 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 document 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/
Paul G. Krohn, Chief Projects Branch 4 Division of Reactor Projects
Docket Nos.
50-387; 50-388 License Nos. NPF-14, NPF-22
Enclosures:
Inspection Report 05000387/2008004 and 05000388/2008004
Attachment: Supplemental Information
REGION I==
Docket No:
50-387, 50-388
License No:
Report No:
05000387/2008004 and 05000388/2008004
Licensee:
Facility:
Susquehanna Steam Electric Station, Units 1 and 2
Location:
Berwick, Pennsylvania
Dates:
July 1, 2008 through September 30, 2008
Inspectors:
F. Jaxheimer, Senior Resident Inspector
A. Rosebrook, Resident Inspector
G. Newman, Resident Inspector
A. Ziedonis, Reactor Inspector
J. Furia, Senior Health Physicist
K. Young, Senior Reactor Inspector
Approved By:
Paul G. Krohn, Chief
Projects Branch 4
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000387/2008004, 05000388/2008004; 07/01/2008 - 09/30/2008; Susquehanna Steam
Electric Station, Units 1 and 2; Operability Evaluations and Adverse Weather Protection.
The report covered a 3-month period of inspection by resident inspectors, two regional reactor inspectors, and a senior health physicist. Two Green NCVs were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a NCV of Susquehanna Unit 1 Technical Specifications (TS) 3.5.1 and 3.5.3 for rendering high pressure coolant injection (HPCI)and reactor core isolation cooling (RCIC) inoperable during a planned shutdown.
Specifically, both the HPCI and RCIC systems were made inoperable to fulfill their TS described safety function when operators raised reactor vessel level above the HPCI and RCIC turbine trip signals in a plant mode and at a plant pressure where both of these systems were required to be fully operable. PPL Susquehanna, LLC (PPL)initiated corrective actions to revise the shutdown procedure to prevent reactor vessel water level from being raised above the trip input level until low pressure (LP)emergency core cooling system (ECCS) are capable of being used.
This finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was evaluated for significance using IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Since the finding did not result in a loss of safety function or the loss of a train for greater than its TS allowed outage time, and was not potentially risk significant due to external event initiators, the finding was determined to be of very low safety significance (Green). This finding is related to the cross-cutting area of Human Performance - Resources because PPL did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. (H.2(c)) (1R15)
Cornerstone: Emergency Preparedness
- Green.
The inspectors identified a NCV associated with emergency planning standard 10 CFR 50.47(b)(4). The inspectors determined that a performance deficiency existed in that inadequate indications were available for operators to determine if a threshold for emergency action levels (EALs) based on sustained wind speed in the protected area, had been met. On the afternoon of July 17, 2008, a severe thunderstorm with winds in excess of 50 miles per hour (mph) passed though the plant site. The storm caused damage to non-vital structures and resulted in the loss of two, 13.2 kilovolts (kV) power lines which interrupted power to several non-power block buildings on site. Inspectors observed operators responding to the event and identified that the wind speed indicator in the control room had indicated the maximum value for several minutes. This recorder only displayed wind speeds up to a maximum of 50 mph. Inspectors also observed that the backup wind speed indication, located 6 miles from the site and which reads from 0-100 mph, was inoperable during the storm. Inspectors identified that the Unit Supervisor had mistakenly read the wind direction trace on the recorder and had determined a 65 mph wind speed in error. Based upon direct observations during this adverse weather event, the inspectors determined that the operators did not have adequate indications available to determine if the threshold, sustained winds of greater than 80 mph, for EALs OA5 or OU5, had been met.
This finding is greater than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding is of very low safety significance (Green) because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. This finding is related to the cross-cutting area of Problem Identification and Resolution - Corrective Action Program because PPL did not take appropriate corrective actions to address a safety issue in a timely manner, commensurate with its safety significance and complexity. Specifically, the NRC had previously identified this potential vulnerability over two years prior to the event and the licensee had entered the concern into their CAP; however, corrective actions were not implemented. P.1(d)
(Section 1R01)
Licensee Identified Violations
A violation of very low safety significance, which was identified by PPL has been reviewed by the inspectors. Corrective actions taken or planned by PPL have been entered into PPL's corrective action program. This violation and corrective action tracking numbers are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Susquehanna Steam Electric Station (SSES) Unit 1 began the inspection period at the authorized power level of 94.4 percent licensed reactor thermal power (RTP). On July 4, 2008, Unit 1 reactor power was reduced to approximately 75 percent RTP for one day to isolate the B feedwater heater string from service and repair a feedwater heater level control valve. Unit 1 returned to operate at 94.4 percent RTP until an unplanned outage on August 15, 2008, to repair an electro-hydraulic (EHC) leak on the # 4 main turbine control valve. Unit 1 reactor power was reduced to approximately 18 percent power and the main generator was taken offline to facilitate repair of the EHC leak. The Unit 1 generator was placed back online on August 17, 2008, and RTP returned to 94.4 percent on August 18, 2008. Unit 1 remained at the authorized power level of 94.4 percent for the remainder of the inspection period with the exception of a one-day planned power reduction to approximately 75 percent power on September 6, 2008, for control rod scram testing and a control rod sequence exchange.
