ML20035F029
| ML20035F029 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/12/1993 |
| From: | Jordan M, Zelig C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20035F014 | List: |
| References | |
| 50-254-OL-93-01, 50-254-OL-93-1, NUDOCS 9304200194 | |
| Download: ML20035F029 (17) | |
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I U. S. NUCLEAR REGULATORY COMMISSION REGION III i
Report No. 50-254/0L-93-01 i
Docket Nos. 50-254; 50-265 Licenses No. DPR-29; DPR-30 Licensee: Commonwealth Edison Company (Ceco)
Opus West III j
1400 Opus P1 ace Downers Grove, IL 60515 Facility Name: Quad Cities Nuclear Power Station l
Examination Administered At: Quad Cities Nuclear Power Station l
i Examination Conducted:
(Initial) March 4, 15 - 19, 1993 l
(Requal) March 4, 19, 1993 j
i Examiners:
K. Shembarger, RIII NRC t
C. Tyner, Idaho National Engineering Laboratories (INEL)
Chief Examiner:
he
/2!i3 C'. Vlfig, RIII Date Approved By:
hh
'/!/2!73 M. J. ppdan, Chief Date j
Operator Licensing Section 1 i
Examination Summary The initial R0/SR0 examination was administered on March 4. 15-19. 1993.
(Report No. 50-254/0L-93-Ol(DRS)). The examination was administered to four (4) Reactor Operator (RO) candidates, two (2) Senior Reactor Operator Instant (SR01) candidates, two (2) Senior Reactor Operator Upgrade (SR00) candidates, and one (1) Senior Reactor Operator Limited to Fuel Handling (LSR0) candidate.
Examinations were administered in accordance with guidelines of NUREG 1021, Operator Licensing Examiner Standards, Revision 6.
A requalification examination was administered to one (1) LSR0 in accordance with guidelines of NUREG 1021, " Operator Licensing Examiner Standards,"
Revision 7.
The decision to use Revision 7 for the LSR0 requalification examination was by mutual agreement of the facility and the NRC.-
Results: Eight candidates taking the initial examination successfully passed the examination. One SROI candidate failed the written section of the initial examination. The LSR0 taking the requalification examination successfully passed the examination.
9304200194 930412 PDR ADOCK 05000254 V
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Examination Summary 2
The following is a summary of the strengths and weaknesses noted during the performance of this examination.
Strenoths:
e Crew communications during dynamic simulator scenarios. (For details see Section 3.d)
Weaknesses:
o Candidate's operational knowledge of radiological control procedures.
(For details see Section 3.c) i e
Training Department's ability to validate written examination questions during pre-exam reviews.
(For details see Section 4.a) i i
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REPORT DETAILS t
1.
Examiners C. Zelig, Chief Examiner, RIII NRC K. Shembarger, RIII NRC C. Tyner, Idaho National Engineering Laboratories 2.
Persons Contacted Facility
- R. Pleniewicz, Site Vice President
- H. Hcntschel, Operations Manager
- J. Kudalis, Services Director
- B. Strub, Assistant Superintendent of Operations
+*J. Hoeller, Training Supervisor
- A. Misak, Regulatory Assurance Supervisor
+*R. Armitage, Licensed Operator Training Group Leader
++K. Rach, CECO BWR Operations Training Supervisor
+*D. Bowman, Initial License Program Coordinator
+*K. Griffith, Point of Contact with NRC for 1993 Exam
- D. Kanakares, Regulatory Assurance Staff
+R. Svaleson, Licensed Operator Instructor
+M. Swegle, Licensed Operator Instructor V. S. Nuclear Reaulatory Commission (NRC)
+*M. Jordan, Chief, Operator Licensing Section 1, RIII
+*T. Taylor, NRC Senior Resident Inspector
+*T. McKernon, RIV NRC, Reactor Inspector
+*C. Zelig, Chief Examiner, RIII NRC
+*K. Shembarger, RIII NRC
+*C. Tyner, Idaho National Engineering Laboratories
+ Denotes those present at the Training Staff Pre-exit Meeting on March 18, 1993.
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- Denotes those present at the Management Exit Meeting on March 18, 1993.
3.
License Trainino Proaram Observations The initial license training program appears to be functioning adequately.
Most candidates appeared well prepared for the examination and only one person failed, a SR01 candidate who failed the written examination.
