IR 05000237/2009301
| ML091140186 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/18/2009 |
| From: | Hironori Peterson Operations Branch III |
| To: | Pardee C Exelon Nuclear |
| References | |
| 50-237/09-301, 50-249/09-301 | |
| Download: ML091140186 (27) | |
Text
ril 18, 2009
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 NRC INITIAL LICENSE EXAMINATION REPORT 05000237/2009301(DRS); 050000249/2009301(DRS)
Dear Mr. Pardee:
On March 23, 2009, the NRC completed the initial operator licensing examination process at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report presents the results of the examination which were discussed on March 13 and March 23, 2009, with Mr. Marik and with other members of your staff.
The NRC examiners administered initial license examination operating tests from March 9, 2009 through March 13, 2009. Members of the Dresden Training Department administered the initial license written examination on March 16, 2009, to the applicants. Three senior reactor operator (SRO) and four reactor operator (RO) applicants were administered license examinations. The results of the examinations were finalized on April 1, 2009. All seven applicants passed all sections of their examinations resulting in the issuance of three senior reactor operator and four reactor operator licenses.
During this examination, one finding of very low safety significance which involved a violation of NRC requirements was identified. However, because of the very low safety significance and because the issue has been entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Dresden Nuclear Power Station. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at Dresden Nuclear Power Station.
The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its 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 Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25
Enclosures:
1. Operator Licensing Examination Report 050000237/2009301(DRS);
050000249/2009301(DRS)
2. Simulation Facility Report 3. Post Examination Comments and Resolutions 4. Written Examinations and Answer Keys (RO and SRO)
REGION III==
Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25 Report No: 05000237/2009301(DRS); 05000249/2009301(DRS)
Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: Morris, IL Dates: March 9 through March 23, 2009 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 05000237/2009301 (DRS); 05000249/2009301 (DRS); 03/09/09 - 03/23/09;
Dresden Nuclear Power Station, Units 2 and 3; 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, Supplement 1.
Examination Summary:
$ Seven examinations were administered (three senior reactor operator and four reactor operator); and
$ All seven applicants passed all sections of their examinations resulting in the issuance of three senior reactor operator and four reactor operator licenses.
A. Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Severity Level IV: The inspectors identified a Severity Level IV Non-Cited Violation (NCV) of 10 CFR 55.40, Implementation, 10 CFR 50.9, Completeness and accuracy of information, and 10 CFR 55.49, Integrity of examinations and tests. For the Dresden Station March 2009 NRC Initial Operator License Examination, the inspectors identified that the examination author and the facility reviewer had initialed Step 2.b and Step 3.a.(3) of Form ES-201-2, Examination Outline Quality Checklist, on August 15, 2008, and August 19, 2008, respectively, and Step 1.c of Form ES-301-3 Operating Test Quality Checklist, on January 15, 2009, and January 20, 2009, respectively, which indicated that the operating test did not duplicate items from the applicants audit test, when, upon NRC review, it was determined that six of the 23 dynamic simulator scenario events, and one of the 15 Job Performance Measures (JPMs) for the Reactor Operator (RO) candidates were duplicated from the applicants audit test.
The finding was determined to be more than minor, because the integrity of the NRC initial operator licensing examination could have been compromised if, but for detection by the NRC examiners, the NRC examination had been administered with the duplication of the operating test items from the applicants audit test. The finding was determined to be of very low safety significance because the duplication of operating test items was discovered by the NRC examiners prior to administration of the NRC examination, the duplicate test items were either removed from the audit test or the NRC exam changed to remove the duplication, and the facility implemented examination security requirements for the audit test similar to that which was required for the NRC examination. The inspectors concluded that this finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not define and effectively communicate expectations regarding procedural compliance and for personnel to follow procedures (i.e., in the development of the NRC initial operator license examination) (H.4(b)). (Section 4OA5.2)
Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA5 Other
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners conducted an announced operator licensing initial examination during the week of March 9, 2009. The facility licensees training staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1, 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 March 9 through March 13, 2009. The facility licensee administered the written examination on March 16, 2009. Three senior reactor operator and four reactor operator applicants were examined. During the on-site validation week of February 16, 2009, the examiners audited two license applications for accuracy.
b. Findings
Written Examination The NRC examiners determined that the written examination, as originally submitted by the licensee, was within the range of acceptability expected for a proposed examination.
