IR 05000334/1998300
| ML20248L051 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/03/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20248L042 | List: |
| References | |
| 50-344-98-300OL, NUDOCS 9806100390 | |
| Download: ML20248L051 (105) | |
Text
{{#Wiki_filter:e I i'
U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No.: 50-334 , Report No.: - 98-300 License No.: DPR-66 - Licensee: Duquesne Light Company Facility: Beaver Valley Unit 1 Nuclear Power Plant Location: Shippingport, Pennsylvania Dates: April 20 24 and May 18,1998 Chief Examiner: T. Kenny, Senior Operations Engineer / Examiner Examiners: J. D' Antonio, Operations Engineer / Examiner T. Fish, Operations Engineer / Examiner Approved By: Richard J. Conte, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety
9806100390 980603 POR ADOCK 05000334 V pg &_________:-__________ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _. - _ __.
J
h i-t EXECUTIVE SUMMARY Beaver Valley Unit 1 Nuclear Powar Plant inspection Report No. 50-334/98-300 Operations Three Unit 1 senior reactor operator instant (SROI) candidates passed all portions of the initial license examination.
Generic strengths were noted during the Unit 1 examination in the area of crew communications, control board awareness, and crew briefings during the simulator portion . of the operating examination. The NRC examiners observed communications to be direct, l succinct, and that all crew members were kept wellinformed. Crew briefings were routinely held during those instances in which time permitted. The written examination was developed at tha appropriate SRO knowledge level, as werc the job performance measures and follow-up questions. Several JPMs, in lieu of questions, were appropriately developed to test the knowledge level of the applicants in the administrative area of the examination.
A;l th aa credidates met or exceeded Regulatory Guide (REGUIDE) 1.8, Rev. 2 for a Senior Reactor Operator instant License.
Some of the site documents did not accurately delineate the current training requirements, however, the licensee was in the process of correcting the errors. This was deemed a minor violation, il t _ _ _ -.. _ _. - - - _ - - - _ _ _ - - - _. ...
e Report Details I. Ooerations
Operator Training and Qualificatiora 05.1 Senior Reactor Ooerator Initial Examinations a.
Scope The NRC examiners reviewed on-site and in-office the examination as prepared by Beaver Valley Power Station (BVPS) and contractor personnel in accordance with the guidelines in interim Revision 8, of NUREG-1021," Examiner Standards," and Revision 1 of NUREG-1122," Knowledge and Abilities Catalog for Nuclear Power
- -
Plant Operators: Pressurized Water Reactors." The NRC examiners administered I-initial operating licensing portion of the examination to three Unit 1 senior reactor operstor instant (SROI) candidates. The facility's training organization administered the written examination.
i ! b.
Observations and Findinas The results of SRO examination for Unit 1 are summarized below: SRO Pass / Fail Written ' 3/0 Operatirig 3/O Overall 3/0 - Overall the entire examination was well written and validated by the licensee prior to the NRC reviewing it. This was evidenced by the few changes that were required by the NRC after their review.
The written portion, job performance measures (JPMs) and simulator scenarios were developed by Beaver Valley Power Station (BVPS) and their contractors in accordance with NUREG-1021. The examination development team was comprised of BVPS training and operation's representatives and a contractor. Allindividuals involved signed a security agreement once the development of the examination commenced. BVPS personnel validated the operation portions of the examination prior to their submitting it to the NRC. The NRC subsequently reviewed and observed the validation of all portions of the proposed examination. During the examination preparation week, the NRC examiners noted that the facility staff had performed good validation of the new simulatos scenarios and JPM's, because little or no changes were required after the demonstrations.
. The written portion of the examination was administered on May 18,1998,and consisted of 100 multiple choice questions. There were minor comments by the NalC concerning the adequacy of four questions on the written examination, however, the licensee promptly corrected them. The results of the written portion of the examination showed that question 51, regarding de bus ground faults and i . - .I
question 85, reaction W rne reactor coolant system hot and cold leg temperatures, during the first few minutes following a reactor trip coincident with a loss of offsite power, were missed by all of the applicants. Discussions with the licensee showed that they were aware of the problem and were taking appropriate actions to: (1) remediate the candidates on missed questions and (2) perform an analysis to determine if training outlines should be enhanced.
The operating portion of the examination was conducted from April 20-23,1998, and consisted of 'three simulator scenarios and ten JPMs. - All JPMs were followed up with two system-related questions. All candidates were also examined using JPMs and/or questions to evaluate the administrative requirement portion of the examination.
Simulator and JPM performance by the candidates was very good.
Com.munications was also good, including the use of repeat backs. The examiners noted that crew briefings were routinely performed by the SROs. Control board awareness by all of the candidates was evident throughout each of the three scenarios.
For the administrative segment of the operating portion of the examination, administrative job performance measures (JPMs) were used in a number of instances in lieu of administrative topic questions. The examiners determined that candidate performance was good as evaluated in this area.
BVPS also exhibited good judgement in asking Region I for a one-month delay in administering the written portion of the examination, because a quality product could not be produced in time to be administered with the operation portion in April 1998, c.
Conclusions The candidates performed well on both the written and operating portions of the examination, and thus were issued licenses. The candidates were well prepared for the examination, indicating that the facility thoroughly evaluated the knowledge and ability of each candidate in an effort to determine their readiness to sit for an initial NRC, SROI examination. Crew communications, control board awareness, and crew briefings were very good. The training department continued to rio an excellent job in adhering to the examiner standards and in developing the examination materials needed to administer the examinations.
l , _ -. _ - - - _ -. - - -. _ _ _ _ _. _ - _ _ _ _ _ - - _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ -
~3 05.2 Angk: ant Trainina and Experience a.. Scope i Regulatory Guide 1.8 (REGUIDE), Rev. 2 requires certain requirements and cer+.ain obligations in the area of training and experience be satisfied by a license candidate - prior to taking the examination for a hot Senior Reactor Operators license. The inspectors reviewed the three candidates' training records and NRC records to verify compliance with these requirements.
b.
Observations and Findinas - REGUIDE 1.8 requires that: l.
. .
e Each candidate, for a senior license, have a high school diploma or equivalent. The inspectors verified that all candidates met or exceeded the requirement.
' e Each candidate, for a senior license, have four years of responsible power plant l experience. The inspectors verified that all candidates met or exceeded the requirement.
e Each candidate, for a senior license, serve three months as an extra person on shift in training for that position. Three months is the equivalent of 520 hours for a 40 hour work week. The inspectors verified that all candidates met or excee%d the requirement.
o Each candidate, for a hot license, should manipulate controls of the facility during a minimum of five reactivity changes. The inspectors verified that all candidates met or exceeded the requirement.
The licensee requires that each candidate maintain a training note book to document the above requirements. The instructions and requirements for maintaining the forms are delineated in the Training Administrative Manual (TAM). The inspectors verified that the inspected portion M the TAM reflected the requirements of Regulatory Guide 1.8, Rev. 2.
Also, the inspectors reviewed the Technical Specifications (TS), The Quality . Assurance Manual (QAM) and The FSAR to determine if these documents
delineated the proper references to the training requirernents. The insp9ctor found D inconsistencies within the documents. The TS referenced, "section 5.5 of ANSI l lN18.1-1971 and CFR Part 55." The QAM referenced, "10 CFR 50 and 10 CFR 55 - and be in agreement with ANSI N18.1-1971." The FSAR referenced, "REGUIDE . 1.8, Rev.1-R, September 1975" and had been updated since the original version.
The TAM referenced, "REGUIDE 1.8."
L ,:; i u_______i._ __.i_. _ _ _ _ _.. -
The licensee was conducting their training of perspective operators in accordance with REGUIDE 1.8, Rev. 2. This is delincated in the TAM. Se licensee issued Condition Report (CR) 980734, on April 9,1998, that descri es the inconsistency.
After discussions and a review of the CR the inspector determined that the licensee was taking corrective actions, and were expected to resolve the issue by June 12,1998, with the exception of the TS change, which may take longer. This failure constitutes a violation of minor significance and is not subject to formal enforcement action.
c.
Conclusions Current operator license training is being conducted in accordance with REGUIDE 1.8, however, site documents were not consistent with the proper reference to the current NRC required training document, REGUIDE 1.8. The licensee was in the process of changing the documents.
i E8 Review of the FSAR.
While performing the preexamination activities discussed in this report, the -] inspectors reviewed applicable portions of the FSAR, that related to the selected examination questions or topic areas. One discrepancy discussed in the previous paragraphs was identified and was being corrected by the licensee.
V. Manaaement Meetinas X1 Exit Meeting Summary On April 23 and May 20, the NRC examiners discussed their observations regarding the examination with Beaver Valley Unit 1 operations and training management representatives. The examiners discussed candidate performance, including communications and briefings among themselves, both were very good. The licensee did not see the need to make comments following the administration of the written examination.
The examiners also expressed their appreciation for the cooperation and assistance that was provided during both the preparation and examination week by licensed operator training personnel and operations personnel. The following participated in the exit meetings.
_ _ _ _ _ _ _ _ _____ _____._________________ _ _
(; l l
PARTIAL LIST OF PERSONS CONTACTED SfAVER VALLEY l.
K. Beatty, General Manager, Nuclear Support R. Brooks, Sr. Nuclear Operations instructor W. Lindsey, Director, Operator Training
S. C-Jain, Vice President, Nuclear Services l-B. Tuite, General Manager, Nuclear Operations L. Shad, Simulator Supervisor NEG T. Kenny, Senior Operations Engineer, Chief Examiner T. Fish, Operations Engineer ~ J. D' Antonio, Operations Engineer Attachments: 1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key 2. Simulation Facility Report _ - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _
f I Attachment 1 BV-1 SRO WRITTEN EXAMINATION W/ ANSWER KEY
_ _. _ _ _ _ _ _ _ _ _ _ -. _ _ _ _ _
i !- Quest 6om Topic: l Temperature trending during cooldown A cooldown is in progress. The milestones listed on Figure 1 of 10M-51.4C,(see attached) were reached at the following times: (1) 0800
(2) 0833- + (3) 0857
.(4) 0917
What action, if any, is required to be taken to comply with Technical Specifications? a.
RCS cooldown is acceptable to this point. RCS cooldown rate will not be exceeded if Figure I time limits are complied with from this point on, b. RCS cooldown is acceptable to this point. RCS cooldown rate may be exceeded even if Figure 1 times are complied with from this point on.
i c.
RCS cooldown exceeded Technical Specifications. RCS temperature must remain constant until , 0927.
' d. RCS cooldown exceeded Technical Specifications. Cooldown rate must be restored to within. Technical Specification limits by 0947.
A.s: Ia l Exam Level: lS l Cognitive Level: l Application l Explanatio nef Answer KA: l2.1.2 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Conduct of Operations Statement: Knowledge of operator responsibilities during all modes of plant operation.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Station Shutdown - 10M-51.4.C IV.A.13.b 10-11 1ss 4 Rev Cooldown From MODE 3 to
, MODE 4 Beaver Valley - Unit 1 3.4.9.2 3/4 4-22,4-27 Amend Technical Specifications No.179 OM 6,7 & 10 Operational LP-SQS-RX IV.D.4
6 Lecture Qrestion Source l New l Question Modification Method l QIestion Source Comments: l M terial Required for Figure 1 of OM-51.4.C - Blowup curve to max 81/2 x 11 Examination: Pagt,1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _
I l Questio2 Tepic: l Core Sifety Limit Curve eval At 20% power, the ruaximum allowable T,,,is limited by the Reactor Core Safety Limit. The basis for l limiting T,,, under these conditions ensures that:
a.
DNBR remains greater than or equal to the safety analysis DNBR limit and the average enthalpy at ' the vessel exit will not exceed saturation.
b. DNBR remaint, greater than or equal to the safety analysis DNBR limit and the highest enthalpy anywhere in the core will not equal saturation.
DNBR remains less than the safety analysis DNBR limit and the average entha;py at the vessel exit c.
l will not exceed saturation.
! d. DNBR remains less than the safety analysis DNBR limit and the highest enthalpy anywhere in the ! core will not exceed saturation.
Ais: la l Eram Level: lS l Cognitive Level: l Memory l Explanatio e cf Answer KA: l 2.1.10 l RO Value: l2.7 l SRO Value: l3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: ,
KA Conduct of Operations i l St:tement: Knowledge of conditions and limitations in the facility license.
Rt.ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Protection System LP-SQS-l.1 II.C.3
6 4.c i l Q estion Source l New l Question Modification Method l Question Source Comments: l M:terial Required for TS Figure 2.1.1 Ex:mination: , . Page 2
_- Questio] Topic: l TS 3.0.5 During power operation the Diesel Generator #1 is declared inoperable. Subsequently the 1B Quench Spray pump is determined to be inoperable.
Assuming all required surveillance are completed satisfactorily, what is the required Technical Specification action? a.
Restore both the 1B Quench Spray and Diesel Generator #1 operable status within 72 hours or be in Hot Standby within the following 6 hours.
j b. Restore either the IB Quench Spray pump or Diesel Generator #1 to operable status within 24 hours or be in Hot Standby within the following 6 hours.
c.
Restore the IB Quench Spray pump to operable status within one hour or be in Hot Standby within the following 6 hours.
d. Restore the 1B Quench Spray pump or Diesel Generator # 1 to operable status within 2 hours or be in Hot Standby within the following 6 hours.
Ars: ld l Exam Level: lS l Cognitive Level: l Application l Explanatio ufAnswer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Conduct of Operations Statement: Ability to apply technical specifications for a system.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Technical Specifications TS 3.0.5,3.6.2.1, 3.8.1.1 Containment LP-SQS-13.01
12 Depressurization Systems Qrestion Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l M:terial Required for Technical Specifications Examination: Page 3
Question Tcpic: l FFD requirements
What are the fitness for-duty requirements, with respect to alcohol, for an unscheduled RO who has been ] l called out? l
a. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be required to pass a breath analysis test.
I < b. The RO may report to work if he/she has consumed alcohol within the past FIVE hours, but will be subject to a breath analysis test only if deemed necessary by the NSS.
The RO must report to work even if he/she has consumed alcohol within the past FIVE hours, but c.
will be required to pass a breath analysis test.
d. The RO shall not report to work if he/she has consumed alcohol within the past FIVE hours.
Ars: la l Exam Level: lS l Cognitive Level: l Application l Explanatio c cf Answer KA: l 2.1.13 l RO Value: l2.0 l SRO Value: l2.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Conduct of Operations St:tement: Knowledge of facility requirements for controlling vital / controlled access.
R;ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fitness-For-Duty Program 1/2 NPDAP 2.14 IV.2 & 3
0 For Duquesne Light Employees i Conduct Of Operations 1/2LP-SQS-48.1 Vill.
10 3,39 ) Q:estion Source l New l Question Modification Method l Qdestion Source Comments: l M;terial Required for 1/2 NPDAP 2.14 Examination: l l l Page 4
Question Topic: l TS SDM & Emerg'ncy Boration Given the following conditions: RCS T,,, - 355 F a RCS pressure - 400 psig a RCS boron concentration - 1000 ppm
Shutdown margin is below Technical Specifications allowable value o Emergency Boration is initiated at 30 gpm boric acid
A 70 ppm RCS boron concentration change is required to restore the
required SDM Of times listed below, which is the MINIMUM emergency boration time that will ensure the required boric acid has been added? a.
15 minutes b.
17 minutes . _.
c. 21 minutes d. 24 minutes Ans: lc l Exam Level: lS l Cognitive Level: l Application l Explanatio A 70 ppm change at Normal Operating Conditions would require 500 gallons boric acid. The correction factor of a cf Answer 1.18 multiplied by 500 would result in 590 gallons of boric acid. 590/30gpm = 19 minutes 40 seconds.
KA: l 2.1.25 l RO Value: l2.8 l SRO Value: l3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution ) ) Title: KA Conduct of Operations Statement: Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain - performance data.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj ] i Emergency Boration IOM-7.4.S IV.A S2 1ss 4 Rev i Beaver Valley Unit 1 - 3.1.1.1 3/4 1 1 Amend Technical Specifications No. 91 CVCS LP-SQS-7.1 IV.E
12 Question Source l New l Question Modification Method l l ' Question Source Comments: l Material Required for IOM-7.5 Figures 7-7,7-8 & Table 7-1.
Examination: l
l Page 5 t i
---_- i
Question Topic: l Parmission for deviation from NSA.
In addition to normal requirements for manipulating components, which of the following describes who is required to approve placing component in other than its Normal System Alignment (NSA)? a.
Two SROs are required to approve the manipulation.
b. Specific permission is required from the NSS.
c.
Either the NSS or ANSS has to approve the manipulation.
d. The General Manager, Nuclear Operations.
, Ans: lc l Exam level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 2.1.29 l RO Value: l3A l SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Conduct of Operations Statement: Knowledge of how to conduct and verify valve lineups.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Question Source l Facility Exam Bank l Question Modification Method l . Qzestion Source Comments: l M;terial Required for Ex:mination: ) l I-Page 6
Questio2 Tcpic: l Procedure change rules for type of procedure While at 100% power, an OMCN is to be written to change IOM-7.4.L " Blender Boration Operation." This change adds a step that directs placing ONE bank of Pressurizer heaters in MANUAL prior to initiating a boration. An Operations Unit Non-Intent Reviewer has determined that this does NOT change the intent of the procedure.
The on the spot change: can be approved by TWO members of management, ONE holding a valid SRO license on Unit 1.
a.
b. becomes effective 14 calendar days following review by the OSC and approval of the GMNO.
c.
cannot be made because use of the procedure is not expected in the next 30 days.
d. cannot be rnade because this is a safety related procedure.
