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| {{Adams
| | #REDIRECT [[IR 05000354/1998003]] |
| | number = ML20217M049
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| | issue date = 04/28/1998
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| | title = NRC Operator Licensing Exam Rept 50-354/98-03OL,(including Completed & Graded Tests) for Tests Administered on 980223-0304
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| | author name =
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| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
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| | addressee name =
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| | addressee affiliation =
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| | docket = 05000354
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| | license number =
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| | contact person =
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| | document report number = 50-354-98-03OL, 50-354-98-3OL, NUDOCS 9805040395
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| | package number = ML20217L976
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| | document type = EXAMINATION REPORT, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
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| | page count = 137
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| }}
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| See also: [[see also::IR 05000354/1998003]]
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| | |
| =Text=
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| {{#Wiki_filter:*
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| j U.S. NUCLEAR REGULATORY COMMISSION
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| REGION I
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| Docket No: 50-354
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| License Nos: NPF-57
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| Report No. 50-354/98-03(OL)
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| Licensee: Public Service Electric and Gas Company
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| Facility: Hope Creek Generating Station
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| l Location: P.O. Box 236
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| Hancocks Bridge, New Jersey 08038
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| Examination Period: February 23,1998 - March 4,1998 (onsite)
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| March 4 - March 12,1998 (inoffice)
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| Chief Examiner: D. Florek, Senior Operations Engineer
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| Examiners: J. Caruso, Operations Engineer
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| T. Fish, Operations Engineer
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| Approved by: R. Conte, Chief, Operator Licensing
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| and Human Performance Branch
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| Division of Reactor Safety
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| 9805040395 990428
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| PDR ADOCK 05000354 l
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| V PDR
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| EXECUTIVE SUMMARY -
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| Examination Report 50-354/98-03(OL)
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| Initial exams were administered to six senior reactor operator (SRO) instant applicants and
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| five reactor operator (RO) applicants during the period of February 23 - March 2,1998, at
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| the Hope Creek Generating Station.
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| OPERATIONS
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| PSE&G staff submit initially an inadequate examination to administer to applicants for an .
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| operator's license. A good majority of the test items of each portion of the examination
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| required replacement or significant modifications. Significant interactions between the
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| NRC and PSE&G and an exam postponement for two weeks were required to develop an
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| exam that was consistent with the NRC Examiner Standards.-
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| Also, there was insufficient controls, criteria, or data recorded in the controlling documents
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| as evidence that the required control manipulations were significant and were properly
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| credited. Because of this, not all of the applicants performed five significant control
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| manipulations which had to be redone. This area is unresolved item pending further
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| enforcement review by NRC staff with respect to meeting 10 CFR 55.31(a)(5)(50-354/98-
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| 03-01).
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| 1
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| ii
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| c
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| .,--
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| Report Details
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| 05 Operator Training and Qualifications
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| 05.1 Operator Initial Exams
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| a. Scope
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| The NRC examiners administered initial exams to five RO and six SRO instant
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| applicants in accordance with NUREG-1021," Examiner Standards," Interim
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| Revision 8. The exams were prepared by PSE&G staff and were approved by the
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| NRC.- PSE&G staff administered and graded the written exam. The NRC
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| administered the operating exam.
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| b. Observations and Findinos
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| -
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| The Hope Creek exam was initially scheduled for the week of February 9,1998, but
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| due to the inadequate submittal by PSE&G, the exam was delayed and rescheduled
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| for the week of February 23,1998. The PSE&G staff involved with the
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| development of these exams signed security agreements to ensure the integrity of
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| the initial exam process.
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| The PSE&G staff submitted their proposed sample plan on December 9,1997,
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| . which was later than requested in the NRC letter dated November 19,1997. The
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| sample plan was generally acceptable. Because of the reduced time for review, the
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| NRC Chief examiner made some general comments regarding low power JPMs and
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| the apparent lack of technical specification assessment on the written exam. The
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| Chief Examiner also informed PSE&G that because of the reduced time for review
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| some comments may also result from the review of the initial proposed exams and
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| these, in the final product, turned out to be minor in nature.
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| The PSE&G proposed SRO and RO exams were submitted for NRC approval on
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| January 5,1998. The PSE&G initial submitted exam was not adequate with
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| respect to discriminating between safe and unsafe license candidates. The exam
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| required significant modification to meet NRC Examiner Standards.
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| PSE&G submitted a revised exam over the period January 20-22,1998. A NRC
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| review of this submittal identified similar difficulties with the exam, but to a slightly
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| lesser degree. Following this submittal, Region i staff discussed in detail each of
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| ; the specific items of the exam at the Hope Creek training center on
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| L January 26-27,1998. The NRC subsequently issued a letter, dated
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| j. February 2,1998 officially delaying the exam and offering PSE&G an additional
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| opportunity to have the NRC administer the exam if PSE&G could submit a adequate
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| ,
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| exam by February 9,1998.
