IR 05000369/1999301
| ML20205H930 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 03/18/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20205H915 | List: |
| References | |
| 50-369-99-301, 50-370-99-301, NUDOCS 9904090121 | |
| Download: ML20205H930 (246) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 11 Docket Nos.: 50-369,50-370 License Nos.: NPF-9, NPF-17
Repod Nos.: 50-369/99-301,50-370/99-301
Licensee: Duke Energy Corporation l
l Facility: McGuire Nuclear Station, Units 1 & 2 i Location: 12700 Hagers Ferry Road Huntersville, NC 28078 -
i Dates: January 25,1999 - February 5,1999 Examiners: D. Charles Payne, Chief License Examiner Edwin Lea, License Examiner Paul M :t'.:ner, Li' ,nse Examiner
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Approved ;;y: Harold O. Christensen, Chief j Operator Licensing and Human Performance Branch Division of Reactor Safet ,
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Enclosure 1 9904090121 990324 ?
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Q ( EXECUTIVE SUMMARY McGuire Nuclear Station, Units 1 & 2 NRC Examination Report Nos. 50-369/99-301,50-370/99-301 During the period January 25 - February 5,1999, NRC examiners conducted an announced operator licensing initial examination in accordance with the guidance of Examiner Standards, NUREG-1021, Interim Revision 8. This examination implemented the operator licensing requirements of 10 CFR 955.41, $55.43, and $55.45.
L Five senior reactor operator candidates and six reactor operator candidates received written l examinations and operating tests. The NRC administered the operating tests during the weeks
, of January 25,1999, and February 1,1999. The licensee administered the written examination l on February 5,1999.
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Ooerations
In general, the examiners found that the as-submitted written examination and operating tests met the requirements of NUREG-1021 and improved on the 1997 examination !
l submittal. The approved written examiriation questions were noted to be excellent test I
items for measuring candidate understanding of systems and administrative knowledg (Section O5.1)-
- Nine of eleven candidates passed the examination. Seven individuals were identified as having weaknesses on the: itten examination and/or the operating test. Six candidates were k atifiea as exhibiting b Tvieds ) weaknesses on the written exarnination and three demonstrated performa. / *. ekoc a= =Nr?" the operating test. (Section O5.1)
Fail 0 2 l 2 18%
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Report Details Summarv of Plant Status i
During the period of the examinations both Unit 1 and Unit 2 were at 100 percent powe l. Operations l
05 Operator Training and Qualifications 0 initial Licensino ExaminNions Sco_gg NRC examiners conducted regular, announced operator licensing initial examinations during the peiiod January 25 - February 5,1999. The examiners administered i examinations developed by members of the McGuire training staff under the !
requirements of an NRC security agreement, in accordance with the guidelines of the Examiner Standards (ES), NUREG-1021, Interim Revision 8. This examination
represented the licensee's second effort at developing written examinations and '
operating tests for the NRC's operator licensing process. Five Senior Reactor Operator ,
(SRO) and six Reactor Operator (RO) license applicants received written examinations l and operating test l
! Observations and Findinas I The facility examination developers submitted 126 multiple choice questions for NRC examiner review. The RO and SRO written examinations were designed to share 74 questions with 26 additional questions for each test that were license level specific. The examiners had few comments where substance was a problem in the question; most comments were to assure clarity in the question stem and to enhance the quality of the incorrect answers. The approved written examination questions were noted to be excellent test items for measuring candidate understanding of systems and i administrative knowledg ;
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The facility examination developers submitted three simulator scenarios and one spare for NRC review. The examiners found the simulator tests followed the guidelint.; of NUREG-1021. The malfunctions were logically sequenced to lead to the major plant ;
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transient and served as a valid measurement tool of candidate abilities. Minor changes were made to enhance the test items of each scenari The walkthrough examinations sets submitted by the facility examination developers contained job performance measures (JPMs) that met the guidelines of the ES and were j of the appropriate level of difficulty. Minor JPM content additions were made to improve j their usability by the examiners. The JPM follow-up questions were good; they were challenging, operationally-oriented tests of candidate knowledge and understanding of {
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The operating tests were validated during the week of January 11-15,1999, arm approved on January 21,1999. The written examinations were finalized and approved the week of February 1,1999, Examination Results and Conclusions in general, the examiners concluded that the quality and level of difficulty of the licensee's draft examination submittal was very good and represented an improving trend since the licensee's development efforts from the last examination in 199 Licensee staff response to examiner changes and comments was prompt and addressed the deficiencie's. This effort directly led to the examination being administered on time and as scheduled. Nine of eleven candidates passed the examinatio One RO candidate failed the written examination. The facility licensee did not submit any post-examination comments. The examiners determined that five other candidates exhibited knowledge weaknesses on the written examination. The licensee conducted a post-examination grading item analysis of both written examinations. This analysis identified five questions wh;te both SRO and RO candidates exhibited knowledge deficiencies. The analysis also identified five additional questions wher" only the SRO candidates exhibited knowledge weaknesses and eleven additional questions where only the RO candidates exhibited knowledge weaknesses. The examiners concluded that no generic knowledge weaknesses existed where rnultiple questions on the same system or topic were missed by a large number of candidates. However, of the 21 questions with high miss rates, five were associated with administrative requirement One RO candidate failed the simulator portion of the operating test and also exhibited performance weakness on the administrative portion of the operating test. The examiners determined that two other candidates exhibited performance deficiencies on the administrative portion of the operating test as well. The examiners concluded there were no significant candidate performance weaknesses during the simulator or plant walkthrough examination Detailed candidate performance comments were transmitted under separate cover for management review and to allow appropriate candidate remediatio Miscellaneous Operations issues 0 (Closed) IFl 50-369.370/98-09-01: Tracking of Corrective Actions for Procedure Weaknesses in Two ARPs for Rod Insertion Limits and Manual Estimated Critical Boron Calculation The examiner reviewed the licensee's corrective actions for the three above procedure weaknesses and found the implemented procedure changes to be adequate. This IFl is closed.
08.2 (Closed) IFl 50-369.370/98 09-02: Tracking of Corrective Actions for Licensed Operators' inability to Perform a Manual Estimated Critical Boron Calculation l
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The examiner reviewed the licensee's corrective actions for the licensed operator's skill !
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weakness described above. The licensee's review and correction of the problem was l adequate. Alllicensed operators were adequately retrained and evaluated. This IFl is {
close i
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V. Manacement Meetinas t-X1 Exit Meeting Summary An exit interview was conducted on February 5,1999, to reiterate the purpose of the site visit and to discuss the findings. The licensee had no comments and the examiner received no dissenting comments. No proprietary information was receive l l
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PARTIAL LIST OF PERSONS CONTACTED l l
Licensee j
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L. Abernathy, Regulatory Complianc * T. Arlow, Operations
- A. Bhatnagar, Operations
- J. Bryant, Regulatory Compliance
- M. Cash, Regulatory Compliance
- B. Dolan, Safety Assurance M. Geddie, Station Manager l
- A. Orton, Operations Training l
- B. Peele, Engineering l
- R. Pope, Operations Training I
- C. Sawyer, Operations
- R. White, Training l i
NRC l
- M. Franovich, Resident inspector
- S. Shaeffer, Senior Resident inspector M. Sykes, Resident inspector
- Attended Exit interview ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Closed 50-369,370/98-09-01 IFl Tracking of Corrective Actions for Procedure Weaknesses in Two ARPs for Rod Insertion Limits and the Manual Estimated Critical Boron Calculation 50-369,370/98-09-02 IFl Tracki.ng of Corrective Actions for Licensed Operators' Inability to Perform a Manual Estimated Critical Boron Calculation Discussed None l
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LIST OF ACRONYMS USED
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CFR Code of Federal Regulations ES Examiner Standards (NUREG-1021) ,
JPM Jcb Performance Measure NRC Nuclear Regulatory Commission I
'RO Reactor Operator SRO Senior Reactor Operator IFl Inspr c ollow-up item i
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SIMULATION FAC!LITY REFC.AT l
l Facility Licensee: McGuire Nuclear Station 1 Facility Docket Nos.: 50-369 and 50-370 g
Operating Tests Administered on: January 25 - February 4,1999 This form is to be used only to report observations. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tests, the following items were observed (if none, so state):
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ITEM DESCRIPTION None
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i Enclosure 2
i NRC Official Use Only
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Mas fa n n- 3 er
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Nuclear Regulatory Commission l Reactor Operator Licensing Examination )
Final Version for Approval McGuire Nuclear Station Questions with Answers This document is removed from Official Use Only category on l
date of exmnination ,
I NRC Official Use Only DISTRIBUTION CODE IE42
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McGuira Writtin Exam i Simpin PI:n I
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I Facility: McGuire Date of Exam: 2/5/99 Exam Level: RO I K/A Category Points Point A A A A Tier Group l 1 2 314 5 6 1 2 3 4 Total Target 1 1 1 3 4 5 2 1 16 16
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Emergency & 2 4 2 4 2 2 3 17 17
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Abnonnal 3 1 0 1 0 0 1 3 3
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Plant Tier Evolutions Totals 6 5 9 7 4 5 36 36 um 1 1 1 3 1 2 2 4 4 1 3 1 23 23 2 2 0 1 4 l1 l2l2l2 5l2 l0l1 20 20 Plant 3 1 2 1 l1 l0l0l0 2l0l1 l0 8 8 Systems Tier l Totals 2 4 8 3l4 4 6 11 l 3 l 4 l2' 51 51 3 Generic Knowledge and l Ca' 1 l Cat 2 Cat 3 l Cat 4 ;
Abilities l 3 l 4 3 l 3 13 13 l Note: * Attempt to distribute topics among all KIA categories; solect Totals:
at least one topic from every K/A category within each tie * Actual point totals must match those specified in the tabl * Select topics from many systems; avoid selecting more then two or three K/A topics from a system unless they relate to plant-specific prioritie * Systems / evolutions within each group are identified on the associated outlin * The shaded areas are not applicable to the category / tie " Denotes plant specific, high priority K/As
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Mc2uire Written Ex*m Sample Plan Eb40f Generic K and Abilities Outline (Tier 3) Form ES-4015 Facility: McGuire Date of Exam: Exam Level: RO Bank Category N/A 8 Topic Im Pointa Question 2.1.2g Knowledge of how to conduct and verify valve lineups 3.4/ Ability to recognize indecations for system operating parameters which are entry 2.1.33 level conditions for technical specifications 3.4/ Conduct of 2.1.20 Ability to execute procedure steps 4.3/ Operatione Total 3 2. Knowledge of design, procedural er d operational differences between units 3.1/ Ability to ananipulate console controls as required to operate facility between 2. shutdown and designated power levels 4.0/ Equipment 2.2.13 Knowledge of tagging and clearance procedures 3.6/ Control 2. Knowledge of refueling process 2.6/ Total 4 Knowledge of radiation exposure hmits and contamination control, including 2. permissible levels in excess of those authorized 2.5/ .3.11 Ability to control radiation releases 2.7/ Radiation 2. Knowledge of process for performing a containment purge 2.5/ Control Total 3 Knowledge of EUP implementation hierarchy and coordination with other support 2.4.18 procedures 3.0/ Knowledge of local operator auxiliary tasks dunng emergency operations including 2.4.35 geography and system implications 3.3/ Emergency 2.4.48 Ability to verify that alarms are consistent with plant conditions 3.5/ Procedures and Plan Total 3 Tier 3 Point Total 13 13l
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ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet j l
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U.S. Nuclear Regulatory Commission ,
Site-Specific ]
Written Examination Applicant Information Name: Region: I /(j'I)/
I III / IV Date: Facility / Unit: Mccuire License Level: h / SRO Reactor Type: (E)/CE/BW/GE Start Time: Finish Time:
Instructions Use the answer sheets provided to document your answer Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination start Applicant Certification All work done on this examination is my ow I have neither given nor received ai Applicant's Signature Results Examination Value Points l
l Applicant's Score Points Applicant's Grade Percent I
NUREG-1021 39 of 39 Interim Rev. 8, January 1997 J
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Question #1 McGuire Nuclear Station RO Exam l
Bank Question: 366 Answer: D 1 Pt(s) Unit I was operating at 100% power with control rods in automatic when the Reactor Operator observed the following indications:
Individual Rod Position Indication for control bank D rods:
M-12 = 220 steps D-4 = 220 steps M-4 = 223 steps D-12 = 210 steps H-8 = 223 steps Control Group D Step counter = 221 steps Which of the follow.ng sets ofimmediate actions (if any) must be taken? No operator action is required The system will maintain automatic control Adjust turbine load to maintain Tave = Tref Monitor AFD and QPTR Piace control rods in manual and position to maintain Tave = Tref Monitor AFD and QPTR Piace control rods in manual Adjust turbine load to maintain Tave = Tref
.
Distracter Analysis: Incorrect: - rod position for CD3 and CD-4 are > 12 steps different Plausible: -if the operator is unable to diagnose from memory the fact that rod CD-4 is out of alignment > 12 steps from any other ro Incorrect: - rods are > 12 steps in position difference Plausible: - if candidate does not remember that rods must be placed in manual Incorrect: - rods are > 12 steps in position difference Plausible: - if candidate does not remember that turbine power is adjusted to maintain Tave and Tref Correct answer For Official Use Only Ques _366 Page1
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Question #2 McGuire Nuclear Station RO Exam Bank Question: 367 Answer: B 1 Pt(s) Which NC pump, in E3-0.2, is the " priority pump" for restarting under natural circulation conditions and what is the EOP basis for this priority? NC Pump 2A - this pump is connected to the pressurizer spray line and restarting will provide the best pressure control NC Pump 2B - this pump is connected to the pressurizer spray line and resbrting will provide the best pressure control NC pump 2A - this pump is connected to the pressurizer surge line and restarting will prevent thermal shock to the PZR spray nozzle
< NC pump 2B - this pump is connected to the pressurizer surge line and restarting will prevent thermal shock to the PZR spray nozzle Distracter Analysis: Incorrect: - 2A does not provide the best spray control Plausible: - because 2A is the next highest priothy pump to be used in ES-0 Correct answer Incorrect: - The surge line will not provide any pressure control Plausible: - if the candidate confuses preventing thermal shock to the spray nozzle is the EOP basis (for a different step) Incorrect: - 2D is not connected to either the surge line, nor the spray line Plausible:-if the candidate, es not remember physical connections and thinks that the concem is thermal shock to the spray nozzle (as in some other EOPs).
..
For Official Use Only Ques _367 Page 2
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Question #3 McGuire Nuclear Station RO Exam Bank Question: 368 Answer: C j 1 Pt(s) Unit I was responding to an ATWS event. Given the following conditions:
. NV pump 1 A is out of service
INV-244A and INV-245B (Charging Line Cont Outside Isol) are open
- INV-221 A (NV Pumps Suct From FWST) was tagged shut
. 1NI-9A and 1NI-10B (NC Cold Leg Inj from NV) are shut
. 1NV-265B (Boric Acid to NV Pumps) is shut Which one of the following choices (if any) would provide adequate emergency boration flow in step 5 of FRS.1, Response to Nuclear Power Generation / ATWS? Start NV pump 1B Open INI-9A and INI-10B Close INV-244A and INV-245B Start NV pump 1B Open INV-265B Start NV pump 1B Open INV-265B Start both boric acid transfer pumps .Not able to provide adequate horation flow without INV-221 A being open Distracter Analysis: Incorrect:- the flow path is not aligned with INV-265B shu Plausible:-if the candidate does not recognize that 1NV-265B is part of the flow pat Incorrect: - boric acid transfer pumps required to provide sufficient boric acid concentration with NV pump suction aligned to the VCT (normallineup)
Plausible: - if the candidate does not recognize that the BA pumps are needed or thinks that normal line up is the FWST.
l Correct answer
- Incorrect: - suction is normally aligned to the VCT - do not need FWST for normalline up.
.
Plausible: - if the candidate thinks that the FWST is the nonnal suction path lineup for the NV pump.
,
For Official Use Only Ques _368 Page 2 l
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Question #4 McGuire Nuclear Station RO Exam Bank Question: 369 Answer: A .
1 Pt(s) What is the nonnal source of makeup water to the KC system if surge tank level is low? What is the attemate flow path for draining from the KC system if surge tank level is high with the KC drain tank inoperable? makeup from YM drain to LIIST sink makeup from RN drain to NCDT makeup from YC drain to WZ sump makeup from NM drain to WEFT
_
Distracter Analysis: Correct answer: YM is the demineralized water tank, LHST is qualified to accept KC water with corrosion inhibitor Incorrect: - YM and the LHST sink are the source / drain for the KC syste Plausible: - if the candidate believes that the KC system woudi use RN water as an altemate source - ahvays available - frequently used as an altemate source foi many other system Incorrect: - YC is make up water for the chilled water system Plausible: - if the candidate believes that non-nuclear grade makeup water is required for the KC system, this is a potential source. The KC system can be lined up to feed and bleed off the WZ sump in OP6400/005 enci 4.11 so it can drain to this sum I morreu. - WM is the nuclear sample system
. sausible: - the WEFT is where the NM system drains to for non-reactor grade water - the KC system provides cooling flow for the NM system but not makeup.
For Official Use Only Ques _369 Page 2
Question #5 McGuire NuclearStation RO Exam ;
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Bank Question: 370 Answer: B i 1 Pt(s) Unit 2 was operating at 100% power with Pressurizer pressure channels in a normal lineup and NC pressure in the nonnal operating band when an instrument failure occurred. Given the following plant conditions and events:
- PORVs NC-32 and NC-36, pen causing NC pressure to decrease
. PORVs close at 2185 psig e Backup heaters energize and pressure increases e PORVs reopen at 2189 psig
. A reactor trip does not occur
. This cycle continues until operator action is taken Assuming no operator action to realign (or swap) instrument channel lineups, which one of the following events would cause these actions? PZR pressure channelI fails high PZR pressure channelII falls high P7R pressure channelIII fails high PZR pressure channelIV falls high Distracter Analysis: Incorrect: - channel I does not control PORVs 32 and 36 Plausible: - plant response is very similar except PORV 34 opens and pressure continues to decrease below 2185 until a plant trip occurs due to spray valves being full open Correct answer - PZR channel III is the interlock channel to channel 11 Incorrect: - PORVs do rot open - interlock channel only Plausible: - if candidates do not recall that this is an interlock channel or think that it controls the interlocked PORVs Incorrect: - PORVs do not open - interlock channel only Plausible: - if candidates do not recall that this is an interlock channel or think that it controls the interlocked PORVs For Official Use Only Ques _370 Page 2
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l Question #6 McGuire Nuclear Station RO Exam Bank Question: 371 Answer: C 1 Pt(s) In the event of a steamline rupture that cannot be isolated, FR-P.2, Response to Anticipated Pressurized Thennal Shock Condition, could be implemented due to an excessive cooldown. What is the limiting component for this PTS event and what is the best indication of the temperature of this component?
Limitine Component Best Indication steam generator tube sheet pressurizer spray nozzle Tw-T , reactor vessel wall Tm i reactor vessel downcomer T,,, !
Distracter Analysis: Most candidates will have grasped the bases for FR-P.1 but FR-P.2 is less well known. Ilowever, the basic j principles regarding NDT and PTS are the same for this l
questio ' Incorrect: - the steam generator tube sheet is not the limiting component for a steam leak event Plausible: - if the candidate corresponds the steam leak to the steam generator and thinks that this is limiting due to the large pressure difference across the steam generator from the rupture (PTS) Incorrect: - the pressurizer spray nozzle is not limiting for this event as there is no spray flow l
Plausible: - if the candidate confuses this event with other events where the pressurizer delta Temp is limiting - or does not recognize l that NC pumps have been stopped l Correct answer Incorrect: - the reactor vessel downcomer is not generally limiting Plausible: - if the candidate does not recognize that Thot is not the most limiting temperature. The Rx vessel wdl and the Rx vessel downcomer are essentially synonymous for the same region.
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For Official Use Only Ques _371 Page 2 l
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Question #7 McGuire Nuclear Station RO Exam Bank Question: 372 Answer: C 1 Pt(s) Unit 2 was respeding to a faulted steam generator event. The operators entered FR-P.1, Response to Imminent PTS, and reach step 15 where they are directed to isolate cold leg accumulator What is the EOP basis for isolating the CLAs in FR-P.l? To prevent injecting the CLA nitrogen bubble into the reactor and creating a gas bubble in the vessel head regio To prevent repressurizing the reactor vessel and adding pressure stress to thermal stres To prevent adding more cold water to the reactor vessel and increasing the thermal stres To prevent depleting CLA volume and to preserve a source of highly borated water to prevent recriticallity during c%idow Distracter Analysis: Incorrect: - the CLAs are isolated to prevent adding to the thennal stress. The gas bubble would not be a limiting concern in FR- Plausible: - this is a valid limiting condition for isolating the CLAs during LOCA depressurizations - good answer, wrong even Incorrect: - adding the CLA volume would not cause an increase in pressure because the addition of the CLA volumc is caused by pressure in the system decreasing below CLA pressur Plausible: - the basis for terminating Si is to prevent adding water to the system and increasing pressure thereby adding pressure stress to thermal stress. Not applicable to the CLA Cc rect answer Incorrect: - the CLAs are not require for Boron addition for this scenari Plausible: - recriticallity this is a valid concern for scenarios that involve adding large quantities of unborated water. Good answer -
wrong event.
For Official Use Only Ques __372 Page 2
Question #8 McGuire Nuclear Station RO Exam Bank Question: 373 Answer: D 1 Pt(s) Unit 2 was at 10% during a plant startup when a loss of condenser vacuum caused a reactor trip to occur. Given the following conditions:
. The steam dump select switch was in steam pressure mode
. Condenser vacuum dropped to 18 inches
. No component or instrument failures occurred
. No operator action taken Which one of the following sequences best describes the actuation of the steam dumps to this event assuming?
Condenser dump valves Atmospheric duma valves open open open shut shut open shut shut Distracter Analysis: Incorrect: - the condenser dumps will not open due to the C9 arming signal not present cau, sed by a low vacuum condition (20 in)
The atmospheric dumps will not open because they do not actuate when the steam dump selector. 'tch is in steam pressure mode Plausible: - this would occur h ale transition had not yet been made to steam pressure mode and C9 had actuated - a normal plant response during a startup < 10% powe Incorrect: - the condenser dumps will not open due to the C9 amling signal not present caused by a low vacuum condition Plausible: - this sequence would occur if there was not a loss of condenser vacuum or if the candidate does not recognize the interaction between condenser pressure and the C9 arming signa Incorrect: - the atmospheric dumps will not open because they do not actuate when the steam dump selector switch is in steam pressure mode Plausible: - this sequence would occur if the plant was in Tave mode - the transition from Tave to steam pressure mode occurs by procedure at 10% power Correct answer For Official Use Only Ques _373 Page 2
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l Question #9 McGuire Nuclear Station RO Exam
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Bank Question: 374 Answer: B
\
1 Pt(s) Unit 1 is responding to a unit blackout. Step 9.g in ECA0.0 that states: J
"9.g IF A T ANY TIME Unit 2 main condenser vactuun goes down significandy, TilEN
_1) Ensure IRC-7 (RC Crossover Supplyfrom Unit 2) is open l _2) WIIEN 1RC-7 is open, TilEN close IRC-5 (RC 1 CrossoverSupplyfrom unit 1)
Which one of the following statemems correctly describes the required i operator action for step 9.g_l) that begins with the sverd " Ensure"7 Determine present status of IRC-7 (no action required other than to check the valve position light in the control room)
I
' Take any necessary action to guarantee that IRC-7 is open but there is no requirement to dispatch personnel to locally verify the valve position if position can be confirmed in the control j room Take any hecessary action to guarantee that 1RC-7 is open including dispatching personnel to locally verify the valve position even ifindication is available in the control roo Carefully appraise the situation - including taking all actions based on evaluation Distracter Analysis:
l Incorrect: - this is the definition for the term " check" Plausible: - requires knowledge of terms - candidate may not know this definition Correct answer Incorrect: - local actions are not required unless the step specifies to dispatch personnel Plausible: - the candidate may not understand the difference between B and C and C is more conservative Incorrect: - this is the definition for the term " evaluate" Plausible: - the candidate may not know the difference between these terms which are very close For Official Use Only Ques _374 Page 2
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y Question #10 McGuire Nuclear Station RO Exam Bank Question: 375 Answer: D 1 Pt(s) What is the concern for nonnal RN system high flow through the KC heat exchangers and what is the basis for this limit? Controlled by mini-ilow protection valves for the train related RN pump to remain less than 2000 gpm which will protect system piping against flow oscillation overcooling in the KC syste Controlled by mini-flow protection s alves for the train related
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RN pump to remain less than 2700 gpm which will protect system piping against flow induced vibratio Controlled by operator action to remain less than 2700 gpm which will protect system piping against flow oscillation and overcooling in the KC syste Controlled by operator action to remain less than 2000 gpm which will protect system piping against flow induced vibratio Distracter Analysis: Incorrect: - the basis is to prevent flow induced vibration. The mini-flow valves maintain flow below 2700 gp Plausible: - the basis is correct for a different limitation - why a super flush must not be performed if KC HX delta pressure is less than 9 psid Incorrect: - the correct flow limit is 2000 gpm Plausible: - each statement is correct on its own - but 2700 gpm is not the correct limit to prevent flow induced vibratio Incorrect: - the limitation is 2000 gpm and the basis is to prevent flow induced vibratio Plausible: - 2700 gpm is the mini-flow valve control setpoint. The basis is correct Gr a di'ferent limitation - why a super flush must not be performed if KC H1 delta pressure is less than 9 psi Correct answer For Official Use Only Ques _375 Page 2
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Bank Question: 376 Answer: A 1 Pt(s) Unit 1 is operating at 100% power when a fire in a room in the Auxiliary ;
Building causes the room heat detector to actuate and the nearby sprinkler I fused link to open. Which one of the following statements describes the correct sequence of events that would occur to provide a flow of water to the sprinkler head? l Water would immediately flow to the sprinkler valve under the normal configuration of the fire header without any further actio Deluge valves IRY-113 and 1RY-114 would open automatically l when the thermostatic detector was actuated and provide water '
to the fire heade Deluge valves IRY-113 and IRY-114 would first need to be opened by operator action in the control room to provide water to the sprinkler syste , Deluge valves IRY-ll3 and IRY-114 would first need to be opened by operator action at the local breakglass stations in the Auxiliary Building to provide water to the sprinkler syste Distracter Analysis: Correct answer - the header is maintained in a filled and pressurized status through bypass lines around RY-113 and RY-114 Incorrect: - RY-113 and RY-114 do not open automatically Plausible: - if the candidate does not know about the bypass lines -
automatic opening should be a conservative feature Incorrect: - RY-113 and RY-144 cannot be opened from the control room Plausible: - the control room must manually open the RY isolation valves to charge the fire main in the Reactor Building Incorrect: - the header is maintained in a filled and pressurized status through bypass valves around RYRY-113 and RY-114 Plausible: - RY-113,114 are opened from the breakglass station in the Auxiliary Building but this action does not provide water the j sprinkler heads -just additional fire main capacity for increased flow For Official Use Only Ques _376 Page 2
Question #12 McGuire Nudear Station RO Exam i
Bunk Question: 377 Answer: D l
1 Pt(s) What is the purpose of the "S" position on the remote / local switch on the auxiliary shutdown panel for NV-1 A? Allows NV-1 A to move to the " safe" position in the event of a i mismatch between ASP local switch position and the actual l valve position Places NV-1 A in the " secure" position by removing power from the valve . Places NV-1 A in the " standby" position which allows positioning from either the control room or the AS Allows operator to check the " status" of NV-1 A position Distracter Analysis; Incorrect: - there is no " safe" position
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Plausible: - S" could stand for " safe" Incorrect: - there is no " secure" position Plausible: "S" could stand for " secure" Incorrect: - there is no " standby" position
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Plausible: - S" could stand for " standby" Correct answer l
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For Official Use Only Ques _377 Page 2 l
Question #13 McGuire Nuclear Station RG Exam !
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Bank Question: 378 Answer: A
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l Pt(s) Unit 2 is in mode 4, cooling down on ND for a refueling outage. A worker reports that both airlock doors are open on the personnel access airlock and the interlock between them is broke {
What is the minimum set of actions that must be taken to comply w.th Tech Specs? Close one airlock door within one hour 1 Close both airlock doors within one hour Close one airlock door within one hour and immediately initiate action to evaluate overall containment leakage ;
l Lock both airlock doors within one hour and immediately initiate action to evaluate overall containment leakage
Distracter Analysis: Correct answer Incorrect: - no requirement to close both doors Plausible: - with the interlock inoperable, the candidate may think it I is necessary to close both doors Incorrect: - no requirement to evaluate overall containment leakage i Plausible: - closing one door will satisfy Tech Specs Incorrect: - no require to lock either door in one hour, - no requirement to evaluate overall containment leakage Plausible: - locking one door is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to satisfy ,
tech specs
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For Official Use Only Ques _378 Page 2
Question #14 McGuire Nuclear Station RO Exam Bank Question: 379 Answer: B 1 Pt(s) Unit 2 is responding to a reactor trip with a total loss of CA flow. The operators entered FR-C.1, Response to Inadecpeate Core Cooling, due to a red path requirement. Step 17 of this procedure states:
Depressurize intact steam generators to atrnosphericpressure What is the purpose of this action?
j To reduce steam generator pressure to allow the line up of a low pressure feedwater system for heat sink restoration i I To re-cover the core via ND flow by reducing NC pressure below l ND pump shutoff head ' To minimize the PTS consequences of the event by reducing the I pressure stress on the reactor vessel wall To reduce steam generator pressure to allow restoration of 450 gpm CA flow Distracter Analysis: Incorrect: - this depressurization occurs even if a heat sink exist Plausible: - this is the correct basis a similar step in for FR- Correct answer Incorrect: - This would increase the thermal stresses not decrease them. PTS is not the concern for this cas Plausible: - reducing SG pressure will reduce NC pressure and thereby reduce pressure stress on the Rx vessel wall Incorrect: - CA pump shutoff head is capable ofinjecting into a hot SG at design pressure - within the SG PORV relieflimits Plausible: - 450 gpm is the total CA flowrate required to provide a heat sink to the reactor - basis for step 11 of FR- i l
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Question #15 McGuire Nuclear Station RO Exam Bank Question: 380 Answer: C 1 Pt(s) Which one of the following statements is a correct description of the capabilities of EMF-48, the Reactor Coolant Monitor at 100% power? Detects beta flux from the NC system coolant. This prevents the detector falsely responding to N16 gamma radiation which would mask a failed fuel even Detects the N16 gamma flux from the NC system coolant which b proportional to the amount of failed fuel claddin C Detects total gamma flux from NC system coolant after a 1 minute sample delay time to allow N16 gamma radiation to decay awa Detects total gamma flux from NC system coolant. The gamma source term from a clad failure would be much greater than the N16 gamma flux at powe Distincter Analysis: Incorrect: - EMF-48 does not detect beta radiation Plausible: - the detector could function if designed this way because N16 gamma would mask the failed fuel problem Incorrect: - N16 gamma is proportional to reactor power level and does not correlate to failed claddin Plausible: - if the candidate was confused over the correlation between N16 gamma levels and power level l Correct answer Incorrect: - The gamma flux from N16 is >> failed fuel at power Plausible: - if the candidate did not know that N16 gamma was >>
than failed fuel source term level ,
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Question #16 McGuire Nuclear Station RO Exam l Bank Question: 381 Answer: C l
1 Pt(s) Unit 1 is operating at 100% power when the output breaker for 1EVIA trips open causing iEKVA to deenergize. Given the following events and conditions:
- All controls are in channel 1 or 1/2 positions
- The operators enter AP/1/A/5500/15 (Loss of Vital or Aux. Control Power).
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The FIRST immediate action step in AP/15 is omitted by the operator e All subsequent steps are performed correctly What consequence would result from a failure to perfonn the FIRST immediate action of AP/15? Pressurizer pressure would increase until PORVs 32 and 36 cycle at 2335 psig S/G levels would increase to the Hi-lli S/G level turbine trip Rods would step in at 72 steps per minute continuously Pressurizer level would increase to the reactor trip setpoint Distracter Analysis: Incorrect: - PORVs do not actuate because operators r 4 tc operable Pzr pressure channel in step 6 of the proc-Plausible: - ifcandidates do not know that Pr . ? hired to a different controlling channel Incorrect: - S/G channels are shined to operable channels as an immediate action -
Plausible: - if candidate does not know that the second immediate action step has the operator shin any effected channels Correct answer: - the missed immediate action is to place rods in manual - this prevents the turbine impulse pressure channel failure from causing Tref to fail to 557 and generating a large +TMM and
, +PMM - rods drive in at 72 steps / min ifin auto l 1) Incorrect: - the failed pressurizer level channel is shined to an operable channel in step 7 of AP/15 - pressurizer level will not increase aller the channel is shifted Plausible: - if the operator does not recognize that the pressurizer level channel is shifted to an operable channel in Ap/15
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For Official Use Only Ques _381 Page 2 l
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1 Question #17 McGuire Nuclear Station RO Exarn Bank Question: 382 Answer: B 1 Pt(s) Unit 2 was conducting a plant start up when a continuous rod withdrawal i accident occurred in shutdown bank A. Given the following conditions and I indications: !
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. The Power Mismatchfremperature Mismatch Meter indicates l .. PMM = +4 F
! * TMM = -2 F e Reactor poweris 50% ,
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e Rods are in automatic l Which one of the following describes control bank rod response?
l rods step in at 72 steps per minute rods step in at 8 steps per minute rods step out at 40 steps per minute rods der out at 8 steps per minute
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Distracter Analysis: Incorrect: - PMM sums with TMM and the combined error signal is +2 F which moves rods in at 8 steps per minute 3ausible: - if the candidate subtracts PMM and TMM the l .ombined error is +6 F which causes rods to move in at 72 SPM Correct: - PMM + TMM = +2 F, rod move at 8 SPM l Incorrect: - rods step at 8 steps per minute PLusible: - if the candidate thinks that the combined error signal is on the proportion part of the rod speed control curve Incorrect: - control rods will respond by stepping in not out i Plansible: - if the candidate thinks that the combined error cause l rods to move out (transposed direction) in his/her mind)
For Official Use Only Ques _382 Page 2 i o
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Question #18 McGuire Nuclear Station RO Exam Bank Question: 383 Answer: D 1Pt(s) Which one of the following choices best describes the bnit 2 response !
