ML17329A573
| ML17329A573 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/24/1992 |
| From: | Bailey R, Burdick T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17329A572 | List: |
| References | |
| 50-315-OL-92-02, 50-315-OL-92-2, NUDOCS 9207310013 | |
| Download: ML17329A573 (15) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION REGION III Report No.
50-315/OL-92-02 Docket Nos.
50-315; 50-316 Licenses No.
Indiana Michigan Power Company One Riverside Plaza
- Columbus, OH 43216 Facility Name:
Donald C.
Cook Nuclear Power Plant Examination Administered At:
Bridgman, Michigan Examination Conducted:
Week of July 6, 1992 Chief Examiner:
B l y Approved By:
Thomas M. Burdick, Chief Operator Licensing Section 2
Date Z ~/9L Date Examination Summar Examination administered on the week of Jul 6
1992 Re ort No. 50-315 OL-OL-92-02 Written and operating examinations were administered to six reactor operator and three senior reactor operator candidates.
An exit meeting was conducted with plant management on July 10, 1992.
Results:
All candidates passed the operational examination.
One reactor operator candidate and one senior reactor operator candidate failed the written examination.
During the administration of the simulator examination, several of the candidates demonstrated a weakness in interpreting the meaning and significance of selected main control board alarms.
Additionally, selected emergency procedure steps lacked a
checklist to aide the operator in verifying that expected actions were carried out properly (i.e. operation at the hot shutdown panel).
9207310013 920724 PDR ADOCK 05000315 V
REPORT DETAILS Examiners
- M. Bailey, NRC T. Burdick, NRC J.
- Hansen, NRC C. Osterholtz, NRC J. Walker, NRC
- Chief Examiner 2 ~
Exit Meetin An exit meeting was held on July 10, 1992, with facility management and training staff representatives, to discuss the examiner's observations.
h The licensee representatives acknowledged the examiner observations discussed in Section 3 of this report as well as the items identified in Enclosure 4, the Simulation Facility Report.
NRC Re resentatives in attendance were:
. M. Bailey, Chief Examiner
'T. Burdick, OLS Section Chief J.
- Hansen, Examiner C. Osterholtz, Examiner D. Hartland, Resident Inspector Facilit Re resenta'tives in attendance were:
3 ~
A. A. Blind, Plant Manager K. Baker, Assistant Plant Manager, Production B. Nichols, Operations Training Supervisor H. F. Runser, Operations Production Supervisor J. Stubblefield, RO/SRO Program Administrator Examiner Observations a ~
Examination Develo ment The licensee's reference material delivered to the NRC for examination preparation was adequately bound and labeled but an additional table of contents had to be requested to facilitate a timely exam preparation.
For the most part, the NRC examiners were able to extract the needed information for examination development.
During the examination development it was noted that training material terminology and drawings were not always consistent with plant procedures and design which hindered the written examination development and review.
The pre-examination review conducted by the licensee on the written examination portions ensured the validity and clarity of information used.-
In addition, the review process insured that the examinations were technically correct and applicable for each license type (RO/SRO) as specified by the licensee's job description.
The licensee training staff provided good support during simulator scenarios and job performance measures validation.
The facility personnel pointed out some procedural changes and revisions which resulted in modification to more than one'ob performance measure.
None of these changes were considered safety significant, but additional examiner effort was required to implement these changes.
0 eratin Examination Administration During the administration of the operating examinations, the examiners observed strengths and weaknesses regarding the senior reactor operator and reactor operator candidate's performance.
The following strengths in the candidates'erformance were observed:
1)
The senior reactor cooperator candidates demonstrated good command and control over the crew's operation and direction.,
2)
The candidates demonstrated good communication skills throughout each event with few exceptions.
3)
The candidates were able to effectively work together as a team and interact with each other to ensure continued safe plant operations.
The following weaknesses in the candidates'erformance were observed:
1)
Operational Simulator Dynamics Examination k
a)
The reactor operator candidates failed to demonstrate an adequate understanding of the
- effect or relationship that P-11
operation/indication has with bistable tripping following the failure of a pressurizer pressure channel.
b)
Majority of the candidates failed to recognize a stuck or misaligned rod during a plant power decrease transient with frequent "Rod Sequence Violation" alarms in.
- Also, some of the candidates attributed this alarm to a low temperature condition which'was inappropriate.
2)
Operational Walkthru (JPM) Examination a)
All candidates were encouraged to use available note paper to make notes of the initiating cues (directions) 'given by the examiner.
In most cases, the candidates relied on memory to perform and accompli.sh the task assigned to them which resulted in some candidates performing additional tasks.
b)
The candidates were not consistent in performing a review/verification of their initial switch lineup during instances of multiple switch operation.
This action was most apparent during operation at the hot shutdown panel which required the candidates to align the panel for local operation.'ome candidates failed to verify all switches were, in fact, aligned properly which resulted in one or more misaligned switch positions.