All changes made to the submitted examination were made in accordance with NUREG-1021, "Operator Licensing Examination Standards for Power Reactors, Revision 1, Supplement 1. The licensee graded the examination on March 16, 2009, and conducted a review of each question to determine the accuracy and validity of the examination questions. The licensee submitted three post-examination question comments on March 23, 2009. 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,"
Revision 1, Supplement 1. The inspectors identified a Severity Level IV NCV associated with the operating test as discussed in Section 4OA5.2 of this report.
Examination Results All seven applicants passed all sections of their examinations resulting in the issuance of three senior reactor operator and four reactor operator licenses.
.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, Revision 9, Supplement 1. 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 examination security identified by the inspectors during the review of the examination. Other than the issue identified below, the licensees implementation of examination security requirements during examination preparation and administration were acceptable and met the guidelines provided in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 9, Supplement 1.
Failure to Provide Complete and Accurate Information to the NRC Associated with Verifying No Operating Test Item Duplication with the Audit Test
Introduction:
The inspectors identified a Severity Level IV NCV of 10 CFR 55.40, Implementation, 10 CFR 50.9, Completeness and accuracy of information, and 10 CFR 55.49, Integrity of examinations and tests. For the Dresden Station March 2009 NRC Initial Operator License Examination, the inspectors identified that the examination author and the facility reviewer had initialed Step 2.b and Step 3.a.(3) of Form ES-201-2, Examination Outline Quality Checklist, and Step 1.c of Form ES-301-3 Operating Test Quality Checklist, indicating that the operating test did not duplicate items from the applicants audit test, when, upon NRC review, it was determined that six of the 23 dynamic simulator scenario events, and one of the 15 Job Performance Measures (JPMs) for the Reactor Operator (RO) candidates were duplicated from the applicants audit test.
Description:
While performing a review of the Dresden Station March 2009 NRC Initial Operator License Examination to verify that the operating test did not duplicate items from the applicants audit test (as required per Examiner Standard (ES) Form ES-301-3, Operating Test Quality Checklist, Step 1.c), the examiners identified that six of the 23 dynamic simulator scenario events, and one of the 15 Job Performance Measures (JPMs) for the RO candidates were duplicated from the applicants audit test.
The audit test is an examination administered by the facility to the applicants for an NRC license prior to the NRC initial licensing examination that is used by the facility to determine whether the applicants are sufficiently trained and ready to take the NRC examination. Examiner Standard (ES)-301, Preparing Initial Operating Tests, Section D.1.a., required that the operating test not duplicate items (simulator scenarios or JPMs) from the applicants audit test, and the facility licensee identify for the NRC chief examiner those simulator events and JPMs that were similar to those that were tested on the applicants audit test. The inspectors identified that the examination author and the facility reviewer had initialed Step 2.b and Step 3.a.(3) of Form ES-201-2, Examination Outline Quality Checklist, and Step 1.c of Form ES-301-3 Operating Test Quality Checklist, which indicated that the operating test did not duplicate any item from the applicants audit test.
Upon determining the duplication of the large number of operating test items on the NRC examination when compared to the audit test, the inspectors contacted the licensee concerning the reason for the duplication of test items. It was determined that the review performed by the licensee was not sufficient to ensure that no test item duplication had taken place. In addition, there was a misconception by the licensee about the requirements of the statement in ES-301, Section D.1.a, where it stated that Operating tests may not duplicate test items (simulator scenarios or JPMs) from the applicants audit test. The facility author and facility representative both believed that individual simulator scenario events could be similar or duplicated provided that the full simulator scenarios were not duplicated, and the duplication of events did not result in the predictability of events contained within the simulator scenario. The personnel involved did not understand that there was to be no duplication of simulator events (as specified later in the same paragraph in ES-301, Section D.1.a) and not just no duplication of entire simulator scenarios.
The Operator Licensing and Human Performance Branch at NRR in Headquarters were contacted for disposition of the examination. With the Operator Licensing and Human Performance Branch approval, it was determined that if the facility developed examination security requirements for the audit test similar to that which was required for the NRC examination (per Form ES-201-3, Examination Security Agreement), and if the simulator events and JPM that were duplicated on the audit test were removed, then the duplicated operating test items could remain on the NRC initial license examination.
The facility implemented examination security requirements for the audit test similar to that which was required for the NRC examination, and the six simulator events that were duplicated on the NRC examination were removed from the audit test. The JPM that was duplicated was left on the audit test and the JPM replaced on the NRC examination.
The NRC then administered the initial operator licensing examination with no duplication of operating test items remaining on the NRC examination.