I caused by a dropped control rod from 100% power with rod control in automatic? Initially the control rods step in due to a negative power mismatch signal until the temperature mismatch signal builds j up and overcomes power mismatch causing rods to step out to eventually stabilize at a T,,,below 585 I The control rods step in as the psitive temperature mismatch signalis caused by an immediate drop in T,,, due to the reactor power decrease until T,,, stabilizes at ~585 { The control rods step out until the positive power mismatch !
signalis overcome by the negative temperature mismatch signal j and the reactor stabilizes at a temperature above 585 * The control rods step out until the positive power mismatch signal decays away (rate sensitive) and the temperature mismatch signal is nulled out by reactor stabilizing at ~585 * i)istracter Analysis: Incorrect: - power mismatch is positive as turbine power > NI power causing rods to step out Plausible. - if the candidate is confused over the derivation of the power mismatch signa Incorrect: - rods do not step in because Trefis derived from turbine impulse pressure not NI power so it is only effected by the slight reduction in turbine power caused by the cooling due to the dropped rod. This is not a large effect nor is it an immediate effect when compared to the positive power mismatch signal Plausible: - if candidate thinks that NI power provides the input for derivation of the Tref signal Incorrect: - the reactor will not stabilize above 585 F because Tref is clipped at 585 F Plausible: - if the candidate does not recognize that Tref cannot exceed 585 F Correct answer l
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Question #19 McGuire Nuclear Station RO Exam Bank Question: 384 Answer: D 1 Pt(s) Unit 2 is operating at 48% power during a plant start up when an electrical power transient causes the 2A NCP to trip on under voltage and the main turbine to trip. Given the following indications prior to the trip:
f Permissive Signal Status Light i
e P6 (SR Block Permissive) lit
. P7 (Lo Power Rx Trips Blocked) not lit j e P8 (Hi PWR Lo Flo Rx Trip Blocked) lit j e
P10 (Nuclear at Power) lit e
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P11 (Pressurizer S/I Block Permissive) not lit 1
- P12 ( Lo-Lo Tavg) not lit l
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Which one of the following statement describes the reactor trip system '
immediate response?
l the reactor will trip on loss of flow 1 the reactor will trip on turbine trip i the reactor will trip on NCP under voltage the reactor will not trip
Distracter Analysis: Examiner Note: There is no status light for P- Power is at the P-8 setpoint (48%) but the P-8 permissve have not l yet picked up (deadband) and the candidates must determine the status of P-8 by understanding what the P-8 status light is telling the Incorrect: - loss of flow in one loop is blocked below P8 Plausible: - if the candidate does not recognize that P8 is not yet in i Incorrect: - the turbine trip / reactor trip is blocked below P8 I Plausible: - if the candidate does not recognize that .P8 is not yet in ! Incorrect: - although this trip is active with P7 in the coincidence for the trip is 2/4
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Plausible: - if the candidate does not recognize that the coincidence for the NCP monitor panel is 2/4 vice 1/4 Correct answer: - P-8 not yet "in" For Official Use Only Ques _384 Page 2
F Question #20 McGuire Nuclear Station RO Exam Bank Question: 385 Answer: A 1 Pt(s) Which one of the following pressurizer conditions would result in the most rapid rate of pressure decrease if a code safety valve had stuck open (assuming the NC system was in a subcooled condition)?
Fluid Released PZR pressure water 1000 psig water 500 psig steam 1000 psig steam 500 psig Distracter Analysis: Correct answer - solid water condition, highest pressure Incorrect: - lower pressure than A, less leak rate Plausible: - if candidate does not recognize that the leak rate is
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lower forlowei pressure Incorrect: - steam release causes lower pressure drop than water Plausible: - if candidate does not know that a steam release has a smaller mass flow rate that a water release Incorrect: - steam release rate < water release rate and pressure low Plausible: - if candidete does not know correlation between change l in leak rate with change in pressure l
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Question #21 McGuire Nuclear Station RO Exam Bank Question: 386 Answer: A 1 Pt(s) While responding to a small break LOCA, the Unit 1 operators reach step 7 of E-1 when the STA reports a valid orange path on suberiticality. Which one of the following statements describes the correct procedure transitions <
for this situation? Enter FR-S.2 and complete until the end. Transition back to step 7 of E- Enter FR-S.2 and complete until the end. Transition back to step 1 of E l Refer to FR-S.2 for information only while continuing on in E- !
l t Complete E-1 and transition to FR-S.2 when complete !
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Distracter Analysis:
i Correct answe Incorrect:- do not retum to step 1 of E-1 Plausible: - If the candidate does not understand usage rules that direct returning to step in effect from CSF Incorrect: - must enter FR-S.2 unless have a higher priority CSF Plausible: - true for yellow pat Incorrect: - do not complete E-1 if you have a valid orange path l
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Question #22 McGuire Nuclear Station RO Exam l Bank Question: 387 Answer: D 1 Pt(s) What is the basis for the foldout page requirement in E-1 to close NV-150B and NV-151 A (NV Pumps Recirc) when NC pressure is less than 1500 psig? To prevent NV pump runout at low pressures To prevent loss of NPSH to the NV pumps To prevent NV pump from being damaged by operating against shutoff head pressure To prevent a reduction of full SI flow to the core Distracter Analysis: Incorrect: - pump runout is not a concern - NV pumps are designed to run under these conditions Plausible: - pump runout can be a concem for situations where the pump discharge pressure is very low Incorrect: - NPSH is not a concern for NV pumps as the FWST supplies adequate suction hea Plausible: - NPSH can be a consideration for some EOP scenarios Incorrect: - the water does not recirc to the VCT but goes back to the NV pump suctions Plausible: - this is the reason for opening the miniflow valves on the foldout page, not closing the Correct answer - maximizes flow into the core For Official Use Only Ques _387 Page 2
Question #23 McGuire Nuclear Station RO Exam Bank Question: 388 Answer: A 1 Pt(s) Unit 1 is responding to a LOCA in the auxiliary building. The operators have implemented ECA-1.2, LOCA Outside ofContainment. Step 2 guides the operators to attempt to find and isolate the leak. Step 2C requires the following sequence:
e Close INI-162A (NI Pumps Cold Leg Isol)
What is the purpose of opening the mini flow valves? Protect the NI pumps from operating against shutoff head Isolation of a potential LOCA path to the FWST Protect the NI pumps from runout conditions upon restart Provide a diversion path to prevent high pressure water from over-pressurizing the ND system during leak isolation procedures Distracter Analysis: Correct answer Incorrect: - the mini flow valves will not isolate any potential path Plausible: - in ES-1.3 (and some other EOPs), the NI mini flow j valves are closed to prevent pumps from recirculating radioactive I water back to the FWST - right reason, wrong procedure ; Incorrect: - Opening mini flow valves will not protect against nmout Plausible: - if the candidate is confused over the difference between runout conditions and shutoff head conditions I Incorrect: - this will not create a diversion path - it will align the NI recirc line to the ND system and eventually to FWST (which is a low pressure system)
Plausible: - the ND system has a design pressure of 600 psig and if NC system pressure was applied,it would rupture. Setting up' a diversion path would be a reasonable thing to do. This alignment does not accomplish this goal.
For Official Use Only Ques _388 Page 2
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Question #24 McGuire Nuclear Station RO Exam Bank Question: 389 Answer: B l
1 Pt(s) Unit 2 is responding to a LOCA and the operators have reached step 26 of ES-1.2, Post LOCA Cooldown andDepressurization, which requires isolating cold leg accumulators. What is the purpose of this step? l To prevent CLA injection when the NC system depressurizes with the resulting thermal stress and PTS concerns from the cold water on the reactor vessel To prevent injecting the CLA nitrogen bubble into the reactor and creating a gas bubble in the vessel head regio I To reserve the CLA water in the event of a recriticallity during j the cooldown due to a loss of shutdown margin To prevent the excessive addition of highly Borated CLA water l into containment with the potential for Boron crystallization after cooling down, possibly fouling the flow paths in the cor l Distracter Analysis: Incorrect: - PTS is not a concern because cold Si flow is already injecting into the reactor vessel Plausible: - good reason - wrong EOP. This is the correct basis for isolating CLAs during FR- Correct answer Incorrect: - the SI system contains sufficient Boron to prevent recriticallity events without the CLA water Plausible: - recriticallity is a concem anytime the system is cooled down. This procedure cools down the core Incorrect: - although Boron recrystalization is a concem, the EOPs provide a method of preventing this problem (Hot Leg Recirculation) and this is not concem Plausible: - Boron recrystalization occurs at low temperatures and
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high Boric acid concentrations. The BA in the CLAs is relatively high - but not enough to cause this problem For Official Use Only Ques _389 Page 2
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Question #25 McGuire Nuclear Station RO Exam j l
1 Bank Question: 390 Answer: B i
l 1 Pt(s) Unit 1 is recovering from a LOCA. The operators started the process of i terminating safety injection at 2:00 AM. Given the following indications at the following times:
I Parameter 2:00 2:05 2:10 2:15 l t
1) Pressurizer level (%) 40 29 15 11 2) NC pressure (psig) 280 285 290 295 3) ND Flow 1000 1025 1075 1085 4) Core exit T/Cs ( F) 690 702 695 685 5) FWST ievel (inches) 183 179 149 113 6) Containment Pressure (psig) .3 .1 What is the earliest time that the operators should transition to ES-l .3, transfer to cold leg recirculatio Distracter Analysis: This question is designed to test the candidate's ability !
to identify the criteria for switchover to cold leg recire from a list of !
plant parameters. A recent change to the switchover criteria was )
changing the FWST level from 150 inches to 180 inches (since last exam).' In addition, he meets the S/I reinitiation criteria at 0205 and this will check if he recognizes that ES-1.3 has priority Incorrect: - does not meet criteria for switchover, FWST > 180 inches j Plausible: .ifcandidate does not know foldout criteria for j switchover Correct answer - FWST level < 180 inches - reaches ACC value for S/I reinitiation criteria - needs to decide if ES-1.3 needs to be l delayed until S/I reinitiation is completed
' Incorrect: - too late Plausible: - if candidate does not know foldout criteria for switchover - recent change (since last exam) to the switchover criteria - used to be 150 inches - now is 180 inches For Official Use Only Ques _390 Page 2
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Question #25 McGuire Nuclear Station RO Exam Incorrect: - too late Plausible: - if candidate thinks that Si termination must be competed before switchover to cold leg recire or if he thinks that S/I reinitiation takes priority over switchover - reaches non-ACC value for S/I reinitiatio ,
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r Question #26 McGuire Nuclear Station RO Exam l i
Bank Question: 391 Answer: D 1 Pt(s) Which one of the following ND pump operating conditions would be most likely to experience cavitation? ND pumps running with both trains crosstled and NC pressure I is 290 psig ND pump running in mid loop operations and reactor vessel levelis low ND pump running in a shutdown cooling lineup and the discharge header ruptures ND pump running and the suction valve closes part way (pump does not trip)
Distracter Analysis: Incorrect: - this is an example of pump deadheading Plausible: - if the candidate does not understand centrifugal pump :
operating conditions Incorrect: - this is an example of conditions for pump vortexin Vortexing and cavitation are have similar indications but they are different operating situations. Vortexing is the entrainment of air in the suction path due to a low level condition. Cavitation is the loss of NPSH due to a low pressure conditio Plausible: - if the candidate does not understand centrifugal pump operating conditions and this is the condition that the candidates are primarily trained to avoid.- they will be most familiar with this condition Incorrect: - this would be an example of conditions for pump runout Plausible: -if the candidate does not understand centrifugal pump operating conditions Correct answer l
l For Official Use Only Ques _391 Page 2
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L Question #27 McGuire Nuclear Station RO Exam !
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Bank Question: 392 Answer: A
- I 1 Pt(s) Unit 1 is operating at 100% power when a station blackout oce icd. The !
RO diagnoses the following conditions:
! e A total loss of onsite and offsite AC power has occurred
. The 4160 VAC buses remain deenergized e The reactor has not tripped, the rods are still out and power is not decreasing ,
What procedure will be used to bring the reactor to a subcritical condition? 1 l
' E-0 FR- )
i ECA- ECA- Distracter Analysis: Incorrect: - do not enter E-0 first before transitioning to ECA-0.0 if you recognize the blackou Plausible: - it is not wrong to enter E-0 first if the blackout is not immediately recognized Incorrect: - do not monitor CSFs in ECA-0.0 so no transition to FR-S.1 is possible Plausible: - may not realize that CSFs are not monitored in ECA- Correct answer - ECA-0.0 will address suberiticality Incorrect: - - ECA-0.0 will address suberiticality, not ECA- Plausible: - ECA-0.1 will handle restoration of power - start monitoring CSFs in ECA-0.1 - which will allow transition to FRS-0.1 i
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Question #28 McGuire Nuclear Station RO Exam Bank Question: 393 Answer: C
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1 Pt(s) Unit I was conducting a reactor startup when source range channel .N-31
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, failed. Given the following conditions and events:
- No reactor trip has occurred prior to this point t * AP/1/A/5500/16 has been completed
. N-31 repairs have been made e N-31 is being retumed to service
. N-32 = 10' CPS
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Immediately upon taking the "leveltrip" switch to the " normal" position a reactor trip occurred Which of the following operator errors would explain this event? " Operation selector" switch was left in " level adj. " position with i
level potentiometer set at a level of 10' CPS after retesting
. "High-flux at shutdown" switch leit in the " blocked" position
, Instrument power fuses were never reinstalled control power fuses were never reinstalled Distracter Analysis:
J Incorrect: - the operation se/ccror switch is taken out of the circuit t
when the leveltrip switch is taken to normal Plausible: - if the candidate thinks that a test signal can be inserted with level trip switch in the normal position Incorrect: - no effect - only blocks out high flux trip at 10' CPS -
trip is separate from level trip function Plausible: - if the candidate does not understand that the leveltrip
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switch in bypass does not effect this trip - some protective system trips will be active if one signal is left in the blocked condition when the circuit is returned to service Correct answer Incorrect: - if control power fuses has been removed, the reactor
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, would have tripped prior to this point Plausible: - if the candidate does not recognize that a trip would have already occurred
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For Official Use Only Ques _393 Page 2
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Question #29 McGuire Nucler Station RO Exam I
Bank Question: 394 Answer: A 1Pt(s) The condenser air ejector monitor, EMF-33, is designed to detect: gaseous activity in the air ejector discharge into the plant vent particulate activity dissolved in the hotwell water drawn into the air ejector tritium activity in the air ejector discharge into the plant vent iodine activity dissolved in the hotwell water drawn into the air ejector Distracter Analysis: Correct answer - EMF-33 uses a single range beta gas plastic scintillation detector Incorrect: - does not detect particulate activity in water Plausible: - if the candidate does not recall that the purpose of the monitor is to detect for the presence of high gaseous activity in the non-condensable gases discharged from the air ejector Incorrect: - does not detect tritium because tritium has low energy beta radiation that requires a different monito Plausible: - Tritium is a gas contained in reactor coolant Incorrect: - does not detect Iodine activity in the water Plausible: - low levels ofiodine activity would be released into the condensate system during a SGTR For Official use Only Ques _394 Page 2
r Question #30 McGuire Nuclear Station RO Exam I Bank Question: 395 Answer: B l
1 Pt(s) What is the reason for EMF-34 automatically closing the S/G sample header radiation monitor inlet isolation valve (1NM-267) when it reaches a trip 2 l condition during a steam generator tube rupture?
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' To prevent contaminated condensate from reaching the l auxiliary building sump To prevent contaminated water from reaching the turbine building sump To reduce gaseous radiation levels in the CT Lab room To prevent the cation sample columns from becoming saturated with boric acid Distracter Analysis: Incorrect: - the water cannot drain into the auxiliary building Plausible: - if the candidate does not understand the drainage path, l
many drains do go into the auxiliary building sump l Correct answer l Incorrect: - this is not the primary concern at the condensate water does not get exposed to the air in the room unless a manual sample is drawn Plausible: - gaseous activity is the concern at the air ejector monitor i Incorrect: - not a significant concem for a tube rupture l Plausible: - boric acid would enter the cation columns if they were in service
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Question #31 McGuire Nuclear Station RO Exam l
Bank Question: 396 Answer: D 1 Pt(s) Unit 1 is responding to a faulted S/G inside containment in E-1 when a red (
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path occurs on FR-H.1 due to a loss of CA pumps. The operators are attempting to recover the heat sink using the condensate system when the following system parameters occur:
0200 0205 0210 0215 NC System Pressure (psig) 1986 1993 1981 1935 Containment pressure (psig) .9 .9 Steam eenerator levels (inches)
1 A S/G (WR) 16 14 18 15 I IB S/G (WR) 11 10 16 16 )
1D S/G (WR) 14 13 18 20 l l
Total feedwater How rate 0 50 100 200 l Core exit thermocouple temp ( F) 936 1039 1012 981 I What is the earliest time (if any) that the operators should establish feed and bleed in accordance with the foldout criteria in FR-H.l?
REFERENCES PRO VIDED - FR-S.1 Foldout Page (36) never need to initiate feed and bleed )
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1 l Distractea Analysis: Incorrect: - criteria are met at 0215 Plausible: - if candidate does not analyze parameters correctly or mistakenly thinks that once containment pressure drops below 3.0
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psig, ACC setpoints no longer apply. AT 0205, condensate water is causing S/G levels to increase and appears to be effectively I
recovering the heat sin Incorrect: - 0205 - only 2 S/Gs < 12 inches, ACC not met l
For Official Use Only Ques _396 Page 2
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Question #31 McGuire Nuclear Station RO Exam Plausible: - Highest CET readings - worst conditions for NC system Incorrect: - 0210 - ACC applies but only 2/4 S/G WR < 17 inches due to feedwater entering the S/G and raising levels Plausible: - if candidate doesn't realize ACC criteria is 3/4 S/G < 17 inches, feedwater is entering the S/Gs and level is rising Correct answer - 3/4 S/G WR < 17 - ACC still applies although containment pressure has dropped back below 3 psig l
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l For Official Use Only Ques _396 Page 3 I L
Question #32 McGuire Nuclear Station RO Exam Bank Question: 397 Answer: B 1 Pt(s) Which one of the following system lineups best describes the proper operation of the DC auxiliary power system during an equalizing charge of battery IDP? The normal battery charger charges IDP separately from the distribution center The standby battery charger supplies distribution center loads Unit DC bus tie breakers are open The normal battery charger supplies distribution center loads The standby battery charger charges IDP separately from the distribution center Unit DC bus tie breakers are closed The normal battery charger charges IDP separately from the distribution center The nandby battery charger in standby Unitous tie breakers are closed The normal battery crarger is in standby The standby battery charger charges 1DP tied to the distribution center Unit bus tie breakers are open Distracter Analysis:
l Incorrect: - bus tie breakers must be shut, cannot allow a battery l charger to supply a bus without being connected to a battery Plausible: - the bus tie breakers are normally required to be open for electrical separation between the 2 unit batteries Correct answer Incorrect: - allows 2DP to supply 1 DP loads without any charger Plausible: - this could physically function but 2DP would be under-sized to handle both units with only one charger Incorrect: - the battery voltage would be too high to allow 1DP to be supply bus loads Plausible: - this would physically work For Official Use Only Ques _397 Page 2
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L l Question #33 McGuire Nuclear Station RO Exam l
Bank Question: 398 Answer: C 1 Pt(s) Unit 2 is in the process of starting up the reactor. Given the following conditions and events:
e 0200 - 2 EMF-3 (Refueling bridge ARM) reaches a trip 2 conditio . 0210 - N-32 is deenergized due to a SR detector failure
- The operators complete all required procedural steps during the startup What is the earliest time (if any) that the containment evacuation alarm will no longer actuate in Unit 2 during the startup?
0200 0205 0210 0215 4 4 N-31 (CPS) 1.2x10 1.7x10 2x10 4 0 3 4 N-32 (CPS) 9.2x10 1.1x10 0 0 N-35 (amps) 1.1x 10-' l .4x10 ' l.610-' i.910-'
N-36 (amps) 1.3x 10-' l.5x10- 1.8x10-' 2.1 x10-' the containment evacuation alarm will actuate at all times Distracter Analysis: Incorrect: - both SR. high flux trips are still active Plausible: - if the candidate thinks that P6 permissive actuation cuts out the alarm Incorrect:- N-31 SR high flux trip still active Plausible: - if the candidate does not recognize that BOTH SR high flux trips must be blocked Correct answer Incorrect: - alarm is blocked at 0215 Plausible: - ifcandidate thinks that some other action or interlock blocks the alarm r
l For Official Use Only Ques._398 Page 2
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Question #34 McGuire Nuclear Station RO Exam Bank Question: 399 Answer: A 1 Pt(s) Unit I was operating at 100% with the pressurizer level controller in the 1-2 position. Given the following initial response:
. Charging How reduces to minimum
. Backup heaters immediately energize
. Actuallevel begins to decrease Which one of the following failures has occurred to cause this plant response? PZR level channel 1 detector reference leg has ruptured PZR level chanel 1 detector variab.le leg has ruptured PZR level raannel 2 detector reference leg has ruptured PZR level channel 2 detector variable leg has ruptured Distracter Analysis: Correct answer - a leak in the reference leg causes the pressurizer channel to sense a high level condition - which causes the system response as indicated. Actual pressurizer level decreases initially due to charging now decreasing while letdown remains in servic Incorrect: - variable leg rupture causes channel 1 to sense a low PZR level - actual level will increase not decrease initially Plausible: - if the candidate thinks that this will cause a high level Incorrect: - will cause channel 2 to sense a high level - will not get heaters deenergizing - only get a high level alarm Plausible: - if the candidate confuses the plant response for channel 2 - thinks that channel 2 controls Incorrect: - pressurizer level would increase not decrease Plausible: - if the candidate confuses the direction of the pressurizer level failure or doesn't recognize plant response - this is very similar to a channel I high failure only the actual level increase instead of decreases For Official Use Only Ques __399 Page 2
Question #35 McGuire Nuclear Station RO Exam Bank Question: 400 Answer: B 1 Pt(s) Unit 2 is responding to a small break LOCA. Given the following plant conditions and events:
e 2D S/G pressure = 1228 psig (all channels indicating the same)
+ 2D S/G NR level = 73%
e no S/G PORVs or code safeties have lifted on any S/Gs
- Operators have entered FR- * Containment pressure = 3.1 psig Which one of the following statements represents actual conditions for the S/G instrumentation under these accident conditions? Actual wide range level > 73% Actual wide range level < 73% Actual S/G pressure < 1228 psig Actual S/G pressure > 1228 psig Distracter Analysis: Incorrect: - narrow range SGWL detectors are calibrated for i normal (hot) operating conditions. Wide range SGWL detectors are calibrated for cold (shutdown) conditions. This means that the wide range SGWL instrument will read lower than actual level under hot ;
conditions because variable leg water density is less than the calibrated setting Plausible: - if the candidate fails to understand the effect of cold calibration on the SGWL or gets the effect reverse Correct answer Incorrect: - SG pressure transmitter accuracy is not sensitive to temperature changes in the S/G - pressure transmitters located outside containment - no ACC valves for steam pressure Plausible: - if the candidate is confused over the accuracy of the S/G pressure transmitters. May confuse pressure transmitters with level transmitters Incorrect: - SG pressure transmitters are not sensitive to temperature changes in the S/G Plausible: - if the candidate is confused over the accuracy of the S/G pressure transmitters. May confuse pressure transmitters with level transmitters For Official Use Only Ques _400 Page 2
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Question #36 McGuire Nuclear St .tlon RO Exam Bank Question: 401 Answer: D 1 Pt(s) Unit 2 was operating at 100% when a single control rod in control bank D drops into the core due to a failed CDRM. The SRO directs that the dropped rod be recovere Which one of the following prevents the remaining rods in the control bank from being withdrawn while the dropped rod is being recovered? The Rod Stop Bypass is used to block control rod bank D outward movement C-11 actuated when the rod dropped and will prevent outward rod motion by control bank D The Lift Coil Disconnect Switch is openen on the dropped rod to electrically isolate it from control bank D The Lift Coil Disconnect Switches are opened on control bank D rods that did not drop Distracter Analysis: Incorrect: - the rod stop bypass only blocks the overpower channel signal to rod control which does not effect the situation under these conditions Plausible: - if the candidate does not recognize the function of this switch Incorrect: - C-11 will not allow any rod motion - could not pick up the dropped rod and it has not actuated under these conditions Plausible: - C-1 I will prevent outward rod motion in bank D Incorrect: - will not be able to pick up the dropped rod Plausible -if the candidate was not familiar with the actions of the lift coil disconnect switch - it will disconnect the rod from bank D but the rod cannot then be withdrawn with the switch open Correct answer For Official Use Only Ques _401 Page 2
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Question #37 McGuire Nuclear Station RO Exam Bank Question: 402 Answer: C 1 Pt(s) Unit 1 is responding to a LOCA. Given the following conditions inside containment:
l 0200 0210 0220 0230 l Containment Pressure (psig) .1 .3 Safety injection occurred at 020 J At what time do the NCPs lose KC and RN cooling flow and when can the operators regain restore cooling flow by using reset pushbuttons?
Lost cooline flow Able to reenin cooline flow I Distracter Analysis: Incorrect: - setpoint is too low Plausible: - this is the hi containment pressure setpoint - causes an SI which causes a phase A containment isolation Incorrect: - pressure too low Plausible: - if the candidate thinks that hi containment pressure causes a loss of flow and this cannot be reset until after the CPCS reset at 0.35 psig Correct answer: - setpoint is hi-hi containment pressure > 3.0 psig gets a phase B isolation - can be reset immediately even with the signal present Incorrect: - actuation setpoint is correct but reset point is wrong Plausible: - if the candidate thinks that pressure must drop to clear the signal before allowing a reset For Official Use Only Ques _402 Page 2
F Question #38 McGuire Nuclear Station RO Exam l
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Bank Question: 403 Answer: C
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1 Pt(s) Which one of the following instruments would be the most reliable indication when core voiding was occurring during a LOCA severe accident? Source Range Excore Nuclear Instruments Intermediate Range Nuclear Instruments Wide Range Neutron Instruments Power Range Nuclear Instruments Distracter Analysis: Incorrect: - not environmentally qualified for LOCA conditions Plausible: - Source Range instruments will detect the change in moderator effectiveness caused by voiding - used during TMI accident Incorrect: - not environmentally qualified for LOCA conditions Plausible: - instruments will detect the change in moderator ,
effectiveness caused by voiding I Correct: - these instruments are environmentally qualified IAW RegGuide 1.97 Incorrect: - not environmentally qualified and not sensitive at low power levels Plausible: - if candidates do not consider the source temi for the core when the voiding commences - i.e. shutdown low neutron flux l levels I l
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For Official Use Only Ques __403 Page 2
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Question #39 McGuire Nuclear Station Rs Exam l
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Bank Question: 404 Answer: C 1 Pt(s) Unit 1 is responding to a small break LOCA. Containment pressure reached l 3.5 psig. The Subcooling Margin Monitor currently indicates -35 Which of the following statements best describes the status of subcochng in
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the~ core? The core is subcooled by 35 F The core is superheated by 35 F The core is superheated by at least 35 *F The core is subcooled by at least 35 F l
i Distracter Analysis:
l l Incorrect: - the core is not subcooled l
Plausible: -if the candidate reverses the meaning of the indication -
believing that it is measuring subcooling not margin to subcooling Incorrect: - subcooling may be greater than 35 F Plausible: - on face value I Correct answer - the subcooling margin monitor is at the high end of the measuring range and subcooling may be greater than 35 F Incorrect: - the core is not subcooled Plausible: - if the candidate reverses the meaning of the indication -
believing that it is measuring subcooling not margin to subcooling l
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For Official Use Only Ques _404 Page 2
n l Question #40 McGuire Nuclear Station RO Exam l
l Bank Question: 405 Answer: D l
1 Pt(s) Unit 2 is conducting a plant start up. The 2D S/G control valve fails open and causes a feedwater system transient. No reactor trip occurs due to effective operator control. Given the following conditions and events:
0200 0210 0220 0230 Turbine output 38 % 39 % 40 % 42 %
S/G CF Control Valves Position 2A S/G (% open) 39 % 25 % 24 % 40 %
2B S/G (% open) 35 % 22 % 23 % 39 %
2C S/G (% open) 42 % 21 % 26 % 45 %
2D S/G (% open) 45 % 100 % 80 % 50%
All 4 CF bypass valves were closed at 020 During what time period (if any) will the main feedwater pumps trip? l i Will not trip Distracter Analysis: Incorrect: - AMSAC trip not active < 40% power Plausible: - loss of feedwater control starts at this time frame Incorrect: - AMSAC trip not active < 40% power Plausible: - 3/4 feedwater paths are blocked (CF valves position s 25% closed) Incorrect: - no longer have 3/4 feedwater pa'hs blocked Plausible: - AMSAC no longer blocked > 40% power Correct answer - the AMSAC logic requires 3/4 flow paths to be blocked (CF valves to close) when turbine impulse pressure > 40% ;
power. This is never met. When AMSAC actuates, the MFP turbine !
is tripped
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l i For Official Use Only Ques _405 Page 2
Question #41 McGuire Nuclear Station RO Exam !
l Bank Question: 406 Answer: D l l
1 Pt(s) Unit 2 is operating at 100% power in a normal lineup. Reactor pressure I drops to 1953 psig because of a spray valve malfunction. Which one of the following events will cause the turbine driven CA pump to start but NOT the motor driven CA pumps?
l AMSAC Low S/G NR level on 2/4 S/Gs ;
i Both main feedwater pumps trip Loss of VI header pressure l
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Distracter Analysis: The drop in reactor pressure is provided for the candidate to have to consider the effect of P-11 circuitry on the CA start logic (no effect). Incorrect: - AMSAC causes only the motor driven CA pumps to start, not the turbine driven CA pump Plausible: - a valid CA pump start signal Incorrect: - cause BOTH the MD and TD CA pumps to start Plausible: - a valid CA pump start signal Incorrect: - causes BOTH MD and TD CA pumps to Plausible: - a valid CA pump start signal Correct answer - causes TD CA pump steam supply valve to open For Official Use Only Ques _406 Page 2
Question #42 McGuire Nuclear Station RO Exam I
Bank Question: 407 Answer; D 1 Pt(s) Unit I has a liquid radioactive waste release in progress from the Ventilation Unit Condensate Drain Tank (VUCDT) through the RC system. All lineups and authorizations have been properly made in accordance with OP/0/B/6200/35 using the normal path. 2 RC pumps are the minimum required under LWR documen Given the following initial conditions:
- 3 RC pumps are mnning ;
e 'l EMF-44 correctly set for trip 1 and trip 2 activity limits )
e MRIRR = 75 GPM based on boron concentration j e No other releases are in progress
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What automatic actions would terminate the release?
l WM-46 will close automatically if 1 RC pump trips ' WM-46 will close automatically when EMF-44 reaches the trip 2 setpoint i WL-320 and WP-35 will close automatically if1 RC pump trips I WL-320 and WP-35 will close automatically when EMF-44 reaches the trip 2 setpoint Distracter Analysis: Incorrect: - RC pump interlock will not actuate - set at 2 pumps 1 (minimum required on LWR document). WM-46 is isolated and not
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used anymore as a release pat Plausible: - this was formerly the normal release path Incorrect: - WM-46 receives a closing signal from EMF-44 but this is not the normal path for a release. WM-46 is isolated and not used ,
anymor Plausible: - this was formerly the normal release path Incorrect: - RC pump interlock will not actuate - set at 2 pumps (minimum required on LWR document).
Plausible: - this was the correct answer on the last NRC exam -
modified the question Correct answer - changed this distracter from the last exam For Official Use Only Ques _407 Page 2
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Question #43 McGuire Nuclear Station RO Exam Bank Question: 408 Answer: C
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1 Pt(s) The following parameters /;e observed for Hydrogen and Oxygen concentration in the Waste Gas Holdup syste ]
0200 0215 0230 0245 Hydrogen Concentration 5.3% . 6.8% 7.5% 8.5% j Oxygen Concentration 1.2% 1.5% 4.2% 7.1%
What is the earliest time that the operator is required, by procedure, to immediately suspend all additions to the Waste Gas Holdup System to prevent flammable limits from being exceeded? ' Distracter Analysis:
The flammability limits are already sampled and explicitly tested in the JPM questions Incorrect: - Oxygen concentration < 4% - no flammability concern Plausible: - if candidate thinks that Hydrogen > 4% alone is sufficient to require tennination Incorrect: - Oxygen concentration < 4% - no flammability concem Plausible: - if candidate thinks that Hydrogen > 4% alone is sufficient to require termination Correct: - Oxygen concentration reaches 4% and Hydrogen > 4% ; Incorrect: - required at 0230 l Plausible: - if the candidate does not know flammability tech spec !
restrictions of TS 3.11.2.5 - an immediate action LCO .
