IR 05000346/2005301
| ML052640561 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/19/2005 |
| From: | Hironori Peterson NRC/RGN-III/DRS/OLB |
| To: | Bezilla M FirstEnergy Nuclear Operating Co |
| References | |
| 50-346/05-301 50-346/05-301 | |
| Download: ML052640561 (18) | |
Text
September 19, 2005
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION NRC INITIAL LICENSE EXAMINATION REPORT 050000346/2005301(DRS)
Dear Mr. Bezilla:
On July 28, 2005, the NRC completed initial operator licensing examinations at your Davis-Besse Nuclear Power Station. The enclosed report documents the results of the examination which were discussed on July 28 and August 8, 2005, with Mr. D. Imlay and Mr. J. House, respectively, and with other members of your staff.
NRC examiners administered the operating test during the weeks of July 18 and 25, 2005.
Members of the Davis-Besse Nuclear Power Station Training Department staff administered the written examination on July 28, 2005. Four Reactor Operator (RO) and eight Senior Reactor Operator (SRO) applicants were administered license examinations. The results of the examinations were finalized on August 17, 2005. All 12 applicants passed all sections of their examinations resulting in the issuance of four operator and five senior operator licenses.
During the NRCs review of initial license applications, three senior operator applicants were granted eligibility deferments so that the applicants could take the examination. These three applicants will need to complete a minimum of six months onsite responsible power plant experience before they will receive a license. The applicants will be granted a license only after you certify in writing to the NRC that the applicants have completed the eligibility requirement that was previously deferred.
Although all 12 applicants performed satisfactorily and passed the NRC initial license examination, the submittal of the written examination material by your training staff was considered outside the acceptable quality range expected by the NRC in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9.
Specifically, the RO written examination material was outside the 20 percent acceptable margin for quality in accordance with NUREG 1021. This determination was based on the observation that 16 out of 75 RO questions (21 percent) and 3 out of 25 SRO questions (12 percent)
required replacement or significant modifications and were identified as unsatisfactory. The minimum requirement to determine an adequate quality range, assessed separately for each RO and SRO examination in accordance with ES-501 of NUREG-1021, was 20 percent or fewer questions identified as unsatisfactory. In addition, during preparation and administration of the NRC license examination, one finding of very low safety significance was identified which involved a violation of NRC requirements.
However, because this violation was of very low safety significance, and because the issue was entered into your corrective action program, the NRC is treating this finding as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of a Non-Cited Violation, 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 Davis-Besse Nuclear Power Station.
In accordance with 10 CFR Part 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 NRC's 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).
We will gladly discuss any questions you have concerning this examination.
Sincerely,
/RA/
Hironori Peterson, Chief Operations Branch Division of Reactor Safety Docket No. 50-346 License No. NPF-3
Enclosures:
1.
Operator Licensing Examination Report 050000346/2005301(DRS)
2.
Simulation Facility Report 3.
Post Examination Comments and Resolutions 4.
Written Examinations and Answer Keys (RO & SRO) In addition, during preparation and administration of the NRC license examination, one finding of very low safety significance was identified which involved a violation of NRC requirements.
However, because this violation was of very low safety significance, and because the issue was entered into your corrective action program, the NRC is treating this finding as a Non-Cited Violation in a
REGION III==
Docket No:
50-346 License No:
NPF-3 Report No:
050000346/2005301(DRS)
Licensee:
FirstEnergy Nuclear Operating Company (FENOC)
Facility:
Davis-Besse Nuclear Power Station Location:
5501 North State Route 2 Oak Harbor, OH 43449-9760 Dates:
July 18 through July 28, 2005 Examiners:
M. Bielby, Chief Examiner B. Palagi, Examiner R. Morris, Examiner Approved by:
H. Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
Enclosure 1
SUMMARY OF FINDINGS
ER 05000346/2005301(DRS); 07/18/2005-28/2005; Davis-Besse Nuclear Station; 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.
A.