ATs: Ia l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio acf Answer - ' KA: l2.2.6 l RO Value: l2.3 l SRO Value: l3.3 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Equipment Control Statement: Knowledge of the process for making changes in procedures as described in the safety analysis report.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Control Of Operating 1/2OM-48.2.B C. I.a B 10 iss 4 Rev Procedures
Conduct Of Operations I/2LP-SQS-48.1 1.H.2
10 8, 9 QIestion Source l New l Question Modification Method l QIestion Source Comments: l M:terial Required for Examination: , Page 7
, uestion Tcpic: l Omissions in OSTs Q A partial OST is to be performed. Which of the following is an acceptable method of blocking the portions of the OST that are NOT applicable? The ANSS blocks the non-applicable portions.
a.
b. The STA blocks the non-applicable portions and the RO verifies they are correct.
c. The system engineer blocks the non-applicable portions and the ANSS verifies they are correct.
d. The PO blocks the non-applicable portions and the RO verifies they are correct.
Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 2.2.12 l RO Value: l3.0 l SRO Value: l3.4 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Equipment Control Statement: Knowledge of surveillance procedures.
Reference l Reference Number Reference Section Page Number (s) Revision Learn.
Obj Adherence and 1/20M-48.2.C VI.B.17
iss 3 Rev Familiarization to operating
Procedures Conduct Of Operations 1/2LP-SQS-48.1
i0 Q:estion Source l New l Question Modification Method l Q estion Source Comments: l M:terial Required for Examination: Page 8
Question Tepic: l Caution Tags Use of a Caution Tag is PROHIBITED for which of the following conditions? a.
Special additional manual actions are required to operate the tagged component.
b. Operation of the tagged component will be affected because a portion of the system is not in NSA.
c. As a temporary replacement for a component label that has fallen off.
d. As a warning that operation of the component will cause erratic indication.
Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Esplanatio c of Answer KA: l 2.2.13 l RO Value: l3.6 l SRO Value: l 3.8 l Section: l PWG l ROGroup: l 1 l SRO Group: l1 System / Evolution Title: KA Equipment Control Statement: Knowledge of tagging and clearance procedures.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Use of Caution Tags 1/20M-48.3.L IV.A 1-2,3 iss 4 Rev
Conduct Of Operations 1/2LP-SQS-48.1 V.P
10
Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Ex:mination: Page 9
Question Tcpic: l SRO control Which of the following describes a responsibility of the Refueling SRO during fuel movement? The Refueling SRO will: a.
initial the Fuel Assembly Handling Deviation Report with NSS concurrence.
b. be located on the manipulator crane structure during most fuel handling activities.
c.
maintain the DLC Master Copy of the Fuel Handling data Sheets.
d. continuously monitor source range count level.
Ars: lb l Exam Level: lS l Cognitive Level: l Memory l Esplanatio c ef Aaswer KA: l2.2.31 l RO Value: l1.6 l SRO Value: l3.8 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Equipment Control Statement: Knowledge of SRO fuel handling responsibilities.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj R: fueling Administrative Book 1 -lRP-12R-1.1 II.DA.b.15)
Iss 0 Rev Section
Fuel Handling Operations LP-SQS-6.13 Ill.B
5 2.b Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Ex::mination: . l Page 10 l
. _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Question Topic: l High Radiation Definition l Technical Specifications requires radiction areas to be isolated by locked doors if the radiation levels are greater than: a.
100 mrem /hr b. 500 mrem /hr c.
1000 mrem /hr i l . ' i d. 5000 mrem /hr Ans: lc l Eman 12 vel: lS l Cognitive Level: l Memory l Explanatio eef Answer KA: l2.3.1 l RO Value: l2.6 l SRO Value: l3.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Radiation Control Statement: Knowledge of 10 CFR: 20 and related facility radiation control requirements.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Technical Specifications 6.12 6-23 188 . Q'estion Source l New l Question Modification Method l Question Source Comments: l Material Required for Verify Section 6 of the Technical Specification is not included in materials Ex:mination: l l Page11
Question Tople: l SRO action for gas release Given the following conditions: Reactor power-100%
Discharge of Waste Gas Decay Tank [lGW-TK-1 A] is planned for 1000 on 4/22/98
The RWDA-G had been approved on 1500 on 4/20/98
The meteorological information indicates Stability Class A for atmospheric conditions
The status of the Gaseous Efiluent Monitors is as follows: .* - Gaseous Waste / Process vent [RM-GW-108A] noble gas channel inoperable - Gaseous Waste / Process vent [RM-GW-108B] noble gas channel inoperable Preparation for the release was then delayed until 2300 on 4/23/98 Which of the following describes the status at the new planned time for release (2300 on 4/23/98), assuming. equipment status and other conditions do NOT change? a.
The release can be initiated without restriction.
. b. The release can be initiated only if sampling of the release stream is analyzed at least one per every FOUR hours.
c.
The release cannot be made because the 72-hour effective time limit for the RWDA-G has elapsed.
, d. The release cannot be made because the Stability Class for release is unacceptable.
Ars: lc l Esam Level: lS l Cognitive level: l Comprehension l Explanatio a cf Answer KA: l2.3.6 l RO Value: l2.1 l SRO Value: l3.1 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: - KA Radiation Control Statement: Knowledge of the requirements for reviewing and approving release permits.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Decay Tank Discharge IOM-19.4.E step 7 NOTE E3 iss 3 Rev
Gaseous Waste Disposal LP-SQS-19.1 11.0, ODCM 3.3.3.10 17 13
9.e System Question Source lNew l Question Modification Method l Question Source Comments: l Material Required for IOM-19.4.E Examination: ,. Page 12 c _
.- - - _ _ _ _. _ _ _ -. - - - _ _ _ _ - _ _ _ - - _ _ _ _ _ _ - _ - _ - _ - _ - _ _ _ - _ - _ . _ -_ _ _ _ - _. _ e Question Topic: l Time to Core Boiling Given the following conditions: The reactor has been shutdown for 2 days.
- RCS temperature is 150 F.
- L RCS pressure is atmospheric.
. PZR is a normal level for shutdov;n cooling.
-* Assume RHR is lost. Which of the following describes the time available until core boiling occurs? ( Using the attached references, AOP 1.10.1 attachments 1,2,3, & 4) a.
Less than 10 minutes.
l b. Il'to 20 minutes.
c.
21 to 30 minutes, d. 31 to 40 minutes.
Ams: ld l Exam Level: lS l Cognitive Level: l Application l Explanatio e of Answer KA: l2.4.9 l RO Value: l3.3 l SRO Value: l 3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Emergency Procedures / Plan Statement: Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR mitigation strategies).
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj ResidualHeat Removal AOP 1.10.1 11, Attachment i Iss 3 A System Loss Rev 5 Residual Heat Removal LP-SQS-10.1
9,10 System Question Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: l Material Required for AOP 1.10.1 Attachments I,2 3 & 4.
Examination:
l - .
i Page 13 i s
f C
- ion Topic: l Implementation of Orange Path l
Given the following conditions: An unisolable steam line break has occurred on SG "B"
SG "A" and "C" levels were overfed.
+ A reactor trip and SI occur.
- Pressurizer pressure is 1180 psig
Pressurizer level is 12%
T.v,is 400 F and slowly dropping
! E-0 " Reactor Trip Or Safety Injection", step 9 is being performed.
I
The STA informs' crew that B loop T u is 283 F and slowly dropping.
- What is the EOP flowpath that will be followed given the above conditions?
Immediately transition to FR-P.1 " Response To imminent Pressurized Thermal Shock Condition", a.
b.
Perform actions of E-0 through diagnosis of steamline break, then transition to E-2 " Faulted Steam Generator Isolation" . Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.1 " Response to c.
Imminent Pressurized Thermal Shock Condition" d.
Perform actions of E-0 through diagnosis of steamline break, then transition to FR-P.2 " Response to Anticipated Pressurized Thermal Shock Condition" Ass: lc l Exam tevel: IS l Cognitive Level: ! Application l Explanation ofAnswer KA: l 2.4.14 l RO Value: l3.0 l SRO Value: l 3.9 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Emergency Procedures / Plan Statement: Knowledge of general guidelines for EOP flowchart use.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Ob] Subcriticality - Status Tree F-0.4 ORANGE PATH IssIB Rev1 Reactor Trip Or Safety IOM-53B.4.E-0 1. Ist paragraph
IssIB Injection Background Rev 5 EOP Introduction LP-SQS-53.1 B.1
I Question Source l New l Question Modification Method l ' Question Source Comments: l Material Required for F-0.4 and Att 5-D Examination: Page 14
Questici Topic: l EOP Usags During Critical Safety Function Status Tree monitoring it was determined that TWO functions had Orange P:ths. One of the Orange paths is FR-H.1, Response to Loss of Secondary Heat Sink.
Which Critical Safety Function, also Orange, would take precedence over FR-H.l? a. FR-C.1, Response to Inadequate Core Cooling b. FR-Z.1, Response to High Containment Pressure c.
FR-P.1, Response to Imminent Pressurized Thermal Shock Condition d. FR-1.1, Response to High Pressurizer Level A s: la l Exam Level: lS l Cognitive Level: l Comprehension l -_ Explanatio a cf Answer KA: l 2.4.16 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Emergency Procedures / Plan Stitement: Knowledge of EOP implementation hierarchy and coordination with other support procedures.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj EOP Executive Volume - 1/20M-53B.2 111.B
IssIB User's Guide Rev 3 EOP Introduction LP-SQS-53.1
Question Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l M;terial Required for Examination: Page 15 _______________-___-____ _ - _ -
Questio] Te pic: l Functionil Recovery Procedure usage During a loss of all Emergency 4KV AC Power, When are Functional Restoration Procedures implemented? a.
Immediately upon electrical power restoration to I AE or IDF.
b.
Immediately upon exiting ECA-0.0 " Loss of all 4KV AC Emergency Power " c.
When directed by ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0.2. " Loss of all AC Power Recovery With SI Required" t 'd.
When ECA-0.1 " Loss of all Power Recovery Without SI Required" or ECA-0 2. " Loss of all AC l Power Recovery With SI Required" is completed.
l l A:s: lc l Exam Level: lS l Cognitive Level: l Memory l . Explanatio a cf Answer ' KA: l 2.4.16 l RO Value: l3.0 l SRO Value: l4.0 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Emergency Procedures / Plan St:tement: Knowledge of EOP implementation hierarchy and coordination with other support procedures.
Reference Reference Number Reference Section Page Number (s) Itevision Learn.
Obj EOP Executive Volume - 1/20M-538.2 VI.D
issIB ) User's Guide Ret 3 EOP Introduction LP-SQS-53.1 IV.C.4
1
Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l ~ Mrterial Required for ) Ezrmination: i . Page 16 )
. - _ - _ - _ _ _ _ - - - - _ _ _ _ _ _ _ _ _ ___ - ___ _ _ __ _ _ __ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
Questio2 Topic: l Fire Brigade Responsibility s During a plant fire, who is responsible for coordinating fire-fighting activities with the offsite fire department l chiefs? l i a.
The ANSS when acting as the Fire Brigade Chief.
b. The ANSS when acting as the Fire Brigade Captain.
h c. The affected Unit's NSS.
' d. The Nuclear Operator when he/she is acting as the Fire Brigade Captain.
Ans: la l Exam Level: lS l Cognitive 12 vel: l Memory l Explanatio o ef Answer KA: l 2.4.27 l RO Value: l3.0 l SRO Value: l3.5 l Section: l PWG l RO Group: l 1 l SRO Group: l1 System / Evolution Title: KA Emergency Procedures / Plan Statement: Knowledge of fire in the plant precedure.
Reference l Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fire Protection NPDAP 3.5 Ill.N
6 Conduct of Operations 1/2LP-SQS-48.1
1 Question Source l New l Question Modification Method l Qrestion Source Comments: l Material Required for Examination: . Page 17 __ - _ _ _ _ - _ _ _ _ _ _ -
a u ,4 .
. Question Topic: l Rod motion control If a power mismatch signal is generated by the Rod Control System, which of the following parameters determines the magnitude of the gain imposed by the variable gain unit? a.
Median Tave i b.
Median delta T c.
N44 Power d.
Turbine Impulse pressure l Ans: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio n of Answer i KA: l 001 Al.02 l RO Value: l3.1 l SRO Value: l3.4 l Section: lSYS l RO Group: l 1 l SRO Group: l1 l l System / Evolution Control Rod Drive System Title: KA Ability to predict and/or monitor changes in parameters associated with operating the Control Rod Drive System Statement: controls including: T-ref _ Reference Reference Number Reference Section Page Number (s) Revision Learn.
Ob) Reactor Control and lOM-l.$.A.51
iss 4 Rev Protection
Re:ctor Control and 10M-l. i.D
iss 4 Rev
Protection
Full Length Rod Control LP-SQS-1.3
l Question Source l NRC Exam Bank l Question Modification Method l l Question Source Comments: l M:terial Reouired for Examination: I ! Page 18 - _ - _ _ _ _. _ _ _ _ - _ - - _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. ....
. . . . . I' Question Topic: l Misoperation of Bank Selector Switch Given the following conditions:
Reactor Power - 72%
Control Rods are at step 210 on Control Bank D . AOP 1.1.1, Failure of RCCA Control Bank to Move, is implemented due to
rod control problems The RO incorrectly places the Control Rod Bank Sel Sw in CONTROL BANK D = l instead of MANUAL ! Rods are withdrawn 5 steps before this is discovered
l If the Control Rod Bank Sel Sw is placed in Manual at this point, which of the following will occur? a.
Upon shutdown, all Control Bank D rods will remain 5 steps withdrawn from the core.
b. Upon shutdown, the ROD BOTTOM / ROD DROP alarm will actuate 5 steps sooner than expected.
While operating, the Rod Insertion Limit alarms (A4-116 and A4-134) for Control Bank D would c.
, l actuate 5 steps lower than the actual alarm setpoint positions.
i l d.
While operating, the Bank Demand Position Indict. tion will read 5 steps lower than the Analog Rod l Position Indication.
l ' Ars: la l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio a cf Answer , KA: l 001 K4.02 l RO Value. l 3.8 l SRO Value: l3.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1 I System / Evolution Control Rod Drive System Title: KA Knowledge of Control Rod Drive System design feature (s) and or interlock (s) which provide for the following: i St:tement: l Control rod mode select control (movement control) R:ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj reactor Control & Protection I OM-1.1.D Bank Overlap 15-16 iss 4 Rev-Instrumentation and
Controls Fulllength Rod Control LP-SQS-1.3 Ill.F.1
4 6.a QYestion Source l New l Question Modification Method l Q estion Source Comments: l Miterial Required for l Examination: ,. l Page 19 _ _ - _ - _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - ___-
Questio] Tepic: l Subcooling Margin l During a natural circulation cooldown the required number of CRDM fans cannot be started.
During the cooldown, upper head voiding is prevented by: e. venting the head via reactor vessel head vents.
b. verifying incore thermocouple temperatures are within an allowable range ofloop tensperatures.
c.
increasing the minimum subcooling margin during portions of the cooldown.
d. periodically injecting cold Safety Injection water intc the Hot legs.
Ars: lc l Exam Level: lS l Cognitive Level: l Comprehension i Espirnatio a cf Answer , KA: l 002 K5.15 l RO Value: l4.2 l SRO Value: l4.6 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Coolant System Title: KA Knowledge of the operational implications of the following concepts as they apply to the Reactor Coolant System: Statement: Reasons for maintaining subcooling margin during natural circulation Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj EOP Generic issues LP-SQS-53.2
13 Natural Circulation IOM 53B.4.ES-0.2
2-3 IssIB Cooldown Background Rev 4 Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M:t: rial Required for Ex:mination: , . i Page 20
Question Topic: l SG temperature effect upon start of RCP ' Given the following conditions: Plant heatup in progress
RCS temperature - 175 F
RCS pressure -325 psig
Pressurizer level-28%
, l
Preparations are underway for the start of the first RCP, RCP 1 A The requirement of having less than 25 *F difference between SG temperature and the primary system temperatures: a.
is not applicable since this is the first RCP to be started.
b. prevents an RCS overpressure event.
c.
prevents exceeding RCS heatup rate =. d. prevents exceeding RCS cooldown rates.
~ ~ Ans: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio
- cf Answer KA: l 003 Kl.10 l RO Value: l3.0 l SRO Value: l3.2 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution l Reactor Coolant Pump System Title:
l KA Knowledge of the physical connections and/or cause-effect relationships between Reactor Coolant Pump System Statement: ard the following: RCS Reference Reference Number Reference Section Page Number (s) Revision ' Learn.
Obj Reactor Coolant Pump IOM-6.4.A II.V
iss 4 Rev Stanup
RCS - Reactor Coolant LP-SQS-6.3 lil.A
4 12.A Pumps Question Source l New l Question Modification Method l Question Source Comments: l M:terial Required for Examination: l Page 21 ._ _ _ - - _ _ - _ _ _ _ _ -
Questio: Tcpie: l RCP power supplies The reactor is at 35% with the electrical busses in NSA. Unit Station Service Transformer ID develops a fault opening [4KV ACB 241D) USST ID Supply to 1C 4KV Bus and [4KV ACB 341D] USST 1D Supply to ID 4KV Bus The auto bus transfer fails to operate on C & D Bus.
Which of the following lists all running RCPs? a.
RCP1A b. RCP 1 A.and 1B c.
RCP IB and IC d.
RCP IC Ats: lb l Esam Level: lS l Cognitive Level: l Memory l Esplanatio e af Answer KA: l 003 K2.01 l RO Value: l3.1 l SRO Value: l3.1 l Section: lSYS l RO Group: l 1 l SRO Group: l1 Syst:m/ Evolution Reactor Coolant Pump System Title: KA Knowledge of electrical power supplies to the following: Statement: _ RCPS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj 4KV Distribution System LP-SQS 36.1 !!1.B.2
Reactor Coolant System - LP-SQS-6.3 1.C. I i1
1 Reactor Coolant Pumps Qrestion Source l Nev/ l Question Modification Method l Qrestion Source Comments: l Material Required for Eemination: l l l
Page 22
l Question Topic: l TS sval for charging pump Given the following conditions: l . * - Plant heatup in progress
- RCS temperature - 175 F'
- RCS pressure - 325 psig -
- Charging pump [1CH-P-1B] is in service.
i e
- Charging pump [lCH-P-1 A) is inoperable.