Unit 2 began the inspection period at RTP. With the exception of a one-day power reduction to approximately 70 percent RTP on September 20, 2008 for control rod scram time testing, circulating water box cleaning, and a rod sequence exchange, Unit 2 remained at full RTP through the remainder of the inspection period.
Note: The licensed thermal power for Unit 1 is 3952 megawatts thermal. The extended power uprate (EPU) License Amendment for SSES was approved on January 30, 2008, and was implemented on Unit 1 in accordance with issued license conditions. For the current operating cycle, the Unit 1 authorized power level is 94.4 percent of the EPU licensed power limit. For the purposes of this report, full RTP for Unit 2 remains at 3489 megawatts thermal since EPU power level increases have not yet been implemented on that unit.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
==1R01 Adverse Weather Protection (71111.01 - 1 Sample)
==
.1 Adverse Weather - Impending Adverse Weather Conditions
a. Inspection Scope
On the afternoon of July 17, 2008, a severe thunderstorm with winds measured in excess of 50 miles per hour (mph) passed though the plant site. The storm caused damage to non-vital structures and resulted in the loss of two, 13.2 kV power lines, which interrupted power to several buildings on-site outside the power block. The inspectors responded to the control room, and observed operator response to the weather conditions and alarm response to the subsequent power loss. Inspectors performed plant walkdowns for selected structures, systems, and components (SSCs)to determine the adequacy of PPLs weather protection features. Inspectors reviewed operator actions to address failures of equipment and compensatory actions during and following the high wind conditions. The inspectors also reviewed and evaluated plant conditions resulting from the high wind and reviewed considerations in PPLs risk assessment and Emergency Plan. Documents reviewed are listed in the Attachment.
The adverse weather protection sample included:
- Windstorm (severe thunderstorm on July 17, 2008) with loss of 13.2 kV offsite power.
b. Findings
Introduction:
The inspectors identified a Green NCV associated with emergency planning standard 10 CFR 50.47(b)(4). The inspectors determined that a performance deficiency existed in that inadequate indications were available for operators in the control room to determine if a threshold for an alert or notice of unusual event (NOUE)emergency action level (EAL), based on sustained wind speed in the protected area, had been met. This issue did not result in the loss or degradation of a risk significant planning standard based on the inspectors assessment of the criteria in IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process.
Description:
On the afternoon of July 17, 2008, a severe thunderstorm with winds measured in excess of 50 miles per hour (mph) passed though the plant site. The storm caused damage to non-vital structures and resulted in the loss of two, 13.2 kV power lines which interrupted power to several buildings on-site outside the power block. The inspectors responded to the control room, and observed the operator response to the weather conditions and alarm response to the subsequent power loss during the event.
The inspectors noted that the Unit Supervisor reported to the Shift Manager that the maximum wind speed was 65 mph. Based on that information, the Shift Manager determined that entry conditions for EAL OU5 (NOUE for Tornado or sustained wind speeds of greater than 80 mph inside the protected area) had not been met.
After the peak of the storm had passed, the inspectors identified that the wind speed indicator on the control panel had indicated its maximum value for several minutes. The recorder only displayed wind speeds up to 50 mph. A backup wind speed indication, which reads from 0 to100 mph receives a signal from a tower in Nescopeck, PA, approximately 5 miles from the site. This tower lost power during the storm and that signal was lost. Site personnel did not report seeing a funnel cloud or any indication of a tornado.
The inspectors identified that the Unit Supervisor had mistakenly read the wind direction instead of wind speed on the panel recorders and had determined the 65 mph wind speed in error. Based upon damage to several non-vital structures (Supplemental Decay Heat Removal enclosure, and non-security related fencing), and other plant parameters (cooling tower basin water levels fluctuating over 12 inches during the thunderstorm, the residents concluded the 65 mph wind call was a reasonable estimate.
Nonetheless, the inspectors determined that the operators did not have adequate indications available in the control room to determine if the threshold for EALs OA5 or OU5 had been met.
PPL procedure EP-TP-001, EAL Classification Levels, Revision I, defines sustained winds as wind speeds which exceed the defined threshold value for greater than one minute. Based on the wind speed trace in the control room and the cooling tower water level fluctuation data, it was established that there were sustained winds in excess of 50 mph for 5 to 10 minute periods inside the protected area. However, maximum sustained wind speed could not be determined due to the Nescopeck Tower losing power and Susquehanna Met tower limitations. Therefore, there was no direct indication to determine whether the EAL threshold was reached and there was no violation for failure to declare an emergency and inform state and local officials.
In April 2006, the NRC had previously identified this vulnerability to PPL and the issue was entered into their corrective action system as CR 754193. PPL had initiated a modification (AR 758005758005 to evaluate the issue and provide direct indication of wind speed from the Met tower. PPL determined that this could be done by upgrading cards to the existing Met tower and rescaling the indications in the control room. However, this modification was not completed.