A requalification program evaluation was not performed as part of the LSR0 requalification examination administered.
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l The following information from both the initial examinations and the LSR0 requalification examination is provided for evaluation by the i
licensee via their SAT based training program. With the exception of the concern addressed in the cover letter, no response is required.
a.
Written Examination Strenoths e
Several candidates did extremely well on this section of the examination.
1 Weaknesses e
Weaknesses in some operationally important areas were evidenced by more than 50 percent of the candidates missing questions in the following areas:
the expected plant response to an Electromagnetic Relief Valve vacuum breaker failing in the "open" position.
[2/4 R0 and 3/4 SRO candidates missed]
the effect of a loss of DC logic and control power on the
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HPCI system [all candidates missed this question].
when MSIVs should be manually closed during a total loss of Instrument A~.r. [3/4 R0 and 3/4 SR0 candidates missed]
the effect on primary containment integrity of operating the Traversing In-Core Probe (TIP) system shear valves. [4/4 SRO i
candidates missed]
Lack of knowledge in these areas could result in inappropriate l
l actions being taken during an operational event.
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b.
Job Performance Measures (JPMs) i No strengths or weaknesses noted in this area.
c.
Administrative Section of Operatina Test Strenoths: None noted.
j Weaknesses:
e Operational knowledge of radiation protection procedures was observed to be a weakness. Two R0 candidates had difficulty interpreting a standard symbol on a survey map at the entrance to l
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t the RCA. One R0 candidate picked up the probe of a portable frisker prior to surveying his hand and another did not know the l
correct probe speed to conduct a proper frisk.
Several R0 l
candidates did not know the background radiation levels at which a portable survey instrument was no longer operable.
In addition, i
t several candidates did not know the source of alpha contamination in the plant.
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e Some weaknesses were noted in the area of the station Emergency Plan. One SRO candidate did not know (without references) the responsibilities that the individual with " command and control" authority can not delegate. One R0 candidate did not know where to take a visitor he was escorting during a site assembly.
i d.
Dynamic Simulator Scenarios Strenaths:
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e Crew communications were clear and concise in almost all cases.
Proper use of repeat-backs was noted and the SR0s did a good job of keeping their crews informed of plant conditions.
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Weaknesses:
o Two candidates (1-RO, 1-SRO) had minor difficulty controlling
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vessel level within the specified band using the feedwater control i
valves.
In both cases, a reactor feed pump trip occurred when level exceeded +48 inches.
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Two R0 candidates were unfamiliar with the operation of the rod l
worth minimizer (RWM). Specifically, one candidate was unable to take a drifting rod out of service on the RWH and another i
candidate failed to bypass the RWM during an ATWS event.
l 4.
Trainina Department Activities a.
Initial License Examinations Strenaths e
The Training Department was very helpful during the prep week and during the examination itself. As a result of_their efforts, the j
Operations Section of the examination ran very smoothly, especially during the dynamic simulator scenarios.
e Quad Cities has recently written a new procedure to address the recently identified industry-wide problem of level indication inaccuracy during rapid vessel depressurizations.
During the 5
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dynamic simulator scenarios, all crews demonstrated proficiency in use of the new procedure. This is a good indication that operational events are incorporated into the Operator Training Program in a timely manner.
Weaknesses A three day pre-exam review of the initial license written examination was conducted by the facility. All facility comments were resolved by the NRC and all proposed revisions to the examination were reviewed by the Quad Cities Training Department before the NRC examination team left the site.
The pre-exam review is the preferred mechanism for identifying concerns with the NRC's written examinations. The licensee's post-exam comments identified 11 questions as being invalid or having more than one correct answer. The fact that Quad Cities' previous initial examination (see NRC Report No. 50-254/0L-92-01 dated April 20, 1992) also resulted in a large number of post-exam comments indicates a generic weakness in the training staff's ability to validate written examination questions during pre-exam reviews.
Of the 11 post-exam comments submitted by the facility, only 3 were evaluated by the NRC examination team as valid (see enclosure 2). This is further evidence that the Training Department has difficulty discriminating between valid and invalid questions.
A LSR0 license application was signed and submitted to the NRC prior to the candidate completing the LSR0 Training Program (see RP-93-008, R. Pleniewicz to M. Bies, letter dated March 5,1993).