The licensee and the inspectors performed an extent of condition review to check for duplication of operating test items between the audit test and the NRC examination for the last two NRC examinations in 2007 and 2008. One JPM was found duplicated on the April 2007 NRC examination and six scenario events were found duplicated on the March 2008 NRC examination.
The NRC concluded that the licensing decisions made following the April 2007 and March 2008 NRC examinations were appropriate. For the April 2007 examination, this conclusion was based on the assessment that even if each candidate had received an unsatisfactory grade on the one JPM that was found duplicated, the candidates would still have passed the JPM portion of the operating test. For the March 2008 examination, this conclusion was based on the following assessment:
- (1) a review was performed of the duplicated events and considering less than full credit for these events, the overall grading for the candidates would not have resulted in an unsatisfactory grade on the simulator portion of the operating test,
- (2) there were no critical tasks in the simulator events that were duplicated from the audit test, and
- (3) the full scenarios were not duplicated and there was no predictability in event sequencing and scenario outcome.
Analysis:
The inspectors determined that the failure to provide complete and accurate information to the NRC regarding duplication of operating test items on the NRC initial operator license examination from the audit test was a significant regulatory issue which potentially could impede or impact the regulatory process. Issues that could potentially impede or impact the regulatory process are dispositioned using traditional enforcement instead of the SDP. Using IMC 0612, Appendix B, Issue Screening, the finding was determined to be more than minor, because the integrity of the NRC initial operator licensing examination could have been compromised if, but for detection by the NRC examiners, the NRC examination had been administered with the duplication of six of the 23 dynamic simulator scenario events, and one of the 15 JPMs for the RO from the applicants audit test. The finding was determined to be of very low safety significance because the duplication of operating test items was discovered by the NRC examiners prior to administration of the NRC examination, the duplicate test items were either removed from the audit test or the NRC exam changed to remove the duplication, and the facility implemented examination security requirements for the audit test similar to that which was required for the NRC examination.
However, the finding was determined to be of significant regulatory importance because the incorrect information could have impacted a licensing decision for the individuals taking the NRC examination. This duplication of operating test items could have, without NRC intervention, affected the equitable and consistent administration of the NRC examination. The excessive duplication of test items could have led to improved performance by the applicants, which may not be indicative of their actual performance without previous exposure to the duplicated operating test items, and thus could have resulted in an unqualified applicant successfully passing the NRC examination. The inspectors concluded that this finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not define and effectively communicate expectations regarding procedural compliance and for personnel to follow procedures (i.e., in the development of the NRC initial operator license examination)
(H.4(b)).
Enforcement:
Part 55.40(b) of Title 10 of the Code of Federal Regulations stated, in part, that power reactor facility licensees may prepare the operating tests required by 10 CFR 55.45, subject to the following conditions:
- (1) Power reactor facility licensees shall prepare the required examinations and tests in accordance with the criteria in NUREG-1021; and
- (2) Pursuant to 10 CFR 55.49, power reactor facility licensees shall establish, implement, and maintain procedures to control examination security and integrity.
Accordingly, NUREG-1021, ES-201, Section C.1.f, stated, in part, that when a licensee writes its own examination, it shall develop the outlines and examinations in accordance with ES-301 and ES-401. ES-301, Section D.1.a., stated, in part, that the operating test shall not duplicate items (simulator scenarios or JPMs) from the applicants audit test, and the facility licensee shall identify for the NRC chief examiner those simulator events and JPMs that were similar to those that were tested on the applicants audit test.
Part 50.9 of Title 10 of the Code of Federal Regulations required, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. ES-301, Section E.1, stated that the facility shall use Form ES-301-3, Examination Outline Quality Checklist, to evaluate the initial operator license examination. Form ES-301-3, Step 1.c, required an assessment that the operating tests on the NRC initial operator licensing examination not duplicate items from the applicants audit test. In addition, ES-201, Section C.1.f, stated, in part, that the facility shall use Form ES-201-2, Examination Outline Quality Checklist. Form ES-201-2, Step 2.b.,
required an assessment that no scenarios were duplicated from the applicants audit test. In addition, Form ES-201-2, Step 3.a.(3), required an assessment that no JPM tasks were duplicated from the applicants audit test. When the examination was developed by the licensee, these steps (Steps 2.b and 3.a(3) of Form ES-201-2 and Step 1.c of Form 301-3) were required to be initialed by the examination author and a facility reviewer, indicating that no operating test items on the NRC initial operator license examination were duplicated from the applicants audit test.