Examination Summary
- Twelve examinations were administered (four Reactor Operator and eight Senior Reactor Operator).
- Twelve applicants passed all sections of their examinations, nine of these applicants were issued respective operator or senior operator licenses. Three senior operator applicants were granted eligibility deferments so that the applicants could take the examination. These three applicants will need to complete a minimum of six months onsite responsible power plant experience before they will receive a license.
(Section 4OA5.1)
B.
Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
A Green finding associated with a Non-Cited Violation of Title 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to remove an abandoned equipment load listed in the Emergency Procedure DB-OP-02000 as part of modification MOD 95-0050. As a result, upon implementation of the modification, the licensee failed to identify the component abandoned by the modification was referenced in the plant emergency procedures. On July 20, 2005, the inspectors observed operators perform Job Performance Measure (JPM), 2005 NRC JPM F, in the simulator during the NRC initial license examination. The inspectors noted that the applicants had difficulty completing the required procedural steps because of a delay in reducing the load on the electrical bus.
The inspectors determined that a primary cause of this finding was related to the cross-cutting area of Human Performance because the licensee failed to verify the appropriate emergency procedure revisions were established based on the equipment modification.
Although simulated as part of an NRC operator license examination, the issue was more than minor because the finding was associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, Phase 1 Screening, Secondary Core Decay Heat Removal Degraded. The inspectors also determined that the finding was of very low safety significance because even though the establishment of feedwater flow to the Once Through Steam Generator (OTSG) was delayed, the applicants did complete the task as assigned and would have been able to start the Motor Driven Feedwater Pump (MDFP). The licensee took prompt action to enter the item into their corrective action process. (Section 4OA4)
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA4 Cross-Cutting Aspects of Findings
.1 Failure to Update Emergency Procedure DB-OP-02000
a. Inspection Scope
On July 20, 2005, the inspectors observed Initial Operator License examination applicants perform JPM, 2005 NRC JPM F, in the simulator during the NRC initial license examination. The inspectors noted that the applicants had difficulty completing the required procedural steps because of a delay in reducing the load on the electrical bus.
b. Findings
Introduction:
A Green finding associated with a Non-Cited Violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to remove an abandoned equipment load listed in the Emergency Procedure DB-OP-02000 as part of modification MOD 95-0050.
Description:
The applicants were directed to restore power to D2 bus from D1 in accordance with Attachment 2, Section 3.0 of Emergency Procedure DB-OP-02000, RPS, SFAS, SFRCS Trip, Or SG Tube Rupture, in order to start the Motor Driven Feed Pump (MDFP). The applicants followed the procedure to step 3.2 and were directed by the procedure to REFER TO Section 6 of the attachment for load reduction guidance.
The guidance provided in Step 6.1
- (1) required the applicants to reduce load on the Emergency Diesel Generator (EDG) by de-energizing loads from the list to allow the MDFP to be started without exceeding the 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> rating of 2946 KW. The third load on the list was the Transfer Pump Primary Water 2. The applicants could not find the load on the boards and believed that the load had been abandoned in place. However, the fact that the load was on the list caused them to look through procedures and associated drawings to verify that it had been abandoned. This load had been removed by plant modification MOD 95-0050. The modification package included the removal of this component in the drawings and in other procedures, but did not reference DB-OP-02000. Although the applicants eventually completed the EDG load reduction, it was significantly delayed. The licensee entered this issue into their corrective action program as Condition Report CR 05-03962.
Analysis:
The inspectors determined that failure to remove the reference to the Primary Water Transfer Pump 2 from the Emergency Procedure caused unnecessary delays in the operators ability to restore feedwater flow to the OTSGs during an emergency condition. The failure to remove the reference was a licensee performance deficiency warranting a significance evaluation.
The inspectors concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued
on May 19, 2005, in that, the finding was associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, if left uncorrected, the finding could become a more significant safety concern.