No procedural checklist was available to aide the operator in verification of expected actions.
Additionally, deficiencies were identified in the setup and performance of the simulator portion of the operational examination.
The following are specific examples of deficiencies associated with simulator operation:
1)
During the setup for group one scenario set, the simulator operator informed the chief examiner that the prearranged method (two-way voice communication) of cuing events was unavailable aqd an alternate method was agreed upon.
During the first scenario, the simulator operator provided the two-way voice unit which required familiarization training and resulted in some delay of the scenario performance.
Use of the two-way voice unit was less than desirable since
it required communication through the. local phone system and accidental disconnect occurred more than once.
2)
During the performance of job. performance
- measures, the. first group of candidates did not observe a stuck rod event due to an oversight of the simulator operator to verify the simulator setup was consistent with the validation setup.
The apparent cause for these difficulties was inattention to detail since the same persons were operating the simulator during both the examination and scenario validation.
c ~
Written Examination Administration The post-examination review identified several weaknesses in the candidates'nowledge as evidenced by the majority of the candidates failing to provide the correct response for each particular knowledge area examined.
This information is being provided as input to the licensee System Approach to Training (SAT) process.
No response is required.
1)
Cause(s) of expected alarm(s) during a dropped rod recovery event.
b)
Steam Dump system arming 'and operation following a reactor/turbine trip event.
c)
Operation of the Eberline Radiation Monitor system block switch arrangement.
d)
Operation of the automatic fire pump start feature.
e)
Operation of the core exit thermocouple's instrumentation reference junction.
f)
Plant effect of an overcooling event on the CVCS demineralizer resin.
2)
RO candidates only a)
Knowledge of the order of preference for the various.methods available to perform a steam generator tube rupture cooldown.
3) b)
Plant response following the failure of a controlling pressurizer level channel.
SRO candidates only a)
Administrative requirements concerning key control for vital area door keys.
b)
Knowledge of conditions that. require the initiation of emergency boration.
c)
Knowledge of source range nuclear instrumentation response to post-transient core voiding event.
d)
Understanding of plant response following. the depressurization to zero (0) psig of a faulted steam generator.
e)
Understanding of diesel generator operability requirements in regard to available fuel on site.
4
~
Written Examination Review Licensee representatives reviewed the written examination prior to administration with appropriate changes being incorporated into the examinations at that time.
Following the administration of the written examinations, the facility was given a copy of the RO and SRO examinations including answer keys for review.
The facilities post examination comments and the NRC resolutions are contained in Enclosure 2 of this report.
ENCLOSURE 2
FACILITY COMMENTS AND NRC RESOLUTION OF COMMENTS RO and SRO UESTION NO. 12:
In accordance with administrative requirements, a change sheet that lacks signatures and dates for QA, PNSRC and Plant Manager approval and was issued 15 days ago...
a.
may not be used until the signatures are obtained.
b.
may be used as approved for use by the Shift Supervisor.
c.
may not be used unless it is stamped "approved by PNSRC and Plant Manager on II date d.
may be used if stamped "Controlled Copy".
ANSWER'EFERENCE:
Standing Order OS0.010, Rev.
4, date 06/24/87 PMI-2010 FACILITY COMMENT'he signatures referenced in (a) indicate review and approval of the change in accordance with PMI 2010 Section 4.6.8 and 4.6.9.
i.e. If these signatures have not been obtained within 14 days the change sheet cannot be used.
Therefore, answer (a) is correct.
Response
(c) is also correct because if, due to time considerations, the control copies cannot be updated within the prescribed times 'a stamp can be used indicating the required signatures have been obtained.
In either case the requirement is'for obtaining signatures prior to use after 14 days which is stated in both (a) and (c).
Accept answer (a) in addition to answer (c) for full credit.
NRC RESOLUTION:
Comment Accepted.
The SRO and RO examination answer key have been modified to indicate answer a or c is correct.
17:
Plant conditions following a Unit 2 reactor trip from 904 power:
Reactor trip breakers:
One turbine stop valve:
Condenser vacuum:
RCS Tave:
OPEN OPEN 11.4 in. Hg.
561 Deg.
F The steam dump system...
a
~
b.
c ~
is armed and dumping steam to the condenser with group 1 and 2 trip open solenoid bistables energized.
is armed and dumping steam to the condenser with group 1
trip open solenoid bistables energized.
P is armed and not dumping steam to the condenser because of a blocking signal.
d.
is not armed and not pumping steam to the condenser.
ANSWER:
REFERENCE:
LP RO-C-PG12, Steam Dump System, Rev.
8, pg.
18 FACILITY COMMENT:
Request that answer (a) and (b) be accepted for full credit.
The information provided in the stem is not clear enough to determine whether the steam dump system is operating in the turbine trip or load rejection mode.
