IR 05000266/2007301
| ML072340427 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/21/2007 |
| From: | Hironori Peterson NRC/RGN-III/DRS/OLB |
| To: | Koehl D Nuclear Management Co |
| References | |
| 50-266/07-301, 50-301/07-301 | |
| Download: ML072340427 (19) | |
Text
ust 21, 2007
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 NRC INITIAL LICENSE EXAMINATION REPORT 05000266/2007301(DRS);
Dear Mr. Koehl:
On July 13, 2007, the NRC completed initial operator licensing examinations at your Point Beach Nuclear Plant. The enclosed report presents the results of the examination which were discussed on July 13 and July 30, 2007, with Mr. McCarthy and Mr. Sizemore, respectively, and with other members of your staff.
The NRC examiners administered initial license examination operating tests during the week of July 9, 2007. Members of the Point Beach Nuclear Plant Training Department administered an initial license written examination on July 13, 2007, to the applicants. Four senior reactor operator and three reactor operator applicants were administered license examinations. The results of the examinations were finalized on August 2, 2007. Seven applicants passed all sections of their examinations resulting in the issuance of four senior reactor operator and two reactor operator licenses. One reactor operator applicant passed all portions of the examination, but had not completed all requirements to receive an operating license. Upon completion of all training requirements, you must certify to the NRC, in writing, using the certification statement in Item 19.b. on the NRC Form 398, that the reactor operator successfully completed all Point Beach Nuclear Plant training program requirements. The applicant will be issued a reactor operator license at that time.
During this examination, one finding of very low safety significance which involved a violation of NRC requirements, was identified. However, because of the very low safety significance and because the issue has been entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Perry Nuclear Power Plant.
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 System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Hironori Peterson, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27
Enclosures:
1. Operator Licensing Examination Report 05000266/2007301(DRS); 05000301/2007301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO/SRO)
DISTRIBUTION See next page U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Perry Nuclear Power Plant.
In a
REGION III==
Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27 Report No: 05000266/2007301(DRS);
Licensee: Nuclear Management Company, LLC Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: Two Rivers, Wisconsin Dates: July 9 through July 13, 2007 Examiners: N. Valos, Chief Examiner C. Zoia, Chief Examiner Certification M. Morris, Examiner Approved by: Hironori Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000266/2007301(DRS); 05000301/2007301(DRS); 07/09/07 - 07/13/07;
Point Beach Nuclear Plant, Units 1 and 2; Initial License Examination Report.
The announced operator licensing initial examination was conducted by regional examiners in accordance with the guidance of NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9.
Examination Summary:
- Seven examinations were administered (four senior reactor operators and three reactor operators).
- Seven applicants passed all sections of their examinations resulting in the issuance of four senior reactor operator and two reactor operator licenses. One reactor operator applicant passed all portions of the examination, but had not completed all requirements to receive an operating license. Upon certification of completion of all training requirements, the applicant will be issued a reactor operator license at that time.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a Non-Cited Violation (NCV) of Technical Specification 5.4, Procedures, for the failure to have adequate procedures to ensure the continued operation of the steam dumps to the condenser to maintain a Reactor Coolant System (RCS) cooldown during a Steam Generator Tube Rupture (SGTR) event. Specifically, the procedures permitted the operators to lock in a Safety Injection (SI) signal and then reset SI more than once, which could cause an automatic closure of the Main Steam Isolation Valves (MSIVs) and a loss of steam dump to the condenser, which could result in a delay in terminating the Primary-To-Secondary Leakage. The licensee has initiated procedure change requests to the SGTR emergency operating procedures as a corrective action for this finding.
