IR 05000271/2005301
| ML051030388 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/13/2005 |
| From: | Conte R NRC/RGN-I/DRS/OSB |
| To: | Thayer J Entergy Nuclear Operations |
| Shared Package | |
| ML042440154 | List: |
| References | |
| IR-05-301 | |
| Download: ML051030388 (16) | |
Text
April 13, 2005
SUBJECT:
VERMONT YANKEE - SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT NO. 05000271/2005301
Dear Mr. Thayer:
This report transmits the results of the Senior Reactor Operator (SRO) licensing examination conducted by the NRC during the period of January 31-February 3, 2005. This examination addressed areas important to public health and safety and was developed and administered using the guidelines of the Examination Standards for Power Reactors (NUREG-1021, Revision 9).
Based on the results of the examination, five (out of five) Senior Reactor Operator applicants passed all portions of the examination. The five applicants included four upgrade SROs and one instant SRO. The results indicated that the applicants were generally well prepared for the examination. Mr. John G. Caruso discussed initial performance insights observed during the examination with Mr. Mike Gosekamp of your staff on February 2, 2005. On March 16, 2005, final examination results, including individual license numbers, were given during a telephone call between Mr. John G. Caruso and Mr. Mike Gosekamp of your staff.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). These records include the final examination and are available in ADAMS (Master File - Accession Number ML042440154; RO and SRO Written - Accession Number ML050470032; SRO Operating Section A - Accession Number ML050470039; SRO Operating Section B - Accession Number ML050470041; and SRO Operating Section C - Accession Number ML050470044), ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
M Should you have any questions regarding this examination, please contact me at (610) 337-5183 or by E-mail at RJC@NRC.GOV.
Sincerely,
/RA/
Richard J. Conte, Chief Operational Safety Branch Division of Reactor Safety Docket No.
50-271 License No.
DPR-28 Enclosure:
Initial Examination Report No. 05000271/2005301 cc w/encl:
M. R. Kansler, President, Entergy Nuclear Operations, Inc.
G. J. Taylor, Chief Executive Officer, Entergy Operations J. T. Herron, Senior Vice President and Chief Operating Officer D. L. Pace, Vice President, Engineering B. OGrady, Vice President, Operations Support J. M. DeVincentis, Manager, Licensing, Vermont Yankee Nuclear Power Station Operating Experience Coordinator - Vermont Yankee Nuclear Power Station J. F. McCann, Director, Nuclear Safety Assurance M. J. Colomb, Director of Oversight, Entergy Nuclear Operations, Inc.
J. M. Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc.
S. Lousteau, Treasury Department, Entergy Services, Inc.
Administrator, Bureau of Radiological Health, State of New Hampshire Chief, Safety Unit, Office of the Attorney General, Commonwealth of Mass.
D. R. Lewis, Esquire, Shaw, Pittman, Potts & Trowbridge G. D. Bisbee, Esquire, Deputy Attorney General, Environmental Protection Bureau J. Block, Esquire J. P. Matteau, Executive Director, Windham Regional Commission M. Daley, New England Coalition on Nuclear Pollution, Inc. (NECNP)
D. Katz, Citizens Awareness Network (CAN)
R. Shadis, New England Coalition Staff G. Sachs, President/Staff Person, c/o Stopthesale J. Sniezek, PWR SRC Consultant R. Toole, PWR SRC Consultant Commonwealth of Massachusetts, SLO Designee State of New Hampshire, SLO Designee State of Vermont, SLO Designee M. Desilets, Manager - Training and Development S. Glenn, INPO (GlennSG@Inpo.org)
M
SUMMARY OF FINDINGS
IR 05000271/2005301; January 31 - February 3, 2005; Vermont Yankee; Initial Operator
Licensing Examination.
Five of five applicants passed the examination (four SRO upgrades and one SRO instant).
The written examinations were administered by the facility and the operating tests were administered by two NRC region-based examiners. There were no inspection findings of significance associated with the examinations.