Part 55.49, Integrity of examinations and tests, of Title 10 of the Code of Federal Regulations stated, in part, that applicants, licensees, and facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. The integrity of a test or examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, would have affected the equitable and consistent administration of the test or examination.
This includes all activities related to the preparation, administration, and grading of the tests and examinations required by this part.
Contrary to the above, on January 26, 2009, the NRC discovered that the licensee had prepared an NRC initial license operator examination, which was not developed in accordance with the criteria contained in NUREG-1021, ES-301, Section D.1.a., as required by 10 CFR 55.40(b), in that there was a large number of operating test items that were duplicated from the applicants audit test. In addition, the examination author and facility reviewer had initialed Step 2.b and Step 3.a.(3) of Form ES-201-2 on August 15, 2008, and August 19, 2008, respectively, and Step 1.c of Form ES-301-3 on January 15, 2009, and January 20, 2009, respectively, which indicated that the operating test did not duplicate items from the applicants audit test; when, in fact, upon NRC review, it was determined that six of the 23 dynamic simulator scenario events, and one of the fifteen 15 JPMs for the RO candidates were duplicated from the applicants audit test. The large number of operating test items that were duplicated from the applicants audit test had the potential, if not detected, for compromising the integrity of the initial operator licensing examination, and thus had the potential for impacting the NRCs ability to perform its regulatory function.
This finding is considered a violation of 10 CFR 55.40(b), 10 CFR 50.9, and 10 CFR 55.49. However, because this issue was not willful, was of very low safety significance, and was entered into the licensees corrective action program (IR 00889736 and IR 00892534), the issue is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000237/2009301-01; 05000249/2009301-01)
4OA6 Meetings
Exit Meeting The chief examiner presented the examination teams preliminary observations and findings with Mr. Marik and other members of the licensee management on March 13, 2009. A subsequent exit was held on March 23, 2009, with Mr. H. Dodd, Operations Training Manager, and D. Carlson, Shift Manager, 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
- S. Marik, Plant Manager
- L. Jordan, Training Director
- H. Dodd, Operations Training Manager
- D. Gronek, Operations Director
- P. Salgado, Shift Operations Supervisor
- D. Carlson, Shift Manager
- G. Morrow, Shift Manager
- M. Knott, Initial License Training Lead Instructor
- P. OConnor, Licensed Operator Requalification Lead Instructor
- S. Taylor, Regulatory Assurance Manager
- R. Rybak, Principal Licensing Engineer
- J. Griffin, Licensing Engineer
- F. Ferrero, Operations Training Instructor/Exam Author
- S. Deprest, Corporate Training Lead
- R. Coon, Corporate Training Director
- J. Ruth, Corporate Training Director
NRC
- N. Valos, Chief Examiner
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000237/2009301-01 NCV Failure to Provide Complete and Accurate Information to
05000249/2009301-01 the NRC Associated with Verifying No Operating Test Item
Duplication with the Audit Test
Closed
05000237/2009301-01 NCV Failure to Provide Complete and Accurate Information to
05000249/2009301-01 the NRC Associated with Verifying No Operating Test Item
Duplication with the Audit Test
Discussed
None.
Attachment
LIST OF ACRONYMS
ADAMS Agency-Wide Document Access and Management System
CFR Code of Federal Regulations
DRS Division of Reactor Safety
ILT Initial License Training
IR Issue Report
MMI Minor Maintenance Issue
NCV Non-Cited Violation
NRC Nuclear Regulatory Commission
PARS Publicly Available Records System
RO Reactor Operator
SDP Significance Determination Process
SRO Senior Reactor Operator
SWR Simulator Work Request
Attachment
SIMULATION FACILITY REPORT
Facility Licensee: Dresden Nuclear Power Station, Units 2 and 3
Facility Licensee Docket Nos. 50-237, 50-249
Operating Tests Administered: March 9 through March 13, 2009
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
On March 11, 2009, the computer assistance execution program for simulator
scenario ILT-N-1 failed to load properly. Because of this issue, this simulator
scenario was delayed from March 11, 2009 to March 12, 2009. The scenario that
was originally scheduled for March 12, 2009, was run on March 11, 2009,
instead. A total delay of approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> occurred on March 11, 2009, due
to this malfunction. [IR 00891280 and SWR # 11581 issued]
The Instrument Air Header Pressure indicator was intermittently sticking.
[Simulator MMI # 6683 issued]
Drywell Floor Drain Sump Flow Recorder 2-2040-1 was inoperable.