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| !
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| L . . .
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| PSE&G submitted their third version of the exam on February 9,1998. The NRC
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| concluded that the quality, while not at the totally acceptable level, was sufficient
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| to proceed with the on-site preparation activities. The PSE&G staff was able to
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| revise the exam materials during this NRC on-site preparation visit to a level that !
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| allowed the exam to be administered. l
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| 2
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| While the written question topic areas were generally acceptable, the difficulty.with
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| -
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| the specific written question generally related to the discrimination validity of the
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| question. The following summarizes the problems noted with the PSE&G written
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| exam submittals (Some examples from the initial submittal are identified): I
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| l
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| -- Poorly written question distractors which were easily eliminated. (38,43,
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| 65,67)
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| -- Questions with multiple correct answers. (15,55,76).
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| !
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| - Questions with no correct answer as written. (50,75,104,110) i
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| l
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| -- Questions that did not correlate with the assigned K/A. (31,98,116)
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| - Awkwardly worded questions. (6,96,102)
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| -- Questions stems that did not solicit the answer in the answer key. (52,59,
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| 90)
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| - Questions not appropriate for the license level. (56,58)
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| The following summarizes the problems noted with the walkthrough portion of the
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| exam submittals:
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| -- Insufficient JPM coverage against the safety function specification.
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| -- Insufficient JPMs to assess low power conditions.
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| : - Inadequate standards in the JPMs.
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| -
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| JPM and administrative questions written as simple memory or direct look up
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| rather than "open reference" use.
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| The simulator scenarios were deficient because they lacked sufficient depth to
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| properly assess applicant performance against the required competencies, as well as
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| details regarding the actions expected of the applicants. Contributing to this was
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| >
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| insufficient description of the scenario objectives, insufficient description of the
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| specific malfunction effects, insufficient critical task specification, and improper
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| completion of the forms in NUREG 1021 to assess the simulator exam.
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| ' The NRC examiners administered the operating exams in the period of
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| February 23-27,1998. PSE&G administered the written exam on March 2,1998.
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| 3
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| By letter, dated March 6,1998, PSE&G staff identified answer key comments on
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| eleven questions. A copy of the PSE&G letter is contained in Attachment 3. The
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| NRC resolution of the PSE&G comments on the written exam is described in
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| Attachment 4. PSE&G also graded the written exam based on answer key revisions
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| consistent with their comments. The NRC regraded the written exam based on the
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| NRC resolution of the facility comments.
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| During the administration of the walkthrough portion of the operating' test, several .
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| items were identified that demonstrated a poor quality product in the exam. JPM I
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| initiation cues and JPM questions contained typos in significant data that confused
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| the applicant and required the examiner to revise on the spot. One JPM and one
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| admin question had incorrect answers in the answer key. The admin JPMs did not
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| contain sufficient cues to provide to the applicant and did not contain all the
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| required attachment material to determine whether the applicant's action was
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| correct. These required considerable post exam interaction between the NRC
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| examiners and the PSE&G staff to resolve .
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| c. Conclusions
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| PSE&G staff submitted initially an inadequate exam to administer to applicants for
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| an operator's license. A majority of the test items of each portion of the
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| examination required replacement or significant modifications. Significant
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| interactions between the NRC and PSE&G, and an exam postponement for two
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| weeks, were required to develop an exam that was consistent with the NRC
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| Examiner Standards.
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| 05.2 Sianificant Control Manipulations
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| a. Scone
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| The examiner reviewed in detail the evidence of significant control manipulations J
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| performed by the applicants. These manipulations were required per 10 CFR
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| 55.31(a)(5). Guidance contained in information notice IN 97-67," Failure to Satisfy
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| ]
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| Requirements for Significant Manipulations of the Controls for Power Reactor
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| Operator Licensing" was also used.
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| b. Observations and Findines
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| PSE&G criteria and supporting documentation were not sufficient to assure that
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| applicants performed five significant control manipulations as required by 10 CFR
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| 55.31(a)(5). The criteria of "at least one' notch for a minimum of eight rods" did not .
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| assure that a manipulation was significant. This could be a very significant
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| manipulation with clearly observable power changes or not significant with no
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| power changes depending on the rods selected and its location and position in the
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| core. In addition PSE&G did not record supporting data ( initial power level, time
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| start, final power level, time end ) to demonstrate that the actual manipulation in
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| Mode 1, whether it was by recirculation flow or control rods, was significant and
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| that multiple credit was not provided for the manipulation.
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| 4
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| The PSE&G control for documenting significant control manipulations was the
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| " Reactivity Manipulations Documentation Guide," dated January 31,1997. The
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| guide documented each manipulation with a signature and date with no additional
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| specific detail provided as to what the applicant specifically performed. All the
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| applicants that took this exam, performed significant control manipulations while
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| the plant was in Mode 1. The PSE&G method and criteria for these manipulations
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| were:
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| -- Core Flow in Mode 1 - a change in reactor power, as indicated by the
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| APRMs, of at least 5%.