NRC-Identified and Self-Revealing Findings
None
Licensee-Identified Violations
None
REPORT DETAILS
REACTOR SAFETY
Mitigating Systems - Senior Reactor Operator (SRO) Initial License Examination
a. Scope
of Review The licensee developed the written examination and the operating initial examination.
The NRC together with Site training and operations personnel verified or ensured, as applicable, the following:
- The examination was prepared and developed in accordance with the guidelines of Revision 9 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors. A review was conducted both in the Region I office and at the plant and training facility. Final resolution of comments and incorporation of test revisions were conducted during and following the onsite preparation week.
- Simulation facility operation was proper.
- A test item analysis was completed on the written examination for feedback into the systems approach to training programs.
- Examination security requirements were met.
The NRC examiners administered the operating portion of the examination to all applicants from January 31 - February 2, 2005. The written examination was administered by the site training staff on February 3, 2005.
b. Findings
Grading and Results Five applicants (four SRO upgrades and one SRO instant) passed all portions of the initial licensing examination.
The facility had two post-examination comments on the SRO portion of the written examination. The NRC resolutions of these comments are attached. Based on these comment resolutions, the NRC re-graded all of the SRO applicants written examinations. The re-grading of the SRO written examinations resulted in two SRO applicants achieving passing grades, rather than failing grades on the SRO only portions of the written exam as originally determined by the licensee (i.e., on the original exam grading both of these applicants had passed these exams on their overall scores but their sub-scores on the 25 question SRO-only portion were below the required minimum passing score of 70%).
Examination Administration and Performance During the exam administration there were no potential simulator fidelity issues or training deficiencies identified.
OTHER ACTIVITIES
4OA6 Exit Meeting Summary
On March 16, 2005, the NRC provided conclusions and examination results to site management representatives via telephone. License numbers for all five applicants were provided since the licensees letter dated March 14, 2005 provided notification of completion of all program requirements.
The NRC expressed appreciation for the cooperation and assistance that was provided during the preparation and administration of the examination by the licensees training staff.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
ATTACHMENT 1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- M. Gosekamp, Operations Training Supervisor
NRC Personnel
- J. Caruso, Senior Operations Engineer
- S. Dennis, Senior Operations Engineer
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened/Closed/Discussed
None
B-1
ATTACHMENT 2
Licensees Post Written Examination Comments Publically Available
ADAMS Accession No. ML0501010097
Note: The licensees post exam comments regarding these two questions were received by the
NRC on February 14, 2005. After initial review, the NRC provided comments to and requested
additional information from the licensee on February 16, 2005. After further review by the NRC
staff and discussions with the licensee, the licensee submitted revised comments for both SRO
questions #77 and 79, dated and received in Region I on March 10, 2005 (see below). During
the exam there were no questions from the applicants regarding either SRO question #77 or
- 79. The NRCs resolution for these two post exam comments is based on the independent
reviews that were conducted by both NRC examiners assigned to the exam team as well as the
Branch Chief.
Original SRO Question 77:
Select the correct answer:
At approximately 20% power during startup and power ascension, Control Room annunciators
alarmed, including the following:
6-D-1 INST AIR HDR PRESSURE LO
5-E-2 FW VLV LOCKUP SIGNAL/AIR FAIL
The power ascension was immediately halted, and the following conditions have been reported:
! Reactor level is slowly lowering
! Scram Air Header pressure is 70 psig and stable
Which of the following describes the required actions and the reason for those actions?
a. Override SA-PCV-1 closed after making an announcement over the Gai-tronics page.
Service air supplies any respirators in use.
b. Scram the reactor and enter OT 3100. Control rods are expected to drift at this
pressure.
c.
Scram the reactor and enter OT 3100. The in-service FWRV has locked up.
d. Place the aux FWRV in service and restore reactor level. The in-service FWRV has
locked up.
B-2
Submitted answer explanation:
Justification
a. Incorrect - SA-PCV-1 begins to automatically close at 85 psig and should be fully
closed at 80 psig.
b. Incorrect - If the Scram Air Header pressure drops to less than 55 psig, then scram the
reactor and enter OT 3100.
c.
