IR 05000324/2015301
| ML16050A274 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 02/19/2016 |
| From: | Eugene Guthrie Division of Reactor Safety II |
| To: | William Gideon Progress Energy Carolinas |
| References | |
| 50-324/15-301, 50-325/15-301 | |
| Download: ML16050A274 (17) | |
Text
UNITED STATES ruary 19, 2016
SUBJECT:
BRUNSWICK STEAM ELECTRIC PLANT - NRC EXAMINATION REPORT 05000325/2015301 AND 05000324/2015301
Dear Mr. Gideon:
During the period of December 1 - 7, 2015, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the Brunswick Steam Electric Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests with those members of your staff identified in the enclosed report. The written examination was administered by your staff on December 15, 2015.
Four Reactor Operators (RO) and three Senior Reactor Operators (SRO) applicants passed both the operating test and written examination. One RO and three SRO applicants failed the written examination. There were five post-administration comments concerning the written examination. These comments, and the NRC resolution to the comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial written RO/SRO examination submitted by your staff failed to meet the guidelines for quality contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 10, as described in the enclosed report.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures 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 Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you have any questions concerning this letter, please contact me at (404) 997-4662.
Sincerely,
/RA/
Eugene F. Guthrie, Chief Operations Branch 2 Division of Reactor Safety Docket Nos: 50-325, 50-324 License Nos: DPR-71, DPR-62
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket No.: 05000325, 05000324 License No.: DPR-71, DPR-62 Report No.: 05000325/2015301 and 05000324/2015301 Licensee: Duke Energy Progress, Inc.
Facility: Brunswick Steam Electric Plant, Units 1 and 2 Location: Southport, NC Dates: Operating Tests - December 1 - 7, 2015 Written Examination - December 15, 2015 Examiners: R. Baldwin, Chief Examiner, Senior Operations Engineer B. Caballero, Senior Operations Engineer D. McNeil, Region III, Senior Operations Engineer D. Lanyi, Senior Operations Engineer Approved by: Eugene F. Guthrie Chief Operations Branch 2 Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS ER 05000325/2015301 and ER 05000324/2015301; operating test December 1 - 7, 2015 &
written exam December 15, 2015; Brunswick Steam Electric Plant, Units 1 & 2; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 10 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
Members of the Brunswick Steam Electric Plant training staff developed both the operating tests and the written examination. The initial written RO/SRO examination submitted by your staff did not meet the guidelines for quality contained in NUREG-1021. The initial operating test submittal met the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period of December 1 - 7, 2015.
Members of the Brunswick Steam Electric Plant training staff administered the written examination on December 15, 2015. Four Reactor Operators (RO) and three Senior Reactor Operators (SRO) applicants passed both the operating test and written examination. Three ROs and 3 SROs were issued licenses commensurate with the level of examination administered. One RO and three SRO applicants failed the written examination. One RO applicant passed the operating test, but passed the written examination with an overall score between 80% and 82%. The RO applicant was issued a letter stating that the applicant passed the examination and issuance of their license has been delayed pending any written examination appeals that may impact the licensing decision for their application.
There were five post-examination comments.
No findings were identified.
REPORT DETAILS 4. OTHER ACTIVITIES 4OA5 Operator Licensing Examinations a. Inspection Scope The NRC evaluated the submitted operating test by combining the scenario events and JPMs in order to determine the percentage of submitted test items that required replacement or significant modification. The NRC also evaluated the submitted written examination questions (RO and SRO questions considered separately) in order to determine the percentage of submitted questions that required replacement or significant modification, or that clearly did not conform with the intent of the approved knowledge and ability (K/A) statement. Any questions that were deleted during the grading process, or for which the answer key had to be changed, were also included in the count of unacceptable questions. The percentage of submitted test items that were unacceptable was compared to the acceptance criteria of NUREG-1021, Operator Licensing Standards for Power Reactors.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.
The NRC administered the operating tests during the period of December 1 - 7, 2015.
The NRC examiners evaluated five RO and six SRO applicants using the guidelines contained in NUREG-1021. Members of the Brunswick Steam Electric Plant training staff administered the written examination on December 15, 2015. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Brunswick Steam Electric Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
The NRC evaluated the performance or fidelity of the simulation facility during the preparation and conduct of the operating tests.
b. Findings No findings were identified.
The NRC developed the written examination sample plan outline. Members of the Brunswick Steam Electric Plant training staff developed both the operating tests and the written examination. All examination material was developed in accordance with the guidelines contained in NUREG-1021, Revision 10.
