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| A a, L 7, 199 7 p
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| NOTE T0: NRC DOCUMENT CONTROL DESK MAIL STOP 0-5-D-24 FROM: gro hons OPE 1A"ING LICEN
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| , LICENSING ASSISTANT BRANCH _ REGION I
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| ==SUBJECT:==
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| OPERATOR LICENSING EXAMINATION ADMINISTERED ON SeK32.u iver _,ArBirchTstn a va {
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| D0hKET N05 Ss-9 77Ss -g 7{
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| ON en 1 2 2 2._ / 9 9 7 OPERATOR LICENSING EXAMINATIONS WERE ADMINISTERED AT THEl REFE'tENCED FACILITY. ATTACHED YOU WILL FIND THE FOLLOWING l INFORMATION FOR PROCESSING THROUGH NUDOCS AND DISTRIBUTION TO THE NRC STAFF, INCLUDING THE NRC PDR. )
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| Item #1 t) FACILITY SUBMITTED OUTLINE AND INITIAL EXAM SUBMITTAL DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE A070.
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| a) AS GIVEN OPERATING EXAMINATION, DESIGNATED F0P. DISTRIBUTION UNDER RIDS CODE A070.
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| Item #2 EXAMINATION REPORT WITH THE AS GIVEN WRITTEN. EXAMINATION ATTACHED, DESIGNATED FOR DISTRIBUTION UNDER RIDS CODE IE42. 1 1
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| 9804150104 DR 980407 ADOCK 05000277 PDR 1
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| a % %._.= h
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| *p >* "'4g t
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| DMb4 [k i
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| 'k
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| ' g UNITED STATES NUCLEAR REGULATORY COMMISSION o
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| j REGloN I 475 ALLENDALE ROAD
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| /g
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| * KING oF PRUSSIA, PENNSYLVANIA 19406-1415
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| ""** February 26, 1998 Mr. D. M. Smith Senior Vice President-Nuclear PECO Energy Nuclear Group Headquarters I- Correspondence Control Desk P. O. Box 195 Wayne, Pennsylvania 19087-0195
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| ==SUBJECT:==
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| EXAMINATION REPORT NOS. 50-277/97-09 AND 278/97-09 (OL)
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| ==Dear Mr. Smith:==
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| During the period of September 22,1997 to September 26,1997, the NRC administered initial examinations to 10 employees of your company who had applied for licenses to operate the Peach Bottom Atomic Power Station, Units 2 and 3.
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| Both of the reactor operator (RO) and seven of the senior reactor operator (SRO) applicants passed the examinations and were subsequently issued licenses. One SRO applicant failed the operating portion of the examination.
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| The examination was prepared by your staff using the Examiner Standards, NUREG 1021 Rev. 8 as guidance. During the examination developmental process, your staff identified similarities between the PECO proposed examination and an audit examination administered at Public Service Electric and Gas Co.'s Hopa Creek Nuclear Generating Station (HCNGS) during the week of September 1,1997. PECO Energy did a satisfactory job of controlling information regarding the similarities and revised some parts of the examination to reduce the similarities. While the similarities represented an unnecessary, unfortunate complication of the examination, the NRC determined the examination to be acceptable. A detailed review of the exam results by PECO and the NRC staff failed to identify indications that an examination compromise occurred. Accordingly, the examination was judged to be a valid basis for issuing licenses.
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| Ir-l l
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| lE j elm n n cy - -
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| my , v - o$ n) c l ,
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| E j
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| : Mr. D. M. Smith 2 We regret that internal processing difficulties caused transmittal of this report to be delayed. Should you have any questions regarding this information, please contact Carl Sisco of my staff at (610) 337-5076 or me at (610) 337-5211.
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| . Sincerely,
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| . I Richard J. Conte, Chief Operato. Licensing and Human Performance Branch Division of Reactor Safety .
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| Docket Nos. 50 277;50-278 cc w/ encl: w/ Attachments 1-3:
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| L. MacEntee, Acting Director - Training ec w/ encl: w/o Attachments 1-3:
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| G. Edwards, Chairman, Nuclear Review Board and Director, Licensing T. Mitchell, Vice President, Peach Bottom Atomic Power Station G. Rainey, Senior Vice President, Nuclear Operations J. B. Cotton, Vice President, Nuclear Station Support T. Niessen, Director, Nuclear Quality Assurance A. F. Kirby, lit, External Operations - Delmarva Power & Light Co.
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| M. Warner, Plant Manager, Peach Bottom Atomic Power Station G. J. Lengyel, Manager, Experience Assessment J. W. Durham, Sr., Senior Vice President and General Counsel T. M. Messick, Manager, Joint Generation, Atlantic Electric W. T. Henrick, Manager, External Affairs, Public S qvice Electric & Gas R. McLean, Power Plant Siting, Nuclear Evaluations -
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| J. Vannoy, Acting Secretary of Harford ' . sunty Council R. Ochs, Maryland Safe Energy Coalition J. H. Walter, Chief Engineer, Public Service Commission of idlaryland Mr. & Mrs. Dennis Hiebert, Peach Bottom Alliance Mr. & Mrs. Kip Adams Commonwealth of Pennsylvania State of Maryland TMI- Alert (TMIA)
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| o U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket Nos.: 50-277;50-278 Report Nos: 97-09;97-09 Licensee: PECO Energy Wayne, PA Facility: Peach Bottom Unit Nos. 2 and 3 Location: Delta, PA Exam Date: September 22-26,1997 Examiners: John Caruso, Operations Engineer Scott Willoughby, LITCO Chief Examiner: Carl Sisco, Operations Engineer, Operator Licensing and Human Performance Bianch-Division of Reactor Safety APPROVED BY: Richard J. Conte, Chief, Operator Licensing and Human Performance Branch l Division of Reactor Safety i
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| i i
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| I i
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| I
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| O Mr. D. M. Smith 3 Distribution w/ encl: w/ Attachments 13:
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| DRS Master Exam File I
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| ' PUBLIC .
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| Nuclear Safety Information Center (NSIC)
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| Distribution w/ encl: w/o Attachments 1-3:
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| Region 1 Docket Room (with concurrences)
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| J. Wiggins, DRS C. Sisco, Chief Examiner, DRS NRC Resident inspector DRS OL Facility File DRS File Distribution w/ encl: w/o Attachments 1-3:
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| : 8. McCabe, OEDO L. M. Padovan, PM, NRR J. Stolz, PDI-2, NRR
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| * inspection Program Branch, NRR (IPAS)
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| DOCDESK t
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| 'h L
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| 9 , . .
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| I EXAMINATION
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| ==SUMMARY==
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| j Examination Report 50-277/97-09and 50-278/97-09(OL)
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| During the period of September 22,1997 to September 26,1997,the NRC administered . l initial examinations of two RO and eight SRO applicants. Both of the RO and seven of the l SRO applicants passed the examinations and were cabsequently issued licenses. One SRO applicant failed the operating portion of the examination. During the cxamination developmental process, PECO identified similarities between the proposed examination and j an audit examination administered at the Public Service Electric and Gas Co.'s Hope Creek Nuclear Generating Station (HCNGS) during the week of September 1,1997. PECO Energy did a satisfactory job of controlling information regarding the similarities and revised some parts of the examination to reduce the similarities. While the similarities represented an unnecessary, unfortunate complication of the examination, the NRC determined the examination to be acceptable. A detailed review of the exam results by PECO and the i NRC staff failed to identify indications that an examination compromise occurred. I Accordingly, the examination was judged to be a valid basis for issuing licenses.
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| 1 I
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| e Report Details 1.0 Introduction
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| . The NRC administered initial examinations to seven SRO instants, one upgrade and two RO applicants.~ The examinations were administered in accordance with NUREG-1021, " Examiner. Standards," Revision 8, 2.0 Preexamination Activities The licensee prepared the examinations in accordance with NUREG-1021,
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| " Examiner Standards," Revision 8. Personnel involved in the development of the examination signed security agreements to assure integrity of the examination process.
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| 3.0 Examination 3.1 Examination Results c The results of the examinations are summarized below:
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| SRO RO Pass / Fail Pass / Fail Written 7'/O 2/O Operating 7/1 2/O Overall 7/1' 2/0
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| * Written examination waived for one SRO applicant 3.2 facility Strenaths and Weaknesses Ooeratina Examination I s All crews demonstrated strong commurications and effective teamwork. Also, the SRO applicants demonstrated very good conimand and control and effectively used crew briefings at appropriate times. The crews were poised during the operating examination and displayed an effective use of procedures. There were no weaknesses observed during the operating examinations.
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| 3.3 Examination Preparatiort f PECO staff prepared an examination sample plan, and the proposed examination was received in the. NRC regional office in a timely manner. The sample plan was comprehensive and needed no changes to the plan as a result of the in-office review. The proposed examinatio 1 was very well organized and added to the ease !
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| of review. The in-office review of the proposed examination did not result in any 1
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| , substantive changes. The NRC conducted an onsite preparation of the examination
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| 2 I the week of September 8.1997. During the onsite review, the NRC made changes to the examination. Thr. written examination changes consisted of rewriting five questions to increase cognitive difficulty level and minor changes to twenty four questions to improve clarity and assure technical adequacy. The JPM changes consisted of rewriting eight questions to improve clarity and cognitive difficulty level. The dynamic scenarios required no changes as a result of NRC review. j However, as a result of the similarities between the Peach Bottom NRC and Hope Creek internal audit examinations, PECO staff developed and validated a j replacement scenario. All axamination changes were reviewed in detail as well as a 1 demonstration of the newly validated scenario during a return on site visit September 16,1997. 1 1
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| l 3.4 Examination Similar%es !
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| l During the conduct of peer evaluations the week of September 1,1997 at Hope i Creek, a PECO instructor noted similarities between the Hope Creek audit j examination and the PECO-proposed Peach Bottom initial examination submittal to the NRC. This instructor was a member of ths PECO examination preparation team and was farraliar with the content of the PECO examination. PECO limited discussion of 'his information to personnel on the security agreem:rd. The NRC was notified c: these similarities on September 5,1997 and also that both of these examinations were prepared by the same contracting firm. On September 9,1997, PECO conducted a comparison of the Peach Bottom examination and the Hope Creek audit examination at the contractor's office. Based on this comparison, PECO stated changes were being made to the Peach Bottom examination to reduce the similarities between the examinations to be consistent with the guidance of NUREG-1021, Examination Standards, Rcv. 8.
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| The changes made to the examination are summarized below:
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| Scenarios: One scenario was replaced with a scenario developed during the preparation week September 8,1997.
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| Administrative Section: The radiation controls section of the SRO and RO examinations were changed by replacing a JPM with two questions.
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| Written Examinatiom Five questions (of 29) were revised significantly to minimize the overlap between the Peach Bottom and Hope Creek exeminations.
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| The revised examination was administered on September 19,1997. Following the administration of the examination, the NRC staff conducted a detailed review of the as-administered Hope Creek audit examination and the Peach Bottom as-administered examinations. Based on this review, the NRC findings concerning examination similarities were consistent with PECO's. Also, a detailed review of the exam o *ults by PECO and the NRC staff failed to identify indications of an examinauon compromise. Based on the above, NRC concluded that the examination represented a valid basis on which to issue licenses.
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| i o
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| 1 3
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| 4.0 EXIT MEETING
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| ' An exit meeting was conducted at the training facility September 26,1997.- The '
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| general findings of the NRC examiners were discussed. : Simulation Facility Report : Reactor Operator Written Examinction : Senior Reactor Operator Written Examination
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| r- 1 I' i l
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| l ATTACHMENT 1 l
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| SIMULATION FACILITY REPORT Facility License: DPR-63 l
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| Facility Docket No: 50 277 Operating Test Administration: Septem.it er 22-24,1997 l This form is to be used only to report observations. These observations do not constitute '
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| audit or inspection findings and are not, without further veri?ication and review, indicative of noncompliance with 10 CFR 55.45(b). These observatio'is do not affect NRC certification or approval of the simulation facility other tha'i to provide information that may be used in future evaluations. No licensee action is raquired in response to these observations. l l
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| JTEM DESCRIPTION l None 4
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| O e
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| ATTACHMENT 2 i
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| )
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| Question: 001 Which of the following individuals has the specific responsibility to ensure the current on-shift crew are u- ' up-to-date regarding their qualifications to be on shift, i e., Licensed Operator Requalification (LOR),
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| activities to maintain their license active, etc.7
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| : a. Senior Manager-Operations
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| : b. Shift Manager
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| : c. Manager- Operations Traming
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| : d. Shift Operations Assistant Answer: b
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| ==References:==
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| OM-P-3.2, " Senior Licensed Operators", Rev. 8, Section 2.2.b, Page 8 LOT-0006, "OM Chapters 0 - 5", Rev. 000, LO - None identified Exam Level: SRO History: New K/A: 294001G101 3.7/3.8 KA Statement: Knowledge of conduct of operations requirements System / Evolution:
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| Exam Section: Plant Wide Generics RO Grwp: SRO Gmup:
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| Mh e
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| -\
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| Questi:n: 002 Which of the following evolutions shall be performed under the direct supervision of a Licensed Senior Reactor Operator.
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| : a. Transfer ofEHC Pressure Regulators
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| : b. Placement of Recire MG Set scoop tube adjustable mechanical stops.
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| : c. Local Recirculation Pump MG Set scoop tube operations
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| : d. Fuel shufBe within the vessel (no fuelis removed from the vesrel)
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| Answer: d
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| ==References:==
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| LOT-0005, " Licensee Obligations And Responsibilities", Rev. 007, Section II.C.1, Page 6, LO - 1 Exam Level: Both History: New K/A: 294001G102 3.0/4.0 KA Statement: Knowledge of operator responsibilities during all modes of plant operation System / Evolution:
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| Exam Section: Plant Wide Generics RO Group: SRO Group:
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| Question: 003 i' . Given the following conditions:
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| - The Control Room Supervisor (CRS) has delegated completion of GP-3, " Normal Plant Shutdown" for Unit 3 to a fully quali6ed Senior Reactor Operator (SRO) l - This has been so logged in the Uni 6ed Control Room Log During the Unit 3 shutdown a problem requires entry into T-103, " Secondary Containment Control"
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| - Unit 2 is operating at 75% power during this time Which of the following dehneates the responshhty for wimed and control authority on each of the two Units for these conditions?
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| : l. a. The CRS shall retain command and control over both Units at all tunes
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| ! b. The Unit 3 SRO retains command and control over Unit 3 until an emergency no longer
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| : i. exists. The CRS retains wi.iwd and control over Unit 2.
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| l
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| : c. The Senior Manager - Operations shall assume w .is4 and control over both Units upon his arrival.
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| : d. The Unit 3 SRO irdi=*ely transfers Unit 3 wwwd and control to the Shift Manager
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| . and provides support and backup to the CRS on both Units.
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| Answer: a l
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| ==References:==
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| OM-P-3.2, " Senior LW Operators", Rev. 8, Section 4.1.a, Page 12 i-
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| !- LOT-0006, "OM Chapters 0 - 5", Rev. 000, LO - 3
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| ! Exam Level: SRO History: New L
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| K/A: 29400lG437 2.0/3.5
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| . KA Statement: Knowledge of the lines of authonty dunng emergency operations l
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| System / Evolution:
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| Exam Section: Plant Wide Generics RO Gesup: SRO Group:
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| 1
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| Question: 004 Which of the following evolutions require Unit 3 to have two (2) Licensed Operators assigned prior to starting?
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| Unit 3 is:
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| : a. perfonning an immediate shutdown required by Technical Spu:ifications.
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| : b. stroking rods for CRD testing.
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| : c. raising power from 10% to 15% with control rods after placing the Reactor Mode Switch in "h''.
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| : d. raising power with recire from 50% to 60%.
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| Answer: c
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| ==References:==
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| OM-P-3.3," Licensed Operators", Rev. 4, Section 6.2 & 6.3, Page 11 LOT-0006, "OM Chapters 0 - 5", Rev. 000, LO - 3
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| . Exam Level: Both History: New K/A: 29400lG201 3.7/3.6 KA Statement: Ability to perform pre-startup procedures for tim facahty, iebding operatmg those controls associated with plant equipment that could affect reactivity.
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| System / Evolution:
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| Exam Section: Plant Wide Generics RO Gmup: SRO Gmup:
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| l
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| L Question: 005 What is the maximum amount of time the Unit Reactor Operator (URO)"at-the-controls" can be
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| ";7 -
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| provided with a temporary reliefi e., a complete tumover does not have to be done?
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| l a. 15 minutes
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| : b. 30 minutes-
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| : c. 60 minutes
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| : d. 90 minutes l
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| Answer: c
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| ==References:==
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| OM-C-6.2, " Temporary Relief', Rev.1, Section 2.2.1, Page 2 LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - 1 Exam Level: Both History: New .
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| K/A: 2940010103 3.0/3.4 KA Statement: Knowledge of shift tumover practices System / Evolution:
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| Exam Section: Plant Wide Generics RO Gmup: SRO Group:
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| Questi:n: 006
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| -i Given the following conditions:
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| - With Unit 2 operating at 50% power, a packing leak is discovered on an accessible motor operated valve in a safety-related system
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| - The leak is not severe and it has been decided to backseat the valve during the next shift.
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| - All plant systems are operating as designed Which of the following describes how this valve should be hadsted? ]
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| : a. The appropriate System Manger should manually backseat the valve using TMT.
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| : b. The Operator in the Main Control Room should electrically backseat the valve.
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| : c. M sintenance personnel shoda manually backseat the valve.
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| : d. A Equipment operator at the motor control center should electrically backseat the valve.
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| Answer: c
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| ==References:==
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| OM-C-7.5, " Valves", Rec. 3, Section 2.4, Page 2
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| .. - LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - None Identified
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| (
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| Exam Level: Both History: New K/A: 294001G130 3.9/3.4 KA Statement: Ability to locate and operate components, including local controls.
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| System / Evolution:
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| Exam Section: Plant Wkle Generics RO Group: SRO Group:
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| i Questi:n: 007 l
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| Which of the following ALLOWS operation of Danger Tagged equipment?
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| : a. Local Leak Rate Test Tag
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| : b. Special Condition Tag
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| : c. Information Tag
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| : d. SuspensionLabel Answer: d
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| )
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| ==References:==
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| Clearance And Tagging Manual, Rev. 3, Section 19.0 Definitions, Page 125 NCT-0200, " Clearance & Tagging", Rev.1, LO - 2 Exam Level: SRO History: New K/A: 29400lG213 3.6/3.8 KA Statement: Knowledge of tagging and clearance procedures System / Evolution: ,
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| l Exam Section: Plant Wide Generics RO Group: SRO Group:
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| f I
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| Question: 008
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| ~
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| Prior to a reactor stanup, a check-offlist (COL) is being performed on the Unit 3 "A" Residual Heat Removal (RHR) Loop. During performance, the operator has discovered that a step in the COL has an incorrect component identification designation.
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| Select the appropriate actions for these conditions.
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| TFe operator shall:-
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| : a. not complete the COL until a temporary change is prepared in accordance with A-3.
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| : b. document the problem with that COL step and continue to completion of the COL with Shift Management approval.
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| : c. make a note of the discrepancy on the specific COL step, initial and date the step and continue to completionin the COL.
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| : d. have an immediate Double Verification performed on the COL step and the component, and then complete the COL with Shift Management approval.
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| Answer: b
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| ( ]i
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| ==References:==
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| OM-C-10.7, " Check-Off Lists", Rev.1, Section 5.2, Page 5 LOT-0008, "OM Chapters 10 - 15", Rev. 000, LO - 1 Exam level: Both History: New K/A: 29400lG129 3.4/3.3 KA Statement: Knowledge of how to conduct and venfy valve lineups.
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| System / Evolution:
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| Exam Section: Plant Wide Generics RO Group: SRO Group:
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| L
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| i l Questi:n: 009 l Given the following assumptions: l l -- Plant systems A and B are Tech Spec systems req ired to support the operation of System !
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| C 1
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| - Ifinoperable, System A has a completion time for testoration to Operability of 7 days !
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| - Ifinoperable, System B has a completion time for restoration to Operability of 14 days
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| - Ifinoperable, System C has a completion time fer restoration to Operability of 3 days :
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| - System B hasjust been determined to be Inope able l Which of the following is the MAXIMUM completion time for restoring the System C to Operable status?
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| : a. 3 days 1
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| : b. 10 days 1
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| 1
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| : c. 14 days
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| : d. 17 days Answer: d i;
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| ==References:==
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| OM-P-12.2, " Safety Function Detemunation Program", Rev.1, Section 4.5 LOT-0008,"OM Chapters 10 - 15, Rev. 000, LO - 1 & 3 Exam Level: SRO History: New K/A: 29400lG223 2.6/3.8 KA Statement: Ability to track knutmg conditioos for operations.
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| System / Evolution:
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| Exam Section: Plant Wide Generics RO Group: SRO Group:
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| _. c i
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| l l
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| 1
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| r l
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| Question: 010-Given the following conditions:
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| A male, fully qualified radiation worker at Peach Bottom hasjust retumed from 4 weeks of outage support at Lunerick
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| - Total Effective Dose Equivalent (IEDE) received at Limerick was 250 mrem.
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| - This workers' current TEDE from Peach Bottom for 1997 is 225 mrem l
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| What is the MAXIMUM annual non-emergency Total Effective Dose Equivalent (TEDE) that can be received at Peach Bottom for the remainder of1997 WITHOUT m= ling the Federal Exposure Limits.
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| : a. 4475 mrem
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| : b. 4525 mrem .
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| : c. 4750 mrem i
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| : d. 4775 mrem
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| ! Answer: b
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| ==References:==
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| HP-C106, " Dosimetry Program", Rev. 3, Sections 7.1.1, Page 3 f.
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| )
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| 10CFR20.1201, Ocmpatinnal Dose Lunits for Adults LOT-1730, " Radiation Exposure Limits", Rev. 012, LO. - 2 Exam Level: Both History: New K/A: 294001G304 2.5/3.1 KA Statement: Knowledge of radiation exposure knuts and contammation control, including permissible levels in excess of those authorized System / Evolution:
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| Exam Section: Plant Wide Generics RO Group: SRO Group:
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| l Justification: Low RO K/A importance value but required knowledge per 10CFR20 ,g i
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| Es
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| Question: 011 Which of the following is the REQUIRED immediate action if a Locked High Radiation Area door is
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| * found open with no control of area access?
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| a Inform Security and establish Positive Access Control
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| : b. Inform the on-shift Health Physics Technician and lock the area after checking for unauthorized personnel
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| . Inform the Secu:ity, lock the area and have Health Physics check for exposures in excess of those expected
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| : d. Inform the Health Physics Supervisor and establish Positive Access Control i Answer: d
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| ==References:==
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| HP-C-202, " Locked High Radiation Area Controls", Rev. 6, Section 7.13.1, Page 9 GETCM-10308, " Radiation Protection For PAAT Workers", Rev. 0A, LO - 10 ,
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| Exam level: Both History: New s K/A: 29400lG310 2.9/3.3 KA Statement: Ability to perform procedures to reduce excessive levels of radiation and guard agamst personnel exposure System / Evolution:
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| i Exam Section: Plant Wide Generics RO Group: SRO Group:
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| L a
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| Questi n: 012 An off-site release is in progress such that the Emergency Director has made a Protective Action Recommendation to the state and local agencies.
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| What is the MLNIMUM emergency classification that must be made for these condition (!
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| : a. Unusual Event
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| : b. Alert
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| : c. Site Area Emergency
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| : d. GeneralEmergency Answer: d
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| ==References:==
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| ERP-101, " Classifications Of Emergencies", Rev.19, Section 6.3 Note, Page 4 PEPP-6010, " Emergency Director Trainmg", Rev. O, LO - 5.c Exam level: SRO History: New K/A: 29400lG444 2.1/4.0 KA Statement: Knowledge of emergency plan protective action recommendations
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| - System / Evolution:
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| Exam Section: Plant Wide Generics RO Group: SRO Group:
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| Question: 013 Given the following conditions:
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| - With Unit 3 at 80% power it has just been discovered that a weekly surveillance was not accomplished within its speciSed frequency
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| - It should has e been completed 3 days ago Which of the following describes the requirements regardmg the associated equipment's operability status?
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| The equipment:
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| : a. must have a Safety Function Determmation performed within 6 hours.
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| : b. may be considered to be operable for a maximum period of 24 hours to allow the surveillance to be performed.
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| : c. must be declared inoperable immediately.
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| : d. may be considered to be operable for the surveillance grace period (1.73 days) to allow the, surveillance to be performed.
| |
| Answer: b Refeances: OM-P-15.1," Operability", Rev. 3, Section 4.12.1, Page 23 LOT-0008, "OM Chapters 10 - 15", Rev. 000, LO. - 3 Exam level: SRO History: New K/A: 29400lG212 3.0/3.4 KA Statement: Knowledge of survedlance procedures System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Gesup:
| |
| j
| |
| | |
| Question: 014 dich of the following must be met prior to taking deliberate action to depart from Peach Bottom fechnical Specifications during an emergency? l l
| |
| This action is:
| |
| : a. immediately needed specifically to protect the public health and safety.
| |
| : b. immediately needed specifically to protect valuable public property.
| |
| : c. immediately needed specifically to maintain capacity factors in excess of corporate standards.
| |
| : d. immediately needed to mamtam grid stability during and Emergency Generation condition.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| 10 CFR 50.54x & y LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - 2 & 3 Exam level: Both History: New K/A: 29400lG412 3.4/3.9 KA Statement: Knowledge ofgeneral operatmg crew responsib:. Sties during emergency operations System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Gmup: SRO Gmup:
| |
| | |
| 1 I
| |
| Question: 015 i
| |
| Which of the following is an appropdate use of a Special Condition Tag (SCT)? j l
| |
| : a. An SCT is being applied to an active clearance to correct a tagging conflict covered by a )
| |
| Suspension label. !
| |
| : b. A second SCT, with the same position as the first, is being applied to the same component. !
| |
| l
| |
| : c. An SCT is being applied to a component requidng manipulation during maintenance. !
| |
| l
| |
| : d. The SCT is being applied on equipment already tagged with a danger tag from the Master Clearance. i l
| |
| Answer: c
| |
| | |
| ==References:==
| |
| Clearance & Taggmg Manual, Rev. 3, Section 4.3, Pages 21 & 22 NCT-0200, " Clearance & Taggmg", Rev.1, LO - 2 Exam Level: Both History: New K/A: 294001G213 3.6/38 KA Statement: Knowledge of taggmg and clearance procedures System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Gmup: SRO Gmup:
| |
| 1 i
| |
| .= an
| |
| | |
| 1 Questirn: 016 Which of the following tasks may be performed by a fully qualified Advancec Radiation Worker (ARW) without prior HP Supervisor approval?
| |
| : a. Free release ofmaterials from an RCA
| |
| : b. Performing surveys in areas posted as " Radiation Area" or less
| |
| : c. Performing surveys in a suspected hot particle area
| |
| : d. Performing contammation surveys in a high airborne radioactisity area Answer: b
| |
| | |
| ==References:==
| |
| PECO Energy Radiation Worker Handbook, Page 40 GETCM-10400, " Basic Radiation Worker Traming", Rev. 2, LO - 61 & 75 Exam Level: Both History: New K/A: 29400lG301 2.6/3.0 1
| |
| 3 KA Statement: Knowledge of 10CFR20 and related facdity radiation control requirements System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Gmup:
| |
| I
| |
| | |
| l !
| |
| I Question: 017 i i !
| |
| ; During a declared emergency, at what point will the Emergency Director have to approve an
| |
| *~'
| |
| " Emergency Exposure Authorization Form"? I l
| |
| Expected individual Total Effective Dose Equivalent (TEDE) exposures will exceed:
| |
| : a. 3000 mrem.
| |
| : b. 4000 mrem. l
| |
| : c. 5000 mrem.
| |
| : d. 25,000 mrem.
| |
| Answer: c i
| |
| | |
| ==References:==
| |
| ERP-670, " Emergency Radiation Exposure Guidelines And Controls", Rev. 4, Section l 2.1.4, Page 2 !
| |
| PEPP-0010, " Emergency Preparedness Tranung", Rev. O, LO - 7 l f
| |
| z Exam level: SRO History: New I J K/A: 29400lG440 2.3/4.0 KA Statement: Knowledge of the SRO's responsibilities in mum sucy plan implementation i System / Evolution:
| |
| l Exam Section: Plant Wide Generics RO Group: SRO Group: i Justification: Blanket approval to exceed the 3000/4000 limits to 5000 during declared emergencies L
| |
| | |
| Questi:n: 018 One of the concems with maintaining proper reactor water level during plant operation is to minimize "carryunder" Which of the following would result if excessive "carryunder" were occumng?
| |
| I
| |
| : a. Steam quality exiting the reactor vessel will decrease.
| |
| {
| |
| : b. Jet pump net positive suction head would 'mcrease.
| |
| : c. Indicated reactor water level will fluctuate.
| |
| : d. Core thermal power would decrease Answer: d
| |
| | |
| ==References:==
| |
| LOT-0010," Reactor Vessel And Intemals", Rev. 007, Section III.C.4.c.2), Page 24, LO. - 2.b Exam Imel: Both Eb4 v: New K/A: 290002K303 3.3/3.4
| |
| .G
| |
| " KA Statement: Reactor power i
| |
| System / Evolution: Reactor VesselIntemals Exam Section: Plant Systems RO Group: 3 SRO Group: 3 Justification: Answer is the effect from incressmg downcomer twwgures from steam coming from the separators into the downcomer i
| |
| | |
| Question: 019 During rapid depressurization, reactor water level indication is susceptible to " outgassing" Which of the following describes how proper ECCS system initiation and operation is assured with potentially unreliable reactor water level indication?
| |
| : a. The ECCS systems receive initiation signals from water level indicators not affccied by this transient.
| |
| : b. The ECCS systems will have already initiated prior to " outgassing" affecting the instrumentation.
| |
| : c. The Condensing Chamber Backfill System ensures accurate water -vel indicators for ECCS initiation during all depressunzation transients.
| |
| : d. ECCS initiation logic is designed such that level differential errors are ef.ectively canceled out during" outgassing" Answer: b
| |
| | |
| ==References:==
| |
| LOT-0050, " Reactor Vessel Instrumentation", Rev. 013, Handout #5, Pages 1 & 2, LO - 2.c & 6.f& g
| |
| .s-f History: New Exam Level: SRO K/A: 216000K405 3.9/4.1 ,
| |
| KA Statement: Initiation of the emergency core cooling systems System / Evolution: Nuclear BoilerInstrumentation l Exam Section: Plant Systems RO Group: 1 SRO Group: 1 I
| |
| | |
| Questirn: 020' Given the following conditions:
| |
| A reactor startup is in progress on Unit 2 . .