The inspectors determined that having no means of indication readily available to the operators in the control room to determine if sustained wind speeds were in excess of the EAL threshold during an actual event was a performance deficiency and that this was reasonably within the licensees ability to foresee and prevent due to the vulnerability being previously identified and entered into their CAP in 2006.
Analysis:
The failure to provide adequate indication for assessment of EAL entry criteria could impact the timely declaration of an emergency and is contrary to 10 CFR 50.54(q)and 10 CFR 50.47(b)(4). Traditional enforcement does not apply because there were no actual safety consequences, the violation was not willful, and it did not impact the NRCs ability to regulate because the inspectors were present and aware of the event.
This finding is greater than minor because it was associated with the Emergency Preparedness (EP) cornerstone attribute of Facilities and Equipment, and affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply. This finding is associated with a failure to meet or implement a regulatory requirement. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. Section 4.4 of IMC 0609, Appendix B, provides examples for use in assessing emergency preparedness related findings.
One example of a Green finding states, The EAL classification process would not declare any Alert or Notification of Unusual Event that should be declared. Since the declaration of an Alert or NOUE based on sustained wind speed in the protected area of greater than 80 mph could have been missed or delayed, this finding was considered consistent with the example provided and was determined to be of very low safety significance (Green).
This finding is related to the cross-cutting area of Problem Identification and Resolution - Corrective Action Program because PPL did not take appropriate corrective actions to address a safety issue in a timely manner, commensurate with its safety significance and complexity. Specifically, the NRC had previously identified this potential vulnerability over two years prior to the event and the licensee had entered the concern into their CAP; however, corrective actions were not implemented. This was determined to be the most significant contributing factor to this issue. P.1(d)
Enforcement:
10 CFR 50.54(q) requires that the facility licensee follow and maintain in effect emergency plans which meet the standards in 10 CFR 50.47(b). 10 CFR 50.47(b)(4) requires, in part, that emergency response plans include a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters. The emergency classification and action level scheme is required to be used by the nuclear facility licensee, and State and local response plans rely on information provided by facility licensees for determinations of minimum initial offsite response measures. Contrary to the above, on July 17, 2008, PPL did not have adequate means of indication in the control room to support an EAL classification based on sustained wind speed in the protected area. PPL entered this issue into its CAP as AR 1053296 and is evaluating the development of permanent corrective actions. Because this issue is of very low safety significance (Green) and has been entered into PPL's CAP, it is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000387 & 05000388/2008004-01, Inadequate Equipment to Assess Threshold for Emergency Action Level)
==1R04 Equipment Alignment (71111.04 - 4 Samples)
==
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial walkdowns to verify system and component alignment and to identify any discrepancies that would impact system operability. The inspectors verified that selected portions of redundant or backup systems or trains were available while certain system components were out of service. The inspectors reviewed selected valve positions, electrical power availability, and the general condition of major system components. The walkdowns included the following systems:
- Unit 1, standby liquid control (SLC) system, skid components, and control room and local indications;
- Station, manhole cable vault inspections; manhole #8 - startup transformer, manhole #19 - EDG, and manhole #31 - emergency service water (ESW).
b. Findings
No findings of significance were identified.
.2 Complete Walkdown
a. Inspection Scope
The inspectors conducted a detailed review of the alignment and condition of the B emergency diesel generator (EDG). The inspectors reviewed operator rounds, check-off lists, system operating procedures, operator logbooks, and system piping and instrumentation diagrams (P&IDs). The inspectors evaluated ongoing maintenance (other divisions) and outstanding condition reports associated with the system.
Inspectors evaluated system health and reliability. Documents reviewed are listed in the
. The walkdown included the following system:
b Findings
No findings of significance were identified.
==1R05 Fire Protection (71111.05Q - 6 Samples)
==
.1 Fire Protection - Tours
a. Inspection Scope
The inspectors reviewed PPLs fire protection program to evaluate the required fire protection design features, fire area boundaries, and combustible loading requirements for selected areas. The inspectors walked down those areas to assess PPLs control of transient combustible material and ignition sources, fire detection and suppression capabilities, fire barriers, and any related compensatory measures to assess PPL's fire protection program in those areas. The inspected areas included:
- Unit 1, upper cable spreading room and upper relay room;
- Unit 1, lower relay room (FP-013-139) and computer room (FP-013-140);
- Unit 2, B residual heat removal (RHR) room during welding operations, FP-113-106;
- Security control center, FP-013-360;
- E EDG building, FP-013-236; and
- Fire protection simplex panel inoperable on September 20, 2008.
b. Findings
No findings of significance were identified.
==1R07 Heat Sink Performance (71111.07A - 1 Sample)
a. Inspection Scope
==
The inspectors reviewed PPLs evaluations for the as-found condition of the Susquehanna Ultimate Heat Sink. The inspectors observed spray pond diving operations and reviewed PPLs reports related to the spray pond structural inspection, spray pond sedimentation study, and spray pond bio-fouling evaluation (zebra mussels).
Documents reviewed are listed in the Attachment. The annual heat sink performance sample included:
- Station ultimate heat sink, spray pond diving operations and inspection.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11 - 1 Sample)
Resident Inspector Quarterly Review
a. Inspection Scope
On August 14, 2008, the inspectors observed licensed operator simulator training during routine operator requalification training. The inspectors compared their observations to TSs, emergency plan implementation, and the use of system operating procedures.
