ML082180007
| ML082180007 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/05/2008 |
| From: | Widmann M Division of Reactor Safety II |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR-08-301 | |
| Download: ML082180007 (22) | |
See also: IR 05000338/2008301
Text
August 5, 2008
Mr. David A. Christian
President and Chief Nuclear Officer
Virginia Electric and Power Company
Innsbrook Technical Center
5000 Dominion Boulevard
Glen Allen, VA 23060
SUBJECT: NORTH ANNA POWER STATION - NRC EXAMINATION REPORT
05000338/2008301 AND 05000339/2008301
Dear Mr. Christian:
During the period of June 2-20, 2008, the Nuclear Regulatory Commission (NRC) administered
operating examinations to employees of your company who had applied for licenses to operate
the North Anna Power Station. At the conclusion of the examination, the examiners discussed
the examination questions and preliminary findings with those members of your staff identified in
the enclosed report. The written examination was administered by your staff on June 24, 2008.
Two Senior Reactor Operator (SRO) applicants and three Reactor Operator applicants passed
both the written and operating examinations. Three SRO applicants and five RO applicants
failed the written examination. There were eleven post examination comments. These
comments and the NRC resolution of these comments are summarized in Enclosure 2. A
Simulation Facility Report is included in this report as Enclosure 3.
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 (PARS) component of NRC's document
system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room). Should you have any questions
concerning this letter, please contact me at (404) 562-4550.
Sincerely,
/RA/
Malcolm T. Widmann, Chief
Operations Branch
Division of Reactor Safety
Docket Nos.: 50-338, 50-339
Enclosures:
1. Report Details
2. NRC Post Examination Comment Resolution
3. Simulation Facility Report
(cc: w/encl - See page 2)
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
SAM NUNN ATLANTA FEDERAL CENTER
61 FORSYTH STREET, SW, SUITE 23T85
ATLANTA, GEORGIA 30303-8931
2
cc w/encl:
Chris L. Funderburk
Director
Nuclear Licensing and Operations Support
Virginia Electric and Power Company
Electronic Mail Distribution
D. G. Stoddard
Site Vice President
North Anna Power Station
Electronic Mail Distribution
Executive Vice President
Old Dominion Electric Cooperative
Electronic Mail Distribution
County Administrator
Louisa County
P.O. Box 160
Louisa, VA 23093
Lillian M. Cuoco, Esq.
Senior Counsel
Dominion Resources Services, Inc.
Electronic Mail Distribution
Attorney General
Supreme Court Building
900 East Main Street
Richmond, VA 23219
Senior Resident Inspector
North Anna Power Station
U.S. NRC
P.O. Box 490
Mineral, VA 23117
Eugene S. Grecheck
Vice President - Nuclear Development
Dominion Resources Services, Inc.
Electronic Mail Distribution
Leslie N. Hartz
Vice President - Nuclear Support Services
Dominion Resources Services, Inc.
5000 Dominion Boulevard
Glen Allen, VA 23061
Eric Hendrixson
Director, Nuclear Safety and Licensing
Virginia Electric and Power Company
Electronic Mail Distribution
Michael M. Cline
Director
Virginia Department of Emergency Services
Management
Electronic Mail Distribution
Mr. Christopher A. McClain
Manager of Nuclear Training
North Anna Power Station
P.O. Box 402
Mineral, VA 23117
3
Letter to David A. Christian from Malcolm T. Widmann dated August 5, 2008
SUBJECT:
NORTH ANNA POWER STATION - NRC EXAMINATION REPORT
05000338/2008301 AND 05000339/2008301
Distribution w/encl:
C. Evans, RII EICS (Part 72 Only)
L. Slack, RII EICS (Linda Slack)
OE Mail (email address if applicable)
RIDSNRRDIRS
PUBLIC
S. P. Lingam, NRR (PM: NA, SUR)
_________________________
XG SUNSI REVIEW COMPLETE
OFFICE
RII:DRS
RII:DRP
RII:DRS
RII:DRP
SIGNATURE
/RA/
/RA/
/RA/
/RA/
NAME
B. Caballero
M. Bates
M. Widmann
James Dodson
DATE
08/01/2008
08/04/2008
08/04/2008
08/04/2008
8/ /2008
8/ /2008
8/ /2008
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
Enclosure 1
NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.:
50-338, 50-339
License No.:
Report No.:
05000338/2008301, 05000339/2008301
Licensee:
Virginia Electric and Power Company
Facility:
North Anna Power Station, Units 1 & 2
Location:
1022 Haley Drive
Mineral, VA 23117
Dates:
Operating Test - June 2-20, 2008
Written Examination - June 24, 2008
Examiners:
M. Bates, Chief Examiner, Operations Engineer
E. Lea, Senior Operations Examiner
B. Caballero, Chief-under-instruction, Operations Engineer
M. Riches, Operations Engineer Trainee
Approved by:
Malcolm T. Widmann, Chief
Operations Branch
Division of Reactor Safety
Enclosure 1
SUMMARY OF FINDINGS
ER 05000338/2008301, 05000339/2008301, 06/02-20/2008 and 06/24/2008; North Anna Power
Station; Licensed Operator Examinations.