This finding was more than minor because it was associated with the attribute of procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the Main Steam (MS) system to respond to initiating events to prevent undesirable consequences. Steam dump to the condenser is the preferred means of cooling the RCS during a SGTR because it minimizes radiological releases, conserves feedwater, and provides the most rapid cooldown capability. The finding is of very low safety significance based on the results of the SDP Phase 1 screening worksheet. The inspectors concluded that this finding was cross-cutting in the area of human performance, resources (H.2(c)), in that the licensee failed to have complete, accurate, and up-to-date procedures for the response to a SGTR event. (Section 4OA5)
Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners conducted an announced operator licensing initial examination during the week of July 9, 2007. The facility licensees training staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the outline and develop the written examination and operating test. The examiners administered the operating test, consisting of job performance measures and dynamic simulator scenarios, during the period of July 9 through July 12, 2007. The facility licensee administered the written examination on July 13, 2007. Four senior reactor operator and three reactor operator applicants were examined. During the on-site validation week of June 18, 2007, the examiners audited one license application for accuracy.
b. Findings
Written Examination The NRC examiners determined that the written examination, as originally submitted by the licensee, was within the range of acceptability expected for a proposed examination.
All changes made to the submitted examination were made in accordance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors.
One post-examination comment was submitted by the applicants and station training department personnel on July 23, 2007. The results of the NRCs review of the comments are documented in Enclosure 3, Post Examination Comments and Resolutions.
Operating Test The NRC examiners determined that the operating test, as originally submitted by the licensee, was within the range of acceptability expected for a proposed examination.
All changes made to the submitted examination were made in accordance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors."
Examination Results Seven applicants passed all sections of their examinations resulting in the issuance of four senior reactor operator and two reactor operator licenses. One reactor operator applicant passed all portions of the examination, but had not completed all requirements to receive an operating license. Upon certification of completion of all training requirements, the applicant will be issued a reactor operator license at that time.
.2 Examination Security
a. Inspection Scope
The NRC examiners briefed the facility contact on the NRCs requirements and guidelines related to examination physical security (e.g., access restrictions and simulator considerations) and integrity in accordance with 10 CFR 55.49, Integrity of Examinations and Tests, and NUREG-1021, Operator Licensing Examination Standard for Power Reactors. The examiners reviewed and observed the licensees implementation and controls of examination security and integrity measures (e.g.,
security agreements) throughout the examination process.
b. Findings
The licensees implementation of examination security requirements during examination preparation and administration were acceptable and met the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors.
However, there were two issues associated with examination security identified during the preparation and administration of the examination.
During the time period of the development of the initial license examination, the licensee notified the NRC of one issue which had the potential to affect the integrity of the operating examination. The issue associated with examination security was identified by the licensee on March 9, 2007, when it was discovered that one Operations person had observed initial license training and provided feedback after signing the NRC Examination Security Form ES-201-3. The individual observed simulator training and provided feedback to the initial license crew and the instructor. About 2 days later, the individual realized that the actions taken were inappropriate and self-identified the issue to the General Supervisor of Operations Training (GSOT). The GSOT then interviewed the individual regarding the specific content of the feedback provided, contacted the Point Beach Exam Project Manager, and began implementing the required corrective actions.
The licensee documented this issue in the corrective action program as Corrective Action Program (CAP) Number 01081428. The NRC examiners were appropriately notified of the issue. The examiners reviewed the licensees investigation and assessed the issue for a possible violation of 10 CFR 55.49, Integrity of Examinations and Tests.
Based on the minimal interface of the individual with the initial license crew and no discussion of examination material, the examiners determined that no actual examination compromise had occurred. The issue was not subject to enforcement action in accordance with NRC enforcement policy.
In addition, there was another issue associated with examination security identified by the examiners during the administration of the operating exam. During the administration of a Job Performance Measure (JPM) to one applicant, it was noted that a yellow arrow used for highlighting affected equipment was pointing to a control switch that was manipulated (i.e., the Containment Spray Eductor Suction Valve) during the simulator JPM. The arrow was placed on the switch by the preceding applicant who had completed the same JPM earlier, but the arrow was not removed as required during the subsequent simulator setup. Failure to remove the arrow was considered potentially significant because highlighting equipment being operated can invalidate a JPM under different circumstances. In this case, it was determined that no actual examination compromise had occurred. The licensee documented the issue in the corrective action program as CAP Number 01101383 after the operating test was completed.