[Simulator MMI # 6684 issued]
Enclosure 2
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 78
Unit 2 was in startup, with Reactor power indicating 35%, when a transient occurred. The
following conditions are observed:
- Turbine pressure indicates zero.
- Main Generator Field Breaker is closed.
Thirty (30) seconds after the transient, what immediate action(s) is/are the Unit Supervisor
required to direct?
- A. Scram the Reactor and close the MSIVs per DEOP 100, REACTOR CONTRO
- L.
B. Control Reactor pressure with the HPCI system per DOA 2300-02, HARD CAR
- D.
C. Control Reactor pressure with the Iso Condenser per DOP 1300-03, HARD CAR
- D.
D. Reduce Recirc flow per DGP 03-01 POWER CHANGES, or insert CRAM rods per
DGP 03-04, CONTROL ROD MOVEMENTS.
Answer: D
Applicant Comment:
An applicant commented that in addition to distractor D, distractors B and C could also be
considered correct. The applicant provided the following comments:
The question stem does not provide enough information with regards to turbine pressure. The
stem could be referring to: 1st stage pressure, Electrohydraulic Control (EHC) pressure, or
throttle pressure.
If 1st stage pressure is assumed: this indicates a turbine trip, and per the question stem Reactor
power is 35%, which meets the requirements of the DOA procedure (<38.5%), which makes the
actions of the questions correct answer D (reduce Recirc flow or insert CRAM rods) NOT
required. This makes D incorrect.
If EHC pressure is assumed: this indicates a turbine trip with the Stop Valves (SVs), Control
valves (CVs), and bypass valves all closed, which is a loss of the main condenser. With a loss
of the main condenser, pressure control will be needed. This makes distractors B and C
correct.
If throttle pressure is assumed: this indicates that the Main Steam Isolation Valves (MSIVs) are
closed, eliminating the bypass valves, making pressure control needed. This makes distractors
B and C correct. Also, if the MSIVs close this would also cause a Reactor scram. If a scram
occurred, DEOP 100 would be entered on Reactor pressure and DEOP 100 directs controlling
Reactor pressure with either the Isolation Condenser or High Pressure Coolant Injection (HPCI).
This makes distractors B and C correct.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Facility Proposed Resolution:
The facility commented that the answer key should be changed to delete the question from the
examination. The facility provided the following comments:
The stem establishes Reactor power at 35% prior to a transient. The indications provided
following the transient are turbine pressure indicates zero and Main Generator Field Breaker
is closed. No other information is provided in the stem of the question. Based on these
indications the candidate cannot eliminate a primary containment group 1 isolation (Main Steam
Isolation Valve closure) as the plant transient. Primary containment group 1 isolation will result
in isolating steam to the main turbine and subsequently all turbine steam pressures will go to
zero and the turbine will subsequently trip. Main turbine bypass valves will no longer be
available to control reactor pressure. DOA 5600-01 directs that if bypass valves are not
controlling Reactor pressure, then control Reactor pressure with the isolation condenser or
HPCI.
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 states that with Reactor power at 35%, a transient occurred, with the
following conditions observed:
- Turbine pressure indicates zero.
- Main Generator Field Breaker is closed.
The question then asks at 30 seconds after the transient, what immediate action(s) is/are the
Unit Supervisor required to direct?
Evaluation of the conditions stated in the question stem revealed the following two plant
conditions that could have resulted in the indications specified:
1) A turbine trip occurred.
For this plant condition, the correct procedure to enter would be DOA 5600-01, Turbine
Trip. Immediate Operator Action C.4 of DOA 5600-01 requires that if a turbine trip
occurs without a Reactor scram, to Reduce recirculation flow OR insert CRAM rods as
needed in accordance with DGP 03-01 or DGP 03-04 to maintain core thermal power
below 38.5% of rated (1137 MWth) and within the capabilities of the Main Condenser
Bypass Valves. These required actions correspond to distractor D being the correct
answer.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
2) A Group 1 isolation has occurred (i.e., all MSIVs have closed).
- For this plant condition, the Group 1 isolation would result in a reactor scram. The
reactor scram would result in an automatic turbine trip. With the reactor initially at
35% power, an entry into DEOP 100, RPV Control, would result due to Reactor
Pressure Vessel (RPV) water level lowering below eight (8) inches due to void
collapse in the RPV and the RPV level control provided by Feedwater Level Setpoint
Setdown.