The NRC examiners conducted operator licensing initial examinations in accordance with the
guidance in NUREG-1021, Revision 9, Operator Licensing Examination Standards for Power
Reactors. This examination implemented the operator licensing requirements of 10 CFR
§55.41, §55.43, and §55.45.
The NRC administered the operating tests during the period of June 2-20, 2008. Members of
the North Anna Power Station training staff administered the written examination on June 24,
2008. The written examination outline was developed by the NRC. The written exam, operating
test outlines and operating test details were developed by the North Anna Power Station training
staff.
Two Senior Reactor Operator (SRO) applicants and three Reactor Operator applicants passed
both the written and operating examinations. Three SRO applicants and five RO applicants
failed the written examination. Two SRO and three RO applicants were issued operating
licenses.
There were eleven post examination comments.
No findings of significance were identified.
Enclosure 1
REPORT DETAILS
4.
OTHER ACTIVITIES
4OA5 Operator Licensing Initial Examinations
a. Inspection Scope
The North Anna Power Station training staff developed the written exam and operating
test. NRC regional examiners reviewed the proposed examination material to determine
whether it was developed in accordance with NUREG-1021, Operator Licensing
Examination Standards for Power Reactors, Revision 9, Supplement 1. Examination
changes agreed upon between the NRC and the licensee were made according to
NUREG-1021 and incorporated into the final version of the examination materials.
The examiners reviewed the licensees examination security measures while preparing
and administering the examinations to ensure examination security and integrity
complied with 10 CFR 55.49, Integrity of Examinations and Tests.
The examiners evaluated five SRO applicants and eight RO applicants who were being
assessed under the guidelines specified in NUREG-1021. The examiners administered
the operating tests during the period of June 2-20, 2008. Members of the North Anna
Power Station training staff administered the written examination on June 24, 2008. The
evaluations of the applicants and review of documentation were performed to determine
if the applicants, who applied for licenses to operate the North Anna Power Station, met
requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings
The NRC determined that the details provided by the licensee for the written exam,
walkthrough, and simulator tests were within the range of acceptability expected for a
proposed examination.
Two Senior Reactor Operator (SRO) applicants and three Reactor Operator applicants
passed both the written and operating examinations. Three SRO applicants and five RO
applicants failed the written examination.
The final RO and SRO written examinations with knowledge and abilities (K/As) question
references/answers and examination references, and licensees post examination
comments may be accessed in the ADAMS system (ADAMS Accession Numbers,
ML082110224, ML082110233 and ML082110241).
Copies of all individual examination reports were sent to the facility Training Manager for
evaluation and determination of appropriate remedial training.
4
Enclosure 1
4OA6 Meetings
Exit Meeting Summary
On June 20, 2008, the examination team discussed generic issues associated with the
operating test with Mr. Sam Hughes, Operations Manager, and members of the North
Anna Power Station staff. The examiners asked the licensee whether any materials
examined during the inspection should be considered proprietary. No proprietary
information was identified.
PARTIAL LIST OF PERSONS CONTACTED
Licensee personnel
E. Hendrixson, Director, Safety & Licensing
S. Hughes, Manager, Operations
J. Leberstien, Technical Consultant, Station Licensing
C. McClain, Manager, Training
J. Scott, Supervisor, Nuclear Training
W. Shura, Supervisor, Nuclear Training
NRC personnel
M. Bates, Operations Engineer
R. Clagg, Resident Inspector
J. Reece, Senior Resident Inspector
M. Riches, Operations Engineer (In-Training)
Enclosure 2
NRC Resolution to the Facility Comments
A complete text of the licensee's post examination comments can be found in ADAMS under
Accession Number ML082110241.
RO QUESTION # 5
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that 1-OP-3.3, Unit Shutdown From Mode 4 to Mode 5, did not
identify a specific lower limit of RCS temperature for when the accumulator discharge isolation
valve breaker was required to be opened. Since the stem question only asked for its required
breaker position for the current plant conditions and did not specify in accordance with 1-OP-
3.3, the licensee contended that the limiting RCS temperature (with respect to the accumulator
discharge isolation valve breakers) was when any RCS cold leg temperature was 280 °F in
accordance with Tech Spec LCO 3.4.12 (Low Temperature Overpressure Protection System).
The licensee contended that when RCS temperature was at 325°F during a shutdown from
Mode 4 to Mode 5, the accumulator discharge isolation valve breakers were allowed to be open
or closed, dependent on cool down rate and other outage activities.
NRC DISCUSSION:
The question asked for the applicant to identify 1) the correct power supply to 1-SI-MOV-1865A
(A Accumulator Discharge Isolation Valve) and 2) its required position when the unit was in
Mode 4, RCS pressure was 720 psig, and RCS temperature was 325°F.