The inspectors also evaluated PPLs critique of the operators' performance to identify discrepancies and deficiencies in operator training. The following training was observed:
- Operations crew E, SA-9 simulator activity.
b. Findings
No findings of significance were identified.
==1R12 Maintenance Effectiveness (71111.12 - 2 Samples)
a. Inspection Scope
==
The inspectors evaluated PPLs work practices and followup corrective actions for selected SSC issues to assess the effectiveness of PPL's maintenance activities. The inspectors reviewed the performance history of those SSCs and assessed PPLs extent-of-condition determinations for these issues with potential common cause or generic implications to evaluate the adequacy of PPLs corrective actions. The inspectors reviewed PPL's problem identification and resolution actions for these issues to evaluate whether PPL had appropriately monitored, evaluated, and dispositioned the issues in accordance with PPL procedures and the requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance." In addition, the inspectors reviewed selected SSC classification, performance criteria and goals, and PPL's corrective actions that were taken or planned, to determine whether the actions were reasonable and appropriate. In addition, the inspectors performed field walkdowns and interviewed PPL staff to verify whether the identified actions were appropriate to correct the extent of condition for identified performance issues. The following issues were reviewed:
- Unit 1, reactor recirculation pump (RRP) seal, Offsite Dose Calculation Manual (ODCM); and
b. Findings
No findings of significance were identified.
==1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 Samples)
a.
==
Inspection Scope
The inspectors reviewed the assessment and management of selected maintenance activities to evaluate the effectiveness of PPL's risk management for planned and emergent work. The inspectors compared the risk assessments and risk management actions to the requirements of 10 CFR Part 50.65(a)(4) and the recommendations of NUMARC 93-01, Section 11, "Assessment of Risk Resulting from Performance of Maintenance Activities." The inspectors evaluated the selected activities to determine whether risk assessments were performed when required and appropriate risk management actions were identified.
The inspectors reviewed scheduled and emergent work activities with licensed operators and work-coordination personnel to evaluate whether risk management action threshold levels were correctly identified. In addition, the inspectors compared the assessed risk configuration to the actual plant conditions and any in-progress evolutions or external events to evaluate whether the assessments were accurate, complete, and appropriate for the emergent work activities. The inspectors performed control room and field walkdowns to evaluate whether the compensatory measures identified by the risk assessments were appropriately performed. The selected maintenance activities included:
- Unit 1, C EDG and Unit 1 RCIC inoperable for maintenance, August 19 2008;
- Unit 2, 2B RHR and ESW Division II inoperable for maintenance, July 29, 2008;
- Unit 2, HPCI inoperable due to missing insulation;
- D and C EDG inoperable, planned and emergent work the week of July 1, 2008;
- E EDG failure to load following scheduled system outage; and
- Temporary loss of secondary containment vacuum, on September 17, 2008.
b. Findings
No findings of significance were identified.
==1R15 Operability Evaluations (71111.15 - 5 Samples)
a. Inspection Scope
==
The inspectors reviewed operability determinations that were selected based on risk insights, to assess the adequacy of the evaluations, the use and control of compensatory measures, and compliance with TSs. In addition, the inspectors reviewed the selected operability determinations to evaluate whether the determinations were performed in accordance with NDAP-QA-0703, "Operability Assessments." The inspectors used the TSs, Technical Requirements Manual, Final Safety Analysis Report (FSAR), and associated Design Basis Documents as references during these reviews.
Documents reviewed are listed in the Attachment. The issues reviewed included:
- Unit 1, ESW leak 1C RHR motor cooler and B ESW loss of offsite power (LOOP);
- Unit 1, outboard scram discharge volume vent valve stroke time greater than inservice testing (IST) criteria;
- Unit 2, HPCI gland seal leak and missing turbine insulation;
- C EDG operability evaluation following operation at full fuel rack conditions.
b. Findings
Introduction:
The inspections identified a Green NCV of Susquehanna Unit 1 TS 3.5.1 and 3.5.3 for rendering HPCI and RCIC inoperable during a planned shutdown.
Specifically, both the HPCI and RCIC systems were made inoperable to fulfill their TS described safety function when operators raised reactor vessel level above the HPCI and RCIC turbine trip signals in a plant mode and at a plant pressure where both of these systems were required to be fully operable. The increased plant risk for this evolution was not modeled or evaluated by plant personnel.
Description:
On March 4, 2008, operations personnel were in the process of shutting down the reactor plant and placing Unit 1 in Mode 4 in accordance with plant procedure GO-100-105, Plant Shutdown to Hot/Cold Shutdown, Revision 41. The operators reached Step 5.26 which states, BEFORE approximately 150 psig, Raise level to 90 inches - 100 inches. When reactor pressure vessel (RPV) level reaches approximately Level 8 (58), a turbine trip signal for the main turbine, the reactor feed pump (RFP)turbines, and the HPCI and RCIC turbines is generated. During the March 4 shutdown, operators reached the Level 8 setpoint with reactor steam dome pressure at 258 psig.