An evaluation of the applicable procedures and the actions required are the following:
- Following the automatic turbine trip, procedure DOA 5600-01, Turbine Trip, was
applicable. The first Immediate Operator Action of DOA 5600-01 is to verify that the
main turbine bypass valves are controlling reactor pressure. Since the main turbine
bypass valves would not be controlling reactor pressure with the MSIVs closed, the
same Immediate Operator Action step of DOA 5600-01 states to then control reactor
pressure with one of the following:
(1) Isolation Condenser
(2) HPCI
Using the Isolation Condenser to control reactor pressure corresponds to selecting
distractor C as the correct answer. Using HPCI to control reactor pressure
corresponds to selecting distractor B as the correct answer.
- Following entry into DEOP 100, the DEOP 100 flowchart would require the operator
to enter procedure DGP 02-03, Reactor Scram, while continuing with the DEOP
100 flowchart actions. The Pressure leg of DEOP 100 would then require that RPV
pressure be stabilized below 1050 psig using the main turbine bypass valves.
However, since the main turbine bypass valves would not be available for stabilizing
RPV pressure with the MSIVs closed, DEP 100 required the operator to use
Alternate Pressure Control Systems in a provided list that included:
- Isolation Condenser
- Alternate Depressurization System (ADS) valves
- HPCI
- Other equipment of lesser pressure control capacity
Though no DEOP 100 actions are specifically designated as Immediate Actions,
using the Isolation Condenser (as specified in the list above) to control reactor
pressure would correspond to selecting distractor C as the correct answer. Using
HPCI (as specified in the list above) to control reactor pressure would correspond to
selecting distractor B as the correct answer.
Based on the above evaluation, distractor D would be correct answer for the case in which a
turbine trip was the initiating event for the plant transient. However, either distractor B or
distractor C would be the correct answer for the case in which a Group 1 isolation was the
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
initiating event for the plant transient. Since there are potentially three correct answers, it was
decided to delete the question from the examination.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 89
Both units are operating at near rated power, when the following occurred:
- At time 05:01 Rx Building D/P was 0.45 inches vacuum, when the Rx Building Exhaust Fan
tripped.
- At time 05:03 Rx Building D/P was 0.35 inches vacuum and becoming less negative.
- At time 05:04 Annunciator 923-5 C-1 RX BLDG DP LO came in.
Assuming a constant rate of change, what is the EARLIEST time that Secondary Containment
will exceed the Tech Spec limit AND what is the Basis for this LCO?
A. 05:05;
Minimize untreated radioactive release during a LOCA
B. 05:05;
Minimize untreated radioactive release due to a breach of the Drywell or the Torus
C. 05:06;
Minimize untreated radioactive release due to a LOCA
D. 05:07;
Minimize untreated radioactive release due to a breach of the Drywell or the Torus
Answer: A
Applicant Comment:
An applicant commented that the correct answer should be changed from distractor A to
distractor C. The applicant provided the following comments:
Question stem clearly asks what is the earliest time that secondary containment delta-pressure
(D/P) will exceed the Tech Spec limit. Provided data reveals at time 05:01 D/P is 0.45 inches
vacuum. At time 05:03, D/P is 0.35 inches and lowering at a constant rate.
Above data indicates rate of change is 0.1 inch of vacuum every 2 minutes. As a result,
conditions at time 05:05 are still within Tech Spec SR 3.6.4.1 of >0.25 inches. In other words,
at time 05:05 D/P is 0.25 inches.
Question is challenged on the basis that at the time 05:06 the Tech Spec Limiting Condition for
Operation (LCO) Surveillance Requirement (SR) is exceeded and this is the earliest of the
provided times. This time and provided basis for the LCO would make distractor C the correct
answer.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Facility Proposed Resolution:
The facility agreed with the applicant and commented that the answer key should be changed
so that the correct answer is changed from distractor A to distractor C. The facility provided
the following comments:
The question asks: Assuming a constant rate of change what is the EARLIEST time that
Secondary Containment will exceed the Tech Spec limit.
This question was originally submitted with the starting D/P value of 0.50 inches (instead of 0.45
inches). The Station elected to change this value as an enhancement based on the
recommendation of the Chief Examiner.