.3 Inadequate Procedure to Ensure Main Steam (MS) System Function During a SGTR
Introduction:
The inspectors identified a finding involving a NCV of Technical Specification (TS) 5.4, Procedures, having very low safety significance (Green) for the failure to have an adequate procedure to ensure the continued operation of the steam dumps to the condenser to maintain a Reactor Coolant System (RCS) cooldown during a Steam Generator Tube Rupture (SGTR) event. Steam dump to the condenser is the preferred means of cooling the RCS during a SGTR because it minimizes radiological releases, conserves feedwater, and provides the most rapid cooldown capability.
Description:
On July 10, 2007, during the administration of an initial license examination, the inspectors identified that procedure EOP-3 U1, Steam Generator Tube Rupture, was inadequate in that the procedure did not ensure the continued operation of the steam dumps to the condenser to maintain an RCS cooldown during a SGTR event. Specifically, the procedure permitted the operators to lock in a Safety Injection (SI) signal and then reset SI more than once, which could cause an automatic closure of the Main Steam Isolation Valves (MSIVs) and a loss of steam dump to the condenser.
During a SGTR event, an RCS cooldown to the target temperature is made using the steam dumps, if possible. To do this, the SI signal must be locked in by manually actuating SI and then tripping the Low RCS Pressure SI bistables. The SI signal is then reset prior to initiating the cooldown. The SI signal is locked in prior to resetting SI to ensure auto SI actuation does not occur once the cooldown is begun. This is necessary because of Point Beach specific differences in the SI reset logic circuitry from the standard Westinghouse plant. At Point Beach, an automatic reactuation of SI does not stay locked out when SI is reset. Therefore, if an SI actuated during the RCS cooldown and depressurization, it may complicate efforts to depressurize the RCS later. Manually tripping SI bistables at step 11 of EOP-3 prior to initiating the RCS cooldown ensures that SI will not automatically reactuate after the SI signal is reset.
At EOP-3 Step 15, while the RCS cooldown to the target temperature has been initiated but is still in progress, there is an additional step to again lock in the SI signal and reset SI. Point Beach has an administrative procedure NP 1.1.4, Use And Adherence Of Procedures And Work Plans, that states that if an action required by a procedure was already completed when the step was reached, the step does not need to be re-performed. However, the placement of Step 15 in EOP-3 to once again lock in the SI signal and then reset SI has the potential for adverse consequences if the Senior Reactor Operator (SRO) performing the procedure does not realize that this step does not need to be re-performed, and re-performs the step.
During an initial license examination, a crew repeated EOP-3 Step 15 after Step 11 was previously performed. This resulted in automatic MSIV closure and a corresponding loss of steam dump to the condenser due to an SI signal being present with Low Average Temperature (Tavg) below 543EF. Closure of the MSIVs required the crew to reestablish the RCS cooldown using the intact SG atmospheric steam dump valve, which resulted in a delay in terminating the Primary-To-Secondary Leakage. In addition, the preferred method of RCS cooldown using the steam dumps was no longer available to minimize radiological releases and to conserve feedwater supply to the condenser.
Analysis:
The inspectors determined that the failure to provide an adequate procedure to ensure the continued operation of the steam dumps to the condenser to maintain an RCS cooldown during a SGTR was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, because it was associated with the attribute of procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the Main Steam (MS)system to respond to initiating events to prevent undesirable consequences.
The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Appendix A, Attachment 1, Determining the Significance of Reactor Inspection Findings for At-Power Situations. Since the inspectors answered No to all the screening questions in the Mitigating System Cornerstone Column, the finding screened as having very low safety significance (Green). The inspectors also determined that this finding was cross-cutting in the area of human performance, resources (H.2(c)), in that the licensee failed to have complete, accurate, and up-to-date procedures for the response to a SGTR event.