The inspectors evaluated the finding using IMC 0609, Appendix A, Phase 1 Screening, Secondary Core Decay Heat Removal Degraded. This event was simulated as part of an NRC operator license examination and did not actually occur. The inspectors determined that the finding was of very low safety significance because even though the establishment of feedwater flow to the OTSG was delayed, the applicants did complete the task as assigned and would have been able to start the MDFP without any adverse consequences. Therefore, the finding was considered to be of very low safety significance (Green).
Enforcement:
Title 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, requires, in part, that the licensee accomplish activities affecting quality in accordance with instructions and procedures of a type appropriate to the circumstances. Contrary to the above, the licensees emergency procedures had a reference to a component that was incorrect and not appropriate for the circumstances.
Therefore, the inspectors determined this finding was a violation of Criteria V, Instructions, Procedures, and Drawings. Because this violation was of very low safety significance, and documented in the licensees corrective action program as CR 05-03962, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 0500346/2005301-01).
.2 Use of Device to Pin SFAS Switch
a. Inspection Scope
On July 18, 2005, the inspectors observed Initial Operator License applicants perform JPM, 2005 NRC JPM C, in the simulator during the NRC initial license examination.
Although not referenced by procedure DB-OP-06014, Section 5.2, the inspectors observed that applicants used a banana plug device to pin and hold the spring return Safety Features Actuation System (SFAS) REACTOR COOLANT (OPER-TEST) Switch in the TEST position to bypass the 800 psig Reactor Coolant Pressure automatic isolation valve open function.
b. Findings
Introduction:
The inspectors identified an Unresolved Item (URI) for an issue raised regarding whether the use of the banana plug device to pin and hold the spring return SFAS panel switch in the test position was consistent with the design and operation of the SFAS system. The inspectors noted that the device was not referenced by the procedure, and there was no identified accountability for the device.
Description:
During performance of a JPM, the simulator initial conditions were Mode 3 with a slowly rising level in Core Flood Tank (CFT) 2 due to check valve leakage. The applicants were directed to perform DB-OP-06014, Core Flooding System Procedure,
Section 5.2, Emergency Closure of CFT 2 Isolation Valve CF1A, to stop the increasing level in CFT 2. The applicants followed the procedure to step 5.2.3.b. which provided direction to Turn and hold the REACTOR COOLANT (OPER-TEST) Switch to TEST position. Prior to performing the procedural step the applicants noted that they would use a device to pin the spring return switch in the TEST position. The device was staged in the simulator key locker for the examination; however, some applicants identified there was no specified location for the device in the plant control room.
Applicants readily identified the device which was an orange electrical lead with a banana plug on the end. Although no procedural direction was provided, applicants inserted the banana plug in a SFAS panel hole located next to the spring return switch to hold it in the TEST position. The inspectors noted that all four SFAS panels had holes drilled next to the respective spring return switches. After performing Steps 5.2.4 through 5.2.6 to close isolation valve CF1A, applicants removed the banana plug and restored the SFAS channel to normal operation. The licensee identified that this same device is also used to pin the switch during the monthly functional test. The licensee was not able to immediately identify any basis documentation for the holes in the SFAS panel or use of the device to hold the spring return switch in place.
On July 26, 2005, the licensee documented the issue in their corrective action program as CR 05-04057 and submitted a request for assistance to Plant Engineering.
Therefore, this issue will be considered an Unresolved Item pending further NRC review of the licensees evaluation (URI 05000346/2005301-02).
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners conducted an announced initial operator licensing examination during the weeks of July 18, 2005, and July 25, 2005. The facilitys training staff used the guidance established in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, to prepare the examination outline and to develop the written examination and operating test. The NRC examiners administered the operating test during the weeks of July 18, 2005, and July 25, 2005. Members of the Davis-Besse Nuclear Power Station Training Department administered the written examination on July 28, 2005. Four Reactor Operator (RO) and eight Senior Reactor Operator (SRO)applicants were examined. Three of the eight SRO applicants were granted eligibility deferments so they could take the examination.
