IR 05000346/2008301
| ML081021162 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/08/2008 |
| From: | Hironori Peterson Operations Branch III |
| To: | Bezilla M FirstEnergy Nuclear Operating Co |
| References | |
| 50-346/08-301 50-346/08-301 | |
| Download: ML081021162 (38) | |
Text
ril 8, 2008
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION NRC INITIAL LICENSE EXAMINATION REPORT 05000346/2008301(DRS)
Dear Mr. Allen:
On March 11, 2008, the NRC completed the initial operator licensing examination process at your Davis-Besse Nuclear Power Station. The enclosed report presents the results of the examination which were discussed on February 28 and March 14, 2008, with Mr. Hovland and with other members of your staff.
The NRC examiners administered initial license examination operating tests from February 20, 2008 through February 28, 2008. Members of the Davis-Besse Training Department administered the initial license written examination on February 29, 2008, to the applicants. Eight senior reactor operator (SRO) and three reactor operator (RO) applicants were administered license examinations. The results of the examinations were finalized on March 19, 2008. Ten applicants passed all sections of their examinations resulting in the issuance of four senior reactor operator and three reactor operator licenses. One SRO applicant failed the written examination and was issued a proposed license denial. One SRO applicant scored less than 82 percent overall on the written examination, and one SRO applicant scored less than 74 percent on the SRO-only portion of the written examination; and in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501, Section D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted. One SRO 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 SRO applicant successfully completed all Davis-Besse Nuclear Power Plant training program requirements. The applicant will be issued a SRO license at that time.
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 SRO 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 10 out of 25 SRO questions (40.0 percent) and 11 out of 75 RO questions (14.7 percent) required replacement or significant modifications and were identified as unsatisfactory. The minimum requirement to determine an adequate quality range, assessed separately for each SRO and RO examination in accordance with ES-501 of NUREG-1021, was 20 percent or fewer questions identified as unsatisfactory.
In accordance with 10 CFR 2.390 of the NRC's "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 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 No. 50-346 License No. NPF-3
Enclosures:
1. Operator Licensing Examination Report 050000346/2008301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO & SRO) replacement or significant modifications and were identified as unsatisfactory. The minimum requirement to determine an adequate quality range, assessed separately for each SRO and RO examination in a
REGION III==
Docket No. 50-346 License No. NPF-3 Report No: 05000346/2008301(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: February 20 through March 11, 2008 Examiners: N. Valos, Chief Examiner K. Walton, Examiner M. Morris, Examiner Approved by: Hironori Peterson, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS
ER 05000346/2008301 (DRS); 02/20/08 - 03/11/08; Davis-Besse Nuclear Power 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.
Examination Summary:
$ Eleven examinations were administered (eight senior reactor operator (SRO) and three reactor operator (RO).
$ Ten applicants passed all sections of their examinations resulting in the issuance of four SRO and three RO licenses. One SRO applicant failed the written examination and was issued a proposed license denial. One SRO applicant scored less than 82 percent overall on the written examination, and one SRO applicant scored less than 74 percent on the SRO-only portion of the written examination; and in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501, Section D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted. One SRO applicant passed all portions of the examination, but had not completed all requirements to receive an operating license.
Upon completion of all training requirements, the facility must certify to the NRC, in writing, using the certification statement in Item 19.b. on the NRC Form 398, that the SRO applicant has successfully completed all Davis-Besse Nuclear Power Station training program requirements. The applicant will be issued a SRO license at that time.
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 weeks of February 18 and 25, 2008. 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 February 20 through February 28, 2008. The facility licensee administered the written examination on February 29, 2008. Eight senior reactor operator and three reactor operator applicants were examined. During the on-site validation week of January 21, 2008, the examiners audited two license applications for accuracy.
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 10 out of 25 SRO questions (40.0 percent) and 11 out of 75 RO questions (14.7 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 21 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.
This was the third consecutive examination in which there were 19 or more questions identified as unsatisfactory on the written exam. The last initial license written examination given in July 2005 was also 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 (21.3 percent) and 3 out of 25 SRO questions (12.0 percent)required replacement or significant modifications and were identified as unsatisfactory.
In addition, on the May 2004 initial license written examination, 19 of 100 questions were identified as unsatisfactory.