The governing plant procedure (1-OP-3.3, Rev 58, Unit Shutdown From Mode 4 to Mode 5),
step 5.8.2 stated to Verify all RCS hot leg temperatures (Th) are less than 350°F before the
operator was directed to open the accumulator discharge isolation valve breakers. Technical
Specification LCO 3.4.12 (Low Temperature Overpressure Protection System) required that the
accumulators must be isolated and the power removed from the isolation valve operators when
any RCS cold leg temperature was 280 °F. Since the 1-OP-3.3 procedure did not specifically
prohibit these breakers from remaining closed when RCS temperature was 325°F, then the
breaker was allowed to be either open or closed at that point in time during the plant cool down,
i.e., there was no breaker position requirement when RCS temperature was 325°F. Therefore,
there was no correct answer for this question.
NRC RESOLUTION
In accordance with NUREG 1021, Rev. 9, Supplement 1, ES-403, Section D.1.c, Question # 5 is
deleted from the exam.
Enclosure 2
RO QUESTION # 15
LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the exam. The original
post exam comments submitted by the licensee can be viewed in ADAMS under ML number
ML082110241. The licensee contended that:
the question had no discriminatory value because 7 out of 8 ROs missed the question;
determining the status of the General Warning Lights, located in the Solid State Protection
System Logic Cabinets (SSPS), was not a job requirement for reactor operators at North
Anna. Instead, the licensee contended that the licensed operators would dispatch
Instrument Technicians to determine the cause if either of the two Safeguards Trouble
annunciators (1K-G1, SFGDS PROT SYS TR A TROUBLE and/or 1K-G2, SFGDS PROT
SYS TR B TROUBLE) were in an alarm condition;
the targeted K/A for this question was not applicable to North Annas Engineered Safety
Features Actuation System (ESFAS) and SSPS because it is only associated with
Programmable Logic Controller (PLC) based ESFAS/SSPS systems; and
this question was not administered equally for the ROs and SROs because the SRO written
exam question #81 provided an unfair advantage to SRO applicants when answering this
question. The licensee stated that 7 out of 8 ROs missed the question whereas 4 out of 5
SROs answered the question correctly.
NRC DISCUSSION:
The question asked for the applicant to 1) identify whether there was only one Safeguards
Trouble control room annunciator, common for both SSPS trains, versus two separate
annunciators - one for each train, A and B and 2) recognize how a General Warning
condition would affect the red light in the associated trains logic cabinet, i.e., illuminated or
extinguished.
Each licensee comment for this question was addressed below.
NUREG-1021, Rev. 9, Supplement 1, Appendix A (Overview of Generic Exam Concepts)
outlines three principle facets of test validity as 1) content validity, 2) operational validity,
and 3) discrimination validity. Furthermore, the Appendix A states that:
Test items that are so difficult that few (if any) of the examinees are expected to answer
correctly do not discriminate and should not be used on an NRC examination. It is expected
that every examination will contain some test items that all or most of the examinees will
answer correctly or incorrectly. This does not necessarily mean that the test items or the
examination are invalid.
Prior to administration of the examination, the licensee performed reviews and conducted
validation of the entire examination to assure that all test items were 1) related to the job, 2)
addressed an actual or conceivable activity performed on the job, and 3) not too difficult.
Based on these reviews and validation activities, this exam item was not anticipated to be so
difficult that few (if any) of the applicants would be able to answer the item correctly.
Enclosure 2
Furthermore, this test item was technically correct and pertinent to the applicants job, (see
next point below). Consequently, the licensees contention did not substantiate deleting the
item from the exam.
The North Anna Reactor Operator lesson plan (77-A, Revision 2, 05/24/2007) for the
Reactor Protection System included the following reactor operator learning objective for
Topic 3.6 (General Warning Reactor Trip):
U 8966
List the following information as it applies to the general warning reactor trip:
o Conditions that result in a general warning alarm
o Local indications of a general warning alarm
The corresponding Section 3.6.2.1 of this lesson plan described that:
Each trains logic cabinet has lights that indicate the absence or presence of a General
Warning condition, for example, the Train A logic cabinet has a red light which is
normally off, and which will be lit if a General Warning exists on Train A.
The licensees training program is based on a systems approach to training (SAT) and the
training material learning objectives are directly linked to those tasks which are analyzed for
the RO job. The North Anna RO lesson plan included a learning objective and the
corresponding information that is required to correctly answer the test item. Therefore, the
licensees contention (that determining the status of the General Warning lights was not a
part of the RO job duties) did not substantiate deleting this item from the exam.
The NUREG 1122, Rev. 2, Supplement 1 (Knowledge and Abilities Catalog for Nuclear
Power Plant Operators - Pressurized Water Reactors) K/A which this question was targeted
to meet is:
013 Engineered Safety Features Actuation System (ESFAS)
K4.15 Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the
following: Continuous testing (2.6 / 3.2)
The question was applicable at North Anna because the Safeguards Protection System
Trouble alarms and General Warning Status lights were considered design features and/or
interlocks which provide the means for continuous testing of the ESFAS and SSPS.