Correct Response - If level is unexpectedly decreasing and the FRVs are locked up
then insert a scram.
d. Incorrect - The aux FWRV will fail closed on loss of air; additionally, it is only 10%
capacity, and there is no procedural direction for this action.
LICENSEES JUSTIFICATION FOR CHANGE
Revised answer explanation:
The original correct answer for this question is C and was justified as correct with the following
statement: If level is unexpectedly decreasing and the FRVs are locked up then insert a
scram. The stem of the question has an alarm (6-D-1) which indicates low air pressure and an
alarm (5-E-2) which indicates a lockup of the Feed Regulating Valves. With these conditions
the plant would enter ON 3146, Low Instrument/Scram Air Header Pressure. Step 2 of this
procedure provides the following direction:
2.
If RPV level is decreasing uncontrollably due to feed reg valve lockup, then perform
the following:
a.
Scram the reactor and enter OT 3100, Scram Procedure.
This question is written at the comprehension level. The candidate must realize that at 20%
power operations, the Feed Regulating Valves are the valves used to control level. If they are
locked up due to a loss of air and level is lowering then it is decreasing uncontrollably. The
actual rate of a level decrease, whether slowly or rapidly, doesnt affect the fact that it cannot be
controlled. After a reactor scram is inserted, other means of level control become available
such as the Emergency Core Cooling Systems (ECCS).
The entry into ON 3146, Low Instrument/Scram Air Header Pressure, and the uncontrollable
level decrease still make C a correct answer.
Choice A was initially justified as wrong with the statement: SA-PCV-1 begins to
automatically close at 85 psig and should be fully closed at 80 psig. The wording of choice A
starts with Override SA-PCV-1 closed which was considered incorrect, the original assumption
being that the valve is designed to automatically close on low pressure and that there was
nothing in the stem to contradict that assumption. However, it was not originally recognized
that the stem statement Scram Air Header pressure is 70 psig and stable could provide
reason to believe that valve SA-PCV-1 had not closed completely. If an air leak was on the
Service Air side and PCV-1 was closed, then Scram Air header air pressure would be expected
B-3
to rise to near its original value, not stabilize at a lower pressure. With Service Air Header
pressure at 70 psig and stable, in combination with other conditions described in the stem, it is
credible to postulate that PCV-1 may not have closed completely. The previously provided
justification did not take into account the fact that the value and trend of Scram Air Header
pressure provided reasonable indications that SA-PCV-1 may not have closed as designed and,
therefore, it may also be appropriate to attempt to mitigate the source of the air leak by using
other, procedurally directed, steps prior to inserting a manual scram.
In ON-3146, Low Instrument/Scram Air Header Pressure, the Caution before Step 7 and Step 7
procedurally direct the operator to perform the action specified in answer A.
CAUTION
Since service air supplies any air respirators in use, an announcement
should be made over the gai-tronics page that the supply will be lost
prior to SA-PCV-1 closing. If conditions permit, allow time for
personnel response before closing the valve manually.
7.
If service air pressure decreases to less than 85 psig, verify air header pressure
control valve SA-PCV-1 closes.
a.
If SA-PCV-1 fails to close when pressure decreases to less than 85 psig,
manually override and close SA-PCV-1.
As previously stated, using information regarding Scram Air Header pressure as an indication of
the operation of SA-PCV-1, it would be appropriate to complete the actions specified in steps 7
and 7.a, thereby making answer A an alternate correct answer.
Conclusion: Answers A and C should be accepted as correct answers. In this case, both
answers are correct since each is a credible, procedurally-directed action from the appropriate
off normal procedure.