The original question prior to the enhancement with a starting value of 0.50 inches (instead of
0.45 inches) resulted in a constant rate of change (0.075 inches vacuum per 1 minute). The
Reactor (Rx) Building D/P timeline was as follows:
- at time 05:04 Rx Building D/P will be 0.275 inches vacuum
- at time 05:05 Rx Building D/P will be 0.20 inches vacuum
- at time 05:06 Rx Building D/P will be 0.125 inches vacuum
- at time 05:07 Rx Building D/P will be 0.05 inches vacuum
Modification to the starting value resulted in a constant rate of change (0.05 inches vacuum per
minute). The Rx Building D/P timeline is as follows:
- at time 05:04 Rx Building D/P will be 0.30 inches vacuum
- at time 05:05 Rx Building D/P will be 0.25 inches vacuum
- at time 05:06 Rx Building D/P will be 0.20 inches vacuum
- at time 05:07 Rx Building D/P will be 0.15 inches vacuum
Based upon the values provided and the constant rate of change, the earliest of the times
provided that the Tech Spec limit would be exceeded is time 05:06. Time 05:05 would result in
the differential pressure being at the Tech Spec limit (>0.25 inches). This indicates that C is
the correct answer. The question was not updated to reflect the correct answer following the
change to the starting value.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
NRC Resolution:
Upon review of the question, the applicant comment, and the facility proposed resolution, it was
decided that the correct answer should be changed from distractor A to distractor
- C.
The question asks: Assuming a constant rate of change, what is the EARLIEST time that
Secondary Containment will exceed the Tech Spec limit AND what is the Basis for this LCO?
The only difference between distractor A and distractor C is in the EARLIEST time that
Secondary Containment will exceed the Tech Spec limit. The Technical Specification limit for
the Secondary Containment vacuum is specified in Technical Specification Surveillance
Requirement (SR) 3.6.4.1.1, which requires a Secondary Containment vacuum of greater than
or equal to 0.25 inches of vacuum water gauge.
The conditions stated in the question stem are:
- At time 05:01 Rx Building D/P was 0.45 inches vacuum, when the Rx Building Exhaust Fan
tripped.
- At time 05:03 Rx Building D/P was 0.35 inches vacuum and becoming less negative.
- At time 05:04 Annunciator 923-5 C-1 RX BLDG DP LO came in.
The Annunciator 923-5 C-1 RX BLDG DP LO alarms at 0.3 inches vacuum water gauge per the
alarm response procedure. The lowering in Secondary Containment vacuum is thus given by:
- 05:01 0.45 inches vacuum water gauge
- 05:03 0.35 inches vacuum water gauge
- 05:04 0.30 inches vacuum water gauge
The Secondary Containment vacuum is thus lowering at a constant rate of 0.05 inches vacuum
water gauge per minute. Assuming a constant rate of change results in the Secondary
Containment vacuum being the following at the given times below:
- 05:05 0.25 inches vacuum water gauge
- 05:06 0.20 inches vacuum water gauge
Since the question asked what was the EARLIEST time that Secondary Containment will
exceed the Tech Spec limit, the earliest of the times provided such that the Tech Spec limit
would be exceeded was time 05:06 when the differential pressure would be 0.20 inches. At
time 05:05 the differential pressure would be 0.25 inches and thus would be at the Tech Spec
limit of greater than or equal to 0.25 inches of vacuum water gauge.
Therefore, the answer key was modified to change the correct answer from distractor A to
distractor C.
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Question Number 90
A LOCA has occurred on Unit 2, concurrently with a LOOP, with the following conditions:
- Reactor Pressure is 300 psig and LOWERING.
- HPCI and SBLC are the ONLY high pressure systems available AND injecting into the
Reactor.
- RPV water level is -193" and LOWERING.
- BOTH loops of Torus Sprays are in operation.
- BOTH loops of Torus Cooling are in operation.
- Drywell sprays are NOT in operation due to valve binding on both loops.
- Drywell Pressure is 19 psig and RISING.
- Torus Bottom Pressure is 24 psig and RISING.
- Torus Level is 14 feet and STABLE.
Complete the following statements.
The SRO is required to direct the NSO to (1) and blowdown is required based upon
(2) .
A. (1) CONTINUE to operate Torus Cooling AND Torus Sprays;
(2) Torus Bottom Pressure ONLY
B. (1) CONTINUE to operate Torus Cooling AND Torus Sprays;
(2) Reactor Water Level AND Torus Bottom Pressure
C. (1) STOP Torus Cooling AND Torus Sprays;
(2) Reactor Water Level ONLY
D. (1) STOP Torus Cooling AND Torus Sprays;
(2) Reactor Water Level AND Torus Bottom Pressure
Answer: C
Enclosure 3
POST EXAMINATION COMMENTS AND RESOLUTIONS
Applicant Comment:
An applicant commented that answer D should also be accepted as correct. The applicant
provided the following comments:
Both C and D answers are correct. See Facility Post-Examination Comments for the Dresden
Initial Exam - April 2007. Torus sprays and torus cooling are required to be secured at low
pressure when below Top Active Fuel. Blowdown required per contingency leg of DEOP 100 at
level -164. With torus bottom pressure rising and no sprays available Figure L of
DEOP 200-01 will be violated. Per Emergency Procedure Guidelines (EPG) definition on page
B-3-3 of Appendix B, if parameter cannot be maintained then the appropriate actions can be
taken.