' Which of the following describes limitations, if any,if[1CH-P-1C] were to be placed in service on AE Bus, and [1CH P-1B] were to be removed from service? a.- [1CH-P-1B] must be stopped and placed in PULL-TO-LOCK prior to taking [lCH P-1C] out of PULL-TO-LOCK.
b. [lCH-P_-1B] must be stopped and placed in AUTO prior to taking [1CH-P-1C) out of PULL-TO-' LOCK.
c. [1CH-P-1B and 1C) r ?v be run simultaneously for up to 15 minutes, after which [1CH-P-1B] must be stopped and placed in PULL-TO-LOCK.
' d. Both Charging Pumps mg be run without restriction until [1CH-P-1B] is removed from service.
Ass: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explomatio i cefAnswer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l 4.0 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System /Evolutica Chemical and Volume Control System Title: KA Conduct Of Operations Statement: Ability to apply technical specifications for a system.
Reference Reference Number Reference Section Page Number (s) Revision 12arn.
Obj Beaver Valley - Unit 1 3.4.9.3 3/4 4-27a Amend Technical Specifications No.193 Placing the Spare Charging l OM-7.4.W IV.C W 9-13 Iss 4 Rev
Pump into Operation
CVCS LP-SQS-7.1 IV.A. B
12 Question Source l New l Question Modincation Method l Question Source Comments: l Material Required for Technical Specifications - Examtmation: Page 23
. - - _ _ _ _ - _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ - Question Topic: l Evd of1:ak in R gin Hx Given the following conditions: ' Reactor power-90%
Pressurizer level-51% stable
- VCT level - 30% rising Letdown flow on [FI-CH-150] - 60 gpm
- Charging flow on [FI-CH-122] - 45 gpm
. Seal Injection flows - 8 gpm (A); 10 gpm (B); 7 gpm (C)
- RCP #1 seal leakoff flows - 4 gpm (A); 4 gpm (B); 2 gpm (C)
Which of the following would result in the conditions above? a.
A leak exists in the Seal Water Heat Exchanger.
b. RCP #1 Seal Bypass Valve [MOV-CH-307] was inadvertently opened.
c. Letdown Pressure Control valve [PCV-CH-145] has failed open.
d. A leak exists in the CVCS Non-Regenerative Heat Exchenger.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio c cf Answer KA: l 004 K6.07 l RO Value: l2.7 l SRO Value: l2.8 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Chemical and Volume Control System Title: KA Knowledge of the of the efreu of a loss or malfunction on the following will have on the Chemical and Volume Statement: Control System: Heat exchange 3 and condensers Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj CVCS LP-SQS-7.1 II.S
6 2, 9 Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: ! l
Page 24 l. - !
Question Topic: l RHR/RCS pressure response Given the following conditions: Plant cooldown is in progress at 20 F/hr
RCS temperature - 155 F
Pressurizer level [LI-l RC-462] Cold Calib - 100%
RHR Pump 1 A is running with flow of 4000 gpm set on [MOV-RH-605] RHR
Flowin AUTO [MOV-RH-758] Residual Heat Removal Hx FCV demand is set at 40% -* [MOV-CH-142] RH LTDN to Non Regen Hx Inle Flow Control Viv demand is
set to 75% Controller for [PCV-CH-145] LP LTDN Back Press Reg Viv is set in MANUAL at the position that is
maintaining 50 psig with charging flow balanced If[ HIC-RH-758] controller causes [MOV-RH-758] to close with NO operator action, which of the following are the results for the first 10 minutes? a.
RHR flow will decrease and RCS pressure will decrease.
b. ' RHR flow will increase and RCS pressure will increase.
c.
RHR flow will remain the same and RCS pressure will decrease.
d. RHR flow will remain the same and RCS pressure will increase.
I Ans: ld l Exam level: lS l Cognitive level: l Comprehension l Explanatio ocf Answer ' KA: l 005 K3.01 l RO Value: l3.9 l SRO Value: l 4.0 l Section: lSYS l RO Group: l 3 l SRO Group: l3 System / Evolution Residual Heat Removal System Title: KA Knowledge of the effect that a loss or mstfunction of the Residual Heat Removal System will have on the Statement: following: RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Residual Heat Removal IOM-10.4.A E, F A 8-9 Iss 4 Rev l System Startup(Plant
cooldown) And Operation RHRS LP-SQS 10.1 D.2.e, f 7-8
5.a. b, f; 10 Question Soutee l New l Question Modification Method j Question Source Comments: l Material Required for Examination: I^ L Page 25 - _._ - - _ -
(.. l l Question Topic: l Los ef ONE St Accum Given the following conditions: Reactor power is 55%
Accumulator [ISI-TK-1 A) level is 85%
Accumulator [1SI-TK-1 A] pressure is 657 psig l
l.
SI Accumulator Isolation Valve [MOV-1SI-865A] is closed
The lockoutjack is removed
Reactor shutdown was initiated due to the accumulator conditions a > l Which of the following states the response of the SI Accumulators if a Design Basis LOCA occurs on the ' Loop B Cold Leg? a.
THREE Accumulators will fully inject into the core.
b. THREE Accumulators will fully inject into the core, provided the operator manually opens [MOV-ISI-865A]. c.
TWO Accumulators,1B and IC, will fully inject to the core.
d. ONE Accumulator, IC, will fully inject to the core.
Ans: ld l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio 'Ihe 1B Accumulator will discharge through the break o gf Answer KA: l 006 K6.02 l RO Value: l3.4 l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Emergency Core Cooling System Title: KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Emergency Core Cooling Statement: System: Core flor 4 tanks (accumulators) Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj i SIS LP-SQS-11.1 Vill.D.), XI.C.2 18,23
7.d,12.a Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Ermination: j l
Page 26 l - _ _ __-______ - -_ -
j.
! Question Topic: l Source of PRT conditions Reactor is a 100% with all systems in NSA. The operator observes that PRT level has increased.
Which of the following can cause the level increase? a.. A relief valve on the CCR system inside containment has lifted.
b. RCP #2 Seal Leak off flow has increased.
c.
A PORV is leaking.
d. RCP #1 Seal Leak off flow has increased.
Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio n ef Answer KA: l 007 A3.01 l RO Value: l 2.7* l SRO Value: l2.9 l Section: lSYS l RO Group:- l 3 l SRO Group: l3 System / Evolution Pressurizer Relief Tank / Quench Tank System Title: KA Ability to monitor automatic operations of the Pressurizer Relief Tank / Quench Tank System including: Statement: Components which discharge to the PRT Reference Reference Number Reference Section Page Number (s) Revision Learn.
O,,bj Alarm - Pressurizer Relief lOM-6.4.AAF PC No. 2 AAF 2-3 iss 4 Rev Tank Levelliigh-Low
i Pressurizer end Pressure LP-SQS-6.4 1.B.2.c 4-5
7 R, lief Systems Reactor Coolant System-LP-SQS-6.3 Reactor Coolant Pumps Question Source l New l Question Modification Method l Question Source Comments: l ) Material Required for Examination: ! l Page 27 l l l
( l Question Topic: l PORV operation [MOV-RC-535] Pressurizer Power ReliefIsolation Valve is closed due to [PCV-RC-455C] PORV leaking.
[PT-RC-445] Pressurize Pressure has failed downscale.
Select the available automatic overpressure protection, if any.
) a.
No PORVs will protect against overpressure.
b. Only PCV-RC-455D will protect against overpressure.
c.
Only PCV-RC-456 will protect against overpressure.
i d. Both PCV-RC-456 and 455D will protect against overpressure.
~Ars: la l Eram Level: IS l Cognitive Level: l Application l Explanatio c(f Answer KA: l 010 K4.03 l RO Value: l3.8 l SRO Value: l4.1 l Section: l SYS l RO Group: l 2 l SRO Group: l2 , System / Evolution Pressurizer Pressure Control System Title: KA Knowledge of Pressurizer Pressure Control System design feature (s) and or interlock (s) which provide for the Statement: following: Over pressure control Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Instrument Failure Procedure AOM-6.4-IF Figure
iss 4 Rev
Pressurizer & Pressure Relief LP-SQS-6.4
11 Syrtem Question Source l New l Question Modification Method l Question Source Comments: l Material Required for IOM-6.4-l P I Ex:mination: Page 28
Questici Tcple: l Pressurizer Lev 1 Rx trip , Pressurizer Level Control Channel Selector is selected to LT 459 & 460. All plant conditions are stable.
Which of the following will result in a reactor trip due to high pressurizer level? a.
At 5% power LT-RC-461 fails low.
b. At 5% power LT-RC-459 fails high.
c.
At 25% power LT-RC-460 fails low.
l l d. At 25% power LT-RC-461 fails low.
Avs: lc l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio o cf Answer 'KA: l 01i K1.04 l RO Value: l3.8 l SRO Value: l3.9 l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Level Control System . .-. Title: KA Knowledge of the physical connections and/or cause-effect relationships between Pressurizer Level Control Statement: System and the following: RPS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj RCS - Instrument failure IOM-6.4.IF ll.a II.C.I.a IF 8-9 iss 4 Rev ~
Pressurizer and Pressure LP-SQs-6.4 1.D.I.f 9-10
12 Rtlief System Q:estion Source l New l Question Modification Method l Q estion Source Comments: l M;terial Required for Ex:mination;
! Page 29
e Question T:pic: l Evil OTDT & OPDT setpoints on input failure During operation at 97% power one Tuo, instrument is reading 4 degrees higher than other Tuo, instruments.
All Ta temperatures are equal.
Which of the following describes the effect on OPdeltaT and OTdeltaT for the loop with the highest T,,,7 Loop deltaT will be a.
closer to both OPdeltaT and OTdeltaT trip setpoints.
b. closer to its OPdeltaT trip setpoint, but will be farther from its OTdeltaT trip setpoint.
c.
farther from its OPdeltaT trip setpoint, but will be closer to its OTdeltaT trip setpoint.
d. farther from both OPdeltaT and OTdeltaT trip setpoints.
A;s: la l Eram Level: lS l Cognitive Level: l Comprehension l Explanatio e cf Answer KA: l 012 A2.05 l RO Value: l 3.l* l SRO Value: l 3.2* l Section: lSYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Protection System Title: KA Ability to (a) predict the impacts of the following on the Reactor Protection System and (b) based on those Statement: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Faulty or erratic operation of detectors and function generators Reference Reference Number Reference Section Page Number (s) Revision Learn.
l Obj RCS-Instrument Failure IOM-6.4.lF !!.B, III.
IF 32-33,35-36 iss 4 Rev
Reactor Protection System LP-SQS-1.1 V.C.16 25-26
8 Reactor Coolant System LP-SQS-6.5 IV.A 17 20 5.a. b Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l MI.terial Required for i Ex mination: ! l t l l l Page 30
_ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ - _ - _ _ - - _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ ___-__ _ - _ _ - - _ - - _ _ _ _ _ _ _ _ _ -. _ _ _ _ _ _ Question Topic: l Operation of BOTli Bypass Trip Br akers RPS testing is in progress for RPS train B and the status of the breakers are as follows: l Reactor trip beakers (RTA and RTB) closed
l Reactor bypass breaker B (BYB) closed
l ' Bypassing both RPS trains simultaneously is prevented by: l a.
tripping only BYA ifit is racked in and its CLOSE pushbutton is depressed.
b. tripping only BYB if BYA is fully racked in, c. preventing closure of BYA ifit is racked in.
d. tripping all reactor trip and bypass breakers if BYA is racked in aad its CLOSE pushbutton is depressed.
A:s: ld l Exam Level: lS l Cognitive Level: l Memory l Esplanatio ' c ef Answer KA: l 012 A3.07 l RO Value: l4.0 l SRO Value: l 4.0 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Protection System Title: KA Ability to monitor automatic operations of the Reactor Protection System including: Setement: Trip breakers Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Control and lO M 1.1.B RP,2nd paragraph
iss 4, Protection - Summary Rev.0 Description Reactor Protection System LP-SQS-l.2 11.1
6 8, 9 Hardware QIestion Source l Facility Exam Bank l Question Modification Method l Qrestion Source Comments: l Material Required for Examination: ,_ Page 31 .. _ _ _.
l Question Topic: l Containment Pressura logics Containment pressure instrument PT-LM-100C has failed downscale. All appropriate actions of lOM-l 1.4.IF, Instrument Failure Procedure, have been completed.
Subsequently PT-LM-100D fails upscale.
Which of the following lists all expected actions? a.
CIA and SI b. CIA, SI and MSLI l c.
CIB and MSLI d. CIA, CIB, SI and MSLI A's: lb l Exam Level: lS l Cognitive Level: ! Comprehension l Explanatio o tf Answer KA: l 013 A2.06 l RO Value: l 3.7* l SRO Value: l4.0 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Engineered Safety Features Actuation System Title: KA Ability to (a) predict the impacts of the following on the Engineered Safety Features Actuation System and (b) Statement: based on those predictions, use procedures to correct, control, or mitigate the consequences of those abncrmal operation: Inadvertent ESFAS actuation Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Instrument Failure Procedure 10M-1.4.lF ll.C
Iss 4 Rev i Reactor Protection Trip LP-SQ-1.1
9 Logics Q estion Source l Facility Exam Bank l Question Modification Method l Q:estion Source Comments: l M;terial Required for Examination:
Page 32 J
Questio2 Tepic: l Operation following Si sign 11 A steam break has occurred causing an SI on high containment pressure. Reactor Trip Breaker BYA will NOT open. The crew has transitioned to ES-1.1, SI Termination. If containment pressure remains above the Si setpoint, which of the following will occur if both SI Reset Pushbuttons are depressed? a.
Neither train of S1 will reset.
b. Only one train of SI will reset.
c.
Both trains of SI will reset but one train will immediately reinitiate.
d. Only one train of S1 will reset. The reset train will immediately reinitiate.
A:s: lc l Exam Level: lS l Cognitive Level: l Application l Esplanatio a cf Answer KA: l 013 A3.02 l RO Value: l4.1 l SRO Value: l 4.2 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Engineered Safety Features Actuation System Title: KA Ability to monitor automatic operations of the Engineered Safety Features Actuation System including: Statement: Operation of actuated equipment Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj FS AR Logic Diagrams Figure 7.2-1 Sheet 8 Reactor Protection System LP-SQS-1.1 VI.E.1.f 34-351
9 j l Qrestion Source l Facility Exam Bank l Question Modification Method j Question Source Comments: l Miterial Required for Figure 7.21 Sheet 8 Examination: l l ) l l l I Page 33 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
I Question Topic: l ROD BOTTOM alarm f-During a reactor startup, when does the ROD BOTTOM / ROD DROP alarm (A4-126) become active for i each control bank? The alarm will actuate for a dropped rod for: a.
any Control Bank whenever Control Bank A RPI output is above 20 steps.
b. each Control Bank whenever that Control Bank demand position is above 35 steps.
c.
Control Banks A, B and C whenever their Control Bank demand position is above 35 steps, and for Control Bank D whenever Control Bank D demand position is abov.: 20 steps.
d. Control bank A whenever Control Bank A RPI output is above 20 steps, and for Control Banks B, C and D whenever their Control Bank RPI output is above 35 steps.
Ans: Id l Exam Level: lS l Cognitive level: l Memory l Explanatio o ef Answer KA: l2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: l SYS l RO Group: l 2 l SRO Group: l1 System / Evolution Rod Position Indication System i Title: ' KA Emergency Procedures / Plan Statement: Knowledge of annunciators alarms and indications, and use of the respoE instructions.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Control & Protection IO M l.l.B RPI, Ist & 2nd
Iss 4 Rev - Summary Description paragraphs
RPI and Insertion Limits LP-SQS-1.4 VI.B. C 5-6
2.b c Reactor Control and lOM-l.2.B
Protection Setpoints Question Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: l Mrterial Required for Ex mination: i l
, Page 34 ! _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _
[. l l-Question Tcpic: l determination of NIS counts by IR/SR status Given the following conditions: Reactor tripped from 100% power
Following transition to ES-0.1 " Reactor Trip Response", Intermediate
l Range NIS is reading 1E-7 amps Five minutes later Intermediate range NIS is reading 2.2E-9 amps
How soon following the last reading will Source Range NIS provide correct readings? a.
4 minutes.
, , b. 8 minutes.
c.
10 minutes.
d.
13 minutes.
A;s: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio P=Pi 10((T)(SUR)) Determine SUR form IRNIS readings over 5 minutes which gives SUR = -1/3 dpm (constant n ef Answer rate). This SUR is used with IR activation setpoint ~ IE-10 gives time of 4.02 minutes.
KA: l 015 K$.06 l RO Value: l3.4 l SRO Value: l3.7 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Nuclear Instrumentation System Title: KA Knowledge of the operational implications of the following concepts as they apply to the Nuclear Instrumentation Statement: System: Subcritical multiplications and NIS indications Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Excore Inst. System IOM-2.1.C IR 2nd paragraph
iss 4 Rev - Major Components
Excore Instrumentation LP-SQS-2.1 IV.C.8
5 5, 8 System Question Source l New l Question Modification Method l Qrestion Source Comments: l Material Required for Examination: l ! l Page 35 - _ __-___-_-__-_-_-_-__a
Question Topic: l Leak in RVLIS A leak has occurred at the inlet to a RVLIS differential pressure transmitter.
Which of the following describes RVLIS system indication and how the leak will be isolated? l RVLIS hydraulic isolator position will indicate a leak has occurred. The leak will automatically a.
isolate.
l b. RVLIS hydraulic isolator position will indicate a leak has occurred. The leak can only be isolated l by closing a manual isolation valve.