Written 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 February 29, 2008, and conducted a review of each question to determine the accuracy and validity of the examination questions. The licensee submitted eight post-examination question comments on March 11, 2008. The results of the NRCs review of the stations comments are documented in Attachment 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 Ten applicants passed all sections of their examinations resulting in the issuance of four senior reactor operator and three reactor operator licenses. One SRO applicant failed the written examination and was issued a proposed license denial. One SRO applicant scored less than 82 percent overall on the written examination, and one SRO applicant scored less than 74 percent on the SRO-only portion of the written examination; and in accordance with the guidelines of NUREG 1021, Operator Licensing Examination Standards for Power Reactors, ES-501, Section D.3.c, these licenses will be withheld until any appeal rights of the failed applicant are exhausted. One SRO applicant passed all portions of the examination, but had not completed all requirements to receive an operating license. Upon completion of all training requirements, the facility must certify to the NRC, in writing, using the certification statement in Item 19.b. on the NRC Form 398, that the SRO applicant has successfully completed all Davis-Besse Nuclear Power Plant training program requirements. The applicant will be issued a SRO 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
There was one issue associated with exam security identified by the licensee during the preparation of the exam. The issue associated with examination security was identified on August 10, 2007, when the key to the Initial License Examination Room was found in the door lock. This room was maintained locked for examination security. The key was in the lock from approximately 7 am until 1:30 pm. The dynamic simulator scenarios were the only part of the initial license examination being worked on in the room at the time of the event. It was verified that the computers in the room were turned off and no hard copies of the scenarios were available in the room. The examination computer files were password protected. The only three people that had logged onto the computers in the examination room were either on or placed on the Examination Security Agreement.
The licensee documented this issue in the corrective action program as Condition Report (CR) Number 07-25052.
The NRC was appropriately notified of this issue. The issue was reviewed and assessed for a possible violation of 10 CFR 55.49, Integrity of Examinations and Tests.
It was determined that no actual examination compromise had occurred. The apparent violation was considered minor in nature and was not subject to enforcement action in accordance with NRC enforcement policy.
Other than the issue identified above, the licensees implementation of examination security requirements during examination preparation and administration was acceptable and met the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors.
4OA6 Meetings
Exit Meeting The chief examiner presented the examination teams preliminary observations and findings with Mr. Hovland and other members of the licensee management on February 28, 2008. A subsequent exit via teleconference was held on March 14, 2008, with Mr. Hovland and other members of the licensee management following receipt of the site post-examination comments. The inspectors stated that they had reviewed proprietary information during the preparation and administration of the examination, but that the proprietary information would not be included in the examination report. The licensee acknowledged the observations provided.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
Enclosure 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- R. Hovland, Training Manager
- D. Saltz, Operations Superintendent
- A. Stallard, Operations Superintendent
- C. Steenburgen, Operations Training Superintendent
- C. Price, Performance Improvement Director
- J. House, Training/Examination Author
- S. Livingston, Initial License Training Supervisor
- P. Timmerman, Training/Examination Author
- A. Baker, Quality Assurance
- G. Wolf, Regulatory Compliance
NRC
- N. Valos, Chief Examiner
- K. Walton, Examiner
- M. Morris, Examiner
- T. Taylor, Acting Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None.
LIST OF ACRONYMS
ADAMS Agency-Wide Document Access and Management System
CFR Code of Federal Regulations
CR Condition Report
DRS Division of Reactor Safety
ILT Initial License Training
NRC Nuclear Regulatory Commission
PARS Publicly Available Records System
RO Reactor Operator
SCR Silicon Controlled Rectifier
SDP Significance Determination Process
SRO Senior Reactor Operator
SWO Simulator Work Order
Attachment
POST EXAMINATION COMMENTS AND RESOLUTIONS
SIMULATION FACILITY REPORT
Facility Licensee: Davis-Besse Nuclear Power Station
Facility Licensee Docket No. 50-346
Operating Tests Administered: February 20 through February 28, 2008
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
The Negative Margin light on Post Accident Monitoring Panel 2 Subcooling
Meter is dimly lit when the light should be off.
[Simulator Work Order (SWO) 08-0019 issued]
The Pressurizer Silicon Controlled Rectifier (SCR) Heater Bank Controller does
not appear to be controlling properly in AUT
- O. In several cases, the crews took
manual control the heaters. [SWO 08-0020 issued]
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 2
Plant conditions:
- The plant is at 100% power.
- Annunciator 2-4-A, LETDOWN OR MAKEUP FILTER P HI, has alarmed.
- High P across Purification Demineralizer Filter 1 is indicated at PDI MU62 on Control
Room Console C5702.
- Letdown flow indication is 18 gpm.
The above conditions could result in ______________ by the in service Letdown Purification
Demineralizer.
A. deborating of the RCS
B. channeling of the demineralizer resin bed
C. leaching of boron from the demineralizer resin
D. removal of impurities from the demineralizer resin
Answer: B
Reference: DB-OP-06006, Makeup and Purification System, Revison 19, Page 7
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Low differential pressure across the demineralizer can cause resin to slough, which can cause
the release of impurities.
With a plugged filter and therefore a high pressure on the inlet side of the filter (outlet side of the
demineralizers), this would cause a low differential pressure across the demineralizers. This
condition, according to PWR Generic Fundamentals for Davis-Besse, can cause a phenomenon
called sloughing as well as channeling. When either of these events occurs, this can cause
the resin to become unsettled and ionic exchanges as well as mechanical filtering are degraded.
Due to the resin being unsettled impurities could be released from the demineralizers and enter
the system.
Reference: PWR Generic Fundamentals for Davis-Besse - Components;
Chapter 4, Demineralizers and Ion Exchangers, Page 9 of 18
POST EXAMINATION COMMENTS AND RESOLUTIONS
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidates request to accept two correct answers. The purification demineralizers used at
Davis-Besse are deep-bed demineralizers and are not susceptible to the "sloughing"
phenomenon as described by the candidate.