| |
| -- The " continuous withdraw" mode for control rod withdrawals is being used While withdrawing control rod 26-31, the Unit Reactor Operator noted that control rod motion had stopped, the " settle" function had been completed and the rod had automatically de-selected Which of the following has occurred?
| |
| : a. The Emergency In/ Notch Override Switch was released
| |
| : b. A Rod Position Information System "Inop" has been received.
| |
| : c. A Rod Block Monitor rod block has occuned
| |
| : d. A Rod Worth Minimizer rod block has occurred Answer: b
| |
| | |
| ==References:==
| |
| LOT-0060, " Control Rod Drive Mechamsm System / Rod Position Indication System",
| |
| s Rev. 009, Section II.B.2.d, Page 18 ARC 211 D-5 LOT-0080, " Reactor Manual Control System", Rev. 007, LO - 2.a & b Exam Level: Both History: New K/A: 214000K303 3.1/3.2 KA Statement: RMCS System / Evolution: Rod PositionInformation System Exam Section: Plant Systems RO Group: 2 SRO Group: 2 L
| |
| I
| |
| | |
| Questi:n: 021 i
| |
| Given the following conditions:
| |
| Unit 2 dofwell pressure has reached: 2.2 psig ;
| |
| Reactor water levelis: -172 inches !
| |
| One Core Spray Pinnp has started with no other ECCS puraps in service
| |
| - No operator arnions have been taken 4 l
| |
| l The Automatic Depressurizat;on System will actuate:
| |
| : a. in 115 seconds.
| |
| : b. in 9.5 minutes.
| |
| : c. 9.5 minutes after any additional Core Spray Pump is started I
| |
| : d. I 15 seconds after a :j RIiR hmo is started. j Answer: d
| |
| | |
| ==References:==
| |
| LOT-0330, " Automatic Depressunzation System", Rev. 008, Section V.A.1 - 8, Pages 15 & 16, LO - 2.a & 6.a ;
| |
| . 1 Exam Icel: Both History: New l
| |
| K/A: 21800CK501 3.8/3.8 i
| |
| I KA Statement: ADS logic operation I
| |
| System / Evolution: Automatic Depressunzation System i Exam Section: Plant Systems RO Group: 1 SRO Group: 1 l
| |
| | |
| Question: 022 Unit 2 is operating at 95% power when z. scram signal is generated from a main turbine trip.
| |
| Which of the following conditions will PREVENT the backup scram valves from venting the scram air header?
| |
| : a. The check valve bypassing the downstream backup scram valve fails closed.
| |
| : b. The upstream backup scram valve in the air flow path fails to reposition.
| |
| : c. The "A" Reactor Protection System trip system does NOT deenergize.
| |
| : d. 125 VDC to ONE of the backup scram valve is deenergized Answer.- c ,
| |
| | |
| ==References:==
| |
| LOT-0070," Control Rod Drive Hydraulic System", Rev. 008, Section III.I.4.d &
| |
| Figure 6, Page 16, LO - 4.d & 6.c
| |
| ~
| |
| Exam Level: Both History: New K/A: 20100lK203 3.5/3.6 r
| |
| KA Statement: Backup SCRAM valve solenoids System / Evolution: Control Rod Drive Hydraulic System Exam Section: Plant Systems RO Group: 1 SRO Group: 2
| |
| | |
| 1 I
| |
| Questi:n: 023 Given the following ccaditions:
| |
| l l
| |
| Reactor power is being raised from 25% to 35% by control rod withdrawals The current group of rods being withdrawn has a withdraw limit of Notch "24" The last rod in the group is withdrawn to Notch "28" -
| |
| The CRS has been notified
| |
| - One minute has elapsed since the rod withdrswal was completed What are the required actions for this control rod?
| |
| : a. Immediately insert the rod to Notch '74" and notify the Reactor Enghicer.
| |
| : b. Leave the rod at Notch "28" and notify the Reactor Engineer l
| |
| : c. Insert the rod to Notch "24" and continue the power increase with CRS approval.
| |
| : d. Immediately insert the rod to Notch "00". j
| |
| \
| |
| Answer: a ;
| |
| i
| |
| | |
| ==References:==
| |
| ON-122, "Mispositioned Control Rod", Rev. 4, Section 2.5, Page 1
| |
| '/ 's j LOT-1550, "Off-Normal Procedures", Rev. 006, LO - 2 l Exam Level: Both History: New K/A: 201002G449 4.0/4.0 KA Statement: Ability to perform without reference to procedures those actions that require immaAnte operation of system Owi-: =.a and controls System / Evolution: Reactor Manual Control System ,
| |
| Exam Section: Plant Systems RO Group: 1 SRO Group: 2
| |
| - G.
| |
| e 8
| |
| | |
| l Question: 024 Given the following conditions: {
| |
| i '
| |
| A cooldown is in progress using bypass valves Level control is in automatic on the startup levd controller Actual water level is being controlled at a constant value Which of the following describes what will happen to INDICATED Wide Range level du:ing the cooldown?
| |
| INDICATED Wide Range level will:
| |
| : a. remain constant.
| |
| : b. decrease towvds actuallevel.~
| |
| : c. increase from actud level.
| |
| : d. decrease from actuallevel.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0050, " Reactor Vessel Instrumentation", Rev, 013, Section III.F & Handout #7,
| |
| ) ages 20 & 21, LO - 6.d .
| |
| Exam Level: Both History: New K/A: 216000K510 3.1/3.3 KA Statement: Indicated level versus x:tual vessel level during vessel heatups or cooldowns System / Evolution: Nuclear Boiler Instre= tion Exam Section: Plant Systems RO Gesup: 1 SRO Group: !
| |
| -i
| |
| /
| |
| | |
| 1 1
| |
| Question: 025 Given the following conditions:
| |
| Drywell ventilation was operating in its normal lineup when a Loss Of Coolant Accident occurred on Unit 2
| |
| -- Drywell pressu::is 4 psig
| |
| - The Reactor and Radwaste Buildings are not accessible Which of the following describes the cooling capabibs of the drywell coolet s for these conuitions?
| |
| The drywell coolers-.
| |
| : a. will automatically be running in" Fast" speed ,
| |
| : b. will automatically be running in " Slow" speed.
| |
| : c. cannot be restarted until the Reactor Building access is restored.
| |
| : d. may only be restarted in " Fast" speed.
| |
| Answer: d -
| |
| I s'
| |
| | |
| ==References:==
| |
| LOT-0140, "Drywell Ventilation", Rev. 009, Sectiota IV.C.2 & D.4, Pages 6 - 8, LO -
| |
| S&5 i
| |
| Exam Level: Both h tory: New K/A: 22300lK103 3.2/3.3 ,
| |
| i KA Stat.ement: Containment /drywell sanosphere control ;
| |
| SysteinAholution: Prirnary Contamawst System AM Amaharies Exam Secdon: Plant Systems RO Group: 1 SRO Group: I l u
| |
| | |
| Question: 026
| |
| - Given the following conditions:
| |
| A Loss OfCoolant Accident has occurred on Unit 2
| |
| - All plant systems automatically operated as designed
| |
| . P.eactorwaterlevelis: -25 inches y - - Drywell pressure is: 1.7 psig
| |
| - The Plant Reactor Operator is attempting to tnnsfer the Reactor Core Isolation Cooling (RCIC) system from the Injection Mode to the CST to CST Mode
| |
| - Each time the Full Flow Test Valve (MO-30) is opened, it imn=Aistely closes Which of the following is the reason why the PRO cannot complete the transfer.
| |
| : a. The PRO has failed to reset the low reactor water level RCIC initiation signal.
| |
| : b. The RCIC Minimum Flow Valve (MO-27) interlock with MO-30 is in effect.
| |
| i
| |
| : c. The RCIC system is op.Gg with suction from the torus. j
| |
| : d. The drywell pressure is still too high to complete the transfer. !
| |
| Answer: c y
| |
| | |
| ==References:==
| |
| SO 13.1.B-2,"RCIC System Manual Operation", Rev. 6, Section 4.3.1 Note, Page 7 LOT-0380, " Reactor Core Isolation Cooling", Rev. 010, LO - 1.c & 2.a Exam lael: Both History: New K/A: 217000A404 3.6/3.6 KA Statement: Manuallyinitiated controls System / Evolution: Reactor Core Isolation Coohng System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| 'f
| |
| | |
| Questian: 027 With Unit 2 at 100% the "B" Core Spray Line Break differential pressure is reading approximately -3.0 psid.
| |
| This reading is:
| |
| : a. indicative of a "B" Core Spray line break outside the shroud.
| |
| : b. considered normal due to the actual differential pressure across the steam dryers and separators at 100% power.
| |
| i
| |
| : c. indicative of a "B" Core Spray line break inside the shroud.
| |
| : d. considered normal due to the differences between 'mstrument cabbration conditions and conditions at 100% power.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0350, " Core Spray", Rev. 010, Section IV.A.4.c, Pages 10 & 11, LO - 5.d Exam level: Both History: New
| |
| , K/A: 209001A205 3.3/3.6 ;
| |
| KA Statement: Core sprayline break System / Evolution: Low Pressure Core Spray System j Exam Section: Plant Systems RO Group: 1 SRO Group: 1 L
| |
| t
| |
| (
| |
| | |
| Questi:n: 028
| |
| ~
| |
| ' Given the following conditions on Unit 2:
| |
| - A half scram exists on Reactor Protection System (RPS)"A" due to APRM testing
| |
| - A fire hasjust caused a loss of RPS Bus "B" resulting in a full scram All control rods fullyinserted
| |
| -- The half scram testing was nopped and the APRMs returned to normal
| |
| - The SDV High Level S' ram c Bypass switch is then taken to " Bypass" When can the half scram on RPS "A" be reset for these conditions?
| |
| The RPS "A" half scram may be reset:
| |
| : a. immediately. -
| |
| : b. after the Scram Discharge Volume Vent and Drain valves are fully open..
| |
| : c. after RPS "B"is reenergized.
| |
| : d. only if the Reactor Mode Switch is in " Shutdown", " Refuel" or "Startup/ Hot Standby".
| |
| Answer: c 1
| |
| | |
| ==References:==
| |
| LOT-0300, " Reactor Protection System", Rev. 013, Section IV.B.4.b, Page 30, LO -
| |
| 5.i & 7.a Exam level: Both History: New K/A: 212000K601 3.6/3.8 KA Statemer.t: A.C. electrical distribution System / Evolution: Reactor Protection System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 6.
| |
| | |
| . Question: 029-Given the following conditions:
| |
| l . .
| |
| - - Unit 3 is perfornung a scheduled shutdown in accordance with GP-3,"Nonnal Plant Shutdown"
| |
| - Control rods are being mserted per the required insertion sequence .
| |
| The Rod Worth E ..;1.er was NOT initinhmi when required by GP-3 Which of the following describes how this will affect plant operation?
| |
| : a. Control rod insertions may contmue provided the Unit Reactor Operator does not deviate .
| |
| ' from the required sequence.
| |
| : h. Control rod insertions may contmue until a total of two " insert" errors are received.
| |
| : c. Control rod insertions will be stopped when steam Sow decreases to less than 17% or feed flow decreases to less than 16%.
| |
| : d. Control rod haertions will be stopped when steam Sow decreases to less than 27%. ,
| |
| Answer d'
| |
| | |
| ==References:==
| |
| LOT-0090, " Rod Worth m;...;ai", Rev. 010, Section V.B.2, Page 24, LO - 5.j &
| |
| 6.c Exam Imel: Both History: New K/A: . 201006A304 3.5/3.4 KA Statement: Controlrod movement blocks System / Evolution: Rod Worth Minimizer System Exam Section: Plant Systems RO Group: 2 SRO Group: 2 l
| |
| | |
| Question: 030 Given the following conditions:
| |
| The Unit 2 main generator hasjust been synchronized to the grid and is carrying 75 MWe load
| |
| -- The Turbine Building operator reports a loud noise coming from the #1 Low Pressure Turbine
| |
| - The Unit Reactor Operator depresses the "All Valves Closed" pushbutton on the EHC Control Panel Select the expected main turbine response.
| |
| : a. No Turbine Control Valve or Turtune Stop Valve motion will occur.
| |
| : b. The Turbine Control Valves and Turbine Stop Valves will close (turbine trip) and the generator output breaker will open from the turbine trip
| |
| : c. The Load Set Potentiometer will pulse back to 0% load but a turbine will NOT trip..
| |
| : d. The Turbine Control Valves and Turtune Stop Valves will close (turbme trip) and the genemtor output breaker will trip on reverse power.
| |
| Answer: a
| |
| -)
| |
| | |
| ==References:==
| |
| LOT-0590, " Electro-Hydrauhc Control Lognf', Rev. 007, Section I.2.B.a, Page 7, LO
| |
| - 1.b & 4.c Exam Level: Both History: New K/A: 241000A419 3.5/3.4 KA Statement: Turbine panel controls System / Evolution: Reactor /Turtune Presmue Regulatmg System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 Justification: With 1800 rpm selected and the output breaker closed this p=Wmon is disabled.
| |
| | |
| I Question: 031-l Given the following conditions:
| |
| - - - Unit 2 is operating at 100% power
| |
| ! -- The "A" Recirculation Pump hasjust tripped
| |
| -- ARer power and flow have stabilized, the Unit Reactor Operator notes the "B" Recirculation Loop Jet Pump Sow is greater than it was prior to the pump trip No operator actions are taken Which of the following describes the reason for this change?
| |
| l "B" Loop Jet Pump flow increased due to:
| |
| I
| |
| : a. the reduced subcooling at the recire pump suction.
| |
| : b. the increased core voiding.
| |
| I c. the lower core pressure drop.
| |
| : d. theincreased reactorwaterlevel.
| |
| Answer: c i.
| |
| | |
| ==References:==
| |
| LOT-0030, " Reactor Recirculation System", Rev, 009, Section VHF, Pages 48-50, )
| |
| LO - 6.h & 9.d l
| |
| CentnfugalPump Head Loss Curves l Exam Level: Both History: New l- i
| |
| ! K/A: 202001A102 3.4/3.4 i l
| |
| l KA Statement: Jet pump flow l 1 System / Evolution: Recirculation system Exam Section: Plant Systems .RO Group: 2 SRO Group: 2 l O
| |
| t
| |
| | |
| . l l
| |
| Question: 032 l
| |
| ,'~,
| |
| Which of the following describes the possible result of plant operation without the Rod Block Monitor (RBM)? !
| |
| With the RBM not available,:
| |
| : a. a control rod drop accident at high power may cause an Average Planar Linear Heat Generation Rate violation.
| |
| i
| |
| : b. a control rod withdrawal error at low power may cause an Average Planar Linear Heat l Generation Rate violation.
| |
| : c. a control rod drop accident at low power may cause a Minimum Critical Power Ratio violation.
| |
| : d. a control rod withdrawal error at high power may cause a Minimum Critical Power Ratio violation.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| PBS T.S. 3.3.2.1, " Control Rod Block Instra-tation" Bases, Page B 3.3-45 LOT-0280, " Rod Block Monitor", Rev. 010, LO - 1.a & 10 Exam Level: SRO History: New K/A: 215002A304 3.6/3.5 KA Statement: . Verification of proper functioning / operability System / Evolution: Rod Block Monitor System Exam Section: Plant Systems RO Group: 2 SRO Group: 2
| |
| | |
| T Questi n: 033 i r Given the following conditions: i w;.>" i,
| |
| - Unit 3 is performing a reactor stanup and heatup
| |
| - Reactor water level control is via Reactor Water Cleanup (RWCU) rejecting to the main j
| |
| ! condenser j - Main condenser vacuum has been established with the vacuum pump l.
| |
| l Why is the operator cautioned to carefully monitor system parameters while in this lineup?
| |
| l l
| |
| l Rejecting water to the main condenser:
| |
| : a. at high Bowrates can affect condaam vacuum whde using the vacuum pump during the startup.
| |
| l i-
| |
| : b. can exceed the Reactor Building Closed Cooling Water system cooling capacity for the non-regenerative heat exchanger.
| |
| l
| |
| : c. will cause damage to the Siter-demm resin strainers and resin "d==Eg" from the high reject flowrates.
| |
| f", d. may exceed the maximum allowed non-regenerative heat aveh=ger tube side to shdl side differentialtemperature.
| |
| Answer: b i-
| |
| | |
| ==References:==
| |
| LOT-0110, " Reactor Water Cleanup System", Rev. 011, "Section V.C.3.c & 4, Page 15, LO - 4.g & 5.b l
| |
| Exam level: SRO History: New l l
| |
| l K/A: 204000A407 3.1/3.1 KA Statement: System temperature l
| |
| l 1
| |
| l System / Evolution: ReactorWater Cleanup System j- Exam Section: Plant Systems RO Group: 2 SRO Group: 2 L
| |
| l l
| |
| l
| |
| | |
| Questi:n: 034 4
| |
| Unit 2 is operating at 85% power with the "A" Recirculation Pump scoop tube locked.
| |
| Select the action REQUIRED if a reactor scram occurs.
| |
| : a. Direct a Licensed Operator to manually position the "A" Recirculation Pump scoop tube' to "mimmum" speed
| |
| : b. ' Reduce the "B" Recirculnion Pump speed to muumum, then trip the "A" pump.
| |
| : c. Trip the "A" Recirculation Pump immediately,
| |
| : d. Direct any operator to position the "A" Recirculation Pump scoop tube to "nunimum" speed then trip the pump from the control room.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| SO 2D.7.B-2," Recirculation MG Set Scoop Tube Lockup And Reset", Rev. 8, Section 4.2 Caution, Page 4 LOT-0040, " Recirculation Flow Control", Rev. 008, LO - 1.c & 5.c
| |
| , .,- Exam Imel: Both History: New K/A: 202002A205 3.1/3.1 KA Statement: Scoop tubelockup System / Evolution: Recirculation Flow Control System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| f l
| |
| Questi:n: 035
| |
| , Given the following conditions:
| |
| I Unit 2 had been operating at 55% power, increasing power to 100% following an outage Main turbine vibrations required placing the Reactor Mode Switch to " Shutdown"
| |
| ~
| |
| All plant response and scram actions were normal l
| |
| l Which of the following PREVENTS control rod withdrawals for these conditions?
| |
| l l a. The Reactor Mode Switch in " Shutdown" maintains a scram signal on RPS until reset by
| |
| ! the operator.
| |
| : b. The Reactor Mode Switch in " Shutdown" inserts a continuous control rod withdrawal block signal.
| |
| : c. The Rod Block Monitor "Downscale" inserts a control rod withdrawal block signal until bypassed.
| |
| i d. Post scram Rod Worth Minimizer insert and withdraw errors will result in a control ro'd withdrawal block signal.
| |
| Answer: b
| |
| ~
| |
| \
| |
| | |
| ==References:==
| |
| SO 62.7.A-2," Receipt OfRod Blocks", Rev.12, Attachment 1, Page 5 l LOT-0080, " Reactor Manual Control System", Rev. 007, LO - 2.b & 4.a Exam Level: Both History: New
| |
| ! K/A: 201002K103 3.4/3.6 l
| |
| ! KA Statement: Control rod block interlocks / power operation / refueling System / Evolution: Reactor Manual Control S; stem i i Exam Section: Plant Systems RO Group: 1 SRO Group: - 2
| |
| ! i l
| |
| I
| |
| | |
| ' Questi:n: '036 Given the following conditions:
| |
| -- The E-2 Diesel Generator (DG) received a valid start signal on loss of voltage on the E-22 bus
| |
| -- The DG started and reenergized the bus as designed
| |
| - The cause of the bus loss ofvoltage has been found and corrected and preparations are being made to remove the DG from the bus and shut it down
| |
| - The operator hasjust depressed the DG Auto StarCypass pushbutton Which of the following is the function of this pushbutton?
| |
| : a. This bypasses the DG auto start signal to penst the diesel to be shutdown after E-22 has
| |
| , been tr.r.fr.nai to the grid.
| |
| : b. This places the DG in the isochronous (unit) mode to allow the E-22 bus to be synched to the grid and the DG unloaded prior to shutdown.
| |
| : c. This bypasses the DG auto start signal for 3 minutes to allow the E-22 bus transfer back to the grid before DG shutdown.
| |
| : d. This places the DG in the synchronous (parallel) nat to allow the E-22 bus to be synched c . to the grid and the DG unloaded prior to shutdown.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0670," Diesel Generators And Auxiharies", Rev. 006, Section IV.D.2.h.4), Page 38, LO - 1.b & 3.a Exam level: Both History: New K/A: 264000K407 3.3/3.4 KA Statement: local operation and control System / Evolution: Emergency Dusel Generators Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| I l
| |
| Question: 037 Assuming the Local Power Range Monitor (LPRM) detector outputs were NEVER calibrated over their expected lifetimes, which of the following would be the result and why?
| |
| I
| |
| : a. Actual reactor power would be GREATER than indicated APRM power due to the l depletion of the LPRM detector U-235 coating over time.
| |
| : b. Actual reactor power would be LESS than indicated APRM power due to the depletion of the LPRM detector U-235 coating over time.
| |
| l c. Actual reactor power would be GREATER than indicated APRM power uue to the build up of the LPRM detector U-235 coatmg over time.
| |
| i l
| |
| : d. Actual reactor power would be LESS than indicated APRM power due to the build up of l the LPRM detector U-235 coating over time.
| |
| Answer: a i l
| |
| | |
| ==References:==
| |
| LOT-0260, " Local Power Range Monitor System", Rev. 007, Section III.C., Page 8, LO - 5 & 6.a l Exam Level: Both History: New l e I K/A: 215005A101 4.0/4.0 l KA Statement: Reactor powerindication System / Evolution: Average Power Range Monitor / Local Power Range Monitor System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 l
| |
| l l
| |
| 1 I
| |
| l
| |
| | |
| Questi:n: 038 During a Unit 2 stanup and heatup in accordance with GP-2, " Normal Plant Startup", the operator is
| |
| *' cautioned to maintain turbine first stage pressure less than 100 psig during shell wamung.
| |
| Which of the following is the consequence of exceeding this value?
| |
| : a. The main turbine will roll off the turning gear.
| |
| : b. The main turbine trip reactor scram may occur.
| |
| : c. The main turbme will trip.
| |
| : d. The high pressure turbine shell heatup and expansion limits will be exceeded.
| |
| Answer: b ,
| |
| | |
| ==References:==
| |
| GP-2, " Normal Plant Startup", Rev. 83, Step 6.2.6 Caution, Page 60 LOT-1530," General Plant Procedures", Rev. 010, LO - 3 & 4 Exam Level: Both History: FEBQ #2959, Eraminer ModiSed (Reword stem, changed each distracter to more operationally oriented format)
| |
| K/A: 241000A313 3.0/3.0 KA Statement: Turbine stanup System / Evolution: Reactor /rurbine Pressure Rag"1= Hag System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 l
| |
| | |
| Question: 039 I
| |
| Given the following conditions:
| |
| The "A" loop of Residual Heat Rernoval (RHR) is bemg placed into Shutdown Cooling (SDC) using the"A" RHR Pump The RHR Torus Suction Valve (MO-13A) has been closed The Shutdown Cooling Suction Valves (MO-17 and 18) have been opened
| |
| - - The "A" RHR Pump SDC Suction Valve (MO-15A) has been opened No other actions have been taken
| |
| - An inadvertent high drywell pressure signal (2.0 psig) occurs With NO operator actions taken, what would be the expected RHR system (LPCI Mode) status?
| |
| : a. The "A" RHR Pump will start and will pump water from the Shutdown Cooling Suction to the reactor vessel via the LPCl injection Sowpath
| |
| : b. The Torus Suction Valve (MO-13 A) will open and all RHR pumps will inject usmg both the torus and the reactor vessel as water sources.
| |
| : c. The RHR Torus Suction Valve (MO-13A) and the pump SDC Suction Valve (MO-15A) will not reposition and the "A" RHR Pump will not start.
| |
| 3 d. The pump SDC Suction Valve (MO-15A) will close, the Torus Suction Valve (MO-13A)
| |
| , will open and the pump will start and inject Answer: c
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal", Rev. 010, Handout H-LOT-0370-2, Section 2, Page 1, LO - 5.a & b.
| |
| - Exam Level: Both History: New i K/A: 203000A308 4.1/4.1 ]
| |
| KA Statement: Systeminitiation sequence l l
| |
| System / Evolution: RHR/LPCI: InjectionMode Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| Questi:n: 040 Given the following conditions:
| |
| - The Unit 2 "B" Standby Liquid Control (SLC) Squib Valve was declared " Inoperable" on September 17,1997 at 0200
| |
| - The "A" SLC Pump failed its surveillance test on September 19,1997 at 0200
| |
| - No other actions have been taken Using the attached Tech Specs, determine when the plant is required to be in Mode 3.
| |
| : a. At 2200 on September 19,1997
| |
| : b. At 1400 on September 20,1997
| |
| : c. At 1400 on September 24,1997
| |
| : d. At 1400 on September 27,1997 Answer: c
| |
| | |
| ==References:==
| |
| T.S. 3.1.7 LOT-0310," Standby Liquid Control System", Rev. 011, LO - 9 & 10 Exam level: SRO History: New K/A: 211000G225 2.5/3.7 KA Statement: Knowledge ofbases in technical specdications for hmitmg conditions for operations and safetylimits System / Evolution: Standby Liquid Control System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| )
| |
| | |
| l
| |
| -)
| |
| ' Question: 069
| |
| ..Given the following conditions:
| |
| - . Unit 2 is operating at 90% power The High Pressure Coolant Injection (HPCI) system is operating for a surveillance
| |
| - Torus water temperature is 108 *F i
| |
| Which of the following describes the TRIP and Technical Specifications (TS) procedural requirements I for these conditions?
| |
| T-102, "Pnmary Co.ht Control", entry:
| |
| a is not required and the TS LCO is not met.
| |
| : b. is not required and the TS LCO is met.
| |
| : c. is required and the TS LCO is not met.
| |
| : d. is required and the TS LCO is met. l Answer: c t
| |
| I-)
| |
| | |
| ==References:==
| |
| T-102, "Pnmary Containment Control", Entry Conditions i
| |
| PBS T.S. 3.6.2.1," Suppression Pool Average Temperature", Page 3.6-23 LOT-1840, " Tech Spec LCOs", Rev. 007, LO - 2 i
| |
| Exam Ievel: Both History: New I K/A: 295013A201 3.8/4.0 KA Statement: Suppression pooltemperature System / Evolution: High Suppression PoolTemperature Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 1 L
| |
| | |
| Question: 070 2
| |
| ~
| |
| ./ A loss of feedwater heating has occurred on a unit requiring PCIOMR surveillance. OT-104," Positive Reactivity Addition" requires the operator to maintain power 20% below the initial value in accordance with GP-9.
| |
| Reducing power as directed, using GP-9, is done specdically to:
| |
| : a. avoid a reactor scram.
| |
| : b. limit the possibility of fuel failure due to Pellet Clad Interaction.
| |
| : c. reduce local power to increase the margm to the MAPRAT limit.
| |
| : d. reduce the subcooling of the feedwater entering the reactor.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| OT-104 Positive Reactivity Addition - Bases LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 4 & 5 Exam Level: SRO History: New
| |
| (.-
| |
| K/A: 295014K202 3.7/4.2 KA Statement: Fuelthemiallimits System / Evolution: Inadvertent Reactivity Addition Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 l
| |
| l
| |
| | |
| Questi:n: 071 Following a complete loss of off-site power with all of the diesel generators mnning normally with their
| |
| ~
| |
| output breakers closed, which of the following components will NOT have cooling water flow AVAILABLE? 1 I
| |
| l
| |
| : a. Control Rod Drive pumps
| |
| : b. Condensate pumps i
| |
| : c. HPCI room coolers
| |
| : d. RHR heat exchangers Answer: b
| |
| | |
| ==References:==
| |
| LOT-0430, " Turbine Building Closed Cooling Water System", Rev. 008, Sections II.A.5 & III.A, Page 5, LO - 2.b Exam Level: Both History: New -
| |
| K/A: 295018K202 3.4/3.6
| |
| (; '- KA Statement: Plant operations System / Evolution: Partial Or Complete Loss Of Component Cooling Water Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2
| |
| | |
| ?
| |
| Question: 072 ,
| |
| u --- During a failure-to-scram (ATWS) on Unit 2, reactor water level was deliberately lowered to control reactor power as required by T-117," Level / Power Control" Which of the following systems is specifically prohibited from maintaining reactor water level?
| |
| : a. Core Spray
| |
| : b. Feedwater
| |
| : c. Reactor CoreIsolation Cooling i
| |
| : d. LowPressureCoolantInjection Answer: .a
| |
| | |
| ==References:==
| |
| T-117, " Level / Power Control", Rev. I1, Step LQ-15 LOT-1560, "PBS TRIP Prevwbres", Rev. 007, LO - 3, 8 & 10 Exam Level: Both History: New iD K/A: 295015G420 3.3/4.0 KA Statement: Knowledge ofoperational implications of EOP warnings, cautions and notes System / Evolution: Incomplete SCRAM Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 L
| |
| | |
| Questi:n: 073 Given the following conditions:
| |
| ...:./.
| |
| - Unit 2 isin Mode 4
| |
| - No recirculation pumps are operating
| |
| - The "A" Loop of Residual Heat Removal was in Shutdown Cooling
| |
| -- The"A" RHR Pump hasjust tripped
| |
| - Reactor waterlevelis-5 inches
| |
| - No operator actions have been taken Which of the following describes the current status ofreactor coolant temperature indication?
| |
| l F a. Recirculation loop "A" temperature is a valid temperature indication.
| |
| 4 1"
| |
| ' b. There are no valid temperature indications available for these conditions.
| |
| : c. RWCU bottom head drain temperature is a valid temperature indication.
| |
| : d. Calculated temperature from steam dome pressure is a valid temperature indication Answer: . b -
| |
| i
| |
| | |
| ==References:==
| |
| ON-125 Loss Of Shutdown Cooling - Basec, Rev.1, Section 2.7.6, Page 7 p]
| |
| LOT-1550, " oft-Normal Procedures", Re v. 006, LO - 2 & 3 Exam Level: Both History: New
| |
| -K/A: 295021A204 3.6/3.5 KA Statement: Reactor water temperature System / Evolution: Loss Of Shutdown Coohng Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 3 SRO Group: 2 L I l
| |
| l
| |
| < i f
| |
| : u. i
| |
| | |
| l l
| |
| Question: 074 -l
| |
| ..~~
| |
| ; During a lowering torus water level transient, the operator is directed to shutdown the High Pressure ,
| |
| Coolant Injection (HPCI) system iflevel reaches 9.5 feet even if HPCI is required to assure adequate I core cooling. I This limitation willprevent:
| |
| : a. exceeding the maximum allowed primary containment pressure.
