IR 05000482/2009302
| ML100420046 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/11/2010 |
| From: | Mark Haire Operations Branch IV |
| To: | Matthew Sunseri Wolf Creek |
| References | |
| 50-482/09-302 50-482/09-302 | |
| Download: ML100420046 (12) | |
Text
UNITED STATES NUC LE AR RE G UL AT O RY C O M M I S S I O N ary 11, 2010
SUBJECT:
WOLF CREEK GENERATING STATION - NRC EXAMINATION REPORT 05000482/2009302
Dear Mr. Sunseri:
On December 15, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license retake examination at Wolf Creek Generating Station. The enclosed report documents the examination findings and licensing decisions. The preliminary examination findings were discussed on December 15, 2009, with Mr. Rick Murray, Exam Supervisor, and other members of your staff. In a telephone conversation on December 17, 2009, NRC licensing decisions were provided to Ms. Jane Hartley. A telephonic exit meeting was conducted on January 6, 2010 with Terry Damashek, Superintendent, Operations Support, and other members of your staff.
The examination included the evaluation of one applicant for instant senior reactor operator license and was a retake of the simulator scenarios. The license examiners determined that the applicant satisfied the requirements of 10 CFR Part 55, and the appropriate license has been issued. The enclosure to this letter contains the details of this report.
This report documents one NRC-identified finding of very low safety significance (Green).
However, because of the very low safety significance and because it is entered into your corrective action program, the NRC is treating the finding as a noncited violation (NCV),
consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest this NCV, you should provide a response within 30 days of the date of this examination report, with the basis for your denial, to the US Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 East Lamar Blvd., Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, US Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Wolf Creek Generating Station. In addition, if you disagree with the characterization of the finding in this report, you should provide a response within 30 days of the date of this examination report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at the
Wolf Creek Nuclear Operating Corporation -2-Wolf Creek Generating Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of 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/
Mark S. Haire, Chief Operations Branch Division of Reactor Safety Docket: 50-482 License: NPF-42
Enclosure:
NRC Examination Report 05000482/2009302
REGION IV==
Docket: 50-482 License: NPF-42 Report: 05000482/2009302 Licensee: Wolf Creek Nuclear Operating Corporation Facility Wolf Creek Generating Station Location: 1550 Oxen Lane NE Burlington, Kansas Dates: December 14, 2009 - January 6, 2010 Inspectors: G. Apger, Operations Engineer S. Hedger, Operations Engineer Approved By: Mark S. Haire, Chief Operations Branch Division of Reactor Safety 1 Enclosure
SUMMARY OF FINDINGS
ER05000482/2009302; December 14, 2009 - January 6, 2010; Wolf Creek Generating Station;
Initial Operator Licensing Examination Report; Identification and Resolution of Problems.
NRC examiners evaluated the competency of one applicant for instant senior reactor operator license at Wolf Creek Generating Station.
The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 9, Supplement 1. NRC examiners administered the operating test December 15, 2009.
The examiners determined that the applicant satisfied the requirements of 10 CFR Part 55, and the appropriate license has been issued.
NRC-Identified and Self-Revealing Findings
Conerstone: Barrier Integrity
- Green.
The NRC examiners identified a Green noncited violation of Technical Specification 5.4.1.b for failure to validate changes made to Emergency Operating Procedures. Specifically, the licensee failed to validate a change made to Emergency Operating Procedure E-0, Reactor Trip or Safety Injection. This unvalidated change to E-0 had the unintended consequence of changing the Emergency Operating Procedure mitigation strategy in the steam generator tube rupture procedure, E-3, in that it resulted in premature direction to close the main steam isolation valves which increases the likelihood and duration of a radioactive release during a tube rupture event. This was an undesirable effect that the licensee had not considered when it made the change to E-0. This was entered into the licensees Corrective Action Program under AR22391, and the licensee removed the change that was made to E-0.