RVLIS high volume sensor position will indicate a leak has occurred. The leak will automatically c.
isolate.
d. RVLIS high volume sensor position will indicate a leak has occurred. The leak can only by isolated by closing a manual isolation valve.
Ars: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio ccf Answer KA: l 016 K3.01 l RO Value: l3.4* l SRO Value: l3.6* l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Non-Nuclear Instrumentation System Title: KA Knowledge of the effect that a loss or malfunction of the Non-Nuclear instrumentation System will have on the Statement: following: RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj RVLIS Hydraulic isolator IOM-6.4.AG IV.A.7, 8 AG2 Iss 4 Rev Malfunction
RVLSI & Core Cooling LP-SQS-6.7 II.B.e, f; ll.G.c; II.H 4-5,16-17.,22-
6 Monitor
Question Source l New l Question Modification Method l Q:estion Source Comments: l M;terial Required for Ermination: i i l l l Page 36 )
i-l Question Topic: l Ev I of Natural Circulation for conditions ' Given the following conditions: A loss of offsite power occurred
, l A natural circulation cooldown was initiated
The five hottest T/Cs average temperature - 555 F l a ' RCS wide range pressure -1275 psig
All RCS Loop Tw - 552 F
All RCS Loop T,oi - 544 F
All SG pressures - 940 psig j
l Adequate natural circulation flow: (Refer to Att. 6A & 2G) ' a.
exists and the RCS is subcooled.
. b. does not exist and the RCS is subcooled.
c.
exists and the RCS is at saturation.
d.' does not exist and the RCS is at saturation.
, , Ans: lb l Exam Level: lS l Cognitive Level: l Application l Explanatio o of Answer KA: l 017 A3.01 l RO Value: l 3.6* l SRO Value: l 3.8' l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution in-Core Temperature Monitor System Title: KA Ability to monitor automatic operations of the in-Core Temperature Monitor System including: Statement: Indications of normal, natural, and interrupted circulation of RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj 0 F Plus Subcooling Based 10M-53A.I.6-A I issIB on Core Exit TCs Rev 2 Natural Circulation EOP Attachment 2-G I
issIB Verification Rev 2 EOP Generic issues LP-SOS-53.2 VIII.C
12 Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Steam tables, EOP att. 2-G and 6A Examination: Page 37 . _ _ _ _. _ _ - _ _ _ - _ _ - _ _ _ - _ _ _ - - _ _ _ _ _ _ - - _ -
Question Topic: l Power supply following CIB The Containment Air Recirculation fans are in NSA prior to a transient which causes CIB.
After CIB occurs, what will be the status of the Containment Air Recirculation fans? a.
Running in fast speed b. Running in slow speed ! l ! c. Tripped but the power supply is energized l l d. Tripped with the power supply deenergized Ans: ld l Exam Level: lS l Cognitive level: l Comprehension l Explanatio c cf Answer KA: l 022 K2.01 l RO Value: l 3.0* l SRO Value: l 3.1 l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Containment Cooling System Title: KA Knowledge of electrical power supplies to the following: Statement: Containment cooling fans Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj CNMT Vent - Summary IOM-44C.I.B CNMT Air
Iss 4 Rev Description Recirculation
Containment Ventilation LP-SQS-44C.I !!.A.1 1-2
5,7 Systems Question Source l New l Question Modification Method l Q'estion Source Comments: l M;terial Required for Ex mination: Page 38
Question T ple: l Quench Spray rIsponse to RWST lev 1 Given the following conditions: Reactor trip, Si and CIB occurred from 100% power due to a LOCA
RWST levei har decreased to 3 feet 9 inches
CIB has not been reset.
- What would be the status of the Quench Spray (QS) system?
(Assume no operator action has been performed in the Quench Spray system.)
a.
BOTil QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs closed, and TWO QS Chemical Injection pumps are running.
b. BOTH QS pumps are running with [MOV-lQS-103 A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs closed, and FOUR QS Chemical Injection pumps are running.
BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle c.
Bypass Isol Vivs open, and TWO QS Chemical Injection pumps are running.
d. BOTH QS pumps are running with [MOV-lQS-103A,103B] QSPP Cavitating Venturi Nozzle Bypass Isol Vivs open, and FOUR QS Chemical Injection pumps are running.
Ans: la l Exam Level: lS l Cognitive level: I Comprehension l Explanatio ocf Answer l KA: l 026 Kl.01 l RO Value: l4.2 l SRO Value: l4.2 l Section: lSYS l RO Group: l 2 l SRO Group: l1 System / Evolution Containment Spray System Title: KA Knowledge of the physical connections and/or cause-effect relationships between Containment Spray System and Statement: the following: ECCS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Loss of reactor Or Secondary E-l step 30
Iss IB Coolant Rev 4 Transfer to Cold Leg ES 1.3 step 6
IssiB Recirculation Rev 4 , CNMT Depressurization LP SQS-13.1 V.D. I 17-18 5.b System Question Source l New l Question Modification Method l l Question Source Comments: l Material Required for Examination: l l l Page 39 l _ - _ _ _ _ - . _ _ - _ _ _.
Question Topic: l Recombiner Ops ' Given the following conditions: A LOCA has occurred 24 hours ago l
ONE Hydrogen recombiner is placed in service when hydrogen concentration reaches 0.5%
With a recombiner in operation, containment pressure: should be maintained at approximately 8.9 PSIA, to prevent excessive recombiner flow.
a.
b.
will be adequate for recombiner operation ifit is maintained between 8.9 PSIA and -3 PSIG should be maintained slightly above atmospheric, to ensure sufficient recombiner flow.
c.
d.
should be maintained at approximately -2PSIG, to ensure sufficient recombiner flow.
Ans: Ic l Exam level: IS l Cognitive Level: l Application l Explanation of Answer ~ KA: l 028 A1.01 l RO Value: l3.4 l SRO Value: l 3.8 l Section: lSYS l RO Group: l 3 l SRO Group: l2 System / Evolution Hydrogen Recombiner and Purge Control System Title: KA ANiity to predict and/or monitor changes in parameters associated with operating the Hydrogen Recombiner and Statement: Purge Control System controls including: Hydrogen concentration Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Post DBA Hydrogen Control 10M-46.1.B 4th paragraph
Iss 44: System - Summary Rev.0 Description Post DBA H2 Control LP-SQS-46.1 11.C.2.d
3 8,9 System System Question Source l New l Question Modification Method l Question Source Comments: l M:.terial Required for OM 46.4.A Examination: Page 40
- - - _ - _ _ - _ _ - _ - - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - . Question Topic: l Ev:luation cf a leak Given the following conditions: Reactor power is 85%
Spent Fuel Pool is aligned for cooling
- - A leak has occurred in the suction of [FC-P-1 A] Fuel Pool Cooling Pump If the leak remains unisolated, Spent Fuel Pool level should stabilize at:
a. ~25 feet above the top of the fuel.
b. ~23 feet above the top of the fuel.
c. ~10 feet above the top of the fuel.
d. the top of the fuel.
Ans: Ic l Eram Level: lS l Cognitive Level: l Memory l Explanatio n of Answer KA: l 033 A2.03 l RO Value: l3.1 l SRO Value: l3.5 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Spent Fuel Pool Cooling System Title: KA Ability to (a) predict the impacts of the following on the Spent Fuel Pool Cooling System and (b) based on those Statement: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Abnormal spent fuel pool water level or loss of water level Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fuel Pool Cooling and lOM-20.1.B
1ss 4 Rev Purification
Fuel Pool Cooling and LP-SQS-20.1
6,9b Purification Question Source l New l Question Modification Method l ' Question Source Comments: l Material Required for Examination: i !' Page 4i L
Question Topic: l Transfer Cart Operation Which of following describes the interlock between the conveyor car drive and the upende s when l transferring the conveyor car from the transfer canal to the refueling cavity? l a.
Both upenders must be in the down position before the conveyor car can be moved.
t . i b. Only the upender in the refueling cavity must be in the down position before the conveyor car can be moved.
c.
Only the upender in the transfer canal must be in the down position before the conveyor car can be moved.
d. If upender in the refueling cavity is not in the down position, movement of the conveyor car can be initiated, however the conveyor car will stop before reaching the upender.
A ns: la l Exam Level: lS l Cognitive Level: l Memory l Esplanatio s ef Answer KA: l 034 K4.02 l RO Value: l2.5 l SRO Value: l3.3 l Section: l SYS l RO Group: l 3 l SRO Group: l2 System / Evolution Fuelliandling Equipment System Title: KA Knowledge of Fuel Handling Equipment System design feature (s) and or interlock (s) which provide for the Statement: following: Fuel movement Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fuel Handling Operations LP-SQS-6.13 XI.11.9.e
4 8.a 1 RP-12R-3.2 II.6.b
iss 0 Rev
Question Source l NRC Exam Bank l Question Modification Method l Question Source Cominents: l Material Required for Ex:mination: Page 42 -- - - -____ _______ ___-_ -_-- _ _ -- -_ ___________ I {; question Top 6es l SG level program Reactor power is 25% and all plant systems are in NSA.
! Which failure would decrease feedwater flow to all SGs? a.
ONE condenser steam dump fails open.
b. Heater Drain receiver Level Control Valve [LCV-ISD-106B] fails open.
I l c. Turbine First Stage Pressure channel [PT-lMS-446] fails low.
l l d. - Combined Feedwater Header Pressure channel [PS-1FW-151] fails high.
Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l l Explanatio c of Answer KA: l 035 Kl.01 l RO Value: l4.2 l SRO Value: l4.5 l Section: lSYS l RO Group: l 2 l SRO Group:,l2 System / Evolution Steam Generator System Title: KA Knowledge of the physical connections and/or cause-effect relationships between Steam Generator System and the Statement: following: MFW/AFW systems Reference Reference Number i Reference Section Page Number (s) Revision Learn.
Obj SG Feedwater System - IOM-24.lD SGWLC 7-8 iss 4 Rev Instrumentation and Controls
SG Feedwater System - 1OM-24.4.lF Attachment 5, II.A.2 IF 38 iss 4 Rcv 'ustrument Failure
Feedwater System LP-SOS-24.1 Ill.E.10.d
1.A ' Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: i l Page 43
Questio1 Topic: l Effect cf MS-PT-464 failing high Given the following conditions: The unit is in MODE 3 preparing for normal plant cooldown
Condenser Steam Dump System is automatically controlling T,y at 547 F in Steam Pressure Mode
[PT-1 MS-464] Main Steam Header Pressure fails high
Which one of the following describes the effect this will have on the Condenser Steam Dump system? a.
Two banks of steam dumps will open and remain open until manually closed.
b. Two banks of steam dumps will open but should reclose with no operator action.
c.
All banks of steam dumps will open and remain open until manually closed.
d. All banks of steam dumps will open but should reclose with no operator action.
Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio o e f Answer KA: l 041 K6.03 l RO Value: l2.7 l SRO Value: l2.9 l Section: l SYS l RO Group: l 3 l SRO Group: l3 System / Evolution Steam Dump System and Turbine Bypass Control Title: KA Knowledge of the of the effect of a loss or malfunction on the following will have on the Steam Dump System and ) Statement: Turbine Bypass Control: ' Controller and positioners, including ICS, S/G, CRDS Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj M:in Steam System IOM-21.5.A.24 I Iss 4 Rev
Main Steam System LP SQS-21.1
3 Question Source lNew l Question Modification Method l Qrestion Source Comments: l Material Required for j Examination: l l l Page 44 L.__
Quesison Topic: l NPS11 for FW Given the following conditions: '
- Reactor power - 100%
- A load rejection occua and the plant stabilizes at 45% power
'* ' Load rejection bistables " LOAD REJ 15-50%" and " LOAD REJ GREATER THAN 50%" are lit How are the Steam Generator Feed Pumps [1FW-P-1 A,1B] protected from a loss of suction pressure during the load rejection? j l a. The Feedwater Heater Bypass Valve [TV-1CN-100] opened and closed FOUR minutes later.
b. The Heater Drain Receiver Level Control Valve [LCV-ISD-106B] was maintained fully open until LOW-LOW level was sensed in the Heater Drain Receiver, c. The Heater Bypass to Heater Drain Pump Suction Valve [TV-CN-125] opened and closed four minutes later.
d. The Condensate Pumps Recirculation Valve [FCV-lCN-101] closed on the 15-50% load rejection and reopened FIVE minutes later.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio e cf Answer KA: l 056 Al.08 l RO Value: l2.3 l SRO Value: l 2.6* l Section: lSYS l RO Group: l 1 l SRO Group: l1 System / Evolution Condensate System Title: KA Ability to predict and/or monitor changes in parameters associated with operating the Condensate System controls Statement: including: MFW pump suction pressure Reference Reference Number Reference Section Page Number (s) Revision Learn.
l ! Obj Load Rejection AOP 1.35.2 step I1
Iss 3A Rev6 Figure 22-6 - Step Load 1OM 22.5.A.6
Iss 4 Rev i Rejection Ckt
Extraction Steam and lleater LP-SQS 23 Ill.C.7 8-9 12.E Drains Question Source l Other Facility [ Question Modification Method l Question Source Comments: l l M;terial Required for Examination: l ' Page 45 _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
. - , 's ' ' ' , - , l Question Topie: l Restoration of FW capability An inadvertent Si signal occurred at 100% power. The condition causing the SI signal is no longer present.
j All systems function as designed and RCS conditions stabilize as expected following the inadvertent St.
Which of the following states the condition (s) that would have to be met to feed via IFCV-lFW-479(489)(499)], SG F d.4ypass FCVs? Only the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
a.
b.
P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
c.
SI would have o be reset and the FW1 FW BYPASS VALVE RESET pushbuttons would have to be depressed.
d.
S' would have to be reset, P-4 would have to be cleared and the FWI FW BYPASS VALVE RESET pushbuttons would have to be depressed.
Ans: Ia l Exam Level: lS l Cognitt/c Level: l Application l-Esplanatio u of Answer KA: l 059 A4.1l l RO Value: l3.1 l SRO Value: l3.3 l Section: lSYS l RO Group: l 1 l SRO Group: ll System / Evolution Main Feedwater System Title: KA Ability to manually operate and/or monitor in the control room: Statement: Recovery from automatic feedwater isolation Reference Reference Numtwr Reference Section Page Number (s) Revision Learn.
Obj Feedwater System LP-SQS-24.1 lil.E. I 1.
15-16
5.j, 7.A.(12) Reactor Protection Systems LP-SQS-1.1 VI.E.5 38-39
9 Updated FSAR Figure 7.2-1 sheet 1&l3 Question Source l Facility Exam llank l Question Modification Method l Question Source Comments: l Material Required for Figure 7.2-1 sheet 1 & 13 Examination: Page 46
., , e - - . l Qrestion Topic: l SGWLC inputs Given the following conditions: Reactor power is 20%
Feedwater has been transferred to the Main Feed Regulating Valves
All systems are NSA
Narrow Range SG 1C levelis 44%
[FCV-lFW-499] 1C SG FW Bypass Viv is manually opened 15%
Aftet plant conditions stabilize, which parameter (s) will be different from those prior to [FCV-1FW-499] l opening? a.
Only [FCV-lFW-498] IC Main FW Reg Viv position b.
[FCV-lFW-498] IC Main FW Reg Viv position and Narrow Range SG IC Level c.
Only Narrow Range SG IC Level d. Narrow Range SG 1C level and Stm Gen 1C Feed Flow indication A:s: l'n l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n of Answer KA: l 059 Kl.04 l RO Value: l3.4 l SRO Value: l3.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1 l System / Evolution Main feedwater System l Title: KA Knowledge of the physical connections and/or cause-efTect relationships between Main Feedwater System and the St:tement: following: S/GS water level control system Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj SG Feedwater System - lOM 24.1.D SGWLC 7-8 Iss 4 Rev Instrumentation and Controls
Feedwater System LP-SQS-24.1 111.E.10.
14-15
1.A Question Source l New l Question Modification Method l , Qrestion Source Comments: l M:terial Required for i ' Examination: . Page 47 _____ -_-_- - _ - ___-
l f t Question Topic: l Relationship of AFW steam supply & feed suppli:s to SG Given the following conditions: Reactor power - 100%
A loss of all AC power occurs
Auxiliar" Feed Pump IFW-P-2 starts and runs o The stean supply line from SG B to IFW-P-2 ruptures at the connection
to the main steam line.
The steam break prevents access to the Main Steam Valve Room
Which of the following describes how the Auxiliary Feed System is affected by the above conditions? All SGs will blowdown through the mpture, and NO auxiliary feed will be available.
a.
b.
SG A and SG B will blowdown through the rupture, but NO auxiliary feed will be available.
c.
SG A and SG B will blowdown through the rupture, but auxiliary feed can be established by opening the manual steam supply isolation valve from SG C.
. d.
Only SG B will blowdown through the rupture, and auxiliary feed can be established from SG A.
A"s: ld l Exam Level: lS l Cognitive Level: l Memory l f.xplanatio c ef Answer KA: l 061 K3.02 l RO Value: l4.2 l SRO Value: l4.4 l Section: l SYS l RO Group: l 1 l SRO Group: l1 System / Evolution Auxiliary / Emergency Feedwater System Title: KA Knowledge of the efTect that a loss or malfunction of the Auxiliary / Emergency Feedwater System will have on I Statement: the following: S/G Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj SG Feedwater System IOM-24.1.C Auxiliary Feed Pumps 2-3 iss 4; Rev 2 Feedwater System LP-SQS-21.1 til.J.9
7 1.13 SG Feedwater System LP-SQS-21.1 Ill.L.3.a
Q :estion Source lNew l Question Modification Method l Question Source Comments: l M;terial Required for Enmination: Page 48 l
_ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ Questici Tepic: l Overcurrent effect on breaker operation The Unit is at 85%. Which of the following conditions will result in bus I AE being maintained deenergized.
a.
[ACB-1 A10] 1 AE Emergency Bus feeder breaker trips on overcurrent.
b.- I AE Emergency Bus reverse phase PT blows a fuse.
c.