Enforcement:
Technical Specification 5.4, Procedures, required, in part, that written procedures be established, implemented, and maintained covering the Emergency Operating Procedures (EOPs) required to implement the requirements of NUREG-0737, Clarification of TMI Action Plan Requirements, and NUREG-0737, Supplement 1.
Item I.C.1 of NUREG-0737, and NUREG-0737, Supplement 1, Section 7, required, in part, the development of EOPs to cover transients and accidents including an event that required the response to a SGTR event.
Contrary to this requirement, on July 10, 2007, it was discovered that procedure EOP-3 U1, Steam Generator Tube Rupture, and procedure EOP-3 U2, Steam Generator Tube Rupture, were inadequate in that the procedures did not ensure the continued operation of the steam dumps to the condenser to maintain an RCS cooldown during a SGTR event. As a corrective action for this finding, the licensee has initiated procedure change requests to EOP-3 U1 (PCR 01101669) and EOP-3 U2 (PCR 01101687) such that the step to lock in the SI signal is only incountered once during the performance of the procedures.
Since this violation was of very low safety significance and because the issue was entered into the licensees corrective action program (CAP01101596), this violation is being treated as an NCV, consistent with Section VI.A.1 of the Enforcement Policy.
(NCV 05000266/2007301-01; NCV 05000301263/2007301-01)
4OA6 Meetings
Exit Meeting The certifying chief examiner presented the examination teams preliminary observations and findings with Mr. J. McCarthy and other members of the licensee management on July 13, 2007. A subsequent exit via teleconference was held on July 30, 2007, with Mr. C. Sizemore following review of the site post-examination comment. The licensee identified one proprietary item during the exit meeting and the item was not kept by the examiners. The licensee acknowledged the observations and the finding presented.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
Enclosure 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- R. Admundson, General Supervisor Operations Training
- C. Butcher, Engineering Director
- A. Capristo, Business Support Manager
- F. Flentje, Regulatory Affairs Supervisor
- R. Harrsch, Operations Manager
- J. Jones, Operations Training Supervisor
- R. Joplin, Senior Operations Training Instructor
- J. McCarthy, Director - Site Operations
- P. Olson, Operations Support Manager
- J. Pierce, Operations Continuing Supervisor
- M. Ray, Regulatory Affairs Manager
- J. Schlief, Assistant Operations Manager
- J. Schweitzer, Projects Manager
- C. Sizemore, Training Manager
- W. Smith, Production Planning Manager
- B. VanderVelde, Maintenance Manager
- G. Young, Nuclear Oversight Manager
- A. Zommers, Operations Training Instructor
NRC
- G. Gibbs, Resident Inspector
- R. Krsek, Senior Resident Inspector
- M. Morris, Examiner
- N. Valos, Chief Examiner
- C. Zoia, Chief Examiner (Certification)
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000266/2007301-01 NCV Failure to Provide Adequate Guidance to Ensure the
05000301/2007301-01 Operability of the MS System during a SGTR
Closed
05000266/2007301-01 NCV Failure to Provide Adequate Guidance to Ensure the
05000301/2007301-01 Operability of the MS System during a SGTR
Discussed
None
Attachment
LIST OF ACRONYMS
ADAMS Agency-Wide Document Access and Management System
CFR Code of Federal Regulations
CAP Corrective Action Program
DRS Division of Reactor Safety
EOP Emergency Operating Proceure
MS Main Steam
NRC Nuclear Regulatory Commission
PARS Publicly Available Records System
PCR Procedure Change Request
RO Reactor Operator
SDP Significance Determination Process
SGTR Steam Generator Tube Rupture
SRO Senior Reactor Operator
Attachment
SIMULATION FACILITY REPORT
Facility Licensee: Point Beach Nuclear Plant, Units 1 and 2
Facility Licensee Docket Nos. 50-266; 50-301
Operating Tests Administered: July 9 through July 13, 2007
The following documents observations made by the NRC examination team during the initial
operator license examination. These observations do not constitute audit or inspection findings
and are not, without further verification and review, indicative of non-compliance with
CFR 55.45(b). These observations do not affect NRC certification or approval of the
simulation facility other than to provide information which may be used in future evaluations.