This demonstrated a weakness in the licensee's ability to ensure all eligibility requirements are met prior to application submittal.
Region III has received Ceco's proposed long term corrective actions to this problem (see RP-93-010, R. Pleniewicz to M. Bies, letter dated March 30, 1993) and has no further comments.
b.
LSR0 Recualification Examination Strenaths: None noted.
Weaknesses:
Two questions on the LSR0 requalification written exsmination were deleted by the facility after the examination was administered.
One of the questions was invalid because of a new procedure promulgated after the examination was prepared and reviewed, but before it was administered. The other question was deleted 6
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because a wording mistake in the stem made three of the answer choices correct. The NRC examination team concurred with the deletions before leaving the site on 3/19/93 and with the follow-up written justifications provided by the facility (see JPH-93-006 and JPH-93-007, J. Hoeller to C. Zelig, letters dated 3/22/93 and 3/23/93).
l 5.
Generic Examiner Comments:
The plant appeared to be very clean during the in-plant portion of the Operating Examination, especially considering that an outage was in progress.
During the JPM (walk-through) Section of the LSR0 exam, it was e
discovered that station procedure QFP 150-17, Rev 1 " Setting the Main Hoist Position Indication System," did not contain an acceptance criteria for the value of the digital readout after cycling the hoist. The licensee accepted this c mment.
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Most of the valve control switches on the Safe Shutdown Makeup i
Pump (SSMP) control panels (local and in the control room) were configured in the opposite direction from the other valve control switches at the Quad Cities Station. Specifically, the SSMP valve l
control switches are turned clockwise to close a valve, and counterclockwise to open a valve. This could create a problem l
during an emergency when operators would be under stress to l
recover the plant. (This comment was inadvertently omitted during the Management Exit meeting but was passed to the licensee's training staff via a phone call the following week).
o.
Exam Materials The exam preparation materials provided were adequate. The enclosed QGA lesson plans and learning objectives were particularly helpful in preparing the written examinations. One problem with the exam material was the condition of the binders on arrival; several were broken open i
and had to be rebound by the NRC examiners.
7.
Plant SDecific Simulation facility The simulator performed consistently well. However, there were some minor simulator problems during the Operating Exam: During the JPMs, one candidate was unable to start a reactor building ventilation exhaust fan because building vacuum was greater than 1 inch water gage following the use of SBGT during the previous JPM. The cause of this unrealistically l
high vacuum was (as described by your staff). the over-estimation of I
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Reactor Building air-tightness by the simulator's computer model.
In addition, the Rod Worth Minimizer locked up during one of the dynamic scenarios. Neither of these events had a significant impact on the examination. provides NRC comments concerning simulator response during the examination.
8.
Exit Meetina An exit meeting was held in the Quad Cities Training Building on March 18, 1993. Those attending the meeting are listed in Section 2 of this report. The following items were discussed during the exit meeting:
Strengths and weaknesses noted in this report.
The general observations relating to the plant noted in Sections 4 l
and 5 of this report.
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L ENCLOSURE 2 l
NRC RESOLUTION OF FACILITY COMMENTS ON WRITTEN EKAMINATIONS l
OUESTION: SRO-5 l
Given the following plant conditions:
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--Unit 1 - 70% power, steady state conditions
--Unit 2 - 3% power, unit startup in progress
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Which of the following operating shift personnel are REQUIRED to be present at the controls in the Control Room for the above a
conditions?
(Assume no emergencies are present).
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Three (3) Senior Reactor Operators l
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b.
Two (2) Senior Reactor Operators l
i c.
Three (3) Reactor Operators I
d.
Two (2) Reactor Operators l
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ANSWER:
d.
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REFERENCE:
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Quad Cities Unit 1 Tech Spec 6.1.C, Page 6.1-1 and Figure 6.1-3.
l QAP 300-3, " Shift Manning," Rev. 14, Pages 1 & 2 I
NO-332, " Normal Operations," Rev.
1, L.O.
B.1.
l FACILITY COMMENT:
l QAP 300-3 STEP C.S. Requires an active RO licensed individual at the controls of each unit not defueled, making "D" correct.
Technical Specification Figure 6.1.3. requires a minimum of 3 1
licensed RO's in the control room, making "C" correct.