The NRC determined that the licensees examination submittal was outside the range of acceptable quality specified by NUREG-1021. The initial written examination submittal was outside the range of acceptable quality because more than 20% [RO exam: 22/75 (29%), SRO exam 8/25 (32%)] of questions sampled for review contained unacceptable flaws. Individual questions were evaluated as unsatisfactory for the following reasons:
statement contained in the examination outline. Several of these questions failed to meet the K/A statement because either the Tier 1 (Plant Evolution) or Tier 3 (Plant-wide Generic) topic was not tested.
- One RO question was deleted from the examination because the stem did not contain sufficient information, discussed in Enclosure 2.
The NRC determined that the licensees initial operating test submittal was within the range of acceptability expected for a proposed examination.
Four ROs and three SROs passed both the operating test and written examination. One RO and three SRO applicants failed the written examination. Three RO and 3 SRO licenses were issued. One RO applicant passed the operating test, but passed the written examination with an overall score between 80% and 82%. The RO applicant was issued a letter stating that the applicant passed the examination and issuance of their license has been delayed pending any written examination appeals that may impact the licensing decision for their application. Copies of all individual examination reports were sent to the Training Manager for evaluation and determination of appropriate remedial training.
The licensee submitted five post-examination comments concerning the written examination. A copy of the final RO written exam and answer key, final SRO written exam and answer key, with all changes incorporated, and licensee post exam comments may be accessed not earlier than December 15, 2017, in the ADAMS system (ADAMS Accession Numbers ML16007A205, ML16007A210 and ML16007A212).
4OA6 Meetings, Including Exit Exit Meeting Summary On December 8, 2015, the NRC examination team discussed generic issues associated with the operating test with you and members of the Brunswick Steam Electric Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
KEY POINTS OF CONTACT Licensee Personnel R. Gideon, VP Brunswick Nuclear Plant J. Nolan, Plant Manager J. Kalamaja, Operations Manager L. Grzeck, Regulatory Affairs Manager S, Byrd, Manager, Work Control J. Ferguson, Manager, NOS J. Hicks, Director Training M. Similey, Superintendent of Operator Training B. Bolin, Exam Writer B. Eavenson, Supervisor ASG J. Morris, Supervisor ILT
FACILITY POST EXAMINATION COMMENTS AND NRC RESOLUTION A complete text of the licensees post-examination comments can be found in ADAMS under Accession Number ML16007A212.
REACTOR OPERATOR (RO)/SENIOR REACTOR OPERATOR (SRO) WRITTEN EXAM QUESTION #2:
The licensee identified the answer key was in error for RO Question #2. Specifically, the licensee contended that the correct answer to the second part of the question was Source Range Monitor (SRM) B [answer choice C] instead of SRM C [answer choice D]. The licensee stated that the answer key error was not detected during the examination validation.
NRC RESOLUTION:
The licensees recommendation was accepted.
RO/SRO Question #2 was a two-part question; the first part of the question tested the applicants knowledge of how many Reactor Manual Control System (RMCS) switches were required to continuously move control rod 30-39 from position 12 to 48 during a reactor startup.
There were no post exam comments related to the first part of the question.
The second part of the question tested the applicants knowledge of which Source Range Monitor (SRM) would indicate the largest power change when control rod 30-39 was being continuously moved from position 12 to 48. The stem of the question included a core matrix diagram, as depicted on the Unit 2 main control board, with the SRM locations only represented by open circles; the core matrix diagram did not indicate the specific SRM (A, B, C, or D) that was represented by the circles.
The SRMs were located at the following core locations:
Since control rod 30-39 was directly adjacent to SRM B (core location 28-41), the largest observable change in neutron count rate was on SRM B. Therefore, the only correct answer was answer choice C (SRM B).
The answer key was changed to reflect that answer choice C is the only correct answer.
Enclosure 2
REACTOR OPERATOR (RO)/SENIOR REACTOR OPERATOR (SRO) WRITTEN EXAM QUESTION #75:
The licensee contended the answer key should be changed for RO/SRO Question #75.
Specifically, the licensee recommended the answer to the second part of the question be changed to answer choice C (The radiation work permit CAN be used) instead of the original answer choice D (The radiation work permit can NOT be used)..
The licensee contended that RWPs used to enter Locked High Radiation Areas routinely indicated radiation alarms and doses below 1 Rem/hour. The licensee provided an example of a RWP for reference.
NRC RESOLUTION:
The licensees recommendation was not accepted. The question was deleted from the examination based on the explanation below.
RO/SRO Question #75 was a two part question; the first part of the question tested the applicants knowledge of what a flashing red light meant, when encountered at the entry location to a valve in the drywell. There was no post exam comment related to the first part of the question.
The second part of the question tested the applicants knowledge of whether the applicable RWP was allowed to be used to perform the valve manipulation, if the highest dose rate listed on the RWP was 350 mr/hr.
