IR 05000280/2017301
| ML18019A516 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/12/2018 |
| From: | Gerald Mccoy Division of Reactor Safety II |
| To: | Stoddard D Virginia Electric & Power Co (VEPCO) |
| References | |
| ER 2017301 | |
| Download: ML18019A516 (28) | |
Text
UNITED STATES ary 12, 2018
SUBJECT:
SURRY NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000280/2017301 AND 05000281/2017301
Dear Mr. Stoddard:
During the period October 30 - November 3, 2017, the Nuclear Regulatory Commission (NRC)
administered operating tests to employees of your company who had applied for licenses to operate the Surry Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests and the written examination submittal with those members of your staff identified in the enclosed report. The written examination was administered by your staff on November 8, 2017.
All ten applicants passed the operating test, while nine applicants passed the written examination. There were 17 post-administration comments concerning the written examination.
These comments, and the NRC resolution of them, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial written examination submitted by your staff met the guidelines for quality contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). If you have any questions concerning this letter, please contact me at (404) 997-4551.
Sincerely,
/RA/
Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Docket Nos: 50-280, 50-281 License Nos: DPR-32, DPR-37
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Docket No.: 50-280, 50-281 License No.: DPR-32, DPR-37 Report No.: 05000280/2017301 and 05000281/2017301 Licensee: Dominion Energy Facility: Surry Nuclear Plant, Units 1 and 2 Location: 5850 Hog Island RD Surry, VA 23883 Dates: Operating Test - October 30 - November 3, 2017 Written Examination - November 8, 2017 Examiners: M. Donithan, Chief Examiner, Operations Engineer P. Capehart, Senior Operations Engineer J. DeMarshall, Operations Engineer Approved by: Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Enclosure 1
SUMMARY
ER 05000280/2017301, 05000281/2017301; October 30 - November 3, 2017 & November 8, 2017; Surry Nuclear Plant; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 11 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
Members of the Surry Nuclear Plant staff developed both the operating test and the written examination. All met the quality guidelines contained in NUREG-1021.
The NRC administered the operating test during the period October 30 - November 3, 2017.
Members of the Surry Nuclear Plant training staff administered the written examination on November 8, 2017. Three Reactor Operator (RO) and seven Senior Reactor Operator (SRO)applicants passed the operating test; two ROs and seven SROs passed the written examination. Nine applicants were issued licenses commensurate with the level of examination administered.
There were 17 post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC evaluated the submitted operating test by combining the scenario events and job performance measures (JPMs) in order to determine the percentage of submitted test items that required replacement or significant modification. The NRC also evaluated the submitted written examination questions (RO and SRO questions considered separately) in order to determine the percentage of submitted questions that required replacement or significant modification, or that clearly did not conform with the intent of the approved knowledge and ability (K/A) statement. Any questions that were deleted during the grading process, or for which the answer key had to be changed, were also included in the count of unacceptable questions. The percentage of submitted test items that were unacceptable was compared to the acceptance criteria of NUREG-1021, Operator Licensing Standards for Power Reactors.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.
The NRC administered the operating tests during the period October 30 - November 3, 2017. NRC examiners evaluated three RO and seven SRO applicants using the guidelines contained in NUREG-1021. Members of the Surry Nuclear Plant training staff administered the written examination on November 8, 2017. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Surry Nuclear Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
The NRC evaluated the performance or fidelity of the simulation facility during the preparation and conduct of the operating tests.
b. Findings
No findings were identified.
The NRC developed the written examination sample plan outline. Members of the Surry Nuclear Plant training staff developed both the operating test and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 11 of NUREG-1021. The NRC examination team reviewed the proposed examination. Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021 and incorporated into the final version of the examination materials.
The NRC determined that the licensees written examination and operating test submittals were within the range of acceptable quality for a proposed examination specified by NUREG-1021.
No issues related to examination security were identified during preparation and administration of the examination.
Nine applicants passed both the operating test and written examination and were issued licenses. One RO applicant failed the written examination and was denied a license.
Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted 17 post-examination comments concerning the written examination, of which 10 were suggestions by the applicants to modify the answer key.
The facility agreed with only 2 of these, and of those recommended changes, only 1 resulted in an answer key change after NRC review. A copy of the final written examination and answer key, with all changes incorporated, may be accessed not earlier than November 8, 2019, and a copy of the licensees post-examination comments, may be accessed in the ADAMS system (ADAMS Accession Numbers ML17354B289, ML17354B295 and ML17354B287).