| |
| : b. exceeding the SRV Tail Pipe Limit. j
| |
| )
| |
| : c. a potentially unrecoverable tnp of HPCI on high turbine exhaust pressure.
| |
| : d. exceeding the Heat Capacity Level Limit.
| |
| l Answer: a
| |
| | |
| ==References:==
| |
| T-102 Pnmary Containment Control - Bases, Rev.12, Step T/L-13 Pages 7 & 8 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3 Eram Level: Both History: New (b
| |
| ~
| |
| K/A: 2950.'0K302 3.5/3.7 KA Statement: HPCI operation System / Evolution: Low Suppression PoolWaterI.evel Exam Section: Emsgg,cy And AbnormalPlant Evolutions RO Group: 2 SCO Group: 1 L
| |
| l t
| |
| | |
| i I
| |
| Q:esti:n: 075 Given the following conditions:
| |
| - Unit 2 had been operating at 100% power A loss of power has resulted in a full Group 1 isolation and reactor scram 3
| |
| - NO control rod movement was noted
| |
| - The power loss has also canW a loss of all nuclear instrumentation
| |
| - HPCI and RCIC are not available ,
| |
| - Reactor pressure is being mair*=i'wl between 950 and 1050 psig with 6 Safety Relief l Valves (SRV) open and a 7th SRV being cycled by the operator 1 What is the approximate reactor power for these conditiens?
| |
| I Poweris between:
| |
| l
| |
| : a. 7 - 15 %
| |
| : b. 20 - 28%
| |
| I
| |
| : c. 33 -41 %
| |
| !.- . d. 46 - 54%
| |
| '( ) ~
| |
| I Answer: c
| |
| | |
| ==References:==
| |
| LOT-0120, " Main Steam And Pressure Rehef System", Rev. 014, Section III.B.2, Page 10, LO - 6.f& 7.a Exam Level.: Both History: New K/A: 295007A202 4.1/4.1 KA Statement: Reactor power i
| |
| System / Evolution: High Reactor Pressure Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1 Justification: Per LOT-0120, each SRV is rated at approx. 5.7%, 6 SRVs open is approl 34 %
| |
| with 7th cycling about 3% more
| |
| | |
| Question: 076 .
| |
| The Main Control Room has just received alarm indications of a Cardox injection into the Cable Spreading Room.
| |
| What are the bases for the SE-2, "Cardox Injection into The Cable Spreading Room", Immediate Operator Actions?
| |
| These actions prepare both Units for;
| |
| : a. the expected loss of both Reactor Protection System motor generator sets.
| |
| : b. the possibility of performmg SE-1," Plant Shutdown From The Remote Shutdown Panel"
| |
| : c. the plant transient as systems become uncontrollable following Cardox discharge.
| |
| : d. the possibility of performing ON-114, " Actual Fire Reported In The Power Block, Diesel Generator Building, Emergency Pump, Inner Screen Or Emergency Cooling Tower Structures" Answer: c
| |
| ( 't
| |
| | |
| ==References:==
| |
| SE-2 Cardox Injection Into The Cable Spreadmg Room - Bases, Rev. 7, Immediate Operator Action 1, Page 1 LOT-1555, "Special Events", Rev. 004, LO - 3.a & 17 Exam Level: SRO History: New K/A: 600000K304 2.8/3.4 KA Statement: Actions contained in the abnormal procedure for plant fire on site System / Evolution: Plant Fire On Site Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 L
| |
| l
| |
| | |
| i i
| |
| i Questi:n: 077 i
| |
| I Given the following conditions: 1
| |
| -- The Main Control Room has been evacuated due to habitability concems
| |
| - There was no 6te or Cardcx system actuation j
| |
| -- No LOCA or electrical power concerns exist
| |
| - All immediate actions were successfully completed prior to the evacuation Select the system that is NOT required to meet the MINIMAL requirements for control of the reactors from the Remote Shutdown Panel following the scrams
| |
| : a. High Pressure Coolant Injection
| |
| : b. 125 VDC pover
| |
| : c. Reactor Core bolation Cooling l i
| |
| : d. Uninterruptible AC power Answer: a
| |
| | |
| ==References:==
| |
| SE-1 Bases, " Plant Shutdown From The Remote Shutdown Panel", Rev.15, Entry
| |
| .. Conditions, Page 1 LOT-1555, "Special Events", Rev. 004, LO - 2.a & q Exam Level: Both History: New K/A: 295016K201 4.4/4.5 KA Statement: Remote shutdown panel System / Evolution: Control Room Abandonment Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 1 1
| |
| j i
| |
| l l
| |
| i I
| |
| I i
| |
| | |
| Questi:n: 078 gj Given the following conditions
| |
| )- - Drywell sprays were initiated on Unit 3 as directed by the Pnmary Contamrnent Pressure leg of T-102, "Pnmary Contamment Control"
| |
| - As drywell pressure and temperature are decreasing the " Unsafe" region of the Drywell Spray Initiation Lunit (DSIL) curve was entered at a drywell temperature of 250 *F Which of the following is the REQUIRED action?
| |
| The operator shall:
| |
| : a. temunate drywell sprays immediately.
| |
| : b. throttle drywell spray Gow to remain in the " Safe" region of the DSIL curve.
| |
| : c. reduce drywell spray flow and monitor the Torus to Drywell vacuum breaker positions.
| |
| : d. termmate drywell sprays when drywell pressure swa below 2.0 psig. .
| |
| Answer: d
| |
| {h
| |
| | |
| ==References:==
| |
| T-102," Primary Containment Control", Rev. I1, Step PC/P-9 TRIP Curves, Tables & Limits - Bases, Rev. 2, Page 7 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Exam Level: Both History: New K/A: 295024K301 3.6/4.0 KA Statement: Drywell spray operation System / Evolution: High DrywellPressure hm Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1 L
| |
| | |
| Question: . 079
| |
| . , Given the following conditions:
| |
| - Unit 3 was operating at 100% power when a Group 1 isolation occurred on High Steam TunnelTemperature
| |
| - A failure to scram (ATWS)leR reactor power at 30%
| |
| - Attempts to inject boron were not successful
| |
| - Power dropped to 8% when T-240 was used to lower level and mamtam it in a band of
| |
| -172" to -195"
| |
| - Torus waterlevelis normal l
| |
| - Torus cooling was late in being started
| |
| - The required Emergency Blowdown was not performed until well aRer the attached Heat Capacity Temperature Limit (HCTL) curve was ~W
| |
| . i Which of the following i:; the direct result of this delayed Emergency Blowdown?
| |
| : a. Steam Dome pressure will exmed 1345 psig.
| |
| : b. The primary containment pressure limit will be ~M~I ]
| |
| : c. The SRV tailpipe hydrodynamic loadmg hmits will be ex-ded i T d. The max mn temperature will be exceeded
| |
| )
| |
| Answer: b
| |
| | |
| ==References:==
| |
| T-102," Primary Containmet Control", Rev.11, Step T/r-10 & 11 LOT-1360, "PBS TRIP Procedures", Rev. 007, LO - 3 Exam Imel: Both History: New l
| |
| K/A: 295026K301 3.8/4.1 I
| |
| KA Statement: Emergency / normal depressurization I
| |
| System / Evolution: Suppression pool high water temperature ;
| |
| l Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Genup 1
| |
| | |
| ~ .:..
| |
| CURVE T/T-1 NEAT CAPACITY TEMP LIMIT 250 l II
| |
| * 250 .
| |
| I I I zu ' ...'
| |
| \1 I I E 230 -
| |
| UNSAFE __
| |
| c,' *i 7 a 220--
| |
| h '
| |
| (BLONDOWN REQUIRED) '
| |
| * 210 200 SAFE N w g
| |
| E 130 \
| |
| 180 0
| |
| 200 400 800 000 1000 1200 REACTOR PRESEURE (PEIO) l
| |
| | |
| i
| |
| ~ Questi:n: 080 Given the following conditions:
| |
| . .e !
| |
| - Unit 2 is operating at 75% power
| |
| - An unexplained Refuel Floor Ventilation Exlaust high radiation alarm has been received
| |
| - T-103, " Secondary Containment Control", has been entered and directs the operator to venfy Reactor Building and Refuel Floor ventilation isolabons and that Standby Gas Treatment hasinitiated What is the bases for this verification?
| |
| This verification ensures the secondary containment
| |
| : a. atmosphere will be treated and controlled as an elevated release.
| |
| : b. atmosphere will be treated and controlled as a ground level release.
| |
| : c. atmosphere will be held up for 90 hours and controlled as an elevated release.
| |
| : d. atmosphere will be held up r or90 hours and controlled as a ground level release "i Answer: a
| |
| | |
| ==References:==
| |
| T-103 Secondary Containment Control - Bases, Rev. 9, Step SCC-1, Page 5 LOT-1560,"PBS TRIP Procedures", Rev, 007, LO - 1 & 3 Exam Level: Both History: New K/A: ' 295034A104 4.1/4.2 KA Statement: SBGT/FRVS System / Evolution: Secondary Containment Ventilation High Radianon .
| |
| Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 L
| |
| I 1
| |
| 1
| |
| | |
| Question: 081-9 Given the following conditions:
| |
| -n;r
| |
| - Unit 2 had been operating at 100% power
| |
| - An Electro-Hydraulic Control (EHC) sy stem logic failure caused the Main Turb'me Stop Valves to close and the Main Tudxne Bypass Valves to remain closed
| |
| -- Reactor pressure peaked at 1340 psig at which time the reactor iciiu..r.ed on high flux Which of the following describes this trannent?
| |
| : a. A safety limit violation occurred and the Main Turbine Stop Valve closure was NOT the only RPS trip failure.
| |
| b; A safety limit violation occurred and the Main Turbine Stop Valve closure was the only RPS trip faihare
| |
| : c. A safety limit violation did NOT occur and the Main Turbine Stop Valve closure was NOT the only RPS trip failure.
| |
| : d. A safety limit violation did NOT occur and the Main Turbine Stop Valve closure was the only RPS trip failure Answer: a
| |
| ~
| |
| | |
| ==References:==
| |
| PBS T.S. 2.1.2, " Reactor Coolant System Pressure SL", Page 2.0-1 LOT-0300, " Reactor Protection System", Rev. 013, Figure 7
| |
| . LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 1 & 3 Exam Imel: SRO History: New K/A: 295025G222 3.4/4.1 KA Statement: Knowledge oflimiting condmons for operations and safety hnuts Systena/ Evolution: High Reactor Pressure Exam Section: Emergency And Abisviu.id Plant Evoh:tions RO Group: 1 SRO Group: 1 L
| |
| l
| |
| | |
| l Question: 082 ,
| |
| . i
| |
| !... Which of the following is the reason why the Main Steam Isolation Valves (MSIV) are closed prior to evacuating the Main Control Room in accordance with SE-1, " Plant Shutdown From The Remote Shutdown Panel"? .
| |
| i a. With MSIVs closed, all reactor inventory and pressure control may take place at the Remote Shutdown Panel. I
| |
| : b. Since plant release points cannot be monitored at the Remote Shutdown Panel, closing the MSIVs precludes any concem for off-site releases. ,
| |
| : c. The MSIV closure outside the Main Control Room requires access to plant areas that may not be accessible during an evacuation.
| |
| : d. If the MSIVs are closed from outside the Main Control Room, there is no method for verification ofcomplete closure.
| |
| Answer: a -
| |
| | |
| ==References:==
| |
| SE-1 Plant Shutdown From The Remote Shutdown Panel - Bases, Rev.15, ImW:re Action 5, Page 2 i l LOT-1555, "Special Events", Rev. 004, LO - 1.c & 2.n Exam Level: SRO History: New I K/A: 295016G411 3.4/3.6 KA Statement: Knowledge of abnormal condition procedures System / Evolution: ControlRoom Abs.ndonment Exam Section: Emergency And AbnormalPlantEvolutions RO Group: 2 SRO Group: 1 1
| |
| I t
| |
| | |
| Question: 083 .
| |
| t
| |
| %.- Given the following conditions:
| |
| -- Unit 2 was performing a normal shutdown During the shutdown a loss of feedwater resulted in a scram
| |
| - A hydraulic ATWS occurred and power stabilized at 10%
| |
| -- The operator was directed to contmue control rod insertions -
| |
| . Which of the following describes how the control rods should be inserted for these conditions?
| |
| The operator should:
| |
| : a. remove the group scram fuses in the RPS cabinets.
| |
| : b. locally vent the Scram Air Header,
| |
| : c. trip the RPS power supply breakers.
| |
| : d. bypass the Rod Worth Minimiw and insert control rods using " Emergency In" Answer: d
| |
| [9 '
| |
| | |
| ==References:==
| |
| LOT-0080, " Reactor Manual Control System", Rev. 007, Section IV.A.2.d, Page 11, LO - 2.b & 4.e Exam Level: Both - History: FEBQ #3637, Evaminer Modi 6ed (Reword stem, change format, change 2 distracters)-
| |
| K/A: 295015A103 3.6G.8 KA Statement: RMCS System / Evolution: Incom#e SCRAM Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1
| |
| .. r
| |
| | |
| Questi::n: 084 Unit 3 was operating at 100% power whm a transient occurred.-
| |
| The following are the peak radiation leve's which existed on Unit 3 during the transient:
| |
| - Main Steam Line Radiation - 2000 mr/hr .
| |
| -Main Stack Radiation- 20 cps
| |
| - Steam Jet Air Ejector Discharge Radation - 20 mc/hr
| |
| - Reactor Building Ventilation Exhaust Radation - 20 mr/hr Which of the following components should have received a PCIS signal in direct response to these radiation levels? 3
| |
| : a. OffGas Lineisolationvalve
| |
| : b. Main Steamisolationvalves
| |
| : c. Mechanical Vacuum Pump suction valve
| |
| : d. Standby Gas Treatment filter train isolation valves Answer: d-
| |
| : g. N
| |
| | |
| ==References:==
| |
| LOT-0180, "Pnmary Containment Isolation System", Rev. 009, Secuons II.D.8 &
| |
| - IV.A2, Page 14, LO - 2.j Exam Level: Both History: New K/A: 295020K211 3.2/3.4 KA Statement: Standby gas treatment system /FRVS System / Evolution: Inadvertent C*ah Isolation Exam Section: Emergency And AbnormalPlant Evolunons RO Group: 2 SRO Group: 2 m e
| |
| | |
| Question: 085 .
| |
| Given the following conditions:
| |
| - The E-3 Diesel Generator (DG)is running in parallel with the E-32 Bus for surveillance testing
| |
| - A loss of 125 VDC Panel 30D23 supplying the E-3 DG has occurred.
| |
| Which of the following describes the expected status of the diesel for this failure?
| |
| : a. The DG output breaker will trip and the engine will overspeed but will NOT trip.
| |
| : b. DG voltage will decrease due to a loss of the field flash supply.
| |
| : c. All DG alarms will be disabled and the engine will wi*3y trip
| |
| : d. The DG will remain at speed and fully loaded Answer: d
| |
| | |
| ==References:==
| |
| LOT-0670, " Diesel Generator And Auxiharies", Rev. 006, Section IV.H.3, Page 50, LO - 2.b & 6.f
| |
| ( '', ham Level: Both History: New K/A: 295004A102 3.8/4.1 KA Statement: Systems necessary to assure safe plant shutdown System / Evolution: Partial Or Com#e Loss ofD.C. Power Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 L
| |
| | |
| Questi:n: 086 During performance of ON-119, " Loss OfInstmment Air", what is the specific reason why an instrument air pressure of 75 psig was chosen as the limit at which GP-3," Fast Power Reduction" must be initiated?
| |
| This pressure was chosen to:
| |
| : a. reduce power to provide a safety margin due to the potential loss of feedwater heating.
| |
| : b. prevent the unit with the failing air system from affectmg the other unit once the header cross-connect valves are opened
| |
| : c. begm the power reduction well before any of the expected random control rod drifts begia
| |
| : d. avoid the potential loss of the main condenser on a Group I isolation signal with the plant at high power.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| ON-119 Loss OfInstmment Air - Bases, Rev.10, Operator Action #2, Page 2 r -- LOT-1550, "OfF-Normal Procedures", Rev. 006, LO - 2 & 3
| |
| ; 1 Exam Level: Both History: New K/A: 295019A202 3.6/3.7 KA Statement: Status of safety-related instmment air system loads System / Evolution: Partial Or Complete Loss OfInstrument Air Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2 L
| |
| I l
| |
| l
| |
| | |
| Question: 087 .
| |
| Given the following conditions:
| |
| Unit 3 is operating at 40% power
| |
| - The "B" Stator Water Cooling (SWC) Pump is out of service for maintenance
| |
| - The"A" SWC Pump hasjust tripped All automatic actions are occumng as designed
| |
| - The Turbine Building Operator reports the "A" SWC Pump will not restart Which of the following are all the required operator actions for these conditions:
| |
| : a. Verify both Recirculation Pumps trip and place the Reactor Mode Switch to " Shutdown"
| |
| : b. Verify the "A". Recirculation Pump trips and power stabilizes within single loop limits.
| |
| : c. Verify turbine generator runback, reduce VARS to minimum and verify the "A" Rectreulation Pump trips.
| |
| : d. Verify turbine generator runback and reduce VARS to minimum.
| |
| Answer: d f i
| |
| | |
| ==References:==
| |
| OT-ll3," Loss Of Stator Cooling", Rev. 4, Section 2.3, Page 2 LOT-0600, "Mam Generator And Amaharies", Rev 008, LO. - 2.i & 6.d Exam level: Both History: New K/A: 295005A205 3.8/3.9 KA Statement: Reactor power System / Evolution: Main Tuitune GeneratorT:ip Exam Section: Emergency And AbnormalPlant Evolunons RO Group: 1 SRO Group: 2 L
| |
| | |
| Questi:n: 088 gj Given the following initial conditions:
| |
| 1
| |
| - Unit 2 is operating at 80% power ;
| |
| - 'A' and 'C' Condensate pumps are runmng ;
| |
| - All three Reactor Feed Pumps are maintainmg reactor level l
| |
| The following transient occurs:
| |
| 1
| |
| - The 'A' Condensate Pump trips
| |
| - Reactorlevelis 19" and dropping i Which of the following actions should be taken immediately?
| |
| : a. Verify the recirculation system automatic runback to 45% speed
| |
| : b. Reduce power in accordance with GP-9-2," Fast Reactor Power Reduction"
| |
| : c. Take manual control of the Reactor Feed Pump Master Level Controller
| |
| : d. Select all three Reactor Feed Pumps to MSC Control g3 Answer: b
| |
| | |
| ==References:==
| |
| OT-100, " Reactor Low Level", Rev. 9, Section 1.0, Page 1 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 1 Exam Level: Both History: New K/A: 295009K202 3.9/3.9 KA Statement: Reactor waterlevel control System / Evolution: Low Reactor Water Level Exam Section: Emergency And Abnormal Plant Evo'ations RO Group: 1 SRO Group: 1 i
| |
| em $
| |
| l
| |
| | |
| Question: 089 .
| |
| Unit 2 has experienced a LOCA which resulted in the Loss of Adequate Core Cooling for 10 minutes- l Containment Parameters are currently:
| |
| - Drywell Pressure - 12 psig
| |
| - Torus Pressure- 10 psig
| |
| -DrywellTemperature-225 F
| |
| - Drywell Hydrogen - 7.3%
| |
| - Drywell Oxygen- 6%
| |
| - Torus Level- 16.5 feet Under these conditions, T-102, "Pnmary Containment Contror', directs that the primary containment be vented and purged regardless of off-site release rates.
| |
| The intent of this action is to:
| |
| : a. limit the total radioactive release due to an anticipated containment failure.
| |
| : b. rapidly lower contamment pressure to prevent acMag the Containment Vent Pressure.
| |
| T T c. limit the total radioactive release by venting before the torus vacuum breakers are submerged.
| |
| : d. rapidly lower containment pressure since the pressure suppression function of the torus has been bypassed Answer: a
| |
| | |
| ==References:==
| |
| T-102 Pnmary Contamment Control - Bases, Rev.12, Steps PC/H 14, Page 26 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Eumm Imel: SRO History: New K/A: 500000K207 3.2/3.7 KA Statement: Drywellvent system ,.
| |
| Systena/ Evolution: High Containment Hydrogen Concentration Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1
| |
| | |
| Question: 090 w Given the following conditions:
| |
| - Unit 2 is operating at 25% power
| |
| - All plant systems are operating as designed for this power level
| |
| - Main condenser vacuum is 28.5 inches Hg Vac and is slowly lowering No operator actions are taken ]
| |
| l Which of the following is the expected sequence of the plant response?
| |
| : a. Reactor sen 2n, Main Tu6ine trip, Reactor Feedwater Pump trip, Main Tukine Bypass Valves close
| |
| : b. Main Turbine trip, Reactor scram, Reactor Feedwater Pump trip, Main Tuhine Bypass Valves close
| |
| : c. Reactor scram, Main Turbme Bypass Valves close, Reactor Feedwater Pump trip, Main Turbine trip.
| |
| : d. Main Turbine trip, Reactor scram, Main Turbine Bypass Valves close, Reactor Feedwater Pump trip Answer: a
| |
| (' }
| |
| | |
| ==References:==
| |
| OT-106, "CW=r Low Vacuum", Rev. 2, Step 4.0, Page 9 LOT-1540, " Operational Tranment Procedures", Rev. 005, LO - 2 & 5 Exam Lael: Both History: New K/A: 295002G448 3.5/3.8 KA Statement: Abihty to interpret control room ichs to venfy the status and operation of system and understand how operator actions and directrves affect plant and system cW ,
| |
| System / Evolution: Loss OfMain Cher Vacuum Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group (2
| |
| | |
| CURVE T/L-3 SRY TAIL PIPE LIMIT 17.8 , , , , ,
| |
| 17.4 !
| |
| UNSAFE-
| |
| ! I i -
| |
| N I , ,
| |
| ' ''I '
| |
| 17 0 ' i ii i i\ t 16.6 !' I d l SAr i ;
| |
| \
| |
| g 16.2 , , , , , , g l l i i i 15.8
| |
| \
| |
| g , , , , , i g ac I
| |
| * 33 4
| |
| , e i ! l
| |
| \1 i i i \,
| |
| . i 15.0-- e i u
| |
| 1 i 1 i 0 200 400 600 800 1000
| |
| - RFY PRESSURE (PSIC)
| |
| .~
| |
| 1 i
| |
| l l
| |
| l
| |
| | |
| Question: 091
| |
| ,"' j While experiencing HIGH torus water level control problems, the Unit Reactor Operator inadvertently i opens one Safety Relief Valve (SRV) while operating in the " Unsafe" region of the attached SRV Tail .
| |
| Pipe Limit curve.
| |
| Which of the following could be expected to result from this curve violation?
| |
| Openmg the SRV under these conditions:
| |
| : a. results in excessive hydrodynamic loadmg due to the excessive water levels in the tailpipe.
| |
| : b. will result in direct suppression d 44 presmuu.dori due to the p= hen being uncovered.
| |
| : c. results in drawing water up into the tailpipe because the vacuum breakers are under water.
| |
| : d. will result in valve seat damage from the excessive flowrates due to the reduced torus waterlevelin the tailpipe.
| |
| Answer: a g; -)
| |
| | |
| ==References:==
| |
| TRIP Curves, Tables & Limits - Bases, Rev. 2, SRV Tail Pipe Limit, Page 11 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 6 & 8
| |
| ]
| |
| i Exam level: Both History: 'New
| |
| ~
| |
| K/A: 295029K2% 3.4/3.5 KA Statement: SRV's and discharge pip'm g l System / Evolution: High Suppression PoolWater Level Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 l
| |
| | |
| Questi:n: 092 Given the following conditions:
| |
| - Unit 3 has entered T-103," Secondary Containment Control" after receiving an alann for high waterlevelin the HPCI Room Which of the following could be used to make the determination that water level is at or above the
| |
| " Action" level WITHOUT physically entering the room.
| |
| Water level may be considered to be above the " Action" level:
| |
| : a. . by computer point verifications ofECCS room levels on SPDS.
| |
| : b. ifone ReactorBuilding floor drain sump pump is runnmg.
| |
| : c. ifbattery grounds are received with indications of a HPCI Cardox discharge.
| |
| : d. if the Turbine Building floor drain sump high-high level alann is received.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| T-103 Secondary Contamment Control - Bases, Rev. 9, Entry Conditions, Pages 2 & 3 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 2,3 & 8 Exam level: Both History: New K/A: 295036A202 3.1/3.1 KA Statement: Waterlevelin the affected area System / Evolution: Sacnndary Containment High Sump / Area Water Level Exam Section: Erregecy And Abnormal Plant Evolutions RO Group: 3 SRO Gmup: 2
| |
| -L
| |
| | |
| Question: 093 Given the following conditions:.
| |
| - Unit 2 is operating at 25% power
| |
| - A feedwater level control malfunction results in a apidly increasing reactor water level Select the conditions REQUIRING the operator to scram the reactor and close the Main Steam !
| |
| Isolation Valves. l l
| |
| Reactor waterlevelis:
| |
| : a. at 100 inches and increasing on Refuel Range level instrument LR-97.
| |
| : b. at 95 inches and increasing on Refuel Range levelinstmment LI-86.
| |
| : c. at 55 inches on Narrow Range level instmments.
| |
| : d. at the top of scale on Wide Range level instruments.
| |
| Answer: b i
| |
| iO
| |
| | |
| ==References:==
| |
| OT-110 Reactor High Level- Bases, Rev. 6, Step 3.1 & Notes, Pages 2 & 3 1
| |
| LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 2,- 3 & 4 l 1
| |
| Exam Level: Both History: New l
| |
| K/A: 295008K101 3.0/3.2 KA Statement: Mo'sture t carryover System / Evolution: High Reactor Water Level .
| |
| Exam Section: Emergency And Abra,6r ! Plant Evolutions RO Group: .2 SRO Gmup: 2 Justification: Rapid overfeedmg transients require this as an immediate actiort LR-97 is valvad out at power L
| |
| | |
| . Question: .094
| |
| ;. Given the following conditions:
| |
| .w
| |
| - Unit 3 is operating at 85% power
| |
| - The Unit Reactor Operator hasjust reported slowly increasing drywell temperatures Investigation has determined that the drywell chilled water system is lined up such that the Reactor Building Closed Cooling Water system is supplying cooling water Which of the following has occurred?
| |
| : a. Service water cooling to am chdler conocaws has been lost.
| |
| : b. An undervoltage condition has occurred on two of the three busses feeding the chillers.
| |
| : c. A loss of power to the chdied water header containment isolation valves has occurred
| |
| : d. Two of the three chillers have tripped on low chilled water flow.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0150, "Drywell Chilled Water System", Rev. 005, Sections IV.C.2 & V.D, Pages 15 & 20, LO - 2.e & 7.e Exam Level: Both . History: New
| |
| --K/A: 295012A102 3.8/3.8 KA Statement: Drywell cooling system System / Evolution: High DrywellTemperature Exam Section: Emergency And Abnormal Plant Evah*ians RO Group:
| |
| 2 SRO Group: 2
| |
| | |
| Question: 095
| |
| ./ Which of the following is the basis for requiring a reactor scram with no Control Rod Drive (CRD)
| |
| Pumps mnning during a reactor startup?
| |
| 1 The reactoris scrammed 1
| |
| : a. to ensure all controls rods are fully inserted before overheating (un affect the mechanism seals andimpact on the scram times
| |
| : b. since no other control rod motion is available (inst:rt/ withdrawal) withcnit the CRD system in op: ration.
| |
| : c. to ensure there is no loss of scram @ility as the hydraulic control unit ecm=dn' ors depressurize
| |
| ]
| |
| : d. in anticipation of tripping the Recirculation Pumps due to potential pump seal package ]
| |
| damage from theloss ofCRD flow. l l
| |
| Answer: c j
| |
| | |
| ==References:==
| |
| ON-107 Loss Of CRD Reieng Function - Bases, Rev. 6, Sections 2.1 & 2.2, Pages i 1&2 l LOT-1550, "OfF-Normal Procedures", Rev. 006, LO - 2 & 3 Exam Ievel: Both History: New i
| |
| K/A: 295022K301 3.7/3.9 l I
| |
| KA Statement:
| |
| Reactor SCRAM System / Evolution: Loss OfCRD Pumps Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 i
| |
| supp
| |
| | |
| Question: 096 .
| |
| Given the following conditions:
| |
| -- Unit 2 had been operating at 80% power
| |
| - - Due to a leak, a high drywell pressure condition occurred
| |
| - All expected automatic actions occurred Which of the following will be the MAXIMUM expected Reactor Building pressure for these conditions?
| |
| : a. -2.0 inches WG
| |
| : b. -0.25 inches WG
| |
| : c. 0.00 inches WG
| |
| : d. +1.25 inches WG Answer: b
| |
| | |
| ==References:==
| |
| I OT-0210, " Standby Gas Treatment System", Rev. 009, Section I.A & B, Page 8, LO
| |
| - 1.a & 5.a Exam Level: Both History: New K/A: 295035K202 3.6/3.8 KA Statement: SBGT/FRVS System / Evolution': Secondary Contamment High Differential Pressure Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 3 SRO Group: 2 L
| |
| | |
| Question: 097 s
| |
| Given the following conditions:
| |
| . Unit 2 is operating at 100% power
| |
| - - The Main Control Room hasjust received Vent Stack Rad Monitor high radiation alarms
| |
| - Further investigation shows the source of the activity to be from the Reactor Building
| |
| -- There are NO secondary containment high temperature alarms present
| |
| - ON-104, " Vent Stack High Radiation", directs the operator to place Reactor Building Area Ventdation on the Standby Gas Treatment (SBGT) system Select the concern for long tenn operation under these conditions. Assume no additional operator actions are taken.
| |
| : a. Reactor Building high pressure from lower SBGT system flows
| |
| : b. Potential loss of control room nonnal ventdation due to high radiation
| |
| : c. Reactor Building high pressure from spiral ductwork failure
| |
| : d. Potential loss of the main condenser on Main Steam Isolation Valve closure Answer: d
| |
| , a
| |
| | |
| ==References:==
| |
| ON-104 Vent Stack High Radiation - Bases, Rev. 8, Step I-2 Note, Page 3 LOT-1550, "OfF-Normal Procedures", Rev. 006, LO - 2 & 3 Exam Level: SRO History: New ,
| |
| K/A: 295017Al11 3.9/4.1 .