Reference: PWR Generic Fundamentals for Davis-Besse - Components;
Chapter 4, Demineralizers and Ion Exchangers, Page 8 of 18
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The "sloughing" phenomenon as described by the applicant is associated with Powdered Resin
Filter Demineralizers. The purification demineralizers used at Davis-Besse are Deep-Bed
Demineralizers and are not susceptible to the "sloughing" phenomenon as described by the
candidate. In addition, Powdered Resin Filter Demineralizers are not subject to the channeling
phenomenom. Distractor A was correct per Precaution 2.2.13 of procedure DB-OP-06006,
Makeup and Purification System, which stated that Minimum allowable flow rate through each
purification demineralizer in service is 25 GPM due to resin bed channeling. Since letdown
flow was only 18 GPM (i.e., less than 25 GPM) per the question stem, and this would be the
flow through the purification demineralizer, channeling of the demineralizer resin bed could
occur. Therefore, distractor B was retained as the only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 7
Plant conditions:
- The plant is at 100% power.
- Normal equipment lineups exist.
- CCW Pump 1 in service.
- CCW Pump 3 in standby on Train 2.
Event:
- CCW Heat Exchanger 1 outlet temperature is 123°F and rising slowly.
- Annunciator 11-1-B, CCW HX 1 OUTLET TEMP HI, is in alarm.
With the above conditions, _______________.
- A. CCW Pump 1 will have to be manually stopped, and CCW Pump 3 allowed to
automatically start IAW DB-OP-02523, Component Cooling Water System Malfunctions
- B. both Emergency Diesel Generators will be rendered inoperable, due to closing their air
start valves, and TS 3.8.1.1 and 3.8.1.2 will have to be entered
- C. CCW Pump 3 will have to be manually started, using DB-OP-06262, Component Cooling
Water System Procedure and CCW Pump 1 will automatically trip on high temperature
D. the Non-Essential CCW Isolation Valves will have to be manually closed for CCW
Pump 1, due to lack of an automatic pump trip, using DB-OP-06262, Component
Cooling Water System Procedure
Answer: A
Reference: DB-OP-02523, Component Cooling Water System Malfunctions, Revision 5,
Pages 30, 31
POST EXAMINATION COMMENTS AND RESOLUTIONS
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
Although DB-OP-02523 does have the operator stop the running pump and then "verify" the
standby pump starts, it would be more consistent with conservative operations to manually start
the standby pump and then secure the running pump once proper start of the standby pump
had been verified. This is based on the operational principle of make before break, it is better
to have degraded cooling capability than no cooling capability.
If the running pump was stopped and the standby pump did not start, the procedure
(DB-OP-02523) would direct restarting the previously running pump. If the standby pump was
started prior to securing the running pump, then the Non-Essential Component Cooling Water
(CCW) Isolation Valves would have to be manually closed for the previously running pump in
accordance with DB-OP-06262.
Per DB-OP-01003, Operations Procedure Use Instructions, Step 6.1.5.d:
In off-normal and emergency situations, and when procedures are inadequate for that
situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and
site personnel, and return the plant to a stable and safe condition.
Per NORM-OP-1002, Conduct of Operations Handbook, Page 21, Step 6:
Procedures are adhered to and followed with a thorough understanding and focus on the
task. Take actions based on sound operational principles, not solely on compliance with the
rules.
References:
DB-OP-06262, Component Cooling Water System Procedure
DB-OP-01003, Operations Procedure Use Instructions
NORM-OP-1002, Conduct of Operations Handbook
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidates request to accept two correct answers. Although technically the order in which the
CCW pumps are started and stopped should not matter, DB-OP-02523 directs the running
pump to be stopped and the verification of the automatic start of the standby pump. Following
the abnormal procedure as written should not cause any damage to plant equipment.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Taking the actions described in distractor "D" would lead to a loss of CCW to many important
components such as RCPs, CRDMs, and the Letdown Coolers until the standby CCW Pump
was started and the Non-Essential CCW Isolation Valves for CCW Loop 2 were opened.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The applicant stated that although abnormal procedure DB-OP-02523 does have the operator
stop the running pump and then "verify" the standby pump starts (consistent with distractor A
being the correct answer), it would be more consistent with conservative operations to manually
start the standby pump and then secure the running pump once proper start of the standby
pump had been verified. The applicant referred to procedure DB-OP-01003, Operations
Procedure Use Instructions which stated that in off-normal and emergency situations, and
when procedures are inadequate for that situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and site
personnel, and return the plant to a stable and safe condition.
However, abnormal procedure DB-OP-02523 was adequate for the conditions stated in the
question stem. Thus, the guidance provided in DB-OP-01003, Operations Procedure Use
Instructions, as stated above was not applicable for the situation and following the abnormal
procedure as written would not have caused any damage to plant equipment.