4OA6 Meetings, Including Exit
Exit Meeting Summary
On November 3, 2017 the NRC examination team discussed generic issues associated with the operating test with Fred Mladen, Site Vice President, and members of the Surry Nuclear Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
KEY POINTS OF CONTACT Licensee personnel Fred Mladen, Site Vice President Terri Cuthriell, Licensing Engineer Mike Haduck, Manager Nuclear Outage & Planning Ron Herbert, Engineering Manager Carl Irwin III, Supervisor Nuclear Training Randy Johnson, Operations Manager Rich Philpot, Training Manager John Rosenberger, Director, Director Nuclear Safety and Licensing Jim Shell, Assistant Operations Manager-Training David Wilson, Maintenance Manager
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the licensees post-examination comments can be found in ADAMS under
Accession Number ML17354B287.
Item #1: RO Written Exam Question #1
Post-Examination Comment
One applicant contended that there was some ambiguity with the term shutdown margin vice
reactivity. He did not suggest that another answer should be accepted, or that the designated
correct answer was not correct.
Given the following:
- A Reactor Trip occurs on Unit 1 from 100% power.
- 1-ES-0.1, Reactor Trip response, has been implemented.
- Control Bank D Rod H14 IRPI indicates 218 steps.
- All other control rods indicate 0 steps.
- TAVG is 547 oF and stable.
Which ONE of the following completes the statements below?
1) One minute after the trip, shutdown margin is rising due to the decay of __(1)__.
2) In accordance with 1-ES-0.1, Reactor Trip Response, Emergency Boration
__(2)__ required.
A. 1) Iodine 2) is NOT
B. 1) Iodine 2) is
C. 1) Xenon 2) is
D. 1) Xenon 2) Is NOT
Answer Key correct answer: A
The facilitys position was that the question was adequate as written.
NRC Resolution
This was a Modified Bank question from the
- V.C. Summer 2013 NRC exam, and that question
also used the term shutdown margin. Shutdown margin in the given circumstances is a
measure of how much negative reactivity is in the reactor core. As Iodine decays it becomes
Xenon, a powerful neutron absorber which adds more negative reactivity to the core, increasing
the margin by which the reactor is shut down. Use of the term shutdown margin is technically
acceptable in this question.
No answer key change is required.
Item #2: RO Written Exam Question #8
Post-Examination Comment
An applicant asked if 50°F was a potential subset of the answer, 30°
- F.
The Crew is responding to a SGTR in accordance with 1-E-3, Steam
Generator Tube Rupture.
- The Rapid RCS cooldown has been completed.
Which ONE of the following describes:
1) The subcooling value for RCS depressurization termination.
2) The purpose for the RCS depressurization.
A. 1) 50 °F.
2) Minimize break flow and refill the pressurizer.
B. 1) 50 °F.
2) Minimize Reactor Vessel stress for PTS concerns.
C. 1) 30 °F.
2) Minimize break flow and refill the pressurizer.
D. 1) 30 °F.
2) Minimize Reactor Vessel stress for PTS concerns.
Answer Key correct answer: C
The facility determined that there was no subset issue and the question was adequate as
written.
NRC Resolution
This is not a subset issue because of the wording of procedure 1-E-3 Step 19.c: RCS
depressurization is to continue until subcooling is LESS THAN 30°F. If the depressurization
were stopped when subcooling reached 50°F (and getting smaller), it would be stopped too
early, and the Westinghouse Owners Group Basis strategy for the step would not have been
met.
No change to the answer key is required.
Item #3: RO Written Exam Question #13
Post-Examination Comment
One applicant asked if there was a Plant Computer System (PCS) alarm to support to support
the 70% distractor.
Initial Condition
- Unit 1 is operating at 100% power.
- Unit 2 is in RSD with core off-load to the SFP complete 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ago.
Current Condition
- CC surge tank level is 55% and rising.
Which ONE of the following describes:
1) CC Surge Tank High Level alarm setpoint is ___(1)___.
2) The CC Surge Tank Vent Valve ___(2)___ automatically close on high level.
A. 1) 93%
2) will not
B. 1) 93%
2) will
C. 1) 70%
2) will not
D. 1) 70%
2) will
Answer Key correct answer: A
The facility noted that there is no PCS alarm for this function, and the question was adequate as
written.
NRC Resolution
The licensee was asked whether PCS displays CC Surge Tank level. The answer was that CC
Surge Tank level does not input to PCS at all.
No change to the answer key or grading is required.
Item #4: RO Written Exam Question #15
Post-Examination Comment
Three applicants commented that the question stem was confusing where under Current
Conditions the 2nd bullet stated that Unit 2 parameters are comparable.