| |
| l l
| |
| KA Statement: PCIS/NSSSS System / Evolution: High OfF-Site Release Rate Exam Section: Emergency And Abrsed Plant Evolutions RO Group: 2 SRO Group: 1
| |
| | |
| Question: -098 ,
| |
| 4.3 Given the following conditions:
| |
| i
| |
| - Unit 3 is performing a plant shutdown per GP-3, " Normal Plant Shutdown"
| |
| - The reactor is being depressurized via the main turbine bypass valves 4
| |
| - The Unit Reactor Operator (URO) notes that reactor water narrow range level indication is
| |
| " notching" .
| |
| - Which of the following represents the most accurate indicated water level from this " notching" level l indicator?
| |
| : a. Water level at the top of the " notch"
| |
| : b. Water level at the bottom of the " notch" i
| |
| : c. An average of the water levels from all indicators that are " notching".
| |
| : d. An average of the water levels from the top and bottom of the " notch" l Answer: b
| |
| | |
| ==References:==
| |
| GP-3, " Normal Plant Shutdown", Rev. 75, Section 6.79 Caution and Appendix 1, Pages 28 & 40-42 LOT-1530, " General Plant Procedures", Rev. 010, LO - 3 & 4 Eum icel: SRO History: New K/A: 295006A203 4.0/4.2 KA Statement: Reactor waterlevel System / Evolution: SCRAM l
| |
| Eu? Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 '
| |
| Justincat ion: "b" correct W== the " notching" is an irviirstad increase in level from actual level then a return to that level on the instrument
| |
| | |
| l i
| |
| QuestiIn: 099 i The torus water level control leg of T-102, " Primary Containment Control", directs the operator to l emergency blowdown if torus level cannot be maintained in the safe zone of the attached Heat Capacity Level Limit (HCLL) curve.
| |
| Which of the following is the reason for operatmg within these limits?
| |
| 1 i
| |
| I This levellimit ensures the:
| |
| : a. Heat Capacity Temperature Linut will not be exceeded dunng the " blowdown" phase of a design bases loss of coolant accident..
| |
| : b. RPV Blowdown Limit will not be ercW dunng the" blowdown" phase of a design !
| |
| basesloss ofcoolant accident. I
| |
| : c. Heat Capacity Temperature Limit will not be exceeded during a reactor depressurization via the Safety ReliefValves.
| |
| : d. RPV Blowdown Limit will not be exceeded on a reactor depressunzation via the Safety ReliefValves. j Answer: c
| |
| | |
| ==References:==
| |
| TRIP Curves, Tables & Linuts - Bases, Rev. 2, Heat Capacity Level Lumt, Page 9 l LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 4 & 6 l l
| |
| Exam level: SRO History: New ]
| |
| K/A: 295026K102 3.5/3.8 KA Statement: Steam condensation System / Evolution: Suppression Pool High Water Temperature Exam Section: Emergency And AbruiruelPlant Evolutions RO Group: 2 SRO Group: 1 L
| |
| : i. CURVE T /L -1 HEAT'CAPAC:TY LEVEL LIMIT J 16
| |
| - 15 u.
| |
| ~ 14 d SAFE C 13
| |
| \N
| |
| ' 1
| |
| $ i g UNSAFE I1 -dREDU1REDI eLovocom \,\
| |
| ' ' I I 20 . ! ! !
| |
| D 20 40 60 80 8I HC ( El 6T HC IN OF OF HEAT CAPACITY TEWP LIMI T FROW CURVE T/T-1 WI NUS ( -) 0F TORUS TEMP EDUALS ( ) of 6T HC M
| |
| | |
| Question: 100 Gven the following conditions:
| |
| - Unit 2 is operating at 100% power All plant systems are operating as designed A loss of all pressure dgnals to the Digital Feedwater Control System hasjust occurred
| |
| - No operator actions are taken Select the expected unit response for these conditions:
| |
| : a. In6cated Narrow Range reactor water level will go upscale due to the loss of the compensating pressure signal.
| |
| : b. Indicated Narrow Range reactor water level will go downscale due to the loss of the compensating pressure signal.
| |
| : c. Indicated Narrow range reactor water level will read slightly lower than actual due to the default pressure value.
| |
| . d. Indicated Narrow Range reactor water level will read slightly higher than actual due to the default pressure value.
| |
| (' ' , Answer: d i
| |
| | |
| ==References:==
| |
| LOT-0550, "Feedwater Control System", Rev. 006, Section IV.B.9.c.2)a), Page 19, LO - 3.a & 7.a Exam Level: SRO History: New K/A: 295025K106 3.5/3.6 KA Statement: Pressure effects on reactor ween level System / Evolution: High Reactor Pressure Exam Section: EHwgwcy And AbnormalPlant Evolubons RO Group: 1 SRO Group: 1 L
| |
| l i
| |
| I l
| |
| | |
| ATTACHMENT 3 l
| |
| \
| |
| 4 I
| |
| i l
| |
| l 1
| |
| I 1
| |
| | |
| I .
| |
| Questizn: 001 .
| |
| Which of the following evolutions shall be performed under the direct supervision of a Licensed Senior
| |
| ~
| |
| Reactor Operator.
| |
| : a. Transfer ofEHC Pressure Regulators
| |
| : b. Placement ofRecire MG Set scoop tube adjustable mechanical stops.
| |
| : c. Local Recirculation Pump MG Set scoop tube operations
| |
| : d. Fuel shufHe within the vessel (no fuelis removed from the vessel)
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0005, " Licensee Obligations And Responsibilities", Rev. 007, Section II.C.1, Page 6, LO - 1 Exam Level: Both History: New K/A: 29400lG102 3.0/4.0 KA Statement: Knowledge of operator responsibilities during all modes ofplant operation System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| L l
| |
| l i
| |
| i l
| |
| I
| |
| | |
| Question: 002 Select th'e conditions under which a non-licensed operator (NLO) may place clearance tags on Main
| |
| ~~
| |
| Control Room switches. j l
| |
| The NLO may: l
| |
| : a. on'.y place tags on non-safety related systems in the Main Control Room.
| |
| : b. place tags after notifying the appropnate reactor operator. l
| |
| : c. place tags but they must be independently verified by a licensed operator.
| |
| : d. only place tags in the Main Control Room if the Unit is at or below Mode 4.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| OM-P-3.3, " Licensed Operators", Rev. 5 Section 1.1.16 Page 6 LOT-0006, "OM Chapters 0 - 5", Rev. 000, LO - 3 -
| |
| Exam Level: RO History: New l i
| |
| (-
| |
| . K/A: 294001G131 4.2/3.9 KA Statement: Ability to locate control room switches, controls and indications and to determme i that they are correctly reflectmg the desired plant lineup.
| |
| System / Evolution: ;
| |
| 1 Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| )
| |
| w
| |
| | |
| Question: 003 .
| |
| Which of the following evolutions require Unit 3 to have two (2) Licensed Operators assigned prior to starting?
| |
| Unit 3 is:
| |
| : a. performing an immediate shutdown required by Technical Specifications.
| |
| : b. stroking rods for CRD testing.
| |
| : c. raising power from 10% to 15% with control rods after placmg the Reactor Mode Switch b "h"
| |
| : d. raising power with recirc from 50% to 60%.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| OM.P-3.3," Licensed Operators", Rev. 4, Section 6.2 & 6.3, Page 11 LOT-0006, "OM Chapters 0 - 5", Rev. 000, LO - 3
| |
| ..g Exam Imel: Both History: New K/A: 29400lG201 3.7/3.6 KA Statement: Ability to perform pre-stanup procedures for the facility, i@ ding operatmg those controls associated with plant equipment that could affect reactivity.
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| L
| |
| | |
| Questi:n: 004 What is the maximum amount of time the Unit Reactor Operator (URO)"at-the-controls" can be provided with a temporary reliefi e , a complete turnover does not have to be done?
| |
| : a. 15 minutes
| |
| : b. 30 minutes
| |
| : c. 60 minutes
| |
| : d. 90 minutes Answer: c
| |
| | |
| ==References:==
| |
| OM-C-6.2, " Temporary Relief", Rev.1, Section 2.2.1, Page 2 ;
| |
| 1 LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - 1 Exam Level: Both History: New K/A: 294001G103 3.0/3.4 g ., KA Statement: Knowledge of shift turnover practices l
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| l O
| |
| ~
| |
| | |
| Question: 005 Given the following conditions:
| |
| - With Unit 2 operating at 50% power, a packing leak is discovered on an accessible motor operated valve in a safety-related system
| |
| - The leak is not severe and it has been decided to backseat the valve during the next shift
| |
| - All plant systems are operating as designed Which of the following describes how this valve should be bekwied?
| |
| : a. The appropriate System Manger
| |
| * auld maranally backseat the valve using TMT.
| |
| : b. The Operator in the Main Control Room should electrically Mek=t the valve.
| |
| : c. Maintenance personnel should manually backseat the valve.
| |
| : d. A Equipment operator at the motor control center should electrically backseat the valve.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| OM-C-7.5, " Valves", Rec. 3, Section 2.4, Page 2 LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - None Identified 7 )
| |
| Exam Imel: Both History: New K/A: 294001G130 3.9/3.4 KA Statement: Ability to locate and operate components, ineWing local controls.
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| | |
| Questi2n: 006 )
| |
| Which of the following describes the limitations regarding operatic.i of a plant component that has ONLY an Equipment Status Tag (EST) installed?
| |
| I i
| |
| The plant component:
| |
| x 2
| |
| : a. may not be operated until the EST has been removed.
| |
| : b. may be operated with permission of the applicable Unit Reactor Operator.
| |
| : c. may not be operated unless it is a part cf the Operations Department daily rounds. 1 l
| |
| : d. may be operated only if a Double Verification is used to verify all operator actions. )
| |
| l Answer: b 1
| |
| | |
| ==References:==
| |
| OM-C-10.6, " Equipment Status Tags", Rev. 3, Section 4.1, Page 5 l l
| |
| LOT-0008, "OM Chapters 10 - 15", Rev. 000, LO - 1
| |
| )
| |
| Exam Level: RO History: New
| |
| ) 3.6/3.8
| |
| / K/A: 294001G213 KA Statement: Knowledge of tagging and clearance procedures l 1
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| .- m
| |
| | |
| Question: 007 .
| |
| ' Prior to a reactor stanup, a check-offlist (COL)is being performed on the Unit 3 "A" Residual Heat Removal (RHR) Loop. During performance, the operator has discovered that a step in the COL has an incorrect component identification designation.
| |
| Select the appropriate actions for these conditions.
| |
| The operator shall:
| |
| : a. not complete the COL until a temporary change is prepared in accordance with A-3.
| |
| : b. document the problem with that COL step and continue to completion of the COL with Shift Management approval.
| |
| : c. make a note of the discrepancy on the specific COL step, initial and date the step and continue to completionin the COL.
| |
| : d. have an imm~L etDouble Verification performed on the COL step and the component, and then complete the COL with Shift Management approval.
| |
| Ansur: b
| |
| | |
| ==References:==
| |
| OM-C-10.7, ' Check-OffLists", Rev.1, Section 5.2, Page 5 LOT-0008, "OM Chapters 10 - 15", Rev. 000, LO - 1 Exam Level: Both History: New K/A: 29400lG129 3.4/3.3 KA Statement: Knowledge of how to conduct and venfy valve lineups.
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| l 1
| |
| | |
| ' Question: 008 Given the following conditions:
| |
| p - A male, fully qualified radiation worker at Peach Bottom hasjust returned from 4 weeks of outage support at Limerick
| |
| - Total Effective Dose Equivalent (TEDE) received at Limerick was 250 mrem. ~
| |
| This workers' current TEDE from Peach Bottom for 1997 is 225 mrem What is the MAXIMUM annual non-emergency Total Effective Dose Equivalent (TEDE) that can be received at Peach Bottom for the remai* of 1997 WITHOUT Wino the Federal Exposure Limits.
| |
| : a. 4475 mrem
| |
| : b. 4525 mrem
| |
| : c. 4750 mrem
| |
| : d. 4775 mrem .
| |
| Answer: b
| |
| | |
| ==References:==
| |
| HP-C106, " Dosimetry Program", Rev. 3, Sections 7.1.1, Page 3 I
| |
| )
| |
| 10CFR20.1201, Occupational Dose Limits for Aduhs LOT-1730, " Radiation Exposure Limita", Rev. 012, LO. - 2 Exam level: Both History: New K/A: 29400lG304 2.5/3.1 KA Statement: Krcwledge ofradiation exposure hmits and contammation control,m' cluding permissible levels in excess of those authonzed -
| |
| System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Gesup:
| |
| Justification: Low RO K/A importance vahae but required knowledge per 10CFR20 ,;
| |
| | |
| Questi:n: 009 ,
| |
| Which of the following is the REQUIRED inunediate action if a Locked High Radiation Area door is found open with no control of area access?
| |
| : a. Inform Security and establish Positive Access Control f
| |
| : b. Inform the on-shift Health Physics Technician and lock the area after checking for unauthorized personnel
| |
| . Inform the Security. %:h the area and have Health Physics check for exposures in excess of
| |
| ' those expected
| |
| : d. Inform the Health Physics Supervisor and establish Positive Access Control Answer: d Rer.rences: HP-C-202, " Locked High Radiation Area Controls", Rev. 6, Section 7.13.1, Page 9 GETCM-10308, " Radiation Protection For PAAT Workers", Rev. 0A, LO - 10 Exam level: Both History: New
| |
| ( ; K/A: 294001G310 2.9/3.3 KA Statement: Ability to perform procedures to reduce exussive levels of radiation and guard against personnel exposure System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| L
| |
| | |
| Questi:n: 010 Which of the following must be met prior to taking deliberate action to depan from Peach Bottom Technical Speci6 cations during an emergency?
| |
| This action is:
| |
| : a. immediately needed speci6cally to p'rotect the public health and safety.
| |
| : b. immediately needed specifically to protect valuable public property.
| |
| : c. immediately needed speci6cally to maintain capacity factors in excess of corprate l
| |
| ! standards.
| |
| i
| |
| : d. immediately needed to maintain grid stability during and Emergency Generation conditiort.
| |
| l l Answer: a
| |
| | |
| ==References:==
| |
| 10 CFR 50.54x & y l
| |
| l LOT-0007, "OM Chapters 6 - 9", Rev. 000, LO - 2 & 3 i
| |
| History: hew
| |
| () Exam Level: Both K/A: 294001G412 3.4/3.9 KA Statement: Knowledge of general operating crew responsibilities desing emergency operations i
| |
| System / Evolution:
| |
| Eum Section: Plant Wide Generics RO Group: SRO Group:
| |
| l l
| |
| i
| |
| | |
| Question: 011 .
| |
| Which of the following individuals is the preferred NRC communicator should the primary choice not
| |
| ,' ' ~
| |
| be available?
| |
| : a. The Shift Technical Advisor / Independent Assessor
| |
| : b. The unaffected Unit's Reactor Operator
| |
| : c. Shift Operations Assistant
| |
| : d. The founh Reactor Operator Answer: d
| |
| | |
| ==References:==
| |
| ERP-200, " Emergency Director", Rev.13, Section 3.3, Page 2 PEPP-0010, " Emergency 1 reparedness Trainmg", Rev. O, LO - 4 Exam Level: RO History: New K/A: 29400lG439 3.3/3.1 (f~ , KA Statement: Knowledge of the RO's responsibilities in emergency plan implementation System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| | |
| Questi:n: 012 Which of the following is an appropriate use of a Special Condition Tag (SCT)?
| |
| : a. An SCT is being applied to an active clearance to correct a tagging conflict covered by a Suspensionlabel.
| |
| : b. A second SCT, with the same position as the first, is being applied to the same component
| |
| : c. An SCT is being applied to a component requuing manipulation during maintenance.
| |
| I
| |
| : d. The SCT is being applied on equipment already tagged with a danger tag from the Master Clearance. I Answer: c l i
| |
| | |
| ==References:==
| |
| Clearance & Tagging Manual, Rev. 3, Section 4.3, Pages 21 & 22 l l
| |
| NCT-0200, " Clearance & Tagging", Rev.1, LO - 2 l Exam Level: Both History: New ;
| |
| I K/A: 29400lG213 3.6/3.8
| |
| (' 4 KA Statement: Knowledge of tagging and clearance procedures System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| | |
| Question: 013 Which of the following tasks may be y* srmed by a fully qualified Advanced Radiation Worker (ARW) without prior HP Supervisor app wl
| |
| : a. Free release ofmaterials from an RCA
| |
| : b. Performing surveys in areas posted as " Radiation Area" or less
| |
| : c. Performing surveys in a suspected hot particle area
| |
| : d. Performing contamination surveys in a high airborne radioactivity area Ar swer: b F.eferences: PECO Energy Rariantion Worker Handbook, Page 40 GETCM-10400, " Basic Radiation Worker Traming", Rev. 2, LO - 61 & 75 Exam level: Both History: New K/A: 29400lG301 2.6/3.0
| |
| (~j KA Statement: Knowledge of 10CFR20 and related facility radiation control requirements System / Evolution:
| |
| Exam Section: Plant Wide Generics RO Group: SRO Group:
| |
| | |
| Questirn: 014
| |
| , ' One of the concerns with maintaining proper reactor water level during plant operation is to mimmize "canyunder" Which of the following would result if excessive "carryunder" were occurring?
| |
| : a. Steam quality exiting the reactor vessel will decrease.
| |
| : b. Jet pump net positive suction head would increase. !
| |
| : c. Indicated reactor water level will fluctuate.
| |
| i
| |
| : d. Core thermal power would decrease Answer: d
| |
| | |
| ==References:==
| |
| LOT-0010," Reactor Vessel And Internals", Rev. 007, Section DI.C.4.c.2), Page 24, LO. - 2.b Exam level: Both History: New K/A: 290002K303 3.3/3.4 1( )
| |
| KA Statement: Reactor power System / Evolution: Reactor VesselIntemals l
| |
| Exam Section: Plant Systems RO Group: 3 SRO Group: 3 l Justification: Answer is the effect from increanng downcomer temperatures from steam coming from the separators into the downcomer e
| |
| | |
| - Questi:n: 015 Given the following conditions:
| |
| - Unit 2 is operating at 37% power
| |
| - Total core flowis 42%
| |
| Which of the following transients / evolutions will place the Unit CLOSER to thermal hydraulic instabilities?
| |
| : a. Recirculation flow is raised in both loops
| |
| : b. A single controlrod is withdrawn
| |
| : c. Reactor pressureislowering
| |
| : d. Feedwater heaters are bemg placed in senice Answer: b
| |
| | |
| ==References:==
| |
| LOT-0040, " Recirculation Flow Control", Rev. 008, Figure 5, LO - 5.a Exam Level: RO History: New K/A: 202002G447 3.4/3.7 KA Statement: Ability to diagnose and rengni= trends in an accurate and timely manner utilmng the appropriate control room i4-oe matenal System / Evolution: Recirculation Flow Control Exam Section: Plant Systems RO Group: 1 SRO Group: 1 e
| |
| | |
| )
| |
| {
| |
| Questi:n: 016 {
| |
| I Given the following conditions: j I
| |
| A reactor stanup is in progress on Unit 2
| |
| - The " continuous withdraw" mode for control rod withdrawals is being used
| |
| - While withdrawing control rod 26-31, the Unit Reactor Operator noted that control rod motion had stopped, the " settle" function had been completed and the rod had automatically de-selected Which ofthe following has occurred?
| |
| : a. The Emergency In/ Notch Override Switch was released
| |
| : b. A Rod Position Information System "Inop" has been received.
| |
| : c. A Rod Block Monitor rod block has occurred
| |
| : d. A Rod Wonh Minunizer rod block has occurred Answer: b
| |
| | |
| ==References:==
| |
| LOT-0060, " Control Rod Drive Mechanism Sydem/ Rod Position Indication System", .
| |
| Rev. 009, Section II.B.2.d, Page 18 i
| |
| ( .
| |
| ARC 211 D-5 LOT-0080, " Reactor Manual Control System", Rev. 007, LO - 2.a & b I
| |
| Exam Level: Both History: New X/A: 214000K303 3.1/3.2 KA Statement: RMCS System / Evolution: Rod PositionInformation System Exam Section: Plant Systems RO Group: 2 SRO Group: 2 6
| |
| | |
| Question: 017 .
| |
| Given the following conditions:
| |
| -- Control rod withdrawal is in progress during a reactor startup on Unit 2
| |
| -- Reactor power: 1%
| |
| Core flow: 30 %
| |
| Reactor pressure: 920 psig Which of the following will directly result in a reactor scram?
| |
| : a. Main twt'ine trip
| |
| : b. FullGroup 1 isolation
| |
| : c. Main condenser vacuum at 20 inches Hg
| |
| : d. Reactor power of17%
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0300, " Reactor Protection System", Rev.13, Figure 4A, LO - 5.q Exam Ievel: RO History: New
| |
| (,
| |
| K/A: 212000A206 4.1/4.2 KA Statement: High reactor power System / Evolution: Reactor Protection System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 l
| |
| 1
| |
| | |
| l Question: 018 Given the following conditions:
| |
| - Unit 2 drywell pressure has reached: 2.2 psig Reactor waterlevelis: -172 inches
| |
| - One Core Spray Pump has started with no other ECCS pumps in service No operator actions have been taken q The Automatic Depressunzation System will actuate:
| |
| : a. in 115 seconds.
| |
| : b. in 9.5 minutes.
| |
| : c. 9.5 minutes aRer any additional Core Spray Pump is started
| |
| : d. I15 seconds aRer any RHR Pump is started.
| |
| i Answer: d
| |
| | |
| ==References:==
| |
| LOT-0330, " Automatic Depressunzation System", Rev. 008, Section V.A.1 - 8, Pages ,
| |
| 15 & 16, LO - 2.a & 6.a i
| |
| , I. .)
| |
| ' Exam level: Both History: New K/A: 218000K501 3.8/3.8 KA Statement: ADS logic operation System /Evolutior.: Automatic Depressunzation System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 l
| |
| l
| |
| [
| |
| | |
| Question: 019 .
| |
| Unit 2 is operating at 95% power when a scram signal is generated from a main turbine trip.
| |
| Which of the following conditions will PREVENT the backup scram valves from venting the scram air header?
| |
| : a. The check valve bypassmg the downstream backup scram valve fails closed.
| |
| : b. The upstream backup scram valve in the air flow path fails to reposition.
| |
| : c. The "A" Reactor Protection System trip system does NOT deenergize.
| |
| : d. 125 VDC to ONE of the backup scram valve is deenergized.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0070, " Control Rod Drive Hydraulic System", Rev. 008, Section III.I.4.d &
| |
| Figure 6, Page 16, LO - 4.d & 6.c Exam Level: Both History: . New K/A: 20100lK203 3.5/3.6
| |
| (' ) KA Statement: Backup SCRAM valve solenoids System / Evolution: ControlRod Dnve Hydrauhc System Exam Section: Plant Systems RO Group: 1 SRO Group: 2 l
| |
| 1
| |
| | |
| . Questisn: 020 l Given the following conditions:
| |
| - Reactor power is being raised from 25% to 35% by control rod withdrawals The current group ofrods being withdrawn has a withdraw limit of Notch "24"
| |
| -- The last rod in the group is withdrawn to Notch "28"
| |
| - The CRS has been noti 6ed
| |
| - One minute has elapsed since the rod withdrawal was completed What are the required actions for this control rod?
| |
| : a. hnmediately insert the rod to Notch "24" and notify the Reactor Engineer
| |
| : b. Leave the rod at Notch "28" and notify the Reactor Fagia .
| |
| : c. Insert the rod to Notch "24" and continue the power increase with CRS approval.
| |
| : d. Immediately insert the rod to Notch "00" l Answer: a
| |
| | |
| ==References:==
| |
| ON-122, "Mispositioned Control Rod", Rev. 4, Section 2.5, Pa8e I l ') LOT-1550, "Off-Nonnal Procedures", Rev. 006, LO - 2 Exam Level: Both History: New K/A: 201002G449 4.0/4.0 KA Statement: Ability to perform without mbe to procedures those actions that require '
| |
| . iWiee operation of system w,ir.pc,r--. and controls System / Evolution: Reactor ManualControl System ;
| |
| i Exam Section: Plant Systems RO Group: 1 SRO Group: 2 6
| |
| | |
| . Question: 021 Given the following conditions:
| |
| A cooldown is in progress using bypass valves Level control is in automatic on the startup level controller
| |
| - Actual water level is betng controlled at a constant value Which of the following describes what will happen to INDICATED Wide Range level dunng the cooldown?
| |
| INDICATEC Wide Rangelevel will: ,
| |
| : a. remam constant.
| |
| : b. decrease towards actuallevel.
| |
| : c. increase from actuallevel.
| |
| : d. decrease from actuallevel.
| |
| Answer: a
| |
| -A
| |
| | |
| ==References:==
| |
| LOT-0050, " Reactor Vessel Instrumentation", Rev. 013, Section III.F & Handout #7, Pages 20 & 21, LO - 6.d Exam Lael: Both History: New K/A: 216000KS10 3.1/3.3 KA Statement: Indi:atM level venus actual vessel level dunng vessel hestups or cooldowns System / Evolution: Nuclear BoilerInstrumentation Exam Section: Plant Systems RO Group: 1 SRO Group: 1 t
| |
| I
| |
| | |
| , l Questi:n: 022 Given the following conditions:
| |
| -- Drywell' ventilation was operating in its normallineup when a Loss Of Coolant Accident occurred on Unit 2
| |
| - Drywell pressureis 4 psig
| |
| - The Reactor and Radwaste Buildings are not accessible Which of the following describes the cooling capabihties ofthe drywell coolers for these conditions?
| |
| The drywell coolers:
| |
| : a. will automatically be mnning in " Fast" speed
| |
| : b. will automatically be running in " Slow" speed
| |
| : c. cannot be restarted until the Reactor Building access is restored.
| |
| : d. may only be restarted in " Fast" speed Answer: d
| |
| | |
| ==References:==
| |
| ' LOT-0140, "Drywell Ventdation", Rev. 009, Sections IV.C.2 & D.4, Pages 6 - 8, LO -
| |
| f r-) -
| |
| 5&6 Exam Isvel: Both History: New K/A: 22300lK103 3.2/3.3 KA Statement: Containment /drywell at==hne control System / Evolution: Primary Containment System And Aux haries Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| Question: 023 .
| |
| The Unit 3 High Pressure Coolant Injection (HPCI) system is running in " Automatic" maintaining reactor water level following a leak. An oil leak occurs on the line supplying the Turbine Governor Valve (HO-4512).
| |
| Which of the following is the expected response of HPCI?
| |
| : a. Turbine speed will decrease as governor valve oil pressure is lost.
| |
| : b. HPCI will receive a turbine trip signal if the Auxiliary Oil Pump cannot maintain oil pressure.
| |
| : c. Turbine speed will increase as governor valve oil pressure is lost.
| |
| : d. HPCI will rapidly accelerate and trip on overspeed.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0340, "High Pressure Coolant Injection", Rev. 009, Sections III.N & V.A.3.d, Pages 17-18 & 26, LO - 5.i & k Exam 12 vel: RO History: New
| |
| ( ,
| |
| K/A: 206000A301 3.6/3.5 KA Statement: Turbine speed System / Evolution: High Pressure Coolant Injection System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 L
| |
| )
| |
| | |
| 024 ;
| |
| - Quesaca:
| |
| ' Given the following conditions:
| |
| A Loss Of Coolant Accident has occurred on Unit 2
| |
| - All plant systems automatically operated as designed Reactor waterlevelis: -25 inches
| |
| - Drywell pressureis: 1.7psig
| |
| - . The Plant Reactor Operator is attempting to transfer the Reactor Core Isolation Cooling ,
| |
| (RCIC) system from the injection Mode to the CST to CST Mode i
| |
| - Each time the Full Flow Test Valve (MO-30)is opened, it mmMi=*ely closes i
| |
| Which of the following is the reason why the PRO cannot complete the transfer.
| |
| : a. The PRO has failed to reset the low reactor water level RCIC initiation signal.
| |
| : b. The RCIC Minimum Flow Valve (MO-27)' m terlock with MO-30 is in effect.
| |
| : c. The RCIC system is operatmg with suction from the torus.
| |
| . d. The drywell pressure is still too high to c@e the transfer.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| SO 13.1.B-2, "RCIC System Manual Operation", Rev. 6, Section 4.3.1 Note, Page 7 LOT-0380, " Reactor Core Isolation Cooling", Rev. 010, LO - 1.c & 2.a Exam Level: Both History: New j
| |
| K/A: 217000A404 3.6/3.6 KA Statement: Manuallyi+=*M controls l 1
| |
| System / Evolution: Reactor Core Isola: ion Coohng System l Exam Section: Plant Systems RO Group: 1 SRO Group: 1 lI j
| |
| b l
| |
| j i
| |
| | |
| Question: 025 - ;
| |
| , With Unit 2 at 100% the "B" Core Spray Line Break differential pressure is reading approximately -3.0 psid. l This reading is:
| |
| : a. indicative of a "B" Core Spray line break outside the shroud.
| |
| : b. considered normal due to the actual differential pressure across the steam dryers and separators at 100% power.
| |
| : c. indicative of a "B" Core Spray line break inside the shroud.
| |
| : d. considered nopnal due to the differences between instrumerit calibration conditions and conditions at 100% power.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0350, " Core Spray", Rev. 010, Section IV.A.4.c, Pages 10 & 11, LO - 5.d Exam Level: Both History: New K/A: 209001A205 3.3/3.6
| |
| (.-)
| |
| KA Statement: Core sprayline break System / Evolution: Low Pressure Core Spray System Exam Section: Plant Systems RO Group: 1 SRO Group: -1 L
| |
| l
| |
| _______ ---------------_-------_--_--_---__J
| |
| | |
| l
| |
| . Questisn: 026 Given the following conditions on Unit 2:
| |
| A half scram exists on Reactor Protection System (RPS)"A" due to APRM testing A fire has just caused a loss of RPS Bus "B" resulting in a full scram
| |
| - Allcontrolrodsfullyinserted
| |
| - The half scram testing was stopped and the APRMs returned to normal
| |
| - The SDV High Level Scram Bypass switch is then taken to " Bypass" When can the half scram on RPS "A" be reset for these conditions?
| |
| The RPS "A" halfscram may be reset:
| |
| : a. unmediately.
| |
| : b. after the Scram Discharge Volume Vent and Dram valves are fully open.
| |
| : c. after RPS"B"is reenergized. -
| |
| )
| |
| : d. only if the Reactor Mode Switch is in " Shutdown", " Refuel" or "Startup/ Hot Standby".