Note pre-l exam review comments were accepted, but further review by the l
facility indicates the stem does not contain wording sufficient l
to clearly determine the desired answer.
Accept "C" and "D" as correct answers.
FACILITY
REFERENCE:
Technical Specification Figure 6.1-3; QAP 300-3, Rev. 15 NRC RESOLUTION:
Comment not accepted.
Technical Specifications Figure 6.1.3 is the minimum shift manning chart, which indicates that a minimum of 3 licensed RO's are required to be on shift.
The question required the candidate to recognize the operating shift personnel required to be at the controls in the control room.
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. _ _ _., _ _ _ _ _ _ _...., _ _ _ _ _ _ _,.., _ __--, _.. _,.., - I
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OUESTION: RO-19 i
What conditions require the use of a second operator or qualified technical person in place of an inoperable Rod Worth Minimizer?
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a.
Power is above 20% with the reactor mode switch in "Run".
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l b.
Power is less than 30% and the Rod Block Monitor is i
t c.
The first group of 20 control rods have been withdrawn to
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position 12.
d.
The first group of 16 control rods have been withdrawn to i
position 48.
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i ANSWER:
d.
REFERENCE:
ILT 0207, " Rod Worth Minimizer," Rev.
O, page 48, L.O.
6& 15.
FACILITY COMMENT:
i QCGP 4-1, Control Rod Movements and Control Rod Sequence, requires a second verifier for all listed conditions except "C".
l Responses "A",
"B",
and "D" are correct.
Note pre-exam comments were accepted for this question, however further review by the facility indicates the comments were not adequate to fix the question.
i Delete question.
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FACILITY
REFERENCE:
j QCGP 4-1 Step B.2.,
page 1.
j NRC RESOLUTION:
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Comment accepted.
Question was deleted from the exam.
OUESTION: SRO-36
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Which of the following describes how use of the shorting links during refueling will affect'the Reactor Protection System (RPS)?
a.
Removal of the shorting links activates the SRM scrams in a coincident (one-out-of-two-twice) logic scheme.
b.
Removal of the shorting links activates the SRM,.IRM and APRM scrams in non-coincident logic schemes.
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i c.
Installation of the shorting links activates the SRM, IRM and APRM scrams in coincident logic schemes.
i d.
Installation of the shorting links activates the SRM scrams and bypasses the IRM and APRM scrams.
i ANSWER:
b.
REFERENCE:
i ILT 0500, " Reactor Protection System," Rev.
O, pages 62-64, L.O.
3a.
FACILITY COMMENT:
Removal of some but not all shorting links will activate the SRM scrams in a one-out-of-two-twice coincidence.
Procedure QCOP 500-2 details steps for aligning shorting links to establish such a condition.
i Accept "A" and "B" as correct answers.
FACILITY
REFERENCE:
I QCOP 500-2, step F.1.a.
i NRC RESOLUTION:
l Comment accepted.
"A" or "B" were accepted as correct answers.
OUESTION: RO-41/SRO-41 Given the following plant conditions:
Reactor Power 45%
Total Core Flow 35 E6 lb/hr APRM flow on channel 1 38%
APRM flow on channel 4 38%
FRP/MFLPD 1
For these conditions, what is the APRM Flow Bias Scram Setting?
(Choices are rounded to the nearest tenth.)
l a.
70.7%
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b.
72.5%
c.
84.5%
d.
82.7%
ANSWER:
c.
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REFERENCE:
ILT 0700-4, " Average Power Range Monitoring System," Rev.
O, page 34, L.O.
16.A.
38/98(100) = 38.8 X.58 = 22.5 + 62 = 84.5%
FACILITY COMMENT:
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An incorrect calculation is used to determine the correct answer.
Note pre-exam comments were accepted, but the calculation was not changed to reflect % of flow vice actual value of flow, which was provided in the original version of the question.
The correct answer is 38*.58 = 22.04+62=84.04 (or 84.0).
This answer is not within a tenth of any of the distractor answers.
Accept "C" and "D" as correct, as they are the closest to the l
actual value.
FACILITY
REFERENCE:
ILT 700-4 pg. 34 of 47 l
NRC RESOLUTION:
Comment partially accepted.
Since the correct answer to the l
question, rounded to the nearest tenth, was not provided to the
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candidate, the questions was deleted from the exam.