The finding was more than minor because it adversely affected the barrier integrity cornerstone attribute of Procedure Quality in that the change to the emergency operating procedure increased the likelihood of an offsite release during a steam generator tube rupture casualty. Manual Chapter 0609, Attachment 4, Initial Screening and Characterization of Findings, was used to evaluate the finding. The finding is of very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool; it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; it did not represent an actual open pathway in the physical integrity of reactor containment; and it did not involve an actual reduction in function of hydrogen ignitors in the reactor containment. The finding had a crosscutting aspect in the area of human performance associated with decision making because the licensee failed to conduct effectiveness reviews of safety-significant decisions to verify the validity of underlying assumptions and identify possible unintended consequences H.1(b). (Section 40A2)
Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
Cornerstone: Barrier Integrity
a. Inspection Scope
As part of the retake examination the inspectors observed performance of emergency operating procedures and reviewed changes that were made to them.
b. Findings
Introduction:
The NRC examiners identified a Green noncited violation of Technical Specification 5.4.1.b for failure to validate changes made to Emergency Operating Procedures (EOP). Specifically, the licensee failed to validate a change made to EOP E-0, Reactor Trip or Safety Injection. This unvalidated change to E-0 had the unintended consequence of changing the EOP mitigation strategy in the steam generator tube rupture procedure in that it resulted in premature direction to close the main steam isolation valves which increases the likelihood and duration of a radioactive release during a tube rupture event. This was an undesirable effect that the licensee had not considered when it made the change to E-0.
Description:
Following the August 19, 2009 Loss of Offsite Power event at Wolf Creek, the licensee initiated Condition Report (CR) 19315. One of the results of this CR was an insertion of main steam isolation valve (MSIV) closure criteria into E-0, Reactor Trip or Safety Injection, at continuous action step 9 Check RCS Cold Leg Temperature -
Stable at or trending to 557F for condenser steam dumps. Under the Response Not Obtained column for this step, item a.4 originally stated, If cooldown continues due to excessive steam flow then close MSIVs. This is consistent with Westinghouse Owners Group guidance and mitigates excessive cooldown due to a possible steam leak.
However, as a result of the CR, item a.4 was changed to state, If any RCS cold leg temperature reaches 555F, then isolate main steam lines. While this would provide the same mitigation for a steam leak event, it has unintended consequences when no steam leak exists. Specifically, during a steam generator tube rupture (SGTR) event, the combination of safety injection flow and auxiliary feedwater flow would likely cause Reactor Coolant System (RCS) cold leg temperature to go below 555F early in the event. Therefore, the new direction in a.4 of E-0 would lead operators to close MSIVs for most SGTR events during the execution of E-0 and before the transition to the SGTR procedure, E-3, which would result in undesirable early and sustained lifting of atmospheric relief valves on the steam generators for most SGTR events (with the MSIVs closed and RCS pressure leaking into the ruptured steam generator), as illustrated by the two simulator scenarios detailed below.
On December 14, 2009, the inspector observed licensed operators in the simulator step through a 500 gallon per minute SGTR event as part of an exam validation. The inspector noted that the crew shut the MSIVs at 555F, as directed by the revision to a.4 of E-0. Note that this shutting of the MSIVs under these conditions would not have been directed by the previous revision of E-0, and is not desirable at this stage of a SGTR event. Primary coolant entering the affected steam generator caused steam generator level to increase which caused steam generator pressure to increase.
Approximately five minutes later, the Atmospheric Relief Valve (ARV) for the affected steam generator began to modulate open to control steam generator pressure. The ARV remained open for approximately twenty minutes until the crew had completed the initial cooldown of the RCS in accordance with Emergency Operating Procedure E-3, Steam Generator Tube Rupture. Opening the affected ARV provided a direct path from the RCS to the atmosphere.
On December 15, 2009, the inspector observed one applicant and two operators respond to the same scenario as part of the retake operating test. The same sequence of events occurred, and the ARV was open for approximately twelve minutes. The difference in time was due to early isolation of feedwater into the affected steam generator and faster transition into E-3 by the operators.
Contrary to the change in E-0 implemented by the CR, both the Westinghouse Emergency Procedure Guidelines (EPG) and the licensees own implementation of E-3, Steam Generator Tube Rupture, indicate that the affected steam generator MSIVs should not be closed until immediately prior to performing steps that isolate feedwater flow into the ruptured generator and then cool the RCS. By keeping the affected steam generators MSIVs open until required to be closed for the RCS cooldown, any radioactive steam in the ruptured generator would be directed to the condenser while the MSIV is open as long as offsite power is available. This desirable delay in closing the MSIVs increases the likelihood that the affected steam generators ARV would remain shut during the time from isolation to reaching the target temperature, which, in turn, reduces the likelihood of a radioactive release through the ruptured steam generators ARV.
The change that was made to E-0 effectively changed the entry condition for E-3.