In addition, distractor D did not state to manually start standby CCW Pump 3, as the applicant
stated would be a better alternative to the guidance provided in DB-OP-02523. Taking the
actions described in distractor "D" would lead to a loss of CCW to many important components
such as RCPs, CRDMs, and the Letdown Coolers until the standby CCW Pump was started and
the Non-Essential CCW Isolation Valves for CCW Loop 2 were opened. Therefore, distractor
A was retained as the only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 8
Plant conditions:
- The plant is at 100% power.
- Normal equipment lineups exist.
Event:
- Annunciator 11-3-A, CCW SURGE TK LVL LO, is in alarm.
- CCW Surge Tank level is 39 inches and slowly lowering.
- CCW temperatures are steady.
Which ONE of the following actions must be taken at this CCW Surge Tank level?
A. Trip the reactor.
B. Lock out Waste Gas Compressors.
C. Stop all four Reactor Coolant Pumps.
- D. Verify CC2645 and CC2649, Aux Bldg Return Valves, are closed.
Answer: B
References: DB-OP-02011, Heat Sink Alarm Panel 11 Annunciators, Revision 7,
Pages 29, 30
DB-OP-02523, Component Cooling Water System Malfunctions, Revision 5,
Pages 4, 16, 17
Applicant Comment:
An applicant commented that answer A should also be accepted as correct.
Forty-five (45) inches in the CCW Surge Tank is the point at which the abnormal procedure
DB-OP-02523 directs the operator to send field operators to shutdown affected equipment,
including the Waste Gas compressors. The equipment specified in the procedure, although
important, is not safety significant. Also, if the surge tank level was at 39 inches, then this order
should have already been directed to the field.
POST EXAMINATION COMMENTS AND RESOLUTIONS
The setpoint to manually trip the Reactor is 35 inches in the CCW Surge Tank. Since the level
was described as 39 inches and lowering, it would be conservative to trip the Reactor, since
inches is imminent and the Auxiliary Building Non-Essential header has already been
isolated (47 inches), and the next isolation of the entire Non-Essential Header (37 inches) is
about to occur which would cause a loss of cooling to the makeup pumps, control rod drives,
letdown coolers, RCP seal return coolers, as well as other important equipment.
Per DB-OP-01003, Operations Procedure Use Instructions, Step 6.1.5.d:
In off-normal and emergency situations, and when procedures are inadequate for that
situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and
site personnel, and return the plant to a stable and safe condition.
Per NORM-OP-1002, Conduct of Operations Handbook, Page 21, Step 6:
Procedures are adhered to and followed with a thorough understanding and focus on the
task. Take actions based on sound operational principles, not solely on compliance with the
rules.
References:
DB-OP-02523, Component Cooling Water System Malfunctions, Page 16
DB-OP-01003, Operations Procedure Use Instructions
NORM-OP-1002, Conduct of Operations Handbook
DWG-OS-021, Sheets 1, 2, 3, Component Cooling Water System
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidates request to accept two correct answers. DB-OP-02523 provided direction for the
actions to be taken at various levels in the CCW Surge Tank. Distractor 'A" (Trip the Reactor),
Distractor "C" (Stop all 4 RCPs) and Distractor "D" (Verify CC2645 and CC2649 are closed)
were all actions required to be taken at 35 inches.
POST EXAMINATION COMMENTS AND RESOLUTIONS
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The applicant referred to procedure DB-OP-01003, Operations Procedure Use Instructions
which stated that in off-normal and emergency situations, and when procedures are inadequate
for that situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and site
personnel, and return the plant to a stable and safe condition.
However, abnormal procedure DB-OP-02523 was adequate for the conditions stated in the
question stem. DB-OP-02523 provided direction for the actions to be taken at various levels in
the CCW Surge Tank. The question stem asked what actions must be taken at the CCW Surge
Tank level of 39 inches. At 45 inches in the CCW Surge Tank, one of the actions required was
to shutdown equipment on the Auxiliary Building Non-Essential CCW header, which included
the Waste Gas Compressors. Since CCW Surge Tank level was at 39 inches, securing the
Waste Gas Compressors was an action that must be taken. Distractor 'A" (Trip the Reactor),
Distractor "C" (Stop all 4 RCPs) and Distractor "D" (Verify CC2645 and CC2649 are closed)
were all actions to be taken at 35 inches.
Thus, the guidance provided in DB-OP-01003, Operations Procedure Use Instructions, as
stated above was not applicable for the situation and following the abnormal procedure as
written would not have caused any damage to plant equipment. Therefore, distractor B was
retained as the only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 31
Plant conditions:
- The plant is at 100% power.
- Selected Pressurizer Level indication is from LT RC14-2.
- Selected Pressurizer Temperature indication is from TT RC15-1.
Event:
- A loss of NNI-X AC Power has occurred.
With the loss NNI-X AC Power, _______________.