Furthermore, the question meets the K/A and this was discussed and appropriate changes
agreed on before the exam was administered.
Because the test item was technically correct and also based on the guidance provided in
NUREG 1021, Rev. 9, Supplement 1, ES-403, Section D.1.b, deletion of this test item from
the exam was not warranted.
The stem for Question #81 on the SRO written exam included the following initial plant
condition statement:
1K-G1, SFGDS PROT SYS TR A TROUBLE annunciator was in the alarm condition.
Enclosure 2
The preceding statement (from Question #81) may have unintentionally aided the SRO
applicants to eliminate distractors on question #15) because it provided cues that there were
two separate annunciators - one for each train, A and B (versus only one Safeguards
Trouble control room annunciator, common for both SSPS trains). This unintentional cue
was not identified during the exam review process. However, the statement in the SRO
Question #81 did not provide a cue for the second portion of the question, i.e., recognizing
how a General Warning condition would affect the red light in the associated trains logic
cabinet (illuminated or extinguished). Although the examination analysis results reflected
that the SRO applicants may have received unintended aid in eliminating distractors on this
question, this does not invalidate this question on the RO exam.
NRC RESOLUTION
Question #15 is valid.
RO QUESTION # 20
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that it was possible that the excess letdown system had been
placed in service at the end of the previous cycle and not drained during the outage. In this
case, the system would contain water with virtually zero boron concentration. Consequently,
the result would be a dilution (versus boration) of the RCS when excess letdown was placed in
service.
NRC DISCUSSION:
The initial conditions of the question stated that the unit was at 100% following a refueling
outage. This test item asked the applicant 1) an excess letdown lineup restriction and 2) the
reason why reactor power must be monitored when placing excess letdown in service in
accordance with 1-OP-8.5. The second part of this test item was targeted to Precaution &
Limitation (P&L) 4.7 at the front of 1-OP-8.5 (Rev 18), which states:
4.7
WHEN Excess Letdown is placed in service, THEN monitor RCS temperature and
Reactor power closely due to the possible reactivity effects. A dilution may be required
to maintain desired RCS temperature and Reactor power level. This is due to a
potentially higher boron concentration in the Excess Letdown piping.
Additionally, the instructions for shifting from normal letdown to excess letdown (Section 5.1)
included the following note:
NOTE: When Excess Letdown is placed in service, RCS temperature and Reactor power level
should monitored closely due to the possible reactivity effects. A dilution may be
required to maintain desired RCS temperature and Reactor power level.
Enclosure 2
If the excess letdown system was placed in service at the end of the previous fuel cycle, then
the piping would contain diluted water (versus a high boron concentration). In this case, a
boration would have been required to maintain RCS temperature and reactor power level
constant. Because the P&L and the note both stated that a dilution MAY be required and
because the licensees postulated scenario was operationally credible, then there were two
correct answers for this test item.
NRC RESOLUTION
In accordance with NUREG 1021, Rev. 9, Supplement 1, ES-403, Section D.1.c, question # 20
will be graded with two correct answers. Either A or D is correct.
RO QUESTION # 23
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that technically, there was no functional difference between the
two sequences of establishing Quench Spray (QS), i.e., discharge valves actuation first and
then pump(s) actuation or vice versus.
NRC DISCUSSION:
The initial conditions for this question stated that a safety injection had occurred and
containment pressure was 23 psia and slowly rising. The stem question asked the applicant to
identify the appropriate operator action and the sequence of the action required.
Since the Containment Depressurization Actuation (CDA) auto actuation set point (27.75 psia)
had not yet been achieved, the applicant was expected to eliminate choices B and D
(because they both stated that CDA was required). Additionally, the applicant was expected to
recognize that the QS function was required to be manually actuated in accordance with 1-E-0,
Reactor Trip or Safety Injection, Step 12, because containment pressure had exceeded 20 psia.
Furthermore, the applicant was expected to recall the sequence (discharge valve first, then
pump) delineated in Step 12 for initiating QS.
1-E-0 provided two sequences for verifying or initiating the same pump(s). Step 11 was
applicable when an auto CDA signal failed to occur and directed the operator to start the pump
first and then manually open the discharge valve. Step 12 was applicable when manual
initiation of the QS function (i.e., radio-isotope control afforded by the injection of NaOH from the
Chemical Addition Tank.) was required and directed the operator to first open the discharge
valve and then start the pump. In both cases, the final configuration of pump(s) running with
discharge valve(s) open was achieved within a very short period of time with negligible short-
term consequences. Since there is no documented technical basis difference between the two
approved manual actuation sequences, either sequence was an acceptable response to the
conditions stated in the question stem.
NRC RESOLUTION
In accordance with NUREG 1021, Rev. 9, Supplement 1, ES-403, Section D.1.c, question # 23
will be graded with two correct answers. Either A or C is correct.
Enclosure 2
RO QUESTION # 26
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that the conditions provided in the stem of the question did not
give positive confirmation of the final PORV 1-RC-PCV-1455C valve position and functionality
status. Consequently, the licensee contended that there were two scenarios, each with different
Tech Spec (TS) required action statements.