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l Question #44 McGuire Nuclear Station RO Exam Bank Question: 409 Answer: B 1 Pt(s) . Unit I was shutting down in mode 3 when a fuel handling accident in the refueling canal area occurred. A spent fuel bundle was smashed into the side of the canal and a section of cladding was damaged. Given the following conditions:
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Gamma radiation levels are 10 R/hr at a distance .of 10 ft from the I bundle l
The refueling bridge area radiation monitor,1 EMF-16, is 2011 away l from the suspended bundle ;
e IEMF-16 trip i setpoint = 1 R/hr
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- 1 EMF-16 trip 2 setpoint = 6 R/Hr
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. A cloud of radioactive particulate gas surrounds the fuel handling bridge e Beta radiation levels from this cloud are 25 R/hr i What is the correct description of the alama status of 1 EMF-16 and the containment evacuation alarm? No alarms will be present on IEMF-16 - the containment evacuation alarm will not sound EMF-16 will be in trip I alarm - the containment evacuation alarm will not sound EMF-16 will be in trip I alarm - the containment evacuation alarm will sound EMF-16 will be in trip 2 alarm - the containment evacuation alarm will sound Distracter Analysis:
The gamma radiation levels at 20 fl will be:
Ifpoint source modelis used:
y = 10 R/hr (10)2 /(20)2 y = 2.5 R/hr > trip 1 but < trip 2 Ifline source model used:
y = 10 R/hr(10/20)
y = 5 R 9 > trip 1 but < trip 2 Incorrect: - y > trip 1 For Official Use Only Ques._409 Page 2
F Question #44 - McGuire Nuclear Station RO Exam l
Plausible: - if the candidate makes an error in computing the point source y radiation levels Correct: - y > trip 1 but the containment evacuation does not sound until 1 EMF-16 exceeds the trip 2 condition Incorrect: - y > trip 1 setpoint - BUT the containment evacuation l alarm will NOT sound until 1 EMF-16 exceeds the trip 2 condition (unless blocked)
Plausible: - if the candidate thinks that the containment evacuation alarm sounds at trip 1 Incorrect: - y < trip 2 = no alarm Plausible: - if the candidate adds in the beta dose rate of l miscalculates the y dose rate. lEMF-16 is a GM tube and will not detect beta radiation l
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i For Official Use Only Ques _409 Page 3
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Question #45 McGuire Nuclear Station RO Exam Bank Question: 410 Answer: C 1Pt(s) Unit 1 is in mode 4 preparing for refueling. Given the following conditions:
e NC pressure = 400 psig
. Tave = 200 F e Pressurizer level = 25%
. Boron concentration = 230 ppm Which one of the following statements correctly describes what must be added to the NC system to change boron concentration to 1120 pp REFERENCES PROVIDED Data Book OP/1/A/6100/22 Enclosure 4.3 Sections 5.1 and add 8501 - 10,000 gallons boric acid add 10,001 - 11,500 gallons boric acid add i1,501 - 13,000 gallons horic acid add 13,001 - 14,500 gallons boric acid Distracter Analysis:
230 - 500 ppm = 2743 gal 500 - 700 ppm = 2106 gal 700 900 ppm = 2174 gal 900 - 1120 ppm = 2475 gal l Total = 9498 gal boric acid 9498 x 1.28 (K) = 12,157 gal boric acid Incorrect: - did not chose any K factor from table Plausible: - if candidate forgets to use K factor Incorrect; - chose incorrect K factor from table 5.2 or some other calculation enor Plausible: - if candidate misreads plant conditions Correct answer
! Incorrect answer - chose incorrect K factor from table 5.2 -
l selected 1.4 for water solid condition or some other calculation error Plausible: - if candidate misreads plant conditions
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t For Official Use Only Ques _.410 Page 2
r Question #46 McGuire Nuclear Station RO Exam Bank Question: 411 Answer: B 1 Pt(s) Unit 2 is has recently restarted after a refueling outage and is operating at 100% power. A previously isolated Mixed Bed Demineralizer (that had been in use extensively prior to the shutdown) was inadvertently placed in service. All primary chemistry parameters are within specification Which one of the following statements best describes the expected effects on the chemistry of the NC system? Boron concentration will increase due to the exchange ofII2 BOi ions from the demineralizer bed back into the coolant causing a net decrease in reactivit Boron concentration will decrease due to II2 BOf ion removt 1 from the coolant into the demineralizer bed causing a net increase in reactivit Boron concentration will not change because the mixed bed demineralizer will be in equilibrium with the NC Boron concentration, having achieved saturation conditions while operating on line in the previous fuel cycl Boron concentration will not change because the mixed bed demineralizer only exchanges impurity ions for II2 BOi ion The chemical impurity levels in the coolant are not significant and current chemistry impurity levels in the coolant are within speclilcation j I
Distracter Analysis: Incorrect: - H BOf 2 ions will not increase and reactivity will increase not decrease - see below j Plausible: - if the candidate confuses the equilibrium conditions {
I Correct: - the demin is saturated with H2 BOfions at equilibrium '
concentration levels from EOL - where NC system Boron concentration was much lower than BOL conditions. Hence, until a ,
l new equilibrium is established, the demineralizer will absorb H2 BOi ions causing a reduction of H2 BOiand an increase in reactivit )
l l l Incorrect: - the equilibrium conditions are not the same - H2 BOi levels are much lower at EOL in the coolant so equilibrium conditions in the demineralizer is also lower For Official Use Only Ques _411 Page 2 l
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Question #46 McGuire Nuclear Station RO Exam
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Plausible: - if the candidate is confused over the establishment of equilib ium conditions in the coolant Incorrect: - mixed bed demineralizers will exchange H2 BOf ions only until they reach saturation conditions.
- Plausible
- mixed bed demineralizers do exchange impurity ions for l H 2BOfions - only part of the chemical process in effect I;
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Question #47 McGuire Nuclear Station - RO Exam Bank Question: 412 Answer: A 1 Pt(s) Unit 1 is responding to a LOCA. Given the following initial conditions:
i e Safety injection actuation occurred at 0150
. MSIVs are shu * Phase B containment isolation has occurred The operators reach step 2 in ES-1.1 requiring a reset of the safety injection J signa j k
Given the following parameter trends at 0200: )
e NC pressure = dropped to 1850 psig then stabilized at 1951 psig
. Steamline pressure = 771psig - decreasing slowly e Containment pressure = 2.2 psig - decreasing slowly Given the following sequence of operator actions:
0202 Blocks the low steam line pressure SI signal 0203 Blocks the low PZR pressure SI signal 0204 Resets the phase B isolation signal What is the earliest time that SI could be reset? Distracter Analysis:
This question will test if a candidate understands that safety injection can be reset even with valid SI actuation signals still present Correct answer - safety injection can be reset afler a 60 minute timer has elapsed and the train related reactor trip breaker has opened (P-4). None of the SI signals being present will prevent reset of St. Once reset, only manual SI is available Incorrect: - SI already reset at 0200 Plausible: - the steam line low pressure SI can be blocked < P-11 For Official Use Only Ques _412 Page 2 l
Question #47 McGuire Nuclear Station RO Exam L Incorrect:- SI already reset at 0202
! Plausible: - this will block the low pressurizer pressure SI signal , Incorrect:- SI already reset at 0202 l
Plausible: - this action will block hi-hi containment pressure SI signal i
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Question #48 McGuire Nuclear Station RO Exam l
Bank Question: 413 Answer: A j l
l 1Pt(s) What is the source of power to the Upper Containment Air Handling Units? l l VAC Essential 1 oad Centers VAC Shared Load Centers VAC Unit Load Centers VAC Auxiliary Load Centers l
Distracter Analysis: Correct answer: - these receive safety grade power off the ETA /ETB busses Incorrect: - they use essential load center power Plausible: - 600 VA.C shared load centers do power some vital equipment like the control room air handling units Incorrect: - they use essential load center power Plausible: - this is the " normal" kind of power - non-safety grade Incorrect: - No such type ofload center at McGuire Plausible: - the term " auxiliary" sounds like a reasonable choice for an auxiliary load like the AHUs.
For Official Use Only Ques _413 Page 2
, Question #49 McGuire Nuclear Station RO Exam Bank Question: 414 Answer: B 1 Pt(s) Unit 2 was operating at 100% power when a double-ended high energy line 1
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break occurred and containment spray operated as designe Proper ftmetioning of the ice condenser will ensure that peak containment pressure, which is reached (A) minutes following the event, remains below (B) psi i I
(A) minutes (B) psie i I Distracter Analysis:
! Incorrect: - peak pressure reached in 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> (105 minutes) I Plausible: - peak pressure is correct,20 minutes is approximately the time that peak pressure is reached in a non-ice condenser plant Correct answer Incorrect: - peak pressure too high, time to reach is too short Plausible: - would be appropriate for some non-ice condenser plants Incorrect:- peak pressure too high Plausible: - if candidate does not know peak pressure for containment - 20 psig is appropriate for some non-ice condenser plants i For Official Use Only Ques _414 Page 2
(I l Question #50 McGuire Nuclear Station RO Exam l
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Bank Question: 415 Answer: B l
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I 1 Pt(s) Unit 1 is operating at 100% power when an electrical transient causes
, several condensate system pumps to trip. The operators take action to regain ;
! the system. Given the following conditions and events:
l start 10 sec 20 sec 30 see CF pump 1 A Suction Pressure (psig) 451 238 235 225 CF pump 1B Suction Pressure (psig) 448 225 235 225
- Hotwell Pumps running 2 2 3 1
- Condensate Booster Pumps running 2 1 0 1 i What is the earliest time (if any) that BOTH main feedwater pumps will have tripped? seconds seconds l seconds feedwater pumps will continue to run under these conditions Distracter Analysis: Incorrect: - CF pump 1 A continues to run as suction pressure temains > 230 psig Plausible: - CF pump 1B will trip on low suction pressure Correct answer: - both CF pumps trip when 3/3 condensate booster pumps trip - trip is irrespective of suction pressure Incorrect: - trip criteria met but trip occurred earlier at 10 sec Plausible: - if candidate does not know that the CBPs are interlocked with the CF pumps Incorrect: - tripping criteria are met at 10 and 15 sec l Plausible: - if candidate does not know CF pump trips
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l Question #51 McGuire Nuclear Station RO Exam Bank Question: 416 Answer: D 1 Pt(s) Unit 2 was operating at 100% power when a reactor trip occurred at 020 T.,, is defeated for loop C while troubleshooting efforts are in progres Given the following conditions and events:
e a T, RTD in loop 2C is out ofservice e at 0202, a T, RTD in loop D fails high (*) i
0201 0202 0203 0204
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T ,, loop A ( F) 556 554 552 551 T,., loop B ( F) 558 556 554 552 T,,, loop D ( F) 552 595* 595* 595* '
At what time will feedwater isolation occur? I ' l Distracter Analysis:
l The Tave signal for loop C is bypassed and does not provide an input to Lo l T, . Coincidence for P-4 is normally 2/4 but becomes 2/3 when loop C l Tave is bypassed l
, Incorrect: - 1/3 logic satisfied for Lo T,,, j Plausible: - if candidate thinks that the bypassed loop C input j satisfied the second channel for Lo T,,, Incorrect: - 0/3 logic satisfied for Lo T,,, i Plausible: - if candidate thinks that the bypassed channel and the i failed channel will input 2/4 signals for Lo T.,, Incorrect: - 1/3 logic satisfied f
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Plausible: -if candidate thinics that 2/3 signals from the failed loop D channel and loop A will satisfy the logic Correct answer - 2/3 signals from loops A and B for Lo T,,, i
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For Official Use Only Ques _416 Page 2 l l
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,. t ( ~ Question #52 McGuire Nuclear Station RO Exam Bank Question: 417 Answer: C 1 Pt(s) Unit 1 is responding to a reactor trip in ES-0.1. Given the following events and conditions:
- NC pumps tripped on under frequency due to an electrical transient
- Pressurizer pressure is stable at 2025 psig
. Stearn generator level is stable at 40 %
Which one of the followinb conditions verifies that natural circulation flow is occumng in the core?
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T,,, L,,a S/G nressure NCS subcooline l Increasing stable increasing <0F l i stable increasing increasing >0 F stable decreasing decreasing >0 F decreasing stable stable <0 F Distracter Analysis: Incorrect: - NCS subcooling <0 F Plausible: - if candidate does not understand criteria for natural i circulation Incorrect: - losing thermal driving head (TDH) - S/G pressure
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increasing Plausible: - TDH is present and increasing Correct answer Incorrect: - NCS subcooling <0 F Plausible: - if candidate does not understand criteria for natural circulation For Official Use Only Ques _417 Page 2
Question #53 McGuire Nuclear Station RO Exam Bank Question: 418 Answer: B 1 Pt(s) Unit 2 was operating at 100% power with all systems in their normal lineu Which one of the following conditions would cause one RV pump to auto start? liigh containment temperature Safety Injection actuation Shifting upper containment AIIUs to fast speed RV pump low AP alarm actuation Distracter Analysis: Incorrect: - does not auto start pumps .
Plausible: - high temp condition would require more RV cooling l flow Correct answer: - SI actuation would auto-start all RV pumps in
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AUTO Incorrect: - does not auto start pumps Plausible: - condition would warrant additional cooling flow Incorrect: - does not auto start pumps Plausible: - a high D/P alarm opens RV bypass valves around the ,
RV pumps to provide mini-flow protection - a low D/P condition I would be indicative of a loss of RN flow to the VL AHUs i
For Official Use Only Ques __418 Page 2
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Question #54 McGuire Nuclear Station RO Exam
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l Bank Question: 419 Answer: C 1 Pt(s) Which one of the following systems could canse the alarm
"lCE COND LOlVERINLETDOORS OPEN" ifoperated improperly? Lower Containment Air Ilandling Units ,
i Upper Containment Air llandling Units Containment Air Return Fans Containment Ilydrogen Skimmer System Distracter Analysis: Incorrect: - will not cause a significant pressure differential between upper and lower containment l
Plausible: - circulates air in lower containment l Incorrect: - will not cause a significant pressure differential between )
upper and lower containment Plausible: - circulates air in upper containment Correct answer Incorrect: - will not cause a significant pressure differential between upper and lower containment Plausible: - pulls air out ofparts ofcontainment and will cause some D/P to occur For Official Use Only Ques _419 Page 2
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Question #55 McGuire Nuclear Station RO Exam Bank Question: 420 Answer: A 1 Pt(s) Unit 1 is operating at 100% power when the MSIV for the IC steam generator closes. Which one of the following statements describes the impact of this malfunction on the plant? Unable to remove heat from the IC S/G through the condenser dump valves Loss of steam supply to the turbine drive CA pump Unable to remove heat from the IC S/G through the IC S/G PORY Remaining MSIVs will close due to steam line pressure rate trip signal from the transient
_ Distracter Analysis: Correct answer Incorrect: - the steam supply line comes off upstream of the MSIV Plausible: - if the car.didate does not recognize that closing the MSIV does not isolate this path Incorrect: - the S/G PORV comes off upstream of the MSIV Plausible: - if the candidate does not recognize that closing the MSIV does not isolate this path Incorrect: - although rate trip signal may be generated, the rate trip signal is blocked above P-11 (NC system pressure >1955 psig )
Plausible: - would actuate if < P-11 and SI is blocked For Official Use Only Ques._420 Page 2 l
F Question #56 McGuire Nuclear Station RO Exam l
i Bank Question: 421 Answer: B 1 Pt(s) Unit 1 is operating at 100% power. If valve CS-18 (CA Supply from UST)
is shut, which one of the following statements correctly describes who would most liDly find this misalignment and when it would be noted? The on-coming shift NCO would discover the valve is out of alignment because he is required to complete the NCO Turnover Checklist computer program prior to relieving the shift.
l The off-going shift NCO would discover the valve is out of alignment because he is required to complete the NCO Turnover Checklist computer program prior to being relieved by the on-coming shif The on-coming shift NCO would discover the valve is out of alignment because he is required to complete the NCO Turnover Checklist computer program immediately after relieving the shif The off-going shift NCO would discover the valve is out of alignment because he is required to complete the NCO Turnover Checklist computer program immediately after being relieved by the on-coming shif Distracter Analysis:
1 Incorrect: - the off-going NCO completes the checklist ,
Plausible: - if the candidate does not know the shift turnover j requirements - this is more typical at other places Correct answer - per OMP 5-6 Incorrect: - the checklist is completed prior to being relieved by the off-going NCO Plausible: - if the candidate does not know the sequence for the sL:ft relief Incorrect: - the checklist is completed prior to relieving not after
, Plausible: - if the candidate does not know the shift tumover l requirements l
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m Question #57 McGuire Nuclear Station RO Exam Bank Question: 422 Answer: D 1 Pt(s) Unit 1 is responding to a loss of main feedwater event. If feedwater to the steam generators is being supplied from the Turbine Driven (TD) CA pump, how is steam generator water level controlled immediately following the auto start? Automatically controlled by controlling the TD CA pump speed using the governor speed signalisolation valve which will match TD CA pump output to maintain steam generator level to maintain 450 gp Manually controlled from the control room by adjusting the TDCA pump speed using the governor speed control which will match TD CA pump output to maintain steam generator leve Automatically controlled by positioning the discharge valve and recire/ mini-flow valve immediately following the auto start signal to maintain 450 gp Manually controlled from the control room after first resetting the modulating control valve reset pushbuttons and positioning the TD CA pump discharge valves and recirc/ mini-flow valve to maintain steam generator water leve Distracter Analysis: ince rect: - no automatic control Plausible: - if candidate thinks that auto control is correct - the operator need not take manual control during the initial phase when the CA pumps start
. Incorrect: - TD CA pump speed is not manually controlled to provide flow rate Plausible: - the main feed pumps use speed control to regulate flow Incorrect: - no automatic control - the discharge and recirc mini-flow valves cannot be operated until aller they have been reset Plausible: - reset is not required unless the start signal is an auto start Correct answer - the modulating control valve reset is required because the control circuitry is not seismically qualified. On an auto start, these valves go to the safe position (discharge open, recirc shut) until the reset is performed For Official Use Only Ques __422 Page 2
Question #58 McGuire Nuclear Station RO Exam i
Bank Question: 423 Answer: B 1 Pt(s) Which one of the following monitors is required to be operable in mode 4 to
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provide a input to the containment ventilation isolation (Sa) signal? EMF-35, EMF-36, EMF-37 unit vent monitors EMF-38, EMF-39, EMF-40 containment airborne monitors EMF-44 containment vent drain monitor EMF-51 A, EMF-51B reactor building activity monitors Distracter Analysis: Incorrect: - does not actuate CVI I Plausible: - monitor a primary release pathway Correct answer Incorrect: - does not actuate CVI Plausible: - monitors containment vent system Incorrect: - does not actuate CVI l
Plausible: - a high range containment radiation monitor
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l For Official Use Only Ques _423 Page 2 i l
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Question #59 McGuire Nuclear Station RO Exam i
Bank Question: 424 Answer: B 1 Pt(s) Unit 1 is responding to a small break LOCA. Given the following conditions and events:
. NCPs have been stopped
. RVLIS DP Range = 2%
. RVLIS Upper Range = 64%
. RVLIS Lower Range = 51%
What is the correct interpretation of these Reactor Vessel Level Indication System (RVLIS) readings? % level % level % level the RVLIS system has malfunctioned and cannot be relied upon for valid levelindication Distracter Analysis:
, Incorrect: - the lower tap on the upper RVLIS range is at 64% level so this reading indicates off-scale low Plausible: - this is one of 3 possible indications provided - if the candidate does not understand how the system works Correct answer Incorrect: - the dynamic range does not indicate accurately when the NCE are not running Plausible: - this is one of 3 possible indications provided - if the candidate does not understand how the system works Incorrect: - the system indications are normal for this condition Plausible: - if the candidate cannot chose between the three indications, this option is possible
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Question #60 McGuire Nuclear Station RO Exam ;
Bank Question: 425 Answer: A
1 Pt(s) Unit 1 is responding to a LOCA. Operators delayed terminating safety I injection upon reaching the applicable step in E- In which one of the following scenarios is the repressurization of the NC system above 1845 psig most likely to occur?
) A 1 inch line break with two trains of ECCS operating A 1 inch line break with one train of ECCS operating <
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! A 2 inch line break with two trains of ECCS operating i A 2 inch line break with one train of ECCS operating Distracter Analysis: l Correct answer - the most How and the smallest break size l Incorrect: - less flow causes no repressurization Plausible: - less ECCS Ilow means less heat removal which could imply more pressure sooner ' Incorrect: - larger break size = more flow out of hole = lower break flow pressure Plausible: - ifcandidate thinks larger break size = more inventory loss = less cooling = more pressure Incorrect: - less flow - larger break = lower break flow pressure Plausible: - if candidate does not understand what is going on in the thermal-hydraulics in the NC system
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For Official Use Only Ques._425 Page 2 l
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Question #61 McGuire Nuclear Station RO Exam I
? Bank Question: 426 Answer: B 1 Pt(s) Unit 1 is operating at 50% power. Given the following conditions:
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e Pressurizer pressure is 2235 psig l
e Pressurizer Relief Tank (PRT) pressure is 25 psig
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.- The PRT is being cooled by spraying from the RMWST
. A pressurizer code safety valve is suspected ofleaking by its' seat l
What temperature would be indicated on the associated safety valve discharge RTD if the code safety was leaking by?
REFERENCESPROVIDED-Steam Tables F F F F Distracter Analysis: Incorrect: too high Plausible: Correct answer - Incorrect: too low Plausible: - if apply psig - psia conversion incorrectly Incorrect: - way too low Plausible:
For Official Use Only Ques _426 Page 2
l Question #62 McGuire Nuclear Station RO Exam Bank Question: 427 Answer: A 1 Pt(s) Unit 1 is conducting a natural circulation cooldown in ES-0.2. NV auxiliary spray has been started in step 18 to depressurize the NC system. How is pressurizer level controlled during this cooldown and what effect will auxiliary spray flow have on pressurizer level control? ' Automatic positioning of NV-238 in response to the ILE pressurizer level control system. The auxiliary spray line taps off downstream of NV-238 and auxiliary spray flow has no effect on level contro Automatic positioning of NV-238 in response to the ILE pressurizer level control system. The auxiliary spray line taps off upstream of NV-238 and total charging flow indicated through NY-238 will be reduced by the amount of auxiliary spray flow through NV-2 Manual control of NV-238 and NV-21 by the operator. The I auxiliary spray line taps off upstream of NV-238 and NV-238 must be manually positioned to prevent auxiliary spray from raising pressurizer leve Manual control of NV-238 and NY-840 by the operator. The auxiliary spray line taps in from the ND system and NV-238 must be manually positioned to prevent auxiliary spray from raising pressurizer leve Distracter Analysis: Correct answer Incorrect: - aux spray flow has no significant effect because auxiliary spray taps off the charging header downstream of NV-238 so ILE will not see any change in charging flow Plausible: - if the candidate does not know where auxiliary spray flow taps off the charging header Incorrect: - auxiliary spray taps off the charging header downstream of NV-238 so there is no need to go to manual control Plausible: - if the candidate does not know the difference between NV aux spray and ND aux spray flows Incorrect: - aux spray flow from the ND system is only used when ND is in shutdown cooling mode 4 Plausible: - if the candidate does not know that auxiliary spray flow comes from the NV system charging header through NV-21 For Official Use Only Ques __427 Page 2
Question #63 McGuire Nuclear Station RO Exam Bank Question: 428 Answer: C 1 Pt(s) Unit 1 is operating at 50% during a plant startup with rod control in automatic. An RO trainee attempts to take manual control of rods but instead mistakenly moves the Bank Selector Switch to the "CBB" position ,
and moves the rods "IN' for 3 steps with the "In-Hold-Out" switc Which one of the following statements describes the situation? No rod motion occurred because control bank B will not insert until the bank overlap sequence requirement with banks C and D are satisfie Bank B moved into the core but this poses no problems as long as AFD and QPTR are satisfied for this condition - which is likely at 50% powe Bank B is below the rod insertion limits and must be restored to the full out position i Control rods are out of alignment and the operator must 1 immediately implement AP/1/A/5500/14 case IV, Control Rod Misalignmen Distracter Analysis: Incorrect: - the bank sequencer does not prevent control rod motion when rods are in manual Plausible: - this would be true with rods in automatic control Incorrect: - control bank B rods are below the RIL for power operation Plausible: - AFD and QPTR are part of the concems - but so is shutdown margi Correct answer - Do not expect candidates to memorize the RIL values in the COLR - but certainly should know that at 50% power, banks A and B are below RIL Incorrect: - does not meet entry conditions for this procedure - rods not misaligned from group indication or from other rods in group Plausible: - rods are misaligned
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For Official Use Only Ques _428 Page 2
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i Question #64 McGuire Nuclear Station RO Exam {
Bank Question: 429 Answer: A 1 Pt(s) Unit 1 is operating at 50% power with one bus line in operation. Given the following indications:
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. IPB fan IB is tagged out ofservice '
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"IPB Trouble" alarm annunciates on AD-11
The NLO reports that the intemal alaml " Vibration High Fan lA "is illuminated on the local alann panel
. What action must be taken to comply with plant limits? Transfer loads to the second bus line and shut down the operating bus line isophase bus within 15 minutes Reduce load on the bus line to 10,000 amps within 15 minutes Reduce load on the bus line to 20,000 amps within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Reduce load to less than 48% reactor power within 30 minutes Distracter Analysis: Incorrect: - this action will not unload the Isophase bus Plausible: - if the candidate thinks that the isophase bus is on the bus line side of the PCBs
. Correct answer Incorrect: - load too high for one bus line - time requirement is too long ;
Plausible: ' the load restriction of 20,000 amps is for two bus lines I in operation. The time requirement is typical of many important LCOs Incorrect: - not required for a loss of i Plausible: - conect required action by AP-03 for a 50% load rejection due to a loss of CF pump except that the required action is to reduce load below 50% - and IPB the loss of an IPB will cause a load rejection I
For Official Use Only Ques __429 Page 2
l Question #65 McGuire Nuclear Station RO Exam j
Bank Question: 430 Answer: D
1Pt(s) Unit 1 is responding to a large break LOCA into containment. Given the following events and conditions:
Containment spray auto-actuated and reduced containment pressure
. Containment pressure now = 0.24 psig - increasing 1
. NS actuation logic has not been rese Which one of the following describes the NS system response to an increase in containment pressur NS pumps will start and discharge valves will open after containment pressure reaches 0.35 psi NS pumps will start and discharge valves will open after containment pressure exceeds 0.8 psi NS pumps will start when containment pressure reaches 0.35 psig and discharge valves will open when containment pressure exceeds 0.8 psi Discharge valves will open when containment pressure exceeds 0.35 psig and NS pumps will start when containment pressure exceeds 0.8 psi Distracter Analysis: Incorrect: - pumps do not start at 0.35 psig Plausible: - valves open at 0.35 psig by CPCS Incorrect: - valves do not open at 0.8 psig Plausible: - pumps start at 0.8 psig by CPCS Incorrect: - pumps do not start at 0.35 psig and valves do not open at 0.8 psig Plausible: - this is backwards to what happens Correct answer For Official Use Only Ques _430 Page 2
Question #66 McGuire Nuclear Station RO Exam Bank Question: 431 Answer: B 1 Pt(s) Unit 1 is responding to a station blackout. Given the following conditions and events:
- Power has been restored to 1 ETA and 1ETB by starting the D/Gs
The KF air handling unit (AHU) control switch is in the AUTO position Which statement describes the starting sequence of the KF pumps and AHUs aller power is restored to the safety buses? KF pumps start automatically KF AIIUs start automatically after the KF pumps start KF pumps must be manually started from the control room KF AHUs start automatically after the KF pumps start KF pumps start automatically KF AllUs must be manually started from the control room KF pumps must be manually started from the control room KF AllUs must be manually started from the control room Distracter Analysis: Incorrect: - KF pumps must be manually started Plausible: - KF is an important load Correct answer Incorrect: - KF pumps must be manually started, AHUs start in auto Plausible: - reverse case of true starting sequence Incorrect:- AHUs start automatically Plausible: - important loads frequently start in auto For Official Use Only Ques _431 Page 2
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Question #67 McGuire Nuclear Station RO Exam Bank Question: 432 Answer: C 1 Pt(s) Unit 1 is shutting down in mode 5 with all equipment hatches installed, preparing for refueling. Given the following conditions and events:
- A containment entry is planned for the next shift
. The SRO directs the RO to purge the containment in preparation for the containment entr *
The VP system is secured with all fans off and containment purge and exhaust valves closed Which one of the following statements is correct regarding the purge of containment? The containment should be purged using the " REFUEL" mode Only the upper containment should be purged using the
" REFUEL" mode The containment should be purged using the " NORMAL" mode Only the lower contamment should be purged using the
" NORMAL" mode Distracter Analysis: Incorrect: - this will over-pressurize the upper containment Plausible: - getting ready to " refuel" the plant Incorrect: - this will over-pressurize the upper containment Plausible: - getting ready to " refuel" the plant Correct answer: - Incorrect: - should not purge lower containment in mode 1-4 Plausible: - the normal mode is the correct mode to use For Official Use Only Ques _.432 Page 2
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Question #68 f4cGuire Nuclear Station RO Exam
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Bank Question: 433 Answer: A 1Pt(s) Unit 2 is responding to a SGTR on the 1C S/G in E-3. Which one of the following sets of S/G level indications is expected for the IC S/G as it fills
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up with NC system coolant?
0200 0201 0202 0203 0204 IC S/G WR level = 72 % 74 % 75 % 75 % 75 %
1C S/G NR level = 98 % 100% >100%>100% >100% C S/G WR level = 72 % 74 % 77 % 83 % 86 %
1C S/G NR level = 98 % 100% >100%>100% >100% C S/G WR level = 90 % 92 % 93 % 93 % 93 %
1C S/G NR level = 98 % 100% >100%>100% >100%
l C S/G WR level = 90 % 92 % 96 % 100 % 100 %
1C S/G NR level = 98 % 100% >100%>100% >100%
Distracter Analysis: I
' Correct: WR level is lower than NR level because it is calibrated for cold conditions. The WR variable leg tap and the NR variable leg tap are at the same level Incorrect: - WR S/G water level variable leg tap is at the same level as the NR tap. WR level will stop increasing when NR level reaches 100 %
Plausible: - ifcandidate does not know that taps for WR and NR l are at same level (design feature) - overlap between NR and WR ranges is correct Incorrect: - WR level and NR level not correctly correlated - WR level does not continue to rise after NR level reaches the upper tap setting ;
Plausible: - if not familia: with cold cal of WR level and overlap l with NR level range Incorrect: - WR level and NR level not correctly correlated with
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cold cal of WR level- WR level too high for NR level Plausible- if not familiar with cold cal of WR level and overlap with l- NR level range For Official Use Only Ques _433 Page 2
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l Question #69 McGuire Nuclear Station RO Exam Bank Question: 434 Answer: C ,
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1 Pt(s) Unit 1 is responding to a faulted S/G inside containment from 100% powe Given the following conditions and events:
- Containment pressure = 3.5 psig - decreasing
- Faulted S/G pressure = 200 psig - decreasing
- Intact S/G pressure = 850 psig - stable e NC system pressure = 1750 psig - stable What is the MINIMUM set of actions that will allow the operator to manually reset and open the S/G PORVs? Reset MSI Reset hi-hi containment pressure Reset MSI Block Low steamline pressure SI Reset MSI Reset hi-hi containment pressure Block low steamline pressure SI Reset MSI Distracter Analysis: Incorrect: - resetting MSI is NOT sufficient to allow opening a PORV for the plant conditions because the low steam line pressure SI signalis stilllocked in Plausible: - if candidate thinks that the low steamline pressure signal is not locked in or has auto-cleared below P-11 Incorrect: - MSI cannot be reset without first resetting steamline low pressure SI Plausible: - hi-hi containment pressure is the only MSI actuation signal that need not be reset to allow resetting MSI Correct: - although the NC system is < 1955 psig and P-11 is not presently "in" - low steamline pressure SI was interlocked with P-11 when the event occurred so the signal is locked in and must be blocked Incorrect: - not the minimum set of actions - no need to reset hi-hi containment pressure as MSI can be reset with this signal still in ( Plausible: - if candidate thinks that hi-hi containment pressure must be reset.
t For Official Use Only Ques _434 Page 2 a
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Question #70 McGuire Nuclear Station RO Exam Bank Question: 435 Answer: B 1 Pt(s) Unit 1 is operating at 50% power when vacuum in the unit Imain condenser
! starts to fall. Unit 2 is in cold shutdown. Which one of the following j conditions would cauce this problem? l AS-9 ("C" Heater Bleed Steam Reducer) has failed l AS-11 (Main Steam Reducer) has failed l AS-4 (40 psia Auxiliary Steam Header Reducer) has failed l l \ I AS-74 (Auxiliary Steam Cross connect) was shut Distracter Analysis:
l A- Incorrect: - at 50% power, the "C" Heater Drain supply to aux.
l steam is secured and l Plausible: "C" Heater Drain supply is used for power levels >
80% to supply the condenser nir ejectors Correct: - aux. steam to the condenser air ejectors is supplied
! through 1 AS-11 from main steam when operating between 15% and 80% power Incorrect: - the condenser air ejectors are operated from aux.170 psig steam not 40 psig aux. steam j
l Plausible: - there are many aux. Steam loads that are operated from l 40 psig aux. steam
! Incorrect: - unit 2 is in shutdown and not supplying steam - AS-74 is the unit cross connect valve for the 170 psig aux. steam header Plausible: - there have been several recent instances when AS-74 was shut and aux. steam was being supplied from the other uni Shutting AS-74 under these conditions would cause a loss of au I l
steam pressure to the condenser air ejectors l
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t For Official Use Only Ques. 435 Page 2 I
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Question #71 McGuire Nuclear Station RO Enam Bank Question: 436 Answer: B 1 Pt(s) Unit 1 is responding to a station blackout. Given the following conditions
.and events:
. A fault occurred on the 1ELXA load center that supplied power to the 1 A D/G battery charger - causing this load center to remain deenergized until after repairs can be made
. The operators are ready to start the 1 A diesel to restore power on 1 ETA What effect will the loss of power on 1 ELXA have on diesel generator 1 The diesel will start and run normally The diesel will only run for approximately 30 minutes The diesel will start but the output breaker will not close The diesel will not start Distracter Analysis: Incorrect: - the diesel will run until the DG battery is exhausted which lasts for only ~30 minutes Pi:usible: - if the candidate thinks that 120 VAC control power can tx restored without powering ELXA Correct: Incorrect: - 125 VDC control power provides the needed electrical power to start and load the DG ;
Plausible: - if the candidate believe that DG output breaker control l power comes from 120 VAC instead of 125 VDC Incorrect: - the DG is designed to start without 120 VAC power being available Plausible: - if the candidate thinks that 120 VAC control power is needed to start the diesel - or is provided by vital instrument power or Official Use Only Ques _.436 Page 2
Question #72 McGuire Nuclear Station RO Exam i
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? Bank Question: 437 Answer: C 1Pt(s) Which one of the following indications are correct for the existence of a ground on the positive leg of the 125 VDC electrical system? Battery ground positive leg light burns dimly Battery ground negative light is out Battery ground positive leg light burns dimly Battery ground negative leg light burns brightly Battery ground positive leg light burns brightly Battery ground negative light is burns dimly Battery ground positive leg light is out Battery ground negative light is burns dimly Distracter Analysis: Incorrect: - a brightly burning lamp = grounded condition Plausible: - the positive leg is lit Incorrect:- reverse of correct answer Plausible: - if candidate reverses the correct answer in his/her mind
- a grounded condition ollen leads to a reduction in voltage and lamps glow dimly Correct Answer: Incorrect:- positive leg lamp is out Plausible: - if the candidate thinks a grounded condition would cause the light to extinguish I
For Official Use Only Ques _437 Page 2 l
Question #73 McGuire Nuclear Station RO Exam
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Bank Question: 438 Answer: D l
1 Pt(s) During a blackout start of a diesel generator, what causes the DG field flashing to occur? Residual magnetism in the DG generator windings VAC vitalinstrument power VDC control power from the battery charger VDC DG control power from the DG battery l
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Distracter Analysis:
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) Incorrect: - field flashing provided by the DG battery i Plausible: - this is the " Navy" answer - and is used by some DGs Incorrect: - field flashing provided by the DG battery l Plausible: - 125 VAC vital instrument power is extremely reliable ! Incorrect: - although the battery charger supplies 125 VDC DG control power which supplies field flashing current during normal 4 starts, this charger is not powered during blackout starts (where 120 VAC control power is not present)
Plausible: - the answer for non-blackout starts Correct answer
For Official Use Only Ques _438 Page 2
l l Question #74 McGuire Nuclear Station RO Exam Bank Question: 439 Answer: B
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1 Pt(s) What automatic action (s) will occur if radioactive particulate levels exceed the 1 EMF-35(L) trip 2 alami setpoints in the unit vent exhaust flow stream? Closes WG-160 to terminate waste gas discharge
, Stops Auxiliary Building unfiltered exhaust fans Stops Spent Fuel Pool ventilation exhaust fans Stops Containment Annulus ventilation fans Distracter Analysis: l Incorrect: - does not effect WG-160 Plausible: - trip 2 alarm on EMF-36(L) (Unit Vent Gas) causes this action Correct answer - 1 EMF-35(L) monitors this exhaust stream Incorrect: - does not stop these fans Plausible: - EMF-35 monitors exhaust from Spent Fuel Pool area Incorrect: - does not stop VE fans Plausible:- monitors VE exhaust
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Question #75 McGuire Nuclear Station RO Exam Bank Question: 440 Answer: D 1 Pt(s) Unit 1 is shutdown to mode 5 in a refueling outage with a normal electrical power lineup. Which one of the fellowing statements describes the proper configuration of IND-2AC under these conditions? IND-2AC is closed and MOV power is removed from 1EMXA IND-2AC is closed and MOV power is removed from SMXG IND-2AC is open and MOV power is removed from SM* G ll IND-2AC is open and MOV power is removed from IEMXA ('
Distracter Analysis: ,
l Incorrect: - 1ND-2AC must be open in mode 5 Plausible: - if candidate does not realize that the valve must be open Incorrect: - IND-2AC must be open in mode 5, the alternate power I supply is SMXG Plausible: - if candidate does not realize that the valve must be open and SMXG is the alternate power supply Incorrect: - 1EMXA is the normal power supply Plausible: - SMXG is the attemative power supply , Correct answer: I
For Official Use Only Ques _440 Page 2
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l Question #76 McGuire Nuclear Station RO Exam j Bank Question: 441 Answer: A l
i 1 Pt(s) Unit 1 is operating at 100% power when the following alarms are received:
I e PRT Hi Temp e PRT Hi Press '
Assuming that the alarms were due to the slow heat up of the PRT from containment ambient conditions, how can the operators clear these alarms? Vent the PRT to the waste gas system Recirculate the PRT water through the NCDT Heat Exchanger j l Vent the PRT into the containment through the HEPA filter !