Specifically, because of the change to step a.4 of E-0, E-3 would be entered with MSIVs closed during all major tube rupture events. The inspectors reviewed the bounding analysis for the STGR casualty in the Final Safety Analysis Report, and this change would not cause the SGTR mitigation strategy to place the plant outside the bounds which assumes offsite power is lost. However, the change implemented by the CR does have the unintended consequence of increasing the likelihood of an extended release of radioactive coolant to the atmosphere because early closure of the MSIVs results in extended lifting of the affected steam generators ARV, as demonstrated by the two simulator scenarios detailed above.
Analysis:
The finding was considered more than minor because it adversely affected the barrier integrity cornerstone attribute of Procedure Quality in that the change to the EOP increased the likelihood of an offsite release during a SGTR casualty. Manual Chapter 0609, Attachment 4, Initial Screening and Characterization of Findings, was used to evaluate the finding. The finding is of very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool; it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere; it did not represent an actual open pathway in the physical integrity of reactor containment; and, it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a crosscutting aspect in the area of human performance associated with decision making because the licensee failed to conduct effectiveness reviews of safety-significant decisions to verify the validity of underlying assumptions and identify possible unintended consequences H.1(b).
Enforcement:
Technical Specification 5.4.1.b states in part that Emergency Operating Procedures shall be maintained in accordance with NUREG-0737 and NUREG-0737, Supplement 1, as stated in section 7.1 of Generic Letter 82-33. NUREG-0737, Supplement 1, section 7.2.b states in part that the procedures generation package (for Emergency Operating Procedures) shall include a Writers Guide that details the specific methods to be used by the licensee in preparing Emergency Operating Procedures based on Technical Guidelines. Procedures AP 15C-006, Emergency Procedure Generation Package, and AP 15C-004, Preparation, Review and Approval of Procedures, Instructions and Forms, were developed by the licensee in response to NUREG-0737, Supplement 1. AP 15C-004 requires Emergency Operating Procedure changes that could change the intent or mitigation strategy be validated. Contrary to the above, the licensee changed E-0, "Reactor Trip or Safety Injection," without properly analyzing the effect the change would have on Emergency Operating Procedure mitigation strategies; therefore, the licensee subsequently failed to validate the change made to E-0. The licensee entered this into their corrective action program under AR22391 and removed the change to E-0. Because this violation is of very low safety significance and was entered into the corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2009302-01, "Inadequately Analyzed Emergency Operating Procedure Change."
4OA5 Other Activities (Initial Operator License Examination)
.1 License Applications
a. Scope
NRC examiners reviewed the license application that was submitted to ensure the applicant satisfied relevant license eligibility requirements. The examiners also audited the license application in detail to confirm that it accurately reflected the subject applicants qualifications. This audit focused on the applicants experience and on-the-job training, including control manipulations that provided significant reactivity changes.
b. Findings
No findings of significance were identified.
.2 Examination Development
a. Scope
NRC examiners reviewed the examination outline and draft examination submitted by the licensee against the requirements of NUREG-1021. The NRC examination team conducted an onsite validation of the operating test.
b. Findings
NRC examiners provided outline, draft examination and post-validation comments to the licensee. The licensee satisfactorily completed comment resolution prior to examination administration.
NRC examiners determined that the operating test initially submitted by the licensee was within the range of acceptability expected for a proposed examination.
.3 Operator Knowledge and Performance
a. Scope
The NRC examination team administered the operating test which consisted of two simulator scenarios to the applicant on December 15, 2009.
b. Findings
No findings of significance were identified.
The applicant passed the retake examination.
.4 Simulation Facility Performance
a. Scope
The NRC examiners observed simulator performance with regard to plant fidelity during examination validation and administration.
b. Findings
No findings of significance were identified.
.5 Examination Security
a. Scope
The NRC examiners reviewed examination security during both the onsite preparation day and examination administration day for compliance with 10 CFR 55.49 and NUREG-1021. Plans for simulator security and applicant control were reviewed and discussed with licensee personnel.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
The chief examiner presented the preliminary examination results to Mr. Rick Murray, Exam Supervisor, and other members of the staff on December 15, 2009. A telephonic exit was conducted on January 6, 2010, between Messrs. Gabriel Apger, Chief Examiner, Terry Damashek, Superintendent, Operations Support, and other members of the staff.
The licensee did not identify any information or materials used during the examination as proprietary.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- R. Murray, Exam Supervisor
- L. Rockers, Licensing
NRC Personnel
- C. Long, Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000482/2009302-01 NCV Inadequately Analyzed Emergency Operating Procedure Change (Section 40A2)