NRC DISCUSSION:
This question asked the applicants to recognize the one hour action statement requirement for
the following plant sequence of events:
Given the following:
Unit 1 is operating at 100% power
An instrument failure causes PORV 1-RC-PCV-1445C to lift.
RCS Pressure is recovering at a much slower rate than expected.
The operator closes the PORV Block Valve, 1-RC-MOV-1536, and pressure is now
2100 psig and increasing as expected.
All other equipment appears to be operating normally.
The correct answer, (A), was based on TS 3.4.11, Condition B:
One or more PORVs inoperable for reason other than Condition A and capable of being
manually cycled.
The licensee contended that choice D was also correct based on TS 3.4.11, Condition C:
One PORV inoperable and not capable of being manually cycled.
The applicants were expected to interpret the third (3rd) bullet (see above) to mean that 1) RCS
pressure was rising before the Block Valve was manually closed and 2) that the 1-RC-PCV-
1455C PORV did not fully seat following its actuation due to the instrument failure. The licensee
contended that the wording for the third (3rd) bullet (see above) implied that the PORV could
have been partially or fully open (vs. only leaking by at the valve seat). The design of the PORV
is such that it is never throttled, only full open or full closed. If the PORV had remained open
following its lift, then the pressurizer heater capacity cannot cause RCS pressure to rise.
The licensee also contended that by not providing a final valve position for 1-RC-PCV-1455C,
that this validated an assumption that the PORV was not capable of being manually cycled.
During the examination administration, one applicant asked the proctor whether 1-RC-PCV-
1455C indicated shut. After consultation with the chief examiner, the licensee told the applicant
to answer the question with the information given, i.e., no further clarification was provided to
the applicant.
Enclosure 2
The following excerpt from NUREG 1021, Rev. 9, Supplement 1, Appendix E, (Policies and
Guidelines for Taking NRC Examinations), Part B: Written Examination Guidelines, was read
verbatim (by the chief examiner) to all the applicants before the exam was administered.
When answering a question, do not make assumptions regarding conditions that are not
specified in the question unless they occur as a consequence of other conditions that
are stated in the question.
The licensees contention that choice D was also correct required an assumption (i.e., the
PORV was not capable of being manually cycled) that was not a consequence of the conditions
stated in the question.
NRC RESOLUTION
For question # 26, the only correct answer is A.
RO QUESTION # 28
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that the applicants answered the question based on their
knowledge of the background document discussions associated with FR-S.1, Response to
Nuclear Power Generation/ATWS; therefore, automatic control rod insertion (C) was also a
correct choice.
NRC DISCUSSION:
This question asked the applicants to identify the action that would insert the most negative
reactivity within the first 30 seconds following a 100% power ATWS at the end of core life. The
choices provided were (answer key correct answer is B):
A. Initiation of Emergency Boration
B. Manual Turbine Trip
C. Automatic Control Rod Insertion
D. Manual Control Rod Insertion
Following a turbine trip, the moderator temperature coefficient (Tmod) reduces the core power as
the coolant temperature rises. Additionally, Tmod becomes more negative at the end of core life.
The reactivity effect of the turbine trip occurs at a much greater rate than the reactivity effect
from automatic (or manual) rod insertion; especially within the first 30 seconds following a 100%
power ATWS.
The licensee admitted in their post-exam comments that the correct answer from a transient
analysis perspective was that the manual turbine trip added the most negative reactivity (vs. the
automatic or manual rod insertion). However, the licensees concern was that some applicants
based their selection with the mindset that the question was asking for what the background
document contained in its step discussions for FR-S.1, Step [2] (Verify Turbine Trip - Immediate
Operator Action) and Step 4 (Initiate Emergency Boration of the RCS). The licensee referred to
the following portions of the background documents discussion for Steps [2] and 4,
respectively:
Enclosure 2
Step [2] discussion:
The turbine is tripped to prevent an uncontrolled cool down of the RCS due to steam
flow that the turbine would require. For an ATWS even where a loss of normal
Feedwater has occurred, analyses have shown that a turbine trip is necessary (within 30
seconds) to maintain SG inventory.
Step 4 discussion:
After control rod trip and insertion functions, boration is the next most direct manner of
adding negative reactivity to the core.
The licensee contended that based only on these discussions provided in the FR-S.1
background document, the only information available stated that the automatic function of the
control rods was the most direct manner of adding negative reactivity to the core (vs. manual
rod insertion and emergency boration).
The question was soliciting the fundamental knowledge associated with the operational
implications of the negative temperature coefficient as it applies to large PWR systems. The
question was not soliciting the detailed knowledge of the licensees background document
associated with FR-S.1. Even so, the background document information associated with FR-
S.1 did not conflict with the transient analysis perspective that the manual turbine trip added
significantly more negative reactivity (vs. the automatic and/or manual rod insertion and
emergency boration) within the first 30 seconds after the ATWS. Furthermore, the first action in
FR-S.1 was to verify the reactor is tripped, which did not occur as stated in the stem. The
second action was to verify that the turbine is tripped. The third action in the procedure was to
verify rods inserting in AUTO at > 48 steps/minute. The FR-S.1 procedure sequence did not
conflict with the transient analysis perspective that the manual turbine trip added significantly
more negative reactivity within the first 30 seconds after the ATWS.