b. Findings
Written Examination The licensee developed the written examination. During their review, NRC examiners determined that the initially proposed 100 question written examination (75 RO questions and 25 SRO only questions), as submitted by the licensee, was outside the acceptable quality range expected by the NRC in accordance with NUREG-1021, Revision 9. This determination was based on the observation that 16 out of 75 RO
questions (21 percent) and 3 out of 25 SRO questions (12 percent) required replacement or significant modifications and were identified as unsatisfactory. The minimum requirement to determine an adequate quality range, assessed separately for each RO and SRO examination in accordance with ES-501 of NUREG-1021, was 20 percent or fewer questions identified as unsatisfactory. Of the 19 questions identified as unsatisfactory, the questions contained various psychometric errors including low level of difficulty, more than one (or no) correct answer, examination questions that did not match the selected outline Knowledge and Ability statements, and two or more question distractors that were not plausible. In addition, 27 questions (24 RO and 3 SRO questions) needed enhancements which were required to be completed prior to administration of the examination.
During June 7 - 8, 2005, and the week of June 27, 2005, examination changes were agreed upon between the NRC and the licensee and were made according to NUREG-1021, Revision 9. The licensee graded the examination on July 28, 2005, and conducted a review of each question to determine accuracy and validity of the examination questions. The licensee submitted one post-examination question comment on August 4, 2005. The results of the NRCs review of the stations comments are documented in Attachment 3, Post Examination Comments and Resolutions.
Operating Test The licensee developed the operating test, including the JPM walkthrough and the dynamic simulator scenarios. The NRC examiners determined that the operating test, as originally submitted by the licensee, was within the acceptable quality range expected by the NRC in accordance with NUREG-1021, Revision 9. The NRC examiners validated the operating test during the validation week and replaced or modified several items in the proposed operating test. Test changes, agreed upon between the NRC and the licensee, were made in accordance with NUREG-1021 guidelines.
Examination/Test Results Twelve applicants passed all sections of their examinations, nine of these applicants were issued respective operator or senior operator licenses. During the NRCs review of initial license applications, three senior operator applicants were granted eligibility deferments so that the applicants could take the examination. These three applicants will need to complete a minimum of six months onsite responsible power plant experience before they will receive a license. The applicants will be granted a license only after the NRC receives written certification from the station that the applicants have completed the eligibility requirement that was previously deferred.
.2 Examination Security
a. Inspection Scope
The NRC examiners observed the licensees implementation of examination security and integrity measures (e.g., physical barriers, sequestering, security agreements, sampling criteria, and test item repetition) throughout the examination process.
b. Findings
The licensees implementation of examination security requirements during examination preparation and administration was acceptable and met the guidelines provided in NUREG-1021, Revision 9. However, during the facility licensees self-validation of a simulator scenario to be used on the NRC initial operating test a security incident occurred which had the potential to affect the integrity of the operating test.
Following validation on April 26, 2005, an individual inadvertently removed hand written notes containing information from an operating test simulator scenario and left them unattended in a non-secure classroom for approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The individual retrieved the notes prior to leaving the Training Center, then returned the notes and informed the training staff the following morning. This was a violation of the licensees examination security procedure for controlling examination material.
The licensee documented this incident in their corrective action program as CR 05-02457. The NRC examiners were appropriately notified of the incident and the affected scenario was replaced. The examiners reviewed the licensees investigation and assessed the overall incident for possible violation of 10 CFR 55.49, Integrity of Examinations and Tests. The examiners determined that no actual examination compromise had occurred. The apparent violation of the licensees examination security procedure was considered minor in nature and was not subject to enforcement action in accordance with NRC enforcement policy.