Recirculate the PRT water through the WDT Heat Exchanger j l Purge the PRT with Nitrogen to the shutdown tank via the waste gas header Spray the PRT from the RMWST while draining to the WDT Purge the PRT with Nitrogen to the VCT Spray the PRT from the RMWST while draining to the NCDT Distracter Analysis: I Correct answer Incorrect: - the WDT is not used to drain the PRT. In addition, this path requires containment entry and would not be approved at 100%
powe Plausible: - WDT accepts radioactive waste from a number of primary systems and has a heat exchanger, the PRT can be vented into containment through a manual valve Incorrect: - cannot drain to the WDT Plausible: - Purging with N2 to the WG system is an acceptable method of reducing pressure Incorrect: - cannot purge to the VCT Plausible: - spraying from the RMWST while draining to the
. NCDT is an alternative method for reducing temp.
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Question #77 McGuire Nuclear Station RO Exam l
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Bank Question: 442 Answer: D 1 Pt(s) Unit 2 is operating at 100% power when a leak develops from an NCP ( thennal barrier heat exchangen. What design feature (s) prevent a small leak i from over-filling the KC surge tank and overflowing into the Auxiliary
- Building? The KC surge tank vent line willisolate on high surge tank level and a relief valve will prevent overpressure of the surge tank by relieving at 100 psig and draining to the floor drain syste The KC surge tank has an overflow line to drain directly into the KC drain tan l The KC surge tank overflow line will isolate on high liquid radiation levels and flow will be redirected from the Au I Building to the KC drain tank l I The KC surge tank vent path willisolate on high liquid f
radiation levels if the system becomes contaminated and the KC ]
relief valve will prevent over-pressure on the surge tank by relieving at 15 psig and draining to the floor drain syste Distracter Analysis: Incorrect: - the surge does not isolate on high level - the relief valve operates at 15 psi Plausible: - if the candidate does not know about EMF-46 Incorrect: - does not have an overflow line that drains to the KC drain tank Plausible: - the KC drain tank accepts drainage from other parts of the KC system Incorrect: - relief valve does not drain to the KC drain tank - relief valve setpoint is 15 psig Plausible: - the KC drain tank accepts drainage from other parts of the KC system Correct answer
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For Official Use Only Ques __442 F uge 2
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Question #78 McGuire Nuclear Station RO Exam Bank Question: 443 Answer: B 1 Pt(s) Unit I was responding to a LOCA inside containment when the following
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alarms were received in the control room:
e lA VE Filter Bed Hi Temp on the annunciator panel e
VE Filter Hi Temp > 220 'F on the HVAC panel e VEFilter Fire > 325 F on the HVAC panel What complete set of automatic actions are expected as a result of these alarms?
f stops VE fans l
l stops VE fans and shuts off filter heaters l
l stops VE fans and starts RFY deluge stops VE fans, shuts off filter heaters and starts RFY deluge Distracter Analysis: Incorrect: - also shuts off filter heaters Plausible: - if candidate does not know that filter pre-heaters are l interlecked to this alarm l Correct arswer Incorrect: - does not start RFY deluge automatically Plausible: - opening deluge valves is a manual action if there is a fire in the filter bed Incorrect: - - does not start RFY deluge automatically Plausible: - opening deluge valves is a manual action if there is a fire in the filter bed I
For Official Use Only Ques __443 Page 2
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Question #79 McGuire Nuclear Station RO Exam Bank Question: 444 Answer: B 1 Pt(s) Which one of the following statements describes the operation of the Containment Annulus Ventilation System (VE) during a large break LOCA into containment? VE fans start in the exhaust mode at +3.0 psig in containment VE fans stop running when annulus pressure =-7.0 in. Hg VE fans cycle on and olibetween -!.2 in Hg and -7.0 in. Hg VE fans start in the exhaust mode at +3.0 psig in containment VE shifts into recire mode at annulus pressure =-4.2 in. Hg VE cyces dampers between recire and exhaust modes between-1.2 and -4.2 in. Hg VE fans start at on an EMF-38,39,40 (containment monitors)
VE fans stop running when annulus pressure = -7.0 in. Hg VE fans cycle on and off between -1.2 in Hg and -7.0 in. Hg VE fans start at on an EMF-3 (containment monitors)
VE shi'h into recire mode at ahuutus pressurc = -4.2 in. Hg VE cycles dampers between recire and exhaust modes between-1.1 and -4.2 in. Hg Distracter Atalysis: Incorreed - VE Fans are not designed to cycle on and off to maintain annulus pressure - they trip off at -7.0 in. Hg but reset at-5.5 in. Hg. They would not recirc through the filter trains if they tripped off to maintain pressure and Iodine removal would not be as effective Plausible: - tripping a: -7.0 in. Hg is a safety trip to ensure annulus pressure is not dropped too low Correct answer Incorrect: - VE fans are not started by EMF monitors, VE Fans are not designed to cycle on and off to maintain annulus pressure - they ,
trip off at -7.0 in. Hg but reset at i-5.5 in. Hg Plausible: - VP and VQ systems are controlled by EMF 38,39,40 Incorrect: - VE fans are not started by EMF monitors Plausible: - VP and VQ systems are controlled by EMF 38,39,40 For Official Use Only Ques _449 Page 2 l
Question #80 McGuire Nuclear Station RO Exam Bank Question: 445 Answer: C 1 Pt(s) What bus supplies power directly to the 1 A KC pump? IELXA load center to MCC IEMXG load center to MCC ETA bus ITA bus Distracter Analysis: Incorrect: - not powered off 600 VAC essential power Plausible: - many large essential motors are powered off this load center- safety grade power Incorrect: - not powered off 600 VAC shared essential power Plausible: - many large important motors powered off this load center- safety grade power Correct answer- 4160 VAC power Incorrect:- not powered offITA Plausible: - 1ETB is powered from 1TA through SATA transfonner - many large loads are powered directly off this 6.9 kV bus i I
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Question #81 McGuire Nuclear Station RO Exam l
l Bank Question: 446 Answer: C 1 Pt . Unit 1 is responding to a loss of all AC power. Unit 2 is operating at 100%
power. One hour into the event, it becomes necessary to cycle the pressurizer PORV on Unit I to reduce NC pressure. VI header pressure in the control room indicates 25 psi )
l What provides the motive force to cycle this valve under these conditions? j Air pressure from the VI system header Air pressure from the Auxiliary Building Instrument Air Tank Header { Nitrogen pressure from the Cold Leg Accumulators VAC shared essential power ft om unit 2 L)istracter Analysis:
l l Incorrect: - 25 psig is not sufficient pressure to cycle a PORV - VI
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Header is not safety grade and will depressurize in a blackout i
Plausible: - many sites have safety grade VI systems Incorrect: - - does not proside air to PZR PORVs Plausible: - the aux. building air tank provides blackout protection for most AOVs that are required to operate during a blackout Correct answer Incorrect: - the PORV is operated by Nitrogen from the CLA Plausible: - many safety related MOVs are powered off shared essential power load centers l
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For Official Use Only Ques __446 Page 2
l Question #82 McGuire Nuclear Station RO Exam i
l-l Bank Question: 447 Answer: D l
1 Pt(s) Unit 1 is shutdown, mode 6, in a refueling outage. Given the followin; conditions:
i e Containment airlock doors are both open e A full shift of qualified maintenance personnel are available inside containment
. The Refueling SRO is in the control room
. The Operations Fuel Handling Supervisor is inside containment Refueling has been completed and the Operations Fuel Handling Supervisor (who is not a qualified SRO) requests permission to latch all control r, .is to prepare for the reactor startup. What additional req 6ements must be met (if any) to proceed with latching rods?
I Latching rods may proceed at the discretion of the Fu21 Handling Supen'iso Latching rods may not proceed until after containment integrity has been restore Latching control rods may not proceed until after the Refueling SRO arrives inside containment to supervis Latching control rods may not proceed until after the Refueling l SRO arrives inside containment and containment integrity has been restored.
l Distracter Analysis: Incorrect: - the Refueling SRO is required to supervise this l
evolution and containment integrity must be restored l
Plausible: - if the candidate does not recognize that latching rods is a core alteration or doesn't recognize that this requires containment l integrity to be established l Incorrect: - the Refueling SRO it cquired to supervise this
! evolution Plausible: - if the candidate does not recognize that latching rods is a core alteration I Incorrect: - containment integrity must first be established Plausible: - core alterations requires SRO coverage and containment integrity Correct answer For Official Use Only Ques _447 Page 2
ra Question #83 McGuire Nuclear Station RO Exam Bank Question: 448 Answer: C 1 Pt(s) Unit 1 is conducting a plant startup and power ascension after a refueling outage when a pressurizer PORY failed open. Given the following events and conditions:
- The reactor had not tripped prior to 0200
. The P-13 light extinguished prior to 0200 e Operators were in the process of transferring CF control from the CF bypass to the main CF control valve on the A S/G at 0200
- The operators did not manually trip the reactor at any time
- *The operators tripped the main turbine at 0220
. Assume that all redundant instrument channels are in service and read !
within 0.5% or 1.0 psig and un trip setpoints are within required tolerances i
Given the following conditions and events: i l
l 0200 0210 0220 0221 )
Reactor Power (%) 12 % 9% 8% 5% l Turbine impulse Press (%) 12 % 8% *0% 0% !
PZR pressure (psig) 1946 1942 1948 1955 1 A S/G level (%) 35 19 18 16 IB S/G level (%) 40 35 39 41 1C S/G level (%) 38 35 40 39 1D S/G level (%) 35 32 38 42 At what time (if any) would the reactor have been in a tripped condition? the reactor would not be tripped Distracter Analysis:
' Incorrect: - P7 is"out" (<10% for Rx power and Turb imp) so the at-power reactor trips are blocked - low pressurizer pressure (<l944 psig)is blocked F r Official Use Only Ques _448 Page 2 l
Question #83 McGuire Nuclear Station RO Exam Plausible: -if the candidate does not recognize the P7 blocks the low pressurizer pressure trip < 1944 psig Incorrect: - the main turbine trip does not cause a reactor trip below P8 (<48%)
Plausible: - if the candidate does not recognize that the main turbine trip is blocked < P8 Correct answer: - the reactor trips on low SGWL in I A S/G - this trip is not blocked Incorrect: - the reactor has tripped Plausible: - if the candidate thinks that low SGWL is blocked by P7/P8 or does not recognize that the low SGWL trip is "in"
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For Official Use Only Ques _448 Page 3 l
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l Question #84 McGuire Nuclear Station RO Exam
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Bank Question: 449 Answer: D l
! 1 Pt(s) Unit I was commencing a plant start up when a flooding event occurred l inside containment. The operators entered E-0 and transitioned to FR-Z.2, Containment Flooding. Chemistry reports that the water inside containment has a high level of free Chlorin Which one of the following systems is the most probable source of the water? RV System RN system KC system RF system Distracter Analysis: Incorrect: - RV system water does not contain free Chlorine Plausible: - if candidate does not know chemistry requirements for RV water Incorrect: - RN system contains water from Lake Norman - does not contain Chlorine Plausible: - Asiatic clams are a problem in the RN system and Chlorination is one way of destroying the larva l h4 correct: - KC contains Chromates but not Chlorine for chemistry i control Plausible: - if the candidate does not know chemistry requirements j for KC water Correct answer - the RF system is Chlorinated to prevent Asiatic Clam degradation
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For Official Use Only Ques _449 Page 2
Question #85 McGuire Nuclear Station RO Exam Bank Question: 450 Answ ::D 1 Pt(s) Unit 1 was operating at 100% power when a loss of offsite power occurred I on Unit 1. Given the following events and condition ;
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a Unit 2 continued to operate at 100% power and was not affected by the LOO . The B and C KR pumps were operating when power was lost j Which one of the following statements describes effect of this event on the i
plant?
l The NF chillers must be manually tripped immediately or else they willice up and damage will occur The VI air compressors will automatically trip and a loss of VI will occur The KR system flow will be restored to operation when the D/Gs sequence on their safety buses The KR system flow will continue at reduced levels during the l
blackout Distracter Analysis: Incorrect: - the NF chiller have an interlock that stops the chillers if KR flow is lost Plausible: - running a chiller without KR flow will cause damage to the chiller Incorrect: - power is not lost to the KR pumps Plausible: - if power lost to the KR pumps, the VI compressors have an interlock that would prevent them from operating Incorrect: - KR pumps are not powered from safety busses Plausible: - the KR system is not safety grade but it is very important to station operation Correct answer - KR power is normally maintained during a unit LOOP because B KR pump is powered from Unit 2 shared essential power For Official Use Only Ques __450 Page 2
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Question #86 McGuire Nuclear Station RO Exam I
Bank Question: 451 Answer: B y l
1 Pt(s) Unit 1 is shutdown in a refueling outage. A VI header rupture occurs and the VI system completely depressurizes. VI-820 was open at the time of the rupture. What effect does a total loss of the VI system have on the VS system? VI-820 will auto-close as VI header pressure decreases below 90 l
psig and the VS air compressor will start automatically at 82 psig to maintain VS header pressure i VI-820 will auto-close as VI header pressure decreases below 82 psig and the VS air compressor must be manually started to maintain VS header pressure
] Check valves in the VI - VS cross-connect line will close to isolate VS system pressure before it drops below 90 psig 1 VS pressure in the Fire Protection Pressurizer Tank will be lost )
until a VS air compressor can be starte Distracter Analysis:
, Incorrect: - the VS air compressor does not automatically start to l maintain pressure - VI-820 auto-closes at 82 psig not 90 psig Plausible: - The system is safety significant, VI-820 does close at 82 4 psig and there is a separate VS air compressor - it just isn't in standby - requires operator action Correct answer Incorrect: - there are no check valves in this line Plausible: - this is another possible method to prevent depresseridng the V5 header Incorrect: - the RF system tank is pressurized with high pressure VS air - which is maintained isolated from the VI header Plausible: - if the candidate does not know that high pressure VS is isolated For Official Use Only ques _451 Page 2 ,
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Question #87 McGuire Nuclear Station RO Exam
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Bank Question: 459 Answen D
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1 Pt(s) Unit 1 is operating at 100% power when there are multiple indi.:anons of a fire in the turbine building. Who is the FIRST person who has the authority by procedure to activate the FULL Fire Brigade without first sending someone to investigate at the scene?
(e.g. Ifthe B, C and D are all valid answers, then the FIRSTperson on the list is answer B) Plant Security Guard on rounds in the turbine building NLO making rounds in the turbine building Work Control Center SRO Operations Shift Manager Distracter Analysis: Incorrect: - cannot call for full activation l Plausible: - available in the building while making rounds Incorrect: - cannot call for full activation Plausible: - responsible for area Incorrect: - not the first person on the list Plausible: - he is in charge of many other important area ) Correct answer - by RP/0/A/5700/025 Enci ,
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For Official Use Only Ques _.459 Page 2
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l Question #88 McGuire Nuclear Station RO Exam Bank Question: 461 Answer: C 1 Pt(s) Unit 1 is cooling ds .m for a refueling outage. Given the following initial conditions and events:
. Tave = 475 F !
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e Pressurizer pressure = 1850 psig
- Pressurizer level = 50%
e S/G levels = 35% in all four S/Gs If PORV NC-32 failed a surveillance test (the valve failed to open and could not be cycled manually) which Tech Spec actions (if any) are required within one hour? h Close the block valve for PORY NC 32 Close and remove power from PORV NC 32 Close the block valve for PORV NC 32 and remove power from the block valve No actions are required within one hour - restore PORY NC 32 to an operable status within 7 days Distracter Analysis: Incorrect: - not an applicable if PORV cannot be cycled manually Plausible: - appropriate for a failed PORV in mode 3 that CAN be manually cycled Tech Spec 3.4.11 condition A Incorrect: - not applicable to failed PORV Plausible: - appropriate for a failed PORV block valve in mode 3 Tech Spec 3.4.11 condition C Correct answer - Tech Spec 3.4.11 condition B > Incorrect: - not in mode 4 Plausible: - appropriate for mode 4 Tech Spec 3.4.12 condition E
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For Official Use Only Ques _461 Page 2
t Question #89 McGuire Nuclear Station RO Exam Bank Question: 462 Answer: C 1 Pt(s) Which one of the following actions is NOT allowed be taken in advance of the SRO reaching and reading the step in the applicable EP?
l CA may be reset and controlled to a S/G when that S/G reaches j i
its normallevel setpoint NC pumps may be tripped on an Sp signal due to a loss of I support funct!ons unless the procedure has directed otherwise Safety injection may be reset Known leaks may be isolated if plant conditions require it to be j done Distracter Analysis: Incorrect: - allowed per OMP 4-3 Plausible: - if candidate does not know the OMP requirements l Incorrect:- allowed per OM 3 Plausible:-if candidate doct m know the OMP requirements l' Correct answer: - not on OMt 4-3 list of approved actions Incorrect: - allowed per OMP 4-3 Plausible: - if candidate does not know the OMP requirements For Official Use Only Ques _.462 Page 2
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Question #90 McGuire Nuclear Station RO Exam
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Bank Question: 463 Answer: D l 1 Pt(s) Which one of the following is the proper method of verifying the position of l an ND system manual rising stem globe valve in the " locked open" position if that valve has a lock and chain installed to prevec elosure? Verify that the lock is installed and prevents movement in the
! closed direction Verify that the valve stem is in the full up position Verify that the lock is installed and prevents movement in the closed direction Turn the valve handwheel to the open direction to verify that it will not move and is securely seated on the backseat l Remove the lock and chain l Turn the valve handwheel in the open direction to verify that it l will not move and is securely seated on the backseat Reinstall the lock and chain to prevent closure i Remove the lock and chain Turn the valve handwheel in the closed direction to verify that the valve will move Turn the valve handwheel to the open direction Distracter Analysis: Incorrect: - must remove the locking clevice and tum the valve in the closed direction Plausible: - this is true for valves that are position tamper sealed Incorrect: - must first remove the locking device Plausible: - locks will generally permit slight movement in the open direction Incorrect: - never check a manual valve in the open direction as the valve may be closed and moving it to the open direction may inadvertently pressurize a system Plausible: - this would check that the valve was open - unless it was jammed on its seat
, Correct answer - if the valve is open, it will close slightly but this l will have no effect on the system. Ifit is closed, it will not mov For Official Use Only Ques __463 Page 2 I
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Question #91 McGuire Nuclear Station RO Exam i
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l Bank Question: 464 Answer: C l
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1 Pt(s) Unit 2 is operating at 59% power. Which one of the following alarm conditions will cause a slow rate turbine runback on Unit 27 Transmission Line Overload Loss of one Main Feedwater Pump l Loss of Cooling Groups Low Stator Water Cooling Flow I
Distracter Analysis:
l Incorrect: - not a runback on Unit 2 !
Plausible: - this is a runbac'k on Unit 1 only - at the slow rate - I when >56% turbine impulse pressure l
' Incorrect: - causes a fast rate runback Plausible: - applicable to both units Correct answer per OP-MC-GEN-EHC page 61 Incorrect: - causes a fast rate runback Plausible: - applicable to both units
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For Official Use Only Ques _464 Page 2
l-Question #92 McGuire Nuclear Station RO Exam Bank Question: 465 Answer: A
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1 Pt(s) Unit 1 is operating at 15% power going to 100% power. The operators just completed synchronizing the main generator on the power grid. Which one of the following sequences is the correct operating procedure for increasing the main generatorload? Select MW IN Raise the GV limit from 17% to 120%
l Verify REFERENCE and DEMAND windows are matched Depress LOAD RATE pushbutton and enter desired load rate Depress the REFERENCE pushbutton and enter the load Depress the GO pushbutton Select MW IN Raise the GV limit from 17% to 120%
Verify REFERENCE and DEMAND windows are matched !
Depress STANDARD pushbutton and enter desired load and
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l load rate using the keypad j Depress the GO pushbutton Select MW OUT Raise the GV limit from 17% to 100%
Verify REFERENCE and DEMAND windows are tratched Depress STANDARD pushbutton and enter desired load and lead rate using the keypad Depress the GO pushbutton Select MW OUT Raise the GV limit from 17% to 100%
Verify REFERENCE and DEMAND windows are matched Depress the REFERENCE pushbutton and enter the load l Depress LOAD RATE pushbutton and enter desired load rate i Depress the GO pushbutton l
I Distracter Analysis: l l Correct answer: Incorrect:- do not use the STANDARD button I Plausible: - this is designed to work correctly but is not used at McGuire J i Incorrect: - MW must be IN - do not use the STANDARD button l
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Question #92 McGuire Nuclear Station RO Exam Plausible: -if the candidate does not understand the MW IN feedback loop Incorrect: - MW must be IN for feedback loop Plausible:-if the candidate does not understand the MW IN
-. feedback loop I
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Ii Question #93 McGuire Nuclear Station RO Exam
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Bank Question: 466 Answer: A )
1 Pt(s) When is the use ofdouble isolation boundaries NOT required to be considered when preparing a tagout in accordance with NSD 5007 Isolation of high voltage electrical components greater than ;
6.9 kV Isolation against a high temperature energy source with suspected internalleakage l Isolation against a high pressure energy source with suspected internalleakage Isolation of a check valve approved for use as a tagout boundary Distracter Analysis: Correct answer - not required Incorrect: - required by NSD-500 Plausible: - if candidate doesn't know the requirements Incorrect:- re"uired by NSD-500 Plausible: - it candidate doesn't know the requirements Incorrect: - required by NSD-500 Plausible: - if candidate doesn't know the requirements For Official Use Only Ques _466 Page 2 l
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Question #94 McGuire Nuclear Station ' RO Exam Bank Question: 467 Answer: A 1 Pt(s) Unit I refueling is in progres Given the following plant conditions:
- 1 EMF-39, Containment Gas Monitor "Hi Rad" alar . 1 EMF-40, Containment Iodine Monitor "Hi Rad" alar * IEMF-16, Containment Refueling Bridge Monitor alar Which one of the following Abnormal Operating Procedures should be implemented? AP-25, Spent Fuel Damage AP-40, Loss of Refueling Canal Level l AP41 Case I, Loss of Spent Fuel Pool Cooling AP-41 Case II, Loss of Spent Fuel Pool Level Distracter Analysis: l Correct answer Incorrect: - would not see EMF 39,40, alarms Plausible: - would see EMF-16 alarm , Incorrect: - would not see EMF alarms - would get other alarms Plausible: - loss of cooling could lead to Spent Fuel damage Incorrect: - would not get EMF 39,40 alarms Plausible: - would see EMF-17 alarm For Official Use Only Ques _467 Page 2
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i Question #95 McGuire Nuclear Station RO Exam i
l' mk Question: 468 Answer: D
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1 Pt(s) A maintenance worker is preparing to disassemble the internals of a valve that was removed from the NC system in the mechanical maintenance sho This valve has been surveyed by Health Physics with the following results e General radiation levels in the shop work area = <.1 mR/her e Radiation levels on contact with the valve = 0.3 mR/hr e Extemal loose surface contamination = 750 dpm/cm2 p,7
- Intemal loose surface contamination = 5000 dpm/cm2 p,7 l
Which one of the following statements correctly characterizes the radiological work controls that are applicable to work on this valve? No RWP is required because the valve in not inside the RC/. No RWP or SRWP is required because the radiation levels in the work area and on the valve are sufficiently low A SRWP is appropriate because the work is a routine operation on a system component and the radiation levels are sufficiently low ' A RWP is required because the valve is not a routine operation and is internally contaminated Distracter Analysis: Incorrect: - an RWP is required Plausible: - if the candidate thinks that RWPs only apply to activities inside the RCA Incorrect: - RWP is required because ofinternal contamination level Plausible: - the er.ernal radiation levels are < 0.5 mR/hr which is the threshold for requiring an RWP Incorrect: - this work is not a routine operation Plausible: - SWRPs may be used for routine, repetitive tasks l Correct answer - the internal contamination requires an RWP l
For Official Use Only Ques _468 Page 2 l
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Ques; ion #96 McGuire Nuclear Station RO Exam i
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Bank Question: 469 Answer: B t
1 Pt(s) Unit 1 is in the process of making a radioactive gaseous waste release from the waste gas decay tank in accordance with OP/0/A/6200/18. Given the following conditions:
- MRIRR = 35 CFM e EMF-50 trip 1 setpoint = 1.0E5 CPM e EMF-50 trip 2 setpoint = 2.0E5 CPM e EMF-36 is out of service
- The operators reset EMF-50 whenever procedural direction allows Time 0200 0215 0230 0245 Release rate (CFM) 22 37 32 41 EMF-50 (CPM) 1.1ES 2.1E5 2.2E5 2.2E5 What was the earliest time that the operators were reauired to terminate (and not restart) the gaseous releas Distracter Analysis: Incorrect: - EMF-50 automatic protection functions do not actuate until trip 2 Plausible: - some EMFs have trip 1 protective actions Correct answer - release rate > MRIRR and EMF-50 trips and closes release path Encorrect: - release already stopped at 0215 Plausible: - and this is only the 2"d trip of EMF-50 - you could reset EMF-50 and restart this release IF EMF-36 was operable Incorrect: - release already stopped at 0215 Plausible: - exceeded 40 CFM (MSORR) max operation system release rate for the Waste Gas System i
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Question #97 McGuire Nuclear Station RO Exam Bank Question: 470 Answer: B 1 Pt(s) Unit 1 is shutdown in mode 5 during a refueling outage. Containment purge
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has been in progress for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Which one of the following conditions would reauire the containment purge to be immediately terminated? EMF-36(L) becomes inoperable j
l IEMF-39(L) becomes inoperable l
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l l A S/G diaphragm was removed and the monitoring EMF W increases by 75% but remains < trip 1 RP reports that the latest containment air sample shows that the concentration of the isotope Rb5has increased by 15 times Distracter Analysis: Incorrect: - EMF-36 (unit vent monitor) is not the correct EMF monitor for a containment purge release - not required Plausible: - used for waste gas release Correct answer - EMF-39(L) is required to be operable because it is i used to monitor the release rate I Incorrect: - not a requirement to terminate the release rate i Plausible: - > 50% is a requirement to resample the contamment atmosphere i Incorrect: - not a requirement to terminate the release rate Plausible: - this is a requirement to resample the containment atmosphere i
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For Official Use Only Ques __470 Page 2 l
Question #98 McGuire Nuclear Station RO Exam i
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Bank Question: 471 Answer: C 1 Pt(s) Unit 1 is responding to a LOCA. Given the following events and conditions:
- Completed E-0 i e Entered E-1 l e The STA reported the following valid critical safety functions:
. Suberiticality - orange path e Core Cooling - red path 3
- Heat Sink - red path e All other CSFs are green or yellow What procedure should be operator select? I Remain in E-1
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! Transition immediately to FR- f Transition immediately to FR- Transition immediately to FR- Distracter Analysis: Incorrect: - must transition to CSFs Plausible: - if candidate does not know restrictions and applicability of F-0 Incorrect: - Orange path does not have priority over red paths Plausible: - if candidate does not know rules of usage Correct answer: - core cooling has priority Incorrect: - Heat sink does not have priority over core cooling Plausible: - if candidate does not know CSF rules of usage j ll l
For Official Use Only Ques _.471 Page 2
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! Question #99 McGuire Nuclear Station RO Exam
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Bank Question: 472 Answer: B
1 Pt(s) Unit 2 is responding to a SGTR. The RO is directed to locally open the 2A S/G PORV, Where does the operator go and what direction does he/she turn this valve to open manually? Goes to the exterior doghouse Turns the PORV in the counter-clockwise direction Goes to the exterior doghouse Turns the PORV in the direction noted on the tag attached to the valve (may be clockwise or counter clockwise)
l Goes to the interior doghouse l Turns the valves in the counter-clockwise direction I Goes to the interior doghouse )
Turns the PORY in the direction noted on the tag attached to the valve (may be clockwise or counter clockwise)
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Distracter Analysis: Incorrect: - this valve opens in the clockwise direction Plausible: - most valves open by turning in the counter-clockwise direction, right location Correct answer - right location, right opening direction Incorrect: - valve opens counter-clockwise, wrong location - 2A S/G PORV is in the exterior doghouse Plausible: - most valves open by turning in the counter-clockwise direction Incorrect: - wrong location - 2A S/G PORV is in the exterior doghouse Plausible: - the operator should refer to the tag for correct opening direction as not all S/G PORVs open in the same direction
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For Official Use Only Ques _472 Page 2
Question #100 McGuire Nuclear Station RO Exam Bank Question: 473 Answer: C 1 Pt(s) Unit 1 was operating at 100%. I&E is conducting calibration of the VCT tank level instrument and causing periodic VCT !evel alarms due to testin Which one of the following statements correctly describes the proper pmcedural response to these alarms by the RO? The RO may use his/her judgement to determine if announcing the alarms is appropriat The RO must announced the alarm the first time that it actuates, I then announcements may be suspended until testing is !
completed or until reaching the end of the shif l
' The RO must announce all occurrences of the alarm during the testing - unless the control room SRO exempts announcement of alarms due to testing (do not need permission from higher authority). The RO must announce all occurrences of the alarm during the testing- only the OSM can exempt announcement of alarms due to testing (the control room SRO does not have authority to suspend alarm announcements at 100% power).
Distracter Analysis: Incorrect: - RO cannot make this decision on his/her own Plausible: - correct for situations when responding to abnormal or emergency conditions Incorrect: - RO cannot make this decision on his/her own Plausible: - correct ifin shutdown modes Correct answer Incorrect: - the control room SRO does have authority to suspend alarm announcements Plausible: - OSM authority may be required for very important emlutions and decisions l
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NRC Official Use Only N Q s -l6 A SAO 9 9- J O/
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Nuclear Regulatory Commission Senior Reactor Operator Licensing Examination l
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Final Version for Approval McGuire Nuclear Station !
l Questions with Answers
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This document is removed from Official Use Only category on !
date ofexamination ;
i NRC Official Use Only DISTRIBUTION CODE r
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McGuire Writt::n Exam Sample Plan Facility: McGuire Date of Exam: 2/5/99 Exam Level: SRO l
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K/A Category Points Point l
Tier Group K K K K K K A A A A 1 2 3 5 1 2 g Target 1 1 2 4 7 5 4 2 24 24 j Emergency & 2 5
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4 5 1 2 16 16 Abnormal 3 1 0 1 0 0 1 3 3
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Plant Tier Evolutions Totals 6 5 12 10 5 5 43 43 j 1 1 1 1 1 2 1 4 5 1 2 0 19 19 1 2 2 1 1 2l1 l2l2l1 5l1 l0l1 17 17 Plant 3 0 2 0l1 l0l0l0 0l0l1 l0 4 4 Systems Tier i Totals 2 4 3l3l4l3 5 10 l 2 l 3 l 1 40 40 )
3 Generic Knowledge and l Cat 1 l Cat 2 Cat 3 l Cat 4 Ablittles l4 l 5 4 l 4 17 17 Note: * Attempt to distribute topics among all KlA categories; select Totals at least one topic from every K/A category within each tie !