| |
| l Answer: c i
| |
| (}
| |
| | |
| ==References:==
| |
| LOT-0300, " Reactor Protection System", Rev. 013, Section IV.B.4.b, Page 30, LO -
| |
| 5.i & 7.a 1
| |
| Exam Ievel: Both History: New K/A: 212000K601 3.6/3.8 KA Statement: A.C. electrical distribution System / Evolution: ReactorProtection System i
| |
| Exam Section: Plant Systems RO Gesup: 1 SRO Group: 1 i
| |
| l' O N i
| |
| l
| |
| | |
| i
| |
| ?
| |
| Questi:n: 027 Given the following conditions:
| |
| - Unit 3 is performing a scheduled shutdown in accordance with GP-3, "Nomial Plant Shutdown" Control rods are being inserted per the required insertion sequence
| |
| - The Rod Worth Minimi- was NOT irutishzed when required by GP-3 Which of the following describes how this will affect plant operation?
| |
| : a. ' Control rod insertions may continue provided the Unit Reactor Operator does not deviate from the required sequence
| |
| : b. Control rod insertions may continue until a total of two " insert" errors are received.
| |
| c Control rod insertions will be stopped when steam flow decreases to less than 17% or feed flow decreases to less than 16%.
| |
| : d. Control rod insertions will be stopped when steam flow decreases to less than 27%
| |
| Answer: d 7 ,
| |
| | |
| ==References:==
| |
| LOT-0090," Rod Worth Minimi=", Rev 010, Section V.B.2, Page 24, LO 5.j &
| |
| -' 6.c Exam level: Both History: New K/A: 201006A304 3.5/3.4
| |
| : KA Statement: Controlrod movement blocks System / Evolution: Rod Worth WJ. 2- System Exam Section: Plant Systems RO Group: 2 SRO Group: 2
| |
| (
| |
| L
| |
| | |
| Question: 028 Given the following conditions:
| |
| - The Unit 2 main generator hasjust been synchronized to the grid and is carrymg 75 MWe load
| |
| - The Turbine Building operator reports a loud noise coming from the #1 Low Pressure Turbine
| |
| - The Unit Reactor Operator depresses the "All Valves Closed" pushbutton on the EHC Control Panel Select the expected main turbine response.
| |
| : a. No Turbine Control Valve or Turbine Stop Valve motion will occur.
| |
| : b. The Turbine Control Valves and Turbine Stop Valves will close (turbine trip) and the generator output breaker will open from the turbine trip
| |
| : c. The Load Set Potentiometer will pulse back to 0% load but a turbine will NOT trip.
| |
| : d. The Turbine Control Valves and Turbine Stop Valves will close (turbine trip) and the generator output breaker will trip on reverse power. -
| |
| ~
| |
| g- Answer: a
| |
| | |
| ==References:==
| |
| LOT-0590, " Electro-Hydraulic Control Logic", Rev. 007, Section 1.2.B.a, Page 7, LO
| |
| - 1.b & 4.c Exam Level: Both History: New ,
| |
| K/A: 241000A419 3.5/3.4 KA Statement: Turbine panel controls System / Evolution: Reactor / Turbine Pressure Regnicing System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 Justification: With 1800 rpm selected and the output breaker closed this pushbutton is disabled.
| |
| .L
| |
| )
| |
| l l
| |
| | |
| 1 Question: 029 .
| |
| Given the following conditions:
| |
| - Unit 2 is operating at 100% power
| |
| - The "A" Recirculation Pump has just tripped 1 After power and flow have #abilimi, the Unit Reactor Operator notes the "B" Recirculation Loop Jet Pump flow is greater than it was prior to the pump trip No operator actions are taken Which of the following describes the reason for this change?
| |
| "B" Loop Jet Pump flow increased due to:
| |
| : a. the reduced subcooling at the recire pump suction.
| |
| : b. the increased core voiding.
| |
| : c. thelower core pressure drop.
| |
| : d. the increased reactor water level.
| |
| Answer:
| |
| (-) c
| |
| | |
| ==References:==
| |
| LOT-0030, " Reactor Recirculation System", Rev. 009, Section VHF, Pages 48-50, LO - 6.h & 9.d CentrifugalPump Head Loss Curves Exam Level: Both History: New K/A: 202001A102 3.4/3.4 KA Statement: Jet pump flow System / Evolution: Recirculation system Exam Section: Phnt Systems RO Group: 2 SRO Group: 2
| |
| | |
| Question: 030 Given the following conditions: j I
| |
| -- Unit 3 is perfonning a reactor startup and heatup l
| |
| - Reactor water level control is via Reactor Water Cleanup (RWCU) rejecting to the main condenser
| |
| - Main condenser vacuum has been established with the vacuum pump
| |
| - The operator is cautioned to carefully monitor system parameters while rejecting Which of the following RWCU system trips /isolations will provide protection while in this lineup?
| |
| : a. Cleanup Drain Header Control Valve (CV-55) closure on low upstream pressure j i
| |
| : b. RWCU system isolation on non regenerative heat exchanger high outlet tengrature.
| |
| : c. Cleanup Dram Header Control Valve (CV-55) closure on non-regenerative heat exchanger high outlet tr.uw.ture.
| |
| i
| |
| : d. RWCU system isolation on low upstream pressure Answer: b l
| |
| .x
| |
| * Refaences: LOT-0110," Reactor Water Cleanup System", Rev. 011,"Section IV.A.I.b, Page 11, !
| |
| LO - 4.c & d Eram level: RO History: New K/A: 204000K403 2.9/2.9 KA Statement: Over temperature protection for system caponents System / Evolution: Reactor Water Cleanup System Exam Section: Plant Systems RO Group: 2 SRO Cup: 2 l
| |
| I l
| |
| L
| |
| | |
| Questi:n: 031 .
| |
| Unit 2 is operating at 85% power with the "A" Recirculation Pump scoop tube locked.
| |
| Select the action REQUIRED if a reactor scram occurs.
| |
| : a. D5 w a i : msed Operator to manually position the "A" Recirculation Pump scoop tube to "nur *; ' speed.
| |
| : b. Reduce the "B" Recirculation Pump speed to minunum. then trip the "A" pump.
| |
| : c. Trip the "A" Recirculation Pump imMetely.
| |
| : d. Direct any operator to position the "A" Recirculation Pump scoop tube to "muumum" speed then trip the pump from t!2 control room.
| |
| Answer: c Referencs: SO 2D.7.B-2, " Recirculation MG Set Scoop Tube Lockup And Reset", Rev. 8, Section 4.2 Caution, Page 4 LOT-0040, " Recirculation Flow Control", Rev. 008, LO - 1.c & 5.c 3
| |
| Exam Level: Both History: New K/A: 202002A205 3.1/3.1 KA Statement: Scoop tubelockup System / Evolution: Recirculation Flow Control System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| 1 Question: 032
| |
| -g Given the following conditions:
| |
| - _ Unit 2 had been operating at 55% power, increasing power to 100% following an outage
| |
| . Main turbine vibrations required placing the Reactor Mode Switch to " Shutdown" All plant response and scram actions were normal 1
| |
| Which of the following PREVENTS control rod withdrawals for these conditions?
| |
| : a. The Reactor Mode Switch in " Shutdown" maintams a scram signal on RPS until reset by the operator.
| |
| : b. The Reactor Mode Switch in " Shutdown" inserts a contmuous control rod withdrawal block signal.
| |
| : c. De Rod Block Monitor "Downscale" inserts a control rod withdrawal block signal until bypassed
| |
| : d. Post scram Rod Worth Minimh'er f.nsert and withdraw errors will result in a control rod withdrawalblock signal.
| |
| Answer: b D
| |
| f)
| |
| | |
| ==References:==
| |
| SO 62.7.A-2,'" Receipt OfRod Blocks", Rev.12, A*-hmant 1, Page 5 l l
| |
| LOT-0080, " Reactor Manual Control System", Rev. 007, LO - 2.b & 4.a l 1
| |
| Exam Imel: Both History New K/A: 201002K103 3.4/3.6 ;
| |
| i KA Statement: Control rod block intedocks/ power operation /mfuehng Systent/ Evolution: Reactor Manual Control System l
| |
| i Fr== Section: Plant Systems RO Group: 1 SROGroup: 2 1
| |
| ' eso em l
| |
| t I
| |
| | |
| 9 Question: 033 .
| |
| Givea the following conditions:
| |
| - The E-2 Diesel Generator (DG) received a valid start signal on loss of voltage on the E-22 bus
| |
| - The DG started and reenergized the bus as designed
| |
| -- The cause of the bus loss ofvohage has been found and corrected and preparations are being made to remove the DG from the bus and shut it down The operator hasjust depressed the DG Auto Start Bypass pushbutton Which of the following is the function of this ge* mon?
| |
| : a. This bypasses the DG auto start signal to pernut the diesel to be shutdown aRer E-22 has been transferred to the grid.
| |
| : b. This places the DG in the isc4. visas (unit) mode to allow the E-22 bus to be synched to the grid and the DG unloaded prior to shutdown.
| |
| : c. This bypasses the DG auto start signal for 3 minutes to allow the E-22 bus transfer back to the grid before DG shutdown.
| |
| : d. This places the DG in the syid. visas (parallel) mode to allow the E-22 bus to be synched
| |
| . to the grid and the DG unloaded prior to shutdown.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0670, "Ihesel Generators And Auxiharies", Rev. 006, Section IV.D.2.h.4), Page 38, LO - 1.b & 3.a Eaam Level: Both History: New K/A: 264000K407 3.3/3.4 KA Statement: Localoparation and control System / Evolution: Emergency DieselGenerators Enam Section.: Plant Systems RO Group: 1 SRO Group: 1 L r f
| |
| | |
| Questirn: 034 Given the following conditions
| |
| - Unit 3 is at 100% power
| |
| - The signal from APRM "B" to Rod Block Monitor (RBM) Channel"B" hasjust failed
| |
| " upscale"
| |
| - Prior to the failure no rod block or half scram signals were present Which of the following is the expected response of the RBM for this failure?
| |
| : a. RBM Channel"B" applies the high trip =*nalat.
| |
| - b. RBM Channel"B" generates an RBM inoperative trip.
| |
| : c. RBM Channel"B" generates a control rod withdraw block.
| |
| : d. RBM Channel"B" switches to the reference APRM backup signal (APRM "D") and goes through a null sequence.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0280, " Rod Block Monitor", Rev. 010, Section V.E.2, Page 14, LO - 7.b I ' Exam level: RO History: FEBQ #3856, Examiner Moddied (Changed fonnat, added speedic channel designations, cleaned up question)
| |
| K/A: 215002A203 3.1/3.3 KA Statement: Loss ofeWM reference APRM channel System / Evolution: Rod Block Monitor System Exam Section: ' Plant Systems RO Group: 2 . SRO Group: 2 L
| |
| I
| |
| \
| |
| 1 l
| |
| | |
| Questi::n: 035 .
| |
| Assuming the Local Power Rane', Monitor (LPRM) detector outputs were NEVER calibrated over their expected lifetimes, whic'. of the following would be the result and why?
| |
| : a. ' Acoubctor power would be GREATER than indicated APRM power due to the aepletion of the LPRM detector U-235 coating over time.
| |
| : b. Actual reactor power would be LESS than indicated APRM power due to the depletion of the LPRM detector U-235 coating over time.
| |
| : c. Actual reactor power would be GREATER than indicated APRM power due to the build up of the LPRM detector U-235 coatmg over time.
| |
| : d. Actual reactor power would be LESS than i*=W APRM power due to the build up of the LPRM detector U-235 coatmg over time.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0260, " Local Power Range Monitor System", Rev. 007, Section IILC, Page 8, LO - 5 & 6.a Exam Level: Both History: New s.
| |
| K/A: 215005A101 4.0/4.0 KA Statement: Reactor powerindication System / Evolution: Average Power Range Monitor / Local Power Range Monitor System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 J
| |
| L
| |
| -___-_-_--___-_--______-___.__a
| |
| | |
| Questi:n: 036 Plant operating experience has shown that indicated reactor water level perturbations may occur during depressurization.
| |
| Which of the following is the cause of these level changes?
| |
| : a. Gases coming out of solution in the vanable leg
| |
| : b. Lowering differential temperature between the vanable and reference legs i
| |
| : c. Gases coming out of solution in the reference leg l
| |
| : d. Lowering differential temperature between the drywell and reactor buildings Answer: c
| |
| | |
| ==References:==
| |
| LOT-0050, " Reactor Vessel Instrumentation", Rev. 013, Section V.E.7, Pages 50 &
| |
| 51, LO - 1.c & 6.o Exam Level: RO History: FEBQ #34%, Exammer Modified (Minor stem l rewording) i l
| |
| f .,
| |
| K/A: 216000A207 3.4/3.5 KA Statement: Reference leg flashing System / Evolution: Nuclear BoilerInstwation Exan Section: Plant Systems RO Group: 1 SRO Group: I l L
| |
| | |
| 1 Question: 037 .
| |
| During a Unit 2 startup and heatup in accordance with GP-2, " Normal Plant Startup", the operator is cautioned to maintain turbine first stage pressure less than 100 psig during sheJ1 warming.
| |
| Which of the following is the consequence of exceeding this value?
| |
| : a. The main turbine will ro.ll off the tuming gear.
| |
| ~
| |
| : b. The main turbine trip reactor scram may occur.
| |
| : c. The main turbine will trip.
| |
| : d. The high pressure turbine shell heatup and expansion limits will be exceeded.
| |
| ~
| |
| Answer: b
| |
| | |
| ==References:==
| |
| GP-2, " Normal Plant Startup", Rev. 83, Step 6.2.6 Caution, Page 60 LOT-1530, " General Plant Procedures", Rev. 010, LO - 3 & 4 Exam Level: Both History: FEBQ #2959, Exanuner Modified (Reword stem, ciged each distracter to more operationally oriented format)
| |
| K/A: 241000A313 3.0/3.0 KA Statement: Turbine startup System / Evolution: Reactor /furbine Pressure Rpa%g System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| Question: 038 Given the following conditions-The Unit 2 "A" Residual Heat Removal (RHR) Pump staned from a valid initiation signal and is still running ,
| |
| The operator has been directed to stop the pump The initiation signal is still present Once the pump is stopped, which of the following will result in the pump rualting?
| |
| The"A" RHR Pump will restan:
| |
| : a. as soon the operator releases the switch when stopping the pump.
| |
| : b. when the Pump Stan Reset pushbutton (S31 A)is depressed with the initiation signal present.
| |
| : c. if a second initiation signal comes in before the first signal clears.
| |
| I
| |
| : d. when the LPCI "A" Lockout Reset pushbutton (SI A) is depressed with the initiation signal present.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal System", Rev. 010, Handout 8, Page 1 of 16, LO
| |
| - 1.b i
| |
| Exam level: RO History: New K/A: 203000K401 4.2/4.2 i
| |
| KA Statement: Automatic system initiation / injection System / Evolution: RHR/LPCI: Injection Mode !
| |
| Exam Section: Plant Systems RO Group: 1 SRO Group: 1 1
| |
| | |
| Question: 039 Given the following conditions:
| |
| The "A" loop of Residual Heat Removal (RHR) is being placed into Shutdown Cooling (SDC)using the"A" RHR Pump The RHR Toms Suction Valve (MO-13 A) has been closed The Shutdown Cooling Suction Valves (MO-17 and 18) have been opened The "A" RHR Pump SDC Suction Valve (MO-15A) has been opened No other actions have been taken An inadvertent high drywell pressure signal (2.0 psig) occurs With NO operator actions taken, what would be the expected RHR system (LPCI Mode) status?
| |
| : a. The "A" RHR Pump will start and will pump water from the Shutdown Cooling Suction to the reactor vessel via the LPCI injection flowpath
| |
| : b. The Toms Section Valve (MO-13 A) will open and all RHR pumps will inject using both the toms and the reactor vessel as water sources
| |
| : c. The RHR Toms Suction Valve (MO-13 A) and the pump SDC Suction Valve (MO-15A) will not reposition and the "A" RHR Pump will not start.
| |
| : d. The pump SDC Suction Valve (MO-15A) will close, the Torus Suction Valve (MO-13A) will open and the pump will start and inject.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal", Rev. 010, Handout H-LOT-0370-2, Section 2, Page 1, LO - 5.a & b.
| |
| Exam level: Both History: New K/A: 203000A308 4.1/4.1 KA Statement: Systemirutiation sequence System / Evolution: RHR/LPCI: Injection Mode ;
| |
| Exam Section: Plant Systems RO Group: 1 SRO Group: 1 I
| |
| | |
| 1
| |
| )
| |
| Question: 040 With Unit 2 at 100% power, Panel E-124-R-C is lost, i How does this power loss affect operation of the Standby Liquid Control (SLC) System?
| |
| : a. No system indicating lights will be illuminated and both pumps and both squib valves are not available.
| |
| : b. The "A" SLC Pump and "A" Squib Valve will not be available.
| |
| : c. Following a SLC system initiation, the operator will not be able to verify pump discharge pressure or SLC storage tank level in the Control Room or locally.
| |
| : d. The "B" SLC Pump and "B" Squib Valve will not be available.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0310, " Standby Liquid Control System", Rev. 011, Section V.3.c, Page 17a, LO
| |
| - 3.a & b Exam level: Both History: New K/A: 211000K202 3.1/3.2 KA Statement: Explosive valves System / Evolution: Standby Liquid Control System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 ,
| |
| l J
| |
| | |
| Question: 041 -
| |
| Given the following conditions:
| |
| - Unit 2 is in Mode 4 The AC input power to the 24 VDC batter / charger supplying Panel 2AD45 has been lost
| |
| - This panel supplies power to Source Range Monitoring (SRM) channels "A" and "C" Which of the following describes the long-term effect on these SRM channels?
| |
| : a. The SRMs will slowly deplete the battery and, over time, channels "A" and "C" will drift "downscale",
| |
| : b. The SRMs will continue to operate due to automatically swapping to the second battery charger.
| |
| : c. The SRMs will slowly deplete the battery and, over time, chanrels "A" and "C" will drift
| |
| " upscale",
| |
| : d. The SRMs will NOT operate until an equipment operator transfers power via the Manual Bus Transfer switch.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0240, " Source Range Monitoring System", Rev. 008, Section III.G, Pages 10 &
| |
| 11, LO - 7.b Exam Imel: Both History: New K/A: 215004K602 3.1/3.3 KA Statement: 24/48 Volt D.C. power System /E;olution: Source Range Monitor System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| Question: 042 Given the following conditions:
| |
| A Unit 2 reactor stanup is in progress with control rod withdrawals occurring 1 Rod Worth Minmuzer (RWM) Group I contains 12 control rods that are to be withdrawn from Notch "00" to Notch "48" 11 rods from this group are withdrawn to Notch "48" and the remainmg rod to Notch "42" A control rod in Group 2 is then selected but not withdrawn )
| |
| J Which of the following is the expected response of the RWM?
| |
| The RWM will display:
| |
| : a. one withdraw error and funher rod withdrawals will be blocked except for the rod with the withdraw error.
| |
| : b. one withdraw error and if a second withdraw error is made funher rod withdrawals will be blocked except for the two rods wit'2 the withdraw errors.
| |
| : c. one insen error and funher rod withdrawals will be blocked except for the rod with the msen error,
| |
| : d. one insert error and if a second insert error is made, funher rod withdrawals will be blocked except for the two rods with the insert errors.
| |
| Answer: d l 1
| |
| | |
| ==References:==
| |
| LOT-0090, " Rod Wonh Mmmuzer", Rev. 010, Sections III.A 7-9 & IV.B, Pages 11
| |
| & 22, LO - 5.a & 6.j & 1 Exam Imel: Both History: New K/A: 201006K513 3.5/3.5 KA Statement: Insert block System / Evolution: Rod Wonh Minimi- System Exam Section: Plant Systems RO Gmup: 2 SRO Gmup: 2
| |
| | |
| Question: 043 During an overpressure transient on Unit 2, the Plant Reactor Operator notes the following Safety Relief Valve (SRV) indications:
| |
| 11 SRV whitelights areilluminated
| |
| -- The "C" and "D" SRV red lights are illuminated
| |
| -- All other SRV green lights are illuminated What was the muumum peak reactor pressure during this transient and what is the approximate current reactor pressure?
| |
| During the transient reactor pressure reached at least:
| |
| : a. 1135 psig and pressure is now approximately 1100 psig.
| |
| : b. I 155 psig and pressure is now approxunately 1135 psig.
| |
| : c. 1260 psig and pressure is now approximately 1135 psig.
| |
| : d. 1325 psig and pressure is now approximately 1155 psig.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0120," Main Steam And Pressure Relief System", Rev. 014, Sections III.B.2 &
| |
| IV.5, Pages 10 & 21-22 Exam level: RO History: New K/A: 239002A302 4.3/4.3 KA Statement: SRV operation on high reactor pressure System / Evolution: Reimf7 Safety Valves Exam Section: Plant Systems RO Group: 1 SRO Gmup: 1
| |
| | |
| Question: 044 l
| |
| 1 If the Main Steam Isolation Valves (MSIV) have one DC and one AC solenoid on each valve, how does a loss of both Reactor Protection System (RPS) buses result in a closure of all eight MSIVs?
| |
| : a. A loss of both RPS buses results in a trip of both primary containment isolation system .
| |
| logic channels and subsequent MSIV closure.
| |
| : b. The RPS buses supply 20Y33 and 20Y34 which supply the AC and DC (via an inverter)
| |
| MSIV solenoids respectively such that a loss of both buses closes the MSIVs.
| |
| : c. A loss of both RPS buses results in a loss of the MSIV position indications giving a reactor scram and MSIV closure.
| |
| : d. The RPS buses supply 20Y33 and 20Y34 which supply the power to the MSIV test solenoids such that a loss of both buses deenergizes them and closes the MSIVs.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0180, " Primary Contamment Isolation System", Rev. 009, Section III.G.1, Pages 26 & 27, LO - 3 & 6.g Exam Level: Both History: New K/A: 223002K309 3.4/3.6 KA Statement: Main steam system System / Evolution: Primary Contamment Isolation System / Nuclear Steam Supply Shut-Off Exam Section: Plant System RO Group: 1 SRO Group: 1
| |
| | |
| Question: 045 .
| |
| Following a valid initiation signal on Unit 2, the Standby Gas Treatment (SBGT) train Bypass Damper (PO-00522) does NOT reposition as designed.
| |
| Which of the following describes how this failure will affect SBGT operation?
| |
| : a. The " Emergency Gas Filter Valve Failure" alarm will be received and the operator will be required to isolate one Ster train in order to maintain the required negative pressure in secondary containment.
| |
| : b. Once the second Ster train is shutdown and isolated, the remairung train may not be sufficient to maintain the required negative secondary contamment pressure.
| |
| : c. The " Emergency Gas Filter Valve Failure" alarm will be received and the operator will be required to start an additional SBGT fan in order to maintain the required negative pressure in secondary contamment.
| |
| : d. Once the second filter train is shutdown and isolated, process flow through the remauung train and fan may not be sufficient for charcoal adsorber bed heat removal.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0210, " Standby Gas Treatment System", Rev. 009, Section IIJ, Pages 11 & 12, LO - 1.b & 6.b Exam Level: Both History: New K/A: 261000K301 3.3/3.6 KA Statement: Secondary containment and environment differential pressure System / Evolution: Standby Gas Treatment System Exam Section: Plant Systems RO Group: 1 SRO Group: 1
| |
| | |
| Question: 046 With Unit 2 at 75% power, the Main Conuol Room has received the " Blowdown Relief Valves Bellows Leaking" alarm The Unit Reactor Operator reports that the "G" SRV amber light is lit.
| |
| Which of the following describes how this impacts the operation of the affected Safety Relief Valve (SRV)?
| |
| The SRV will:
| |
| : a. operate in the Automatic Depressurization mode but NOT in the Safety mode.
| |
| : b. operate in the Safety mode but NOT in the Automatic Depressurization mode. l
| |
| : c. operate only in the manual mode (i.e., switches on the C03 Panel). ,
| |
| : d. not operate in any mode.
| |
| Answer: a I
| |
| | |
| ==References:==
| |
| LOT-0330, " Automatic Depressunzation System", Rev. 008, Section IV.A.2.a.7, Page l 14, LO - 2.f & 7.a Exam Level: Both History: New K/A: 218000K106 3.9/3.9 l l
| |
| KA Statement: Safety /ReliefValve l System / Evolution: Automatic Depressunzation System Exam Section: Plant System RO Group: 1 SRO Group: 1
| |
| | |
| 1 43 Question: 047 - i Given the following conditions:
| |
| Unit 2 experienced a loss of coolant accident (LOCA) 10 minutes ago All 4 Residual Heat Removal (RHR) pumps are running and injecting into the vessel
| |
| --. Reactor water level: -100 inches and rising
| |
| -- Reactor pressure: 100 psig and steady
| |
| - Drywellpressure: 9.5 psig and rising l l
| |
| Which of the following logic interlocks must be met to initiate drywell spray using the "A" RHR Loop (i.e , open MO-31 A and 26A)?
| |
| : a. A least one HPSW Pump must be runrung
| |
| : b. The Lack ofLOCA signal must be bypassed
| |
| : c. The Contamment Spray Valve Control Switch (S17) must be momentarily placed in
| |
| " Manual"
| |
| : d. The Contamment Spray Override 2/3 Core Coverage Switch (SIS) must be placed in
| |
| " Manual Override" Answer: c
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal System", Rev. 010, Section IV.C.3.b, Page 24, .
| |
| LO - 1.c & 5.q Exam level: Both History: _New K/A: 226001A105 3.1/3.4 KA Statement: System lineup Syst?m/ Evolution: RHR/LPCI: Contamment Spray System Mode Exam Section: Plant Systems RO Group: 2 SRO Group: 1 i
| |
| l
| |
| | |
| Question: 048 Given the following conditions:
| |
| Both Units are operating at 100% power All startup power sources are energized and are available No bus breakers are blocked All four diesel generators (DG) are in their normal standby lineups An operator in the Control Room opens the E-212 breaker with its handswitch No other operator actions are taken.
| |
| Select the expected 4 KV system response.
| |
| The E-12 bus:
| |
| : a. will not be energized from 3 SUE but the E-1 Diesel starts and energizes the bus.
| |
| : b. will not be energized from 3 SUE and the E-1 Diesel does NOT start.
| |
| : c. will be energized from 3 SUE and the E-1 Diesel starts but the output breaker does NOT close
| |
| : d. will be energized from 3 SUE and the E-1 Diesel does NOT start.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0660, "4 KV Distribution", Rev. 010, Section V.C.1, Pages 23 & 24, LO - 4.e &
| |
| 7 Exam Level: Both History: New K/A: 262001A401 3.4/3.7 KA Statement: Allbreakers and disconnects System / Evolution: A C. ElectricalDistribution Exam Section: Plant Systems RO Group: 2 SRO Group: 1 i
| |
| | |
| 4 Question: 049 Given the following conditions:
| |
| - Unit 2 was operating at 75% power A break in the "A" Recirculation loop has occurred The "B" loop of Residual Heat Removal (RHR) is not available Reactor levelis-120" and going up
| |
| - . Reactor pressure is 350 psig and dropping
| |
| - Drywell pressureis 10 psig going up All other automatic actions have occurred as designed Select the reason why maximum toms cooling cannot be IMMEDIATELY started.
| |
| : a. High Pressure, Service Water system capacity will not support the additional heat load of torus cooling for ten (10) minutes.
| |
| : b. The RHR pumps cannot be stopped for a transfer to torus cooling for 6ve (5) minutes following aninitiation signal.
| |
| : c. The Transient Response Implementation Plan procedures require maximum injection flow for the first ten (10) minutes fcilowing a LOCA.
| |
| : d. The valve repositionings required for torus cooling cannot be completed for five (5) minutes following a LOCA.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal System", Rev. 010, Handout 0370-8, Pages 1 &
| |
| 2, LO - 1.c & 5.a Exam Leveh Both History: New K/A: 219000A214 4.1/4.3 KA Statement: Loss ofcoolant accident System / Evolution: Torus / Suppression Pool Coohng Mode Exam Section: Plant Systems RO Group: 2 SkO Group: 2
| |
| | |
| Question: 050 With the Unit 2 Refueling Platform unloaded and near the core, which of the following ALOhT will prevent further platform movement toward the core? I
| |
| : a. The Reactor Mode Switch is in "Stanup/ Hot Standby" ,
| |
| : b. The Main Grapple"Not Full Up".
| |
| : c. The Reactor Mode Switch is in " Refuel" -
| |
| : d. One control rod is withdrawn to Notch "08" Answer: a
| |
| | |
| ==References:==
| |
| FH-0762, " Refueling Bridge And Operations", Rev. 00a, Section E.4.b.(2), Page 42, s l
| |
| LO-14 LOT-0080, " Reactor Manual Control System", Rev. 007, LO - 2.b & 4.a ,
| |
| Exam Level: Both History: New K/A: 234000K401 3.3/4.1 KA Statement: Prevention of core alterations during control rod movements System /Evolutira: Fuel Handling Equipment Exam Section: Plant Systems RO Group: 3 SRO Group: 2 l l
| |
| l I
| |
| e
| |
| | |
| Question: 051 -
| |
| Given the following conditions:
| |
| - Unit 3 is operating at 70% power Main Steam Isolation (MSW) stroke testing is in progress
| |
| - The Outboard MSN on the "D" Main Steam Line test pushbutton has been depressed When the Operator depresses the test pushbutton it sticks in the " depressed" position The Operator does not notice and no other actions are taken Which of the following is the expected final position of this MSW?
| |
| : a. 10% closed position
| |
| : b. 50% closed position
| |
| : c. 90% closed position
| |
| : d. 100% closed position Answer: d
| |
| | |
| ==References:==
| |
| LOT-0120, " Main Steam And Pressure Relief System", Rev. 014, Section W.B.2 &
| |
| Figures 5 & 6, Pages 22 & 23 Exam level: Both History: New K/A: 239001A401 4.2/4.0 KA Statement: M S N 's System / Evolution: Main And Reheat Steam System Exam Section: Plant Systems RO Group: 2 SRO Group: 3 1
| |
| | |
| Question: 052 Which of the following describes how a ics3 ofinstrument air can cause a control rod drift?
| |
| Decreasing instrument air pressure causes:
| |
| : a. both scram backup valves to slowly fail to the "open" position.
| |
| : b. the hydraulic control unit scram inlet and outlet valves to begin to open.
| |
| : c. the Alternate Rod Insertion valves to all slowly fail to the "open" position.