NRC RESOLUTION
For question # 28, the only correct answer is B.
RO QUESTION # 43
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that the status of the SAFE/RESET green light was irrelevant and
inconsequential during the plant conditions provided in the question. The licensees comment
was that the question required a level of recall that was beyond what is required to identify and
verify proper response of the Containment High Range Radiation Monitors when an actual high
radiation condition exists. Consequently, the licensees comment was that both answers, i.e.,
LIT and NOT LIT, were correct.
Enclosure 2
NRC DISCUSSION:
The initial conditions provided in the question were that a LOCA had occurred on 30 minutes
ago on Unit and both of the Containment High Range Radiation Monitors (1-RM-RMS-165 and
1-RM-RMS-166) had amber and red lights illuminated in the control room. The question asked
the applicants to recognize 1) whether or not the radiation monitors SAFE/RESET green light
was lit and 2) the location and status of the Unit 1 CONT HI RANGE RADIATION TROUBLE
alarm.
The licensee contended that the level of recall required by the applicants to ascertain the status
of the SAFE/RESET green light for the given plant conditions was both irrelevant and too
difficult. Consequently, the licensee contended that both answers, LIT and NOT LIT, were
acceptable and correct.
The reactor operator lesson plan, STUDENT GUIDE FOR RADIATION MONITORING SYSTEM
(46), (Revision 3, 09/19/2007) included the following learning objectives:
6.3 Objective
U 5250
List the means provided in the control room to determine the following abnormal
conditions as they apply to the CHRRMS (Victoreen) containment radiation monitors.
- Loss of detector and signal cable integrity
- High containment radiation
6.4 Objective
U 5251
Explain how the detector and signal cable integrity of the CHRRMS (Victoreen)
containment radiation monitors is normally monitored.
The licensees training program is based on a systems approach to training (SAT) and the
training material learning objectives are directly linked to those tasks which are analyzed for the
RO job. The North Anna RO lesson plan included learning objectives and the corresponding
information that was required to correctly answer the test item.
According to the lesson plan, the SAFE/RESET pushbutton/light had two important functions:
1) It was energized (green) when all conditions were normal and the system was
operating normally. If a loss of detector or signal cable integrity for the Containment
High Range Radiation Monitors (CHRRMS) occurred, the green SAFE/RESET light
de-energized and the audible alarm annunciated on Unit-2. [Note: One common
annunciator alarm per unit (UNIT 1 CONTAINMENT HI RANGE RADIATION
TROUBLE, Window 2A-B3, and UNIT 2 CONTAINMENT HI RANGE RADIATION
TROUBLE, Window 2A-C3) actuated for alert, high radiation, and failure. Both alarms
were located on the Unit 2 annunciator panel 2-EI-CB-3.]
2) The SAFE/RESET pushbutton/light was used for acknowledging the ALERT (amber)
and HIGH (red) alarms on the CHRRMS. The lights were acknowledged and reset
by depressing the SAFE/RESET button.
Enclosure 2
The SAFE/RESET light was important because it alerted the operator that a problem had
occurred with the detector or signal cable. This was important information both during normal
operating conditions and during the conditions provided in this test item, i.e., during a LOCA.
For the conditions provided in the stem of the question, the SAFE/RESET light was LIT, i.e.,
energized, because there was not a problem with the detector or signal cable. Therefore, the
licensees contention (that ascertaining the status of this light during high radiation conditions is
irrelevant and/or required too much recall) did not substantiate accepting two answers for this
item.
NRC RESOLUTION
For question # 43, the only correct answer is A.
RO QUESTION # 50
LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the examination. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that when placing the Charcoal Filter Inlet From Decay Tanks
Controller, 1-GW-FCV-101, in service; the operator was directed by procedure to adjust the
controller in MANUAL first before placing the controller in AUTOMATIC. The licensees
comment was that there was no correct answer for this question.
NRC DISCUSSION:
This question required the applicant to identify the preferred method for controlling the Waste
Gas Decay Tank (WGDT) release flow rate in accordance with 0-OP-23.2, WGDT and Waste
Gas Diaphragm Compressors.
0-OP-23.2, Precaution & Limitation 4.13 and the Note preceding step 5.4.11 both stated the
following:
It is preferred to operate 1-GW-FCV-101 in AUTO control to maintain flow < 3 SCFM.
IF Manual control is required, THEN it is to only be done with SRO permission.
(Reference 2.4.3)
The procedure specifically prohibited controlling the release in MANUAL without SRO
permission. When the controller was in AUTOMATIC, the flow control valve would respond to
pressure changes in the WGDT to maintain flow < 3 SCFM. The stem question specifically
asked the applicant to identify the preferred method for controlling the release in accordance
with the procedure. The licensees contention was that the first few procedure steps for placing
the controller in service constituted controlling the release. The first few steps of the
procedure were associated with placing the controller in service (vs. controlling); therefore,
these steps were not considered controlling in MANUAL.