- Actual point totals must match those specified in the tabl * Select topics from many systems; avoid selecting more than two or three KIA topics from a system unless they relate to plant specific prioritie * Systems / evolutions within each group are identified on the associated outlin * The shaded areas are not applicable to the category / tie ** Denotes plant specific, high priority K/As
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McGuire Written Exam Sample Plan ES-401 Generic Knowtedge and Abilities Oucine (Tier.1) Form ES-401-6 Facility: McGuire Date of Exam: Exam Level: SRO Bank Categcry K/A 8 Topic im Pointe Queettors 2.1.29 Knowledge of how to conduct and verify valve lineups 3.4/ ,
6, to r%, e hhinai. for system operstmg parameters which are entry 6 wel conditions for technical specsfications l 2.1.33 3.4/ t Conduct of 2.1.20 Ability to execuw procedure steps 4.3/ Acety to recogswze ir& iu,,e for system operating parameters wtuch are entry Operatione 2.1.33 level conditions for technical specr6 cations 31/ l
!
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,
Total 4 2. Knowledge of design, procedural and operabonal differences between units 3.1/ Nw.c of the promse for managing maintenance actMbos dunne shutdown l 2.2.18 operanov 2.3/3,6 1 400 Equipment 2.2.13 Knowledge of tagomo and clearance promdures 3.6/ Control 2.2.27 Knowledge of refueling process 2.6/ Knowledge of the process for determerwng if the proposed change, test or 2. experiment involves and unreviewed safety question 1.8/ l Total 8 Nw-@ of i.J ;~r, exposure limits and contamina9on control, including 2. pomussible levels in excess of those authortred 2.5/ * At4ty to control radiation releases .7/ l Radiation 2. Knowledge of process for performing a containment purge 2.5/ ADWty to psib,,, procedures to reduce excessive levek of FMlaDon and guard Control 2.3.10 against personnel exposure 2.9/ l l
Total 4 Kr=w;.6,w of t;UP implementabon hserarchy and coordmabon with other Support 2.4.16 procedures 3.0/40 1 478 1 4.44 Knowledge of emergency plan protective action recommendabons 2 il4 0 1 491 Emergency 2.4.46 Ately to vertfy that alarms are consistent with plant conditions 3.5/36 1 473 l
Procedures 2.4.11 Knowledge of abnormal condition procedures 30/ a86 l I
and Plan I Total 4 Tier 3 Point Total 17 17l l
I l
l
.
ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet i
U.S. Nuclear Regulatory Commission Site-Specific Written Examination
,
Applicant Information Name: Region:
I/@/III/IV Date: Facility / Unit: Mccuire License Level:
Start Time:
R0
/@ Reactor Type:
Finish Time:
@/CE/BW/GE
,
Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of-the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination start Applicant Certification All work done on this examination is my ow I have neither given nor received ai Applicant's Signature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent NUREG-1021 39 of 39 Interim Rev. 8, January 1997
,
Question #1 McGuire Nuclear Station SRO Exam l
Bank Quesdon: 366 Answer: D
'
1Pt(s) Unit I was operating at 100% power with control rods in automatic when the Reactor Operator observed the following indications:
l Indir . I Rod Position Indication for control bank D rods:
.
M-12 = 220 steps D-4 = 220 steps M-4 = 223 steps D-12 = 210 steps H-8 = 223 steps Control Group D Step counter = 221 steps Which of the following sets ofimmediate actions (if any) must be taken? No operator action is required j The system will maintain automatic control ;
1 Adjust turbine load to malatain Tave = Tref Monitor AFD and QPTR Place control rods in manual and position to maintain Tave = Tref Monitor AFD and QPTR Place control rods in manual Adjust turbine load to maintain Tave = Tref Distracter Analysis: Incorrect: - rod position for CD3 and CD-4 are > 12 steps difTerent Plausible: - if the operator is unable to diagnose from memory the fact that rod CD-4 is out of alignment > 12 steps from any other ro Incorrect: - rods are > 12 steps in position difference Plausible: - if candidate does not remember that rods must be placed in manual Incorrect: - rods are > 12 steps in position difference l Plausible: - if candidate does not remember that turbine power is i
adjusted to maintain Tave and Tref
- Correct answer l
j For Official Use Only Ques _366 Page 1
Question #2 McGuire Nuclear Station SRO Exam Bank Question: 367 Answer: B
_
1 Pt(s) Which NC pump, in ES-0.2, is the " priority pump" for restarting under natural circulation conditions and what is the EOP basis for this priority? NC Pump 2A - this pump is connected to the pressurizer spray line and restarting will provide the best pressure control NC Pump 2B - this pump is connected to the pressurizer spray line and restarting will provide the best pressure control i NC pump 2A - this pump is connected to the pressurizer surge line and restarting will prevent thermal shock to the PZR spray nozzle NC pump 2B - this pump is connected to the pressurizer surge line and restarting will prevent thermal shock to the PZR spray nozzle Distracter Analysis: Incorrect: - 2A does not provide the best spray control Plausible: - because 2A is the next highest prio . y pump to be used
.
in ES-0 Correct answer Incorrect: - The surge line will not provide any pressure control Plausible: - if the candidate confuses preventing thermal shock to the spray nozzle is the EOP basis (for a different step) Incorrect: - 2D is not connected to either the surge line, nor the spray line Plausible: - if the candidate does not remember physical connections and thinks that the concem is thermal shock to the spray nozzle (as in some other EOPs).
i l
l For Official Use Only Ques _367 Page 2 i
Question #3 McGuire Nuclear Statica SRO Exam i
Bank Question: 368 Answer: C i
i Pt(s) Unit I was responding to an ATWS event. Given the following conditions: )
)
. NV pump 1 A is out ofservice j e 1NV-244A and 1NV-245B (Charging Line Cont Outside Isol) are open
]
- 1NV-221 A (NV Pumps Suct From FWST) was tagged shut j e INI-9A and INI-10B (NC Cold Leg Inj from NV) are shut
. 1NV-265B (Boric Acid to NV Pumps)is shut Which one of the following choices (if any) would provide adequate j emergency boration flow in step 5 of FRS.1, Response to Nuclear Power Generation / ATWS? Start NV pump 1B Open 1NI-9A and 1NI-10B Close INV-244A and 1NV-245B Start NV pump 1B Open 1NV-265B Start NV pump 1B Open 1NV-265B Start both boric acid transfer pumps Not able to provide adequate bon ation flow without INV-221 A being open !
Distracter Analysis: Incorrect: - the flow path is not aligned whh INV-265B shu Plausible: - if the candidate does not recognize that 1NV-265B is part of the flow pat Incorrect: - boric acid transfer pumps required to provide sufficient boric acid concentration with NV pump suction aligned to the VCT (normallineup)
Hausible: - if the candidate does not recognize that the BA pumps are needed or thinks that normal line up is the FWS Correct answer Incorrect: - suction is normally aligned to the VCT - do not need FWST for normalline u Plausible: - if the candidate thinks that the FWST is the normal suction path lineup for the NV pump.
For Official Use Only Ques _368 Page 3
_
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Question #4 McGuire Nuclear Station SRO Exam i l
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i Bank Question: 369 Answer: A 1 1 Pt(s) What is the normal source of makeup water to the KC system if surge tank level is low? What is the alternate flow path for draining from the KC system if surge tank level is high with the KC drain tank inoperable? makeup from YM drain to LHST sink makeup from RN drain to NCDT l makeup from YC drain to WZ sump makeup from NM l drain to WEFT Distracter Analysis: Correct answer: YM is the demineralized water tank, LHST is qualified to accept KC water with corrosion inhibitor Incorrect: - YM and the LHST sink are the source / drain for the KC syste Plausible: - if the candidate believes that the KC system woudl use RN water as an alternate source - always available - frequently used as an altemate source for many other system Incorrect: - YC is mcke up water for the chilled water system Plausible: - if the candidate believes that non-nuclear grade makeup water is required for the KC system, this is a potential source. The KC system can be lined up to feed and bleed off the WZ sump in OP6400/005 enci 4.11 so it can drain to this sum Incorrect: - NM is the nuclear sample system Plausible: - the WEFT is where the NM system drains to for non-reactor grade water - the KC system provides cooling flow for the NM system but not makeup.
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l Question #5 McGuire Nuclear Station SRO Exam Bank Question: 370 Answer: B 1Pt(s) Unit 2 was operating at 100% power with Pressurizer pressure channels in a normal lineup and NC pressure in the normal operating band when an instrument failure occurred. Given the following plant conditions and events:
- PORVs NC-32 and NC-36 open causing NC pressure to decrease
- PORVs close at 2185 psig e Backup heaters energize and pressure increases
- PORVs reopen at 2189 psig
- A reactor trip does not occur e This cycle continues until operator action is taken Assuming no operator action to realign (or swap) instrument channel j lineups, which one of the following events would cause these actions? PZR pressure channelI fails high PZR pressare channelII fails high PZR pr essure channelIII falls high PZR pressure channelIV falls high Distracter Analysis: Incorrect: - channel I does not control PORVs 32 and 36 Plausible: - plant response is vwj similar except PORV 34 opens and pressure continues to decrease below 2185 until a plant trip occurs due to spray valves being full open Correct answer - PZR channel III is the interlock channel to channelII Incorrect: - PORVs do not open - interlock channel only Plausible: - if candidates do not recall that this is an interlock channel or think that it controls the interlocked PORVs Incorrect: - PORVs do not open - interlock channel only Plausible: - if candidates do not recall that this is an interlock channel or think that it controls the interlocked PORVs For Official Use Only Ques _370 Page 5
I Question #6 McGuire Nuclear Station SRO Exam
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l Bank Question: 371 Answer: C
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1 Pt(s) In the event of a steamline mpture that cannot be isolated, FR-P.2, Response to Anticipated Pressurized Thermal Shock Condition, could be implemented )
due to an excessive cooldown. What is the limiting component for this PTS J '
event and what is the best indication of the temperature of this component?
Limitine Component Best I_ndication steam generator tube sheet I pressurizer spray nozzle Tw- reactor vessel wall reactor vessel downcomer Tw Distracter Analysis: Most candidates will have grasped the bases for FR-P.1 but FR-P.2 is less well known. However, the basic principles regarding NDT and PTS are the same for this questio Incorrect: - the steam generator tube sheet is not the limiting component for a steam leak event Plausible: - if the candidate corresponds the steam leak to the steam generator and thinks that this is limiting due to the large pressure difference across the steam generator from the mpture (PTS) Incorrect: - the pressurizer spray nozzle is not limiting for this event as there is no spray flow Plausible: - if the candidate confuses this event with other events where the pressurizer delta Temp is limiting - or does not recognize that NC pumps have been stopped Correct answer Incorrect: - the reactor vessel downcomer is not generally limiting Plausible: - if the candidate does not recognize that Thot is not the most limiting temperature. The Rx vessel wall and the Rx vessel downcomer are essentially synonymous for the same regio For Official Use Only Ques _371 Page 6
Question #7 McGuire Nuclear Station SRO Exam Bank Question: 372 Answer: C l
1 Pt(s) Unit 2 was respond.ug to a faulted steam generator c.ent. The operators 1 entered FR-P.1, Response to Imminent PTS, and reach step 15 where they are directed to isolate cold leg accumulator What is the EOP basis for isolating the CLAs in FR-P.l? To prevent injecting the CLA nitrogen bubble into the reactor and creating a gas bubble in the vessel head regio To prevent repressurizing the reactor vessel and adding pr ssure stress to thermal stres To prevent adding more cold water to the reactor vessel and increasing the thermal stres To prevent depleting CLA volume and to preserve a source of highly borated water to prevent recriticallity during cooldow Distracter Analysis: Incorrect: - the CLAs are isolated to prevent adding to the thermal stress. The gas bubble would not be a limiting concem in FR- Plausible: - this is a valid limiting condition for isolating the CLAs during LOCA depressurizations - good answer, wrong even Incorrect: - adding the CLA volume would not cause an increase in pressure because the addition of the CLA volume is caused by pressure in the system decreasing below CLA pressur Plausible: - the basis for terminating SI is to prevent adding water to the system and increasing pressure thereby adding pressure stress to thermal stress. Not applicable to the CLA Correct answer Incorrect: - the CLAs are not require for Boron addition for this scenari Plausible: - recriticallity this is a valid concern for scenarios that involve adding large quantities of unborated water. Good answer -
wrong event.
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Question #8 McGuire Nuclear Station SRO Exam I
Bank Question: 373 Answer: D
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! 1 Pt(s) Unit 2 was at 10% during a plant startup when a loss ofcondenser vacuum caused a reactor trip to occur. Given the following conditions:
. The steam dump select switch was in steam pressure mode
- Cond nser vacuum dropped to 18 inches
. No component or instrument failures occurred
- No operator action taken Which one of the following sequences best describes the actuation of the steam dumps to this event assuming?
Condenser dump valves Atmospheric dumD Valves open open open shut shut open shut shut Distracter Analysis: Incorrect: - the condenser dumps will not open due to the C9 arming signal not present caused by a low vacuum condition (20 in)
The atmospheric dumps will not open because they do not actuate when the steam dump selector switch is in steam pressure mode Plausible: - this would occur if the transition had not yet been made to steam pressure mode and C9 had actuated - a normal plant response during a startup < 10% powe Incorrect: - the condenser dumps will not open due to the C9 arming signal not present caused by a low vacuum condition Plausible: - this sequence would occur if there was not a loss of condenser acuum or if the candidate does not recognize the interaction veen condenser pressure and the C9 arming signa Incorrect: - ti. atmospheric dumps will not open because they do not actuate when . te steam dump selector switch is in steam pressure l mWe Plausible: - this sequence would occur if the plant was in Tave mode - the transition from Tave to steam pressure mode occurs by procedure at 10% power ] Correct answer For Official Use Only Ques _373 Page 8 l
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c Question #9 McGuire Nuclear Station SRO Exam I
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Bank Question: 374 Answer: B l
l l 1 Pt(s) Unit 1 is responding to a unit blackout. Step 9.g in ECA0.0 that states: '
"9.g IF A T ANY TIME Unit 2 main condenser vacuum goes down significantly. THEN
_1) Ensure IRC-7 (RC Crossover Supplyfrom Unit 2) is open j
_2) WHEN 1RC-7 is open, THEN close IRC-5 (RC Crossover Supplyfrom unit 1)
Which one of the following statements correctly describes the required operator action for step 9.g_l) that begins with the word " Ensure"?
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1 Determine present status of IRC-7 (no action required other than to check the valve position light in the conirol room) Take any necessary action to guarantee that IRC-7 is open but there is no requirement to dispatch personnel to locally verify the valve position if position can be confirmed in the control room Take any necessary action to guarantee that IRC-7 is open including dispatching personnel to locally verify the valve position even ifindication is available in the control roo Carefully appraise the situation - including taking all actions based on evaluation Distracter Analysis: Incorrect: - this is the defmition for the term " check" Plausible: - requires knowledge of terms - candidate may not know this defmition Correct answer Incorrect: - local actions are not required unless the step specifies to dispatch personnel Plausible: - the candidate may not understand the difference between B and C and C is more conservative Incorrect: - this is the definition for the term " evaluate" Plausible: - the candidate may not know the difference between these terms which are very close For Official Use Only Ques _374 Page 9
r Question #10 McGuire Nuclear Station SRO Exam l
Bank Question: 375 Answer: D 1 Pt(s) What is the concem for normal RN system high flow through the KC heat exchangers and what is the basis for this limit? Controlied by mini-flow protection valves for the train related l RN pump to remain less than 2000 gpm which will protect system piping against flow oscillation overcooling in the KC syste Controlled by mini-flow protection valves for the train related RN pump to remain less than 2700 gpm which will protect system piping against flow induced vibratio Controlled by operator action to remain less than 2700 gpm which will protect system piping against flow oscillation and overcooling in the KC syste Controlled by operator action to remain less than 2000 gpm which will protect system piping against flow induced vibrntio Distracter Analysis: Incorrect: - the basis is to prevent flow induced vibmtion. The mini-flow valves maintain flow below 2700 gp Plausible: - the basis is correct for a different limitation - why a super flush must not be performed if KC HX delta pressure is less than 9 psi Incorrect: - the correct flow limit is 2000 gpm Plausible: - each statement is correct on its own - but 2700 gpm is not the correct limit to prevent flow induced vibratio Incorrect: - the limitation is 2000 gpm and the basis is to prevent flow induced vibratio Plausible: - 2700 gpm is the mini-flow valve control setpoint. The basis is correct for a different limitation - why a super flush must not be performed if KC HX delta pressure is less than 9 psi Correct answer For Official Use Only Ques _375 Page 10
I-Question #11 McGuire Nudear Station SRO Exam l
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Bank Question: 376 Answer: A 1 Pt(s) Unit 1 is operating at 100% power when a fire in a roc n in the Auxiliary Building causes the room heat detector to actuate and the nearby sprinkler fused link to open. Which one of the following statements describes the correct sequence of events that would occur to provide a flow of water to the sprinkler head? Water would immediately flow to the sprinkler valve under the normal configuration of the fire header without any further actio Deluge valves 1RY-113 and 1RY-114 would open automatically when the thermostatic detector was actuated and provide water to the fire heade Deluge valves 1RY-113 and IRY-114 would first need to be opened by operator action in the control room to provide water to the sprinkler syste Deluge valves 1RY-113 and 1RY-114 would first need to be opened by operator action at the local breakglass stations in the Auxiliary Building to provide water to the sprinkler syste Distracter Analysis: Correct answer - the header is maintained in a filled and pressurized status through bypass lines around RY-113 and RY-114 Incorrect: - RY-113 and RY-114 do not open automatically Plausible: - if the candidate does not know about the bypass lines -
automatic openine should be a conservative feature ;
' Incorrect: - RY-113 and RY-144 cannot be opened from the control room Plausible: - the control room must manually open the RY isolation valves to charge the fire main in the Reactor Building Incorrect: - the header is maintained in a filled and pressurized status through bypass valves around RYRY-113 and RY-114 Plausible: - RY-113,114 are opened from the breakglass station in the Auxiliary Building but this action does not provide water the
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sprinkler heads -just additional hre mam capacity for increased flow l l
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i ( Question #12 McGuire Nuclear Station SRO Exam l
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Bank Question: 377 Answer: D 1 Pt(s) What is the purpose of the "S" position on the remote / local switch on the l auxiliary shutdown panel for NV-1 A?
l i Allows NV-1 A to move to the " safe" position in the event of a mismatch between ASP local switch position and the actual valve position Places NV-1 A in the "cecure" position by removing power from the valve Places NV-1 A in the " standby" position which allows positioning from either the control room or the AS Allows operator to check the " status" of NV-1 A position Distracter Analysis: Incorrect: - there is no " safe" position Plausible: "S" could stand for " safe" Incorrect: - there is no " secure" position Plausible: "S" could stand for " secure" Incorrect: - there is no " standby" position Plausible: "S" could stand for " standby" Correct answer i
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' Question #13 McGuire Nuclear Station SRO Exam Bank Question: 379 Answer: B
1 Pt(s) Unit 2 is responding to a reactor trip with a total loss of CA flow. The operators entered FR-C.1, Response to Inadequate Core Cooling, due to a red path requirement. Step 17 of this procedure states:
I Depressurize intact steam generators to atmosphericpressure l
i What is the purpose of this action? To reduce steam generator pressure to allow the line up of a low pressure feedwater system for heat sink restoration To re-cover the core via ND flow by reducing NC pressure below ND pump shutoff head To minimize the PTS consequences of the event by reducing the pressure stress on the reactor vessel wall To reduce steam generator pressure to allow restoration of 450 gpm CA flow
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Distracter Analysis: Incorrect: - this depressurization occurs even if a heat sink exist Plausible: - this is the correct basis a similar step in for FR- Correct answer Incorrect: - This would increase the thermal stresses not decrease them. PTS is not the concern for this cas Plausible: - reducing SG pressure will reduce NC pressure and thereby reduce pressure stress on the Rx vessel wall Incorrect: - CA pump shutoff head is capable ofinjecting into a hot SG at design pressure - within the SG PORV relieflimits Plausible: - 450 gpm is the total CA flowrate required to provide a heat sink to the reactor - basis for step 11 of FR- For Official Use Only Ques _379 Page 13
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l Question #14 McGuire Nur. lev Station SRO Exam l
l Bank Question: 380 Answer: C l 1 Pt(s) Which one of the following statements is a conect description of the capabilities of EMF-48, the Reactor Coolant Monitor at 100% power?
l Detects beta flux from the NC system coolant. This prevents the l detector falsely responding to N16 gamma radiation which would mask a failed fuel even Detects the N16 gamma flux from the NC system coolant which is proportional to the amount of failed fuel claddin l Detects total gamma flux from NC system coolant after a 1 l minute sample delay time to allow N16 gamma radiation to ;
decay awa Detects total gamma flux from NC system coolant. The gan ma source term from a clad failure would be much greater than the N16 gamma flux at powe Distracter Analysis: Incorrect: - EMF-48 does not detect beta radiation Plausible: - the detector could function ifdesigned this way because N16 gamma would mask the failed fuel problem Incorrect: - N16 gamma is proportional to reactor power level and docs not correlate to failed claddin Plausible: - if the candidate was confused over the correlation between N16 gamma levels and power level Correct answer Incorrect: - The gamma flux from N16 is >> failed fuel at power Plausible: - if the candidate did not know that N16 gamma was >>
than failed fuel source term levels.
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Question #15 McGuire Nuclear Station SRO Exam Bank Question: 381 Answer: C l
1 Pt(s) Unit 1 is operating at 100% power when the output br::aker for 1EVIA trips open causing IEKVA to deenergize. Given the following events and conditions:
- All controls are in channel 1 or 1/2 positions
- The operators enter AP/1/A/5500/15 (Loss of Vital or Aux. Control Power).
- The FIRST immediate action step in AP/15 is omitted by the operator
. All subsequent steps are performed correctly What consequence would result from a failure to perform the FIRST immediate action of AP/15? Pressurizer pressure would increase until PORVs 32 and 36 cycle at 2335 psig S/G levels would increase to the Hi-Hi S/G level turbine trip Rods would step in at 72 steps per minute continuously Pressurizer level would increase to the reactor trip setpoint Distracter Analysis: Incorrect: - PORVs do not actuate because operators shift to operable Pzr pressure channel in step 6 of the procedure Plausible: - if candidates do not know that Pzr pressure is shifted to a ditTerent controlling channel Incorrect: - S/G channels are shifted to operable channels as an immediate action Plausible: - if candidate does not know that the second immediate action step has the operator shift any effected channels , Correct answer: - the missed immediate action is to place rods in '
manual - this prevents the turbine impulse pressure channel failure l from causing Tref to fail to 557 and generating a large +TMM and i
+PMM - rods drive in at 72 steps / min ifin auto Incorrect: - the failed pressurizer level channel is shifted to an !
operable channel in step 7 of AP/l 5 - pressurizer level will not j increase afler the channel is shifled j Plausible: - if the operator does not recognize that the pressurizer level channel is shifted to an operable channel in Ap/15
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For Official Use Only Ques _381 Page 15 I l
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l Question #16 McGuire Nuclear Station SRO Exam i l
Bank Question: 382 Answer: B 1 Pt(s) Unit 2 was conducting a plant start up when a continuaus rod withdrawal accident occurred in shutdown bank A. Given the following conditions and indications:
- The Power Mismatchfremperature Mismatch Meter indicates e PMM = +4 F e TMM = -2 F e Reactor poweris 50%
- Rods are in automatic Which one of the following describes control bank rod response? rods step in at 72 steps per minute rods step in at 8 steps per minute rods step out at 40 steps per minute rods step out at 8 steps per minute Distracter Analysis: Incorrect: - PMM sums with TMM and the combined error signal is +2 *F which moves rods in at 8 steps per minute Plausible: - if the candidate subtracts PMM and TMM the combined error is +6 F which causes rods to move in at 72 SPM ! Correct: - PMM + TMM = +2 F, rod move at 8 SPM ! !
Incorrect: - rods step at 8 steps per minute Plausible: - if the candidate thinks that the combined error signal is I on the proportion part of the rod speed control curve ! Incorrect: - control rods will respond by stepping in not out l Plausible: - if the candidate thinks that the combined error cause i rods to move out (transposed direction) in his/her mind)
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F Question #17 McGuire Nuclear Sta' don SRO Exam Bank Question: 383 Answer: D 1 Pt(s) Which one of the following choices best describes the Unit 2 response caused by a dropped control rod from 100% power with rod control in automatic? Initially the control rods step in due to a negative power mismatch signal until the temperature mismatch signal builds up and overcomes power mismatch causing rods to step out to eventually stabilize at a T,,below 585 ' The control rods step in as the positive temperature mismatch signal is caused by an immediate drop in T,, due to the reactor power decrease until T,, stabilizes at ~585 * The control rods step out until the positive power mismatch signalis overcome by the negative temperature mismatch signal and the reactor stabilizes at a temperature above 585 ' The control rods step out until the positive power mismatch signal decays away (rate sensitive) and the temperature mismatch signalis nulled out by reactor stabilizing at ~585 Distracter Analysis: Incorrect: - power mismatch is positive as turbine power > NI power causing rods to step out Plausible: - if the candidate is confused over the derivation of the power mismatch signa Incorrect: - rods do not step in because Trefis derived from turbine impulse pressure not NI power so it is only effected by the slight reduction in turbine power caused by the cooling due to the dropped rod. This is not a large effect nor is it an immediate effect when compared to the positive power mismatch signal Plausible: - if candidate thinks that NI power provides the input for derivation of the Tref signal Incorrect: - the reactor will not stabilize above 585 *F because Tref is clipped at 585 F Plausible: - if the candidate does not recognize that Tref cannot exceed 585 F Correct answer
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For Official Use Only Ques _383 Page 17
I Question #18 McGuire Nuclear Station SRO Exam i
Bank Question: 384 Answer: D 1 1Pt(s) Unit 2 is operating at 48% power during a plant start up when an electrical power transient causes the 2A NCP to trip on under voltage and the main turbine to trip. Given the following indications prior to the trip:
Permissive Sienal Status Light
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P6 (SR Block Permissive) lit
- P7 (Lo Power Rx Trips Blocked) not lit
. P8 (Hi PWR Lo Flo Rx Trip Blocked) lit
P10(Nuclear at Power) lit
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P11 (Pressurizer S/I Block Permissive) not lit
> 912 ( Lo-Lo Tavg) not lit Which one of the following stuement describes the reactor trio system immediate response? the reactor will trip on loss of flow the reactor will trip on turbine trip the reactor will trip on NCP under voltage the reactor will not trip Distracter Analysis: Examiner Note: There is no status light for P- Power is at the P-8 setpoint (48%) but the P-8 pemlissive have not yet picked up (deadband) and the candidates must determine the status of P-8 by understanding what the P-8 status light is telling the Incorrect: - loss of flow in one loop is blocked below P8 Plausible: - if the candidate does not recognize that P8 is not yet in Incorrect: - the turbine trip / reactor trip is blocked below P8 Plausible: - if the candidate does not recognize that P8 is not yet in Incorrect: - although this trip is active with P7 in the coincidence
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for the trip is 2/4 Plausible: - if the candidate does not recognize that the coincidence for the NCP monitor panel is 2/4 vice 1/4 Correct answer: - P-8 not yet "in" For Official Use Only Ques _384 Page 18
n Question #19 McGuire Nuclear Station SRO Enam Bank Question: 385 Answer: A ,
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1 Pt(s) Which one of the following pressurizer conditions would result in the most rapid rate of pressure decrease if a code safety valve had stuck open i (assuming the NC system was in a subcooled condition)? )
Fluid Released PZR pressure water 1000 psig water 500 psig steam 1000 psig steam 500 psig I
DistracterAnalysis: Correct answer - solid water condition, highest pressure Incorrect: - lower pressure than A, less leak rate Plausible: - if candidate does not recognize that the leak rate is lower for lower pressure Incorrect: - steam release causes lower pressure drop than water Plausible: - if candidate does not know that a steam release has a smaller mass flow rate that a water release Incorrect: - steam release rate < water release rate and pressure low Plausible: - if candidate does not know correlation between change in leak rate with change in pressure l
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For Official Use Only Ques _385 Page 19
p Question #20 McGuire Nuclear Station SRO Exam Bank Question: 387 Answer: D 1 Pt(s) What is the basis for the foldout page requirement in E-1 to close NV-150B and NV-151 A (NV Pumps Recirc) when NC pressure is less than 1500 psig? To prevent NV pump runout at low pressures To prevent loss of NPSH to the NV pumps To prevent NV pump from being damaged by operating against shutoff head pressure To prevent a reduction of full SI flow to the core Distracter Analysis: j i Interrect: - pump runout is not a concern - NV pumps are designed l to ran under these conditions !
Plausible: - pump runout can be a concem for situations where the pump discharge pressure is very low Incorrect: - NPSH is not a concem for NV pumps as the FWST supplies adequate suction hea !
Plausible: - NPSH can be a consideration for some EOP scenarios ! Incorrect: - the water does not recirc to the VCT but goes back to !
the NV pump suctions l Plausible: - this is the reason for opening the miniflow valves on the foldout page, not closing the Correct answer - maximizes flow into the core
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Question #21 McGuire Nuclear Station SRO Exam I
Bank Question: 388 Answer: A 1 Pt(s) Unit 1 is responding to a LOCA in the auxiliary building. The operators have implemented ECA-1.2, LOCA Outside ofContainment. Step 2 guides the operators to attempt to find and isolate the leak. Step 2C requires the following sequence:
. Check open 1NI-l15A and 1NI-144B (NI pump mini flow valves)
e Close INI-162A (NI Pumps Cold Leg Isol)
What is the purpose of opening the mini flow valves? Protect the NI pumps from operating against shutoff head Isolation of a potential LOCA path to the FWST Protect the NI pumps from runout conditions upon restart Provide a diversion path to prevent high pressure water from over-pressurizing the ND system during leak isolation procedures Distracter Analysis: Correct answer j Incorrect: - the mini flow valves will not isolate any potential path Plausible: - in ES 1.3 (and some other EOPs), the NI mini flow valves are closed to prevent pumps from recirculating radioactive ,
water back to the FWST - right reason, wrong procedure l Incorrect: - Opening mini flow valves will not protect against I runout ;
Plausible: - if the candidate is confused over the difference between
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runout conditions and shutoff head conditions Incorrect: - this will not create a diversion path - it will align the NI recirc line to the ND system and eventually to FWST (which is a low ;
pressure system) l Plausible: - the ND system has a design pressure of 600 psig and if NC system pressure was applied,it would mpture. Setting up a ;
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l diversion path would be a reasonable thing to do. This alignment does not accomplish this goal.
l For Official Use Only Ques _388 Page 21 i
I Question #22 McGuire Nuclear Station SRO Exam Bank Question: 389 Answer: B
1 Pt(s) Unit 2 is responding to a LOCA and the operators have reached step 26 of {
1 ES-1.2, Post LOCA Cooldown and Depressurization, which requires \
l isolating cold leg accumulators. What is the purpose of this step? I
l To prevent CLA injection when the NC system depressurizes l with the resulting thermal stress and PTS concerns from the cold water on the reactor vessel To prevent injecting the CLA nitrogen bubble into the reactor and creating a gas bubble in the vessel head regio To reserve the CLA water in f t . event of a recriticallity during the cooldown due to a loss of shutdown margin To prevent the excessive addition of highly Borated CLA water into containment with the potential for Boron crystallization after cooling down, possibly fouling the flow paths in the cor Distracter Analysis: Incorrect: - PTS is not a concern because cold SI flow is already injectinginto the reactor vessel Plausible: - good reason - wrong EOP. This is the correct basis for isolating CLAs during FR- Correct answer Incorrect: - the SI system contains sufficient Bolon to prevent recriticallity events without the CLA water Plausible: - recriticallity is a concern anytime the system is cooled
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down. This procedure cools down the core Incorrect: - although Boron recrystalization is a concem, the EOPs provide a method of preventing this problem (Hot Leg Recirculation) and this is not concem Plausible: - Boron recrystalization occurs at low temperatures and high Boric acid concentrations. The BA in the CLAs is relatively high - but not enough to cause this problem f
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Question #23 McGuire Nuclear Station SRO Exam Bank Question: 390 Answer: B l
l 1 Pt(s) Unit 1 is recovering from a LOCA. The operators sta*ed the process of !
terminating safety mjection at 2:00 AM. Given the following indications at I the following times:
Parameter 2:00 2:05 2:10 2:15 1) Pressurizer level (%) 40 29 15 11 2) NC pressure (psig) 280 285 290 295 l 3) ND Flaw 1000 1025 1075 1085 4) Core exit T/Cs ( F) 690 702 695 685 ,
5) FWST level (inches) 183 179 149 113 6) Containment Pressure (psig) .3 .1 What is the earliest time that the operators should transition to ES-1.3, transfer to cold leg recirculatio Distracter Analysis: This question is designed to test the candidate's ability to identify the criteria for switchover to cold leg recire from a list of plant parameters. A recent change to the switchover criteria was changing the FWST level from 150 inches to 180 inches (since last exam). In addition, he meets the S/I reinitiation criteria at 0205 and this will check if he recognizes that ES-1.3 has priority Incorrect: - does not meet criteria for switchover, FWST > 180 inches Plausible: .ifcandidate does not know foldout criteria for switchover Correct answer - FWST level < 180 inches - reaches ACC value for S/I reinitiation criteria - needs to decide if ES-1.3 needs to be delayed until S/I reinitiation is completed Incorrect: - too late Plausible: - if candidate does not know foldout criteria for switchover - recent change (since last exam) to the switchover criteria - used to be 150 inches - now is 180 inches For Official Use Only Ques _390 Page 23
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Question #23 McGuire Nuclear Station SRO Exam Incorrect: - too late Plausible: - ifcandidate thinks that SI termination must be competed before switchover to cold leg recire or if he thirds that S/I reinitiation takes priority over switchover - reaches non n'C value for S/Ireinitiatio >
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1 Question #24 McGuire Nuclear Station SRO Exam
l Bank Question: 391 Answer: D j
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1 Pt(s) Which one of the following ND pump operating conditions would be most likely to experience cavitation?
l ND pumps running with both trains crosstied and NC pressure is 290 psig ND pump running in mid loop operations and reactor vessel levelis low ND pump running in a shutdown cooling lineup and the discharge header ruptures ND pump running and the section valve closes part way (pump does not trip) !