| |
| : d. the in-service Control Rod Drive Flow Control Valve to drift fully open.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0730," Compressed Air System", Rev. 007, Section V.B.2.a, Page 18, LO - 2.f
| |
| &7 Exam Level: RO History: New K/A: 300000K302 3.3/3.4 KA Statement: Systems having pneumatic valves and controls System / Evolution: Instmment Air System
| |
| ]
| |
| Exam Section: Plant Systems RO Group: 2 SRO Gmup: 2
| |
| | |
| Question: 053 Which of the following is the reason for the Jet Compressor 05 Gas Inlet Valve (MO-2991) automatic closure on LOW steam flow of<7500 lbm/ hour?
| |
| This low steam flow isolation prevents:
| |
| : a. hydrogen concentrations in excess of the 4% limit downstream of the recombiner.
| |
| : b. moisture buildup in the off-gas charcoal adsorber beds.
| |
| : c. high o$ gas flow resulting in channeling of the charcoal adsorber bed.
| |
| : d. stagnant o5 gas flowrates in the system, specifically within the holdup pipe. 1 Answer: a
| |
| | |
| ==References:==
| |
| LOT.0510, "J5 Gas Recombiner System", Rev. 011, Sections III.D.5 & IV.B.4, Pages 10 & 21, LO - 1.d and 4.a & k Exam level: Both History: New K/A: 271000K408 3.1/3.3 KA Statement: Automatic systemisolation System / Evolution: Offgas System Exam Section: Plant Systems RO Group: 2 SRO Gmup: 2 {
| |
| l I
| |
| 1
| |
| | |
| Question: 054 One of the stated functions of the Traversing In-Core Probe (TIP) system is to calibrate the LPRM instrumentation while operaung at power.
| |
| Which of the following describes how the TIP insuumentation itselfis calibrated?
| |
| ' a~ The det' ecto- desgm is unh that routine calibration is not required, however, they are physically checked 12 r:ompared with a reference value each refuehng outage.
| |
| : b. A comparison is mMe between the detector electronic output and a " standard" value based on the detector age, the number of hours it has been inserted in the core and core power whileinsened.
| |
| : c. A comparison of each of three TIP detectors is made by inserting them in tum into the same LPRM stnng position in the core and then normahzing to 100% power.
| |
| : d. Since the detectors are not routmely subjected to neutron Bux, the iM@ly verified factory calibration settmgs are considered valid between rephe cycles, typically every other refueling outage.
| |
| Answer: c t ~,;
| |
| | |
| ==References:==
| |
| LOT-0290, "Traversmg In-Core Probe", Rev. 008, Sections II.B & IV.F, Pages 5-6 & ;
| |
| 5 14, LO - 1.b & 7 Exam level: Both Histon: New K/A: 21500lG128 3.2/3.3 KA Statement: Knowledge of the purpose and function of major system wTgs. cats and controls System / Evolution: TraversingIn-Core Probe Exam Section: Plant Systems RO Group: 3 SRO Group: 3
| |
| | |
| Questi:n: 055 Which of the following is the altemate source of cooling water to the Fuel Pool Cooling and Cleanup System heat exchangers?
| |
| : a. High Pressure Senice Water
| |
| : b. Residual Heat Removal System
| |
| : c. Emergency Cooling Water System
| |
| : d. Reactor Building Closed Cooling Water Answer: d
| |
| | |
| ==References:==
| |
| LOT-0750, " Fuel Pool Cooling And Cleanup", Rev. 007, Section V.5, Pages 10 & 11, LO-5 Exam Level: RO History: New K/A: 233000A208 2.9/3.1 KA Statement: Closed cooling water failure
| |
| ! \
| |
| System / Evolution: Fuel Pool Cooling And Cleanup Exam Section: Plant Systems RO Group: 3 SRO Group: 3
| |
| )
| |
| | |
| Question: 056
| |
| .- Given the following conditions:
| |
| - Unit 3 is shutdown Both Recirculation Pumps are shutdown with their discharge valves closed *
| |
| - The "A" loop of Residual Heat Removal (RHR) is being lined up for shutdown cooling What protects the "A" RHR Pump from damage due to no or low Bow as it is being started?
| |
| : a. The operator will manually operate the pump minimum Bow valve (MO-16A) as necessary
| |
| : b. A discharge flow-path must be established for the pump as soon as possible after it is started.
| |
| : c. ' The "A" RHR Pump will trip on low Bow if flow is not greater than 500 gpm in 10 i seconds.
| |
| : d. The pump mmimum flow valve (MO-16A) will automatically open until system flow
| |
| * exceeds 500 gpm for 10 Wm Answer: b' l
| |
| | |
| ==References:==
| |
| SO 10.1.B-2, " Residual Heat Removal system Shutdown Cooling Mode Manual Start", Rev.16, Section 4.7.2 & Caution 4.7.11, Pages 4 & 8 LOT-0370, " Residual Heat Removal", Rev. 010, LO. - 5 d
| |
| - Exam Level: RO History: New K/A: 205000G102 3.0/4.0 KA Statement: Knowledge ofoperator responshhties during all modes of plant operation
| |
| -System/ Evolution: Shutdown Coohng (RHR Shutdown Cooling Mode) l Exam Section: Plant Systems RO Group: 2 SRO Group: 2 e
| |
| | |
| Question: 057 Given the following conditions:
| |
| -- Unit 2 is perfonning a unit startup
| |
| - The main turbine is rolling at 1500 rpm with "1800" rpm selected at the " Fast" startmg rate
| |
| - The Unit Reactor Operator reports that the turbine is ad 4.g at approximately 60 rpm
| |
| , Which of the following desc:ibes the current main turbine /EHC operation?
| |
| : a. The speed circuit has failed to the "1500" speed set rpm output
| |
| : b. A loss ofboth EHC speed signals has occurred
| |
| : c. The acceleration circuit has faded to the " Slow" startup rate.
| |
| : d. The turbine is ad..img at the .pyrupiiete rate as selected by the speed control unit.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0590, " Electro-Hydraulic Control Logic", Rev. 007, Section III.C, Pages 12-15, LO - 2.n & 3.1
| |
| -f i Exam Ievel: RO History: New K/A: 245000K108 3.4/3.5 KA Statement: Reactor / turbine pressure control system System / Evolution: Main Turbine Generator And Auxiliary Systems Exam Section: Plant Systems RO Gesup: 2 SRO Group: 2 l
| |
| i ens N 4
| |
| | |
| 1 s l Question: 058 Given the following conditions:
| |
| The Unit 2 - 2A 250 Volt battery charger has been lined up and is performing an "equahze" charge on its battery
| |
| - During the charge, AC power to the charger is lost and subsequently restored when the bus is reenergized by the diesel generator Which of the following describes the expected response of this battery charger?
| |
| The 2A Battery Charger will:
| |
| : a. retum in the"equahze" charge mode.
| |
| : b. trip and must be manually restored as permitted by diesel generator loading.
| |
| : c. retumin the" float" charge mode.
| |
| : d. trips and cannot be restored with the diesel generator powering the bus. .
| |
| q Answer: a
| |
| | |
| ==References:==
| |
| LOT-0690, "DC Distribution", Rev. 009, Section III.A.2.b.2), Pages 7 & 8, LO - 2.b
| |
| & 7.a Exam Level: RO History: New K/A: 263000K601 3.2/3.5 KA Statement: A.C. electrical distribution System / Evolution: D.C. Electrical Distnhition i
| |
| Exam Section: Plant Systems RO Group: 2 SRO Group: 2 i I
| |
| | |
| Question: 059 -
| |
| wi During a loss of Control Room Chilled Water, the operator is directed to place Control Room ventilation in the" Purge" mode Which of the following describes what physically occurs to the Control Room ventilation system when it transfers to " Purge" for these conditions? -
| |
| The normalventilation fans:
| |
| : a. continue to run with fresh air being directed through the high efficiency fiher and exhaust air exiting via the normal path.
| |
| : b. continue to run with fresh air entering via the "A/C" Supply Fan and exhaust air bemg directed to the roof.
| |
| : c. fans stop, the Emergency Var *H4 Fans start with fresh air directed through the high efficiency fiker and exhaust air exiting via the normal path.
| |
| : d. fans stop, the Emergency Ventilation Fans sta:t and fresh air is supplied to the Control Room with exhaust air directed to the roof.
| |
| / i Answer: b
| |
| | |
| ==References:==
| |
| LOT-0450 " Control Room Ventilation", Rev. 010, Sections IV.B.8 & V.C, Pages 10
| |
| & 11, LO - 2.b & 4.b Exam Level: RO History: New K/Ai 290003A40'l 3.2/3.2 KA Statement: Initiate / reset system System / Evolution: ControlRoom HVAC Exam Section: Plant Systems RO Group: 2 SRO Group: 2 L
| |
| | |
| i
| |
| [
| |
| Question: 060 a ,.i Select the plant system that does NOT receive an input from the Main Steam The Radia''-
| |
| Monitoring System.
| |
| : a. Primary Contamment Isolation System i
| |
| : b. Reactor Protection System
| |
| : c. Off-Gas Recombiner System
| |
| : d. Main Condenser Air Removal System Answer: c .
| |
| i Refertnces: LOT-0720," Process Radiation Monitoring", Rev. 010, Section III.C.1.b.8), Page 21 i Exam Ixvel: RO History: New K/A: 272000K101 3.6/3.8 l KA Statement: Main steam system System / Evolution: Rariiation Monitoring System
| |
| ( i Exam Section: Plant Systems RO Gesup: 2 SRO Group: 2 j
| |
| | |
| Question: 061 -
| |
| With Unit 2 at 100% power, a trip ofone Condensate Pump occurs.
| |
| Which of the foUowing is the reason why a Reactor Feedwater Pump does not trip as well? Assume all plant systems are operating as designed.
| |
| The Condensate Pump trip:
| |
| : a. does not result in a significant change in feedwater pump suction header pressure.
| |
| : b. opens a bypass valve around the condensate demmerahzers maintaining feed pump suction pressure
| |
| : c. results in a tim'e delay on the feedwater pump low suction pressure trips to allow system pressures and flows to stahhze while ri. hy full power.
| |
| : d. reduces reactor power and hmits reactor fer.swater pump speed to prevent the pump trip.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0520,"CWwe System And Demmerahzers", Rev. 006, Sections IV.B.3 &
| |
| (" 1 V.D, Pages 20 & 21, LO - 2.b & 4.d Exam Level: RO History: ,New-K/A: 256000K304 3.6/3.7 KA Statement: Reactor feedwater system System / Evolution: Reactor Condensate System Eram Section: Plant Systems RO Group: 2 SRO Group: 3 L
| |
| i
| |
| ~
| |
| | |
| . Question: 062 The E-3 Diesel Generator is being run for a surveillance test loaded to 2500 KW on the E-33 bus when aloss ofoff-site power occurs.
| |
| Select the expected response of the E-3 diesel and its output izeakers.
| |
| : a. E-3 will trip and lockout, Operator action is necessary to restert the diesel generator and closeits output breakers.
| |
| : b. E-3 will trip and automatically restart after 'l minute, its output breakers will close automatically.
| |
| : c. E-3 will continue to run, the output breakers will trip and Operator action is necessary ;
| |
| to reclose them. !
| |
| I
| |
| : d. E-3 will continue to run, the E-33 output breaker will stay shut and the E-23 output breakerwillclose automatically. j Answer: c
| |
| | |
| ==References:==
| |
| LOT-0670, " Diesel Generator And Atuaharies", Rev. 006, Section IV.B.1, Page 31, I LO - 6.e
| |
| /i ' , SO 52A.I.B, " Diesel Generator Operations", Rev.15, Precautions 3.6 and 3.7, page 2 )
| |
| Exam level: RO History: New I K/A: 264000K608 3.6/3.7 KA Statement: A.C. power System / Evolution: Emergency Generators (Diesel / Jet)
| |
| Exam Section: Plant Systems RO Group: 1 SRO Group: 1 M
| |
| | |
| e Question: 063 -
| |
| .e~ . j While performing a reactor startup on Unit 2, the Unit Reactor Operator has met the requirements for and is withdrawing the Source Range Monitor (SRM) detectors.
| |
| Which of the following describes the power monitoring capabilities of the SRMs while the detectors are moving?
| |
| : a. SRM power and period indications are rehable.
| |
| : b. SRM power is reliable but period will NOT be rehable.
| |
| : c. SRM power and period indications are NOT reliable.
| |
| : d. SRM power is NOT reliable but period is rehable.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0240, " Source Range Monitoring System", Rev. 008, Section V.A.6, Page 15, LO - 6.b Exam Imel: RO History: New
| |
| ~
| |
| K/A: 215004A102 3.6/3,7
| |
| ( !)
| |
| KA Statement: - Reactor powerindication System / Evolution: Source Range Monitor System Exam Section: f/lant Systems RO Group: 1 SRO Group: 1 L
| |
| | |
| Question: 064 Given the following conditions:
| |
| A reactor startup is in progress on Unit 3 Reactor pressureis 900 psig Scram time testingisin progress The scram INLET valve FAILS TO OPEN during the test Which of the following is the expected response of this control rod?
| |
| : a. The rod will fully insert on acemnteor pressure.
| |
| : b. The rod will fully insert on reactor pressure.
| |
| : c. The rod will partiallyinsert
| |
| : d. The rod willNOTinsert Answer: b
| |
| | |
| ==References:==
| |
| LOT-0060, " Control Rod Drive Mechanism", Rev. 009, Section III.B.4 & Figure 7, Pages 21 & 22
| |
| ]
| |
| Exam 1mel: Both History: New K/A: 2010]3K601 3.3/3.3 KA Statement: Control rod drive hydraulic system System / Evolution: ControlRod And Drive Mechanism Exam Section: Plant Systems RO Group: 2 SRO Group: 3 L
| |
| | |
| f Question: 065 Wnich of the following describes the final status of the Unit 2 feedwater system and Reactor Feedwater Pumps (RFP) following a normal reactor scram? Assume that all expected Unit Reactor Operator actions were taken.
| |
| : a. "A" and "B" RFPs emergency stopped, "C" RFP running on the start-up level controller
| |
| : b. "A" and "B" RFPs tripped, "C" RFP runnmg on the start-up level controller I
| |
| : c. "B" and "C" RFPs emergency stopped, "A" RFP running on the mmimum flow valve
| |
| : d. "B" and "C" RFPs tripped. "A" RFP runnmg on the mmimum flow valve Answer: a Referencer: Exhibit OM-P-16.1:5,"OSPS: Reactor Operator Response To Reanor Scram", Rev.
| |
| 4, Section 5, Page 2 Exam Level: Both History: New K/A: 295006A102 3.9/3.8
| |
| : i. .
| |
| KA Statement: Reactor waterlevel control system System / Evolution: Scram Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1
| |
| | |
| i Question: 066 i
| |
| Table DWTf-1, "RPV Level Instrument Status" states that a reactor water level instrument may not be used if drywell temperature is at or above the RPV Saturation Curve.
| |
| Which of the following is the reason for this requirement?
| |
| The instrument's:
| |
| : a. reference leg is assumed to have flashed, causing level to read falsely high.
| |
| : b. vanable leg is assumed to have flashed, causing level to read falsely low.
| |
| : c. reference and variable legs are assumed to have flashed, causing level to read falsely low.
| |
| : d. reference and variable legs are assumed to be undergo'mg " outgassing", cauung level to -
| |
| read falsely high.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| - T-102 Pnmary Containment Control- Bases, Rev.12, Step DW/f-4, Pages 18 & 19 I I LOT-0050, " Reactor Vessel Instrumentation", Rev. 013, Section V.A.2, Pages 31 &
| |
| 32, LO - 6.b.1 & 6.o F_mam Level: Both History: New K/A: 295028K101 3.5/3.7 KA Statement: Reactor waterlevel measurement System / Evolution: High DrywellTernperature Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2 L
| |
| | |
| t Question: 067 Given the following conditions for Unit 2:
| |
| - A small LOCA has occurred resultmg in a reactor scram on high drywell pressure
| |
| - A failure-to-scram (ATWS) occurred,130 control rods did not insert
| |
| - Reactor water level was lowwed to and maintained between -100 and -120 inches with feedwater
| |
| - Reactor pressure is cunently 950 psig
| |
| - Standby Liquid Control (SLC) Tank level has decreased by 36% and SLC is still irjecting Which of the following could cause a reactor iwiikality for the conditions?
| |
| : a. Placing RCIC in service to maintain reactor water level
| |
| : b. Failuretoinhibit ADS
| |
| : c. Continued pressure drop due to the LOCA
| |
| : d. Decreasing Xenon concentration from decay over the first two hours after the failure to l
| |
| Answer: c i I
| |
| | |
| ==References:==
| |
| Trip Curves, Tables and Limits - Bases, Rev. 2, Section 2, Page 2 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 6 Exam level:' Both History: New K/A: 295037K104 3.4/3.6 KA Statement: Hot shutdown boron weight System / Evolution: SCRAM Condition Present And Reactor Power Above APRM Downscale Or Unknown Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1 I
| |
| ; j 1
| |
| I i
| |
| l
| |
| | |
| Question: 068
| |
| . Per ON-124," Fuel Floor And Fuel Handling Problems", which of the following requires the operators to take the actions tbr "~it cality" i during fuel handling operations? Assume the fuel movement is inside the reactor vessel.
| |
| : a. An unexpected Refuei Floor Area high radiation alarm is received.
| |
| : b. Refuel Bridge reverse motion (towards the core) interlock activates.
| |
| : c. Source Range nuclear instrumentation counts are uvMiy increa.sg.
| |
| : d. Source Range nuclear instmmentation counts are spiking r-Wly.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| ON-124, " Fuel Floor And Fuel Hnadling Problems", Rev.1, Symptoms Section 1, Page1 LOT-1550, "Off-Normal Procedures", Rev. 006, LO - 1 Exam level: Both History: New
| |
| : -, K/A: 295023A106 3.3/3.4 KA Statement: Neutron monitoring System / Evolution: Refueling Accidents Fumm Section: Emergency And AbnormalPlant Evolutions RO Group: 3 SRO Group: 1
| |
| )
| |
| i
| |
| -m i
| |
| * 1 5
| |
| | |
| i Question: 069 -
| |
| w Given the following conditions:
| |
| - Unit 2 is operating at 100% power
| |
| - A failure ofone Jet Pump has occurred
| |
| - No operator actions have been taken Which of the following is the expected response of the " Core Plate D/P" and the "Recire Drive Flow To Loop Containing The Defective Jet Pump"?
| |
| RECIRC DRIVE FLOW TO CORE PLATE LOOP CONTAIN:NG THE D/P DEFECTIVE JET PUMP
| |
| : a. Increase Decrease
| |
| : b. Decrease Increase
| |
| : c. Increase Increase
| |
| : d. Decrease Decrease
| |
| (- )
| |
| Answer: b
| |
| | |
| ==References:==
| |
| ON-100, " Failure Of A Jet Pump", Rev. 3, Symptoms #1 & 2, Page 1 LOT-0050," Reactor VesselInstrumentation", Rev. 013, Section III.L Page 25 LOT-0030, " Reactor Recirculation System", Rev. 009, LO - 2.f& 6.b Fumm Level: RO History: New K/A: 295001A203 3.3/3.3 KA Statement: Actual core flow Systein/ Evolution: Parnal Or C@e 1.oss Of7orced Core Flow Cir=1+
| |
| Fum= Section: Emergency And AbnormalPlant Evolunons RO Group: 2 SRO GriupI 2
| |
| | |
| i Question: 070 l
| |
| Following a transient on Unit 2, the following conditions exist:
| |
| I -- Diywell pressure: 4.5 psig and rising
| |
| - Diywelltemperature: 140 'F and nsing
| |
| - Torus pressure: 8.4 psig and riang
| |
| - Torus water temperature: 82 *F and stable Which of the following has occurred?
| |
| : a. A safety reliefvalve has opened and reclosed
| |
| : b. A pipe break into the drywell has occurred with a torus to flywell vacuum banker open.
| |
| : c. A safety reliefvalve tail pipe has broken above the torus water level while the valve is open.
| |
| : d. A recirculation line break has occurred with all containment parameters ie-;-:.rd% as designed.
| |
| Answer: c
| |
| (
| |
| | |
| ==References:==
| |
| LOT-0130," Primary C1 d= - =^", Rev. 010, Section 5.D and Figure Sc, Page 26, LO - 5.a & 7.c Enam level: Both History: New K/A: 295024A206 4.1/4.1 KA Statement: Suppression pooltemperature SystenvEvolution: HighDrywellPressure Fum= Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 M
| |
| l
| |
| | |
| i 1
| |
| Question: 071 With all control rods fully inserted, T-111, " Level Restoration", in the steam cooling section, directs an Emergency Blowdown when reactor water level reaches -210 incher with no injection systems available.
| |
| What is the bases for this action?
| |
| : a. Core fuel temperatures will decrease allowing additiox.1 time for restoration of some source ofinjection.
| |
| : b. Reactor prescre will be reduced causing core voiding to increase adding negative reactivity resulting in a power drop.
| |
| : c. Reactor core differential pressure will decrease assistmg the thermal driving head for natural circulation flow.
| |
| : d. The total amount ofenergy that could be released when fuel melt begins will be reduced while some waterinventory remams Answer: a
| |
| | |
| ==References:==
| |
| T-111 Level Restoration - Bases, Rev. 8, Step LR-20, Page 7 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Exan Level: Both History: New K/A: 295031K305 4.2/4.3 KA Statement: Emergency depressurization System / Evolution: Reactor Low Water Level Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1
| |
| | |
| I l
| |
| l l
| |
| l Questi::n: 072 Given tne following conditions
| |
| -- Unit 2 has had a failure-to-scram (ATWS) on high drywell pressure
| |
| -- The Automatic Depressunzation System (ADS) initiation was not inhibited Which of the following will result wh n reactor water level reaches the ADS initiation setpoint?
| |
| : a. A rapid, uncontrolled reactor depressuruation, following a 9 minute time delay, resulting in a rapid power drop
| |
| : b. An increase in the injection rate of cold, unborated water resulting in a rapid power rise.
| |
| : c. Loss of boron due to entramment in the steam from the reactor to the torus, increasmg the time required to shutdown the reactor.
| |
| : d. A large reduction in effective control rod worth because of voiding, leadmg to a power excursion.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| T-101 RPV Control - Bases, Rev.19, Step RC/Q-11, Page 6 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3 Exam Imel: Both History: New K/A: 295037K209 4.0/4.2 KA Statement: Reactor waterlevel System / Evolution: SCRAM Condition Present And Reactor Power Above APRM Downscale Or Unknown Exam Section: Emergency And AbnormalPlant Evolutions RO Gesup: 1 SRO Group: 1 i
| |
| i 1
| |
| I
| |
| | |
| 9 Question: 073 During a plant startup, what is the earliest point at which entry into OT-112," Recirculation Pump Trip" is required if a pump trip occurred?
| |
| : a. The Reactor Mode Switch has been placed in "Startup/ Hot Standby".
| |
| : b. The reactor is at or above criticality.'
| |
| : c. The Reactor Mode Switch has been placed in "Run"
| |
| : d. Reactor coolant temperature is > 212 *F.
| |
| Answer; a
| |
| | |
| ==References:==
| |
| OT-112, " Recirculation Pump Trip", Rev. 22, Section 3.4, Page 5 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 1 & 3 Exam Level: Both History: New K/A: 29500lG404 4.0/4.3 KA Statement: Ability to recognize abnormal indications for system w.iks parameters which are entry level conditions for emergency and abnormal operstmg procedures System / Evolution: Partial Or Complete loss OfForced Core Flow Circulation Exam Section: Emergency And Abnonnal Plant Evolutions RO Group: 2 SRO Group: 2
| |
| | |
| l Quotion: 074 l
| |
| Given the following conditions:
| |
| l Unit 2 has experienced a loss ofoff-site power and a Loss OfCoolant Accident
| |
| -- it has been deternuned that the running diesel generators (DG) do NOT have exhg water -
| |
| available
| |
| ~ Cooling water will be available in approxunately 8 minutes
| |
| - - The Control Room Supervisor has directed the running DGs to be shutdown before the 3 minute limit with no cooling is reached l
| |
| _Which of the followmg describes how these DGs are required to ha shutdown and mair*=inad shutdown with the initiatmg signal still present?
| |
| : a. The local operator is directed to depress the Emergency Stop knob located behind the
| |
| ' Diesel Gauge Panel.
| |
| : b. The Plant Reactor Operator instal!s ajumper that directly inserts a diesel generator differential overcurrent trip signal. ,
| |
| : c. The Plant Reactor Operator places the Diesel Control Switch in " Pull-To-lock" in the Main ControlRoom.
| |
| : d. The local operator is directed to depress the Stop :=**an on the Diesel Gauge Panel.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| SE-11 Loss OfOff-Site Power - Bases, Rev. 7, Steps LP-10 & 11, Pages 6 & 7 LOT-1555, "Special Events", Rev. 004, LO - 12.b 1
| |
| Exam Imel: Both History: New i
| |
| K/A: 295003A102 4.2/4.3 l KA Statement: Emeq;ency generators Systesa/ Evolution: Partial Or Complete Loss Of A.C. Power Enm Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 1 i
| |
| l l
| |
| | |
| (
| |
| Question: 075
| |
| * I Given the following conditions:
| |
| Unit 2 is operating at 90% power The High Pressure Coolant Injection (HPCI) system is open ating for a surveillance Torus water temperatureis 108 'F Which of the following describes the TRIP and Technical Specifications (TS) procedural requirements for these conditions?
| |
| T-102," Primary Containment Control", entry:
| |
| : a. is not required and the TS LCO is not re.
| |
| : b. is not required and the TS LCO is met.
| |
| : c. is required and the TS LCOis not met.
| |
| : d. is required and the TS LCOis met.
| |
| Answer: c P.eferences: T-102, " Primary Contamment Control", Entry Conditions PBS T.S. 3.6.2.1, " Suppression Pool Average Temperature", Page 3.6-23 LOT-1840, " Tech Spec LCOs", Rev. 007, LO - 2 Exam Level: Both History: New K/A: 295013A201 3.8/4.0 KA Statement: Suppression pooltemperature System / Evolution: High Suppression Pool Temperature Exam Section: Emergency And AbnormalPlant Evolutions RO Group:
| |
| 2 SRO Group: 1 i
| |
| | |
| Question: 076 Following a complete loss of off-site power with all of the diesel generators runnmg normally with their output breakers closed, which of the following components will NOT have cooling water Dow AVAILABLE?
| |
| : a. Control Rod Drive pumps
| |
| : b. Condensatepumps
| |
| : c. HPCI room coolers
| |
| : d. RHR heat exchangers Answer: b
| |
| | |
| ==References:==
| |
| LOT-0430, " Turbine Building Closed Cooling Water System", Rev. 008, Sections II.A.5 & III.A, Page 5, LO - 2.b Exam Level: Both History:. New K/A: 295018K202 3.4/3.6 KA Statement: Plant operations Syst?m/ Evolution: Partial Or Complete Loss Of Cortponent Cooling Water Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Gmup: 2 i
| |
| 4 1
| |
| | |
| l Question: 077 During a failure-to-scram (ATWS) on Unit 2, reactor water level was deliberately lowered to control reactor power as required by T-i l7,"12 vel / Power Control" Which of the following systems is specifically prohibited from maintaining reactor water level?
| |
| l
| |
| : a. Core Spray
| |
| : b. Feedwater
| |
| : c. Reactor Core Isolation Cooling
| |
| : d. LowPressureCoolantInjection Answer: a
| |
| | |
| ==References:==
| |
| T-117, " Level / Power Control", Rev. I1, Step LQ LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3, 8 & 10 Exam Level: Both History: New K/A: 2950lx>. 'I 3.3/4.0 KA Statement: Knowledge ofoperational implications ofEOP wamings, cautions and notes System / Evolution: Incomplete SCRAM Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1
| |
| | |
| i Question: 078 l
| |
| Given the following conditions:
| |
| l 1
| |
| Unit 2 is in Mode 4 I No recirculation pumps are operating l
| |
| - The "A" Loop of Residual Heat Removal was in Shutdown Cooling l
| |
| - The"A" RHR Pump hasjust tripped l
| |
| - Reactor waterlevelis-5 inches !
| |
| No operator actions have been taken l Which of the following describes the current status of reactor coolant temperature indication?
| |
| : a. Recirculation loop "A" temperature is a valid temperature indication.
| |
| l i
| |
| : b. There are no valid temperature indications available for these conditions. ):
| |
| : c. RWCU bottom head drain temperature is a valid temperature indication.
| |
| : d. Calculated temperature from steam dome pressure is a valid temperature indication Answer: b
| |
| | |
| ==References:==
| |
| ON-125 Loss Of Shutdown Cooling - Bases, Rev.1, Section 2.7.6, Page 7 LOT-1550, "Off-Normal Procedures", Rev. 006, LO - 2 & 3 Exam Level: Both History: New K/A: 295021A204 3.6/3.5 KA Statement: Reactor water temperature System / Evolution: Loss Of Shutdown Cooling Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 3 SRO Group: 2 l
| |
| 1 I
| |
| )
| |
| | |
| i Question: 079 During a lowering torus water level transient, the operator is directed to shutdown the High Pressure Coolant Injection (HPCI) system iflevel reaches 9.5 feet even if HPCI is required to assure adequate core cooling.
| |
| This limitation will prevent:
| |
| : a. exceeding the maximum allowed pnmary containment pressure.
| |
| : b. exceeding the SRV Tail Pipe Linnit.
| |
| : c. a potentially unr ecoverable trip of HPCI on high turbine exhaust pressure.
| |
| : d. exceeding the Heat Capacity Level Limit.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| T-102 Primary Containment Control- Bases, Rev.12, Step T/L-13, Pages 7 & 8 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3 Exam Level: Both History: New K/A: 295030K302 3.5/3.7 KA Statement: HPCI operation System / Evolution: Low Suppression PoolWaterImel Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 1 I
| |
| i l
| |
| | |
| Question: 080 Which of the following is the MAXIMUM number of control rods that are allowed to drift %1THOUT REQUIRING the Reactor Mode Switch be placed in " Shutdown"?
| |
| : a. Two control rods contained in a five by five array moving in oi .4it. .
| |
| : b. . Two control rods without regard to location or whether the rods are dnfhng in or out.
| |
| : c. - One control rod, ifit is drifting out or two control rods, if they are dnfhng in.
| |
| : d. One control rod without regard to whether the rod is dnfhng in or out.-
| |
| Answer: d
| |
| | |
| ==References:==
| |
| ON-121, "Dnfhng Control Rod", Rev. 7, Step 2.1, Page 1 LOT-1550, "Off-Nonnal Procedures", Rev. 006, LO - 2 & 3 Exam level: RO History: New K/A: 295014A101 4.0/4.1 KA Statement: RPS System / Evolution: Inadvertent Reactivity Addition Fumm Section: Emergency Aid AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 f
| |
| | |
| Question: 081 .