NRC RESOLUTION
Question #50 is valid.
Enclosure 2
RO QUESTION # 53
LICENSEE COMMENT:
In summary, the licensee requested that this question be deleted from the examination. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that an applicant who correctly implemented 1-AP-19, Loss of
Bearing Cooling Water, (using the initial conditions provided in the stem) would not be directed
to start the standby bearing cooling tower fan or to shift fans to fast speed because these
actions were not included in the procedure. Furthermore, the licensees comment was that
since the annunciator response (ARs) procedures were normally performed in parallel with the
1-AP-16, then the question had three potentially correct answers because three of the choices
were specifically directed by the AR procedures.
NRC DISCUSSION:
The question provides the initial conditions that the crew had entered 1-AP-19 following receipt
of two control room alarms, 1T-C1 and 1A-F4. The question states:
Which ONE of the following describes the initial action required in accordance with 1-
AP-19?
Choice C was identified as correct on the answer key, i.e., Start available Bearing Cooling
Tower Fans or shift fans to high speed. However, this action was directed by the 1A-F4 AR
procedure (BASIN TEMP HI/LOW) versus being directed by 1-AP-19. Choices A and D were
both actions directed by the 1T-C1 annunciator procedure (HYDROGEN TEMP OR CORE
MONITOR). Choice B was a subsequent action directed by 1-AP-19, and thus was not
considered an initial action. Therefore, there was no correct answer for this test item due to the
wording of the question, i.e., ..in accordance with 1-AP-19.
NRC RESOLUTION
In accordance with NUREG 1021, Rev. 9, Supplement 1, ES-403, Section D.1.c, question #53
is deleted from the exam.
SRO QUESTION # 92
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensees comment was that under normal conditions, the Service Water (SW) system is
operated with the Closed Cooling (CC) Heat Exchanger Outlet Valves throttled. Furthermore,
the licensees comment was that when removing a SW Pump from service for planned or
emergent maintenance, the associated SW Pump Maintenance Operating Procedure (MOP),
used to remove the pump from service, required the establishment or verification of proper SW
System throttling.
Enclosure 2
NRC DISCUSSION:
The initial conditions provided in the question were that both units were at 100% power with two
SW pumps running (1-SW-P-1A and 1-SW-P-1B) and 2-SW-P-1B was out of service.
Subsequently, the 1-SW-P-1B SW Pump tripped and the 2-SW-P-1A was started. The question
asked the applicants to identify the TS action required and also the TS bases for the action.
The configuration for both units operating at 100% power was one pump running per loop, with
the following normal pump alignment:
1-SW-P-1A aligned to the A loop (running)
1-SW-P-1B aligned to the B loop (initially running but subsequently tripped)
2-SW-P-1A aligned to the B loop
2-SW-P-1B aligned to the A loop (out of service)
Given the initial conditions provided in the question, i.e., both units at 100% power with the 2-
SW-P-1B SW Pump out of service, the CC Heat Exchanger outlet valves were required to be
throttled for two reasons:
1. Limiting Condition for Operation (LCO) 3.7.8 required two SW System loops to be
operable in Modes 1, 2, 3, and 4. The LCO 3.7.8 bases stated:
A SW loop is considered OPERABLE during MODES 1, 2, 3, and 4 when:
a. Either
a.1 Two SW pumps are OPERABLE in an OPERABLE flow path;
or
a.2 One SW pump is OPERABLE in an OPERABLE flow path
provided two SW pumps are OPERABLE in the other loop
and SW flow to the CC heat exchangers is throttled
Because 2-SW-P-1B was initially out of service (inoperable), the a.1 statement was
not satisfied and the licensees LCO compliance was based on satisfying the a.2
statement.
2. The Maintenance Operating Procedure (MOP) 2-MOP-49.02 used to remove the 2-
SW-P-1B SW Pump from service required that the SW outlet of the CC heat
exchangers so that each SW Pump discharge pressure is > 54 psig using 0-OP-
49.6, Service Water System Throttling Alignment.
The following excerpt from NUREG 1021, Rev. 9, Supplement 1, Appendix E, (Policies and
Guidelines for Taking NRC Examinations), Part B: Written Examination Guidelines, applied:
When answering a question, do not make assumptions regarding conditions that
are not specified in the question unless they occur as a consequence of other conditions
that are stated in the question.
Enclosure 2
Given the initial conditions provided in the question, i.e., both units at 100% power with the 2-
SW-P-1B SW Pump out of service, the applicant was required to assume that the CC Heat
Exchanger outlet valves were throttled since this condition occurred as a consequence of the
initial conditions stated in the question.