Initial Conditions:
- Unit 1 and Unit 2 are operating at 100%.
- The plant has been notified by SOC that there are significant grid instabilities due to
numerous base load plants being out of service.
- The SOC has requested maximum power generation from both units.
Current Conditions:
- The SOC reports that the 500 KV system is most impacted and the Emergency Low
Limit for that network has been reached.
- The BOP reports the following Unit 1 changes (Note: Unit 2 parameters are
comparable):
GEN GEN Grid Gen Gen Gen
MWe MVARS Freq Amps Volts H2
Press
Initial 907 + 10 60 Hz 23,500 22.3 KV 75 psig
Current 1020 - 210 58.0 Hz 27,500 21.6 KV 75 psig
Which ONE of the following completes the statements below?
1) Based on the information given the Reactor __(1)__ required to be tripped.
2) Which Emergency buses are most affected by the current conditions stated above?
(REFERENCE PROVIDED)
A. 1) is 2) Buses 1J and 2H.
B. 1) is not 2) Buses 1J and 2H.
C. 1) is not 2) Buses 1H and 2J.
D. 1) Is 2) Buses 1H and 2J.
Answer Key correct answer: A
The facility noted that the current conditions indicate a grid frequency below the RCP breaker
trip setpoint, so the question is adequate as written
NRC Resolution
The statement was simply meant to indicate that Unit 2 conditions were similar, without having
to provide a specific data table for Unit 2 to this already busy question. The conditions given for
Unit 1 clearly indicate that a reactor trip should have already occurred on bus low frequency.
No grading change is warranted.
Item #5: RO Written Exam Question #18
Post-Examination Comment
One applicant noted that Lesson Plan ND-95.3-LP-41 and a Caution in procedure FR-H.1 state:
If feed flow is reduced due to operator action to minimize feed flow as instructed in these
guidelines and the capability of providing the minimum feed flow is available, then a transfer
from these procedures is not required and FR-H.1 should not be performed.
Initial Conditions:
- Unit 1 reactor operating at 100% power.
- The reactor is tripped and safety injection actuated.
- A steam leak is reported in Unit 1 Safeguards.
Current Conditions:
- The Crew has entered ECA-2.1, Uncontrolled Depressurization of All Steam
Generators.
- All SG NR levels are off-scale low.
- RCS Cooldown rate is 155°F/hour.
Which ONE of the following states:
1) What is the MINIMUM AFW flow (gpm) to each SG?
2) Will a transition to 1-FR-H.1 be required after AFW has been throttled?
A. 1) 60
2) No
B. 1) 100
2) No
C. 1) 60
2) Yes
D. 1) 100
2) Yes
Answer Key correct answer: C
The facilitys position was that it is correct that FR-H.1 should not be performed, but the crew is
still required to transition to FR-H.1 to read the caution, and then exit FR-H.1. Furthermore,
license candidates are trained to transition to the appropriate FR and then exit if appropriate, so
the question was adequate as written.
NRC Resolution
Per procedure rules of usage, an Orange or Red terminus on the Critical Safety Function Status
Trees must transition to a functional recovery procedure. FR-H.1 recognizes that other
Emergency Operating Procedures may have purposely lowered feedwater flow, and takes this
into account with the Caution prior to Step 1. But you have to be in FR-H.1 for that Caution to
be applicable.
No change to the grading is required.
Item #6: RO Written Exam Question #21
Post-Examination Comment
1) One applicant stated that the labeling for choices A2 and C2, Vent Vent 1 Gas was
confusing, and that operators dont call it that.
2) A different applicant asked if flow couldnt have been aligned to Vent Stack #1.
Given the following:
- Fuel movement is in progress in the Fuel Building.
- Fuel Handlers report an assembly has been dropped and appears to be
damaged.
- 1-RM-RI-153, FUEL PIT BRIDGE, HIGH Alarm is received.
- The Crew has initiated 0-AP-22.00, Fuel Handling Abnormal Conditions.
Which ONE of the following identifies:
1) The MCR must be ISOLATED within ___(1)___ minutes.
2) The release is monitored using ___(2)____.
A. 1) 2
2) 1-VG-RI-104, Vent Vent 1 Gas
B. 1) 2
2) 1-VG-RI-131 A/B/C, Vent #2 Gas/Particulate
C. 1) 60
2) 1-VG-RI-104, Vent Vent 1 Gas
D. 1) 60
2) 1-VG-RI-131 A/B/C, Vent #2 Gas/Particulate
Answer Key correct answer: B
The facilitys position was:
1) This is how the instrument is labeled in the plant.