[ACB-41C) 1 A Normal 4KV Bus Feeder Breaker trips on overcurrent.
d. [ACB-41C) 1 A Normal 4KV Bus Feeder Breaker trips on Unit Statioa Service Tranformer 1C Differential Trip.
) Ars: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio ucf Answer KA: l 062 K4.01 l RO Value: l2.6 l SRO Value: l3.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2 i System / Evolution A.C. Electrical Distribution Title: KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following: St:tement: Bus lockouts Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj 4160V Emergency Bus I AE lOM-36.4.ACZ iss 3 Rev ACB 1 A10 Auto Trip i Diesel Generators LP-SQS-36.2
6 Qrestion Source l Previous 2 NRC Exams l Question Modification Method l Question Source Comments: l - Material Required for Examination: Page 49
_ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - - _ - _ _ _ - - _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ - _ _ _ _. _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ Question Tcpic: l Br:aker interlock (s) Reactor power is 25% during a startup. Electrical loads have been transferred to the Unit Station Service Transformer (USST).
In order for Bus I A to be setup for Auto Bus Transfer to the System Station Service Transformer, which of the following lists the required position of the Live Bus Transfer switch and the control switch for ACB 41A7 a.
Live Bus Transfer Switch - OFF ACB 41 A Control Switch - After Close b. Live Bus Transfer Switch - OFF ACB 41 A Control Switch - After Trip c.
Live Bus Transfer Switch - ON ACB 41 A Control Switch - After Close d. Live Bus Transfer Switch - ON ACB 41 A Control Switch - After Trip ATs: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio a cf Answer KA: l 062 K4.01 l HO Value: l2.6 l SRO Value: l3.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution A.C. Electrical Distribution Title: KA Knowledge of A.C. Electrical Distribution design feature (s) and or interlock (s) which provide for the following: St:tement: Bus lockouts R:,fer ace Reference Number Reference Section Page Number (s) Revision Learn.
Obj MV Station Service System IOM-36.1.E 20-21 iss 4 Rev - Specific Instrumentation I and Controls 4KV Distribution LP-SQS 36.1
7 3.
, QIestion Source l New l Question Modification Method l Qrestion Source Comments: l M:terial Required for i Enmination: l Page 50 l -- -_-
Questio2 Topic: l Response to ground indication l l DC Bus 1-2 ground voltmeter went from 0 volts to -105 volts. The DC Bus is in NSA for 100% power operations.
Which of the following describes the effect the ground will have on DC bus operations? a.
The ground has caused actual voltage to the DC loads to decrease to 105 Volts.
b. The affected battery will discharge significantly faster than designed.
c. The bus will operate as required but the bus reliability has decreased.
d. Another ground on the same polarity of the bus will cause a short circuit.
Ars: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 063 A2.01 l RO Value: l2.5 l SRO Value: l 3.2* l Section: lSYS l RO Group: l 2 l SRO Group: l1 System / Evolution D.C. Electrical Distribution Title: KA Ability to (a) predict the impacts of the following on the D.C. Electrical Distribution and (b) based on those Statement: predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Grounds ~ Reference Reference Number Reference Section Page Number (s) Revision Learn.
j Obj 125 V DC Control System-IOM 39.2 A.16
Iss 3 Rev Precautions & Setpoints
125 V DC Control System IOM-39.1
iss 4 Rev
125 VDC LP-SQS-39.1
Q:estion Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: ! i ! l i Page 51 - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
' Question Topic: l Reverse power trip of DG Diesel Generator No.1 is paralleled to 4160V Bus I AE for testing. The operator is in the process of adjusting load and voltage wb: the Governor Control switch sticks in the LOWER position.
If NO operator action is taken, what will be the Diesel Generator response to this condition? DG frequency will: a.
decrease and the diesel will trip on reverse power.
b. decrease and the diesel will trip on overcurrent.
c. remain constant but the diesel will trip on reverse power.
d. remain constant but the diesel will trip on overcurrent.
A*s: lc l Exam Level: lS l Cognitive Level: l Comprehension l ] Explanatio c ef Answer KA: l 064 Al.08 l RO Value: l 3.1 l SRO Value: l3.4 l Section: lSYS l RO Group: l 2 l SRO Group: l2 . System / Evolution Emergency DieselGenerators Title: KA Ability to predict and/or monitor changes in parameters associated with operating the Emergency Diesel Setement: Generators controls including: Maintaining minimum load on ED/G (to prevent reverse power) Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Transferring Emergency IOM-36.4.Q IV.A.9 & CAUTION Q2 Iss 4 Rev Feed
Transferring Emergency Busses I AE And IDF From Emergency Feed To Normal Feed Alarm DIESEL IOM-34.ADU A8-127 ADUl iss 3 Rev GENERATOR NO. I
REVERSE POWER Diesel Generators LP-SQS-36.2 VI.B
6 Q'estion Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 52 _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
Questiol Tepic: l Diesel Generator Trips A loss of off-site power occurred and the diesel generators are supplying the emergency buses.
Which of the following will trip a diesel generator? a.
The governor control switch in the control room is held in the RAISE position.
b. A governor failure causes engine speed to increase to 1050 RPM.
c.
Thejacket cooling water pump trips.
d. The coupling fails on the lube oil pump.
Ass: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio a c f Answer l KA: l 064 K4.02 l RO Value: l3.9 l SRO Value: l4.2 l Section: l SYS l RO Group: l 2 l SRO Group: l2 System / Evolution Emergency DieselGenerators Title: KA Knowledge of Emergency Diesel Generators design feature (s) and or interlock (s) which provide for the following: , Statement: Trips for ED/G while operating (normal or emergency) Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Local-Overspeed Trip IOM 36.4.AFN
Iss 3 Rev
Diesel Generators LP-SQS-36.2
6 Technical Specifications 4.8.1.1.2.b.4 3/4 8.4a Q;estion Source l Facility Exam Bank l Question Modification Method l Qrestion Source Comments: l M;terial Required for Examination: . i I Page 53 ) )
Question Topic: l Drain Tank Isol: tion Given the following conditions: Low Level Waste Drain Tank level is 110 inches
, I' The discharge permit has been approved at discharge rate of 15 gpm
The discharge is in progress at 15 gpm a What condition will automatically stop the release? Both [TV-LW-105] Liquid Waste Effluent Trip valve and [FCV-LW-104-2] High Range Liquki a.
Waste Efiluent Flow Control Valve closing on high-high radiation signal from [RM-LW-104).
b.
[FCV-LW-104-2] High Range Liquid Waste Effluent Flow Control Valve closing on low flow rate, [FCV-LW-104-1] Low Range Liquid Waste Effluent Flow Control Valve closing on low Waste c.
Drain Tank level.
d. The Low Level Waste Drain pump tripping on low flow rate.
Ans: la l Esam 12 vel: lS l Cognitive Level: l Memory l Explanatio o c.f Answer KA: l 068 A4.04 l RO Value: l3.8 l SRO Value: l3.7 l Section: lSYS j RO Group: l 1 l SRO Group: l1 System / Evolution Liquid Radwaste System Title: KA Naility to manually operate and/or monitor in the control room: Statement: Aut(,matic isolation Reference Reference Number Reference Section Page Number ('.) Revision Learn.
Obj Liquid Waste Disposal LP-SQS-17,1 II.C.7. 8 & 10 11-13
2.b System _ Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: I Page 54 . _ _ _ - -__-_________-___j
f Questici Tepic: l Annunciator Operation l Due to a Steam Generator Tube Leak a Condenser Air Ejector Vent Monitor [RM-1SV-100] High alarm ' occurs causing Annuciator" Radiation Monitoring High"(A4-71) alarm to be received. Annuciator (A4-71) is acknowledged. Which of the following will cause Annuciator " Radiation Monitoring High"(A4-71) to reflash? a.
Condenser Air Ejector Vent Monitor [RM-1SV-100] rising to the High-High alarm setpoint.
b. Sto.'m Generator Blowdown Samnle Monitor [RM-ISS-100] rising to the High alarm Setpoint.
c.
Steam Generator N-16 Monitor [RM-1MS-102] rising to the High alarm Setpoint.
d.
High Capacity Steam Generator Blowdown Monitor [RM-1BD-101] rising to the High alarm Setpoint.- Ais: lb l Exam Level: lS l Cognitive Level: l Memory l Esplanatio a of Answer KA: l 073 A4.02 l RO Value: l3.7 l SRO Value: l3.7 l Section: lSYS l RO Group: l 2 l SRO Group: l2 Systert/ Evolution Process Radiation Monitoring System Title: KA Ability to manually operate and/or monitor in the control room: Stuement: Radiation monitoring system control panel R:ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Rui Monitoring System - I OM-43.1.D
Iss 4 Rev Instrumentation and Controls
Radiation Monitoring System LP-SQS-43.1
Q:estit,n Source l New l Question Modification Method l QIestion Source Comments: l Mrterial Required for Examination: i ! I Page 55 ._ _ _ __________________ -
_ Qrestion Topic: l Evaluation of available tir sources A leak has occurred in the Station Air System in the Fuel Building. [PI-ISA-101] Station Air Main IIeader and [PI-ll A-106] Station Instrument Air 11eader pressure indications are both lowering.
When Station Air pressure decreases to a specific setpoint, (TV-ISA-105] Station Air Header Trip Valve will: a.
open to supply instrument air loads.
b. open to supply contailunent air loads.
c.
close to ensure all station air will be supplied to the instrument air loads.
I d. close to maintain air to all station loads.
Ans: lc l Exam Level: lS l Cognitive Level: l Memory l Explanatio a of Answer KA: l 078 K4.02 l RO Value: l3.2 l SRO Value: l3.5 l Section: lSYS l RO Group: l 3 l SRO Group: l3 System / Evolution 3 instnament Air System Title: l KA Knowledge of instrument Air System design feature (s) and or interlock (s) which provide for the following: St:tement: Cross-over to other air systems Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Compressed Air Systems - IOM-34.1.D Station Air Header Trip
iss 4 Rev instrumentation.: and Controls D Valve
VOND 34-1 Compressed Air LP-SQS-34.1 IV.A & D
5 Qrestion Source l New l Question Modification Method l Qrestion Source Comments: l M terial Required for Ex:mination: Page 56
Question Topic: l Containm nt Building Penetrations during refurling Which of the following is NOT part of the Technical Specification. definition of CONTAINMENT INTEGRITY a..The containment leakage monitoring system is OPERABLE.
b. All equipment hatches are closed and sealed.
c.
The sealing mechanism associated with each penetration is OPERABLE.
d. The containment leakage rates are within their LCO limits.
' Ans: la l Esam Level: lS l Cognitive Level: l Comprehension l Esplanatio e Cf Ahswer KA* l 103 Kl.02 l RO Value: l3.9 l SRO Value: l 4.l* l Section: ISYS l HO Group: l 3 l SRO Group: l2 System / Evolution Containment System Title: _ KA Knowledge of the physical connections and/or cause-efTect relationships between Containment System and the Statement: following: Containment isolation / containment integrity Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Technical Specification 3/4.9.4 3/4 9-4 Containment System LP-SQS-47.1 VI.B
4 8.h _.
Question Source l New l Question Modification Method l QIestion Source Comments: l Material Required for Examination: . ! l i l Page 57 lL .._ _ __....___ _
( -] l Question Topic: l Determination cf pow;r increae Given the following conditions: EOL
Reactor power is 80% steady state
RCS T.,, i; on program
Control Rod position - 160 steps on Control Bank D
Control Rods begin to withdraw e When Control Bank D is at 170 steps the Control Rod Bank Sel Sw is e placed in MANUAL stopping rod motion If N0 further operator action is taken, what would be the affect on actual power level and RCS T,,, after conditions stabilize? Reactor power and RCS T,,, would both rise equally by an amount equivalent to the reactivity a.
addition.
b. Reactor power would rise by an amount equivalent to the reactivity addition and RCS T,,, would , remain approximately 571 F.
Reactor power would remain approximately 80% and RCS T,,, would rise by an amount equivalent c.
to the reactivity addition.
d. Neither reactor power nor RCS T.,, would be significantly affected.
Ans: lc l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio Reactivity addition by rod movement would add power to RCS. Since turbine load controls power level on NIS, acf Answer RCS would heat up. By using Power defect curves could determine the equivalent power level the reactivity would allow and the associated Tavg at that power will approximate the temperature of the RCS (Use of Power Defect Curves provides an approximation because it includes Fuel temp / Doppler coefficient, but impact is relatively small compared to moderator temp coefficient over area of concern) KA: l 001 AKl.03 l RO Value: l3.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution Continuous Rod Withdrawal Title: KA Knowledge of the operational implications of the following concepts as they apply to Continuous Rod St:tement: Withdrawal: Relationship of reactivity and reactor power to rod movement Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Full Length Rod Control LP-SQS 1.3
16 Question Source l New l Quesilon Modification Method l Question Source Comments: l htterial Required for Examination: Page $8 _ _ _ _ _ _ _ _ _ _ _ _ _
Question Topic: l Operation cf Disconnect Switch Given the following conditions:
- Reactor power - 5%
Control rod F-6 in Control Bank D has fully dropped.
- Recovery of the dropped rod is in progress per AOP 1.1.5 " Dropped RCCA"
- All Disconnect Switches in Control Bank D are in DISCONNECT except for F-6 l
Which of the following describes alarms that will be received and their effect on recovering the dropped ' control rod? ! a. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank Sel Sw in Control Bank D.
b. An urgent failure will be received, however rod recovery can proceed with the Control Rod Bank Sel Sw in Manual.
c.
A non-urgent failure will be received which will not affect control rod movement.
d. An urgent failure will be received, however rod recoverj can proceed after depressing the Rod Control Alarm Reset pushbutton.
Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio ] o c.f Answer KA: l 003 AK2.05 l RO Value: l2.5 l SRO Value: l2.8 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution Dropped Control Rod Title: KA Knowledge of the interrelations between Dropped Control Rod and the following: Statemer.t: Control rod drive power supplies and logic circuits Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Dropped RCCA A O P 1.l.5
5 iss 3A Rev 7
' Alarm - ROD CONTROL IOM 1.4.AAR A4105 Corrective AARI Iss3 Rev SYSTEM URGENT Action NOTE
FAILURE Full Length Rod Control LP-SQS l.3 !!.O.3 & IV.A.3 14 & 16 10;16 Question Source l New l Question Modification Method l l Question Source Comments: l Material Required for Examination: I l Page 59 ____- ________-_ -_-
_ _. _ _ _ _ _ _ _ _ - - _ _ - - _. - _ _ - _ _ _ _ _ - _ - _ _ _ - - - - - _ - - - _ -. - - - _ _ __.
- - - - - _ _ _ _. - _ Quest 6on Topic: l Operational limits & basis with given stuck rod Given the following condition,s:
- Reactor power - 85%
- Load increase is in progress
- Control Bank D is 2 steps above the RIL a Control rod K-6 indicates 15 steps below the remaining rods in Control Bank D e Control rod trippability is confirmed +
- Shutdown Margin is verified to be satisfied If the NSS decides to continue power operation with the control rod misaligned, which of the following describes required power reduction and the associated reason?
Reactor power must be reduced to at least: 75% power within ONE hour to remain in compliance with Rod Insertion Limit restrictions.
a.
b. 75% power within ONE hour to provide assurance of fuel rod integrity during continued operations.
50% power within FOUR hours to remain in compliance with Rod Insertion Limit restrictions.
c.
I d.
50% power within FOUR hours to provide assurance of fuel rod integrity during continued operations.
Ans: lb l Exam Level: lS l Cognitive Level: l Application l Explanatio e cf Answer KA: l 005 AKl.% l RG Value: l2.9 l SRO Value: l 3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Inoperable / Stuck Control Rod Title: KA Knowledge of the operational implications of the following concepts as they apply to Inoperable / Stuck Control Statement: Rod: Bases for power limit, for rod misalignment Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Beaver Valley - Unit 1 3.1.3.1 (ACTION C.3) 3/4 1-18-19 Amend Technical Specifications No.154 Be;.ver Valley - Unit i Bases 3/4.1.3 B 3/4 l-4 Amend Technical Specifications No.141 Full Length Rod Control LP-SQS 1.3 111.1.1
15 Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Maternal Required for Technical Specifications Examination:
l
l l Page 60 .j i i
_ _ _. ______ _ __ _____ _ __ _ _- - ____ _ _ _- _ - _ _ _ - _ _ _ ___-_____ _ ___ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ Question Topic: l Steam Dump AiTects ~ Given the following conditions: l , Reactor tripped from 100% power l .* l Reactor trip breaker (RTB), which provides P-4 input to Reactor Trip
Controller, CANNOT be opened after the trip
- Reactor trip breaker (RTA) opened l
Which of the following identifies where the RCS temperature should stabilize prior to placing the Steam l Pressure Mode Selector Switch in Steam Pressure Mode? a.
543 F.
b. 547 F.
c.
549 F.
d. 554 F.
Ans: le l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio c of Answer KA: l 007 EA2.03 l RO Value: l4.2 l SRO Value: l 4.4 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Trip Title: ! ) KA Ability to determine and interpret the following as they apply to Reactor Tnp.
Statement: Reactor trip breaker position Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj M in Steam Systems IOM 21.5.A.24
iss 4 Rev
M:in Steam System - IOM-21.1.D various 3-6 iss 4 Rev instrumentation and Controls
Main Steam Supply / Steam LP-SQS-21.1 Ill.D, Ill.E, V.C.5, 12-14,27 28, i.e, 3.a Dump System V.E.1 30-31 Question Source l New l Question Modification Method l j Question Source Comments: l
M terialRequired for Examination: I l l i l Page 61
' . - - _ _ _ _. _
Questio1 Tr pic: l Operation of controt rods during an ATWS A manual reactor trip was inititted at 100%, however the reactor will not trip. Step 1 of FR-S.1 is being performed. Control rods are in AUTOMATIC.