The HPCI and RCIC systems are both required by TSs to be operable in Modes 1, 2, and 3, when steam dome pressure is greater than 150 psig. The HPCI turbine will automatically reset when RPV level reaches -38, and the RCIC turbine will reset once level drops below Level 8. Both can be manually reset by the operators. However, when RPV level is greater than Level 8, both the HPCI and RCIC systems can not be started manually or automatically. HPCI is designed to automatically provide rated flow to the core within 30 seconds of receipt of a RPV Level 2 signal or a drywell high pressure signal. RCIC is not considered an ECCS system. RCIC is designed to ensure that on a main steam isolation valve (MSIV) closure and loss of feedwater, no low pressure (LP) ECCS actuation is required. HPCI is a backup to RCIC for this design function.
On May 22, 2008, an operator questioned whether HPCI had been operable during the March 4, 2008 shutdown. AR 103449103449was written to capture this concern. PPL evaluated this concern and determined that the HPCI system was still operable. This was based upon the position that the increased RPV ensured adequate core cooling and that HPCI is designed to cycle on and off between Level 2 and Level 8.
The inspectors reviewed PPLs evaluation and concluded that the HPCI evaluation did not adequately assess all of the potential operability concerns created by the plant condition. Specifically, PPL did not consider the impact of plant conditions on the RCIC system, or the fact that HPCI was unable to provide backup to the RCIC functions as designed. Furthermore, on a loss of vacuum or Station Blackout transient initiators, with RPV level greater than Level 8 once MSIVs automatically close neither the RCIC or HPCI systems would be available to start and provide core cooling for these transients.
Due to high decay heat load following the March 4 reactor shutdown, plant temperature and pressure would increase until the low pressure coolant injection system (LPCI)shutoff pressure of 433 psig was reached. Plant pressure and temperature would have to be controlled by feed and bleed using safety relief valves (SRVs). Limiting Condition for Operation (LCO) 3.5.3.A for RCIC System Inoperable requires the operators to verify by administrative means that HCPI is operable immediately. Since HPCI also can not perform this function, HPCI would also have to be considered inoperable. Although not formally recognized or evaluated, all TS actions were met since the plant shutdown was already in progress.
Raising RPV level above Level 8, thus making two technical specification systems inoperable and unavailable during plant conditions where they are required to be operable is a performance deficiency which was reasonably within PPLs ability to foresee and correct.
Analysis:
The finding was not subject to traditional enforcement because there were no actual consequences, it was not willful, and did not impact the NRCs ability to regulate.
The finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was evaluated for significance using IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Since the finding did not result in a loss of safety function or the loss of a train for greater than its TS allowed outage time, and was not potentially risk significant due to external event initiators, the finding was determined to be of very low safety significance (Green).
This finding is related to the cross-cutting area of Human Performance - Resources because PPL did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, operating procedure GO-100-105, Plant Shutdown to Hot/Cold Shutdown, Revision 41 did not contain sufficient detail regarding the reactor steam dome pressure at which raising RPV water level was permissible. This was determined to be the most significant contributing factor to this issue. H.2(c)
Enforcement:
Susquehanna Unit 1 Technical Specifications 3.5.1 and 3.5.3 require that HCPI and RCIC be operable while in Modes 1, 2, and 3 and when reactor steam dome pressure is greater than or equal to 150 psig. Contrary to the above, on March 4, 2008, both HPCI and RCIC were rendered inoperable to perform the required safety function when operators raised RPV level above 58 in Mode 3 with plant pressure at 258 psig.
Because this finding is of very low safety significance (Green) and has been entered into PPLs corrective action program (CR 1045441), this violation is being treated as an NCV consistent with section VI.A of the NRC Enforcement Policy. (NCV
===05000387/2008004-02, HPCI and RCIC Made Inoperable Due to Operator Actions during Shutdown)
==1R18 Plant Modifications (71111.18 - 1 Sample)
Permanent Modification
a. Inspection Scope
==
The inspectors reviewed the following engineering changes (modifications): EC 739001; EC 738965; and EC 739040. This modification review concentrated on the RHR motor cooler ESW pipe replacement which was implemented to make the C and D RHR pumps available for a 10 CFR Part 50, Appendix R safe shutdown event. The inspection verified that this modification was in accordance with licensing and design bases, license commitments, and the Updated Final Safety Analysis Report (UFSAR) and the performance capability of associated structures, systems, or components was not degraded through plant modifications.
The inspectors assessed the adequacy of the 10 CFR 50.59 screenings and permanent plant modifications through interviews with PPL personnel and review of supporting information, such as calculations, engineering analyses, design change documentation, modification packages, drawings, plant procedures, the UFSAR, and technical specifications. The inspectors also performed field walkdowns of the installed changes.
The plant modification was selected based on the safety significance of the affected SSCs. The inspectors also reviewed issues that were entered into the corrective action program to determine whether PPL was effective in identifying and resolving problems associated with modification implementation. The permanent modification sample included:
b. Findings
No findings of significance were identified.