2) Flow could be aligned to Vent stack #1 but there was nothing in the question to indicate a
different alignment than the normal alignment.
No change required, the question is adequate as written.
NRC Resolution
1) Whether Vent Vent 1 Gas was confusing or not, the equipment designator 1-VG-RI-104
specifies the instrument exactly.
2) Applicants are to assume normal alignment for all systems unless specifically directed
otherwise by the question. This was covered in the Policies and Guidelines for Taking NRC
Examinations brief that was held with the applicants in accordance with NUREG 1021
Appendix E, therefore the answer should be based on the normal alignment of the ventilation
system.
A grading change is not justified.
Item #7: RO Written Exam Question #24
Post-Examination Comment
One applicant noted that Attachment 1, System Alignment Verification, also has direction to
check the position of 1-RM-TV-100C.
Given the following:
- Unit 1 is operating at 100% power.
- The reactor is tripped and SI actuated.
- Containment pressure is 25 psia and rising slowly.
Which ONE of the following completes the statement:
1-RM-TV-100C, CTMT ATMOS RM INLET I/S TV, closes on a ___(1)___
containment isolation signal, and is checked in the required position using
___(2)___ of 1-E-0, Reactor Trip or Safety Injection.
A. 1) Phase I
2) Attachment 1, System Alignment Verification
B. 1) Phase I
2) Attachment 4, CLS Component Verification
C. 1) Phase II
2) Attachment 4, CLS Component Verification
D. 1) Phase II
2) Attachment 1, System Alignment Verification
Answer Key correct answer: C
The facilitys position was that while Attachment 1 step 8 Response Not Obtained (RNO) does
check the positions of 1-RM-TV-100A/B/C, this is only done if step 8a is answered as N
- O.
The question stem puts containment pressure at 25 psia, which is above the step 8a setpoint of
psia. Therefore step 8 RNO would not be performed. The question was adequate as written,
no change required.
NRC Resolution
does check the position of 1-RM-TV-100C, but that RNO step isnt reached with
the containment conditions given in the question. The question was broadly trying to test
whether applicants knew that Att. 1 checks Phase I valves, while Att. 4 checks Phase II & III
valves. The intent was not to be tricky by having the subject valve in both Attachments, and
from the question-development documentation it doesnt appear that the facility exam writers or
the Chief Examiner noticed that 1-RM-TV-100C was even in Att. 1.
Because the question is technically correct, no change to the grading is necessary.
Item #8: RO Written Exam Question #28
Post-Examination Comment
One applicant asked why TAVE was lower, and couldnt it be higher instead.
Given the following:
- Unit 1 is operating at 29% power.
- 1-RC-P-1A breaker spuriously trips open on overcurrent.
With no operator actions which of the following describes:
1) How Loop A Tavg will change as compared to previous Tave Loop A.
2) The reason for the Loop A Tavg change?
A. 1) Lower 2) due to no forced flow in the loop A.
B. 1) Lower 2) due to reverse flow in the loop A.
C. 1) Higher 2) due to reverse flow in the loop A.
D. 1) Higher 2) due to no forced flow in the loop A.
Answer Key correct answer: B
The facility noted that TAVE is lower because at this power level B and C loops will provide load
demand (the reactor would not have tripped). TAVE is lower because as flow is reversed, TCOLD
rises, THOT lowers, and TCOLD is greater than THOT. They ran this scenario on the classroom
simulator obtained those results. Question is adequate as written, no change required.
NRC Resolution
The source for this question was the 2008 Diablo Canyon NRC exam, which had the same TAVE
will be lower answer. This is a fundamental concept for Pressurized Water Reactors, and a
fairly common exam question. The facilitys description of the thermodynamic behavior is
correct, and is supported by the results from their glass-top simulator.
No question or grading changes are warranted.
Item #9: RO Written Exam Question #31
Post-Examination Comment
One applicant thought the stem of the question was confusing by using the term Station
Instrument Air in the last bullet of the Initial Conditions.
Initial Conditions:
- Unit 1 is in CSD with the PRZR solid.
- 1-RC-P-1C, C RCP running for crud burst cleanup.
- RCS temperature is 158 °F and lowering.
- RCS pressure is 305 psig.
- Station Instrument Air is supplying containment, the CMT IA
compressors are in OFF.
Current Conditions:
- Unit 1 Instrument Air header ruptures.
Which ONE of the following completes the statements below:
1) RCS pressure will ___(1)___.