Distracter Analysis: Incorrect: - this is an example of pump deadheading Plausible: -if the candidate does not understand centrifugal pump operating conditions Incorrect: - this is an example of conditions for pump vortexin Vortexing and cavitation are have similar indications but they are different operating situations. Vortexing is the entrainment of air in the suction path due to a low level condition. Cavitation is the loss of NPSH due to a low pressure conditio Plausible: - if the candidate does not understand centrifugal pump operating conditions and this is the condition that the candidates are primarily trained to avoid.- they will be most familiar with this ,
condition Incorrect: - this would be an example of conditions for pump runout Plausible: - if the candidate does not understand centrifugal pump operating conditions Correct answer For Official Use Only Ques _391 Page 25
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Question #25 McGuire Nuclear Station SRO Exam Bank Quesdon: 393 Answer: C i
1 Pt(s) Unit I was conducting a reactor startup when source range channel N-31 l failed. Given the following conditions and events:
- No reactor trip has occurred prior to this point
. AP/1/A/5500/16 has been completed )
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. N-31 repairs have been made
- N-31 is being returned to service e N-32 = 10' CPS e immediately upon taking the " level trip" switch to the " normal" position a reactor trip occurred ;
i Which of the following operator errors would explain this event?. ; " Operation selector" switch was left in "leveladj." position with level potentiometer set at a level of 10* CPS after retesting "High-flux at shutdown " switch left in the " blocked" position Instrument power fuses were never reinstalled control power fuses were never reinstalled Distracter Analysis: Incorrect: - the operation selector switch is taken out of the circuit when the level trip switch is taken to normal Plausible: - if the candidate thinks that a test signal can be inserted with level trip switch in the normal position Incorrect: - no effect - only blocks out high flux trip at 105 CPS -
trip is separate from level trip function Plausible: - if the candidate does not understand that the level trip switch in bypass does not effect this trip - some protective system trips will be active if one signal is left in the blocked condition when the circuit is returned to service Correct answer Incorrect: - if control power fuses has been removed, the reactor I would have tripped prior to this point Plausible: - if the candidate does not recognize that a trip would have already occurred l .
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. Question #26 McGuire Nuclear Station SRO Exam Bank Question: 394 Answer: A
- 1 Pt(s) The condenser air ejector monitor, EMF-33, is designed to detect
! gaseous activity in the air ejector discharge into the plant vent particulate activity dissolved in the hotwell water drawn into the air ejector tritium activity in the air ejector discharge into the plant vent iodine activity dissolved in the hotwell water drawn into the air ejector Distracter Analysis:
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1 Correct answer - EMF-33 uses a single range beta gas plastic scintillation detector Incorrect: - does not detect particulate activity in water Plausible: - if the candidate does not recall that the purpose of the monitor is to detect for the presence of high gaseous activity in the non-condensable gases discharged from the air ejector Incorrect: - does not detect tritium because tritium has low energy beta radiation that requires a different monito Plausible: - Tritium is a gas contained in reactor coolant Incorrect: - does not detect Iodine activity in the water Plausible: - low levels ofiodine activity would be released into the condensate system during a SGTR i
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1 Question #27 McGuire Nuclear Station SRO Exam
l Bank Question: 395 Answer: B !
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1 Pt(s)- What is the reason for EMF-34 automatically closing the S/G sample header radiation monitor inlet isolation valve (INM-267) when it reaches a trip 2 condition during a steam generator tube rupture? To prevent contaminated condensate from reaching the auxiliary building sump To prevent contaminated water from reaching the turbine l building .. a l i
' To reduce gaseous radiation levels in the CT Lab room To prevent the cation sample columns from becoming saturated with boric acid Distracter Analysis:
l Incorrect: - the water cannot drain into the auxiliary building i Plausible: - if the candidate does not understand the drainage path, many drains do go into the auxiliary building sump Correct answer Incorrect: - this is not the primary concem at the condensate water does not get exposed to the air in the room unless a manual sample is i drawn l Plausible: - gaseous activity is the concem at the air ejector monitor Incorrect: - not a significant concem for a tube rupture i Plausible: - boric acid would enter the cation columns if they were !
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For Official Use Only Ques _395 Page 28
Question #28 McGuire Nuc! ear Station SRO Exam I
Bank Question: 396 Answer: D 1 Pt(s) ' Unit 1 is responding to a faulted S/G inside containment in E-1 when a red path occurs on FR-H.1 due to a loss of CA pumps. The operators are attempting to recover the heat sink using the condensate system when the following system parameters occur:
0200 0205 0210 0215 l i
NC System Pressure (psig) 1986 1993 1981 1935 i Containment pressure (psig) .9 .9 Steam nenerator levels (inches _)
l A S/G (WR) 16 14 18 15 l IB S/G (WR) 11 10 16 16 1C S/G (WR) 10 11 14 16 ID S/G (WR) 14 13 18 20 Total feedwater flow rate 0 50 100 200 l Core exit thermocouple temp (*F) 936 1039 1012 981 l
What is the earliest time (if any) that the operators should establish feed and bleed in accordance with the foldout criteria in FR-H.l?
REFERENCES PRO VIDED - FR-S.] Foldout Page (36)
l never need to initiate feed and bleed l l l l I l
Distracter Analysis:
< Incorrect: - criteria are met at 0215 Plausible: -if candidate does not analyze parameters correctly or l mistakenly thinks that once containment pressure drops below I psig, ACC setpoints no longer apply. AT 0205, condensate water is
.ausing S/G levels to increase and appears to be effectively recovering the heat sin Incorrect:- 0205 - only 2 S/Gs < 12 inches, ACC not met For Official Use Only Ques _396 Page 29 I
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Question #28 McGuire Nuclear Station SRO Exam Plausible: - Highest CET readings - worst conditions for NC l system l Incorrect: - 0210 - ACC applies but only 2/4 S/G WR < 17 inches due to feedwater entering the S/G and raising levels l
Plausible: - ifcandidate doesn't realize ACC criteria is 3/4 S/G < 17 inches, feedwater is entering the S/Gs and level is rising Correct answer - 3/4 S/G WR < 17 - ACC still applies although containment pressure has dropped back below 3 psig For Official Use Only Ques _396 Page 30 l
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Question #29 McGuire Nuclear Station SRO Exam Bank Question: 397 Answer: B 1Pt(s) Which one of the following system lineups best describes the pmper operation of the DC auxiliary power system during an equalizing charge of battery IDP? The normal battery charger charges 1DP separately from the distribution center The standby battery charger supplies distribution center leads Unit DC bus tie breakers are open The normal battery charger supplies distribution center loads The standby battery charger charges IDP separately from the distribution center Unit DC bus tie breakers are closed The normal battery charger charges IDP separately from the distribution center The standby battery charger in standby Unit bus tie breakers are closed The normal battery charger is in standby The standby battery charger charges 1DP tied to the distribution center Unit bus tie breakers are open Distracter Analysis: Incorrect: - bus tie breakers must be shut, cannot allow a battery charger to supply a bus without being connected to a battery Plausible: - the bus tie breakers are normally required to be open for electrical separation between the 2 unit batteries Correct answer Incorrect: - allows 2DP to supply IDP loads without any charger Plausible: - this could physically function but 2DP would be under-sized to handle both units with only one charger Incorrect: - the battery voltage would be too high to allow 1DP to be supply bus loads Plausible: - this would physically work l
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Question #30 McGuire Nuclear Station SRO Exam Bank Question: 398 Answer: C l
1 Pt(s)- Unit 2 is in the process of starting up the reactor. Given the following conditions and events:
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e 0200 - 2 EMF-3 (Refueling bridge ARM) reaches a trip 2 conditio . 0210 - N-32 is deenergized due to a SR detector failure e The operators complete all required procedural steps during the startup Wlat is the earliest time (if any) that the contaisunent evacuation alann will no longer actuate in Unit 2 during the startup?
0200 0205 0210 0215 N-31 (CPS) 1.2x10' 1.7xlu' 2x10' O N-32 (CPS) 9.2x10' l.1x104 0 0 N-35 (wnps) 1.1x104 1.4x10" 1.6104 1.910
4 4 N-36 (amps) 1.3x 10" 1.5x10 1.8x10 2.1x104* the containment evacuation alarm will actuate at all times Distracter Analysis: Incorrect: - both SR high flux trips are still active Plausible: - if the candidate thinks that P6 permissive actuation cuts out the alarm Incorrect: - N-31 SR high flux trip still active Plausible: - if the candidate does not recognire that BOTH SR high flux trips must be blocked Correct answer Incorrect: - alarm is blocked at 0215 Plausible: - ifcandidate thinks that some other action or interlock blocks the alarm For Official Use Only Ques __398 Page 32
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Question #31 McGuire Nuclear Station SRO Exam Bank Question: 399 Answer: A 1 Pt(s) Unit I was operating at 100% with the pressurizer level controller in the 1-2 i position. Given the following initial response: I e Charging flow reduces to minimum e Backup heaters immediately energize l e Actuallevel begins to decrease Which one of the following failures has occurred to cause this plant response? l PZR level channel 1 detector reference leg has ruptured PZR level channel 1 detector variable leg has ruptured PZR level channel 2 detector reference leg has ruptured PZR level channel 2 detector variable leg has ruptured Distracter Analysis: Correct answer - a leak in the reference leg causes the pressurizer channel to sense a high level condition - which causes the system response as indicated. Actual pressurizer level decreases initially due to charging flow decreasing while letdown remains in servic Incorrect: - variable leg rupture causes channel 1 to sense a low PZR level - actual level will increase not decrease initially Plausible: - if the candidate thinks that this will cause a high level Incorrect: - will cause channel 2 to sense a high level - will not get heaters deenergizing - only get a high level alarm Plausible: - if the candidate confuses the plant response for channel 2 - thinks that channel 2 controls Incorrect: - pressurizer level would increase not decrease Plausible: - if the candidate confuses the direction of the pressurizer level failure or doesn't recognize plant response - this is very similar to a channel I high failure only the actual level increase instead of decreases
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Question #32 McGuire Nuclear Station SRO Exam
Bank Question: 400 Answer: B ,
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1 Pt(s) Unit 2 is responding to a small break LOCA. Given the following plant l conditions and events:
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e 2D S/G pressure = 1228 psig (all channels indicating the same) l
. 2D S/G NR ievel = 73% l
. no S/G PORVs or code safeties have lifted on any S/Gs e Operators have entered FR- e Containment pressure = 3.1 psig Which one of the following ctatements represents actual conditions for the S/G instrumentation under these accident conditions? Actual wide range level > 73% Actual wide range level < 73%
' Actual S/G pressure < 1228 psig Actual S/G pressure > 1228 psig Distracter Analysis: Incorrect: - narrow range SGWL detectors are calibrated for
- nonnal (hot) operating conditions. Wide range SGWL detectors are calibrated for cold (shutdown) conditions. This means that the wide range SGWL instrument will read lower than actual level under hot conditions because variable leg water density is less than the calibrated setting l Plausible
- - if the candidate fails to understand the effect of cold l calibration on the SGWL or gets the effect reversed.
l Correct answer Incorrect: - SG pressure transmitter accuracy is no: sensitive to temperature changes in the S/G - pressure transmitters located outside containment - no ACC valves for steam pressure l Plausible: - if the candidate is confused over the accuracy of the S/G pressure transmitters. May confuse pressure transmitters with level transmitters Incorrect: - SG pressure transmitters are not sensitive to temperature changes in the S/G Plausible: - if the candidate is confused over the accuracy of the S/G pressure transmitters. May confuse pressure transmitters with level transmitters For Official Use Only Ques _400 Page 34
Question #33 McGuire Nuclear Station SRO Exam Bank Question: 401 Answer: D 1 Pt(s) Unit 2 was operating at 100% when a single control rod in control bank D drops into the core due to a failed CDRM. The SRO directs that the dropped l rod be recovered.
! Which one of the following prevents the remaining rods in the control bank l
from being withdrawn while the dropped rod is being recovered? The Rod Stop Bypass is used to block control rod bank D outward movement C-11 actuated when the rod dropped and will prevent outward rod motion by control bank D The Lift Coil Disconnect Switch is opened on the dropped rod to electrically isolate it from control bank D The Lift Coil Disconnect Switches are opened on control bank D rods that did not drop Distracter Analysis:
l Incorrect: - the rod stop bypass only blocks the overpower channel signal to rod control which does not effect the situation under these conditions Plausible: - if the candidate does not recognize the function of this switch Incorrect: - C-11 will not allow any rod motion - could not pick up the dropped rod and it has not actuated under these conditions Plausible: - C-11 will prevent outward rod motion in bank D Incorrect: - will not be able to pick up the dropped rod Plausible -if the candidate was not familiar with the actions of the lift coil disconnect switch - it will disconnect the rod from bank D but the rod cannot then be withdrawn with the switch open Correct answer l
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Question #34 McGuire Nuclear Station SRO Exam
l Bank Question: 402 Answer: C 1 Pt(s) - Unit 1 is responding to a LOCA. Given the following conditions inside containment:
0200 0210 0220 0230 a
Containment Pressure (psig) .1 .3 Safety injection occurred at 020 At what time do the NCPs lose KC and RN cooling flow and when can the operators regain restore cooling flow by using reset pushbuttons?
Lost cooline flow Able to reenin cooline flow l I l Distracter Analysis: Incorrect: - setpoint is too low Plausible: - this is the hi containment pressure setpoint - causes an SI which causes a phase A containment isolation Incorrect: - pressure too low l Plausible: - if the candidate thinks that hi containment pressure causes a loss of flow and this cannot be reset until after the CPCS reset at 0.35 psig Correct answer: - setpoint is hi-hi containment pressure > 3.0 psig gets a phase B isolation - can be reset immediately even with the signal present Incorrect: - actuation setpoint is correct but reset point is wrong Plausible: - if the candidate links that pressure must drop to clear
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the signal before allowing a reset For Official Use Only Ques._402 Page 36
m Question #35 McGuire Nuclear Station SRO Exam Bank Question: 403 Answer: C 1 Pt(s) Which one of the following instmments would be the most reliable indication when core voiding was occurring during a LOCA severe accident? ; Source Range Excore Nuclear Instruments Intermediate Range Nuclear Instruments Wide Range Neutron Instruments Power Range Nuclear Instruments Distracter Analysis: Incorrect: - not environmentally qualified for LOCA conditions Plausible: - Source Range instmments will detect the change in moderator effectiveness caused by voiding - used during TMI accident Incorrect: - not environmentally qualified for LOCA conditions Plausible: - instruments will detect the change in moderator effectiveness caused by voiding Correct: - these instruments are environmentally qualified IAW RegGuide 1.97 Incorrect: - not environmentally qualified and not sensitive at low power levels Plausible: - if candidates do not consider the source temi for the core when the voiding commences - i.e. shutdown low neutron flux levels
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For Official Use Only Ques __403 Page 37
e Question #36 McGuire '.Juclear Stat 9n SRO Exam Bank Question: 404 answer: C 1Pt(s) Unit 1 is responding to a small break LOCA. Containment pressure reached 3.5 psig. The Subcooling Margin Monitor currently indicates -35 # Which of the following statements best describes the status of subcooling in the core? The coreis subcooled by 35 F The core is superheated by 35 *F The core is superheated by at least 35 *F The core is subcooled by at least 35 *F Distracter Analysis: Incorrect: - the core is not subcooled Plausible: - if the candidate reverses the meaning of the indication -
believing that it is measuring subcooling not margin to subcooling
, Incorrect:- subcooling may be greater than 35 *F
- Plausible: - on face value Correct answer - the subcooling margin monitor is at the high end of the measuring range and subcooling may be greater than 35 *F Incorrect: - the core is not subcooled l Plausible: - if the candidate reverses the meaning of the indication -
believing that it is measuring subcooling not margin to subcooling l
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- Question #37 McGuire Nudear Station SRO Exam Bank Question: 405 Answer: D 1 Pt(s) Unit 2 is conducting a plant start up. The 2D S/G control valve fails open and causes a feedwater system transient. No reactor trip occurs due to efTective operator control. Given the following conditions and events:
0200 0210 Ol2_Q 0230 Turbine output 38% 39% 40 % 42 %
S/G CF Control Valves Position 2A S/G (% open) 39 % 25 % 24 % 40 % I 2B S/G (% open) 35 % 22 % 23 % 39 %
2C S/G (% open) 42 % 21 % 26 % 45 %
2D S/G (% open) 45 % 100 % 80 % 50%
All 4 CF bypass valves were closed at 020 During what time period (if any) will the main feedwater pumps trip? ) Will not trip Distracter Analysis: Incorrect: - AMSAC trip not active < 40% power Plausible: - loss of feedwater control starts at this time frame Incorrect:- AMSAC trip not active < 40% power Plausible: - 3/4 feedwater paths are blocked (CF valves position s 25% closed) Incorrect: - no longer have 3/4 feedwater paths blocked Plausible: - AMSAC no longer blocked > 40% power Correct answer - the AMSAC logic requires 3/4 flow paths to be
~ blocked (CF valves to close) when turbine impulse pressure > 40%
l power. This is never met. When AMSAC actuates, the MFP turbine is tripped For Official Use Only Ques __405 Page 39
Question #38 McGuire Nuclear Station SRO Exam Bank Question: 406 Answer: D 1 Pt(s) Unit 2 is operating at 100% power in a normal lineup. Reactor pressure drops to 1953 psig because of a spray valve malfunction. Which one of the following events will cause the turbine driven CA pump to start but NOT the motor driven CA pumps? AMSAC Low S/G NR level on 2/4 S/Gs Both main feedwater pumps trip i Loss of VI header pressure Distracter Analysis: The drop in reactor pressure is provided for the candidate to have to consider the effect of P-11 circuitry on the CA start logic (no effect). Incorrect: - AMSAC causes only the motor driven CA pumps to i start, not the turbine driven CA pump Plausible: - a valid CA pump start signal l Incorrect: - cause BOTH the MD and TD CA pumps to start Plausible: - a valid CA pump start signal Incorrect: -causes BOTH MD and TD CA pumps to Plausible: - a valid CA pump start signal Correct answer - causes TD CA pump steam supply valve to open
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l Question #39 McGuire Nuclear Station SRO Exam I l
Bank Question:407 Answer: D ;
1 Pt(s) Unit I has a liquid radioactive waste release in progress from the Ventilation Unit Condensate Drain Tank (VUCDT) through the RC system. All lineups and authorizations have been properly made in accordance with OP/0/B/6200/35 using the normal path. 2 RC pumps are the minimum required under LWR documen Given the following initial conditions: .
. 3 RC pumps are running
- 1 EMF-44 correctly set for trip 1 and trip 2 activity limits e MRIRR = 75 GPM based on boron concentration e No other releases are in progress What automatic actions would terminate the release? WM-46 will close automatically if 1 RC pump trips WM-46 will close automatically when EMF-44 reaches the trip 2 setpoint i WL-320 and WP-35 will close automatically if 1 RC pump trips WL-320 and WP-35 will close automatically when EMF-44 reaches the trip 2 setpoint j
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Distracter Analysis: ! Incorrect: - RC pump interlock will not actuate - set at 2 pumps (minimum required on LWR document). WM-46 is isolated and not used anymore as a release pat ,
Plausible: - this was formerly the nomial release path l Incorrect: - WM-46 receives a closing signal from EMF-44 but this is not the normal path for a release. WM-46 is isolated and not used anymor I Plausible: - this was formerly the normal release path Incorrect: - RC pump interlock will not actuate - set at 2 pumps !
(minimum required on LWR document).
Plausible: - this was the correct answer on the last NRC exam -
modified the question , Correct answer - changed this distracter from the last exam l
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l Question #40 McGuire Nuclear Station SRO Exam l
Bank Question: 408 Answer: C 1 Pt(s) The following parameters are observed for Hydrogen and Oxygen concentration in the Waste Gas Holdup syste Hydrogen Concentration 5.3% 6.8% 7.5% 8.5%
Oxygen Concentration 1.2% 1.5% 4.2% 7.1%
What is the earliest time that the operator is required, by procedure, to immediately suspend all additions to the Waste Gas Holdup System to prevent flammable limits from being exceeded?
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The flammability limits are already sampled and explicitly tested in the JPM questions Incorrect: - Oxygen concentration < 4% - no flammability concern Plausible: - if candidate thinks that Hydrogen > 4% alone is sufficient to require termination Incorrect: - Oxygen concentration < 4% - no flammability concern Plausible: - if candidate thinks that Hydrogen > 4% alone is sufficient to require termination Correct: - Oxygen concentration reaches 4% and Hydrogen > 4% Incorrect:- required at 0230 Plausible: - if the candidate does not know flammability tech spec restrictions of TS 3.11.2.5 - an immediate action LCO l
For Official Use Only Ques _408 Page 42
- Question #41 McGuire Nuclear Station SRO Exam i
Bank Question: 409 Answer: B
1 Pt(s) Unit I was shutting down in mode 3 when a fuel handling accident in the refueling canal area occurred. A spent fuel bundle was smashed into the side of the canal and a section of cladding was damaged. Given the I following conditions:
- Gamma radiation lev, are 10 R/hr at a distance of 10 ft from the .
bundle l
- The refueling bridge area radiation monitor,1 EMF-16, is 20 ft away from the suspended bundle
- 1 EMF-16 trip 1 setpoint = 1 R/hr
- IEMF-16 trip 2 setpoint = 6 R/Hr I
e A cloud of radioactive particulate gas surrounds the fuel handling bridge i e Beta radiation levels from this cloud are 25 R/hr What is the correct description of the alarm status of 1 EMF-16 and the containment evacuation alann? No alarms will be present on IEMF-16 :he containment l evacuation alarm will not sound EMF-16 will be in trip I alarm - the containment evacuation alarm will not sound EMF-16 will be in trip 1 alarm - the containment evacuation alarm will sound
! EMF-16 will be in trip 2 alarm - the containment evacuation alarm will sound Distracter Analysis:
The gamma radiation levels at 20 ft will be:
If point source modelis used:
y = 10 R/hr (10)2 /(20)2 y = 2.5 R/hr > trip 1 but < trip 2 Ifline source model used:
y = 10 R/hr(10/20)
y = 5 R/hr > trip 1 but < trip 2 Incorrect: -y > trip 1 For Official Use Only Ques.409 Page 43
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i Question #41 McGuire Nuclear Station SRO Exam Plausible: - if the candidate makes an error in computing the point source y radiation levels Correct: - y > trip 1 but the containment evacuation does not sound until IEMF-16 exceeds the trip 2 condition Incorrect: - y > trip 1 setpoint - BUT the containment evacuation alarm will NOT sound until 1 EMF-16 exceeds the trip 2 condition (unless blocked)
Plausible: - if the candidate thinks that the containment evacuation alarm sounds at trip 1 Incorrect: - y < trip 2 = no alarm Plausible: - if the candidate adds in the beta dose rate of miscalculates the y dose rate. 1 EMF-16 is a GM tube and will not detect beta radiation .
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i Question #42 McGuire Nuclear Station SRO Exam
! I Bank Question: 410 Answer: C 1 Pt(s) Unit 1 is in mode 4 preparing for refueling. Given the following conditions:
e NC pressure = 400 psig
- Tave = 200 'F e Pressurizerlevel = 25%
. Boron concentration = 230 ppm Which one of the following statements correctly describes what must be added to the NC system to change boron concentration to 1120 pp REFERENCES PROVIDED Data Book OP/1/A/6100/22 Enclosure 4.3 Sections 5.1 and add 8501 - 10,000 gallons boric acid l add 10,001 - 11,500 gallons boric acid add 11,501 - 13,000 gallons boric acid l add 13,001 - 14,500 gallons boric acid Distracter Analysis:
230 - 500 ppm = 2743 gal 500 - 700 ppm = 2106 gal 700 - 900 ppm = 2174 gal 900 - 1120 ppm = 2475 gal Total = 9498 gal boric acid 9498 x 1.28 (K) = 12,157 gal boric acid Incorrect: - did not chose any K factor from table Plausible: - ifcandidate forgets to use K factor Incorrect: - chose incorrect K factor from table 5.2 or some other calculation error Plausible: - if candidate misreads plant conditions ! Correct answer
! Incorrect answer - chose incorrect K factor from table 5.2 -
I selected 1.4 for water solid condition or some other calculation error Plausible: - if candidate misreads plant conditions j t
For Official Use Only Ques _410 Page 45
Question #43 McGuire Nuclear Station SRO Exam Bank Question: 411 Answer: B 1 Pt(s) Unit 2 is has recently restarted afler a refueling outage and is operating at 100% power. A previously isolated Mixed Bed Demineralizer (that had ,
been in use extensively prior to the shutdown) was inadvertently placed in l service. All primary chemistry parameters are within specification )
Which one of the following statements best describes the expected effects on the chemistry of the NC system? Boron concentration will increase due to the exchange of H,BOi ions from the demineralizer bed back into the coolant causing a net decrease in reactivit Boron concentration will decrease due to H,BOi ion removal from the coolant into the demineralizer bed causing a net increase in reactivit Boron concentration will not change because the mixed bed demineralizer will be in equilibrium with the NC Boron concentration, having achieved saturation conditions while operating on line in the previous fuel cycl Boron concentration will not change because the mixed bed demineralizer only exchanges impurity ions for H,BOi ion The chemical impurity levels in the coolant are not significant and current chemistry impurity levels in the coolant are within specification Distracter Analysis: Incorrect: - H BOfions will not increase and reactivity will increase
not decrease - see below Plausible: - if the candidate confuses the equilibrium conaitions Correct: - the demin is saturated with H2 BOfions at equilibrium concentration levels from EOL - where NC system Boron concentration was much lower than BOL conditions. Hence, until a new equilibrium is established, the demineralizer will absorb H2 BOi ions causing a reduction of H2 BOiand an increase in reactivit Incorrect: - the equilibrium conditions are not the same - H2 BOi levels are much lower at EOL in the coolant so equilibrium conditions in the demineralizer is also lower l
For Official Use Only Ques _411 Page 46
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l l- Question #43 McGuire Nuclear Station SRO Exam l
L Plausible: - if the candidate is confused over the establishment of
- , equilibrium conditions in the coolant
- Incorrect
- - mixed bed demineralizers will exchange H2 BOfions only until they reach saturation condition Plausible: mixed bed demineralizers do exchange impurity ions for H 2BOfions - only part of the chemical process in effect I
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Question #44 McGuire Nuclear Station SRO Exam Bank Question: 412 Answer: A 1 Pt(s) Unit 1 is responding to a LOCA. Given the following initial conditions:
. Safety injection actuation occurred at 0150
- MSIVs are shu . Phase B containment isolation has occurred The operators reach step 2 in E3-1.1 requiring a reset of the safety injection signa Given the following parameter trends at 0200:
e NC pressure = dropped to 1850 psig then stabilized at 1951 psig
. Steamline pressure = 771psig - deci sing slowly e Containment pressure = 2.2 psig - decreasing slowly Given the following sequence of operator actions:
0202 Blocks the low steam line pressure SI signal 0203 Blocks the low PZR pressure SI signal 0204 Resets the phase B isolation signal What is the earliest time that SI could be reset? Distracter Analysis: (
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This question will test if a candidate understands that safety injection can be reset even with valid SI actuation signals still present Correct answer - safety injection can be reset after a 60 minute timer has elapsed and the train related reactor trip breaker has opened (P-4). None of the SI signals being present will prevent reset of SI. Once reset, only manual SI is available Incorrect:- SI already reset at 0200 Plausible: - the steam line low pressure SI can be blocked < P-11 i
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Question #44 McGuire Nuclear Station SRO Exam L Incorrect:- SI already reset at 0202 Plausible: - this will block the low pressurizer pressure SI signal Incorrect:- Si already reset at 0202 l Plausible: - this action will block hi-hi containment pressure SI l signal I
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m Question #45 McGuire Nuclear Station SRO Exam Bank Question: 413 Answer: A 1 Pt(s) What is the source of power to the Upper Containment Air Ilandling Units? ' VAC Essential Load Centers VAC Shared Load Centers VAC Unit Load Centers VAC Auxiliary Load Centers i
Distracter Analysis: ,
i Correct answer: - these receive safety grade power off the ETA /ETB busses Incorrect: - they use essential load center power Plausible: - 600 VAC shared load centers do power some vital equipment like the control room air handling units Incorrect: - they use essential load center power Plausible: - this is the " normal" kind of power - non-safety grade l Incorrect: - No such type ofload center at McGuire Plausible: - the term " auxiliary" sounds like a reasonable choice for an auxiliary load like the AHUs.
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Question #46 McGuire Nuclear Station SRO Exam Bank Question: 414 Answer: B 1 Pt(s) Unit 2 was operating at 100% power when a double-ended high energy line break occurred and containment spray operated as designe Proper functioning of the ice condenser will ensure that peak containment pressure, which is reached (A) minutes following the event, remains i below (B) psi (A) minutes (B) osie i l i
Distracter Analysis:
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' Incorrect: - peak pressure reached in 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> (105 minutes)
Plausible: - peak pressure is correct,20 minutes is approximately the time that peak pressure is reached in a non-ice condenser plant Correct answer Incorrect: - peak pressure too high, time to reach is too short Plausible: - would be appropriate for some non-ice condenser plants Incorrect:- peak pressure too high Plausible: - if candidate does not know peak pressure for containment - 20 psig is appropriate for some non-ice condenser plants For Official Use Only Ques _414 Page 51
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Question #47 McGuire Nuclear Station SRO Exam Bank Question: 415 Answer: B
- 1 Pt(s) Unit 1 is operating at 100% power when an electrical transient causes ;
several condensate system pumps to trip. The operators take action to regain l
the system. Given the following conditions and events: I start 10 sec 20 sec 30 see CF pump 1 A Suction Pressure (psig) 451 238 235 225 CF pump 1B Suction Pressure (psig) 448 225 235 225
- Hotwell Pumps running 2 2 3 1
- Condensate Booster Pumps running 2 1 0 1
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What is the earliest time (if any) that BOTH main feedwater pumps will have tripped? seconds seconds seconds l feedwater pumps will continue to run under these conditions Distracter Analysis:
l Incorrect: - CF pump 1 A continues to run as suction pressure
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remains > 230 psig
! Plausible: - CF pump IB will trip on low suction pressure l Correct answer: - both CF pumps trip when 3/3 condensate booster j pumps trip - trip is irrespective of suction pressure l Incorrect: - trip criteria met but trip occurred earlier at 10 sec l
Plausible: - if candidate does not know that the CBPs are interlocked with the CF pumps Incorrect: - tripping criteria are met at 10 and 15 sec Plausible: - if candidate does not know CF pump trips For Official Use Only Ques __415 Page 52
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l Question #48 McGuire Nuclear Station SRO Exam Bank Question: 416 Answer: D 1 Pt(s) Unit 2 was operating at 100% power when a reactor trip occurred at 020 T,, is defeated for loop C while troubleshooting efforts are in progres Given the following conditions and events:
e a T RTD in loop 2C is out ofservice e at 0202, a T. RTD in loop D fails high (*)
0201 0202 0203 0204 T,,, loop A ( F) 556 554 552 551 j T,,, loop B ( F) 558 556 55t 552 T,,, !oop D (*F) 552 595* 595* 595* I
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At what time will feedwater isolation occur?
l Distracter Analysis:
The Tave signal for loop C is bypassed and does not provide an input to Lo T,,,. Coincidence for P-4 is normally 2/4 but becomes 2/3 when loop C Tave is bypassed Incorrect: - 1/3 logic satisfied for Lo T,,,
Piaus ble: - if candidate thinks that the bypassed loop C input satisfied the second channel for Lo T,,, Incorrect: - 0/3 logic satisfied for Lo T,,,
Plausible: - if candidate thinks that the bypassed channel and the
- failed channel will input 2/4 signals for Lo T,,,
l Incorrect: - 1/3 logic satisfied Plausible: - if candidate thinks that 2/3 signals from the failed loop D channel and loop A will satisfy the logic Correct answer - 2/3 signals from loops A and B for Lo T,,,
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For Official Use Only Ques _416 Page 53
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l Question #49 McGuire Nuclear Station SRO Exam l
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Bank Question: 425 Answer: A l
l 1P..i) Unit 1 is responding to a LOCA. Operators delayed terminating safety l injection upon reaching the applicable step in E- l In which one of the following scenarios is the repressurization of the NC system above 1845 psig most likely to occur?
' A 1 inch line break with two trains of ECCS operating A 1 inch line br. :. sith one train of ECCS operating A 2 inch lins break with two trains of ECCS operating A 2 inch line break with one train of T.CCS operating Distracter Analysis: Correct answer - the most flow and the smallest break size Incorrect: - less flow causes no repressurization Plausible: - less ECCS flow means less heat removal which could imply more pressure sooner Incorrect: - larger break size = more flow out of hole = lower break flow pressure Plausible: - if candidate thinks larger break size = more inventory loss = less cooling = more pressure Incorrect: - less flow - larger break = lower break flow pressure Plausibic: - if candidate does not understand what is going on in )
the thermal-hydraulics in the NC system !
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l For Official Use Only Page 54 I Ques _425
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Question #50 McGuire Nuclear Station SRO Exam l
l Bank Question: 426 Answer: B j 1 Pt(s) Unit 1 is operating at 50% power. Given the following conditions:
- Pressurizer pressure is 2235 psig o' Pressurizer Relief Tank (PRT) pressure is 25 psig i e PRT temperature is 115 *F
! * PRT levelis 81%
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e The PRT is being cooled by spraying from the RMWST 4 e A pressurizer code safety valve is suspected ofleaking by its' seat .
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, What temperature would be indicated on the associated safety valve
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discharge RTD if the code safety was leaking by?