OUESTION: RO-45/SRO-46 l
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How is a loss of main condenser vacuum prevented when the Unit 1 Reactor Water Cleanup (RWCU) system is lined up for coolant i
reject operations?
i a.
Administrative control on the operation of the " Reject To l
Condenser Shutoff Valve" (1201-78) and the " Reject to WCT Shutoff Valve" (1201-77).
b.
Interlocks preventing the simultaneous opening of the
" Reject To Condenser Shutoff Valve" (1201-78) and the
" Reject to WCT Shutoff Valve (1201-77).
c.
The RWCU system Non-Regenerative Heat Exchanger high outlet j
temperature isolation.
d.
The automatic closing features of the RWCU Reject Flow Control Valve (1201-1239).
ANSWER:
a.
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REFERENCE:
QCOP 1200-7, "RWCU System Coolant Rejection," Rev.
O, Pages 3 & 4.
ILT 1200, " Reactor Water Cleanup System, "Rev.
1, L.O.
6.G.
FACILITY COMMENT:
The RWCU reject flow control valve auto closes on low inlet pressure (5 psig), in order to prevent depressurizing the RWCU system.
Since the RWCU may be rejecting to the condenser, such draining could lead to condenser vacuum problems.
Accept "A" and "D" as correct answers.
FACILITY
REFERENCE:
ILT 1200, RWCU Lesson Plan, page 14 of 53.
NRC RESOLUTION:
Comment not accepted.
Although the RWCU reject flow control valve closes on low inlet pressure, valve closure occurs to prevent depressurizing the RWCU system, not to prevent a loss of main condenser vacuum when the RWCU system is lined up for coolant reject operations.
OUESTION: RO-56/ BRO-55 Given the following plant conditions on Unit 2:
High Pressure Coolant (HPCI) system is running DC po>er to the HPCI logic and controls has just been lost Reserve power is not available 1
SELECT the expected HPCI system response to these conditions.
a.
The HPCI suction will swap from the Containment Condensate Storage Tank to the Torus, b.
HPCI will begin to accelerate until controlled by the Motor Gear Unit high speed stop.
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The HPCI turbine will trip and the AC powered isolation valve (MO 4) will close.
d.
HPCI will decelerate until controlled by the Motor i
Speed Changer low speed stop.
I ANSWER:
a.
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REFERENCE:
QCAN 901(2)-3 G-12, "HPCI Control Power Failure,"
Rev.
O, page 1.
ILT 2300, "High Pressure Coolant Injection," Rev.
1, L.O.
6.k.5 & 6.k.6.
FACILITY COMMENT:
The stated facility objective does not cover this question.
Objectives 6.k.5 and 6.k.6 are for the operator to " state the purpose and/or explain the operation of the suppression pool and CCST suction valves" and the material does not include response to the particular failure.
The only place the particular failure mode is described, is in the annunciator procedures; operators are not required to memorize the annunciator procedures.
FACILITY
REFERENCE:
ILT 2300, HPCI Lesson Plan, TKO 6.k.5. and 6.k.6.,
page 38 of 107.
j NRC RESOLUTION:
Comment not accepted.
The K/A applied to this question was l
206000K602 (3.3/3.7), which states " Knowledge of the effect that
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a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM:
D.C.
power: BWR-2,3,4".
The annunciator response procedure associated with annunciator "HPCI Control Power Failure" was used as the reference for the question.
As indicated in the facility comment, the lesson plan for the HPCI system fails to address the affects of a loss of DC power to the HPCI system.
An inadequate lesson plan 1) does not restrict the NRC from examining in given area and 2) does not eliminate the need for licensed operators to possess the know'. edge and abilities necessary to operate the facility.
l OUESTION: R0-58/SRO-57 Both trains of the Standby Gas Treatment System are inoperable.
l How does this affect ooth units?
a.
An immediate shutdown of both units is required.
b.
Both units must be shutdown within seven (7) days.
c.
Primary Containment Integrity is not longer assured.
d.
Secondary Containment Integrity is no longer assured.
ANSWER:
d.
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REFERENCE:
ILT 1600-1, " Containment Systems," Rev.
1, page 82, L.O.
15.c.
Quad Cities Tech Spec definition 1.0 X, page 1.0-4 FACILITY COMMENT:
"A" is a correct answer.