==1R19 Post-Maintenance Testing (71111.19 - 6 Samples)
a. Inspection Scope
==
The inspectors observed portions of PMT activities in the field to determine whether the tests were performed in accordance with the approved procedures. The inspectors assessed the test adequacy by comparing the test methodology to the scope of maintenance work performed. In addition, the inspectors evaluated acceptance criteria to determine whether the test demonstrated that components satisfied the applicable design and licensing bases and TS requirements. The inspectors reviewed the recorded test data to determine whether the acceptance criteria was satisfied.
Documents reviewed are listed in the Attachment. The PMT activities reviewed included:
- Unit 2, 2B RHR testing following valve and motor cooler piping replacement;
- Diesel-driven fire pump testing after overhaul and maintenance, SO-13-001;
- Reactor building ventilation Zone 1 equipment exhaust fan flow controller replacement.
b. Findings
No findings of significance were identified.
==1R22 Surveillance Testing (71111.22 - 4 Samples)
a. Inspection Scope
==
The inspectors observed portions of selected surveillance test activities in the control room and in the field and reviewed test data results. The inspectors compared the test results to the established acceptance criteria and the applicable TS or Technical Requirements Manual operability and surveillance requirements to evaluate whether the systems were capable of performing their intended safety functions. The observed or reviewed surveillance tests included:
- Unit 1, 92 day scram discharge volume vent and drain valve, SO-155-02;
- Units 1and 2, degraded grid voltage surveillance, SO-104-001;
- Units 1 and 2, quarterly RCIC flow verification, SO-150-002 and SO-250-002; and
- Unit 1 and 2 reactor coolant system (RCS) leakage.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06 - 1 Sample)
a. Inspection Scope
The inspectors reviewed the combined functional drill scenario (2008 Gold Crew Emergency Drill) that was conducted on August 19, 2008, and observed selected portions of the drill in the simulator control room and technical support center. The inspection focused on PPLs ability to properly conduct emergency action level classification, notification, and protective action recommendation activities and on the evaluators ability to identify observed weaknesses and/or deficiencies within these areas. Ten performance indicator (PI) opportunities were included in the scenario.
The inspectors attended the evaluators post-drill critique and compared identified weaknesses and deficiencies including missed PI opportunities against those identified by PPL to determine whether PPL was properly identifying weaknesses and failures in these areas. The drill evaluation sample included:
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (PS)
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 6 Samples)
a. Inspection Scope
The inspectors reviewed all licensee PIs for the occupational exposure cornerstone for followup.
The inspectors reviewed PPLs self-assessments, audits, Licensee Event Reports, and Special Reports related to the access control program since the last inspection. The inspectors determined if identified problems were entered into the corrective action program for resolution.
The inspectors reviewed corrective action reports (ARs) related to access controls. The inspectors interviewed staff and reviewed documents to determine if the followup activities are being conducted in an effective and timely manner commensurate with their importance to safety and risk:
- Initial problem identification, characterization, and tracking;
- Disposition of operability/reportability issues;
- Evaluation of safety significance/risk and priority for resolution;
- Identification of repetitive problems;
- Identification of contributing causes;
- Identification and implementation of effective corrective actions;
- Resolution of NCVs tracked in the corrective action system; and
- Implementation/consideration of risk significant operational experience feedback.
For repetitive deficiencies or significant individual deficiencies in problem identification and resolution identified above, the inspectors determined if PPLs self-assessment activities were also identifying and addressing these deficiencies. The inspectors discussed with the radiation protection manager (RPM) high dose rate-high radiation area, and very high radiation area controls and procedures. The inspectors focused on any procedural changes since the last inspection. The inspectors verified that any changes to PPLs procedures did not substantially reduce the effectiveness and level of worker protection.
The inspectors discussed with health physics supervisors the controls in place for special areas that have the potential to become very high radiation areas during certain plant operations. The inspectors determined if these plant operations required communication beforehand with the health physics group, so as to allow corresponding timely actions to properly post and control the radiation hazards. The inspectors evaluated licensee performance against the requirements contained in 10 CFR 20 and Plant Technical Specification 5.7.
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02 - 3 Samples)
a. Inspection Scope
Utilizing PPL records, the inspectors determined the historical trends and current status of tracked plant source terms. The inspectors determined if PPL was making allowances or developing contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry.
The inspectors observed radiation worker and radiation protection (RP) technician performance during work activities being performed in radiation areas, airborne radioactivity areas, or high radiation areas. The inspectors concentrated on work activities that present the greatest radiological risk to workers. The inspectors determined if workers demonstrated the as low as is reasonably achievable (ALARA)philosophy and whether there were any procedure compliance issues. The inspectors observed radiation worker performance to determine whether the training/skill level was sufficient with respect to the radiological hazards and the work involved.
The inspectors determined if there have been any declared pregnant workers during the current assessment period. The inspectors reviewed the exposure results and monitoring controls employed by PPL with respect to the requirements of 10 CFR 20.
The inspector evaluated PPL performance against the requirements contained in 10 CFR 20.1101.
b. Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation (71121.03 - 1 Sample)
a. Inspection Scope
The inspectors verified the calibration expiration and source response checks currently on radiation detection instruments staged for use. The inspectors observed RP technicians for appropriate instrument selection and self-verification of instrument operability prior to use.