Facility Proposed Resolution:
The facility agreed with the applicant and commented that the answer key should be changed to
accept both distractors C and D as correct answers. The facility provided the following
comments:
The author inappropriately allowed this question to be placed on the examination without
revisions following NRC acceptance of previous challenge in 2007. Basis for this challenge is
documented in Facility Post-Examination Comments for the Dresden Initial Exam - April 2007.
Facility Post-Examination Comments for the Dresden Initial Exam - April 2007:
With Figure L, Pressure Suppression Pressure just below the limit (shaded area), and
indication in the question stem, that Torus Bottom Pressure is RISING (and with the Drywell
Sprays inoperable, the ability to recover is gone), the SRO is allowed to make the determination
that the limit will ultimately be exceeded, as allowed in the Dresden Emergency Operating
Procedure (DEOP) Bases document EPG/SAG, page B-3-3. This allows the SRO to blowdown
based on not being able to stay inside Figure L. The SRO also may blowdown based on the
requirements of the level leg of DEOP 100, RPV Control.
Subsequently, the Torus Cooling and Torus Sprays are to be stopped. DEOP 100 states when
RPV water level drops to -143" and no subsystem (Detail F) is lined up with a pump running, to
maximize injection with an alternate injection system (Detail E - Low Pressure Coolant Injection
with Condensate Storage Tank suction).
Based on the flexibility referred to in DEOP 0010-00 discussion section, two different Operators
are allowed to make a different decision based on analysis of the current situation, and both
would be correct.
Therefore, distractors C and D are both correct since, based on the SROs determination on
whether or not Figure L, Pressure Suppression Pressure will be exceeded, the Operating team
can blowdown on both RPV water level and Torus Bottom Pressure.
Enclosure 3
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 the facilitys comment and accept both answer C and D as correct
answers.
The only difference between distractor C and distractor D is in what parameters the
requirement to blowdown the Reactor Pressure Vessel (RPV) is based on.
For the condition of RPV water level at -193" and lowering as stated in the question stem,
distractor C is a correct answer, since RPV blowdown is required in DEOP 100 when RPV
level drops to -164" (i.e., blowdown is required based upon Reactor Water Level ONLY). Per
other conditions stated in the question stem, torus pressure does not exceed the Pressure
Suppression Pressure limit in Figure L, Pressure Suppression Pressure, and so RPV
blowdown per the Primary Containment Pressure leg of DEOP 200-1, Primary Containment
Control is not required.
However, for the conditions stated in the question stem (i.e., torus level 14 feet and stable, torus
bottom pressure at 24 psig and rising, and drywell sprays not in operation due to valve binding
on both loops), one would be just below the Pressure Suppression Pressure limit in Figure L,
Pressure Suppression Pressure. Since torus bottom pressure was at 24 psig and rising, and
drywell sprays were not in operation due to valve binding on both loops, one could reasonably
assume that the Pressure Suppression Pressure limit in Figure L would ultimately be exceeded.
On page B-3-3 of Appendix B of the BWROG EPGs/SAGs (the DEOP Bases document), it
states that if a parameter can not be maintained below a specified limit, then the appropriate
action can be taken if it is anticipated that the limit will ultimately be exceeded. Based on this
flexibility, distractor D is also a correct answer (i.e., blowdown is required based upon Reactor
Water Level and Torus Bottom Pressure), if the SRO determines based on parameter values
and trends that the Pressure Suppression Pressure limit in Figure L would ultimately be
exceeded.
Therefore, the answer key was modified to accept both C and D as correct answers.
As identified by the facility, the examination author inappropriately allowed this question to be
placed on this examination without revisions following NRC acceptance of two correct answers
C and D for the identical question that was on the Dresden 2007 examination. The issue of
not incorporating comments on this question from the previous examination was entered into
the licensees corrective action program (IR 00897239).
Enclosure 3
WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)
RO/SRO Initial Examination ADAMS Accession #ML090890525
Enclosure 4