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| -- Individual Control Rod Manipulation in Mode 1 - at least one notch for a
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| minimum of eight rods.
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| All applicants had at least five significant control manipulations documented. Many
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| of the applicants had several of the significant control manipulations performed on
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| the same day. The data in the summary were not sufficient to determine if an
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| applicant took multiple credit for an extended continuous power change, an issue
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| identified in Information Notice 97-67. PSE&G was requested to provide additional
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| data as to what was the extent of each of the significant control manipulations.
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| The initial PSE&G response provided on February 4,1998 provided some data
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| (control room logs and control rod pull sheets) on some of the manipulations, but
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| the data was not sufficient to determine if all the control manipulations were
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| significant. Additional discussions with the PSE&G staff on February 13,1998
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| provided no new additional data. As a result, on February 18,1998, the NRC
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| informed PSE&G that many of these manipulations were not acceptable because
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| PSE&G could not provide supporting data on the extent of many of the
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| manipulations and provide information that these manipulations were significant.
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| On February 19,1998 PSE&G staff met in the Regional office and were able to
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| provide data using reactor engineering logs, additional control rod pull sheets, which
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| were not provided in earlier discussions, which allowed many of the significant
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| control manipulations to be accepted. The reactor engineering logs provided data
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| on power history when many of the manipulations were performed. Some of these
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| significant control manipulations performed on the same day were acceptable and
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| ,
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| some were not.
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| l
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| ; in the final analysis, five applicants from the February 1998 exam did not have the
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| required five significant control manipulations and one applicant had seven
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| acceptable significant control manipulations, but one of the submitted control
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| manipulation did not meet PSE&G criteria. The problems with the applications
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| were:
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| -
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| No supporting documentation was available to conclude that the
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| manipulation resulted in an observable affect on power or that the
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| manipulation was not part or a continuous power change.
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| -
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| The supporting documentation indicated that the manipulation was part of a
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| continuous power change and multiple credit was taken when only one
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| manipulation should have been credited.
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| --
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| PSE&G credited partial withdrawal of control rods following a single rod
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| scram test. This was not considered significant since this type of
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| manipulation provided little, if any, integrated response and training value.
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| )
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| -
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| Credit was taken for movement of the same four rods twice when the l
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| PSE&G criteria was to move eight rods.
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| The following summarizes these applicant's significant control manipulations. The
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| details are contained as Attachment 5.
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| Docket No. Credited Acceptable Additional Required i
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| I
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| 55-62176 7 2 3 !
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| 55-62178 5 2 3 I
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| 55-62183 5 2 3
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| 55-62187 5 2 3
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| 55-62175 5 4 1
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| 55-62174 9 7 0 (1 not reviewed)
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| 55-60813 6 4 1 1
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| Based on the concerns and findings of the NRC, the five applicants and the one
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| operator performed the required additional significant control manipulations on Hope 3
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| Creek on February 21,1998 by lowering or raising reactor power by at least 5% by 1
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| adjusting recirculation flow.
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| c. Conclusion
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| There was insufficient controls, criteria or data recorded in the controlling document
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| to assure that the control manipulations were significant and were properly credited.
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| Because of this, not all of the applicants performed five significant control
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| manipulations which had to be redone. This area is unresolved item pending further l
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| enforcement review by NRC staff with respect to meeting 10 CFR 55.31(a)(5)(50- l
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| 354/98-03-01).
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| E.8 Review of UFSAR Commitments ,
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| i
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| A recent discovery of a licensee operating their facility in a manner contrary to the l
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| updated final safety analysis report (UFSAR) description highlighted the need for a
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| special focused review that compares plant practices, procedures and/or parameters
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| to the UFSAR descriptions. While performing the exam activities discussed in this
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| report, the examiner reviewed portions of the UFSAR that related to a control rod ]
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| withdrawal accident exam question. The selected exam question reviewed was
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| '
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| '
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| consistent with the UFSAR.
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| !
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| 'I
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| a.-
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| 6
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| V. Manaaement Meetinas
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| X1 Exit Meeting Summary
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| On March 4,1998, the examiners discussed their observations of the exam process with
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| members of PSE&G management. The examiners noted that no simulator fidelity concerns
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| had been observed or identified. PSE&G management acknowledged the examiner.
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| observations.
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| LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
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| NUMBER TYPE DESCRIPTION
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| 50-354/98-03-01 URI Significant control manipulations is unresolved item pending
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| further enforcement review by NRC staff with respect to
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| meeting 10 CFR 55.31(a)(5).