2) Procedure ___(2)___ is used to mitigate the effects of this transient.
A. 1) rise
2) 1-AP-27.00, Loss of Decay Heat Removal Capability
B. 1) lower
2) 1-AP-27.00, Loss of Decay Heat Removal Capability
C. 1) rise
2) 0-AP-40.00, Non-Recoverable Loss of Instrument Air
D. 1) lower
2) 0-AP-40.00, Non-Recoverable Loss of Instrument Air
Answer Key correct answer: C
The facility stated that the question was adequate as written.
NRC Resolution
Containment Air at Surry is normally provided by dedicated air compressors adjacent to the
containment building, which draw air from containment and return it to containment air
systems. But as given in the Initial Conditions, Unit 1 was in Cold Shutdown and these air
compressors were off, so in that situation the air supply is from outside containment.
The terminology Station Instrument Air is used throughout Lesson Plan ND-92.1-LP-1, Station
Air Systems, specifically at D.4.(d) on page 27:
In the event that containment instrument air compressors cannot supply the system needs,
the header can be kept pressurized by the Station Instrument Air System.
This is exactly the situation described in the question. So the normal Containment Air system is
not in service, and now its backup system cant maintain air pressure due to a header rupture.
No grading modification is necessary.
Item #10: RO Written Exam Question #42
Post-Examination Comment
One applicant asked if it was true that TAVE would rise, and referred to IMP IN vs. IMP OUT.
Given the following:
- Unit 1 Turbine Control is in IMP IN.
- Control rods are in MANUAL.
- The following parameters are given:
Rx. Pwr Delta T Tave T ref. Gen MWe
89% 90% 569.7 569.7 784
- 1
-MS-SOV-104, MSR Steam Supply fails CLOSED.
With no operator action which ONE of the following parameters would be higher
when steady-state conditions are reached?
A. Reactor Power.
B. Generator MWe.
C. T ref.
D. Tave.
Answer Key correct answer: D
The facility re-ran the question scenario on the classroom simulator, which supported TAVE rising
once steady-state conditions were reached. No change in the question required.
NRC Resolution
When this question was submitted for approval, its supporting documentation included before
and after data obtained from the glass-top simulator. That data shows TAVE rising from
569.7°F to 574.8°F, a change of 5.1°F, which is quite significant.
No changes to the question or grading are warranted.
Item #11: RO Written Exam Question #53
Post-Examination Comment
Two applicants contended that answer choice B is also correct, because the sub-steps in
Abnormal Procedure 0-AP-12.00, Service Water System Abnormal Conditions, Step 4 RNO are
not bulleted and therefore can be performed in any order.
Initial Conditions:
- 1-SW-P-10A, CHG Pump SW Pump, is running in HAND
- 1-SW-P-10B, CHG Pump SW Pump is in AUTO and OF
- F.
Current Conditions (2 minutes later):
- 1-SW-P-10A, CHG Pump SW Pump, continues running in HAN
- D.
- 1-SW-P-10B, CHG Pump SW Pump is in AUTO and OFF.
- The Service Building Inside Operator reports the following readings:
Parameter Previous Current Log spec
1-SW-DPI-100A, 1-VS-S-1A, SW Strainer D/P 0.8 psid 0.8 psid Max: 1.5 psid
1-SW-DPI-27, 1-SW-S-2A, SW Suct. Strainer D/P 0.2 psid 1.8 psid Max: 1.5 psid
1-SW-PI-140A, 1-SW-P-10A, Suction press 3.8 psig 0.8 psig Min: -5 psig
1-SW-PI-26, 1-SW-P-10A, discharge pressure 27.5 psig 6 psig Min: 15 psig
Which of the following describes the actions the operator must take per 0-AP-12.00, Service Water
System Abnormal Conditions?
- A. Start 1-SW-P-10B, place duplex strainer 1-SW-S-10 in service, and open 1-SW-263
Suction Cross-tie.
- B. Place 1-SW-S-2A, suction strainer standby basket in service, and vent 1-SW-P-10
- A.
C. Start 1-SW-P-10B, stop and vent 1-SW-P-10A, place 1-SW-S-2A, suction strainer standby
basket in service.
- D. Place additional SW header in service, vent all three SW headers, vent 1-SW-P-10A Chg
SW pump.
Answer Key correct answer: C
The facility agreed that choice B is also correct, and provided the following marked-up page
from the procedure, with subsequent justification:
Sub-steps (marked A and B above) are of the same level of indent and although A is the
more appropriate action, B contains corrective actions for this scenario. It would be
reasonable for a team, performing a scenario with the given conditions, to choose either the A
or B path.