If turbine trip is assumed based on 2/3 EHC low oil pressures then (b) is correct.
Reference RO-C-PG12 page 17 and 19. If turbine is assumed not to have tripped based on 1/4 stop valves open then (a) is correct.
NRC RESOLUTION:
The information provided in the stem is sufficient enough for a competent operator to ascertain the fact that a turbine trip occurred with the exception that one turbine stop valve failed to close.
Reference RO-C-PG12, page 19 clearly states that either 2/3 low oil pressure OR 4/4 stop valves is required for a steam dump arming signal.
Logic would indicate that a turbine trip occurred since no information contrary to that fact is presented.
NRC standard practice is to provide any information regarding incorrect, component or system operations.
Therefore, this comment is not accepted.
The south SI pump was started from ambient conditions for post maintenance testing then secured after 2 minutes of run time.
Five minutes later the pump was restarted for continued testing and then, secured after 6 minutes of run time.
Which of the following is the minimum amount of time that must elapse before
,the pump is allowed to be started again?
a \\
b.
c ~
d 4 15 minutes 30 minutes 45 minutes 60 minutes ANSWER:
REFERENCE:
Standing Order OSO.019, Motor Starting Limitations, Rev.
2 FACILITY COMMENT:
Request deletion of question.
This level of detail (memorization is not consistent with the operating philosophy of this facility and this question is NOT supported by facility objectives.
- Further, the KA 006000K4.01 (NUREG 1122, Pg. 3.2-10)
RO 2.6/SRO 2.9 requires knowledge of ECCS design feature(s) and/or interlocks which provide for the following:
K4.01 ".Cooling of Centrifugal
~Pum Bearing" There is no match between 006000K4.01 relating to guum
~bearin cooling design factors and a question relating to administrative requirements for motor starts which is concerned with motor w~indin cooling.
NRC RESOLUTION Comment accepted.
The SRO and RO examinations and answer keys have been modified to indicate this question as deleted.
Unit 2 Technical Specifications require the safety injection pump to be d'eclared"inoperable on low discharge pressure during testing on recirculation flow.
Which of the following pressures is the MAXIMUMpressure at which the pump should be declared inoperable?
a ~
b.
1300 psig.
1350 psig.
c ~
d.
1400 psig.
1450 psig.
ANSWER:
REFERENCE:
Tech.
Specs.
3/4 5-5 (Unit 2)
FACILITY COMMENT:
The word "should" in the second sentence of the stem makes this question extremely confusing to a candidate.
Even though the candidate knows the inoperability setpoint in accordance with T.
S. is 1409 psig he must decide on a conservative versus a non-conservative approach of answering the question for examination purposes.
The conservative approach would indicate answer (d) is correct and consistent with operating philosophy and the non-conservative approach would indicate answer (c) is correct based on literal compliance with Technical Specifications.
If the work "must" would have been used instead of "should" the answer (c) would be the only correct answer.
Request that answer (d) and (c) be accepted for full credit.
NRC RESOLUTION:
Comment Accepted.
The SRO and RO examination answer keys have been modified to indicate that answer c or d is correct.
With Pressurizer level selection switch in position No.
1, a
failure of the pressurizer level transmitter NLP-153 (channels 1
and
- 3) in the LOW direction resulted in zero output.
Which of the following statements describes the plant response expected?
, NOTE:
NLP-152 was taken out of service for repair.
a 0 b.
All pressurizer heaters are OFF; letdown isolation valve QRV-112 is closed; charging flow remains the same.
All pressurizer heaters are OFF; letdown isolation valve QRV-111 is closed; charging flow goes to minimal.
c ~
d.
Both letdown isolation valves are closed; charging flow remains the same; pressurizer level is increasing.
Both letdown isolation valves are closed; charging flow goes to minimal; pressurizer level remains the same.
ANSWER:
REFERENCE:
LP RO-C-NS03 PZR and Pressure Relief, TP-17 FACILITY COMMENT:
Request deletion of question. from examination.
The examination review team suggested inserting "(Channels 1 and 3)" directly following "...Position No. 1" in the original stem.
The requested change,"(Channels 1 and 3)" was accepted but inadvertently placed following "...
NLP 153" in the stem of the question.
This resulted in confusion of the candidates requiring clarification by the proctor.
The proctor provided the following clarification:
"NLP 153 feeds channel 1 and 3".
Since this is not possible, students could not determine correct, answer.
NRC RESOLUTION:
Comment Accepted.
The SRO and RO examinations and answer keys have been modified to inidicate this question as deleted.
The plant has been in ATWS casualty, and boration is in progress.
Which of the following is a criteria that MUST be satisfied prior to returning to the procedure and step from which you entered?
a.
Reactor trip verified.
b.
Power range less than 54.
c.
Source range detectors energized.
d.