With the turbine tripped, which of the following describes required action concerning control rod insertion? Control rods should be inserted in: i ' a.
MANUAL even if they are inserting in AUTOMATIC.
b. AUTOMATIC provided rods are inserting in AUTOMATIC.
c.
AUTOMATIC until reactor power is less than 15% where the rods will stop, requiring MANUAL insertion.
d. AUTOMATIC until the Rod Insertion Limit is reached where the rods will stop, requiring MANUAL insertion.
Ais: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio acf Answer KA: l 007 EK3.01 l RO Value: l4.0 l SRO Value: l4.6 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Reactor Trip Title: KA Knowledge of the reasons for the following responses as they apply to Reactor Trip: Setement: ) I Actions contained in EOP for reactor trip R ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Response To Nuclear Power FR-S.1 step 1. RNO
issIB G;neration-ATWS Rev 4 Response To Nuclear Power IOM-53.4.FR S.I 111.1 Knowledp
1531B Generation-ATWS Rev 4 Background EOPs LP-SQS-53.3
1,3 Q estion Source l New l Question Modification Method l QIestion Source Comments: l j M:.terial Required for Examination: Page 62 _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ -
Question Topic: l Eval cf vapor space leak -Tech Spec limit Given the following conditions: - * '.'Ihe reactor is operating at 100% power
- ' A 1.2 gpm valve packing leak has occurred on [PCV-RC-455B] PRZR Spray Viv
- The Primary Drains Transfer Tank level is increasing -
Which of the following describes what type ofleakage this is and based on the leak size what action is. required per Technical Specifications? This leak is considered: a. Primary boundary LEAKAGE that requires Technical Specification entry, b.- Identified LEAKAGE that does not require Technical Specification entry.
c.
Unidentified LEAKAGE that requires Technical Specification entry.
d. Unidentified LEAKAGE that does not require Technical Specification entry.
Ams: lb l Exna level: lS l Cognitive level: l Comprehension l Esplanatio n ef Answer ~ KA: l 2.2.22 l RO Valae: l3.4 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Pressurizer Vapor Space Accident Title: KA Equipment Control Statement: Knowledge of limiting conditions for operations and safety limits.
Reference Reference Number Reference Section Page Number (s) Revision learn.
06) Beaver Valley -Unit i 1.14,3.4,6.2 l 3,3/4 4-13 Technical Specifications RCS LP SQS-6.5 Vll.A
4 8.g Qrestion Source - l New l Question Modincation Method l QIestion Source Comments: l __ M;terial Required for Examination: l l
l l Page 63
._ - ___ - __ _ - - - _-__ _ _ _ __ _ _ _ _ _ _ _ - _ - - _ - - _ _ _ - _ _ _ _ - _ __-_. _ - - - - _ _ - Question Tople: l Basis for use cf ADVERSE Cnmt v11uis Given the following conditions: + A LOCA has occurred + Containment pressure increased to 6.0 psig a : Containment radiation has increased to 1.5E+5 R/hr.
Ninety minutes later containment pressure decreases to 3.0 psig and containment radiation has decreased to 4E+4 R/hr. Integrated CNMT tadiation dose is 2.3E+5 Rads.
Which of the following describes whether the use of adverse containment parameters can be discontinued? a.
Use of adverse containment parameters can be discontinued.
b. Continued use of adverse containment parameters is required only due to the containment radiation readings.
. c.
Continued use of adverse containment parameters is required only due to the contaimnent pressure - conditions.
d. Continued use of adverse containment parameters is required due to both the containment pressure and radiation conditions.
Ams: la l Exam Level: lS l Cognitive Level: l Application l Esplanatio c af Answer KA: l 009 EK3.16 l RO Value: l3.8 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Small Break LOCA Title: ~ KA Knowledge of the reasons for the following responses as they apply to Small Break LOCA: Stat: ment: Containment temperature, pressure, humidity and level limits Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj Generic Instrumentation IOM 538.5.Gi-2 II.D 12-13 IssIB Rev 2 _EOP Generic issues LP-SQS-53.2 X.B.6. 8 22-23
15 i Question Source l New l Question Modification Method l Question Source Comments: l
Material Required for Subcooling Attachment 6-A Ex mination: I i I Page 64 - p, !. V _ _ __ - . _ _ _ _ - _ _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ -
Question Topic: l Ev:1 e f conditions for tripping RCPs Given the following conditions: l A LOCA has occurred
Containment pressure is 9.2 psig and lowering
RCS pressure has stabilized at 325 psig
Steam generator pressures are 800 psig and lowering
All ECCS equipment has responded as required
Which of the following describes when the RCPs should be tripped?
a.
Immediately b.
When the highest steam generator pressure reaches 700 psig.
c.
When the highest steam generator pressure reaches 525 psig.
d.
When the lowest steam generator pressure reaches 700 psig.
~ Ans: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio c of Answer KA: l 011 EA1.03 l RO Value: l4.0 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l1 System / Evolution Large Break LOCA Title: KA Ability to operate and / or monitor the following as they apply to Large Break LOCA: Statement: Securing of RCPs Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Trip Or S1 IOM 53.A.E-0 Foldout IssIB Rev 5 EOP Generic issues LP-SQS-$3.2 Terminal Obj.
Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 65
'
l ! l
Question Topic: l Determination of RCP/r: actor trip i i f Reactor power is 357o. Which of the following combinations ofloop flow conditions indicates that a reactor l trip should have occurred? ! a.
[FI-lRC-414] RCL. l A Flow indicates 80%. [FI-lRC#'q RCL IB Flow indicates 80%. l ' l l b. [FI-IRC-414] RCL 1 A Flow indicates 80%. i (FI-lRC-415] RCL 1 A Flow indicates 80%. c.
[FI-l RC-414] RCL 1 A Flow indicates downscale.
l [FI-lRC-435] RCL 1C Flow indicates 80%. d. [FI-lRC-414] RCL 1 A Flow indicates upscale.
[FI-1RC-415] RCL 1 A Flow indicates 80%. ' Ars: lb l Exam level: lS l Cognitive Level: l Memory l Explanatio a cf Answer , KA: l 015 AAl.03 l RO Value: l 3.7* l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1, System / Evolution Reactor Coolant Pump Malfunctions Title: , KA Ability to operate and / or monitor the following as they apply to Reactor Coolant Pump Malfunctions: Statement: Reactor trip alarms, switches, and indicators Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj R: actor Coolant System - 10M-6.4 IF lli.A
Iss 4 Rev Instrument Failure Procedure
Reactor Coolant System LP-SQS-6.5
5,6 Question Source l New l Question Modification Method l Qrestion Source Comments: l M:t: rial Required for Enmination: Page 66
_ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ - - _ _ _ _ - _ - _ _ _ - _ _ - _ _ - -. __ __ _ - Question Topic: l Failure cf makeup Giv:n the following conditions: - VOLUME CONTROL TANK LEVEL HIGH LOW (A3-53) has alarmed - [LI-lCH-115] Volume Control Tank Level (VB-A) failed offscale high Actual VCTlevel will: l a.
remain constant.
b. decrease until automatic makeup initiates.
c.
decrease until the charging pump suction transfers to the RWST.
d. decrease until the VCT is empty.
Ars: ld l Esam level: lS l Cognitive Level: l Application l Esplanatio o sf Answer KA: l 022 AA1.08 l RO Value: l3.4 l SRO Value: l3.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Reactor Coolant Makeup Title: , KA Ability to operate and / or monitor the following as they apply to Loss of Reactor Coolant Makeup: Statement: g VCTlevel Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Alarm - A3 53 VCT Level IOM-7.4.AAX PC 4,5 A 2-3 Iss 4 Rev High Low
CVCS - Instrumentation and IOM 7.1.D Auto M/U, LCVs 1-2, 8-9 Iss 4 Rev Controls
CVCS LP-SQS-7.1 Ill.D.2.b
2.g. 6.a Question Source l New l Question Modification Method l QIestion Source Comments: l Material Required for OM Figure 7-39 Examination: . l Page 67 L__----___-_____________-_--__-__-_____--___
- _ _ _ _- -. - - _ - - - _ _ _ _ _ _ _ - - - _ - _. _ - -.. - - _ - _ _ -. - - - _ _. - _ _ - _ - - _ _ - - _ - - - _ _ - - - . ! Question Topic: l Boration and SDM Tech Spect.
l i Given the following conditions: [1CH-P-2A] Boric Acid Transfer Pump is out of service
RCS Temperature is 420 *F
SDM is 167 delta K/K
S/D Banks are fully withdrawn l
if[lCH P-2B] Boric Acid Transfer Pump trips, HOW will required Technical Specification Shutdown Maryn be restored?. ! a.
BORATE, by gravity feeding the in-service Boric Acid tank to the blender.
b.
Emergency borate through the Emergency Boration valve [MOV-CH-350]. c.
Align the suction of the charging pump to the RWST.
d. Open the reactor trip breakers.
Ans: lc l Exam Level: lB l Cognitive Level: l Application l Explanatio o of Answer KA: l 2.1.12 l RO Value: l2.9 l SRO Value: l 4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1 ~ System / Evolution Emergency Boration Title: KA Conduct Of Operations i Statement: J Ability to apply technical specifications for a system.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj ~ Technical Specifications 3.1.1.1, 3.1.2.2, and 3.1.2.6 ! CVCS LP-SQS-7.1
11 , Question Source l New l Question Modification Method l _ Question Source Comments: l ~ Material Required for Technical Specifications Examination: Page 68
Question Topic: l Emerg;ncy Boration requirements Following a turbine load rejection, control rods are automatically inserted causing ROD CONTROL BANK D LOW LOW alarm (A4-124) to be received.
Which of the following is the required action by procedure? a.
Place the rods in manual and withdraw them until the alarm clears.
b. Place the rods in manual and allow temperature to stabilize.
c.
Emergency borate.
d. Borate via the normal flow path until the CONTROL BANK D LOW-LOW alarm clears.
A*s: lc l Esam Level: lS l Cognitive Level: l Memory l Explanatio a af Answer , , KA: l 2.4.31 l RO Value: l3.3 l SRO Value: l3.4 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Emergency Boration Title: KA Emergency Procedures / Plan Statement: ' Knowledge of annunciators alarms and indications, and use of the response instructions.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Emergency Boration 10M-7.4.S I
iss 4 Rev
Rod Control Bank D Low 10M-l.4.ABF
Iss 3 Rev Low I CVCS LP-SQS-7.1 10.p Q:estion Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination: . Page 69 - _ _ _ _
._-_-. - _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ - _ __ Question Topic: l Evt! ofloss cf RHR condition While operating at 175 *F and the RCS depressurized, the running RHR pump trips. The other RHR pump is available to be immediately started.
Which of the following describes when the other RHR pump should be started and the basis for this decision? The second RHR pump should be started: a.
immediately, to avoid any heatup of the RCS.
b. only after investigating the cause of the running pump trip, to avoid losing the second pump.
c. only after observing an RCS heatup, to avoid unnecessary starts of the RHR pump.
d. within five minutes, which is the most limiting time until boiling will occur.
Ans: lb l Exam Level: lS l Cognitive Level: l Memory l ' Esplanatio o cf Answer KA: l 025 AK1.01 l RO Value: l3.9 l SRO Value: l 4.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Residual Heat Removal System Title: KA Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Statement: Removal System: Loss of RilRS during all modes of operation R:ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Residuallicat Removal AOP 1.10.1 Caution
iss 3A System Loss Rev5 OM 53C-AOPs LP-SQS-53.C
4 Question Source l New l Question Modification Method l Qrestion Source Comments: l Miterial Required for Examination: Page 70
I Question Te ple: l Loss of CCW during a loss of power 1B and 1C Component Cooling Water Pumps [lCC-P-1B & IC] are BOTH racked to the Connect position on the DF bus.
Which of the following control switch positions describes when BOTH [lCC-P-IC] and [lCC-P-1B] will fail to restart on a D/G load sequence signal, following a DF bus undervoltage condition? ! a.
[1CC-P-1B]- After START, [lCC-P-1C]- After START l b.
[lCC-P-1B]- PULL-TO-LOCK, [1CC-P-1C]- After Start c.
[1CC-P-1B]- After STOP, [1CC-P-1C]- PULL-TO-LOCK d.
[1CC-P-1B]- After STOP, [1CC-P-1C]- After STOP ATs: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio o cf Answer KA: l 026 AA2.02 l RO Value: l2.9 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Loss of Component Cooling Water Title: KA Ability to determine and interpret the following as they apply to Loss of Component Cooling Water: Statement: The cause of possible CCW loss Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Reactor Plant Component IOM-15.1.D
issue 4 and Neutron Tank Cooling Rev i Water (CCRS) l Itactor Plant Component LP-SQS-15.1
5 l and Neutron Tank Cooling Water (CCRS) , l Qrestion Source l New l Question Modification Method l Q:estion Source Comments: l M:terial Required for j Examination: l Page 71
_ _ __ - -__ ___- -_ - _ _ _ _ - _ _ -. _ _ _. - _ - - - _ - - _ _ _ - _ _ _ _ _ - _ _ - - _ _ _ _ _ _ .. Question Topic: l Effect cf r;ference leg br;ak Given the following conditions: Reactor power - 100%.
A leak develops on the reference leg for the controlling Pressurizer level sensor
l How will charging flow respond over next five minutes? Charging flow will: ) a.
decrease to the minimum value.
! b. decrease and then return to the initial value, c.
increase to makeup for the loss through the leak.
d. increase to the maximum flow value.
Ars: la l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio a rf Answer ] KA: l 028 AK1.01 l RO Value: l 2.8' l SRO Value: l3.1 l Section: l EPE l RO Group: l 3 l SRO Group: l3 System / Evolution Pressurizer Level Control Malfunction Title: KA Knowledge of the operationa! implications of the following concepts as they apply to Pressurizer Le al Control St-tement: Malfunction: PZR reference leak abnormalities Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Patssurizer and Pressure LP-SQS-6.4 1.D.I.f 9-10
14 Rr, lief System Reactor Coolant System - lOM-6.4.lF
4
Instrument Failure Procedure Question Source l New l Question Modification Method j Qrestion Source Comments: l Miterial Required for Ex:mination: i Page 72 - i !' ! - - _ _ .-______________________-________-__-________L
_ _ _ - _ - _. _ . _ _ _ _ - _ l Question Topic: l AFW Actuation due to AMSAC ' Given the following conditions: Reactor power - 100%
Both feedwater pumps trip { e The reactor fails to trip '
) i Which of the following describes when AMSAC should trip the turbine? , a.
Immediately after the feedwater pumps trip.
b. Immediately after feedwater flow decreases below 25% flow.
j c.
150 seconds after the feedwater pumps trip.
d. 25 seconds after feedwater flow decreases below 25% flow.
I Ans: ld l Exam Level: lS l Cognitive Level: l Memory l 'Explanatio e of Answer KA: l 029 AA2.09 l RO Value: l4.4 l SRO Value: l4.5 l Section: l EPE j RO Group: l 2 l SRO Group: l1 System / Evolution Anticipated Transient Without Scram Title: KA Ability to determine and interpret the following as they apply to Anticipated Transient Without Scram: Statement: Occerrence of a main turbine / reactor trip Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj ATWS Mitigation System 10M-458.1.B 1,2 1ss 4 Rev Actuation Circuitry
AMSAC LP-SQS-45.2 II.D.2.e
1
Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 73
l f Question Topic: l Evaluation of SR NIS voltage failure What would be the plant response to the following conditions? f o The plant is operating at 100% power call systems are NSA oThe "A" train Source range RESET /BT OCK switch is inadvertently turned to the BLOCK position.
a.
The reactor would trip, and N31 SR would energize b. The reactor would not trip, and N31 SR would not energize.
I c.
The reactor would trip, and N31 SR would not energize d. The reactor would not trip, and N31 SR would energize Ars: lb l Exam Level: lS l Cognitive Level: l Application l ' Explanatio n cf Answer l l KA: l 032 AKl.01 l RO Value: l2.5 l SRO Value: l3.1 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Source Range Nuclear Instrumentation Title: KA Knowledge of the operational implications of the following concepts as they apply to Loss of Source IUmge St:tement: Nuclear Instrumentation: Effects of voltage changes on performance RIference Reference Numher Reference Section Page Number (s) Revision Learn.
Obj UFSAR fig. 7.2 sheet 3
Reactor Excore instrument 10M 2.1.c lit.E.3
Iss 4 Rev System I Question Source l New l Question Modification Method l Question Source Comments: l M:terial Required for UFSAR fig. 7.2 sheet 3 Examination:
( I Page 74 L - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- --- - _ - _ _ -_
i Question Tople: l Eval of failed IR channel on SU Given the following conditions: Plant startup is in progress.
- All power range channels indicate 6% reactor power.
- Intermediate channel N-36 fails HIGH.
- Reactor power remains at 6%.
Which of the following describes required operator actions? a.
Initiate a reactor trip, enter E-0, and FR-S.I.
b.
Immediately commence a controlled reactor shutdown.
c.
Raise power to greater than PIO and block both intermediate ranges.
d.
Continue power operations.
A~s: Ib l Exam Level: lS l Connitive Level: l Memory' l Explanation of Answer KA: l2.1.1 l RO Value: l3.7 l SRO Value: l3.8 l Section: lEPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Intermediate Range Nuclear Instrumentation Title: KA Conduct Of Operations St'tement: Knowledge of conduct of operations requirements.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Excore Instrumentation LP-SQS-2.1 V.C.3.c & c 16-17
5,8,12 System Conduct of Operations 1/20M-48.1.B VI.H.S
iss 3 Rev
l Conduct of Operations I/2LP SQS-48.1
QTestion Source l New l Question Modification Method l Question Source Comments: l l ! Meterial Required for l Et mination: Page 75 - _ _ - _ _ _ _ _ - _ _ _ _ _ - _ _ - _ - _ - _ _ _ _ _ _ _ _ - . _ _ _ _ _ - - _ - - _ _ .-
. - - - _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ - _ _ _ --. Question Tepic: l Fuel Handling accident syst:ms risponse A fuel assembly was suptured during movement in the fuel building.