NRC Resolution
Placing the standby strainer basket in service would correct the issue presented by the
question, and is procedurally allowed (the B option above). It appears that the question writers
and Chief Examiner were focused on pump 1-SW-P-10B not auto-starting as designed, and the
most direct and quickest way to remedy the situation would be to start the B CHG Pump SW
Pump from the main control room (the A option above). But locally swapping strainer baskets,
while taking more time, would restore normal operating conditions to the A pump, so the start
of B would no longer be necessary.
Therefore, Answer Choice B is also correct, so two correct answers were accepted for
Question 53: B and C
Item #12: RO Written Exam Question #57
Post-Examination Comment
One applicant asked if pulling input fuses would also allow 1-CH-LCV-1460A&B to be opened,
making B also a correct answer.
Given the following:
- A steam generator tube rupture has occurred.
- Due to multiple failures, the PRZR spray and PORVs are unavailable to
depressurize the RCS.
- PRZR level is offscale low.
- The Crew is attempting to re-establish Letdown flow in accordance with ECA-
3.3, SGTR with loss of Pressure Control.
In accordance with 1-ECA-3.3, which ONE of the following identifies:
1) The ______ fuses for LC-1-460C and LC-1-459C are removed.
2) 1-CH-LCV-1460A and B _______.
A. 1) input
2) open automatically
B. 1) input
2) are opened manually
C. 1) output
2) open automatically
D. 1) output
2) are opened manually
Answer Key correct answer: D
The facilitys position was that ECA-3.3 specifically states to pull the output fuses, so the
question was adequate as written.
NRC Resolution
The Chief Examiner asked the same question during question development. The answer from
the facility was that removing the input fuses would cause the Pressurizer low level signal to be
locked in, and letdown valves 1-CH-LCV-1460A & B could not be opened, so Letdown flow
could not be re-established.
And regardless of what physically happens, the question asks for which set of fuses is removed
in accordance with 1-ECA-3.3, and Attachment 8 Step 1 clearly directs removing the output
fuses (with OUTPUT being capitalized).
Furthermore, the facility advised during question development that this action is trained during
both Classroom and Simulator training, and that there is a Job Performance Measure covering
the fuse-removal task, with failure warranted if the applicant incorrectly removes the input fuses.
No answer key change is required for this question.
Item #13: RO Written Exam Question #62
Post-Examination Comment
One applicant asked if choice C would also be correct, because at this vacuum wouldnt steam
dumps be lost.
Given the following:
- Unit 1 is operating at 100%.
- Condenser Vacuum is 27.6 inches Hg and is lowering 0.4 inches Hg/minute.
- Turbine 1 and 2 operators dispatched to look for leaks.
- Generator Megawatts are 890 Mwe and lowering rapidly.
- Annunciator 1A-G1, Traveling Screens Hi Diff Lvl is Lit.
Which of the following completes the following statements:
1) __(1)__ is required to be entered.
2) 1-E-0 is required to be entered if Main Condenser Vacuum lowers to __(2)__.
- A. 1) 0-AP-12.01, Loss of Intake Canal Level
2) 25.0 in-Hg
- B. 1) 0-AP-12.01, Loss of Intake Canal Level
2) 22.5 in-Hg
- C. 1) 1-AP-14.00, Loss of Main Condenser Vacuum
2) 25.0 in-Hg
- D. 1) 1-AP-14.00, Loss of Main Condenser Vacuum
2) 22.5 in-Hg
Answer Key correct answer: D
The facilitys position was that 1-AP-14.00 specifically states that if Condenser vacuum lowers
to 22.5 in - Hg, the turbine must be tripped (if power is > 10%), so the question is adequate.
NRC Resolution
The correct answer is definitively supported by 1-AP-14.00 Step 2 (if condenser vacuum is not
>22.5 Hg, trip the reactor). The question asks when 1-E-0 is required to be entered, so per this
step in AP-14.00, tripping at 25.0 Hg would be much too soon, and would bypass steps that
might restore condenser vacuum and thus avoid an unnecessary reactor trip.
But the applicants comment seems to ask if it would ever be correct to trip the reactor at 25 in-
Hg, specifically citing loss of the Steam Dumps at 25 in-Hg. It is true that one of the Steam
Dump permissives is that condenser vacuum be available, with a setpoint of 25 Hg (ref.