Specifically, Question #75 was:
In preparation for a valve manipulation in the drywell, the applicable RWP indicates that the highest dose rate in the area is 350 mr/hr. A flashing red light is encountered at the entry location to the valve in the drywell.
Which one of the following completes both statements below?
The flashing red light indicates the area is a ___(1)___.
This RWP ___(2)___ be used to perform the valve manipulation.
A. (1) high radiation area (2) can B. (1) high radiation area (2) can NOT C. (1) locked high radiation area (2) can D. (1) locked high radiation area (originally correct)
(2) can NOT
The answer key indicated that answer choice D was the correct answer.
The licensees contention was that operators were routinely allowed to enter a locked high radiation area (LHRA) under an RWP that listed dose rate alarm settings less than 1000 mr/hr.
However, the stem of the question did not specify whether the 350 mr/hr value was a dose rate alarm set point or whether 350 mr/hr was an expected radiological condition listed on the RWP.
The applicants were not provided with an RWP for this question during the examination. There-fore, the applicants were required to make an assumption about what the 350 mr/hr value meant on the applicable RWP.
Furthermore, the question stem was confusing because it stated that the highest dose rate in the area was 350 mr/hr; however, it also stated that the entry location to the valve was posted with a flashing light. In accordance with AD-RP-ALL-2017, Access Controls For High, Locked High, and Very High Radiation Areas, a flashing light was required to be installed for accessible individual areas with radiation levels greater than 1000 mr/hr, where no enclosure exists for purposes of locking and where no enclosure can be reasonably constructed around the individual area. Therefore, since no specific valve number [or component location] was identified, the phrases in the area and entry location to the valve were not clear.
Because the stem for Question #75 was unclear, confusing, and required the applicants to make assumptions, this question was deleted from the examination and the total points were adjusted to reflect this deletion.
Senior Reactor Operator (SRO) Written Exam Question #83 The licensee contended the stem of the question did not specify the magnitude or duration of the cool down rate, and that the NRC immediate reportability requirements were indeterminate based on the stem conditions. Specifically, the licensee contended that immediate reportability to the NRC would be required if the cool down rate was severely exceeded because this could result in a serious degradation of the nuclear power plant, including its principal RCS safety barrier [10CFR50.72(b)(3)(ii)]. The licensee contended that there were two potential correct answers (choices C and D); however, the licensee recommended that the question be deleted since the two potentially correct answers conflicted with one another.
NRC Resolution The licensees recommendation was not accepted.
SRO Question #83 was a two part question; the first part of the question tested the applicants knowledge of the maximum allowable cool down rate required in accordance with Technical Specification (TS) 3.4.9, Reactor Coolant System Pressure and Temperature Limits. There was no post exam comment related to the first part of the question. The second part of the question evaluated the applicants knowledge of whether or not an immediate notification to the NRC was required if the cool down rate was exceeded. No reference was provided to the applicants for this test item. During examination administration none of the applicants asked for clarification on the magnitude or duration of the cool down rate.
The initial conditions indicated that Unit 1 was in Mode 3 and performing a reactor shutdown in accordance with GP-05, Unit Shutdown, in preparation for a refueling outage. When GP-05 was in effect, Attachment 3, Controlled Cooldown Guidance, provided guidance to assist the crews in controlling the cooldown below the point of adding heat. Therefore, the conditions provided in the stem implied that a normal cool down was in progress. There was no information in the stem which should have caused the applicants to assume that an excessive cooldown had, or would have occurred; therefore, it was inappropriate to assume that that the principal safety barrier was degraded.
If during a reactor shutdown, the cool down rate was exceeded, Tech Spec Limiting Condition of Operation (LCO), 3.4.9 applied:
The Tech Spec Bases for Required Action A.1 Completion Time stated that most violations would not be severe, and the parameter restoration could be completed in a controlled manner within the 30 minute Completion Time requirement. Additionally, the bases for Required Action A.2 stated that the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was reasonable to accomplish the engineering evaluation. The stem did not specify the results of Required Action A.2, that is, the stem did not provide that results of the engineering evaluation, which would have been required to determine that the condition of the nuclear power plant was degraded. Therefore, it was not operationally valid to immediately determine that the RCS principal safety barrier was degraded until after the completion of an engineering evaluation, which was not indicated in the stem.
NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Appendix E, Policies and Guidelines for Taking NRC Examinations, Section B.7 states:
When answering a question, do not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question.
Therefore, answer choice D was the only correct answer. No answer key changes were required to be taken based on the above discussion.
Senior Reactor Operator (SRO) Written Exam Question #88 The licensee contended the answer key should be changed for SRO Question #88.