| |
| Given the following conditions:
| |
| Unit 2 had been operating at 100% power A loss of power has resulted in a full Group 1 isolation and reactor scram
| |
| - NO control rod movement was noted The power loss has also caused a loss of all nuclear instrumentation
| |
| - HPCI and RCIC are not available Reactor pressure is bemg maintained between 950 and 1050 psig with 6 Safety Relief Valves (SRV) open and a 7th SRV bemg cycled by the operator What is the approximate reacter power for these conditions?
| |
| Poweris between:
| |
| : a. 7 - 15 %
| |
| : b. 20 - 28%
| |
| : c. 33 -41 %
| |
| : d. 46 - 54%
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0120, " Main Steam And Pressure Relief System", Rev. 014, Section III.B.2, Page 10, LO - 6.f& 7.a Exam Level: Both History: New K/A: 295007A202 4.1I4.1 KA Statement: Reactor power System / Evolution: High Reactor Pressure Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1 Justification: Per LOT-0120, each SRV is rated at approx. 5.7%, 6 SRVs open is approx. 34%
| |
| with 7th cycling about 3% more
| |
| | |
| 1
| |
| \
| |
| l 4
| |
| Question: 082 Given the following conditions:
| |
| - The Main Control Room has been evacuated due to habitability concems 1
| |
| - There was no fire or Cardox system actuation ,
| |
| -- No LOCA or electrical power concerns exist i All immediate actions were successfully completed prior to the evacuation Select the system that is NOT required to meet the MINIMAL requirements for control of the reactors i from the Remote Shutdown Panel following the scrams.
| |
| : a. High Pressure Coolant Injection j
| |
| : b. 125 VDC power 1
| |
| : c. Reactor Core Isolation Cooling i
| |
| : d. Uninterruptible AC power Answer: a .
| |
| | |
| ==References:==
| |
| SE-1 Bases, " Plant Shutdown From The Remote Shutdown Panel", Rev.15, Entry Conditions, Page 1 LOT-1555, "Speaal Events", Rev. 004, LO - 2.a & q Exam Level: Both History: New K/A: 295016K201 4.4/4.5 KA Statement: Remote shutdown pai.el System / Evolution: Control Room Abandonraent Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Genup: 1 l
| |
| t I
| |
| | |
| Question: 083 Given the fNiowing conditions Drywell sprays were initiated on Unit 3 as directed by the Primary Contamment Pressure
| |
| - leg of T-102, " Primary Containment Control" As drywell pressure and temperature are decreasing the " Unsafe" region of the Drywell Spray Initiation Limit (DSIL) curve was entered at a drywell temperature of 250 F Which of the following is the REQUIRED action?
| |
| The operator shall;
| |
| : a. terminate drywell sprays immediately.
| |
| : b. throttle drywell spray flow to remain in the " Safe" region of the DSIL curve.
| |
| : c. reduce drywell spray flow and monitor the Torus to Drywell vacuum breaker positions.
| |
| : d. ' terminate drywell sprays when drywell pressure decreases below 2.0 psig.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| T-102, "Pnmary Containment Control", Rev,11, Step PC/P-9 TRIP Curves, Tables & Limits - Bases, Rev. 2, Page 7 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Eum Level: B'oth History: New K/A: - 295024K301 3.6/4.0 KA Statement: Drywell spray operation System / Evolution: High DrywellPressure Exam Section: Emergency And Abnonnal Plant Evolutu m RO Group: 1 SRO Group: 1
| |
| | |
| Question: 084 i
| |
| Given the following conditions-
| |
| - . Unit 3 was operating at 100% power when a Group i isolation occurred on High Steam Tunnel Temperature
| |
| - A fhilure to scram (ATWS) left reactor power at 30%
| |
| - Attempts to inject boron were not successful
| |
| - Power dropped to 8% when T-240 was used to lower level and maintain it in a band of
| |
| -172" to -195"
| |
| -- Toms water levelis normal Torus cooling was late in bemg started
| |
| -- The required Emergency Blowdown was not performed until well after the attached Heat Capacity Temperature Limit (HCTL) curve was exceeded
| |
| - Which of the following is the direct result of this delayed Emergency Blowdown?
| |
| : a. Steam Dome pressure will exceed 1345 psig. ,
| |
| : b. The primary containment pressure limit will be ecW.
| |
| : c. The SRV tailpipe hydrodynamic loadmg limits will be exceeded.
| |
| : d. ' The max mn temperature will be exceeded.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| T-102," Primary Containment Control", Rev. I1, Step T/T-10 & 11 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3 Exam level: Both History: New K/A: 295026K301 3.8/4.1 KA Statement: Emergency /normaldepressunzation System / Evolution: Suppression pool high water temperature Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 1
| |
| | |
| CURVE T/T-1 l ,
| |
| HEAT CAPACI1Y TEMP LluIT
| |
| - 250 II
| |
| * I 2e I I w
| |
| I :
| |
| K!
| |
| E 230 -
| |
| UNSAFE 3 220 0" UI 210 ,
| |
| w 200 SAFE N A
| |
| = %
| |
| E 150 N-180 . , ,
| |
| A
| |
| - 0 200 400 800 000 1000 1200 REACTOR PRE!!URE (PSIG) l
| |
| | |
| I
| |
| ~
| |
| Question: 085 l
| |
| 'Given the following conditicns: I Unit 2 is operating at 75% power An unexplained Refuel Floor Ventilation Exhaust high radiation alann has been received .
| |
| T-103, " Secondary Containment Control", has been entered and directs the operator to l verify Reactor Building and Refuel Floor ventilation isolations and that Standby Gas l Treatment has initiated What is the bases for this verification?
| |
| This verification ensures the secondary co.ht:
| |
| : a. atmosphere will be treated and controlled as an elevated release.
| |
| 1 i
| |
| : b. atmosphere will be treated and controlled as a ground level release.
| |
| : c. atmosphere will be held up for 90 hours and controlled as an elevated release.
| |
| : d. atmosphere will be held up for 90 hours and controlled as a ground level release.
| |
| Answer; a
| |
| | |
| ==References:==
| |
| . T-103 Secondary Contamment Control- Bases, Rev. 9, Step SCC-1, Page 5 LOT-1560,"PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Exam Level: Both History: New l
| |
| K/A: 295034A104 4.1/4.2 j KA Statement: SBGT/FRVS l
| |
| System / Evolution: Secondary Contamment Ventilation High Radiation !
| |
| Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 1
| |
| | |
| i Question: 086 Given the following conditions:
| |
| Both Units were operating at 100% power Toxic gas concerns have required the Main Control Room to be evacuated
| |
| - The immediate actions of SE-1," Plant Shutdown From The Remote Shutdown Panel",
| |
| have been completed Which of the following Unit systems are available for reactor water level control from the Remote Shutdown Panel?
| |
| : a. Condensate and Core Spray pumps
| |
| : b. Control Rod Drive Hydraulic Pumps and Reactor Core Isolation Cooling
| |
| : c. Reactor Core Isolation Cooling and Reactor Feed Pumps
| |
| : d. High Pressure Coolant Injection and Reactor Core Isolation Cooling Answer: b
| |
| | |
| ==References:==
| |
| SE-1, " Plant Shutdown From The Remote Shutdown Panel", Rev.15, Immediate Operator Actions, Page 1 Exam level: RO History: New
| |
| {
| |
| K/A: 295016A106 4.0/4.1 KA Statement: Reactor waterlevel System / Evolution: ControlRoom Abandonment Exam Section: Emergency And Abnormal Plant Evolutions RO Givup: 2 SROGroup: 1
| |
| | |
| Question: 087 Given the following conditions:
| |
| -- Unit 2 is operating at 100% power The Unit Reactor Operator notes drywell pressure is 0.9 psig and increasing slowly Which of the following actions are REQUIFID to be taken?
| |
| : a. Tenninate surveillance testing that may be adding energy to the contamment.
| |
| : b. Reduce reactor power in accordance with GP-9 and enter T-102.
| |
| : c. Isolate Drywell inerting and increase drywell cooling.
| |
| : d. Make preparations to vent the drywell to mamtam pressure less than 1.2 psig.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| . OT-101, "High Drywell Pressure", Rev. 9, Section 2.1 & 2.2, Page 1
| |
| ' LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 2 Exam level: RO History: New K/A: 295010K205 3.7/3.8 !
| |
| I KA Statement: Drywell cooling and ventilation )
| |
| i System / Evolution: . High DrywellPressure .j i
| |
| I Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: I '
| |
| l l
| |
| l
| |
| | |
| Question: 088 3 Given the following conditions:
| |
| Unit 2 was performing a normal shutdown
| |
| -- During the shutdown a loss of feedwater resulted in a scram l A hydraulic ATWS occurred and power stabilized at 10%
| |
| - The operator was directed to continue control rod insertions Which of the following describes how the control rods should be inserted for these conditions?
| |
| The operator should:
| |
| : a. remove the group scram fuses in the RPS cabinets.
| |
| : b. locally vent the Scram Air Header.
| |
| : c. uip the RPS power supply breakers.
| |
| : d. bypass the Rod Worth Minimi= and insert control rods using " Emergency In" Answer: d
| |
| | |
| ==References:==
| |
| LOT-0080, " Reactor Manual Control System", Rev. 007, Section IV.A.2.d, Page 11, LO - 2.b & 4.e Exam level: Both History: FEBQ #3637, Examiner Modified (Reword stem, change format, change 2 distracters)
| |
| K/A: 295015A103 3.6/3.8 KA Statement: RMCS System / Evolution: Incomplete SCRAM Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 I
| |
| | |
| r 4
| |
| I i
| |
| Question: 089 l
| |
| Unit 3 was operating at 100% power when a transient occurred.
| |
| l The following are the peak radiation levels which existed on Unit 3 during the transient:
| |
| 1
| |
| -- Main Steam Line Radiation - 2000 mr/hr
| |
| - Main Stack Radiation - 20 cps
| |
| - Steam Jet Air Ejector Discharge Radiation - 20 mr/hr i
| |
| -- Reactor Building Ventilation Exhaust Radiation - 20 mr/hr Which of the following components should have received a PCIS signal in direct response to these radiation levels?
| |
| : a. OffGas Line isolation valve
| |
| : b. Main Steam isolation valves
| |
| : c. Mechanical Vacuum Pump suction valve
| |
| : d. Standby Gas Treatment filter train isolation valves Answer: d
| |
| | |
| ==References:==
| |
| LOT-0180, " Primary Containment Isolation System", Rev. 009, Sections II.D.8 &
| |
| IV.A2, Page 14, LO - 2j Exam Level: Both History: New K/A: 295020K211 3.2/3,4 l
| |
| KA Statement: Standby gas treatment system /FRVS System / Evolution: Inadvertent Contamment Isolation ,
| |
| Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2 i
| |
| i
| |
| | |
| Question: 090 -
| |
| Given the following conditions:
| |
| - The E-3 Diesel Generator (DG) is running in parallel with the E-32 Bus for sun'eillance testing
| |
| -- A loss of 125 VDC Prael 30D23 supplying the E-3 DG has occurred.
| |
| Which of the following describes thi expected status of the diesel for this failure?
| |
| : a. - The DG output breaker will trip and the engine will overspeed but will NOT trip.
| |
| : b. DG voltage will decrease due to a loss of the field flash supply.
| |
| : c. All DG alarms will be disabled and the engine will imMietely trip
| |
| : d. The DG will remam at speed and fully loaded.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0670, " Diesel Generator And Auxtharies", Rev. 006, Section IV.H.3, Page 50, LO - 2.b & 6.f Exam level: Both History: New K/A: 295004A102 3.8/4.1' KA Statement: Systems - y to assure safe plant shutdown System / Evolution: Partial Or Complete Loss of D.C. Power Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 e
| |
| | |
| 4 Questirn: 091 During performance of ON-119," Loss OfInstrument Air", what is the specific reason why an
| |
| . instrument air pressure of 75 psig was chosen as the limit at which GP-3, " Fast Power Reduction" must be initiated?
| |
| This pressure was chosen to:
| |
| l l a' reduce power to provide a safety margm due to the potential loss of feedwater heatmg.
| |
| i
| |
| : b. prevent the unit with the failing air system from affecting the other unit once the header cross-connect valves are opened
| |
| : c. begin the power reductie well before any of the expected random control rod drifts begin
| |
| : d. avoid the potential loss of the main condenser on a Group I isolation signal with the plant i
| |
| at high power, Answer: d -
| |
| | |
| ==References:==
| |
| ON-119 Loss OfInstrument Air - Bases, Rev.10, Operator Action #2, Page 2 LOT-1550, "Off-Normal Procedures", Rev 006, LO - 2 & 3 Exam level: Both History: New K/A: 295019A202 3.6/3.7 KA Statement: Status of safety-rebted instrument air system loads System / Evolution: Partial Or Complete loss OfInstrument Air l
| |
| Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2
| |
| | |
| Question: 092 Given the following conditions:
| |
| - Unit 3 is operating at 40% power
| |
| - The "B" Stator Water Cooling (SWC) Pump is out of service for maintenance The"A" SWC Pump hasjust tripped
| |
| - All automatic action are occumng as designed I
| |
| - The Turbine Building Operator reports the "A" SWC Pump will not restart i
| |
| Which of the following are all the required operator actions for these conditions:
| |
| : a. Verify both Recirculation Pumps trip and place the Reactor Mode Switch to " Shutdown"
| |
| : b. Verify the "A" Recirculation Pump trips and power stabilizes within single loop limits.
| |
| : c. Verify turbine generator runback, reduce VARS to minunum and verify the "A" Recirculation Pump trips.
| |
| : d. Verify turbine generator runback and reduce VARS to muumum.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| OT-113, " Loss Of Stator Cooling", Rev. 4, Section 2.3, Page 2 LOT-0600, " Main Generator And Auxdiaries", Rev. 008, LO. - 2.i & 6.d Exam Ixvel: Both History: New K/A: 295005A205 3.8/3.9 KA Statement: Reactor power System / Evolution: Main Turbine Generator Trip
| |
| , , , Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 2 V
| |
| | |
| Question: 093 Given the following initial conditions: I
| |
| -- Unit 2 is operating at 80% power
| |
| 'A' and 'C' Condensate pumps are running )
| |
| - . All three Reactor Feed Pumps are maintaining reactor level ,
| |
| i The following transient occurs:
| |
| - The ' A' Condensate Pump trips
| |
| -- Reactor levelis 19" and dropping Which of the following actions should be taken imMietely?
| |
| : a. Verify the recirculation system automatic runback to 45% speed
| |
| : b. Reduce power in accordance with GP-9-2," Fast Reactor Power Reduction" l
| |
| : c. Take manual control of the Reactor Feed Pump Master Level Controller
| |
| : d. Select all three Reactor Feed Pumps to MSC Control j
| |
| Answer: b
| |
| | |
| ==References:==
| |
| OT-100, " Reactor Iow Level", Rev. 9, Section 1.0, Page 1 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 1 Exact level: Both History: New j l
| |
| K/A: 295009K202 3.9/3.9 I l
| |
| KA Statement: Reactor waterlevelcontrol i System / Evolution: Low Reactor WaterIxvel 1
| |
| Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1
| |
| | |
| Question: 094 -
| |
| Given the following conditions:
| |
| Unit 2 is operating at 25% power
| |
| - All plant systems are operating as designed for this power level
| |
| -- Main condenser vacuum is 28.5 inches Hg Vac and is slowly lowering 1 I
| |
| - No operator actions are taken Which of the following is the expected sequence of the plant response?
| |
| : a. Reactor scram, Main Turbine trip, Reactor Feedwater Pump trip, Main Turbine Bypass Valves close
| |
| : b. Main Turbine trip, Reactor scram, Reactor Feedwater Pump trip, Main Turbine Bypass Valves close
| |
| : c. Reactor scram, Main Turbine Bypass Valves close, Reactor Feedwater Pump trip, Main Turbine trip.
| |
| : d. Main Turbine trip, Reactor scram, Main Turbine Bypass Valves close, Reactor Feedwater Pump trip Answer: a
| |
| | |
| ==References:==
| |
| OT-106, " Condenser Low Vacuum", Rev. 2, Step 4.0, Page 9 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 2 & 5 Exam level: Both History: New K/A: 295002G448 3.5/3.8 KA Statement: Ability to interpret control room indications to verify the status and operation of system and understand how operator actions and directives affect plant and system conditions System / Evolution: Loss OfMain Condenser Vacuum Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2
| |
| | |
| o Question: 095 While experiencing HIGH torus water level control problems, the Unit Reactor Operator inadvertently opens ont Safety Relief Valve (SRV) while operating in the " Unsafe" region of the attached SRV Tail Pipe Limit curve Which of the following could be expected to result from this curve violation?
| |
| Opening the SRV under these conditions:
| |
| : a. results in excessive hydrodynamic loadmg due to the excessive water levels in the tailpipe.
| |
| : b. will result in direct suppression chamber pressurization due to the quenchers being uncovered.
| |
| : c. results in drawing water up into the tailpipe because the vacuum breakers are under water.
| |
| : d. will result in valve seat damage from the excessive flowrates due to the reduced toms waterlevelin the tailpipe.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| TRIP Curves, Tables & Limits - Bases, Rev. 2, SRV Tail Pipe Limit, Page 11 )
| |
| I LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 6 & 8 Exam Level: Both History: New K/A: 295029K206 3.4/3.5 KA Statement: SRV's and discharge piping System / Evolution: High Suppression PoolWater Level Esam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 l 1
| |
| I l
| |
| i
| |
| )
| |
| | |
| CURVE T/L-3 -
| |
| SRY T AIL PIPE L1WII 17.8 ;
| |
| ' ' ' {
| |
| 17.4 i 8 -
| |
| N I UNS AFE -
| |
| g 17.0 '
| |
| - I '
| |
| i i i i # i -
| |
| i i\ l c I ' '
| |
| ~
| |
| 16.6 '
| |
| ' ! \"
| |
| SAFE ' I C 18.2 ,
| |
| l , j ,
| |
| f i , ,
| |
| g
| |
| - e i i g 15.8 ,
| |
| l i l
| |
| \
| |
| - i , i i g; o 33 4 '
| |
| o l ,
| |
| s i l i.
| |
| 15.0 , , i i i ,
| |
| 8 , i s 0 200 400 S00 800 1000
| |
| - RFY PRESSURE (PSIC) a
| |
| | |
| o a
| |
| Questi:n: 096 Given the following conditions:
| |
| Unit 3 has entered T-103," Secondary Containment Control" after receiving an alarm for high water levelin the HPCI Room Which of the following could be used to make the determination that water level is at or above the
| |
| " Action" level WITHOUT physically entering the room.
| |
| Water level may be considered to be above the " Action" level:
| |
| : a. by computer point verifications of ECCS room levels on SPDS.
| |
| : b. ifone Reactor Building Door drain sump pump is running.
| |
| : c. if battery grounds are received with indications of a HPCI Cardox discharge.
| |
| : d. if the Turbine Building floor drain sump high-high level alarm is received. -
| |
| Answer: c
| |
| | |
| ==References:==
| |
| T-103 Secondary Containment Control - Bases, Rev. 9, Entry Conditions, Pages 2 & 3 LOT-1560,"PBS TRIP Procedures", Rev. 007, LO - 2,3 & 8 Exam level: Both History: New K/A: 295036A202 3.1/3.1 KA Statement: Waterlevelin the affected area System / Evolution: Senad=y Containment High Sump / Area Water Level Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 3 SRO Group: 2
| |
| . 1 4
| |
| | |
| i Question: 097 -
| |
| Given the following conditions:
| |
| Unit 2 is operating at 25% power A feedwater level control malfunction results in a rapidly increasing reactor water level Select the conditions REQUIRING the operator to scram the reactor and close the Main Steam Isolation Valves.
| |
| Reactor waterlevelis:
| |
| : a. at 100 inches and increasing on Refuel Range level instrument LR-97.
| |
| : b. at 95 inches and increasing on Refuel Range level instrument LI-86.
| |
| : c. at 55 inches on Narrow Range level instmments.
| |
| : d. at the top of scale on Wide Range level instruments.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| OT-110 Reactor High Level - Bases, Rev. 6, Step 3.1 & Notes, Pages 2 & 3 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 2, 3 & 4 Exam Level: Both History: New K/A: 295008K101 3.0/3.2 KA Statement: Moisture canyover System / Evolution: High Reactor WaterIevel Exam Section: Emergency And AbnormalPlant Evolutions RO Gmup: 2 SRO Group: 2 Justification: Rapid overfeedmg transients require this as an immediate action. LR-97 is valved out at power
| |
| | |
| a a
| |
| Question: 098 Given the following conditions-Unit 3 is operating at 85% power The Unit Reactor Operator hasjust reported slowly increasing drywell temperatures Investigation has detemuned that the drywell chilled water system is lined up such that the Reactor Building Closed Cooling Water system is supplying cooling water Which of the following has occurred?
| |
| : a. Service water cooling to the chiller condensers has been lost.
| |
| : b. An undervoltage condition has occurred on two of the three busses feedmg the chillers.
| |
| : c. A loss of power to the chilled water header containment isolation valves has occurred.
| |
| : d. Two of the three chillers have tripped on low chilled water flow.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0150, "Drywell Chilled Water System", Rev. 005, Sections IV.C.2 & V.D, Pages 15 & 20, LO - 2.e & 7.e Exam Level: Both History: New K/A: 295012A102 3.8/3.8 KA Statement: Drywell cooling system System / Evolution: High Drywell Temperature Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2 i
| |
| 1
| |
| | |
| i
| |
| \'
| |
| i Question: 099 Which of the following is the basis for requiring a reactor scram with no Control Rod Drive (CRD)
| |
| Pumps running during a reactor startup?
| |
| The reactor is scrammed
| |
| : a. to ensure all controls rods are fully inserted before overheating can affect the mechanism seals and impact on the scram times.
| |
| : b. since no other control rod motion is available (insert / withdrawal) without the CRD system {
| |
| in operation.
| |
| : c. to ensure there is no loss of scram capability as the hydraulic control unit accumulators depressunze
| |
| : d. in anticipation of tripping the Recirculation Pumps due to potential pump seal package damage from theloss ofCRD flow.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| ON-107 Loss Of CRD Regulating Function - Bases, Rev. 6, Sections 2.1 & 2.2, Pages 1&2 LOT-1550, "Off-Normal Procedures", Rev. 006, LC - 2 & 3 Exam Level: Both History: New K/A: 295022K301 3.7/3.9 KA Statement: Reactor SCRAM System / Evolution: Loss OfC'iD Pumps Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Genup: 2 i
| |
| | |
| i e
| |
| s Questim: 100 l
| |
| l Given the following conditions:
| |
| Unit 2 had been operating at 80% power Due to a leak, a high drywell pressure condition occurred All expected automatic actions occurred l l
| |
| Which of the following will be the MAXIMUM expected Reactor Building pressure for these l conditions?
| |
| : a. -2.0 inches WG
| |
| : b. -0.25 inches WG
| |
| : c. 0.00 inches WG
| |
| : d. +1.25 inches WG Answer: b
| |
| | |
| ==References:==
| |
| LOT-0210, " Standby Gas Treatment System", Rev. 009, Section I.A & B, Page 8, LO
| |
| - 1.a & S.a Exam level: Both History: New K/A: 295035K202 3.6/3.8 ,
| |
| I KA Statement: SBGT/FRVS !
| |
| 1 l
| |
| . System / Evolution: Secondary Containment High Differential Pressure i l
| |
| Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 3 SRO Group: 2
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE. ;
| |
| i i
| |
| I APPLICABILITY: MODES 1 and 2. l ACTIONS ,
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. Concentraticn of boron A.1 Verify the 8 hours in solution > 9.82% concentration and weight. temperature of boron &HQ in solution and pump .
| |
| suttion piping Once per temperature are 12 hours within the limits of thereafter i Figure 3.1.7-1.
| |
| , 2
| |
| ~
| |
| &HQ i A.2 Restore concentration 72 hours l of boron in solution-to s 9.82% weight. &HQ :
| |
| 10 days from discovery of failure to meet the LCO i
| |
| i B. One SLC subsystem B.1 Restore SLC subsystem 7 days inoperable for reasons to OPERABLE status.-
| |
| other than AHQ Condition A.
| |
| 10 days from discovery of failitre to meet the LCO (continued) 3.1-20 Amendment No. 210 PBAPS UNIT 2
| |
| | |
| SLC System .
| |
| 3.1.7 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Two SLC subsystems C.1 Restore one SLC 8 hours inoperable for reasons subsystem to OPERABLE other than status.
| |
| Condition A.
| |
| D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met.
| |
| SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
| |
| ( :., 24 hours SR 3.1.7.1 Verify level of sodium pentaborate solution in the SLC tank is k 46%.
| |
| SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours solution is k 53*F.
| |
| SR 3.1.7.3 Verify temperature of pump suction piping 24 hours is k 53*F.
| |
| SR 3.1.7.4 Verify continuity of explosive charge. 31 days icentinued)
| |
| J PBAPS UNIT 2 3.1-21 Amendment No. 210
| |
| | |
| SLC System !
| |
| 3.1.7 l
| |
| ; SURVEILLANCE REQUIREMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify the concentration of boron in 31 days solution is s 9.82% weight and within the limits of Table 3.1.7-1. A!(Q !
| |
| Once within 24 hours after !
| |
| water or boron is added to solution AtiD ,
| |
| Once within 24 hours after solution temperature is restored within limits
| |
| ? i SR 3.1.7.6 Verify each SLC subsystem manual and power 31 days i operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can ;
| |
| be aligned to the correct position. l SR 3.1.7.7 Verify the quantity of C-10 stored in the 31 days SLC tank is a 162.7 lbs.
| |
| SR 3.1.7.8 Verify each pump develops a flow rate In accordance !
| |
| = 43.0 gpm at a discharge pressure with the I a 1255 psig. Inservice Testing Program L
| |
| (continued) l 3.1-22 Amendment No. 210 PBAPS UNIT 2
| |
| | |
| SLC System .
| |
| 3.1.7 SURVEILLANCE RE001REMENTS (continued)
| |
| SURVEILLANCE FREQUENCY SR 3.1.7.9 Verify flow through one SLC subsystem from 24 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.10 Verify sodium pentaborate atos percent B-10 Once within 8 ]
| |
| enrichment is within the limits of hours after 3 Table 3.1.7-1. addition to {
| |
| SLC tank i
| |
| l l
| |
| )
| |
| PBAPS UNIT 2 3.1-23 Amendment No. 210
| |
| | |
| SLC System 3.1.7 Table 3.1.7-1 (page 1 of 1)
| |
| Standby Liquid Control System Boron Concentration, Pump Flow Rate, and Boron Enrichment Limits The combination of SLC System boron concentration, pump flow rate, and boron enrichment shall be in accordance with the following equation:
| |
| C Q E x x = 1 13% weight 86 gpm 19.8% atom where, ;
| |
| i C = % weight sodium pentaborate solution concentration, l l
| |
| Q = Pump flow rate (gpm) at a discharge pressure of a 1255 psig, and 1
| |
| E = Boron-10 enrichment (% atom Boron-10). l j
| |
| ?. . . 1 I
| |
| PBAPS UNIT 2 2.~2-24 Amendment No. 210
| |
| | |
| 51.C Systtm -
| |
| 3.1.7
| |
| " U'O (24 %,11 100.0
| |
| /f !
| |
| /4 I f
| |
| ELD ,
| |
| ~/
| |
| ACCEPTABLE g l ,
| |
| E O 80.0
| |
| /
| |
| m A
| |
| _ X
| |
| /l f [ NOT ACCEPTABb
| |
| ~ / I L 1
| |
| /
| |
| J Vl fl (o.a2% 51.s*F) 9D 10.0 11D 12.0 13D 14D 15D 1SD 17D 18D iSD 210 21D 22.0 23.0 24 0 CONCENTRATION (Weight Percent Pentaborate Soluuon)
| |
| Figure 3.1.7-1 (page 1 of 1)
| |
| Sodium Pentaborate Solution Temperature Versus Concentration Requirements PBAPS UNIT 2 3.1-25 Amendment No. 210
| |
| | |
| )
| |
| Question: 041 With Unit 2 at 100% power, Panel E-124-R-C a lost.
| |
| How does this power loss affect operation of the Standby Liquid Control (SLC) System?
| |
| : a. No system indicating lights wd be illuminated and both pumps and both squib valves are ;
| |
| not available. l
| |
| : b. The "A" SLC Pump and "A" Squib Valve will not be available.
| |
| : c. Following a SLC system initiation, the operator will not be able to venfy pump discharge pressure or SLC storage tank level in the Control Room or locally. l
| |
| : d. The "B" SLC Pump and "B" Squib Valve will not be avadable.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0310, " Standby Liquid Control System", Rev. 011, Section V.3.c, Page 17a, LO
| |
| - 3.a & b Exam Level: Bodi History: New g',
| |
| K/A: 211000K202 3.1/3.2 KA Statement: Explosive valves System / Evolution: Standby Liquid Control System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 i
| |
| e 'm e
| |
| L
| |
| | |
| - Question: - 042 Given the following conditions:
| |
| -- Unit 2 isin Mode 4
| |
| -- The AC input power to the 24 VDC battery charger supplying Panel 2AD45 has been lost
| |
| - This panel supplies power to Source Range Monitoring (SRM) channels "A" and "C" W
| |
| . hich of the following describes the long-term effect on these SRM channels?
| |
| : a. The SRMs will slowly deplete the battery and, over time, channels "A" and "C" will drift "downscale"
| |
| : b. The SRMs will continue to operate due to automatically swapping to the second battery charger.
| |
| : c. The SRMs will slowly deplete the battery and, over time, channels "A" and "C" will drift
| |
| " upscale"
| |
| : d. The SRMs will NOT operate until an equipment operator tr.r.f=5 power via the Manual Bus Transfer switch.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0240, " Source Range Monitoring System", Rev. 008, Section III.G, Pages 10 &
| |
| l l, LO - 7.b Exam Level: Both History: New K/A: 215004K602 3.1/3.3 KA Statement: 24/48 Volt D.C. power System / Evolution: Source Range Monitor System Exam Section: Plant Systems RO Gmup: 1 SRO Gmap: 1 E.