A. transferring the Pressurizer level transmitter and the Pressurizer temperature transmitter
to NNI-Y powered transmitters will restore compensated Pressurizer level indication to
the recorder
B. Pressurizer level control will remain in automatic with compensated level indication
available
C. Compensated Pressurizer level will have to be obtained from the recorder since the
computer point for compensated Pressurizer temperature indication is not available
D. transferring the Pressurizer level transmitter and the Pressurizer temperature transmitter
to NNI-Y powered transmitters will allow Compensated Pressurizer level to be obtained
from the computer point since the recorder is not available
Answer: D
Reference: DB-OP-02532, Loss Of NNI/ICS Power, Revision 6, Page 15
Applicant Comment:
An applicant commented that the question has no correct answer.
The question was asking how, upon a loss of NNI-X AC Power, the operator would monitor
actual compensated Pressurizer level.
The procedure (DB-OP-02532, Step 4.1.4) stated that:
Transferring PZR level and temperature control to NNI-Y WILL restore:
POST EXAMINATION COMMENTS AND RESOLUTIONS
- Compensated PZR level to the PZR level controller
- Compensated PZR Heater cutoff at 40 inches
- Annunciator alarms
The procedure also stated that:
Compensated PZR level MAY be available from Computer Point L768 or use uncompensated
PZR level and refer to DB-PF-06703, Misc Operational Curves
Answer D stated that transferring the inputs WILL ALLOW compensated Pressurizer level to
be monitored from the computer point. During training instruction was not given to utilize this
computer point, which has to be pulled up from a different source than normally used. Instead,
instruction was to use the uncompensated level indicators and reference the curve from
DB-PF-06703.
Therefore answers B, C, and D were eliminated immediately and answer A was left as the
only choice. As shown by the test results, approximately 50% of the candidates selected A
and missed this question.
References:
DB-OP-02532, Loss of NNI/ICS Power
DB-PF-06703, Miscellaneous Operational Curves
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidate's claim that there are no correct answers. Unless another malfunction occurred,
compensated Pressurizer level would be available from the computer point described in
DB-OP-02532 when the NNI Y powered level and temperature transmitters were selected. The
electronic components in the compensated Pressurizer level instrument string are powered from
NNI X-DC and are not affected by a loss of NNI X-AC.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The applicant stated that during training instruction was not given to utilize the computer point to
obtain Compensated Pressurizer level (as stated in procedure DB-OP-02532). Instead
instruction was given to obtain Compenensated Pressurizer level from the uncompensated level
indicators and to reference the curve from DB-PF-06703, Miscellaneous Operational Curves.
However, procedure DB-OP-02532, Step 4.1.4 stated that Compensated Pressurizer level could
be obtained from either (1) the computer point as stated in procedure DB-OP-02532, or
POST EXAMINATION COMMENTS AND RESOLUTIONS
(2) using uncompensated level and referencing the curve from DB-PF-06703, Miscellaneous
Operational Curves, for guidance. Even if the applicant was not trained on the first method
mentioned above to obtain Compensated Pressurizer level (i.e., using the computer point) and
was only trained on the second method (i.e., using uncompensated level and referencing the
curve from DB-PF-06703), the first method was an acceptable method of obtaining
Compensated Pressurizer level and was specified in procedure DB-OP-02532. Therefore,
distractor D was retained as the only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 33
Which ONE of the following will prevent the Main Fuel Handling Bridge from positioning over the
transfer basket?
A. Transfer Carriage stopped in the transfer tube.
B. Transfer Carriage Control Switch in the Containment Building is ON.
C. During frame-up operation of the upending frame.
D. During frame-down operation of the upending frame.
Answer: D
Reference: DB-NE-06306 Fuel Transfer System Operating Procedure (Rev 1)(Page 12)
Applicant Comment:
An applicant commented that answer C should also be accepted as correct.
The logic drawings for the Main Fuel Handling Bridge (MFHB) shows that the bridge can not
move in the upender region during operation (up or down) of the upender.
The purpose of this interlock was to ensure that the MFHB was not moving in the transfer area
during operation of the upender. The logic drawings show that in order for this interlock to be
clear and allow movement of the MFHB into the transfer area, the upender must be locked in
the full frame-up position. Both answers C and D implied that the upending frame was
moving, and therefore the interlock would not be satisfied in either answer.
Only 18% of the candidates guessed correctly on this question, 46% chose answer C.
References:
V Drawings:
2040-27, -30, -31
2038-10, -11, -12
M-519-281-3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Facility Proposed Resolution:
The facility disagrees with the candidates request to accept two correct answers. In
DB-NE-06306, Note 3.2.7.b, it stated that "during frame-down operation the MFHB can not be
positioned over the up-ending frame." In actuality the interlock prevents frame-down operation
when the MFHB is positioned over the up-ending frame. The performance of DB-PF-10137,
Main Fuel Handling Bridge Acceptance Test, verifies proper operation of the fuel handling
equipment interlocks prior to use. Section 6.8 of DB-PF-10137 verifies proper operation of the
interlocks for the fuel transfer system baskets: Steps 6.8.9 through 6.8.12 verifies the interlock
for the East Transfer System, and Steps 6.8.19 through 6.8.22 verifies the interlock for the West
Transfer System.
Based on this information the facility recommends that the question should be deleted, since an
interlock does not exist to prevent positioning the MFHB over the transfer basket.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to delete the question from the examination.