Lesson Plan ND-93.3-LP9, Steam Dumps - Handouts, Slide 19). There is no requirement to
trip the reactor simply because the steam dumps would not actuate if called upon to do so. In
the event that steam release were necessary from the Steam Generators, the one Power
Operated Relief Valve and five Code Safety Valves per Steam Generator have sufficient
capacity to meet design-basis heat removal requirements without reliance on the steam dumps
[ref. Surry Technical Specification 3.6-5 Basis].
No change to the answer key for this question is warranted.
Item #14: RO Written Exam Question #63
Post-Examination Comment
One applicant asked why choice A wouldnt also be correct, because isnt 1-SD-LCV-106 the
divert valve.
Unit 1 is operating at 100% when the Reactor Operator notices that Main Feed
Pump suction pressure has lowered on both Main Feed pumps and is now
reading 400 psig (20 psig lower than before).
Which one of the following identifies a possible cause for this lower reading.
- A. 1-SD-LCV-106, HP Heater Drain Pump Level Control valve fails OPE
- N.
B. 1-CN-FCV-107, Condensate Recirculation valve inadvertently opened.
- C. 1-CP-MOV-100, Condensate Polisher Bypass valve fails OPE
- N.
D. 1-CN-126, Condensate bypass for LP FW Heaters 2, 3 and 4
inadvertently opened.
Answer Key correct answer: B
The facility responded that 1-SD-LCV-106 is the normal Discharge LCV to the Feed Pump
Suctions. 1-SD-LCV-107 is the divert valve. Opening 1-SD-LCV-106 further will cause a rise in
Feed pump suction pressure. No change is required, the question is adequate.
NRC Resolution
The noun name for 1-SD-LCV-106 given in the question stem is HP Heater Drain Pump Level
Control Valve. Simulator Drawing SD1C provided to the NRC as a technical basis for the
question indicates that 1-SD-LCV-106 is in the main flowpath from the Heater Drain Pumps to
the Main Feed Pumps. This valve failing open (from its normal partly-throttled position) would
cause more of the Heater Drain Pumps discharge pressure to be applied to the suction of the
Main Feed Pumps, making their suction pressure higher.
No change to the answer key is warranted.
Item #15: RO Written Exam Question #67
Post-Examination Comment
One applicant contended that answer choice C is also correct per GOP-1.4, Attachments 6
and 7.
Which ONE of the following completes the statement from 1-GOP-1.4, Unit Startup, HSD
to 2% Reactor Power, concerning the Reactivity Plan provided by Reactor Engineering?
The plan shall contain recommendations for control of Delta Flux, rod height
and/or Boron adjustments, and _____.
B. startup rate limitations
C. RCS temperature control
D. source range counts at doubling points
Answer Key correct answer: A
The facility agrees that choice C is also correct, because 1-GOP-1.4 also references the
distractor RCS temperature control in attachment 7. The following supporting material was
provided:
Supporting Facts:
- This question references step 5.1.3 to support the correct answer:
And in this reference, Tave is not listed, but upon review of attachment 7 (Reactivity Control and
Monitoring During Startup), the following is given:
The statement Use the Reactivity Plan as a guide infers that the plan contains this information.
- Upon review of an engineering supplied reactivity plan, Tave is a parameter given as a
reference:
Conclusion:
- Question 67 contains two correct answers as the reference given (GOP-1.4) contains
supportive information for A and C.
NRC Resolution
NRC does not agree that C is also a correct answer. The question specifically asks which
answer choice completes the statement from 1-GOP-1.4 [emphasis added] That
statement, as submitted by the facility during exam development as the explanation for the
correct answer, is Step 5.1.3 on p. 12 of 1-GOP-1.4, Rev. 60, which the licensee included
above. The third bullet is the correct answer. Note that RCS temperature is not included in the
list.
Furthermore, 1-GOP-1.4 in three places (Step 5.2.5.m, Step 5.4.15, and Attachment 4) provides
direction to maintain RCS temperature at approximately 547°F. There is no need to infer that
the Reactivity Plan is controlling the desired temperature.
Additionally, later when the main generator is connected to the grid, the main turbine impulse
chamber pressure provides a signal for calculation of the reference RCS temperature, TRE
- F.
GOP-1.5, Unit Startup, 2% Reactor Power to Max Allowable Power, gives direction in 4 places
(5.3.22, 5.5.6, & 5.5.43, and Attachment 9 Step 4) to maintain TAVE close to TRE
theres no need to look to a Reactivity Plan for guidance on how to maintain TAVE: its to be
maintained close to TRE
- F.