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REFERENCESPROVIDED-Steam Tables l *F *F i l F *F i
l Distracter Analysis: Incorrect: too high Plausible: Correct answer Incorrect: too low Plausible: - if apply psig - psia conversion incorrectly Incorrect: - way too low Plausible:
i
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For Official Use Only Ques _426 Page 55
r Question #51 McGuire Nuclear Station SRO Exam l
Bank Question: 427 Answer: A 1 Pt(s) Unit 1 is conducting a natural circulation cooldown in ES-0.2. NV auxiliary spray has been started in step 18 to depressurize the NC system. How is pressurizer levei controlled during this cooldown and what effect will auxiliary spray flow have on pressurizer level control? Automatic positioning of NY-238 in response to the ILE pressurizer level control system. The auxiliary spray line taps off downstream of NV-238 and auxiliary spray flow has no effect on level contro Automatic positioning of NV-238 in response to the ILE pressurizer level crutrol system. The auxiliary spray line taps off upstream of NV-238 and total charging flow indicated i through NV-238 will be reduced by the amount of auxiliary I spray flow through NV-2 Manual control of NV-238 and NV-21 by the operator. The auxiliary spray line taps off upstream of NV-238 and NV-238 l must be manually positioned to prevent auxiliary spray from ,
raising pressurizer leve Manual control of NV-238 and NV-840 by the operator. The auxiliary spray line taps in ' 9m the ND system and NV-238 must be manually positic' 2 to prevent auxiliary spray from raising pressurizer leve Distracter Analysis: Correct answer Incorrect: - aux spray flow has no significant effect because auxiliary spray taps off the charging header downstream of NV-238 so ILE will not see any change in charging flow Plausible: - if the candidate does not know where auxiliary spray flow taps off the charging header Incorrect: - auxiliary spray taps off the charging header downstream of NV-238 so there is no need to go to manual control Plausible: - if the candidate does not know the difference between NV aux spray and ND aux spray flows Incorrect: - aux spray flow from the ND system is only used when ND is in shutdown cooling mode 4
_ Plausible: - if the candidate does not know that auxiliary spray flow comes from the NV system charging header through NV-21 l For Official Use Only Ques _427 Page 56
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Question #52 McGuire Nuclear Station SRO Exam {
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J Bank Question: 428 Answer: C 1 Pt(s) Unit 1 is operating at 50% during a plant startup with rod control in automatic. An RO trainee attempts to take manual control of rods but instead mistakenly moves the Bank Selector Switch to the "CBB" positicn l and moves the rods "IN' for 3 steps with the "In-Hold-Out" switch.
l Which one of the following . atements describes the situation? No rod motion occurred because control bank B will not insert until the bank overlap sequence requirement with banks C and D are satisfie Bank B moved into the core but this poses no problems as long l
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as AFD and QPTR are satisfied for this condition - which is likely at 50% powe l
' Bank B is below the rod insertion limits and must be restored to the full out position Control rods are out of alignment and the operator must immediately implement AP/1/A/5500/14 case IV, Control Rod Misalignmen Distracter Analysis: Incorrect: - the bank sequencer does not prevent control rod motion when rods are in manual Plausible: - this would be true with rods in automatic control Incorrect: - control bank B rods are below the RIL for power operation Plausible: - AFD and QPTR are part of the concerns - but so is shutdown margi Correct answer - Do not expect candidates to memorize the RIL values in the COLR - but certainly should know that at 50% power, banks A and B are below RIL Incorrect: - does not meet entry conditions for this procedure - rods not misaligned from group indication or from other rods in group
, Plausible: - rods are misaligned I
For Official Use Only Ques _428 Page 57
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I Question #53 McGuire Nuclear Station SRO Exam I i
l Bank Question: 429 Answer: A l
1 Pt(s) Unit 1 is operating at 50% power with one bus line in operation. Given the following indications:
- IPB fan 1B is tagged out ofservice j e "IPB Trouble" alarm annunciates on AD-11 i e The NLO reports that the intemal alarm " Vibration High Fan IA"is illuminated on the local alarm panel What action must be taken to comply with plant limits? Transfer loads to 11.2 second bus line and shut down the operating bus line isophase bus within 15 minutes Reduce load on the bus line to 10,000 amps within 15 minutes Reduce load on the bus line to 20,000 amps within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Reduce load to less than 48% reactor power within 30 minutes Distracter Analysis: Incorrect: - this action will not unload the Isophase bus Plausible: - if the candidate thinks that the isophase bus is on the bus line side of the PCBs Correct answer Incorrect: - load too high for one bus line - time requirement is too ;
long l Plausible: - the load restriction of 20,000 amps is for two bus lines in operation. The time requirement is typical of many important LCOs Incorrect: - not required for a loss of Plausible: - correct required action by AP-03 for a 50% load rejection due to a los-s of CF pump except that the required action is to reduce load below 50% - and IPB the loss of an IPB will cause a load rejection
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For Official Use Only Ques _429 Page 58
r Question #54 McGuire Nuclear Station SRO Exam
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Bank Question: 432 Answer: C l
l 1 Pt(s) Unit 1 is shutting down in mode 5 with all equipment hatches installed, preparing for refueling. Given the following conditions and events: !
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! e A containment entry is planned for the next shift.
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e The SRO directs the RO to purge the containment in preparation for the ;
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containment entr . The VP system is secured with all fans off and containment purge and I exhaust valves closed Which one of the following statements is correct regarding the purge of
- containment? The containment should be purged using the " REFUEL" mode Only the upper containment should be purged using the
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" REFUEL" mode The containment should be purged using the " NORMAL" mode l
l Only the lower containment should be purged using the
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" NORMAL" mode Distracter Analysis: I
! Incorrect: - this will over-pressurize the upper containment Plausible: - getting ready to " refuel" the plant Incorrect: - this will over-pressurize the upper containment
- Plausible
- - getting ready to " refuel" the plant l Correct answer: -
! Incorrect: - should not purge lower containment in mode 1-4 Plausible: - the normal mode is the correct mode to use l
For Official Use Only Ques _432 Page 59 L
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l Question #55 McGuire Nuclear Station SRO Exam l
l Bank Question: 433 Answer: A 1Pt(s) Unit 2 is responding to a SGTR on the 1C S/G in E-3. Which one of the following sets of S/G level indications is expected for the IC S/G as it fills up with NC system coolant?
0200 0201 0202 0203 0204 C S/G WR level = 72 % 74 % 75 % 75 % 75 %
1C S/G NR level = 98 % 100% >100%>100% >100% C S/G WR level = 72 % 74 % 77 % 83 % 86 %
1C S/G NR level = 98 % 100% >100%>100% >100% C S/G WR level = 90 % 92 % 93 % 93 % 93 %
IC S/G NR level = 98 % 100% >100%>100% >100% C S/G WR level = 90 % 92 % 96 % 100 % 100 %
IC S/G NR level = 98 % 100% >100%>100% >100%
Distracter Analysis: Correct: WR level is lower than NR level because it is calibrated for cold conditions. The WR variable leg tap and the NR variable leg tap are at the same level Incorrect: - WR S/G water level variable leg tap is at the same level as the NR tap. WR level will stop increasing when NR level reaches 100 %
Plausible: - if candidate does not know that taps for WR and NR are at same level (design feature) - overlap between NR and WR ranges is correct Incorrect: - WR level and NR level not correctly correlated - WR level does not continue to rise after NR level reaches the upper tap ,
setting l
Plausible: - if not familiar with cold cal of WR level and overlap i
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with NR levelrange l Incorrect: - WR level and NR level not correctly correlated with j l
cold cal of WR level - WR level too high for NR level Plausible- if not familiar with coid cal of WR level and overlap with NR level range j
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l For Official Use Only Ques _433 Page 60 1
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Question #56 McGuire Nuclear Station SRO Exam Bank Question: 434 Answer: C l
1 Pt(s) Unit 1 is responding to a faulted S/G inside containment from 100% powe Given the following conditions and events:
. Containment pressure = 3.5 psig - decreasing e Faulted S/G pressure = 200 psig - decreasing
. Intact S/G pressure = 850 psig - stable
- NC system pressure = 1750 psig - stable What is the MINIMUM set of actions that will allow the operator to manually reset and open the S/G PORVs? Reset MSI Reset hi-hi containment pressure Reset MSI Block Low steamline pressure SI Reset MSI Reset hi-hi containment pressure Block low steamline pressure SI Reset MSI Distracter Analysis: Incorrect: - resetting MSI is NOT sufficient to allow opening a PORV for the plant conditions because the low steam line pressure Si signalis stilllocked in Plausible: - ifcandidate thinks that the low steamline pressure signal is not locked in or has auto-cleared belew P-11 Incorrect: - MSI cannot be reset without first resetting steamline low pressure SI Plausible: - hi-hi containment pressure is the only MSI actuation signal that need not be reset to allow resetting MSI 1 Correct: - although the NC system is < 1955 psig and P-11 is not !
pret atly "in" - low steamline pressure SI was interlocked with P-11 when the event occurred so the signal is locked in and must be blocked Incorrect: - not the minimum set of actions - no need to reset hi-hi containment pressure a.; MSI can be reset with this signal still in l Plausible: - if candidate thinks that hi-hi containment pressure
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must be rese For Official Use Only Ques._434 Page 61
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Question #57 McGuire Nuclear Station SRO Exam Bank Question: 437 Answer: C
Which one of the following indications are correct for the existence of a
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1 Pt(s)
ground on the positive leg of the 125 VDC electrical system? Battery ground positive leg light burns dimly Battery ground negative light is out Battery ground positive leg light burns dimly Battery ground negative leg light burns brightly Battery ground positive leg light burns brightly .
Battery ground negative light is burns dimly l l
l Battery ground positive leg light is out '
Battery ground negative light is burns dimly Distracter Analysis: Incorrect: - a brightly burning lamp = grounded condition Plausible: - the positive leg is lit Incorrect: - reverse of correct answer Plausible: - if candidate reverses the correct answer in his/her mind
- a grounded condition often leads to a reduction in voltage and l lamps glow dimly Correct Answer: Incorrect:- positive leg lamp is out Plausible: - if the candidate thinks a grounded condition would cause the light to extinguish
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l For Official Use Only Ques _437 Page 62
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l Question #58 McGuire Nuclear Station SRO Exam Bank Question: 438 Answer: D l
1 Pt(s) During a blackout start of a diesel generator, what causes the DG field flashing to occur?
l . Residual magnetism in the DG generator windings i VAC vitalinstrument power VDC control power from the battery charger VDC DG control power from the DG battery l
Distracter Analysis: ' Incorrect: - field flashing provided by the DG battery Plausible: - this is the " Navy" answer - and is used by some DGs Incorrect: - field flashing provided by the DG battery Plausible: - 125 VAC vital instrument power is extremely reliable Incorrect: - although the battery charger supplies 125 VDC DG control power which supplies field flashing current during normal starts, this charger is not powered during blackout starts (where 120 VAC control poweris not present)
Plausible: - the answer for non-blackout starts Correct answer l
For Official Use Only Ques _438 Page 63 Question #59 McGuire Nuclear Station SRO Exam Bank Question: 439 Answer: 8 l
l 1 Pt(s) What automatic action (s) will occur if radioactive particulate levels exceed the 1 EMF-35(L) trip 2 alarm setpoints in the unit vent exnaust flow stream?
l l Closes WG-160 to terminate waste gas discharge Stops Auxiliary Building unfiltered exhaust fans Stops Spent Fuel Pool ventilation exhaust fans 1 Stops Containment Annulus ventilation fans l Distracter Analysis:
l Incorrect: - does not effect WG-160 1 Plausible: - trip 2 alarm on EMF-36(L)(Unit Vent Gas) causes this l action ' Correct answer - 1 EMF-35(L) monitors this exhaust stream Incorrect: - does not stop these fans Plausible: - EMF-35 monitors exhaust from Spent Fuel Pool area Incorrect: - does not stop VE fans Plausible:- monitors VE exhaust i
l For Official Use Only Ques _439 Page 64
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Question #60 McGuire Nuclear Station SRO Exam '
Rank Question: 441 Answer: A 1Pt(s) Unit 1 is operating at 100% power when the following alarms are received:
. PRT Hi Temp e PRT Hi Press Assuming that the alarms were due to the slow heat up of the PRT fium containment ambient conditions, how can the operators clear these alarms? Vent the PRT to the waste gas system Recirculate the PRT water through the NCDT Heat Exchanger Vent the PRT into the containment through the HEPA fliter Recirculate the PRT water through the WDT Heat Exchanger Purge the PRT with Nitrogen to the shutdown tank via the waste gas header Spray the PRT from the RMWST while draining to the WDT Purge the PRT with Nitrogen to the VCT Si iray the PRT from the RMWST while draining to the NCDT Distracter Analysis: Correct answer Incorrect: - the WDT is not used to drain the PRT. In addition, this path requires containment entry and would not be approved at 100%
powe Piausible: - WDT accepts radioactive waste from a number of primary systems and has a heat exchanger, the PRT can be vented imo containment through a manual valve Incorrect: - cannot drain to the WDT Plausible: - Purging with N2 to the WG system is an acceptable method of reducing pressure l Incorrect:- cannot purge to the VCT l Plausible: - spraying from the RMWST while draining to the
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NCDT is an attemative method for reducing temp.
i For Official Use Only Ques _441 Page 65
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I Question #61 McGuire Nuclear Station SRO Exam l I
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Bank Question:443 Answer: B 1 Pt(s) Unit I was responding to a LOCA inside containment when the following l
alarms were received in the control room: j (
- 1A VEFilter BedHi Temp on the annunciator panel
- VE Filter Hi Temp > 220 *F on the HVAC panel
- VEFilter Fire > 325 F on the HVAC panel l
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What complete set of automatic actions are expected as a result of these alarms? stops VE fans :
I o stops VE fans and shuts off filter heaters
! stops VE fans and starts RFY deluge l l stops VE fans, shuts off filter heaters and starts RFY deluge j Distracter Analys!s:
' Incorrect: - also shuts off filter heaters Plausible: - if candidate does not know that filter pre-heaters are interlocked to this alarm Correct answer Incorrect: - does not start RFY deluge automatically Plausible: - opening deluge valves is a manual action if there is a fire in the filter bed ; Incorrect: -- does not start RFY deluge automatically Plausible: - opening deluge valves is a manual action if there is a fire in the filter bed I
i For Official Use Only Ques _443 Page 66 o
I Question #62 McGuire Nuclear Station SRO Exam
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Bank Question: 444 Answer: B 1 Pt(s) Which one of the following statements describes the operation of the Containment Annulus Ventilation System (VE) during a large break LOCA into containment? VE fans start in the exhaust mode at +3.0 psig in containment VE fans stop running when annulus pressure = -7.0 in. Hg VE fans cycle on and off between -1.2 in Hg and -7.0 in. Hg VE fans start in the exhaust mode at +3.0 psig in containment VE shifts into recire mode at annulus pressure = -4.2 in. Hg VE cycles dampers between recire and exhaust modes between-1.2 and -4.2 in. Hg l VE fans start at on an EMF-38,39,40 (containment monitors)
VE fans stop running when annulus pressure = -7.0 in. Hg VE fans cycle on and off between -1.2 in Hg and -7.0 in. Hg l VE fans start at on an EMF-38,39,40 (containment monitors) l
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VE shifts into recire mode at annulus pressure = -4.2 in. Hg VE cycles dampers between recire and exhaust modes between-1.2 and -4.2 in. Hg l
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Distracter Analysis: Incorrect: - VE Fans are not designed to cycle on and off to maintain annulus pressure - they trip off at -7.0 in. Hg but reset at
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-5.5 in. Hg. They would not recire through the filter trains if they tripped off to maintain pressure and Iodine removal would not be as effective Piausible: - tripping at -7.0 in. Hg is a safety trip to ensure annulus pressure is not dropped too low Correct answer Incorrect: - VE fans are not started by EMF monitors, VE Fans are not designed to cycle on and off to maintain annulus pressure - they trip off at -7.0 in. Hg but reset at-5.5 in. Hg Plausible: - VP and VQ systems are controlled by EMF 38,39,40 Incorrect: - VE fans are not started by EMF monitors l
Plausible: - VP and VQ systems are controlled by EMF 38,30,40 For Official Use Only Ques _444 Page 67
i Question #63 McGuire Nuclear Station SRO Exam Bank Question: 445 Answer: C 1 Pt(s) What bus supplies power directly to the l A KC pump? IELXA load center to MCC EMXG load center to MCC ETA bus ITA bus l
Distracter Analysis: Incorrect: - not powered off 600 VAC essential power Plausible: - many large essential motors are powered off this load center - safety grade power Incorrect: - not powered off 600 VAC shared essential power Plausible: - many large important motors powered off this load center- safety grade power Correct answer-4160 VAC power Incorrect:- not powered off ITA Plausible: - 1ETB is powered from ITA through SATA transformer - many large loads are powered directly off this 6.9 kV bus
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For Official Use Only Ques _445 Page 68 l Question #64 McGuire Nuclear Station SRO Exam Bank Question: 446 Answer: C 1 Pt(s) Unit 1 is responding to a loss of all AC power. Unit 2 is operating at 100%
power. One hour into the event, it becomes necessary to cycle the pressurizer PORV on Unit I to reduce NC pressure. VI header pressure in the control room indicates 25 psi What provides the motive force to cycle this valve under these conditions? Air pressure from the VI system header Air pressure from the Auxiliary Building Instrument Air Tank ,
Header Nitrogen pressure from the Cold Leg Accumulators l VAC shared essential power from unit 2 Distracter Analysis: 1 Incorrect: - 25 psig is not suflicient pressure to cycle a PORV - VI Header is not safety grade and will depressurize in a blackout i Plausible: - many sites have safety grade VI systems Incorrect: -- does not provide air to PZR PORVs Plausible: - the aux. building air tank provides blackout protection for most AOVs that are required to operate during a blackout i Correct answer l Incorrect: - the PORV is operated by Nitrogen from the CLA Plausible: - many safety related MOVs are powered off shared essential power load centers l
For Official Use Only Ques _446 Page 69
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Question #65 Mc(:uire Nuclear Station SRO Exam Bank Question: 447 Answer: D l
1 Pt(s) Unit 1 is shutdown, mode 6, in a refueling outage. Given the following conditions:
. Containment airlock doors are both open
. A full shift of qualified maintenance personnel are available inside l
. containment
. The Refueling SRO is in the control room
. The Operations Fuel Handling Supervisor is inside containment Refueling has been completed and the Operations Fuel Handling Supervisor (who is not a qualified SRO) requests permission to latch all control rods to prepare for the reactor startup. What additional requirements must be met (if any) to proceed with latching rods? Latching rods may proceed at the discretion of the Fuel Handling Superviso Latching rods may not proceed until after containment integrity ,
l has been restore !
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' Latching control rods may not proceed until after the Refueling SRO arrives inside containment to supervise.
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1 Latching control rods may not proceed until after the Refueling SRO arrives inside containment and containment integrity has been restore )
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Distracter Analysis: l l Incorrect: - the Refueling SRO is required to supervise this j evolution and contaimnent integrity must be restored !
l Plausible: - if the candidate does not recognize that latching rods is I a core alteration or doesn't recognize that this requires containment
- integrity to be established Incorrect
- - the Refueling SRO is required to supervise this evolution j
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, Plausible: - if the candidate does not recognize that latching rods is a l core alteration Incorrect: - containment integrity must first be established Plausible: - core alterations requires SRO coverage and containment integrity Correct answer For Official Use Only Ques _447 Page 70
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Question #66 McGuire Nuclear Station SRO Exam Bank Question: 448 Anawer: C 1 Pt(s) Unit 1 is conducting a plant startup and power ascension after a refueling outage when a pressurizer PORV failed open. Given the rollowing events '
and conditions:
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- The reactor had not tripped prior to 0200 e The P-13 light extinguished prior to 0200 ,
e Operators were in the process of transferring CF control from the CF l
bypass to the main CF control valve on the A S/G at 0200
. The operators did not manually trip the reactor at any time
- *The operators tripped the main turbine at 0220 e Assume that all redundant instrument channels are in service and read within 0.5% or 1.0 psig and all trip setpoints are within required tolerances Given the following conditions and events:
l l 0200 0210 0220 0221 Reactor Power (%) 12 % 9% 8% 5% ;
Turbine Impulse Press (%) 12 % 8% *0% 0% I PZR pressure (psig) 1946 1942 1948 1955 '
1 A S/G level (%) 35 19 18 16 IB S/G level (%) 40 35 39 41 1C S/G level (%) 38 35 40 39 ID S/G level (%) 35 32 38 42 At what time (if any) would the reactor have been in a tripped condition?
l I ' l the reactor would not be tripped Distracter Analysis: Incorrect: - P7 is"out" (<10% for Rx power and Turb imp) so the at-power reactor trips are blocked - low pressurizer pressure (<1944 psig)is blocked
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For Official Use Only Ques _448 Page 71
Question #66 McGuire Nuclear Station SRO Exam Plausible: - if the candidate does not recognize the P7 blocks the low pressurizer pressure trip < 1944 psig Incorrect: - the main turbine trip does not cause a reactor trip below P8 (<48%)
Plausible: - if the candidate does not recognize that the main turbine trip is blocked < P8 Correct answer: - the reactor trips on low SGWL in I A S/G - this trip is not blocked
{ Incorrect:- the reactor has tripped Plausible: - if the candidate thinks that low SGWL is blocked by P7/P8 or does not recognize that the low SGWL trip is "in" l
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l Nr Mclal Use Only Ques _448 Page 72
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Question #67 McGuire Nuclear Station SRO Exam l.
l Bank Question: 449 Answer: D l
l 1 Pt(s) Unit I was commencing a plant start up when a flooding event occurred inside containment. The operators entered E-0 and transitioned to FR-Z.2, Containment Flooding. Chemistry reports that the water inside containment has a high level of free Chlorin Which one of the following systems is the most probable source of the water? RV System RN rystem KC system RF system Distracter Analysis: Incorrect: - RV system water does not contain free Chlorine Plausible: - if candidate does not know chemistry requirements for RV water Incorrect: - RN system contains water from Lake Norman - does not contain Chlorine Plausible: - Asiatic clams are a problem in the RN system and Chlorination is one way of destroying the larva Incorrect: - KC-contains Chromates but not Chlorine for chemistry control Plausible: - if the candidate does not know chemistry requirements for KC water Correct answer - the RF system is Chlorinated to prevent Asiatic Clam degradation
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For Official Use Only Ques _449 Page 73
Question #68 McGuire Nuclear Station SRO Exam I
Bank Question: 451 Answer: B 1 Pt(s) Unit 1 is shutdown in a refueling outage. A VI header rupture occurs and i the VI system completely depressurizes. VI-820 was open at the time of the
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rupture. What effect does a total loss of the VI system have on the VS system?
! VI-820 will auto-close as VI header pressure decreases below 90
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psig and the VS air compressor will start automatically at 82 psig to maintain VS header pressure
\ VI-820 will auto-close as VI header pressure decreases below 82 l
psig and the VS air compressor must be manually started to (
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maintain VS header pressure Check valves in the VI- VS cross-connect line will close to j isolate VS system pressure before it drops below 90 psig ' VS pressure in the Fire Protection Pressurizer Tank will be lost -
until a VS air compressor can be starte Distracter Analysis: Incorrect: - the VS air compressor does not automatically start to maintain pressure - VI-820 auto-closes at 82 psig not 90 psig Plausible: - The system is safety significant, VI-820 does close at 82 psig and there is a separate VS air compressor - it just isn't in standby - requires operator action Correct answer Incorrect: - there are no check valves in this line Plausible: - this is another possible method to prevent depressurizing the VS header Incorrect: - the RF system tank is pressurized with high pressure VS air - which is maintained isolated from the VI header Plausible: - if the candidate does not know that high pressure VS is isolated
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For Official Use Only ques _451 Page 74
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Question #69 McGuire Nuclear Station SRO Exam i
Bank Question: 452 Answer: A 1 Pt(s) Unit 1 is conducting a plant startup and power ascension after a refueling outage when a pressurizer spray valve failed open. Given the following l l events and conditions:
I i
e The reactor had not tripped prior to 0200 e The P-13 light. extinguished prior to 0200
. The operators did not manually trip the reactor at any time
- *The operators tripped the main turbine at 0220 .
- Assume that all redundant instnunent channels are in service and read within 0.5% or 1.0 psig and ah trip setpoints are within required tolerances ;
l Given the following conditions and events:
0200 0210 0220 0221 Reactor Power (%) 12 % 11 % 8% 7%
Turbine Impulse Press (%) 12 % 9% *0% *0%
PZR pressure (psig) 1949 1949 1942 .1955 NCPs running 4 0 0 0 At what time (if any) would the reactor have been in a tripped condition? the reactor would not be tripped Distracter Analysis:
l Correct: - the NCP trip occurs before P7 auto-blocks the at-power reactor trips - BOTH turbine impulse pressure and reactor power must be to be below 10% to actuate P Incorrect: - P7 auto-blocks the at-power trip for Pzr low pressure (<l945 psig)
Plausible: - if the candidate does not recognize the P7 interlock Incorrect: <P8 blocks the turbine trip - reactor trip signal
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Plausible: - if the candidate does not recognize the effect of P8 Incorrect:- the trip occurs at 0210 For Official Use Only Ques _452 Page 75
o Question #70 McGuire Nuclear Station SRO Exam Bank Question: 453 Answer: A 1 Pt(s) Unit I was operating at 100% power when condenser vacuum started slowly degrading. All attempts to correct the problem proved unsuccessfu Given the following trends and conditions:
0200 0210 0220 0230 Condenser vacuum (inches of Hg) 2 .5 2 .0 Which of the following statements correctly describes the responsibility to decide to shutdown the plant? The NCO (who is the OATC)is required to manually trip the reactor at 0230 because the automatic turbine trip has not actuated and the main condenser vacuum continues to decrease uncontrollably toward the automatic trip setpoint The NCO (who is the OATC) is required to wait until the automatic turoD e f.ip setpoint has been exceeded before tripping the main turbin The NCO (who is the OATC)is required to obtain direction from the Shift Supervisor prior to tripping the reactor because I
the automatic turbine trip has failed to actuat Only the Shift Supervisor, not the NCO, can trip the unit at any time, ifin his/her opinion, conditions warrant this actio Distracter Analysis: Correct answer Incorrect: -OMP 4-3 requires the NCO to trip the reactor if pressure is decreasing uncontrollably toward the setpoint prior to reaching ,
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automatic action Plausible: - if candidate does not OMP 4-3 requirements. These conflict with the statement in OMP 2-2 - but are more conservative and take priority Incorrect: - the automatic turbine trip has not failed to actuate Plausible: - the plant is not required to be shutdown by any j procedure j Correct answer - this is the operations policy statement in OMP2-2 l (except the words say " shutdown" not " trip" but it conflicts with the guidance in OMP 4-3 - which takes priority i
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For Official Use Only Ques __453 Page 76
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Question #71 McGuire Nuclear Station SRO Exam l i
i Bank Question: 454 Answer: D l
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I 1 Pt(s) Unit 1 is responding to a LOCA inside containment. Given the following events and conditions:
- The operators transitioned from E-1 to ECA-1.1, Loss of Emergency Coolant Recirculation,
- The Control Room Operators have reached step 4 of ECA-1.1 which reads: j Depress "SS RESET"for thefollowing valves: l e INI-184B (RB Sump to Train B ND & NS)
What is the basis for taking this action in ECA-1.l? To ensure that the recirculation flow is not bypassing the containment sump through this flow pat l To align a flow path for makeup water to the FWST to extend l the time that it can be used for a suction source '
- To prevent the recirculation of radioactive water from the containment sump to the FWST To prevent inadvertent loss of FWST inventory due to automatic switchover and damage to the S/I pumps Distracter Analysis
- Incorrect: - this will not align a bypass flow path Plausible: - these valves are in the flow path to the FWST Incorrect: - this will not align a fill path to the FWST Plausible: - this is the basis for the next step in EAC- Incorrect: - this will not align a flow path for water to recirc to the FWST Plausible: - this is the basis for closing the NI mini-flow valves in another part of the EOPs Correct answer For Official Use Only Ques _454 Page 77
l Question #72 McGuire Nuclear Station SRO Exam
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- Bank Question
- 455 Answer: C 1 Pt(s) Unit I was operating at 100% power when 1NV-241 failed shu *
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The operators entered AP/1/A/5500/12 Case II (Loss of Charging)
. They transitioned to Case I when it became necessary to isolate normal letdown due to a loss of charging flow e Excess letdown was placed in service
. Enclosure I to AP/1/A/5500/12 (Operator Actions with Excess Letdown in Service) contains a requirement to assign a licensed Reactor Operator the specific duty ofisolating excess letdown Which one of the following statements correctly characterizes the reason for this compensatory action and describes the responsibilities of the operator assigned? In the event of a reactor trip, the failure to isolate excess letdown will cause a rapid loss of pressurizer inventory and a loss of i
pressurizer level during the outsurge. The operator assigned j this responsibility may have no other dutie ] In the event of a loss of KC flow, the failure to isolate excess letdown will cause over heating of the demineralizers. The operator assigned this responsibility may be assigned other dutie In the event of a safety injection, the failure to isolate excess letdown will result in voiding in associated KC system piping and subsequent water hammer upon restoration of flow. The operator assigned may be assigned other dutie In the event of a safety injection, the failure to isolate excess letdown will result in voiding in the seal water heat exchanger and water hammer in the piping to the VCT. The operator ;
assigned may have no other dutie Distracter Analysis: I Incorrect: - a reactor trip does not cause isolation of excess letdown Plausible: - loss of PZR inventory is a concem for many events i Incorrect: - a reactor trip does not cause isolation of excess letdown Plausible: - cooling the excess letdown to a lower temp is listed as a l criteria in the training material
' I For Official Use Only Ques _455 Page 78
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Question #72 McGuire Nuclear Station SRO Exam Correct answer Incorrect: - the voiding problem is in the KC Hx - not the seal water Hx - the operator MAY have other duties assigned under the compensatory action Plausible: - the excess letdown flow path is coc'ed by the seal water Hx I
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I For Official Use Only Ques _455 Page 79 l
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Question #73 McGuire Nuclear Station SRO Exam l
Bank Question: 456 Answer: C 1Pt(s) The Turbine Driven CA (TDCA) Pump Govemor Speed Signal Isolation Valve, SAEP5020, is cunently closed and locked on Unit I to prevent use during emergency operations.
l How can the TDCA pump speed be adjusted during an emergency, and why are the TDCA pump speed control potentiometers isolated by SAEP50207 TDCA pump speed is controlled automatically by an input signal from the steam generator water level control system to the speed control potentiometer circuit. SAEP5020 is isolated to prevent interference with the automatic speed control syste I TDCA pump speed is adjusted using a manualloader in the control i aom that bypasses the speed control potentiometers and provides a control air signal directly to the TDCA pump governor valve. This is necessary because SAEP5020 is not environmentally qualified and may failin the event of a CA Pump room flooding even TDCA pump speed is adjusted locally by dispatching an NLO to !
the pump and using the Woodward Governor speed setting
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knob to control speed if required. SAEP5020 is not used !
because the TDCA pump speed control system is not safety I grade and does not meet seismic qualificatio l TDCA pump speed is adjusted using a manualloader on the TDCA pump local panel that bypasses the control room speed control potentiometers. This is necessary because SAEP5020 does not meet Appendix "R" requirements for train separatio t Distracter Analysis: Incorrect: - there is no input signal from SGWLC to the TDCA pump govemor Plausible: - if candidate does not know how CA speed is controlled ; Incorrect: - There is no manual loader that bypasses this valve. The manual loader in the control room uses SAEP 5020 to control the TDCA pump speed Plausible: - a manual loader does exist in the control room i Correct answer:
For Official Use Only Ques _456 Page 80
Question #73 McGuire Nuclear Station SRO Exam Incorrect: - although there is a manual loader at the local control panel, it also control SAEP 5020 directly and is not used during emergencies. SAEP 5020 is not an Appendix R concern Plausible: - Appendix R requires train separation of safety grade control circuits
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J For Official Use Only Ques _456 Page 81
Question #74 McGuire Nuclear Station SRO Exam -
Bank Quesdon: 457 Answer: D 1Pt(s) Unit 1 is in mode 4 with the NC system at 205 F preparing for a refueling outage. At 0600, Maintenance is planning on entering containment to stage equipment for a repairjob on the NS system spray heade Given the following conditions:
. Train A of NS is tagged out
. Pressurizer hatch plug is installed e Containment air lock doors are both shut and locked e Personnel access door between upper and lower containment is closed and locked Which one of the .ollowing situations or conditions would be allowed? Tagging out train B of the NS system at 0200 to prepare for the repair Removal of the pressurizer hatch plug between upper and lower containment at 0400 to prepare for repair team and equipmen Opening both containment sirlock doors at 0400 to allow passage of the repair team and equipmen Opening a personnel access hatch door between upper and lower containment at 0530 to allow passage of the repair team and equipmen Distracter Analysis:
Requires knowledge ofI hour LCOs to answer this question Incorrect: - violates tech spec 3.6.2 in mode 4 and would require entry into tech spec 3. Plausible: - if candidate does not know this tech spec requirement Incorrect: - violates tech spec 3.6.14.- removal of the uluipment hatch not allowed in mode 4 for >2 hours Plausible: - removal allowed for " personnel transit entry" - but this clearly would not apply in this case (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO) Incorrect: - violates tech spec 3.6.2 - not able to open both airlock doors at the same time in mode 4 - I hour LCO Plausible: - this tech spec allows entry into containment in mode 4 as long as one door remains shut Correct answer - as long as it is open for " personnel transit entry" l For Official Use Only Ques _457 Page 82
Question #75 McGuire Nuclear Station SRO Exam
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Bank Question: 458 Answer: D '
1 Pt(s) Unit 1 is responding to a small break LOCA. Given the following
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l conditions and events:
- e A reactor trip and safety injection have occurred l e The operators have transitioned to E-1 e The VI system has been slowly depressurizing
. A loss of RN has occurred Which one of the following statements is correct regarding the response to a loss ofVI? The use of any AP is not authorized after E-1 has been entered because actions taken could impact the EP assumptions for plant configuration. All APs are written assuming that SI has NOT actuated and AP implementation would not correctly j address the plant configuratio j The EPs are written to symptomatically address all potential j safety significant events that may occur through the FRP network. Use of APs is not required when EPs are in plemente j The SRO may elect to implement one AP in addition to E- Multiple concurrent use of APs is prohibited. The SRO should i implement AP-20 (Loss of RN) first and complete this procedure before implementing AP-22 (Loss of VI) The SRO may elect to implement more than one AP r.oncurrently with EPs depending on the priority and manpower available. Caution should be used if SI has actuated because the APs are written as if SI has NOT actuate Distracter Analysis: Incorrect:- use of APs is allowed Plausible: - APs are written assuming that SI has not actuated Incorrect: - Not all problems are addressed within the EP networ Plausible: - this is a common misperception Incorrect: - more than one AP may be ir'plemented at the discretion of the SRO Plausible: - multiple use of APs - especially when SI has actuated can be risky Correct: - per OMP-4-3 7.19 page 18 For Official Use Only Ques _458 Page 83
n Question #76 McGuire Nuclear Station SRO Exam l Bank Question: 460 Answer: B l
1 Pt(s) If a fire was reported in the McGuire OfTice Complex (MOC), which one of
- the following responses is correct by station procedures?
l Offsite fire departments are responsible for all fire suppression activities at the scene. The Fire Brigade must not leave the protected area and must be held in reserve fer station fire respons I Offsite fire departments are responsible for all fire suppression j activities at the scene. The Fire Brigade may provkle limited support if resources allo The Fire Brigade is responsible for the initial response at the scene. They are required to turn over control of the scene as soon as an offsite fire departe::::r-ives and immediately return to the protected area.