Upon review there is no interlock that would prevent positioning the MFHB over the transfer
basket. The actual interlock prevents frame-down operation when the MFHB is positioned over
the up-ending frame (not the transfer basket). The performance of DB-PF-10137, Main Fuel
Handling Bridge Acceptance Test, verifies proper operation of the fuel handling equipment
interlocks prior to use. Section 6.8 of DB-PF-10137 verifies proper operation of the interlocks
for the fuel transfer system baskets. In Steps 6.8.9 through 6.8.12 of DB-PF-10137, the
interlock is verified for the East Transfer System, and in Steps 6.8.19 through 6.8.22, the
interlock is verified for West Transfer System.
Since an interlock does not exist to prevent positioning the MFHB over the transfer basket,
there was no correct answer, and it was decided to delete the question from the examination.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 41
Plant conditions:
- The reactor tripped from 100% power due to a large break LOCA which was followed by
a loss of offsite power.
- Both Emergency Diesel Generators are running powering the C1 and D1 4160V busses.
- RCS Pressure is 188 psig.
- LPI flow in Loop 1 is 1350 gpm.
- LPI flow in Loop 2 is 950 gpm.
- BWST Level is at 9.2 feet and lowering at 3 feet per hour.
Given these conditions, the HPI pumps can _____________.
A. be stopped since LPI flow exists in both loops
B. be stopped since LPI Loop 1 has reached 1350 gpm
C. NOT be stopped since they will be needed for Piggyback Operation
D. NOT be stopped and the HPI Alternate Minimum Recirc Flowpath must be placed in
service
Answer: C
Reference: DB-OP-02000, Revision 20, Pages 165, 171, 242
Applicant Comment:
An applicant commented that answer D should also be accepted as correct.
If dose rates are not preclusive to placing the alternate minimum recirculation in service, then it
would be more conservative to do so to ensure minimum flow protection for the High Pressure
Injection (HPI) pumps.
Although the procedure stated that it was required to place the Alternate Minimum Recirculation
Flowpath in service only if the depletion rate of the Borated Water Storage Tank was less than
feet per hour, it would be consistent with conservative operations to do so as long as the dose
rates within the areas required were not preclusive to sending an operator to do so. As the HPI
Pumps are extremely important to the safety of the core, ensuring that they are not damaged
after swapping suctions was a valid concern.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Per DB-OP-01003, Operations Procedure Use Instructions, Step 6.1.5.d:
In off-normal and emergency situations, and when procedures are inadequate for that
situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and
site personnel, and return the plant to a stable and safe condition.
Per NORM-OP-1002, Conduct of Operations Handbook, Page 21, Step 6:
Procedures are adhered to and followed with a thorough understanding and focus on
the task. Take actions based on sound operational principles, not solely on compliance
with the rules.
References:
DB-OP-02000, RPS, SFAS, SFRCS Trip or SG Tube Rupture, Revision 20
DB-OP-01003, Operations Procedure Use Instructions
NORM-OP-1002, Conduct of Operations Handbook
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidates request to accept two correct answers. The HPI Pumps can not be stopped per
Specific Rule 3 since Low Pressure Injection (LPI) flow was less than 1350 gpm per line. This
required HPI to be piggybacked with LPI in accordance with the "Response Not Obtained"
guidance for step 10.12 of DB-OP-02000. Placing the HPI alternate minimum recirculation in
service was only required if the BWST level was decreasing at less than 2 feet per hour.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The applicant agreed with the first part of the distractors C and D that given the conditions in
the question stem, that the HPI pumps can NOT be stopped. The HPI Pumps can not be
stopped per Specific Rule 3, MU/HPI/LPI Control, since Low Pressure Injection (LPI) flow was
less than 1350 gpm (i.e., 950 gpm) in Loop 2, and LPI flow was required to be at least
1350 gpm in both loops for 20 minutes or more in order to be able to secure the HPI Pumps.
The applicant stated that although emergency procedure DB-OP-02000 does not place the
Alternate Minimum Recirculation Flowpath in service unless the depletion rate of the Borated
Water Storage Tank (BWST) was less than 2 feet per hour (the BWST was lowering at 3 feet
per hour per the question stem), it would be consistent with conservative operations to place the
Alternate Minimum Recirculation Flowpath in service (as per distractor D) as long as the dose
POST EXAMINATION COMMENTS AND RESOLUTIONS
rates within the areas required did not prevent sending an operator to perform the required
alignments. The applicant referred to procedure DB-OP-01003, Operations Procedure Use
Instructions which stated that in off-normal and emergency situations, and when procedures
are inadequate for that situation, the Shift Manager shall:
1. Take actions necessary to minimize personnel injury and damage to the plant.
2. Take actions necessary to protect the health and safety of the general public and site
personnel, and return the plant to a stable and safe condition.