IR has zero start-up rate.
ANSWER:
b
REFERENCE:
FR-S.1, PG.
11 FACILITY COMMENT:
Request that both answers (a) and (b) be accepted for full credit.
Under the conditions stated in the stem, step 13 of FR-S.1 is the criteria that must be met in order to "...return to
the procedure and step in from which you entered."
Step 13 requires that both conditions "Power range channels less than 5%"
AND "Intermediate range channels negative start up rate" to be satisfied.
If the statement "a criteria" in the stem is assumed to mean a single condition that allows transition then answer'- (b) could not be correct and the candidate could choose answer (a) assuming "Reactor Trip Verified" ensures BOTH requirements of step 13 are satisfied.
NRC RESOLUTION:
Comment accepted.
The SRO and RO examination answer keys have been modified to indicate that answer a or b is correct.
69:
During a reactor start-up to criticality, the Spray Additive Tank level at 4000 gals annunciator alarms.
Which of the following actions should be taken?
a
~
Trip the reactor and enter E-O, Reactor Trip or Safety Injection.
b.
Continue the reactor/plant start-up to the point of adding heat and stabilize.
c ~
Discontinue the startup until operability requirements are met.
d.
Continue the start-up but remain less than 104 power until operability requirements are restored.
ANSWER
REFERENCE:
TS 3 '.2.2 FACILITY COMMENT:
Accept answer (b) in addition to answer (c) for full credit.
Actual plant mode (i.e.
3 or 2) was not established in the stem.
If it is assumed that mode 2 has been entered as stated in stem then answer (b) is correct because the LCO 6.2.2.2 was entered after the mode change from 3 to 2. If it is assumed that mode 2
has not been entered then answer (c) is correct because of TS 3.0.4.
NRC RESOLUTION:
Comment accepted.
The SRO and RO examination answer keys have been modified to indicate that answer b or c is correct.
With the plant operating at 100% power, turbine impulse pressure transmitter MPC-254 fails high.
What effect, if any, will this have on the rod control system with rods in automatic and no operator action?
(Assume Tavg/Tref initially matched.)
a.
Rods drive in.
b.
Rods drive out.
c.
Rods do not move.
d.
Rod bottom lights energize.
ANSWER:
REFERENCE:
C&E RX19 FACILITY COMMENT:
The keyed response (c) is correct with respect to the Tave/Tref mismatch circuit.
At 100% power Tref is at a maximum value and does NOT increase further as power is increased.
- However, when MPC 254 fails high (1204 power) initially a power mismatch (rate of change) is seen between NI power (100>)
and turbine power (1204).
This power mismatch will cause the rods to drive out making answer (b) correct.
See highlighted areas on the attached pages from lesson plan RO-C-NS04, Student Handout gl (Rod Control Lesson),
and C&E document, RX19 MPC 254 Failure.
Request that answer (b) be accepted as the only acceptable answer.
NRC RESOLUTION:
Comment accepted.
The SRO and RO examination answer keys have been modified to indicate that answer b or c is correct.
RO ONLY UESTION NO. 85:
Hydrogen recombiner system operation is permitted until the hydrogen concentration in containment reaches...
a.
3.04 b.
- 4. O~o c ~
- 5. 04 d.
ANSWER REFERENCE LP RO-C-NS15, Rev.
22, Pg.
6 FACILITY COMMENT:
Request deletion of question.
The stem implies that the hydrogen recombiners must be turned off when H2 concentration reaches a
predetermined
- value, however, all of the sited references state that the recombiners are turned on below 4% but never direct securing the 'unit as concentration increases above this value.
Therefore, there is not a correct answer to question as stated.
NRC RESOLUTION Comment accepted.
The RO examination and answer key has been modified to indicate this question as deleted.
Enclosure 4
SIMULATION FACILITY REPORT Facilit Licensee:
D.C.
Cook Nuclear Power Plant Facility Licensee Docket No.
50-315 Operating Tests Administered on:
07/06/92 through 07/10/92 During the conduct of the simulator portion of the operating
- tests, the following items were observed:
ITEM DESCRIPTION Eberline Radiation Monitor Hardware the panel's key pad numbers are unreadable in most cases.
2.'iesel Generator Software During performance of a JPM, the diesel generator KW meter output was fluctuating some 200 KW while attempting to parallel with the reserve bus.
3.
Auxiliary Feedwater Software During the performance of a steam line break scenario, both motor driven aux feedwater pumps were required to be steam bound.
Both pumps indicated lower than normal amps and their indication was stable.
Actual steam bindin'g condition should not indicate stable amps with no discharge flow indicated.
4 ~
Nuclear Instrumentation Software During the performance of six scenarios, the source range nuclear instrumentation's audible indicator did not consistently function upon source range NI energizing following a reactor trip.
In only two scenarios did the audible indication turn on as anticipated.