Finally, even though the Reactivity Plan has a column titled RCS TAVE, this is essentially the
value of TREF at the corresponding Core Power level. Reactor Engineering creates a Reactivity
Plan to take the reactor from shutdown to some power level, typically 100%. The waypoints
(typically at 15-minute intervals) are chosen based on nuclear fuel constraints related to rate of
power rise, heat removal capabilities, fission product poisons, and other factors. To calculate
the amount of control rod withdrawal and/or dilution water or boric acid to specify, the engineer
has to consider the current power level and the next desired power level. Since these two
power levels will have different TREF values, the engineer has to take into account the expected
change in TREF, and will provide reactivity variables (rod height, dilution water) that will match
TAVE with TREF at the new power level. So while TAVE values are listed on the Reactivity Plan,
theyre not the driver for maintaining a particular TAVE. That comes from the GOP guidance to
maintain TAVE close to TREF. This column could be left off of the Reactivity Plan and not impact
the crews ability to operate, because they are always expected to maintain TAVE near TREF.
No change to the answer key is justified.
Item #16: RO Written Exam Question #69
Post-Examination Comment
One applicant contended that choice B is also correct since VCT level is used in the
calculation for identified leakage.
According to surveillance procedure 1-OPT-RC-10.00, Reactor Coolant Leakage-
Computer Calculated, which of the following parameters is an input into the calculation
for IDENTIFIED RCS leakage?
A. Pressurizer Level.
B. VCT Level.
C. Pressurizer Relief Tank Level.
D. Containment Sump Level.
Answer Key correct answer: C
The facility noted that VCT level and Pressurizer level are used to calculate Total Leakage, but
that only PRT and PDTT levels are used to calculate Identified Leakage. Their position was that
the question is adequate as written.
NRC Resolution
RCS Leakage is categorized as Total, Identified, or Unidentified.
To determine TOTAL RCS leakage, RCS temperature and Pressurizer level are held constant,
and the decrease in VCT level in gallons divided by the duration time of the test gives the
TOTAL leak rate. This is the basis for Distractor B, VCT Level.
IDENTIFIED leakage is determined by measuring leakage that is captured and conducted to
collection systems (terminology from Technical Specifications). The Pressurizer Relief Tank is
one such collection system; if level increases in this tank then it can only be because of RCS
leakage from a known set of components. This is the support for the correct answer, C. Note
that VCT level is not needed for this calculation; only the change in PRT level is required.
UNIDENTIFIED leakage is simply the amount of TOTAL leakage that is not IDENTIFIED, and is
obtained by subtracting IDENTIFIED leakage from TOTAL leakage.
One does not need to know TOTAL leakage (derived from VCT level change) to calculate
IDENTIFIED leakage, therefore B is not a correct answer. No change to the answer key is
necessary.
Item #17: RO Written Exam Question #75
Post-Examination Comment
One applicant commented that the CAP page of 1-ES-1.1 affirms the answer of D.
Given the following:
0130 Unit 1 was at 100% power when a spurious Safety Injection occurred.
0131 The crew enters 1-E-0, Reactor Trip or Safety Injection.
0145 The crew transitions to 1-ES-1.1, SI Termination.
0155 The RCS is solid and PRZR PORV,1-RC-PCV-1456 begins to cycle.
0158 HHSI flow has been isolated and charging placed in service.
212 While placing letdown in service 1-RC-PCV-1456 opens and will not close.
217 The block valve for 1-RC-PCV-1456 will not close.
- RCS pressure is 1000 psig and lowering rapidly.
- All steam generators are at 900 psig and slowly lowering.
Which ONE of the following describes the correct actions and procedure implementation?
- A. Initiate Safety Injection and go to 1-E-0, Reactor Trip or Safety Injection.
- B. Transition to 1-E-1, Loss of Reactor or Secondary Coolant, and use 1-E-1
guidance to reinitiate Safety Injection.
- C. Manually start charging pumps and align HHSI flow path, and go to 1-ES-1.2
Post LOCA Cooldown and Depressurization.
- D. Manually start charging pumps and align HHSI flow path, and go to 1-E-1, Loss
of Reactor or Secondary Coolant.
Answer Key correct answer: D
The facility agreed.
NRC Resolution
The intent of this comment is not clear, but the Continuous Actions Page (the CAP page) for 1-
ES-1.1 was the basis for the correct answer that was given in the question development
package. No changes to the question or answer key are required.
SIMULATOR FIDELITY REPORT
Facility Licensee: Surry Nuclear Plant
Facility Docket No.: 050000280, 050000281
Operating Test Administered: October 30 - November 3, 2017
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 Inspection
Procedure 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 issues were identified.
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