Which of the following describes how the fuel building evacuation alarm is actuated? a. The alarm must be manually initiated from the control room.
b. [RM-1RM-206] and [RM-1RM-207] Fuel Pool Bridge Area Monitors will sound the evacuation alarm.
c.
[RM-IVS-103A, B] Fuel Building Ventilation Exhaust monitors will sound the evacuation alarm.
i d. The alarm must he manually initiated from either the fuel building or the control room.
' A!s: lc l Exam Level: lS l Cognitive Level: l Memory l Esplanatio o cf Answer , ~KA: l 036 AA2.02 l RO Value: l3.4 l SRO Value: l4.1 l Section: l EPE l RO Group: l 3 l SRO Group: l3 System / Evolution Fuel Handling incidents Title: KA-Ability to determine and interpret the following as they apply to Fuel Handling Incidents: Statement: ) Occunence of a fuel handling incident
R,ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Irradiated Fuel Damage AOP 1.49.1 C.2
1ss 3A Rev 3 OM 53C-AOPs LP-SQS-53C.1
6 Q estion Source l Facility Exam Bank l Question Modification Method i QYestion Source Comments: l Material Required for Examination:
Page 76 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _
i Questio2 Tepic: l R:sponse of SG 1:ak detection monitors At what power level will the steam generator leakage N-16 Radiation Monitors [RM-MS-102A,B, & C] BEGIN to provide valid leak rates, in GPD7 a.
5% b. 20% c.
30% d. 50 % Ars: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio Def Answer KA: l 037 AAl.06 l RO Value: l 3.8* l SRO Value: l 3.9* l Section: l EPE l RO Group: l 2 l SRO Group: l2
' System / Evolution Steam Generator Tube Leak Title: KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Leak: Statement: Main steam line rad monitor meters R'.ference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Radiation Monitoring 10M-43.1.C
Iss 4 Rev Systems - Major components
OM $3C-AOPs LP SQS-53C.I
QIestion Source l Facility Exam Bank l Question Modification Method l Q estion Source Comments: l M:terial Required for Examination: Page 77
Question Topic: l Evaluation of coo:down temperature /cooldown ' Given the following conditions-A Steam Generator Tube Rupture has occurred
E-3, Steam Generator Tube Rupture, is being performed . i The RCS has been cooled down to the target temperature.
In order to maintain RCS subcooling, intact steam generator pressure must be maintained: a.
greater than the ruptured generator.
b. equal to the ruptured generator.
, c.
greater than the saturation pressure of the RCS.
d. less than the ruptured generator.. Ans: ld l Exam Level: lS l Cognitive Level: l Application l Explanatio, c cf Answer KA:, l 038 EA1.36 l RO Value: l4.3 l SRO Value: l 4.5 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Steam Generator Tube Rupture Title: KA Ability to operate and / or monitor the following as they apply to Steam Generator Tube Rupture: Statement: Cooldown of RCS to specified temperature Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Steam Generator Tube
iss 1B Rupture Background Rev 5 EOPs LP-SQS 53.3
Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M terial Required for Examination: l l . Page 78
' Question Topic: l Ev;luation c.f FW condition Given the following conditions: A steam break has occurred on SG "A"
l A reactor trip was manually initiated
I A Si has NOT been initiated o ' No operator actions have oeen performed on the feedwater system.
o Only SG "A" narrow range level has decreased below 12%.
RCS T,., are (A) 542 F,(B) 550 'F,(C) 550 F
i Which of the following is the expected status of feedwater? l a.
The feedwater regulating valves will be shut. The Turbine Driven AFW pump will be running.
b. The feedwater regulating valves will be shut. All AFW pumps will be running.
c.
A complete FWI isolation will be initiated. All AFW pumps will be running.
d. The feedwater system will be in the same lineup as prior to the reactor trip, except the FRVs will be throttled closed.
Ans: la l Exam Level: lS l Cognitive Level: l Application l Explanatio ocf Answer KA: l 040 AA1.02 l RO Value: l4.5 l SRO Value: l4.5 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Steam Line Rupture Title: KA Ability to operate and / or monitor the following as they apply to Steam Line Rupture: Statement: Feedwater isolation Reference Reference Numher Reference Section Page Number (s) Revision Learn.
Obj SG Feedwater System - 10M-24.lD Feedwater Isolation 2, 6 iss 4, instrumentation and Controls Rev.2 Reactor Protection System LP-SQS-1.1 V.E.5
6
~ Question Source l New l Question Modification Method l Q'estion Source Comments: l Material Required for Examination:
i I Page 79 l
Q::estio3 Topic: l Effect & mitigation techniques Given the following conditions: An uncontrolled depressurization of all steam generators has occurred
Current RCS cooldown rate is 125 F/hr - Which of the following describes how drying out, of the steam generators, is avoided while trying to limit cooldown rate? a.
A minimum AFW flow to all steam generators is maintained.
b. SGs are intermittently fed to assure that a wide range levels remain above 10%. c.
Only reducing AFW flow as necessary to reduce the cooldown rate to less than 100 F.
d. AFW feed rate is limited to maintain constant level, provided the level is above 10% wide range.
] Ans: Ia l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n of Answer KA: l 040 AKl.07 l RO Value: l3.4 l SRO Value: l4.2 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Steam Line Rupture Title: KA Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: Statement: Effects of feedwater introduction on dry S/G Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Uncontrolled ECA-2.1 STEP 6
issIB, , Depressurization of all SGs rev.4 Uncontrolled 10M-538.4.ECA-2.1 IV.6
issIB; Depressurization of all SGs Rev.4 Background EOPs LP-SQS-53.3
Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 80
__- __ _ ____ _ _ _ _ - _- -- Question Topic: l Block of steam dumps on turbine trip A loss of condenser vacuum has occurred due a leak in the condenser. Main Condenser Steam Dumps are open following a turbine trip.
As vacuum decreases, at what condenser vacuum will Main Condenser Steam Dumps close? a.
25" Hg Vacuum b. 20" Hg Vacuum c.
10" Hg Vacuum d.
5" lig Vacuum Ans: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio n of Answer KA: l 051 AK3.01 l RO Value: l2.8* l SRO Value: l 3.I' l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Loss of Condenser Vacuum Title: KA Knowledge of the reasons for the following responses as they apply to Loss of Condenser Vacuum: Statement: Loss of steam dump capability upon loss of condenser vacuum h Reference Reference Nuinber Reference Section Page Number (s) Revision Learn.
Obj Main Steam Supply / Steam LP-SQS-21.1
4
i ! Dump System 10M-26.2.B
4
' Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: _ l l i ! Page 81 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ -
I i Question Tople: l D. termination cf Feedline break ' A break has occurred on the feedwater line to SG "A" downstream of [MOV-FW-156A], Main Feed Line Containment Isolation valve. Containment pressure increases to the SI setpoint.
Following the reactor trip and SI, which of the following SG pressure indications would exist? a.
Only SG "A" pressure would be decreasing from the break.
b. All SG pressures would be decreasing from the break via the main steam lines.
c.
All SG pressures would be decreasing from the break via the main feedwater lines.
d. All SG pressures would be decreasing fro:n the break via the auxiliary feedwater lines.
A*s: la l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio e cf Answer , KA: l 054 AKl.01 l RO Value: l4.1 l SRO Value: l4.3 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Main Feedwater Title: KA Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater: Stat ment: MFW line break depressurizes the S/G (similar to a steam line break) Reference Reference Number Reference Section l Page Number (s) Revision Learn.
Obj M in Steam Supply / Steam LP-SQS-21.1
1,4g Dump System Miin Steam System I OM-21.1.C
iss 4 Rev
VOND 24-1 Q estion Source l New l Question Modification Method l Q:estion Source Comments: l Mit: rial Required for Ex:mination: Page 82 _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ _ - _ _ -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-_- Questi:a Tcpic: l Load required to be left in AUTO l A loss of all 4KV busses has occurred. ECA-0.0 has been implemented to the point of placing deenergized l equipment in PULL TO LOCK. The iDF emergency bus has been selected to cross tie to Unit 2.
Which of the following l AE Emergency Bus loads shall remain in the AUTO position and the basis for leaving that pump in AUTO? a.
Reactor River Water Pump to assure that the diesel has cooling upon startup.
b.
Charging Pump to restore seal flow.
c.
Charging Pump to restore Pressurizer level.
d. Component Cooling Water Pump to restore cooling to the thermal barrier.
Avs: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio a cf Answer KA: l 2.4.20 l RO Value: l3.3 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Station Blackout Title: KA Emergency Procedures / Plan St:tement: Knowledge of operational implications of EOP warnings, cautions, and notes.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Loss of All Emergency 4KV 10M-53 A. I.ECA-0.0 Caution Step 14
1ss1B AC Power Rev 4 Emergency Operating LP-SQS-53.3
3 Procedures Qrestion Source l Facility Exam Bank l Question Modification Method l QIestion Source Comments: l M terial Required for Examination: Page 83 . _ - _.
Question Topic: l Purpose of Si R set If an SI actuation signal is received when performing ECA-0.0, " Loss of All Emergency 4KV Power", the SI signal should be: a. ' reset to prevent lockout of the stub busses.
b. reset to permit manual loading of equipment of an Emergency bus.
allowed to remain active to ensure rapid injection of core cooling water when power is restored.
c.
d allowed to remain active to ensure the load sequencer re-initiates when the DG starts.
Ass: lb l Exam Level: lS l Cognitive Level: l Memory l Explanatio e cf Answer KA: l 055 EK3.02 l_ RO Value: l4.3 l SRO Value: l4.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Station Blackout Title: KA Knowledge of the reasons for the following responses as they apply to Station Blackout: St::tement: Actions contained in EOP for loss of offsite and onsite power Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Loss of All Emergency 4KV ECA-0.0 steps 31 & 37 22 &25 iss 1B; AC Power Rev 4 Loss of All Emergency 4KV 10M-53 B.4.ECA-0.0 Step 31, Basis 127 Iss IB; AC Power Background Rev 4 EOPs LP-SQS-53.3
Q estion Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l M terial Required for Examination: . Page 84 __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
_. _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ Question Teple: l RCS temperatures What is the expected response of RCS Ilot and Cold leg temperatures during the first few minutes following a reactor trip froml00% power COINCIDENT with a loss of offsite power? Ilot leg temperatures will rise, and Cold leg temperatures will remain relatively constant, until a.
natural circulation flow is established.
b. Ilot leg temperatures and Cold leg temperatures will both rise, until natural circulation flow is established.
Ilot leg temperatures will remain relatively constant and Cold leg temperatures will drop, until c.
natural circulation flow is established.
d. Ilot leg temperatures will rise and Cold leg temperatures will drop, until natural circulation flow is established.
ATs: la l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KAt l 056 AA2.18 l RO Value: l3.8 l SRO Value: l4.0 l Section: l EPE l RO Group: l 3 l SRO Group: l3 lystem/ Evolution Loss of Off-Site Power Title: KA Ability to determine and interpret the following as they apply to Loss of Off-Site Power: St:tement: __ Reactor coolant temperature, pressure, and PZR level recorders Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj R: actor Trip Response ES-0.1 Note before step 3 3-4 Iss lil Rev 4 EOPs LP-SQS-53.3
6 l Q:estion Source l Facility Exam llank l Question Modification Method l Q:estion Source Comments: l Miterial Required for Examination:
! Page 85 l - - _ - - - _ _ -. - _ - - - - _ - - - _ _ _ \\
I-Question Topic: l Effect of a loss of Vital AC on Feedw:ter Given the following conditions: l Reactor power is 74%
Feedwater control is in automatic
Loss of a single 120 VAC Vital bus has occurred
Which of the following describes the expected response of Main Feedwater Regulating Valves which do NOT remain in AUTO? - a. The FRVs willimmediately fail cpen.
b. The FRVs will immediately fail closed.
c.
The FRVs will drift shut.
d. ' The FRVs will transfer to either MANUAL or AUTO HOLD.
As: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio ocf Answer KA: l 057 A A2.19 l RO Value: l4.0 l SRO Value: l4.3 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Loss of Vital AC Instrument Bus Title: KA Ability to determine and interpret the following as they apply to Loss of Vital AC Instrument Bus: Statement: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Alarm VitalIlus I,II,111,IV 10M-38.4.AAA, AAC,
Trouble AAE. AAG 120V AC Distribution LP SQS 38.1
6
System Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 86
- -__ Question Tcpic: l Effect of a loss of DC on RCPs l Which of the following is the effect that losing 125 VDC Bus I will have on the Reactor Coolant Pumps? l a, One or two RCPs will trip on undervoltage, b. One or two RCP breakers will ONLY be able to be tripped using the mechanical trip at breaker.
l l c. Component cooling water will be lost to all RCPs.
d.- Seal water flow to the RCPs will be isolated.
Ans: lc l Eram Level: lS l Cognitive Level: l Application l Esplanatio oaf Answer KAt. l 058 AA2.03 l RO Value: l3.5 l SRO Value: l3.9 l Section: l EPE l RO Group: l 2 l SRO Group: l2 SystIm/ Evolution Loss of DC Power Title: KA Ability to determine and interpret the following as they apply to Loss of DC Power: St:tement: DC loads lost; impact on to operate and monitor plant systems Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj OM 39 10M-39. 5.B.6 Table 39-6 all iss 4 Rev $ QIestion Source lNew l Question Modification Method l Q:estion Source Comments: l M: trial Required for IOM-39.5.B.6(28 pages) Enmination: Page 87
Question Topic: l Evalef Tech Spec Given the following conditions: Unit 1 is in MODE 6 e Unit 2 is in MODE 1
- Movement ofirradiated fuel is ongoing in the Unit 1 Containment only
- Monitor RM-1RM-218A Control Room Area - Unit I has failed low What action is required for the above conditions?
a.
No action is required because the monitor is not required to be operable, b. Within ONE hour the respective Unit 2 control room monitor train shall be verified operable.
Within ONE hour, verify that Control Room Area - Unit I monitor [RM-1RM-218B] is operable.
c.
d.
Within ONE hour, suspend all operations involving movement ofirradiated fuel.
A s: lb l Exam Level: lS l Cognitive Level: l Application l Explanatio ' o e f Answer KA: l 061 AA2.06 l RO Value: l3.2 l SRO Value: l4.1 l Section: l EPE l RO Group: l 2 l SRO Group: ~l2 System / Evolution Area Radiation Monitoring System Title: KA Ability to determine and interpret the following as they apply to Area Radiation Monitoring System: Statement: Required actions if alenn channel is out of service Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Be;ver Valley - Unit i 3.3.3.1, Table 3.3-6,1.c, 3/4 3-33-3-35 Amend Technical Specifications Action 41 119 Radiation Monitoring System LP-SQS-43.1 VI.A
4 7.a Qyestion Source l New l Question Modification Method l Q:estion Source Comments: l Material Required for Tech Specs Examination: Page 88
._ _ ^ Question Topic: l EfYect cf restoring air using IIA-90.
During a loss of containment air, which of the following is the possible effect of opening [llA-90] Instrument Air to Containment Air Isol Valve too quickly? a.
Station Air compressor trips I b. CVCS letdown isolation c.
SG Main FW Feed Reg Vivs failing open d. Main Steam Line Trip Valve closure Ass: ld l Exam Level: lS l Cognitive Level: l Memory l Explanatio o cf Answer KA: l 065 AK3.08 l RO Value: l3.7 l SRO Value: l3.9 l Section: l EPE l RO Group: l 3 l SRO Group: l2 System / Evolution Loss ofinstrument Air Title: KA Knowledge of the reasons for the following responses as they apply to Loss ofinstrument Air: Statement: Actions contained in EOP for ' ass ofinstrument air Reference Reference Numaer Reference Section Page Number (s) Revision Learn.
Obj Loss of Containment AOP 1.34.2 Caution before step 4
iss 3A Instrument Air Rev 3 OM $3C-AOPs LP-SQS-53C.l
4 Question Source l New l Question Modification Method l Qrestion Source Comments: l Miterial Required for Examination: i . Page 89 . _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ ..
_ _ _ _ _. _ - _ - = _ _ _ _ _ _ _ - _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _. _- __ _ ___ - - -.. _ _ _ _ - _ - _ _.. _ _ _ _ _ _ _ _ _ _ _
Question Tepic: l Type of detection / extinguishing eqpt for use ' Which of the following describes the fire protection afforded for the primary process rack area? j a.
Carbon Dioxide is released to the area by manual actuation only.
b.
Carbon Dioxide is released to the area by automatic actuation of smoke detection or by manual actuation.
c.
Halon is released to the area by manual actuation only.
d.
Halon is released to the area by automatic actuation of smoke detection or by manual actuation.
l Ars: ld l Exam Level: lS l Cognitive Level: l Memory l Espinnatio cef Answer KA: l 067 AA1.08 l RO Value: l3.4 l SRO Value: l3.7 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Plant Fire on Site Title: KA.
Ability to operate and / or monitor the following as they apply to Plant Fire on Site: Statement: Fire fighting equipment used on each class of fire Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Fire Protection System - IOM-33.1.B flalon paragraphs 1 & 4
iss 4; Summary Description Rev.3 Fire Protection System LP-SQS-33.1 E. l b.4
5 3.e Qrestion Source l New l Question Modification Method l Qrestion Source Comments: l M;terial Required for Ermination: Page 90 _ _ _ _ _ . _ _ _ - _ _ _ _ - - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ - _ -
Question Tepic: l Pressurizer level control A fire in the control room has resulted in control room evacuation. Plant control has been transferred to local control panels as requir.:d by lOM-56C.1, Altemate Safe Shutdown from Outside the Control Room.
Until a cooldown is initiated from the BIP, pressurizer level is maintained by charging via: a.