With the CC Heat Exchanger outlet valves throttled, the applicant was expected to determine
the TS action required following the trip of another SW pump, i.e., 1-SW-P-1B. TS 3.7.8,
Condition B was applicable when two SW pumps were inoperable. This required action was:
B.1 Throttle SW System flow to CC heat exchangers within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
AND
B.2 Restore one SW pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
Because the CC Heat Exchanger outlet valves were throttled, compliance with B.1 had
previously been established. Therefore, the TS action required following the trip of the 1-SW-P-
1B pump was B.2, i.e., Restore one SW pump to operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This eliminated
choices B and D.
The second part of each choice required the applicant to recognize the basis for the required TS
action. According to the TS 3.7.8 bases for action B.2, with only two operable SW Pumps, the
safety function of providing design SW flow to the Recirc Spray (RS) Heat Exchangers following
a LOCA was still met assuming NO additional failures. The bases stated that restoring one SW
pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> together with the throttling ensured that design flow to
the RS Heat Exchangers was achieved following an accident. In this configuration, a single
failure disabling a SW pump would not result in loss of the SW system function. Consequently,
the correct answer is A.
NRC RESOLUTION
For question # 92, the only correct answer is A.
SRO QUESTION # 96
LICENSEE COMMENT:
In summary, the licensee requested that this question be graded with two correct answers. The
original post exam comments submitted by the licensee can be viewed in ADAMS under ML
number ML082110241.
The licensee commented that the governing procedure (VPAP-1403, Temporary Modifications)
required that temporary modification jumper packages received final approval by the FSRC
whereas the actual jumper installation (in the plant) received final approval by the Shift
Manager. The licensee contended that the stem question was not specific with respect to
asking who provided final approval of the temporary modification jumper package versus asking
who provided the final approval for the jumper to be installed in the plant. Consequently, the
licensee contended that both choices (i.e., Shift Manager and FSRC) were acceptable and
correct.
Enclosure 2
NRC DISCUSSION:
The question asked the applicant to identify 1) the governing procedure for installation of a
jumper on a multiple input annunciator (1D-H5, HIGH CAPACITY S/G BLOWDOWN TROUBLE)
and 2) who, by title, must provide final approval of the jumper. [This was a multiple input
annunciator which required an interim jumper installation to restore functionality of the alarm.]
OP-AA-100, Conduct of Operations, Rev 0, Attachment 2 (Shift Operations) stated that In the
case of a multiple-input annunciator, for example, implementation of a temporary modification
may be considered to restore functionality to unaffected circuitry. The stem of the question
specifically stated that a jumper was required to restore functionality of the alarm.
Consequently, the governing procedure for this jumper installation at North Anna was VPAP-
1403, Temporary Modifications (versus OP-NA-200-1001, Equipment Clearance Process).
The jumper approval and installation process (in accordance with VPAP-1403, Revision 11) was
as follows:
Originator forwards the temporary mod package to the Shift Technical Advisor (Step 6.4.12)
Shift Technical Advisor reviews the package and forwards to the Shift Supervisor (Step 6.5.3)
Shift Supervisor approves the package (Step 6.6.10)
Shift Supervisor obtains the Manager of Nuclear Operations or Operations Manager on call approval
(Step 6.6.11)
Shift Supervisor forwards the package to the Originator (Step 6.6.12)
Originator obtains SNSOC Chairmans signature for SNSOCs approval of the package (Step 6.7.2)
Originator forwards the package to the Temporary Modification Installer (Step 6.7.3)
Installer obtains the Shift Supervisors approval for installation (Step 6.8.1.a)
Installer installs the approved modification (Step 6.8.1.c)
The stem question asked Which procedure governs the installation of the jumper, and who, by
title, must provide FINAL approval of the jumper? The question stem referred to the jumper
installation (versus jumper package) and also asked for the very last approval. In accordance
with VPAP-1403, Temporary Modifications, Rev 11, Step 6.8.1.a, the installer must get the Shift
Supervisors approval. This was the FINAL approval of the jumper in accordance with the
process described in the procedure.
The licensee provided documentation that the Job titles for Shift Supervisor/Nuclear Shift
Supervisor and Assistant Shift Supervisor/Nuclear Assistant Shift Supervisor at the North Anna
and Surry Power Stations were changed to Shift Manager and Unit Supervisor, respectively.
Although the change was effective June 1, 2003, dual titles were maintained until regulatory
relevant documents were revised and until the old titles were removed from Station/ISFSI
Technical Specifications, Quality Assurance Topical Report, Emergency Plan, Safety Analysis
Report, and Station/Corporate procedures, programs and standards when the need arises for a
document revision.
NRC RESOLUTION
For question # 96, the only correct answer is B.
Enclosure 3
SIMULATION FACILITY REPORT
Facility Licensee: North Anna Power Station
Facility Docket Nos.: 05000338/05000339
Operating Tests Administered on: June 2-20, 2008
This form is to be used only to report observations. These observations do not constitute audit
or inspection findings and, without further verification and review in accordance with IP
71111.11, are not indicative of noncompliance with 10 CFR 55.46. No licensee action is
required in response to these observations.
No simulator fidelity or configuration items were identified.