The inspector evaluated PPL performance against the requirements contained in 10 CFR 20.1501, 10 CFR 20.1703, and 10 CFR 20.1704.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
(71151 - 13 Samples)
a. Inspection Scope
The inspectors reviewed PPLs PI data for the period of July 2006 through July 2007 to determine whether the PI data was accurate and complete. The inspectors examined selected samples of PI data, PI data summary reports, and other plant records. The inspectors compared the PI data against the guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline. The documents reviewed are listed in the Attachment. The following PIs were included in this review.
Initiating Events Cornerstone===
- Unit 1 and Unit 2 Unplanned Scrams per 7000 Critical Hours (IE01);
- Unit 1 and Unit 2 Complicated Plant Scrams (IE02); and
- Unit 1 and Unit 2 Unplanned Power Changes per 7000 Critical hours (IE03)
Mitigating Systems Cornerstone (2 samples)
- Unit 1 and Unit 2 Safety System Functional Failures (MS05)
Barrier Integrity Cornerstone (4 samples)
- Unit 1 and Unit 2 RCS activity (BI01); and
- Unit 1 and Unit 2 RCS identified leak rate (BI02).
b. Findings
No findings of significance were identified.
Occupational Radiation Safety Cornerstone a.
Inspection Scope (1 sample)
The inspectors reviewed all licensee PIs for the Occupational Exposure Cornerstone for followup. The inspectors reviewed a listing of licensee ARs for the period January 1, 2008 through August 28, 2008, for issues related to the occupational radiation safety PI, which measures non-conformances with high radiation areas greater than 1R/hr and unplanned personnel exposures greater than 100 mrem total effective dose equivalent (TEDE), 5 rem skin dose equivalent (SDE), 1.5 rem lens dose equivalent (LDE), or 100 mrem to the unborn child.
The inspectors determined if any of these PI events involved dose rates >25 R/hr at 30 centimeters or >500 R/hr at 1 meter. If so, the inspectors determined what barriers had failed and if there were any barriers left to prevent personnel access. For unintended exposures >100 mrem TEDE (or >5 rem SDE or >1.5 rem LDE), the inspectors determined if there were any overexposures or substantial potential for overexposure. The inspectors determined that no PI events had occurred during the assessment period.
b. Findings
No significant findings or observations were identified.
4OA2 Identification and Resolution of Problems
(71152 - 1 Sample)
.1 Annual Sample:
Review of Panel Annunciator Power Supply Failures and High Equalizing Voltage on Battery Chargers
a. Inspection Scope
The inspectors selected CR 838244, 838807, 912601, and 1028113 as a problem identification and resolution (PI&R) sample for a detailed follow-up review.
CR 838244 was initiated to address high equalizing voltage output from station battery charger 1D613. CR 838807 and CR 912601 were initiated to assess a trend of premature failures of panel annunciator power supplies. CR 1028113 was initiated because of the failure of the annunciator power supply for the 1C103 panel.
The inspectors assessed PPLs problem identification threshold, cause analyses, extent-of-condition reviews, operability determinations, and the prioritization and timeliness of corrective actions to determine whether PPL was appropriately identifying, characterizing, and correcting problems associated with these issues and whether the planned or completed corrective actions were appropriate to prevent recurrence.
Additionally, the inspectors interviewed cognizant plant personnel regarding the identified issues. Specific documents reviewed are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
The inspectors determined that PPL properly implemented their corrective action process regarding the initial discovery of the above issues. The CR packages were complete and included cause evaluations, operability determinations, extent-of-condition reviews, and corrective actions. Additionally, the elements of the CR packages were detailed. Corrective actions appeared appropriate to prevent recurrence of the above issues. Corrective actions addressed immediate equipment concerns as well as improvements to procurement requirements and specifications for future component purchases.
The inspectors determined that corrective actions included reducing the 125 volts direct current (VDC) battery chargers equalizing voltage setpoint to between 140 and 141 VDC, performing independent/vendor failure analyses of failed panel annunciator power supplies, and replacement of panel annunciator power supplies with new input voltage and environmental requirements from a new vendor. The first panel annunciator power supply from a new vendor was installed for the 1C103 panel in May 2008. The licensee planned to replace the highest failure rate panel annunciator power supplies (1C101, 1C102, 1C121, 2C101, 2C102, 2C103, 2C121) with new power supplies as they become available. The inspectors determined that adequate tracking mechanisms were in place to ensure all corrective actions will be completed. Also, the licensee planned to track the performance of the new panel annunciator power supplies to ensure their effectiveness.
.2 Routine Review of Items Entered into the CAP
a. Inspection Scope
As required by IP 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures and human performance issues for follow-up, the inspectors performed routine screening of issues entered into PPLs CAP. The review was accomplished by selectively reviewing copies of CRs and accessing PPLs computerized database.
b. Findings
No findings of significance were identified.