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l The Fire Brigade is responsible for fire suppression activities at the scene. An offsite fire elepartment may be called to provide limited support if additional resources are require Distracter Analysis: Incorrect: - the Fire Brigade is allowed to leave the protected area Plausible: - they have restrictions regarding leaving the site Correct answer Incorrect: - the Fire Brigade is not responsible fbr initial response and they are not required to return to *.he protected are Plausible: - if the candidate thinks that they cannot remain outside the protected are Incorrect: - The Fire Brigade is not primarily responsible to fight fire outside the protected are Plausible:
For Official Use Only Ques _460 Pag: 84
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Question #77 McGuire Nuclear Station SRO Exam Bank Question:462 Answer: C 1 Pt(s) Which one of the following actions is NOT allowed be taken in advance of the SRO reaching and reading the step in the applicable EP?
l CA may be reset and controlled to a S/G when that S/G reaches
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its normallevel setpoint
- NC pumps may be tripped on an Sp signal due to a loss of support functions unless the procedure has directed otherwise Safety injection may be reset Known leaks may be isolated if plant conditions require it to be done Distracter Analysis: Incorrect: - allowed per OMP 4-3 Plausible: - if candidate does not know the O'S requirements Incorrect: - allowed per OMP 4-3 Plausible: - if candidate does not know the OMP requirements Correct answer: - not on OMP 4-3 list of approved actions Incorrect:- allowed per OMP 4-3 Plausible: - if candidate does not know the OMP requirements i
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For Official Use Only Ques _462 Page 85
Question #78 McGuire Nuclear Station SRO Exam Bank Question: 463 Answer: D 1 Pt(s) Which one of the following is the proper method of verifying the position of an ND system manual rising stem globe valve in the " locked open" position if that valve has a lock and chain installed to prevent closure? Verify that the lock is installed and prevents movement in the closed direction Verify that the valve stem is in the full up position Verify that the lock is installed and prevents movement in the closed direction Turn the valve handwheel to the open direction to verify that it will not move and is securely setied on the backseat Remove the lock and chala Turn the valve handwheelin the open direction to 4 y that it will not move and is securely seated on the backseat Reinstall the lock and chain to prevent closure Remove the lock and chain Turn the valve handwheel in the closed direction to verify that the valve will move Turn the valve handwheel to the open direction Distracter Analysis: Incorrect: - must remove the locking device and tum the valve in the closed direction Plausible: - this is true for valves that are position tamper scaled Incorrect: - must first remove the locking device Plausible: - locks will generally permit slight movement in the open direction Incorrect: - never check a manual valve in the open direction as the valve may be closed and moving it to the open direction may inadvertently pressurize a system Plausible: - this would check that the valve was open - unless it was jammed on its seat l Correct answer - if the valve is open, it will close slightly but this i
will have no effect on the system. Ifit is closed, it will not move.
I For Official Use Only Ques _.463 Page 86 i:
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Question #79 McGuire Nuclear Station SRO Exam l
Bank Question: 464 Answer: C 1 Pt(s) Unit 2 is operating at 59% power. Which one of the following alarm conditions will cause a slow rate turbine runback on UJt 27 Transmission Line Overload Loss of one Main Feedwater Pump I Loss of Cooling Groups Low Stator Water Cooling Flow l
Distracter Analysis: Incorrect: - not a runback on Unit 2 i Plausible: - this is a runback on Unit 1 only - at the slow rate - !
when >56% turbineimpulse pressure i Incorrect: - causes a fast rate runback
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Plausible: - applicable to both units Correct answer per OP-MC-GEN-EHC page 61 Incorrect: - causes a fast rate runback Plausible: - applicable to both units l
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Question #80 McGuire Nuclear Station SRO Exam (
l Bank Ques 6on: 468 Answer: D
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1 Pt(s) A maintenance worker is preparing to disassemble the internals of a valve that was removed from the NC system in the mechanicc! maintenance sho This valve has been surveyed by Health Physics with the following results
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! * General radiation levels in the shop work area = <.1 mR/her e Radiation levels on contact with the valve = 0.3 mR/hr e External loose surface contamination = 750 dpm/cm2 p,7 e Intemal loose surface contamination = 5000 dpm/cm2 p,7 i
Which one of the following statements correctly characterizes the radiological work controls that are applicable to work on this valve? No RWP is required .hecause the valve in not inside the RCA No RWP or SRWP is required because the radiation levels in the work area and on the valve are sufficiently low A SRWP is appropriate because the work is a routine operation on a system component and the radiation levels are sufficiently low 1 A RWP is required because the valve is not a routine operation and is internally contaminated Distracter Analysis: Incorrect: - an RWP is required Plausibie: - if the candidate thinks that RWPs only apply to activities inside the RCA Incorrect: - RWP is required because ofintemal contamination levels Plausible: - the extemal radiation levels are < 0.5 mR/hr which is the threshold for requiring an RWP Incorrect: - this work is not a routine operation Plausible: - SWRPs may be used for routine, repetitive tasks Correct answer - the internal contamination requires an RWP l
For Official Use Onl Ques _468 Page 88 m-
i Question #81 McGuire Nuclear Station SRO Exam Bank Question: 470 Answer: B 1 Pt(s) . Unit 1 is shutdown in mode 5 during a refueling outage. Containment purge
' has been in progress far 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Which one of the following conditions would reautre the containment purge to be immediately terminated? j EMF-36(L) becomes inoperable
] EMF-39(L) becomes inoperable l l A S/G diaphragm was removed and the monitoring EMF increases by 75% but remains < trip 1 RP reports that the latest containment air sample shows that the concentration of the isotope Rb'" has increased by 15 times Distracter Analysis: Incorrect: - EMF-36 (unit vent monitor) is not the correct EMF monitor for a containment purge release - not required Plausible: - used for waste gas release Correct answer - EMF-39(L) is required to be operable because it is used to monitor the release rate Incorrect: - not a requirement to terminate the release rate Plausible: - > 50% is a requirement to resample the containment atmosphere Incorrect: - not a requirement to terminate the release rate I'lausible: - this is a requirement to resample the containment atmosphere For Official Use Only Ques __470 Page 89
Question #82 McGuire Nuclear Station SRO Exam l
Bank Question: 473 Answer: C
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1 Pt(s) Unit I was operating at 100%. I&E is conducting calibration of the VCT l tank level instrument and causing periodic VCT level alarms due to testin Which one of the following statements correctly describes the proper ;
procedural response to these alarms by the RO?
l The RO may use his/herjudgement to determine if announcing the alarms is appropriat < The RO must announced the alarm the first time that it actuates, then announcements may be suspended until testing is completed or until reaching the end of the shif The RO must announce all occurrences of the alarm during the testing - unless the control room SRO exempts announcement of alarms due to testing (do not need permission from higher authority). The RO must announce all occurrences of the alarm during the testing - only the OSM can exempt announcement of alarms due to testing (the control room SRO does not have authority to suspend alarm announcements at 100% power).
Distracter Analysis: I Incorrect: - RO cannot make this decision on his/her own Plausible: - correct for situations when responding to abnormal or !
emergency conditions l Incorrect: - RO cannot make this decision on his/her own Plausible: - correct ifin shutdown modes Correct answer Incorrect: - the control room SRO does have authority to suspend alarm announcements ;
Plausible: - OSM authority may be required for very important evolutions and decisions
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For Official Use Only Ques _.473 Page 90
Question #83 McGuire Nuclear Station SRO Exam Bank Question: 474 Answer: D 1 Pt(s) While responding to a small break LOCA, the operators reach step 7 of E-1 when the RO reports a red path on subcriticality has activated on SPD After investigating the condition, the STA reports that the source range NI indication is spiking causing the SPDS to intermittently display a red path and a suberiticality condition is not valid in F- Which one of the following statements describes the correct procedure transitions for this situation? Enter FR-S.1 and complete uutil the end. Transition back to step 7 of E- Enter FR-S.1 and complete until the end. Transition back to step 1 of E- Refer to FR-S.1 for information only while continuing on in E- Proceed on in E-1 without referring to FR- Distracter Analysis: Incorrect: - do not enter FR-S.1 because the condition is not valid Plausible: - entry into FRPs is normally expected in response to a valid red pat Incorrect: - do not enter FR-S.1 because the condition is not valid l
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Plausible: - entry into FRPs is normally expected in response to a valid red path. The return to step 1 is inserted to cause the candidate to think about the answer because there are 2 " enter into FR- distracters - normally means one of them is correct Incorrect: - no requirement to refer to FR-S.1 while in E-1 because condition is not valid Plausible: - this is true if ECA-0.0 has been entered. This provides an option shon ofignoring the procedure. Appears to be a conservative approach if candidate does not know correct answe Correct answer i
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For Official Use Only Ques._474 Page 91
p p 4 Question #84 McGuire Nuclear Station SRO Exam l ,
Bank Question: 475 Answer: D 1 Pt(s) Unit 1 is operating at 100% power when a station blackout occurred. The SRO diagnoses the following conditions:
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l . A total loss of onsite and offsite AC power has occurred {
l e The 4160 VAC buses remain deenergized !
e The reactor has not tripped, the rods are still out and power is not j decreasing What is the correct procedure flow path for this event? Enter E-0 and complete immediate actions l Transition to FR-S.1 and complete FR- i Transition to ECA- l l Enter ECA-0.0 directly and complete immediate actions Transition to FR-S.1 and complete FR- Return to ECA-0.0 and complete ' Enter FR-S.1 and complete FR- Transition to ECA-0.0 and complete Transition to ECA-0.1 to restore power l Enter ECA-0.0 directly and complete Transition to ECA-0.1 to restore power When directed to monitor CSFs, transition to FR- Distracter Analysis: Incorrect: - do not enter E-0 first before transiting to ECA-0.0 if you recognize the blackout. Do not monitor CSFs in ECA-0.0 so no transition to FR-S.1 is possible l
Plausible: - while it is not wrong to enter E-0 first if the blackout is not immediately recognized, it is not proper to transition to CSFs out l of ECA- Incorrect: - do not monitor CSFs in ECA-0.0 so no transition to FR-S.1 is possible Plausible: - may not realize that CSFs are not monitored in ECA-0.0 L Incorrect: - must enter ECA-0.0 as soon as blackout is recognized ;
Plausible: - may not know that ECA-0.0 must be entered
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immediately upon recognizing blackout Correct answer
i For Official Use Only Ques _475 Page 92 )
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Question #85 McGuire Nuclear Station SRO Exam l l
l Bank Question: 476 Answer: D 1 Pt(s) Unit 1 is shutdown in mode 4 operating on the ND cooling system. Given the following conditions and events:
. A fire damages the normal power supply MCC for ND-2AC )
- ND-2AC remains in the open position after the fire l e Unit I must remain on shutdown cooling for many days What action (if any)is required?
] ND-2AC is required to remain open with power removed for shutdown cooling alignment - no further action is require Enter containment and manually close ND-2AC - shift to the other train of ND for shutdown cooling - via ND-1 ! ND-2AC may be operated from the ASP using alternate power !
for MOV control if remote closure is require Emergency power may be connected to ND-2AC from the emergency D/Gs via emergency power feeder Distracter Analysis:
, Incorrect: - ND-2AC is required to be operable in mode 4 to bring the unit to cold shutdown and to isolate in the event of a LOC Plausible: - ND-2AC is maintained in the SuUT position with power removed when ND is not in service Incorrect: - ND-2AC is required to be operable in mode 4 to bring the unit to cold shutdown in event of a blackout Plausible: - shifting to the opposite train of ND would provide sufficient shutdown cooling Incorrect: - the ASP uses the normal power for valve operation Plausible: - the valve can be controlled from the ASP Correct answer t
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l For Official Use Only Ques _476 Page 93
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l Question #86 McGuire Nuclear Station SRO Exam l
l Bank Question: 477 Answer: D I
1 Pt(s) Unit 1 is cooling down for a refueling outage. Given the following i conditions and events:
. Tave = 275 *F e Pressurizer pressure = 290 psig
. Pressurizer level = 75%
e S/G levels = 35% in all four S/Gs If PORV NC-32 failed a surveillance test (the valve failed to open and could not be cycled manually) which of the following Tech Spec actions (if any) are required within one hour?
l Close the block valve for PORY NC 32 l Close and remove power from the Llock valve for PORY NC 32 Immediately suspend all actions that could lead to a water solid s pressurizer I No actions are required within one hour- restore PORY NC 32 ,
to an operable status within 7 days !
Distracter Analysis: Incorrect: - not an applicable Tech Spec requirement in mode 4 Plausible: - appropriate for a failed PORV in mode 3 that can be manually cycled Tech Spec 3.4.11 conditior A Incorrect: - not an applicable Tech Spec requirement in mode 4 Plausible: - appropriate for a failed PORV in mode 3 that cannot be ;
manually cycled - Tech Spec 3.4.11 condition B > Incorrect: - not an applicable tech spec requirement in mode 4 !
Plausible: - appropriate for mode 5,6 Tech Spec 3.4.12 condition E j Correct answer- Tech Spec 3.4.12 condition D l l r
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For Official Use Only Ques _477 Page 94 {
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Question #87 McGuire Nuclear Station SRO Exam Bank Question: 478 Answer: D 1 Pt(s) Unit 1 is responding to a reactor trip from 100% power. Given the following events and conditions:
l e 0200 - a reactor trip occurred, SRO entered E-0 l e 0210- transitioned to 'a-0.1, Reactor Trip Recovery
. 0215- entered FR-Z.2, Oftainment Flooding, on a valid orange path e 0216 - safety injection actuated e 0220 - completed FR-Z.2 l If the SRO determined that they were responding to the event in the wrong procedure, which one of the following statements is correct regarding the appropriate action to take to transition to ES-0.0, Rediagnosis? The SRO can enter to ES-0.0 at any time during the event based solely upon SRO judgement The SRO can enter ES-0.0 between 0210 and 0215 or after 0220 l The SRO can enter ES-0.0 after 0216 l The SRO can enter ES-0.0 after 0220 I
Distracter Analysis: Incorrect: - cannot enter ES-0.0 while in E-0 or while in FR- ES-0.0 is only applicable after SI actuates Plausible: - if candidate does not know restrictions and applicability of ES- Incorrect: - canrat enter ES-0.0 while in ES-0.1 becausc ES-0.0 is not applicable until after SI actuation Plausible: - if candidate does not know that ES-0.0 is applicable
- only after SI actuation l Incorrect: - cannot enter .ie in FR-Z.2 l Plausible: - if candidate notes Si has actr. rd and does net knotv l that he/she must complete FR-2.2 before ES-0.0 can be entered .
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For Official Use Only Ques _478 Page 95
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Question #88 McGuire Nuclear Station SRO Exam l
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l Bank Question: 479 Answer: A l l 1 Pt(s) Unit 1 is in the process of making a radioactive gaseous waste release from l l the waste gas decay tank in accordance with OP/0/A/6200/18. Given the following conditions:
- MRIRR = 21 CFM l e MOSRR = 40 CFM l e 1 EMF-50 trip 1 setpoint = 1.0E5 CPM e 1 EMF-50 trip 2 = 2.0E5 CPM e 1 EMF-36 is in service Time _ 0200 0215 0230 0'245 i Release rate (CFM) 22 25 41 37 EMF-50 (CPM) 1.8E5 2.2E5 2.1E5 3.2E5 I i
If the operators reset 1 EMF-50 whenever allowed by procedure, what is the earliest time (if any) that the operators are reauired to terminate the gaseous i releas Distracter Analysis: ' Correct answer - the release rate (21 CFM) > MRIRR (20 CFM)
(most restrictive instantaneous release rate) Incorrect:- execeded MRIRR at 0200 Plausible: - EMF-50 tripped WG-160 fcr the first time Incorrect: - exceeded MRIRR at 0200 Plausible: - exceeded MOSRR (maximum observed system release rate)- the operator is allowed to reset EMF-36 and restart the release 3 times before being required to terminate the release - this is the 2*'
time EMF-36 has reached trip 1 Incorrect:- exceeded MRIRR at 0200 I
Plausible: - exceeded trip 2 on EMF-50 for the 3"' time - allowed to reset this trip 3 times before terminating release For Official Use Only Ques _479 Page 96
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i Question #89 McGuire Nuclear Station SRO Exam
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Bank Question: 480 Answer: A 1 Pt(s) Unit 1 is operating in mode 3 on February 5* preparing for a reactor startu Given the following events and conditions:
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- Isotopic analysis of S/G samples at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> indicated the following primary to secondary leak rates:
- S/G A = 0.03 gpm e S/G B = 0.025 gpm e S/G C = 0.04 gpm e S/G D = 0.02 gpm
. The NC system is at normal operating temperature and pressure Which c:.e of the following statements correctly describes the actions required (if any) based on these values?
REFERENCES PRO VIDED - Tech Spec 3.4.13 No action required l Be in mode 3 by 2400 on 5 February Be in mode 3 by 0200 on 6 February Be in mode 5 by 1600 on 8 February Distracter Analysis:
Leakage from the C S/G = 0.04 gpm x 60 min /hr x 24 hr/ day = 57.6 gpd Combined leakage from all 4 S/Gs S/G A = 0.03 gpm x 60 min /hr x 24 hr.~ day = 43.2 gpd S/G B = 0.025 gpm x 60 min /hr x 24 hr/ day = 36.0 gpd S/G C = 0.04 gpm x 60 min /hr x 24 hr/ day = 57.6 gpd S/G D = 0.02 gpm x 60 hr/ min x 24 hr/ day = 28.8 cod 165.6 gpd total leakage Correct answer: each S/G < 135 gpd and (Tech Spec 3.4.13)
combined leakage through all 4 S/Gs < 389 gpd (Tech Spec 3,4.13) Incorrect: leakage < Tech Specs Plausible: action statement A -ifcandidate miscalculates leakage Incorrect: leakage < Tech Specs
- Plausible
- action statement B (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) except already in mode 3 l Incorrect: - leakage < Tech Specs i
Plausible: - action statement B - if candidate miscalculates leakage l
For Official Use Only Ques _.480 Page 97
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! l Question #90 McGuire Nuclear Station SRO Exam Bank Question: 481 Answer: A 1 Pt(s) Unit 1 is shutdown in mode 6 refueling. A Radwaste Operator brings a I liquid radiological release permit to the SRO for approval. Given the following infomiation on the permit:
- Release ID = Unit 2 CVUCDT l e RC Pumps running = 8
- RC Pumps assigned to release = 1 ;
e Total RC Pumps required = 1
- Allowable release rate = 1.61 E+05 gpm
. Recommended release rate = 6.00E+01 gpm
. EMF-44L in service = yes
- Monitor background = 4.49E+03
- Trip 1 setpoint = 8.97E+03
- Trip 2 setpoint = 1.34E+04
- Remarks: RC minimum flow interlock operable. 2 RC pumps required for all releases If no other releases are in progress, which one of the following actions are correct for approval of this release permit?
l The release may not be approved because there is an error in the number of RC pumps required ,
I The release may not be approved because the EMF-44L trip setpoints are not correct The release may not be approved because the release rate is not i correct l The release may be approved as presented i l
Distracter Analysis: Correct answer - the remarks section states 2 RC pumps are required but the number of RC pumps required is listed as 1 in the RC pump data section Incorrect: - nothing wrong with EMF-44L trip setpoints Plausible: - background < trip 1 < trip 2 . Incorrect: - allowable release rate < recommended release rate i Plausible: - if candidate does not understand this requirement ! Incorrect: - the RC pumps required is not correct j
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Question #90 - McGuire Nuclear Station SRO Exam i:
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Plausible: - if the candidate does not know how to evaluate this data t-
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Question #91 McGuire Nuclear Station SRO Exam Bank Question: 482 Answer: C-1 Pt(s) Which one of the following statements describes the requirements to allow using a pneumatic valve as an isolation boundary? Tags for the pneumatic valve are to be placed on the air supply to the valve with the air regulator petcock open. Tags should be firmly fixed around the remote operating switch on the main control board using wir Tags for the pneumatic valves are to be placed on the air supply to the valve with the air regulator petcock closed. Tags should be firmly fixed around the remote operating switch on the main control board using waxed strin Tags for the pneumatic valve are to be placed on the air supply to the valve with the air regulator petcock open. Switch labels should be firmly fixed around the remote operating switch on the main control boar Tags for the pneumatic valves are to be placed on the air supply to the valve with the air regulator petcock closed. Tags should be firmly fixed around the remote operating switch on the main control board using facing cord
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Distracter Analysis: Incorrect: - switch labels are used on main control boards for tagging remote switches - not red tags. Wire is explicitly never to be used to attach tags - NSD-500.7.2.5.1.5 due to electrical safety hazards Plausible: - the position of the components is correct - wire is strong and would hold up well Incorrect: - the air regulator petcock must be tagged open not closed Plausible: - if the candidate does not know the proper position for tagging the regulator petcock Correct answer- per NSD-500 Incorrect: - the air regulator petcock must be tagged open not closed - tags are not hung on mair control board switches l Plaushole: lacing cord is an acceptable means of hanging tags i
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Question #92 - McGuire Nuclear Station SRO Exam Bank Question: 483 Answer: C l
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l 1Pt(s) Unit 1 is in mode 6 and refueling operations are in progress. C ven the
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following conditions and events:
- The Fuel Handling Manipulator Crane Operator has lowered the mast onto the fuel assembly located at H-8 ;
- All conditions and indications on the Fuel Handling Manipulator Crane !
are satisfied for latching the fuel assembly Which one of the following statements describes the responsibility of the Fuel Handling SRO associated with latching the fuel assembly? Operates the Latch /La ..ch switch and verifies the fuel handlers are documenting satisfactory completion of each rod latch. Must give his permission to the OATC prior to latching rod ! Operates the Latch / Unlatch switch and documents satisfactory l
completion of each rod latch. Must obtain permission from the l OATC prior to latching rod Controls reactivity management and verifies maintenance i
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technicians are continuously monitoring the load cell. Must outain permission from the OATC prior to latching rod Controls reactivity management and personally monitors the l load cell and during alllatching operations. Must give his i permission to the OATC prior to latching any ro l i
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Distracter Analyals: Incorrect: - operating the latch / unlatch switch is not an SRO responsibility - must get permission from the OATC l Plausible: - this description sounds like a supervisor's l Incorrect: - does not operate the latch / unlatch switch - does not do i the actual documentation of the latching - supervises thejo !
Plausibie: - importantjob during latching rods l Correct answer Incorrect: - monitoring the load cell is thejob of the maintenance 1 technician - needs to get permission to latch rods from the OATC - !
not give it l Plausible: - the SRO coordinates reactivity management and murt i ensure that the maintenance technician monitors the load cel l t
I For Official Use Only Ques _483 Page 101
r 1 Question #93 McGuire Nuclear Station SRO Exam .
l Bank Question: 484 Answer: D l 1 Pt(s) Which one of the following exposure events meets the requirements of i
10CFR20.2202 for irmnediate notification?
l An NLO received an exposure of 23 Rem TEDE while working too long in a high radiation are l An utility maintenance worker received an exposure of a 26 Rem to the cornea of the eye while working on a heavily contaminated valv An contractor received an exposure of 79 Rad to the right l forearm while carrying an exposed radiography source that had l net been properly house An NRC Resident Inspector received an exposure of 266 Rad to the skin from a hot particle that was lodged on his left ankl l Distracter Analysis: Incorrect: - < required 25 Rem TEDE requirement Plausible: - close to limit - exceeds 24 hr notification limit Incorrect: - less than 75 Rem EDE requirement Plausible: - over TEDE limit - exceeds 24 hr notification limit Incorrect: -less than 250 Rad CEDE requirement Plausible: - over TEDE and EDE limits - exceeds 24 hr notificaiton limit Correct answer - > CEDE limit = 250 Rad for dose to extremities and skin I
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n Question #94 McGuire Nuclear Station SRO Exam l
l Bank Question: 485 Answer: C
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1 Pt(s) Unit 1 Engineering has asked the Operation Shift Manager to authorize commencing work to install a temporary modification as a compensatory action on VI-820 that will elimit - long standing problem. VI-820 is currently in a operable-but-de tus. A 10CFR50.59 evaluation has not been perfonned. Which one on une following statements is correct regarding the 10CFR50.59 requirement to determine if an Unresolved Safety Question (USQ)is involved prior to approving the temporary modification? A USQ is NOT required because VI-820 is not a safety-related component required by Tech Spec A USQ is NOT required because VI-820 is act inoperabl A USQ is required because VI-820 is a safety-significant component in an operable-but-degraded statu A USQ is required because VI-820 is a safety-related component required by Tech Specs in an operable-but-degraded statu Distracter Analysis: Incorrect: - USQ is required Plausible: - if the candidate does not know the 10CFR50.59 requirements - VI-820 is not required by Tech Specs Incorrect: - USQ is required for inoperable-but-degraded SSCs Plansible: -it is not inoperable Correct answer- NSD-20 Incorrect: - VI-820 is not required by Tech Specs Plausible: - a USQ is truly required - right answer, wrong reason l
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Question #95 McGuire Nuclear Station SRO Exam Bank Question: 486 Answer: D 1 Pt(s) Unit 1 is operating at 50% power widi one of the 120 VAC vital instmment l buses, EKVC, deenergized for maintenance. Control switches are in the 1-2 positio ;
Given the following conditions: l 1) Bistables on instmments powered off EKVC have not been tripped 2) The reactor trip breakers have failed in the closed position due to improper maintenance and retesting (unknown to the operators)
3) NCP Loop A Channe '. flow instrument fails low Which one of the following sequences accurately describes the plant response? The reactor protection system will NOT generate a trip signal with the loss of CH I loop flow instrument ' The reactor protection system will NOT generate a trip signal because the plant is below P-8 The reactor protection system will geverate a trip signal under these conditions and an ATWS wou;d occu ! The reactor protection system will generate a trip signal under these conditions and a failure of RPS would occu Distractar Analysis: Incorrect: - a trip signal will be generated Plausible: - if the candidatu does not know the trip logic for low flow
, Incorrect: - a trip will be generated - the plant is > P8
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Plausible: - if the candidate does not know the P8 setpoint l Incorrect: - no ATWS will occur - there is no anticipated transient in progress l Plausible: - the reactor should have tripped Correct answer- per OMP 4-3.
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l Question #96 McGuire Nuclear Station SRO Exam Bank Question: 487 Answer: A
1 Pt(s) What will be the symptoms and operability status ofICCM due to a loss of .
power from EKVB? l
l ICC MONITOR TRN "B" TROUBLE annunciator will illuminat I
"B" Train ofICCM is inoperabl ' MALFUNCTION ALARM ctatus for all three leyel sensors is displayed l
"B" Train ofICCM is operabl DATA LINK FAILURE is displayed ICCM is operabl LOSS OF SUBCOOLING annunciator is illuminated SUBCOOLING MARGIN ALERT annunciator is illuminated ICCM is inoperabl REFERENCES PRO VIDED - Tech Spec 3. Distracter Analysis: Correct answer - loss of EKVB causes the trouble alarm and B train to be inoperable Incorrect: - will not get alarm status indication on level sensors, the B train ofICCM is inoperable Plausible: - occurs if power is lost to the sensor alone Incorrect: - Will not get data failure alarm Plausible: - loss of power could conceivably cause a data failure (link between local cabinet and remote indication has failed for > 10 seconds) - if candidate does not know what causes a data link failure, ICCM is operable - but B train is inoperable Incorrect: - train A feeds these alarms, not train B, ICCM remains operable - but B train ofICCM is inoperable Plausible: - would get alarms iflost power to EVKA For Official Use Only Ques _.487 Page 105
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Question #97 McGuire Nuclear Station SRO Exam Bank Question: 488 Answer: A 1Pt(s) Unit 1 is operating in mode 1 at 100% power. Performance is in the process of conducting a valve stroke timing test on INS 32A when a loss of the l 1EMXA bus occurs. INS 32A was in the full open position when 1EMXA i was deenergize !
The following partial load list was extracted from the 1EMXA one line !
diagram :
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INS-20A (NS PUMP A SUCT FROM RWST BLK VLV)
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INS-29A (NS PUMP A DISCH CONT ISOL VLV)
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INS-32A (NS PUMP A DISCH CONT ISOL VLV)
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1NS-43A (ND PUMP A DISCH TO NS NOZZLES CONT ISOL VLV)
INI-185A (CONT SUMP LINE 1 A ISOL VLV)
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IRN-134A (NS HX A SUPPLY ISOL VLV) I e
IRN-137A (NS HX A CONTROL VLV) ;
e CONTAINMENT SPRAY PUMP AIR HDLG UNIT 1 A I e RHR PUMP 1 A AIR HDLG UNIT Five minutes later, a large break LOCA into containment occur Which one of the following sequences describes the status of the Containment Spray System Train "A" Containment Spray Pump will not star Train "A" Containment Spray Pump will start and run without a suction flow pat Train "A" Containment Spray Pump will start and supply spray flow into containmen Train "A" Containment Spray Pump will start and run dead-heade Distracter Analysis: Correct Answer: - the discharge valves have power removed and the pumps will not stan due to relaying Incorrect: - A NS pump will not start l
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, Question #97 McGuire Nuclear Station SRO Exam ,
Plausible: - received a valid start signal from the LOCA and discharge valve is open - suction valve has power removed Incorrect: - A NS pump will not start Plausible: - received a valid start signal from the LOCA and >
discharge valve is open - suction valve is normally open Incorrect: A NS pump will not start Plausible: - power is removed from the discharge valve - candidate
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Question #98 McGuire Nuclear Station SRO Exam Bank Question: 489 Answer: C 1 Pt(s) Unit 2 was operating at 100% power following a refueling shutdow Ifit is discovered that a spent fuel element was incorrectly stored in the wrong region of the spent fuel pool, which one of the following statements describes the cerrect action (s) required by Tech Specs? Initiate action to move the noncomplying fuel assembly to the correct location within one hou Initiate actions to move the noncomplying fuel assembly to the correct location and to shutdown Unit 2 within I hou Immediately initiate action to move the noncomplying fuel assembly to the correct locatio I Immediately initiate actions to move the noncomplying fuel assembly to the correct location and to shutdown Unit 2 within I hou Distracter Analysis:
, Incorrect: - must initiate action immediately Plausible: - I hour LCOs for important Tech Specs are common and the fuel assembly has clearly decayed more than 16 days Incorrect: - must initiate action immediately - Tech Spec 3.0.3 does not apply Plausible: - I hour LCOs for important Tech Specs are conunon and the fuel assembly has clearly decayed more than 16 days; Tech ,
Spec 3.0.3 requires initiating action to shutdown the unit within 1 l'
hour Correct answer: Incorrect: - Tech Spec 3.0.3 does not apply Plausible: - If the candidate thinks that Tech Spec 3.0.3 applies l
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Question #99 McGuire Nuclear Station SRO Exam I
Bank Question: 490 Answer: D 1 Pt(s) Unit 1 is in mode 5 conducting a heatup for plant startup following a refueling outage. All systems are in a nonnal alignment for this operating ,
mode. Given the following conditionc:
. Heatup is in progress at 5"F/hr with current temperature at 180 F e No Tech Spec LCOs are app.licable Which of the following statements describes the assessment for shutdown risk?
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REFERENCES PROVIDED MSD 403 SECTIONS 1. APPENDIX B, DEFENSE IN DEPTH ASSESSMENTSHEET '
2. APPENDIX C, SYSTEM EQUIPMENT A VAILABILITYSHEET Shutdown risk managemtat does not apply because the plant is heating u Unit 1 is in code " Red" for the safety function for decay heat remova Unit 1 is in code "Y.iiow" for the safety function for decay heat remova The unit is in code " Green" for the safety function for decay heat remova Distracter Analysis: Incorrect: - shutdowa risk management and defense in depth applies until the unit enters mode 4 Plausible: - if the candidate does not know when shutdown risk tr&,. agement stops being applicable during a startup Inerk c:.t: -in code Green Pladsibie: - if the candidate does not understand how to determine the shutdown risk color codes Incorrect:- in code Green Plausible: -if the candidate does not understand how to determine the shutdown risk color codes l Correct answer For Official Use Only Ques _490 Page 109
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Question 0100 McGuire Nuclear Station SRO Exam Bank Question: 491 Answer: B 1 Pt(s) Unit i experienced a LOCA with a breach ofcontainment at 0200. The OSM assumed the duties of the Emergency Coordinator and declared a general emergency at 0210. The initial recommended protective actions at 0215 were as follows-
- Shelter zones E, F, G, H, I, J, K, P, Q, S e Evacuate zones L, B, M, C, N, A, D,0, R At 0225, the initial dose projection information was presented to the OS Given the following conditions at 0230:
e Wind direction = 210*
. Contamment radiation levels e IEMF51 A = 950 R/hr e 1 EMF 51B = 955 R/hr Which one of the following protective action recommendations are correct in accordance with RP/0/A/5700/04 on the 0230 ENS upda+.e message?
REFERENCES PRO VIDED - RP/0/5700/004
. Change the protective action recommendation in zones O and R l from evacuate to shelter Extend the evacuation to zones G, H, I, J, K, P Extend the evacuation to zones E, Q, G Cancel protective action recommendations based upon dose ,
projection information !
Distracter Analysis: Incorrect: - wrong zones for evacuation Plausib",e: - if the candidate enters the wrong table - the one that has l wind speed > Smph i Correct answer i
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t Question #100 McGuire Nuclear Station SRO Exam Incorrect: - wrong zones for evacuation Plausible: - if the candidate enters the PAR table with a reciprocal wind direction
. Incorrect: - Although dose projections are < PAGs, the PARS would not be cancelled Plausible: - if the candidate thinks that dose projection information is sufficient grounds to terminate PARS
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