Technical Specification 3.7.C.3.
requirec " action initiated to establish the conditions" for which secondary containment integrity is not required, i.e. both Units must be immediately shutdown.
Note pre-exam comments were for the stem to read "...how does this affect plant conditions?".
Accept "A" and "D" as correct answers.
FACILITY
REFERENCE:
Technical Specification 3.7.C NRC RESOLUTION:
Comment not accepted.
Technical Specification 3.6.B.b.
states:
If both standby gas treatment system circuits are not operable, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> the reactor shall be placed in a condition for which the standby gas treatment system is not required in accordance with Specification 3.7.C.1.(a) through (d).
The plant condition in which the standby gas treatment is not required as stated in Technical Specification 3.7.C.1.(a) through (d) is:
The reactors are subcritical and Specification 3.3.A is a.
met.
b.
The reactor water temperature is below 212 degrees F and the reactor coolant systems are vented.
No activity is being performed which can reduce the c.
shutdown margin below that specified in Specification 3.3.A.
d.
The fuel cask or irradiated fuel is not being moved in the reactor building.
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i Technical Specification Definition D.
states:
Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
Based on the Quad Cities Technical Specifications, with both trains of the standby gas treatment system inoperable, action to bring both units to a cold shutdown condition is not required to be initiated as soon as practicable.
Specifically, the operators have 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to bring the units to a cold shutdown condition.
Tech Spec 3.7.C.1 requires procedures to be initiated to establish cold shutdown if secondary containment can not be maintained.
However, Tech Spec 3.7.B.1 states two separate and independent SBGT system circuits shall be operable when secondary containment integrity is required, except as specified in sections 3.7.B.1.(a) and (b).
Thus, the case where procedurec need to be initiated to bring the reactor to cold shut down (T.S.
3.7.C.1) is distinguished from the case in which they do not l
(T.S. 3.7.B.1).
OUESTION: RO-71/SRO-65 Immediately upon entry into QGA 101, "RPV Control (ATWS)", the operator is directed to inhibit the Automatic Depressurization System ( ADS).
IDENTIFY the justification for this action.
a.
Uncontrolled depressurization will put a considerable amount of energy into the suppression pool well before it is necessary or required.
b.
Depressurization would drive plant conditions above the RPV Saturation Temperature curve where reactor water level instrumentation may not be usable.
c.
Uncontrolled depressurization under these conditions would cause a large loss of RPV inventory lowering water level below the top of active fuel (TAF).
I d.
Once depressurized below the shutoff head of the low pressure injection systems, the relatively cold water may cause large positive reactivity additions.
ANSWER:
d.
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REFERENCE:
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QGA 101, "RPV Control (ATWS)" Flowchart, Rev. 2 QGA 101, "RPV Control (ATWS)", Rev.
5, Page 6, L.O.
3 FACILITY COMMENT:
One basis for inhibiting ADS is to prevent a transient on the vessel which could complicate level control actions.
Accept "A" and "D" as correct answers.
EACILITY
REFERENCE:
QGA 101 Lesson Plan, page 6 of 63.
Emergency Procedure Guidelines page B-6-38.
NRC RESOLUTION:
Comment not accepted.
Although one basis for inhibiting ADS is to prevent a transient on the vessel which could complicate level control actions, distractor a. does not address this basis.
OUESTION:
SRO-66 Given the following conditions for Unit 1:
A scram condition exists but the reactor did not shutdown Reactor power is 45%
The Main Steam Isolation Valves are closed Core cooling is adequate IDENTIFY the component (s) being most severely challenged for these conditions.
a.
l b.
c.
HPCI system piping d.
Reactor water level instrumentation ANSWER:
a.
REFERENCE:
QGA 101, "RPV Control (ATWS)", Rev.
5, page 40, L.O.
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FACILITY COMMENT:
For the listed condition, reactor pressure would be elevated above the relief setpoint, since the stated power is greater than the capacity of the ADS valves.
Elevated pressure places the fuel closer to the MCPR thermal limit, thereby endangering the fuel cladding.
There is not enough information to eliminate "B" i
as a correct answer.
Accept "A" and "B" as correct answers.
FACILITY
REFERENCE:
ILT HTFF Chapter IX, Thermal Limits, page 29 and 35 of 56.
NRC RESOLUTION:
Comment not accepted.