4OA3 Event Followup
(71153 - 3 Samples)
.1 Unit 1 Reactor Power Reduction and Turbine Generator Taken Offline to Repair an
Electro-Hydraulic Control Leak
On the afternoon on August 15, 2008, plant operators identified a small control oil leak on one of the Unit 1 turbine control valves. The leak was noticed by trends in operator log readings over a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. PPL personnel made a brief entry to the area around the turbine to observe the leak which was quantified at approximately 0.25 gpm. The hydraulic fluid leak was identified to be at the location of the D main turbine control valve. Later that day, PPL operators reduced reactor power and took the main generator off-line to facilitate repairs. The resident inspectors followed the work preparation, the reduction in reactor power (reactivity manipulation), the leak repairs, and the return to full power operation on August 18, 2008. Inspectors reviewed corrective maintenance planning, initiated corrective action reports, and the operating procedures utilized for the emergent work and the manipulation of the reactor plant.
b. Findings
No findings of significance were identified.
.2 Loss of Station Startup Bus 10
On September 9, 2008, with Susquehanna Unit 1 operating at 94 percent power and Unit 2 operating at 100 percent power, the station lost power to Startup Bus (SUB) 10, one of two TS required offsite power sources. This was caused by a lightning strike resulting in a momentary loss of the Montour electrical power line to the site. The associated T-10 transformer had been radially fed from the Montour line because the Mountain line was previously out-of-service for maintenance. Operations entered TS 3.8.1 for loss on one offsite circuit which has a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time for the limiting condition of operation (LCO).
SUB 10 normally provides power to four of eight 4 kV safety busses between Units 1 and 2. These busses automatically swapped to their alternate feeds from the SUB 20 powered from the T-20 transformer. The emergency diesel generators were not called upon to start or power the safety busses. A number of non-safety-related common loads (e.g., service water and river intake pumps) were lost initially as a result of the event since tie breaker 0A10502 between SUB 10 and SUB 20 failed to close. Both units remained at full power throughout the event.
Operations reenergized SUB 10 from the T-10 transformer, energized the engineered safeguards system (ESS) transformers, transferred the 4 kV safety busses to their normal power supplies, and cleared the LCO in less than two hours. The Mountain line was restored to service the same day.
During the event, the inspectors responded to the control room to observe operator performance. The inspectors verified locally and remotely that the safety busses were being powered from their alternate power source during the event and from their primary source after the event. The inspectors compared their observations against alarm response procedures and off-normal procedures. Additionally, the inspectors reviewed the initiated corrective actions and troubleshooting activities relating to the failure of tie breaker 0A10502. The documents reviewed are listed in the Attachment.
b. Findings
No findings of significance were identified.
.3 (Closed) Licensee Event Report (LER) 05000387/2008-001-00 & 01, Irradiated Fuel
Movement Without All Required Radiation Monitoring Operable
On February 5, 2008, PPL operators made irradiated fuel moves (six irradiated bundles)with both the Susquehanna Unit 1 and Unit 2 refuel floor exhaust high radiation monitors bypassed. This loss of safety function and violation of TSs (Table 3.3.6.2-1 & 3.3.7.1-1)were reported in the subject LER and through an update to this report (LER Revision 01) issued on August 29, 2008. This performance issue was previously documented as a licensee-identified violation of TSs in Section 4OA7 of NRC inspection report 05000387; 05000388/2008002.
Inspectors reviewed this LER, the update to this LER, and PPL condition report 967350 including PPL's follow-up actions. No additional findings of significance were identified.
Inspectors found that the corrective actions addressed all the causal factors and contributors to this loss of safety function. This LER is closed.
b. Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
During the inspection period the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
On October 16, 2008, the resident inspectors presented their findings to Mr. Ronald Smith, and other members of your staff, who acknowledged the findings.
On August 29, 2008, the inspector presented inspection results to Mr. N. Gannon and other members of his staff. PPL acknowledged the findings. The inspectors confirmed that proprietary information was not included in the inspection report.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by PPL and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
$
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that deviations from such standards are controlled. Contrary to this requirement, on September 18, 2008, PPL identified that a portion of insulation was not properly installed on the Unit 2 HPCI turbine casing which challenged the environmental qualification of the nearby, safety-related electronic governor and ramp generator signal convertor electrical components. This issue was entered into PPLs CAP as CR 1075455. This violation is very low safety significance (Green) because it was a qualification deficiency (not previously analyzed) that was confirmed not to result in loss of operability.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- D. Borger, Shift Manager - Operations
- D. Brophy, Regulatory Affairs
- L. Casella, System Engineer
- R. Fry, Supervisor - Operator Training
- N. Gannon, Vice President - Operations
- D. Roth, Supervisor, Programs and Testing Group
- C. Young, Unit Supervisor - Operations
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Opened/Closed
- 05000387; 388/2008004-01 NCV Inadequate Equipment to Assess Threshold for Emergency Action Level (Section 1R01)
- 05000387/2008004-02 NCV HPCI and RCIC Made Inoperable due to Operator Actions during Shutdown (Section 1R15)
Closed
- 05000387/2008-001-00 & 01 LER Irradiated Fuel Movement Without All Required Monitoring Operable (Section 4OA3.3)
BASELINE INSPECTION PROCEDURE PERFORMED