B-4
NRC RESOLUTION:
The NRC conducted detailed reviews of all references provided and concluded that the original
correct answer for this question C should still be considered a correct answer. Alarm (6-D-1)
indicates possible instrument air system leakage and refers the operator to ON 3146, Low
Instrument /Scram Air header Pressure. Alarm (5-E-2) together with the given condition of
reactor level is slowly lowering indicates loss of control air to the feed water regulating valves
(FWRVs) and a rupture of air piping which would result in automatic actions causing the
FWRVs to lockup and refers the operator to OT 3113, Reactor Low Level. It is reasonable to
assume, given that reactor level is slowly lowering with the FWRVs locked-up, that reactor level
could be lowering uncontrollably (for that moment in time) since the operator no longer has the
ability to control and restore level. Therefore, answer C is acceptable based on operator
action Step 2, ON 3146, which states, If RPV level is decreasing uncontrollably due to feed
reg. valve lockup, then perform the following: Scram the reactor and enter OT 3100, Scram
Procedure.
It is also reasonable to assume, given that reactor level is slowly decreasing due to the FWRV
lockup, the reason for the level decrease was known and level was not decreasing
uncontrollably at that moment in time. Given the stem statement Scram Air Header pressure is
psig and stable could provide reason to believe that valve SA-PCV-1 had not closed
completely. If an air leak existed on the Service Air side and PCV-1 was then closed, then
Scram Air header air pressure would be expected to rise and would not stabilize at a lower
pressure. With Service Air Header pressure at 70 psig and stable, it is credible to postulate that
PCV-1 may not have closed completely. The NRC agrees it would be appropriate to attempt to
mitigate the source of the air leak by using other, procedurally directed, steps prior to inserting
a manual scram. Based on the procedural guidance in ON 3146, the Caution before Step 7
and Step 7, it would also be appropriate to make a page announcement regarding the loss of
service air (for personnel safety reasons), to verify the automatic closure of control valve
SA-PCV-1, and to take manual actions to override and close SA-PCV-1, if the valve failed to
close. Therefore, answer A should also be considered correct.
Answer D, although procedurally driven in follow-up action Step 2 of OT 3113, requires the
applicant to assume that the FWRV failed to reset. There is not enough information in the
question stem to make that assumption. In addition, the aux FWRV would not provide
adequate FW flow capacity at 20% power in order to stabilize and restore level. Therefore,
answer D is not acceptable.
Answer B states in part that control rods are expected to drift at this pressure. Given that
the stem states scram air pressure is 70 psig and control rod drifts are not expected until air
pressure lowers to less than 55 psig (ON 3146, symptoms step 2.d), answer B is not
acceptable.
Conclusion
The NRC staff accepts the licensees comment to accept both answers A and C as
correct answers for SRO question #77.
B-5
Original SRO Question 79:
Select the correct answer:
During a Rx Startup at. 60% power, a spurious MSIV closure (MSIV 86A) causes a high
pressure condition.
The current plant conditions are:
! Rx pressure 990 psig
! Steam flow 4.7 x 106 lbs/hr
At this point:
a. Core Thermal Limits evaluations are valid and power may be held steady for an
indefinite period of time.
b. Core Thermal Limits evaluations are valid and MSIV 86A must be reopened within 2
hours.
c.
Core Thermal Limits are suspect and power must be reduced to < 25% before MSIV
86A can be reopened.
d. Core Thermal Limits are suspect & MSIV 86A should be reopened within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or
power shall be reduced to < 25% in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Submitted answer explanation:
Justification
a. Correct Response - The old standard called for a power reduction or MSIVs must be
reopened within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. A recent analysis at the end of 2004 allows us to maintain
power.
b. Incorrect - Thermal Limits are valid indefinitely if < 4.8 x 10E6 lbs/hr
c.
Incorrect - MSIVs may be reopened at current power level
d. Incorrect - An old standard that no longer applies
LICENSEES JUSTIFICATION FOR CHANGE
Revised answer explanation:
The original correct answer for this question is A and was justified as correct by the following
statement: The old standard called for a power reduction or MSIVs must be reopened within 2
hours. A recent analysis at the end of 2004 allows us to maintain power. This analysis was
captured in OT 3115, Reactor Pressure Transients (Rev 9), and was effective December 16,
2004. It was part of an effort to combine two procedures into one procedure: OT 3116, Reactor
High Pressure, and OT 3115, Reactor Low Pressure. These were combined together into OT
3115 and re-titled Reactor Pressure Transients. The change kept the immediate required
actions for an MSIV closure as per the previous OT 3116, Reactor High Pressure. It did,
B-6
however, delete the actions as required in choice D and replaced those actions with a note.