No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM DESCRIPTION
There was one simulator exam scenario delay of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> on the
morning of July 10, 2007, until the Plant Process Computer System (PPCS)
could be restored in the simulator. CAP 01101209 was written associated with
the issue.
Another simulator exam scenario delay of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> occurred on the
afternoon of July 12, 2007, until a failed power supply could be restored in the
simulator. CAP 01101582 was written associated with the issue.
During a simulator scenario, there was a loss of the PPCS Computer which could
not be reset. CAP 01101500 was written associated with the issue.
During a simulator scenario, Nuclear Instrument System Recorder NR-45 blue
pen did not respond to repositioning of its control switch. Simulator Action
Request (SAR) # 233 was written associated with the issue.
During a simulator scenario, hotwell level began rising unexpectedly and almost
required additional operator actions to address. SAR # 234 was written
associated with the issue.
During a simulator scenario, recorders 1TR-401 and 1FR-175 had paper feed
problems. SAR # 235 was written associated with the issue.
During a simulator scenario, Letdown Orifice Isolation Switch 1CV-200B
malfunctioned such that the valve would not open. SAR # 236 was written
associated with the issue.
Post Examination Comments and Resolutions
Question Number 83
After commencing a discharge of T-10A, A Monitor Tank, RE-218, Waste Disposal System
Liquid Monitor, goes into ALERT due to high radiation.
What automatic actions, if any, will occur, and what actions will the OS direct?
- A. WL-18, Waste Liquid Overboard valve will automatically close. The OS will direct the
PAB AO to secure the discharge lineup; the chemistry department will sample the A MT
and issue a new discharge permit if tank levels allow for release.
- B. WL-18, Waste Liquid Overboard valve will automatically close. The OS will direct the
PAB AO to isolate RE-218 and purge Rad Monitor for 5 minutes. If the alarm clears
during the purge, direct the AO to continue the discharge. Otherwise, direct chemistry
to resample the tank and continue the discharge after RE-218 ALERT setpoint is
adjusted.
C. No automatic actions will occur. The OS will direct the PAB AO to secure the discharge
lineup and place the tank on recirc; the chemistry department will sample the A MT and
issue a new discharge permit if tank levels allow for release.
Answer: A
Facility Proposed Resolution:
The question grading for the exam should be changed to have C as the only correct answer.
The question was written to ask about automatic and follow-up actions when an ALERT alarm
is received on RE-218, Waste Disposal System Liquid Monitor. The question as submitted has
A selected as the correct answer. This is incorrect. Choice C is correct in that WL-18,
Waste Liquid Overboard valve does not shut when an ALERT alarm is received.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to change the Answer Key to have distractor C as the only correct answer.
Since WL-18, Waste Liquid Overboard valve, does not automatically close when an ALERT is
received on RE-218, Waste Disposal System Liquid Monitor, distractor A is not a correct
answer (NOTE: Valve WL-18 automatically closes when a High Alarm is received on RE-218).
Distractor C correctly stated that no automatic actions will occur when an ALERT alarm is
received on RE-218. The other actions stated in distractor C for the Operating Supervisor
(OS) to direct (i.e., for the Plant Auxiliary Building (PAB) Auxiliary Operator (AO) to secure the
discharge lineup and place the A Monitor Tank (MT) on recirculation, and for the chemistry
department to sample the A MT and issue a new discharge permit if tank levels allow for
release) are also valid.
Enclosure 3
Since distractor C correctly stated that no automatic actions will occur and also included the
correct actions for the OS to direct, it was decided to change the Answer Key to have distractor
C as the only correct answer.
Enclosure 3
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML072210589.
4