However, emergency procedure DB-OP-02000 was adequate for the conditions stated in the
question stem. Procedure DB-OP-02000 provided direction for the actions to be taken
depending on the drawdown rate of the BWS
- T. Thus, the guidance provided in DB-OP-01003,
Operations Procedure Use Instructions, as stated above was not applicable for the situation
and following the emergency procedure as written would not have caused any damage to plant
equipment. Therefore, distractor C was retained as the only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 53
Which ONE of the following would preclude using the Temporary Diesel Air Compressor during
a Severe Loss of Instrument Air event?
A. The station air system is being used for breathing air.
B. The leak or rupture is in the Station Air System piping.
C. The air will be used to supply the Instrument Air System.
- D. The air compressor is an oil flooded air compressor (i.e., The Worthington Blue Brute).
Answer: A
References:
DB-OP-06251, Station and Instrument Air System Operating Procedure, Revision 17, Page 57
DB-OP-02528, Loss of Instrument Air, Revision 11, Page 12
Applicant Comment:
An applicant commented that answer B should also be accepted as correct.
If the leak or rupture was in the Station Air piping prior to the Temporary Diesel Air Compressor
inlet isolation valves (SA54 and SA55) to the Station Air Compressor Receivers (which would
separate the temporary connection from the Station Air Receivers), then this would also
preclude the use of the Temporary Diesel Air Compressor. The rupture in this area of the
Station Air system piping would not allow the Temporary Diesel Air Compressor to pressurize
the Instrument or Station Air systems.
Reference:
Drawing OS-019B, Sheet 1
Facility Proposed Resolution:
The question grading for the exam should not change. The facility disagrees with the
candidates request to accept two correct answers. The location of a leak in the Station Air
system does not preclude placing the Temporary Diesel Air Compressor in service. In the leak
location described by the candidate, the following would prevent a loss of Instrument Air:
(1) check valve SA 17, and (2) normally closed solenoid valve SA 6445 (Instrument Air to
POST EXAMINATION COMMENTS AND RESOLUTIONS
Station Air Crosstie). If SA 6445 was open, then it would automatically close on lowering
Instrument Air header pressure (see Control Logic 6 on print OS-019B Sheet 3). Solenoid valve
SA 6445 Manual Bypass Valve, SA 236, is normally closed. Also, the Emergency Instrument
Air Compressor would automatically start and maintain Instrument Air pressure.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to accept only the original correct answer.
The applicant stated that if the leak or rupture was in the Station Air piping prior to the
Temporary Diesel Air Compressor inlet isolation valves (SA54 and SA55) to the Station Air
Compressor Receivers (which would separate the temporary connection from the Station Air
Receivers), then this would also preclude the use of the Temporary Diesel Air Compressor. The
applicant stated that rupture in this area of the Station Air system piping would not allow the
Temporary Diesel Air Compressor to pressurize the Instrument or Station Air systems.
However, in the leak location described by the candidate, the following would prevent a loss of
Instrument Air event: (1) check valve SA 17, and (2) normally closed solenoid valve SA 6445
(Instrument Air to Station Air Crosstie). If SA 6445 was open, then it would automatically close
on lowering Instrument Air header pressure. Also, valve SA 236, Solenoid Valve SA 6445
Manual Bypass Valve, is normally closed. In addition, the applicant made an assumption that
there was a leak in a specific location in the Station Air system. The location of the leak that
was causing the Severe Loss of Instrument Air event was not specified in the question stem. In
NUREG-1021, Appendix E, Policies and Guidelines for Taking NRC Examinations, it was
stated, in part, that When answering a question, do not make assumptions regarding conditions
that are not specified in the question .... The applicants were briefed verbatim on the contents
of NUREG-1021, Appendix E prior to the administration of the written examination, and were
provided a copy of Appendix E. The applicants did not ask for a clarification of the question
during the administration of the written examination.
Since in the leak location described by the candidate, there would not be a loss of Instrument
Air event, and there was no discussion in the question stem concerning the location of the leak
that was causing the Severe Loss of Instrument Air event, distractor A was retained as the
only correct answer.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 83
Plant conditions:
- The plant has been tripped due to a serious fire in the Control Room.
- All four RCPs are running.
- The Control Room staff has manned the appropriate stations outside the Control Room.
- DB-OP-02501, Serious Station Fire, has been implemented.
- DB-OP-06903, Plant Shutdown and Cooldown, is being referred to.
- RCS Pressure is 1675 psig and stable.
- RCS Tave is 450°F and lowering at 77°F/hr.
Event:
- HPI Pump 1 starts automatically
- Pressurizer level is 97 inches
- RCS Pressure is 1710 psig
- Cooldown rate is 80°F/hr
With the above conditions, _______________.
A. HPI Pump 1 will be stopped and RCS pressure will be reduced to 1650 so the SFAS
signal can be blocked
B. HPI Pump 1 will be stopped and Pressurizer level will be reduced to <85 inches to
prevent an overpressurization potential
C. HPI Pump 1 will be left running and RCS cooldown rate will be lowered to < 50°F/hr to
prevent the development of a head bubble
D. HPI Pump 1 will be left running and RCS cooldown rate will be raised to 100°F/hr to
return the operating point to the desired envelope
Answer: B
References:
DB-OP-06903, Plant Shutdown and Cooldown, Revision 27, Page 10
TS 3.4.2, Safety Valves
POST EXAMINATION COMMENTS AND RESOLUTIONS
DB-OP-02501, Serious Station Fire, Revision 12, Page 23
Applicant Comment:
An applicant commented that answer A should also be accepted as correct.