The actions and note in OT 3115, Reactor Pressure Transients, pages 3 and 4, reads:
If pressure change was due to an MSIV closure, THEN:
a.
Reduce reactor power at a rate not to exceed 10% RTP/min. until the
following criteria are satisfied:
1) Reactor Pressure < 1000psig.
AND
2) No. of Isolated Steam Lines
Total Steam Flow
<4.8 Mlbs/hr (<75% RTP)
<3.2 Mlbs/hr
NOTE
Long term steady state operation with 1 MSIV isolated is permissible provided reactor
power is limited to <75% of rated thermal power.
(GENE-L12-00-068)
The stem of the question states that a high pressure condition occurred. This would require
the plant to enter OT 3115, Reactor Pressure Transients. However, the stem of the question
also states that the plant is in a startup. This would require the plant to be in OP 0105, Reactor
Operations. The analysis was not captured in OP 0105, Reactor Operations, where a caution
(page 76) states:
If any of the following conditions exist when CTP >25% RTP, then consider core
thermal limits (MCPR and APLHGR) evaluation suspect: (EDCR 97-422)
Less than 4 main steam lines passing steam (except surveillances).
MTS-2 tripped or any bypass valve inoperable.
Feed water heater string bypassed.
Then exit the condition within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce CTP to less than 25% RTP
within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
B-7
This caution requires an action to be taken for the conditions as stated in the question stem.
Choice A is incorrect since the old standard still applied in OP 0105, Reactor Operations.
Choice D was originally justified as incorrect with the statement: An old standard that no
longer applies. As stated above, the standard still applied as a caution in OP 0105, Reactor
Operations. This makes choice D correct.
The plant would have entered OT 3115, Reactor Pressure Transients, when the high pressure
condition occurred. The conditions in the stem require no further action per this procedure.
The plant would exit OT 3115, Reactor Pressure Transients, and return to OP 0105, Reactor
Operations. This procedure would require the actions of choice D.
If it is assumed both procedures are utilized at the same time then choice D is still the only
correct answer. Choice A can be restated as no further actions are necessary. Choice D
is a procedurally directed action which thus makes choice A incorrect. With the industry
emphasis on conservative decision making, the actions listed in choice D are the most
conservative, procedurally directed actions.
Conclusion: Answer D should be accepted as the correct answer. In this case, two approved
procedures, OP 0105, Reactor Operations (Rev 11), and OT 3115 (Rev 9), Reactor Pressure
Transients, had conflicting guidance for the given condition. (A condition report, CR-VTY-2005-
00423, was generated to resolve the conflict.) The procedural conflict was not identified during
the examination validation process. Choice D is correct because it is still a procedurally-
directed action from an approved procedure. Choice A is incorrect because it states that the
plant may operate indefinitely and this is not allowed per OP 0105, Reactor Operations.
B-8
NRC RESOLUTION:
The NRC agrees with and accepts the licensees recommendation. The plant would have
entered OT 3115, Reactor Pressure Transients, when the high pressure condition occurred.
The conditions in the stem require no further action per this procedure. The plant would then
exit OT 3115 and return to OP 0105, Reactor Operations. This procedure would require the
actions of choice
- D. In this case, two approved procedures, OP 0105, Reactor Operations
(Rev 11), and OT 3115 (Rev 9), Reactor Pressure Transients, had conflicting guidance for the
given condition. (A condition report, CR-VTY-2005-00423, was generated to resolve the
conflict.). Choice D is correct because it is still a procedurally-directed action from an
approved procedure and these actions are the most conservative. Choice A is incorrect
because it states that the plant may operate indefinitely and this is not allowed per OP 0105,
Reactor Operations.
Conclusion
Accept the licensees recommendation to change the correct answer for SRO Question
- 79 from A to D.