Specifically, the licensee contended that the answer to the second part of the question was answer choice C (Unusual Event) instead of answer choice D (Alert).
The licensees original distracter analysis indicated that the highest emergency action level (EAL) classification for the event was an Alert; however, this classification was based on 0PEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions, EAL-2, Modes 4, 5, & De-fueled. The licensee contended since the plant was initially operating in Mode 3, EAL-1, Modes 1, 2, and 3, was the correct emergency plan chart, upon which to base the EAL classification.
NRC Resolution The licensees recommendation was accepted.
SRO Question #88 was a two part question; the first part of the question tested the applicants knowledge of whether a mode change occurred after a loss of offsite power occurred, with the plant initially operating in Mode 3, and subsequent heat up of 30°F. There was no post exam comment related to the first part of the question. The second part of the question tested the applicants ability to identify the highest required EAL classification for the event, using 0PEP-02.1, Brunswick Nuclear Plant Initial Emergency Actions, EAL-1 and EAL-2 as a reference.
The initial conditions in the stem of the question, prior to the loss of offsite power, stated that the Mode Switch was in the SHUTDOWN position and that Residual Heat Removal (RHR) Heat Exchanger (Hx) inlet water temperature was 214°F. Based on Tech Spec Table 1.1-1, and 0PEP-2.2.1, Emergency Action Level Technical Basis, Section 2.8, Operating Mode Applicability, the initial plant conditions was Hot Shutdown.
0PEP-2.2.1, Section 2.8, stated:
The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action is initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.
Since the plant was initially in Hot Shutdown, EAL-1, Modes 1, 2, & 3, was the correct emergency plan chart to use for the EAL determination. Based on all four diesel generators starting and remaining loaded for 30 minutes, the highest required EAL classification was SU1.1, Unusual Event.
Therefore, answer choice C was the only correct answer. The answer key was changed to reflect that answer choice C is the only correct answer.
Senior Reactor Operator (SRO) Written Exam Question #96 The licensee contended the question was not tied to the licensing duties of an SRO and was a job performed outside of the main control room. The licensee contended the on-line work management process was not specific to the SRO position and was performed by several different staff members, including Maintenance and Work Control staff. The licensee recommended that the question be deleted from the exam because it was not SRO level knowledge.
NRC Resolution The licensees recommendation was not accepted.
SRO Question #96 was a two part question; the first part of the question tested the applicants knowledge of when the work week schedule was required to be frozen in accordance with AD-WC-ALL-0200, On-Line Work Management. There was no post exam comment related to the first part of the question.
The second part of the question tested the applicants knowledge whether or not an On-line/Outage Scope Change Request was required to perform FIN (Fix-It-Now) after the work week was frozen.
The licensees distracter analysis identified the following as the basis for why Question #96 was tied to the licensing duties of an SRO:
The following items were identified regarding the tie between knowledge of the on-line work management process and the SRO job:
- NUREG-1123, Boiling Water Reactor Knowledge and Abilities Catalog for Nuclear Power Plant Operators, Revision 2, Supplement 1, Tier 3 generic knowledge and ability (K/A) statement:
G2.2.17: Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. CFR 41.10 / 43.5 / 45.13 RO 2.6 / SRO 3.8
- Brunswick Nuclear Plant Initial Licensed Operator Lesson Plan LOI-CLS-LP-201-E, Maintenance Documents, SRO/STA only Learning Objective #29:
Discuss the responsibilities of the following in accordance with WCP-NGGC-0500, Work Activity Integrated Risk Management Program:
a. FIN Team SRO b. Implementing Supervisor c. Operations Planner/SRO/STA d. Operations Shift Manager e. Work Control Center SRO
- AD-WC-ALL-0200, On-Line Work Management, Attachment 2, On-Line Work Process Roles and Responsibilities, included Operations Activities & Deliverables throughout the rolling work week schedule, including:
o Actively participate in the schedule freeze meeting o Review and commit to the final schedule o Operations Shift Managers review and approve scheduled work o Affirm the Operations Department is sufficiently staffed to support the schedule Based on the licensees distracter analysis and the previously listed items, the licensees recommendation was not accepted.
Therefore, no answer key changes were required to be taken; answer choice D was the only correct answer.
SIMULATOR FIDELITY REPORT Facility Licensee: Brunswick Steam Electric Plant Facility Docket No.: 05000325 and 05000324 Operating Tests Administered on: December, 1 - 7, 2015 This form is to used only to report observations. This observation does not constitute audit or inspection findings and, without further verification and review in accordance with Inspection Procedure 71111.11, is not indicative of noncompliance with 10 CFR 55.46. No licensee action is required in response to these observations.
While conducting the simulator portion of the operating test, examiners observed the following:
Item Description None Enclosure 3