The action to block the Safety Features Actuation System (SFAS) would also be taken. RCS
pressure would be reduced as well as Pressurizer level and SFAS would be blocked, if possible.
This question was unclear and unrealistic at best. It stated that the plant had been tripped due
to a Serious Fire in the Control Room and then stated that all four Reactor Coolant Pumps
(RCPs) were running; this would be in conflict with DB-OP-02519, Serious Control Room Fire,
which directed tripping RCPs 1-1 and 2-2. Then the question stated that DB-OP-02501,
Serious Station Fire, had been implemented, which would have been the incorrect procedure
for the stated initial conditions of a Serious Control Room Fire. The event stated that High
Pressure Injection (HPI) Pump 1 auto started but the question did not state why this occurred
(assume due to fire?). So if SFAS had not been blocked and the plant was in a Plant Shutdown
and Cooldown (per DB-OP-06903), then the proper action to take would be to block SFAS, if it
is possible to do so. Tthe question does not provide adequate, accurate information from which
to make a decision between answers A and B.
50% of the candidates chose answer A and 50% chose Answer B
References:
DB-OP-02501, Serious Station Fire
DB-OP-02519, Serious Control Room Fire
DB-OP-06903, Plant Shutdown and Cooldown
Facility Proposed Resolution:
The facility disagrees with the candidates request to accept two correct answers. With a serious
Control Room fire as described in the question stem, the Control Room staff would have
implemented DB-OP-02519, Serious Control Room Fire. DB-OP-02519 does not implement
DB-OP-02501, Serious Station Fire as described in the question stem. The RCPs would not be
running as the candidate described, as the pumps are either stopped from the Control Room
prior to evacuation or stopped locally at the breakers after evacuation in accordance with
DB-OP-02519. The HPI pump would be deenergized in accordance with DB-OP-02519 to
prevent an automatic start. Additionally, blocking SFAS would not be possible, as the candidate
described, since SFAS can only be blocked from the Control Room.
POST EXAMINATION COMMENTS AND RESOLUTIONS
Based on this information, the facility recommends that the question should be deleted, since
DB-OP-02501 would not have been implemented and the RCPs and HPI Pump would not be
running in accordance with DB-OP-02519.
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided to delete the question from the examination.
The question stem lacked focus and provided several pieces of information which were
inconsistent with the procedural directions required to be followed for a Serious Fire in the
Control Room. Examples included the following:
- The question stem stated that the plant had been tripped due to a Serious Fire in the Control
Room and that procedure DB-OP-02501, Serious Station Fire, had been implemented.
However, for a Serious Fire in the Control Room, procedure DB-OP-02519, Serious Control
Room Fire, would have been the correct procedure to implement instead of DB-OP-02501
(if one went initially to DB-OP-02501, then Attachment 1, Page 5, would have directed one
to the correct procedure DB-OP-02519 to address the event).
- The question stem stated that all four RCPs were running. Having four RCPs running would
be in conflict with DB-OP-02519, Step 4.1.2.b.2, which directed tripping RCPs 1-1 and 2-2.
- The question stem stated that DB-OP-06903, Plant Shutdown and Cooldown, was being
referred to, and that a plant cooldown of 77°F/hour was in progress. However, per
DB-OP-02519, Attachment 1, Step 4.0, procedure DB-OP-06903 was only referenced in
order to maintain HOT STANDBY conditions until directed to cool down the plant by the
Technical Support Center with concurrence from Operations Management. If cooling down
to COLD SHUTDOWN, DB-OP-02519, Attachment 1, Step 6.0, referred one to
DB-OP-02519, Attachment 15, Cooldown to Cold Shutdown (and not to DB-OP-06903).
Thus, Precaution and Limitation 2.2.18 from DB-OP-06903, Plant Shutdown and
Cooldown, (which required that Pressurizer level be limited to less than 85 inches when the
Reactor Coolant System (RCS) temperature was less than 500°F to prevent an RCS
overpressure potential) was not required to be referred to for this event.
- The question stem stated that HPI Pump 1 had auto started. However, for a fire in the
Control Room, procedure DB-OP-02519, Attachment 4, Step 10, tripped the breaker for HPI
Pump 1. Only later during the RCS cooldown did DB-OP-02519, Attachment 15, Cooldown
to Cold Shutdown, address using Attachment 16, Utilizing HPI Spray, to use HPI Pump 1
to spray the Pressurizer during the RCS cooldown for depressurization of the RCS.
Since the question stem lacked focus and provided several pieces of information which were
inconsistent with the procedural directions required to be followed for a Serious Fire in the
Control Room, there was no correct answer, and it was decided to delete the question from the
examination.
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML080920724.
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