[MOV-RC-556A, B, C) Reactor Coolant Loop Fill Valves to the RCS loops.
b. the normal charging connection.
c.
the RCP seals.
d. the BIT.
A7s: lc l Exam level: lS l Cognitive Level: l Memory l Explanatio n cf Answer KA: l 068 AA1.30 l RO Value: l3.4 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1
System / Evolution Control Room Evacuation Title: KA Ability to operate and / or monitor the following as they apply to Control Room Evacuation: St:tement: Operation of the letdown system Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Alternate Safe Shutdown LP-STA-56C.1 VI.A.3
2 6.a from Outside the Control Room _ Question Source lNew l Question Modification Method l Q estion Source Comments: l M terial Pequired for Examination: Page 91
Ouestion Tepic: l Controller locition Which of the following identifies the components used by the operator stationed at the BIP (Backup Indicating Panel) to lower pressurizer level? a.
[SOV-1RC-102B] RCVS Reactor Vessel Vent Viv [SOV-1RC-103B] RCVS Pressurizer Vent Viv [SOV-1RC-105] RCVS Vent to Containment Isolation Viv b. [LCV-1CH-460A and B] Ltdn to Regen Hx Isol [TV-CH-200B] 60 GPM Ltdn Orifice Cnmt Isol Viv Letdown will flow to the degasifier via [LCV-115A], which has failed to the degasifier position.
I c.
[MOV-CII-201] Excess Ltdn HX Inlet Isolation Viv ! [MOV-lCH-137] Excess Ltdn HX Flow Control Viv d.
[PCV-1RC-455D] PZR PORV Relief Viv [PCV-IRC-456] PZR PORV Relief Viv ~ , Ans: la l Exam Level: lS l Cognitive level: l Memory l Explanatio ecf Answer KA: l 068 AK2.01 l RO Value: l3.9 l SRO Value: l4.0 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Control Room Evacuation Title: KA Knowledge of the interrelations between Control Room Evacuation and the following: St:tement: Auxiliary shutdown panellayout Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Misc. Safety-Related 10M-45.1.B (BIP) Indications
1ss 4 Rev Systems Summary
Description Alternate Safe Shutdown LP-STA 56C.1
2
Outside the Control Room R: actor Coolant System - lOM-6.1.D 5-6 Instrumentation and Controls Qyestion Source l New l Question Modification Method l QIestion Source Comments: l M terial Required for Examination: Page 92 ....
. _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _. __- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ Question Topic: l Basis f r starting rn RCP
An RCP is started in FR-C.1, " Response to Inadequate Core Cooling", in order to: a.
allow using RVLIS Dynamic Range indication to determine core void content.
b.
temporarily improve core cooling until some form of makeup flow to the RCS can be established.
c.
enhance the cooling caused by rapid depressurization of the steam generators.
d.
establish pressurizer spray flow to reduce RCS pressure to cause low pressure systems to inject.
Ans: Ib l Esam Exvel: IS l Cognitive Level: I Comprehension l Explanation of Answer i l KA: l 074 EK2.01 l RO Value: l3.6 l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution inadequate Core Cooling Title: KA Knowledge of the inter;lations between Inadequate Core Cooling and the following: Statement: RCP Reference Reference Number Reference Section l Page Number (s) Revision Learn.
Obj Response to inadequate Core 10M 538.4.FR-C.I I Iss IB Cooling Background Rev4 Emergency Operating LP-SQS-53.3
3 Procedures Question Source l Facility Exam Bank l Question Modification Method l Question Source Comments: l M:terial Required for Examination: l Page 93
l Question Topk: l Actions to lower R/A levels Given the following conditions: Reactor power hasjust been raised from 20% to 100%
Dose Equivalent Iodine hasjust been reported as 5.0 ci/ gram.
- Which of the following explains why operation can continue with Dose Equivalent Iodine above the Technical Specification LCO limit?
a.
To allow for CVCS removal of the crud released by the power change.
b. The Technical Specification LCO limit is conservative enough, to allow extended periods (> 7 days) of exceeding the limit.
c. To accommodate the iodine that was released during the power change.
d. The probability of a Large break LOCA occurring during the time period Iodine is above the limit, presents an acceptable risk.
Ans: ic l Exam Level: lS l Cognitive Level: l Memory l Explanatio c ef Answer KA: l 076 AK3.05 l RO Value: l2.9 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution liigh Reactor Coolant Activity Title: KA Knowledge of the reasons for the following responses as they apply to High Reactor Coolant Activity: Statement: Corrective actions as a result of high fission-product radioactivity level in the RCS Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Technical Speci0 cations LP-SQS-TS
4 Bc:ver Valley Unit i Bases 3/4 4-4 83/44-4 Amend No 102 Question Source l NRC Exam Bank l Question Modification Method l Qrestion Source Comments: l M;terial Required for Extmins*lon: Page 94
Question Topic: l Securing Si flow Which of the following describes the required subcooling requirements before ternt nating SI in ES-1.1, S1 i Termination? The required subcooling: a.
is based on saturation conditions plus instrument errors, b. is based on the expected pressure after SI is terminated.
c.
is based on the expected temperatures after SI is terminated.
d. provides for a 50 F margin to saturation to avoid reinitiation.
A's: la l Enam Level: lS l Cognitive Level: l Memory l Esplanatio o ef Answer _KA: l E02 EK3.2 l RO Value: l3.3 l SRO Value: l3.8 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution S1 Termination Title: KA Knowledge of the reasons for the following responses as they apply to S1 Termination: Statement: Normal, abnormal and emergency operating procedures associated with (S1 Termination).
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj SI Termination /Reinitiation IOM053B.5.Gl-11 II.A.1
issIB Rev1 EOP Generic issues LP-SQS-53.2 i LO 3 Question Source iNew l Question Modification Method l Q estion Source Comments: l Miterial Required for Examination: i , Page 95 _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _
- ___ Question Topic: l Basis for required Pressmaer Level A reactor trip and SI have occurred, and the control room operators are responding to a small-break LOCA.
All RCPs are tripped. The operators have proceeded to the recovery stage in ES-1.2, " Post-LOCA Cooldown and Depressurization". A PZR PORV is used to depressurize the RCS until PZR level is greater than 18% [50% ADVERSE CONTAINMENT]. In addition to ensuring that RCS conditions are under adequate operator control, the basis for this pressurizer level ensures: a.
that a reduction in subcooling does not occur when SI flow is reduced.
b.
sufficient inventory such that PZR level does not drop low when an RCP is started.
pressurizer level indication is not due to a void in the vessel head.
c.
d. adequate PZR stearn space to absorb pressure fluctuations during RCP start.
Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n of Answer KA: l E03 EK2.2 l RO Value: l3.7 l SRO Value: l4.0 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution LOCA Cooldown and Depressurization Title: KA Knowledge of the interrelations between LOCA Cooldown and Depressurization and the following: Statement: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of tL facility.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Post LOCA Cooldown and ES-1.2 step 15
issIB Depressurization Rev5 Post LOCA Cooldown and l OM-53 B.4.ES-1.2
issIB Depressurization Rev 5 EOP Generic issues I LP-SQS-53.2 II.B.1, Ill.A 5,10 3, 4 Question Source l NRC Exam Bank l Question Modification Method l Question Source Comments: l Material Required for Examination: l Page 96
Question Topic: l Purpose of ECA 1.2 Given the following conditions: o A small break LOCA has occurred due to a break at some unknown location outside containment.
o Performance of ECA - 1.2 "LOCA Outside Containment" did not isolate the break.
o At the completion ECA - 1.2 "LOCA Outside Containment", RCS pressure is still dropping At the conclusion of ECA - 1.2 "LOCA Outside Containment" the operating crew should transition to a.
E-0 "Rx Trip or SI" in order to reverify that all automatic actions have been completed.
b. ' E-3 "SGTR", since there are adequate steps within this procedure to deal with these conditions.
c.
ES-0.0 "Rediagnosis" in an attempt to diagnosis the break location.
d. ECA-l.1 " Loss of Emergency Coolant Recirculation", in order to deal with the loss of available inventory for core cooling.
A s: ld l Exam Level: lS l Cognitive Level: l Comprehension l Esplanatio o cf Answer ~ K- " 204 EK2.2 l RO Value: l3.8 l SRO Value: l4.0 l Section: lEPE l RO Group: l 2 l SRO Group: l1 System / Evolution LOCA Outside Containment Title: KA Knowledge of the interrelations between LOCA Outside Containment and the following: Statement: Facility's heat removal systems, including primary coolant, emergency coolant. the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj LOCA Outside Containment IOM-53B.ECA-1.2
IssIB Background Rev 3 Emergency Operating LP-SQS-53.3
1 Procedures Q estion Source l New l Question Modification Method j Qrestion Source Comments: l M;terial Required for Examination: s Page 97
. _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ Questics Tcpic: l Apply procedural direction for cooldown ) l During a natural circulation cooldown with RVLIS unavailable, it is likely that voids will form in the upper ' head region. ES-0.4 " Natural Circulation Cooldown With Steam Void in the Vessel (Without RVLIS)", limits the size of these volds in the RCS head ri.gion by : a.
Requiring all CRDM fans to be runnung.
b. Limiting the allowable increase in pressurizer level.
. Limiting the maximum temperature on Core Exit Thermocouple.
c.
l d.
Requiring a minimum of 200F subcooling.
, l Ans: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio a cf Answer KA: l E10 EA2.2 l RO Value: l3.4 l SRO Value: l3.9 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution Natural Circulation with Steam Void in Vessel with/without RVLIS Title: KA Ability to determine and interpret the following as they apply to Natural Circulation with Steam Void in Vessel StItement: with/without RVLIS: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
Ref;rence Reference Number Reference Section Page Number (s) Revision Learn.
Obj N:tural Circulation ES-0.4 step 9
Iss1B Cooldown With Steam Void Rev 4 in Vessel (Without RVLIS) EOPs LP-SQS-53.3
Qr.',stion Source l Facility Exam Bank l Question Modification Method l Qxestion Source Comments: l M;terial Required for Exr.mination: Page 98 _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ -
_ _ _ _ - _ - _ - _ _ _ _ _ - _ - _ _ _ - - _ _ _ - _ _ - _ - _ - - - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ - f l Question Topic: l Condition resulting in loss ef recirc ' Given the following conditions: A LOCA has occurred
Due to low RWST level a transfer to Cold Leg Recirculation has occurred.
- All automatic actions for the transfer to Cold Leg Recirculation are
complete.
' - * [ISI-P-1B] LHS1 pump is not available .
- Containment pressure - 12.4 psig Which of the following would result in a loss ofinjection flow?
a.
RCS pressure - 450 psig [MOV-ISI-862A] 1 A LHSI Pump RWST Suct Viv fails open b. RCS pressure -250 psig [MOV-1SI-863A] 1 A LHS1 to Chg Pumps Sup Viv fails closed . c.
RCS pressure - 380 psig (CH-P-1 A] 1 A Charging /HHSI Pump trips [MOV-ISI-863B] 1B LHSI To Chg Pumps Sup Valve fails closed.
d. RCS pressure - 180 psig [MOV-ISI-885A] 1 A LHSI PP Mini Flow Isol Valve fails open A s: lb l Exam Level: lS l Cognitive Level: l Comprehension l Explanatio n ef Answer , KA: l ElI EA2.1 l RO Value: l3.4 l SRO Value: l4.2 l Section: l EPE l RO Group: l 2 l SRO Group: l2 System / Evolution Loss of Emergency Coolant Recirculation Title: KA Ability to determine and interpret the following as they apply to Loss of Emergency Coolan: Recirculation: Statement: Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Transfer To Cold Leg ES 1.3 step 4
issIB Recirculation Rev 4 EOP Attachment 1-0 IOM-53 A.I.1-G step 2
IssIB Rev 2 EOPs LP-SQS 53.3
Question Source l New l Question Modification Method l Question Source Comments: l Material Required for Examination: Page 99 _ _ _ _ _ _ ______ _ _ _ - _ -
Question Topic: l CIB setpoints How long after a CIB signal is received will the quench spray and containment spray pumps start? [QS-P-1 A,B] Quench Spray pumps - 5 seconds a.
[lRS-P-2A, B] Outside Recirc Spray Pumps = 120 seconds [lRS-P-I A, B] Inside Recirc Spray Pumps = 225 seconds b.
[QS-P-1 A,B] Quench Spray pumps - 60 seconds [lRS-P-l A) Inside Recirc Spray Pump, [lRS-P-2B] Outside Recirc Spray Pump = 120 seconds [lRS-P-1B] Inside Recirc Spray Pump, [lRS-P-2A] Outside Recire Spray Pump = 210 seconds c.
[QS-P-1 A,B] Quench Spray pumps - 60 seconds [1RS-P-1 A, B] Inside Recirc Spray Pumps = 210 seconds [lRS-P-2A, B] Outside Recirc Spray Pumps = 225 seconds d.
[QS-P-1 A,B] Quench Spray pumps - 5 seconds [lRS-P-1 A] inside Recire Spray Pump, [lRS-P-2B] Outside Recirc Spray Pump = 210 seconds [lRS-P-1B] Inside Recirc Spray Pump, [lRS-P-2A] Outside Recire Spray Pump = 225 seconds Ans: Id I Esam Level: IS l Cognitive Level: l Memory l Explanation of Answer KA: l E14 EKl.3 l RO Value: l3.3 l SRO Value: l3.6 l Section: l EPE l RO Group: l 1 l SRO Group: l1 System / Evolution liigh Contaimnent Pressure Title: KA Knowledge of the operational implications of the following concepts as they apply to High Contaimnent Pressure: Setement: Annunciators and conditions indicating signals, and remedial actions associated with the (High Containment Pressure).
Reference Reference Number Reference Section Page Number (s) Revision Learn.
Obj Containment IOM 13.2.B
Iss 4 Rev Depressurization System
Containment LP-SQS-13.01
5
Depressurization System Question Source l New l Question Modification Method l Question Source Comments: l Material Required for { Examination:
Page itxt _ _ _ _ _ _ _ - - -. _ _ _ - -. _ - _ _ - - _ - - _
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ - - - _ _ - - _ _ - _. _ _ __. - __ __ _ _ _-___ _ _ _ - _ Senior R=ct:r Operet:r Answ:r Key 26. d 1. o 2. a 27. c 3. d-28. a 4. o 29. c 30. a 5. c' 31. d 6. c 32. b 7. o , 33. c 8. c M. d 9. c 35. a 10. b- . 11. d 36. a 37. b '12, c 38. d 13. d 14. ' c 39. a 40. g d [[J Ty[0 666 o m M A,4 15. a b O [////(8 16. c
c , 17. a 42. a 18. d 43. c 19. a 44. b 20. c-45. a ,21' b 46, a . 22. b 47. a-23. c 48. d l l 24. c: 49. a (- ,25. d 50. a Page1
Senior Reactor Operator Answer Key 51. c 76. c 52. c 77. b 53. b 78. d 54, a 79. a 55. b 80. a 56. c 81. b G7. a 82. a 58. c 83. a 59. a 84. b 60. b 85. a . 86. d 61. c 62. b 87. c 63. b 88. b 64. a 89. d 65. a
c' 66, b 9' - ' 67. d ./. a
68. c 93. b ! ! 69. c 94. c 70. b 95. a .71 b 96. b , 72. a 97. d 73. d 98. b 74. b 99. b 75. b 100 d Page 2 _
--- - _- - _ _ - - _ _ - - _ - - _ _ _ _ - ___- - _ _ - _ _ _ - _ _ _ _ - - _ - _ _ - _ _ _ - 10M-46.4.A
Beaver Valley Power Stat;on UnN1 Issue 4 Revision 2
Post-DBA Hydrogen Control System Page A 1 of 5 , Operating Procedures . Hydrogen Recombiner Startup ~ 1.
PURPOSE This procedure describes the startup of the Post DBA Hydrogen Recombiner following the unlikely occurrence of a loss of coolant accident. This is accomplished by first Jetting up the Hydrogen Analyzer and monitoring Containment hydrogen concentration. When the concentration level reaches a preset value, the Hydrogen Recombiner is aligned and started.' This procedure is entered from an EOP.
l ' l-II.
PRECAUTIONS AND LIMITATIONS - If hydrogen concentration is it 5%, consult TSC before placing Recombiners in L A.
operation.
During accident conditions, radiation levels may be high in the Recombiner area.
B.
Limit the time spent in this area.
C.
In order for the Hydrogen Recombiners to operate with sufficient flow, Containment - ..- pressure must be controlled as close as possible to -2 psig (13 psla). However.
Containment pressure must remain below -2 psig (13 psia) to ensure Containment g i l remains substmospheric.
' N# lit.
INITIAL CONDITIONS A. The EOPs require the Hydrogen Recombiners to be placed in service.
' B. The NSS has approved the performance of this procedure.
. C. The 480 VAC distribution system is operable.
D. The following procedure is available: 10M-46.4.G. "Placir.g Wide Range Containment Hydrogen Monitoring System in 1.
Operation".
IV.
INSTRUCTIONS Valves for the A Recombiner are given in procedure, valves for the B Recombiner Note: are in parenthesis.
A.
Place the Hydronen Rembiner in Service
Contact Radcon to determine what type of protective apparel is to be worn and , . 1.
any shielding required.
2. Obtain the following keys to unlock [1HY-101,~ 102,103,104,110, iii,196 and 197]. a. SRIO.C.
b. SR/O.D.
.
- -. - _. _ --_ ____ _ . _ _ _ _ _ _ _ _ - - _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _. _ _ _ _ _ - _ _ _ - _ Attachment 2 l SIMULATION FACILITY REPORT Facility Licensee: Beaver Vallev Unit 1 Facility Docket No: 50-334 Operating Tests Administered from: April 20-24,1998 ! This form is used only to report simulator 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 observations.
No simulator deficiencies, that affected the scenario examinations or JPMs, were identified during the execution of the examination.
! l l }}