4OA6 Meetings
.1 Exit Meeting
The chief examiner presented the examination team's preliminary observations and findings on July 28, 2005, to Mr. D. Imlay and other members of the Operations and Training Department staff. A subsequent exit via teleconference was held on August 8, 2005, with Mr. J. House following review of the site post examination comments. The licensee acknowledged the observations and findings presented. No proprietary information was identified by the stations staff during the exit meetings.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- D. Imlay, Operations Superintendent
- J. House, Training Instructor
- S. Laeng, Training Instructor
- S. Loehlein, Site Engineering Director
- S. Martin, Training Instructor
- R. Patrick, Operations Services Superintendent
- A. Shallard, Operations Training Supervisor
- J. Sturdavant, Regulatory Compliance Engineer
- M. Trump, Training Manager
Nuclear Regulatory Commission
- J. Rutkowski, Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Failure to Implement Plant Modification Into Emergency
Procedure (Section 4OA4.1)05000346/2005301-02;
Use of Device to Pin SFAS Switch (Section 4OA4.2)
Closed
Failure to Implement Plant Modification Into Emergency
Procedure (Section 4OA4.1)
Discussed
None
LIST OF ACRONYMS USED
Agency-Wide Document Access and Management System
Core Flood Tank
Division of Reactor Safety
kW
Kilowatts
MDFP
Motor Driven Feedwater Pump
Non-Cited Violation
NRC
Nuclear Regulatory Commission
Once Through Steam Generator
Publicly Available Records
Reactor Operator
SFAS
Safety Function Actuation System
SFRCS
Steam Feed Rupture Control System
Senior Reactor Operator
SIMULATION FACILITY REPORT
Facility Licensee:
Davis-Besse Nuclear Power Station
Facility Docket No.:
50-346
Operating Tests Administered:
July 18 - July 27, 2005
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
None
RO and SRO Post Examination Comment and Resolution
Question No. 52:
The plant is at 100 percent power.
The RCS pressure low trip bistable (BA 304) in SFAS Channel 3 has been tripped to comply
with a Tech. Spec action statement.
Which one of the following describes how a subsequent loss of Y1 bus will affect the Makeup
and Purification System?
- A. Seal Return will be lost due to MU 59A, MU 59B, MU 59C and MU 59D going closed.
B. Seal Injection will be lost to RCP 1-1 and RCP 2-2 due to MU 66B and MU 66C going closed.
C. RCS Makeup will be lost due to MU 6422 going closed.
D. Letdown will be lost due to MU 2A going closed.
Original correct answer: D.
Facility Comment:
During the examination a question was asked by 2 different candidates pertaining to
Question 52. In the stem of the question BA 304, the SFAS Low Pressure Bistable in SFAS
Channel 3, is tripped. Since the bistable number begins with the number 3," the candidates
ask(ed) if this should be labeled as the SFAS Low-Low Pressure Bistable in SFAS Channel 3.
The low pressure bistable will actuate SFAS level 2 components which would make distractor D
correct as shown on the answer key. The low-low pressure bistable will actuate SFAS level 3
components which would make distractors A and B correct. The examination proctor incorrectly
told the candidates that the bistable was mislabeled in the question stem and should be labeled
as SFAS Low-Low Pressure Bistable in SFAS Channel 3." Three candidates had already
completed and turned in their examination when the information was incorrectly supplied to the
rest of the candidates. This led to some candidates answering the question as written and the
rest of the students answering the question based on the incorrect information. Some students
changed their answers from the original correct answer based on the incorrect information.
Facility Recommendation: Since Question 52 has three correct answers depending upon what
information was used by the candidates, the question should be deleted in accordance with
ES 403, Step D.1.b.
NRC Resolution:
As originally written, the correct answer to Question 52 was D, and the first three applicants that
left the examination based their answers on the original Question 52. However, the stem as
modified by the proctors comments to the remaining candidates, resulted in distractors A and B
as correct answers. Based on review of the original question, discussions with the facility, and
review of supporting references for the modified Question 52, it was decided to grade the first
three candidates on Question 52 with D as the correct answer, and grade the remaining
candidates on the modified Question 52 (Question 52a) with both A and B as correct answers.
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO Initial Examination ADAMS Accession # ML052640495
SRO Initial Examination ADAMS Accession # ML052640504