| |
| | |
| Questirn: 043 i
| |
| Given the following conditions:-
| |
| ~ ;W
| |
| -- A Unit 2 reactor stanup is in progress with control rod withdrawals occurring
| |
| -- Rod Worth Minimtzer (RWM) Group I contains 12 control rods that are to be withdrawn from Notch"00" to Notch"48"
| |
| - 1 I rods from this group are withdrawn to Notch "48" and the remaimng rod to Notch "42"
| |
| - A control rod in Group 2 is then selected but not withdrawn Which of the following is the expec'ed response of the RWM7 The RWM will display;
| |
| : a. one withdraw error and further rod withdrawals will be blocked except for the rod with the withdraw error,
| |
| : b. one withdraw error and if a second withdraw error is made further rod withdrawals will be blocked except for the two rods with the withdraw errors.
| |
| : c. one insert error and funher rod withdrawals will be blocked except for the rod with the -
| |
| l insert error, t
| |
| ' d. one insert error and if a second insert error is made, further rod withdrawals will be l blocked except for the two rods with the insert errors.
| |
| Answer: d
| |
| | |
| ==References:==
| |
| LOT-0090, " Rod Wmth Minimi=", Rev. 010, Sections III.A 7-9 & IV.B, Pages 11
| |
| & 22, LO - 5.a & 6.j & I l Eaam Level: Both History: New K/A: 201006K513 3.5/3.5 KA Statement: Insert block System / Evolution: Rod Worth Mirumizer System RO Group: 2 SRO Group: 2 L Eaam Section: Plant Systems
| |
| | |
| Questi:n: 044
| |
| . ?
| |
| , Given the following conditions:
| |
| - The Transfer / Isolation Switch for the 71K Safety Relief Valve (SRV) in the E22 Bus Room has been placed in" Emergency"
| |
| - The red indicating light for that SRV on the HPCI Alternative Shutdown Panel is illuminated What is the status ofSRV 71K?
| |
| : a. The SRV tailpipe acoustic monitor is picking up flow noises
| |
| : b. The SRV tailpipe temperature is high.
| |
| : c. The SRV solenoid is energized
| |
| : d. The SRV has opened on an overpressure condition.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0120, " Main Steam And Pressure Relief System", Rev. 014, Section V.B.7.b, Page 25, LO - 1.b & 5.i
| |
| (-
| |
| Exam level: SRO History: New K/A: 239002K405 3.6/3.7 KA Statement: Allows for SRV operation from more than one location System / Evolution.: Relief 7 Safety Valves Exam Section: Plant Systems RO Group: 1 SRO Gesup: 1 L
| |
| 1 I
| |
| | |
| Question: 045 If the Main Steam Isolation Valves (MSIV) have one DC and one AC solenoid on each valve, how does a loss of both Reactor Protection System (RPS) buses result in a closure of all eight MSIVs?
| |
| : a. A loss of both RPS buses results in a trip of both primary contamment isolation system logic channels and subsequent MSIV closure.
| |
| : b. The RPS buses supply 20Y33 and 20Y34 which supply the AC and DC (via an inverter)
| |
| MSIV solenoids respectively such that a loss of both buses closes the MSIVs.
| |
| : c. A loss of both RPS buses results in a loss of the MSIV position indications giving a reactor scram and MSIV closure.
| |
| : d. The RPS buses supply 20Y33 and 20Y34 which supply the power to the MSIV test solenoids such that a loss of both buses deenergizes them and closes the MSIVs.
| |
| l Answer: a i
| |
| )
| |
| | |
| ==References:==
| |
| LOT-0180, "Pnmary Containment Isolation System", Rev. 009, Section III.G.1, Pages 26 & 27, LO - 3 & 6.g Exam level: Both History: New K/A: 223002K309 3.4/3.6 KA Statement: Main steam system System / Evolution: Primary Contamment Isolation Syste n/ Nuclear Steam Supply Shut-Off j Exam Section: Plant System RO Group: 1 SRO Group: 1 l
| |
| l l
| |
| L
| |
| | |
| Question: 046 Following a valid initiation signal on Unit 2, the Standby Gas Treatrnent (SBGT) train Bypass Damper (PO-00522) does NOT reposition as designed.
| |
| Which of the following describes how this failure will affect SBGT operation?
| |
| : a. The " Emergency Gas Filter Valve Failure" alarm will be received and the operator will be required to isolate one filter train in order to mamtain the required negative pressure in secondary containment
| |
| : b. Once the second filter train is shutdown and isolated, the remaining train may not be sufficient to mamtain the required negative Way containment pressure.
| |
| : c. The " Emergency Gas Filter Valve Failure" alarm will be received and the operator will be required to start an additional SBGT fan in order to maintain the required negative pressure in secondary contamment.
| |
| : d. Once the second filter train is shutdown and L*~i, process flow through the imeirdrig train and fan may not be sufficient for charcoal adsorber bed heat removal.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| LOT-0210, " Standby Gas Treatment System", Rev. 009, Section II.J, Pages 11 & 12, i LO - 1.b & 6.b Exam 12 vel: Both History: New K/A: 261000K301 5.3/3.6 KA Statement: Secondary tantainment and environment differential pressure System / Evolution: Standby Gas Treatment System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 L.
| |
| | |
| Question: 047 Given the following conditions:-
| |
| - The Unit 3 High Pressure Coolant Injection (HPCI) system is runrung in the full flow test mode (CST to CST)
| |
| - System flow rate is 5000 gpm with the flow controller in " Automatic" -
| |
| - The HPCI ramp generator output just failed to its " low" limit Select the expected HPCI system response and the reason for that response.
| |
| The HPCIturbine:
| |
| : a. begins to slow down since the failed ramp generator transfers the flow controller from
| |
| " automatic" to " manual" and slowly drives it to "0" output
| |
| : b. remains at its original speed since the flow controller output ovenides the ramp generator output when the controller is in " automatic".
| |
| : c. begins to slow down since the turbine control valve is pa=tiaani based upon the lowest of the two signals it receives, flow controller output versus ramp generator output.
| |
| . d. remains at its original speed since the ramp generator output signal is disabled when the flow controller is controlling and remains out of the drcut until the turbme stop valve closure resetsit to" idle".
| |
| Answer: c
| |
| | |
| ==References:==
| |
| LOT-0340, "High Pressure Coolant Injection", Rev. 009, Section III, C, Pages 12 &
| |
| 13, LO. - 6.a & d Exam Level: SRO History: New K/A: '206000K505 3.3/3.3 .
| |
| KA Statement:- Turbine speed control System / Evolution: High Pressure Coolant Injection System Exam Section: Plant Systems RO Group: 1 SRO Group: 1 _;
| |
| | |
| Questi:n: 048 With Unit 2 at 75% power, the Main Control Room has received the " Blowdown Relief Valves Bellows Leaking" alarm. The Unit Reactor Operator reports that the "G" SRV amber light is lit.
| |
| Which of the following describes how this impacts the operation of the affected Safety Relief Valve (SRV)? l The SRV will: )
| |
| : a. operate in the Automatic Depressunzation mode but NOT in the Safety mode.
| |
| : b. operate in the Safety mode but NOT in the Automatic Depressunzation mode.
| |
| : c. operate only in the manual mode (i.e., switches on the C03 Panel).
| |
| : d. not operate in any mode.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0330, " Automatic Depressunzation System", Rev. 008, Section IV.A.2.a.7, Page 14, LO - 2.f& 7.a
| |
| '' Eum Level: Both History: New K/A: 218000K106 3.9/3.9 KA Statement: Safety /ReliefValve System / Evolution: Automatic Depressunzation System Exam Section: Plant System RO Group: 1 SRO Group: 1 1
| |
| \
| |
| | |
| )
| |
| Question: 049 Given the following conditions:
| |
| - Unit 2 experienced a loss of coolant accident (LOCA) 10 minutes ago
| |
| - All 4 Residual Heat Removal (RHR) pumps are mnning and injecting into the vessel ,
| |
| Reactor waterlevel: -100 inches and rising !
| |
| -- Reactor pressure: 100 psig and steady
| |
| - Drywell pressure: 9.5 psig and rising Which of the following logic interlocks must be met to initiate drywell spray using the "A" RHR Imop (i.e., open MO-31 A and 26A)?
| |
| : a. Aleast one HPSW Ptunp must be running
| |
| : b. The Lack of LOCA signal must be bypassed
| |
| : c. The Containment Spray Valve Control Switch (S17) must be momentanly placed in
| |
| " Mand"
| |
| : d. The Containment Spray Override 2/3 Core Coverage Switch (S18) must be placed in
| |
| " Manual Override" .
| |
| . Answer: c ;
| |
| i 1 i
| |
| | |
| ==References:==
| |
| LOT-0370, " Residual Heat Removal System", Rev. 010, Section IV.C.3.b, Page 24, !
| |
| LO - 1.c & 5.q ;
| |
| i Exam Level: Both History: New K/A: 226001A105 3.1/3.4 l
| |
| KA Statement: Systemlineup System / Evolution: RHR/LPCI: Containment Spray System Mode Exam Section: Plant Systems RO Group: 2 SRO Group: 1 M
| |
| 4
| |
| | |
| I l
| |
| Questi:n: 050 l Given the following conditions:
| |
| - Both Units are operating at 100% power All startup power souren tre energized and are available No bus breakers are blocked
| |
| - All four diesel generators (DG) are in their normal standby lineups I
| |
| - An operator in the Control Room opens the E-212 breaker with its handswitch
| |
| - No other operator actions are taken.
| |
| Select the expected 4 KV system response.
| |
| The E-12 bus:
| |
| : a. will not be energized from 3 SUE but the E-1 Diesel s' arts and energizes the bus.
| |
| : b. will not be energized from 3 SUE and the E-1 Diesel does NOT start.
| |
| : c. will be energized from 3 SUE and the E-1 Diesel starts but the output bmiker does NOT close
| |
| : d. will be energized from 3 SUE and the E-1 Diesel does NOT start.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| - LOT-0660, "4 KV Distribution", Rev. 010, Section V.C.1, Pages 23 & 24, LO - 4.e &
| |
| 7 Exam level: Both History: New K/A: 262001A401 3.4/3.7 KA Statement: Allbreakers and disconnects System / Evolution: A.C. Electrical Distnhation Exam Section: Plant Systans RO Group: 2 SRO Group: 1 L
| |
| | |
| n 1
| |
| Question: 051 Given the following conditions:
| |
| Unit 2 was operating at 75% power -
| |
| A break in the "A" Recirculation loop has occurred i The "B" loop of Residual Heat Removal (RHR) is not available Reactor levelis-120" and going up <
| |
| - Reactor pressure is 350 psig and dropping Drywell pressureis 10 psig going up l All other automatic actions have occurred as designed Select the reason why maximum torus cooling cannot be IMMEDIATELY started.
| |
| : a. High Pressure Service Water system capacity will not supi.oit the additional heat load of torus cooling for ten (10) minutes. J
| |
| : b. The RHR pumps cannot be stopped for a transfer to torus coobg for five (5) minutes following aninitiation signal.
| |
| : c. The Transient Response Implementation Plan procedures require maximum injection flow for the first ten (10) minutes following a LOCA.
| |
| , d. The valve repositionings required for torus cooling cannot be completed for five (5) i ,i minutes following a LOCA.
| |
| l Answer: d !
| |
| Refesences: LOT-0370," Residual Heat Removal System", Rev. 010, Handout 0370-8, Pages 1 &
| |
| 2, LO - 1.c & 5.a ,
| |
| 1 Exam 12 vel: Both History: New K/A: :19000A214 4.1/4.3 KA Statement: Loss ofcoolant accident Synem/ Evolution: Torus / Suppression PoolCoohng Mode Exam Section: Plant Systems RO Group: 2 SRO Group: 2 l
| |
| | |
| Question: 052 .
| |
| With the Unit 2 Refueling Platform unloaded and near the core, which of the following ALONE will prevent further platform movement toward the core?
| |
| : a. The Reactor Mode Switch is in "Startup/ Hot Standby"
| |
| : b. The Main Grapple"Not. Full Up".
| |
| : c. The Reactor Mode Switch is in " Refuel"
| |
| : d. One control rod is withdrawn to Notch"08" Answer: a
| |
| | |
| ==References:==
| |
| FH-0762, " Refueling Bridge And Operations", Rev. 00a, Section E.4.b.(2), Page 42, LO-14 LOT-0080, " Reactor Manual Control Systein", Rev. 007, LO - 2.b & 4.a Exam Level: Both History: New K/A: 234000K401 3.3/4.1
| |
| / ' 'i KA Statement: Prevention of core alterations during control rod movements System / Evolution: Fuel F=dling Equipment Exam Section: Plant Systems RO Group: 3 SRO Group: 2 l
| |
| | |
| Questi2n: 053
| |
| [ Which of the following is the purpose of the refueling interlocks?
| |
| The refuelinginterlocks will:
| |
| l
| |
| ~
| |
| : a. ensure a correctly loaded reactor core during refueling.
| |
| : b. prevent the simultanmus chaapag of reactivity by two or more means
| |
| : c. ensure a minimum water level is mair*=iaad above the fuel assemblies at all times.
| |
| I d. prevent exceedmg predetermmed reactivity rate of change limits during refueling.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| FH-0762, " Refueling Bridge And Operations", Rev. 00a, Section E.1, Page 39, LO-14 i
| |
| Exam Level: SRO History: New ,
| |
| l K/A: 234000A201 3.3/3.7 l
| |
| ; KA Statement: Intedock failure l ( \
| |
| System / Evolution: FuelHandling Equipment Exam Section: Plant Systems RO Grsup: 3 SRO Group: 2 l
| |
| l -
| |
| l c 1
| |
| l l
| |
| \
| |
| J e
| |
| I
| |
| | |
| Question: 054
| |
| .3, Given the following conditions:
| |
| - Unit 3 is operating at 70% power
| |
| -- Main Steam Isolation (MSIV) stroke testing is in progress
| |
| - The Outboard MSIV on the "D" Main Steam Line test pushbutton has been depressed
| |
| - When the Operator depresses the test pushbutton it sticks in the " depressed" position
| |
| - The Operator does not notice and no other actions are taken Which of the following is the expected final position of this MSIV?
| |
| : a. 10% closed position
| |
| : b. 50% closed position
| |
| : c. 90% closed position
| |
| : d. 100% closed position Answer: d
| |
| | |
| ==References:==
| |
| LOT 4120," Main Steam And Pressure Relief System", Rev 014, Section IV.B.2 &
| |
| Pi .
| |
| Figures 5 & 6, Pages 22 & 23 Exam level: Both History: New K/A: 239001A401 4.2/4.0 KA Statement: MSIV's System / Evolution: Main And Reheat Steam System Exam Section: Plant Systems RO Group: 2 SRO Group: 3
| |
| .r i 4
| |
| 1
| |
| | |
| Question: 055 Which of the following is the reason for the Jet Compressor Off-Gas Inlet Valve (MO-2991) automatic closure on LOW steam flow of <7500 lbm/ hour?
| |
| This low steam Bow isolation prevents:
| |
| : a. ' hydrogen concentrations in excess of the 4% limit downstream of the recombiner.
| |
| : b. moisture buildup in the off-gas charcoal adsorber beds.
| |
| : c. high off-gas flow resulting in channeling ofthe charcoal adsorber bed.
| |
| : d. stagnant off-gas flowrates in the system, @cally withm the holdup pipe.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| LOT-0510,"Off-Gas Recombiner System", Rev. 011, Sections III.D.5 & IV.B.4, Pages 10 & 21, LO - 1.d and 4.a & k Exam level: Both History: New ,
| |
| K/A: 271000K408 3.1/3.3 1
| |
| 3 KA Statement: Automatic systemisolation
| |
| - System / Evolution: Offgas System Exam Section: Plant Systems RO Group: 2 SRO Group: 2 O
| |
| | |
| Questi:n: 056 One of the stated functions of the Traversing In-Core Probe (TIP) system is to calibrate the LPRM instrumentation while operating at power.
| |
| Which of the following describes how the TIP instrumentation itselfis calibrated?
| |
| : a. The detector design is such that routine calibration is not required, however, they are physically checked and compared with a reference value each refueling outage.
| |
| : b. A comparison is made betwi the detector electronic output and a " standard" value based on the detector age, the number of hours it has been inserted in the core and core power whileinserted.
| |
| : c. A comparison ofeach of three TIP detectors is made by insertmg thern in tum into the same LPRM string position in the core and then normalMng to 100% power.
| |
| : d. Since the detectors are not routinely subjected to neutron flux, the independently verified factory calibration settings are considered valid between rnlacement cycles, typically every other refueling outage.
| |
| Answer: c i ,
| |
| | |
| ==References:==
| |
| LOT-0290, " Traversing In-Core Probe", Rev. 008, Sections II.B & IV.F, Pages 5-6 &
| |
| 14, LO - 1.b & 7 Exam Imel: Both History: New K/A: 215001G128 3.2/3.3 KA Statement: Knowledge of the purpose and function ofmajor system components and controls System / Evolution: TraversingIn-Core Probe Exam Section: Plant Systems RO Group: 3 SRO Group: 3 L.
| |
| | |
| J Question: 057 l Given the following conditions:
| |
| -- A reactor startup is in progress on Unit 3 Reactor pressureis 900 psig
| |
| - Scram time testingisin progress
| |
| - The scram INLET valve FAILS TO OPEN during the test Which of the following is the expected response of this control rod?
| |
| : a. The rod will fully insert on acnimiihtor pressure. .
| |
| : b. The rod will fully insert on reactor pressure.
| |
| : c. The rod willpartiallyinsert
| |
| : d. He rod willNOTinsert Answer: b
| |
| | |
| ==References:==
| |
| LOT-0060," Control Rod Drive Mdiuilsm", Rev. 009, Section III.B.4 & Figure 7, Pages 21 & 22 i, , -
| |
| Exam level: Both History: New l
| |
| K/A: 201003K601 3.3/3.3 KA Statement: Controlrod drive hydraulic system System / Evolution: ControlRod AndDriveMechanism Exam Section: Plant Systems RO Group: 2 SRO Group: 3 L
| |
| | |
| Questi:n: 058 Which of the following describes the final status of the Unit 2 feedwater system and Reactor Feedwater Pumps (RFP) following a normal reactor scram? Assume that all expected Unit Reactor Operator actions were taken.
| |
| : a. "A" and "B" RFPs emergency stopped, "C" RFP runnmg on the start-up level controller
| |
| : b. "A" and "B" RFPs tripp' ed, "C" RFP running on the start-up level controller
| |
| : c. "B" and "C" RFPs emergene; ; , ,wi, "A" RFP running on the rr.inimum flow valve
| |
| : d. "B" and "C" RFPs tripped. "A" RFP running on the mimmum flow valve Answer: a
| |
| | |
| ==References:==
| |
| Exhibit OM-P-16.1:5,"OSPS: Reactor Operator Response To Reactor Scram", Rev.
| |
| 4, Section 5, Page 2 Exam Level: Both History: New K/A: 295006A102 3.9/3.8
| |
| ( i KA Statement: Reactor waterlevel control system System / Evolution: Scram Exam Section: Emergeacy And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 L.
| |
| | |
| 1 Question: 059 Table DW/T-1, "RPV Level Instrument Status" states that a reactor water level instrument may not be used if drywell temperature is at or above the RPV Saturation Curve.
| |
| Which of the following is the reason for this requirement? .
| |
| The instrument's:
| |
| : a. reference leg is == mad to have flashed, caumng level to read falsely high.
| |
| : b. variable leg is assumed to have flashed, causmg level to read falsely low.
| |
| : c. reference and variable legs are assumed to have flashed, causing les o read falselylow.
| |
| : d. reference and vanable legs are == mad to be undergoing " mag ==Ag", causing level to read falsely high.
| |
| Answer: a
| |
| | |
| ==References:==
| |
| T-102 Prmwy Contamment Control- Ba es, Rev.12, Step DW/f-4, Pages 18 & 19
| |
| ( LOT-0050, " Reactor Vessel Instrumentation", Rev. 013, Seebon V.A.2, Pages 31 &
| |
| 32, LO - 6.b.1 & 6.o Exam Level: Both History: New K/A: 295028K101 3.5/3.7 KA Statement: Reactor waterlevelmeasurement System / Evolution: High DrywellTemperature Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 2 SRO Group: 2 e
| |
| d
| |
| .- e.
| |
| l
| |
| | |
| Q:estio::: ' 060 Given the following conditions for Unit 2:
| |
| - A small LOCA has occurred resulting in a reactor scram on high drywell pressure l
| |
| - Afailure-to-scram (ATWS) occurred,130controlrodsdidnotinsert
| |
| - Reactor water level was lowered to and maintained between -100 and -120 inches with feedwater
| |
| - Reactor pressure is currently 950 psig
| |
| - Standby Liquic' Control (S14) Tank level has decreased by 36% and SLC is still injecting Which of the following could cause a reactor recriticality for these conditions?
| |
| : a. Placing RCIC in service to maintain reactor water level
| |
| : b. Failure toinhibit ADS
| |
| : c. Continued presmire drop due to the LOCA
| |
| : d. DK,resing Xenon concentration from decay over the first two hours aAer the failure to scram Answer: c.
| |
| i
| |
| | |
| ==References:==
| |
| Trip Curves, Tables and limits - Bases, Rev. 2, Section 2, Page 2 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 6 Exam Imel: Both History: New K/A: 295037K104 3.4/3.6 KA Statement: Hot shutdown boron weight System / Evolution: SCRAM Condition Prescrit And Reactor Power Above APRM Downscale Or Unknown Exam Section: Emergency And AbnormalPlantEvolutions RO Group: 1 SRO Group: 1 a
| |
| | |
| 1 Questi::n: 061 Per ON-124, " Fuel Floor And Fuel Handling Problems", which of the following requires the operators
| |
| -- to take the actions for " criticality" during fuel handling operations? Assume the fue; movement is inside the reactor vessel.
| |
| : a. An unexpected Refuel Floor Area high radiation alarm is received.
| |
| : b. Refuel Bridge reverse motion (towards the core) interlock activates.
| |
| : c. Source Range nuclear instrumentation counts are urspectedly increasing.
| |
| : d. Source Range nuclear instrumentation counts are spiking repeatedly.
| |
| Answer: c
| |
| | |
| ==References:==
| |
| ON-124," Fuel Floor And Fuel Handhng Problems", Rev.1, Symptoms Section 1, Page1 <
| |
| l I
| |
| LOT-1550, "Off-Normal Procedures", Rev. 006, LO - 1 Exam Level: Both History: New K/A: 295023A106 3.3/3.4 !
| |
| \
| |
| KA Statement: Neutron monitoring l System / Evolution: Refueling Accidents Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 3 SRO Group: 1 l L
| |
| | |
| Questi3n: 062
| |
| . . , Following a transient on Unit 2, the following conditions exist:
| |
| - Drywell pressure: 4.5 psig and rising
| |
| - Drywell temperature: 140 *F and rising Torus pressure:- 8.4 psig and nsmg
| |
| - Torus water temperature: 82 *F and stable Which of the fo'Jowing has occurred?
| |
| : a. A safety reliefvalve has opened and rem
| |
| : b. A pipe break into the drywell has occurred with a torus to drywell vacuum breaker open.
| |
| : c. A safety relief valve tail pipe has broken above the toms water level while the valve is open.
| |
| : d. A recirculation line break has occurred with all containment parameters responding as designed.
| |
| Answer: c
| |
| ()
| |
| | |
| ==References:==
| |
| LOT-0130,"Pnmany Containment", Rev. 010, Section 5.D and Figure Sc, Page 26, LO - 5.a & 7.c Exam level: Both History: New K/A: 295024A206 4.1/4.1 KA Statement: Suppression poolternperature System / Evolution: High DrywellPressure Exam Section: Eirepry And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 I
| |
| | |
| 1 a
| |
| Questi:n: 063 t.,3u/ With all control rods fully inserted, T-111, " Level Restoration", in the steam cooling section, directs an
| |
| ~
| |
| ' Emergency Blowdown when reactor water level reaches -210 inches with no injection systems available.
| |
| What is the bases for this action?
| |
| : a. Core fuel temperatures will decrease allowing additional time for restoration of some source ofinjection.
| |
| : b. Reactor pressure will be reduced causmg core voiding to increase addmg negative reactmty resultmgin a power drop.
| |
| : c. Reactor core differential pressure will decrease assisting the thermal dnving head for natural circulation flow.
| |
| : d. The total amount of energy that could be released when fuel melt begms will be reduced while some waterinventory remains Answer; a
| |
| | |
| ==References:==
| |
| T-111 Level Restoration - Bases, Rev. 8, Step LR-20, Page 7 I (y
| |
| LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Exam I.evel: Both Histor';: New 1
| |
| K/A: 295031K305 4.2/4.3 KA Statement: Emergency depressunzation System / Evolution: Reactor Low Water Level Enam Section: Emergency And AbnormalPlantEvolutions RO Group: 1 SRO Group: 1
| |
| | |
| i Question: 064 ,
| |
| i
| |
| _/ Given the following conditions
| |
| - Unit 2 has had a failure-to-scram (ATWS) on high drywell pressure 3
| |
| - The Automatic Depressurization System (ADS) initiation was not inhibited Which of the following will result when reactor water level reacles the ADS initiation setpoint?
| |
| a- A rapid, uncontrolled reactor depressurization, following a 9 minute time delay, resulting in' a rapid power drop
| |
| : b. An increase in the injection rate of cold, unborated water resulting in a rapid power rise.
| |
| : c. Loss of boron due to entrainment in the steam from the reactor to the torus, increanteum time required io shutdown the reactor.
| |
| : d. A large reduction in effective control rod worth because ofvoiding, leadmg to a power excursion.
| |
| Answer: b
| |
| | |
| ==References:==
| |
| T-101 RPV Control- Bases, Rev.19, Step RC/Q-11, Page 6 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 3 Exam Level: Both History: New K/A: 295037K209 4.0/4.2 KA Statement: Reactor waterlevel System / Evolution: SCRAM Condition Piment And Reactor Power Above APRM Downscale Or Unknown Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 1 SRO Group: 1 L
| |
| | |
| 1 0
| |
| Question: 065
| |
| .c Unit 2 has a high drywell pressure requiring primay containment venting in accordance with T-102,
| |
| " Primary Contamment Control" Per T-200,"Pnmary Contamment Venting", the preferred vent path !
| |
| is via the Torus 2" vents to the Standby Gas Treatment System.
| |
| l Which of the following is the reason for venting the primary containment via the toms instead of the drywellifpossible?
| |
| Venting via the torus will:
| |
| : a. allow better control of the release rate due to valve and piping sizing considerations.
| |
| 1
| |
| : b. reduce the overall plant radioactivity release as it passes through the torus water I I
| |
| : c. give the operator the ability for a more rapid pressure reduction. 1
| |
| : d. reduce the hydrogen concentration to prevent a fire hazard.
| |
| I l
| |
| Answer: b J 1
| |
| | |
| ==References:==
| |
| T-200-2, " Primary Contamment Venting'', Rev. 5, Table 4.17, Page 7 BWROG Emergency Plant Guidelines, Rev. 4 LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 4 ,
| |
| Exam Level: SRO History: New j l
| |
| K/A: 295010K301 3.8/4.0 l l
| |
| 1 KA Statement: Drywellventing System / Evolution: High DrywellPressure Exam Section: Emergency And Abnormal Plant Evolutions RO Group: 1 SRO Group: 1 L
| |
| O
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| Questi:n: 066 Unit 3 is perfonning T-118, " Primary Contahment Floodmg" Which of the following is the reason why the reactor vessel is vented periodically during the primary containment flooding?
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| Reactor vessel venting ensures:
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| : a. the primary containment will be vented during the floodmg process.
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| : b. actual reactor water level will have reached the main steam lines when containment floodingis complete.
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| : c. the level instrumentation reference legs will be free ofnon-condensable gasses.
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| : d. adequate core cooling via core submergence will be achieved dunng conta' mment iloodmg.
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| Answer: d
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| ==References:==
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| T-118 Primary Contamment Floodmg - Bases, Rev. O, Step PCF-9
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| (: 'g LOT-1560, "PBS TRIP Procedures", Rev. 007, LO - 1 & 3 Exam Level: SRO History: New K/A: 295031K101 4.6/4.7 KA Statement: Adequate core cooling System / Evolution: Reactor Low WaterI.evel Exam Section: Emergency And AbnormalPlant Evohtions RO Group: 1 SRO Group: 1 L
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| l
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| .-____--_______________-------______-_a
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| D Question: 067 During a plant startup, what is the earliest point at which entry into OT-112 " Recirculation Pump Trip" is required if a pump trip occurred?
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| : a. The Reactor Mode Switch has been placed in "Startup/ Hot Standby".
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| : b. The reactor is at or above criticality.
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| : c. The Reactor Mode Switch has been placed in "Run".
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| : d. Reactor coolant temperature is > 212 'F.
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| Answer: a
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| ==References:==
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| OT-112, "Rectreulation Pump Trip", Rev. 22, Section 3.4, Page 5 LOT-1540, " Operational Transient Procedures", Rev. 005, LO - 1 & 3 Exam Level: Both History: New K/A: 29500lG404 4.0/4.3 .
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| KA Statement: Ability to recognize abnormal i**4ns for system operating parameters which (3 are entry level conditions for emergency and abnormal operating procedures System / Evolution: Partial Or Complete Loss OfForced Core Flow Circulation Exam Section: Emergency And AbnormalPlant Evolutions RO Group: 2 SRO Group: 2 e
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| M
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| Question: 068 i Given the following conditions: i q.g,; - - - ~ .
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| Unit 2 has eper.ced a loss of off-site power and a Loss Of Coolant Accident
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| - It has been determined that the nmning diesel generators (DG) do NOT have cooling water available Cooling water will be available in approximately 8 minutes The Control Room Supervisor has directed the mnning DGs to be shutdown before the 3 minute limit with no coohng is reached Which of the following describes how these DGs are required to be shutdown and maintained shutdown with the initiating signal still present?
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| : a. The local operator is directed to depress the Emergency Stop knob located behind'the Diesel Gauge Panel.
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| : b. The Plant Reactor Operator installs ajumper that directly inseits a diesel generator differential overcurrent trip agnal
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| : c. The Plant Reactor Operator places the Dent Control Switch in " Pull-To-Lock" in the Main ControlRoom.
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| ,. . d. The local operator is directed to depress the Stop pushbutton on the Diesel Gauge Panel.
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| t.
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| Answer: b
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| ==References:==
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| SE-11 Loss OfOff-Site Power . Bases, Rev. 7, Steps LP-10 & 11, Pages 6 & 7 .
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| LOT-1555, "Special Events", Rev. 004, LO - 12.b Exam level: Both History: New K/A: 295003A102 4.2/4.3 KA Statement: Emergency generators System / Evolution: Partial Or Complete Loss Of A.C. Power Exam Section: Emergency And AbnormalPlantEvolutions RO Group: 2 SRO Group: 1
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| -L
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| _}}
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