One of the conditions established in the stem of the question was " core cooling is adequate".
Therefore, MCPR is not a concern.
OUESTION: RO-78/SRO-75 QGA 200, " Primary Containment Control", Torus Level Control leg, directs the operator to " prevent HPCI operation even if core cooling will be lost" if torus level cannot be held above 11 feet.
Which of the following describes the effect of continued operation of the High Pressure Coolant Injection (HPCI) system?
a.
will result in the failure and loss of the final fission product barrier.
b.
will reduce suppression pool level below the safety relief valve T-quenchers.
c.
will cause HPCI turbine damage due to the high temperature cooling water to the oil system, d.
results in a loss of the last high pressure injection system l
on high exhaust backpressure.
l ANSWER:
a.
REFERENCE:
QGA 200, " Primary Containment Control", Rev.
3, page 42, L.O.
22 10 l
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FACILITY COMMENT:
1 Quad Cities teaches the fission product barriers to be, in order:
fuel pellet, clad, primary system, primary containment, secondary containment.
Since the " final fission product barrier" may be assumed to be the reactor building by the candidate, the next logical choice would be "D" since high exhaust pressure will trip the HPCI turbine.
Accept "A" and "D" as correct answers.
FACILITY
REFERENCE:
ILT 1600-1, Containment Lesson Plan, page 8 of 103.
NRC RESOLUTION:
Comment not accepted.
The question asked for the effect of.
continued operation of the HPCI system when executing QGA 200,
" Primary Containment Control".
From the QGA 200 Lesson Plan :
"When a decision must be made between maintaining core cooling and assuring primary containment integrity, the QGAs preferentially choose to protect against the uncontrolled release of radioactivity to the general public."
An uncontrolled release of radioactivity to the ceneral oublic could not occur unless secondary containment integrity was also lost.
Therefore, if HPCI continued to operate in this condition, the final fission product barrier (i.e. Secondary Containment) would be lost.
HPCI operation is not prevented to prevent a trip of the HPCI turbine.
OUESTION: SRO-98 SELECT the reason for turbine bypass valve automatic response following a main turbine trip from 100% power.
l l
a.
Main steam line pressure exceeds controlling EHC pressure regulator setpoint.
b.
Controlling EHC pressure regulator pressure setpoint exceeds turbine throttle pressure.
c.
Controlling EHC pressure regulator setpoint exceeds reactor pressure.
d.
Turbine first stage pressure exceeds controlling EHC pressure regulator setpoint.
ANSWER:
a.
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REFERENCE:
i 1
ILT 5650-2, "EHC Logic System", Rev.
O, pages 10-14, L.O.
12.d.
FACILITY COMMENT:
l If turbine throttle pressure drops below 920 psig, (EHC setpoint) due to cold water injection or heat loss to ambient, the bypass i
valves will close.
Note pre-exam comments were accepted, but further review by the facility indicates there is not enough l
information in the stem to eliminate "B" as the correct answer.
j Accept "A" and "B" as correct answers.
i FACILITY
REFERENCE:
ILT 5650-2, EHC Logic Lesson Plan, page 66 and 68 of 79.
NRC RESOLUTION:
Comment not accepted.
The question asked the candidate to recognize the reason for turbine bypass valve automatic response following a main turbine trip from 100% power.
Following a main turbine trip from 100% power, the bypass valves open as main steam line pressure rapidly increases above EHC pressure regulator pressure because of the loss of the main turbine (the primary heat load).
This effect is the turbine bypass valve automatic response to the turbine trip.
While it is true that the bypass valves will eventually close, it is not in response to the turbine trip from 100% power.
Rather, the bypass valves close in response to the cooldown of the primary (via the open bypass valves) which lowers turbine throttle pressure below EHC pressure regulator pressure.
OUESTION: LSRO-43 The following conditions exist:
- A fuel assembly is being moved from the Fuel Pool to the reactor vessel.
- The assembly is located in the spent fuel storage pool.
l
- The Fuel Pool Storage Low Level Alarm is received.
- Fuel Pool Level is confirmed to be decreasing.
- The Refuel Floor ARM is NOT alarming.
The required fuel handling action is to:
a.
stop all fuel movement and evacuate the reactor building.
I b.
complete the move to the specified core location.
12
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c.
return the assembly to the nearest spent fuel location.
d.
place the assembly in the nearest core location.
ANSWER:
d.
REFERENCE:
1 1
(
QOA 1900-1, p.2, item C.2.
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I l
FACILITY COMMENT i
"C" is the correct answer to this question.
The stem was l
reworded per pre-exam comments, but the key was not changed to reflect the correct answer of "C" vice "D".
l Accept "C" as the only correct answer.
)
1 FACILITY
REFERENCE:
QOA 1900-1, Step C.2.,
p.2 QCFHP 110-5, Step C.4.a.,
p.1 l
NRC RESOLUTION:
l Comment accepted.
"C" accepted as the only correct answer.
l OUESTION: LSRO-46 l
If the reactor has been defueled, which one of the following may be moved inside the core area (above the core support plate, below the upper grid, and within the shroud) without a Nuclear Component Transfer List?
a.
Fuel Loading Chamber (Dunker) b.
Fuel Support Piece c.
Blade Guide d.
Control Blade ANSWER:
c.
REFERENCE:
QFP 100-1, p.10, item 12.
13
. - a FACILITY COMMENT:
If a control rod is moved by the CRD system (normal rod motion) then the NCTL is not required.
The question does not specify movement by fuel handling equipment.
Accept "C" and "D" as correct answers.
FACILITY
REFERENCE:
Technical Specifications definition 1.0.A NRC RESOLUTION:
Comment not accepted.
While Tech Specs do not define movement of control rods with the normal CRD system as an " Alteration of the Reactor Core", QFP 100-1, p.10, item 12 does not distinguish between control rod movements that are core alterations and those that are not.
IAW QFP 100-1, p.10, item 12, all control rod movements performed as part of refueling have to be logged on a NCTL.
OUESTION: LSRO-53 Given the following conditions:
- You are performing a whole body " frisk" using a portable frisker (HP-210).
- Background radiation count rate in the area is 300 cpm.
Under these conditions, select the MINIMUM count rate at which you are considered to be " contaminated?"
a.
100 counts per minute b.
200 counts per minute c.
300 counts per minute d.
400 counts per minute ANSWER:
c.
REFERENCE:
QRP 1470-1, " Routine Personnel Decontamination", Rev.
3, p.2.
FACILITY COMMENT:
The key is equal to the given background radiation level.
The correct answer to this question should be "D".
Accept "D" as the only correct answer.
14
J FACILITY
REFERENCE:
QRP 1470-1, p.2.
NRC RESOLUTION:
Comment partially accepted.
"D" is the correct answer to the question.
However, since only one of the four choices was plausible, the question was invalid and was therefore deleted.
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e ENCLOSURE 4 SIMULATION FACILITY REPORT Facility Licensee: Commonwealth Edison Company, Quad Cities Nuclear Power Etation Facility Licensee Docket Nos.
50-254; 50-265 Operating Tests Administered On: March 15-18, 1993 During the conduct of the simulator portion of the operating tests, the simulator was noted to work well. One simulator infidelity was noted:
During the JPMs, one candidate was unable to start a reactor building ventilation exhaust fan because building vacuum was greater than 1 inch of water following the use of SBGT. The cause of this unrealistically high vacuum was over-estimation of Reactor Building air-tightness by the simulator's computer model. Another minor problem occurred during one of the dynamic scenarios when the Rod Worth Minimizer locked up. Neither of these events had a significant impact on the examination.
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ENCLOSURE 5 RE00ALIFICATION PROGRAM EVALUATION REPORT (LSR0 ONLY)
Facility:
Quad Cities Nuclear Power Station Examiner:
Craig Zelig Dates of Evaluation:
March 4, 19, 1993 Areas Evaluated: X Written X
JPMs Examination Results:
R0 SRO Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S or U)
Written Examination N/A 1/0 1/0 S
JPMs (walkthrough)
N/A 1/0 1/0 S
Simulator N/A N/A N/A N/A Evaluation of facility written examination grading S
Crew Examination Results: N/A Overall Program Evaluation: N/A
- NOTE: Due to the small sampling size no training program review was performed.
Submitted:
Forwarded:
Approved:
.h (14 C. I(M g f
M. Jfr/an M. iring 1
Examiner Section Chief Branch Chief 04/j2 /93 04//2./93 04/%/93
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