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| PARTIAL LIST OF PERSONS CONTACTED
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| Licensee
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| P. Doran, Operations Training
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| H. Hanson Jr., Operations Superintendent
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| K. Krueger, Assistant Operations Manager
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| J. McMahon, Director Training, QA and EP
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| { M. Swartz, Simulator Supervisor
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| B. Thomas, Licensing
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| Attachments:
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| 1. SRO Exam and Answer Key
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| 2. RO Exam and Answer Key
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| l 3. PSE&G Comments on the Written Exam
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| 4. NRC Resolution of PSE&G Comments on the Written Exam
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| 5. Significant Control Manipulation Details
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| 1
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| 3"
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| ATTACHMENT 1
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| SRO EXAM AND ANSWER KEY
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| U.S. Nuclear Regulatory Commission
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| .S.ite-Specific
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| Written Examination
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| Applicant information
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| Name: Region: I
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| Date:. Date: 2/23/98 Facility: Hope Creek
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| License L'evel: SRO ReactorType: GE
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| '
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| Start Time: Finish Time:
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| Instructions
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| Use the answer sheets provided to document your answers. Staple this cover sheet
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| on top of the answer sheets. The passing grade requires a final grade of at least
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| 80.00 percent. Examination papers will be collected four hours after the examination
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| starts.
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| Applicant Certification
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| All work done on this examination is my own. I have neither given nor received aid.
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| Applicant's Signature
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| Results
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| Examination Value Points
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| Applicant's Score Points
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| Applicant's Grade Percent
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| e
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| Sini::r Rrct::r Operat:r An:w r Sh:ct3
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| :
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| Circle the correct answer. If an answer is changed write it in the blank.
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| 1. a b c d 26. a b c d
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| l 2: a b c 'd: 27 a b c d
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| -
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| !' 3. a b c d. 28' . a b c ' d
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| 4. a b c d 29. a b c d
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| .5. a b c d 30.. a b c d
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| 6. a b c d 31. a b c d .
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| '
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| 7. a'b c d 32. a b c d '
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| ' ~
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| 8. a b c d 33. a b c d -
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| , .. 9. a b c d 34. a b c d
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| 10. a b c d 35. a b c d
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| 36. a b c d
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| '
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| 11. a b c d 1
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| 12. a b c d 37. a b c d l
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| 13. a b c d 38. a b c d
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| 1
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| 14. a b c d 39. a b c d
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| 15. a b c d 40. a b c d '
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| 16. a b c d 41. a b c d
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| 17. a b c d 42. a b c d
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| 18. a b c d 43. a b c d
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| 19. a b c d 44, a b c d
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| 20. a b c d 45. a b c d
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| 21. a b c d 46. a b c d
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| 22. a b c d 47. a b c d ,
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| 23. a b c d 48. a b c d
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| 24. a b c d 49, a b c d i
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| ; 25. a b c d 50. a b c d
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| ,
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| Page 1
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| u.--.-.----- . . . . . . _ _
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| r
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| Senior R:cctor Oper;t:r Answ:r ShIct3
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| ..
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| Circle the correct answer, if an answer is changed write it in the blank.
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| 51. a b c d 76. a b c d
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| ' 52 'a b"c d
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| - - 77, a b c d -
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| 53. a b c d ' 78. a b c d
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| 54. a b c d 79. a b c d
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| 55.-a b c d 80. a b 'c d
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| '
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| 56. a b c d 81. a b c d
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| '82. a b c d '
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| '
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| 57.'& b c d
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| 58. a'b~ c 'd' 83. a b c d -
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| 59. . a b c . d .
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| , 84. a b c d
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| 60. a b 'c d 85. a b c d
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| 61. a b c d 86, a 'b c d
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| 62.- a b c d 87. a b c d
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| 63. a b c d 88. a b c d _
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| 64. a b c d 89. a b c d
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| 65, a b c d 90. a b c d
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| 66. a b c d 91. a b c d
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| l
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| 67. a b c d 92. a b ' c d
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| 68 a b c d 93. a b c d
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| l 69. a b c d 94, a b c d
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| 70. a b c d 95 a b c d
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| 71. a b c d 96. a b c d
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| 72. a b c d 97. a b c d
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| 73. a b c d 98. a b c d
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| 74. a b c d 99. a b c d
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| 75. a b c d 00. a b c d
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| l Page 2
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| l
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| e
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| S:ni::r Reactor Op::rator Examination
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| # 1. Which of the following evolutions is NOT cllow:d to be perform d by ths Rscctor Building
| |
| Equipment Operator?
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| a. Transferring an RPS bus to its alternate power supply with the reactor at power.
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| ~ '
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| b. Test scramming a control rod from the individual test switches'on ths hydraulic control'
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| . unit.
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| c. Operating the Standby Liquid Control system in'the Test Tank to Test Tank' mode.
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| d. Reducing hydraulic control unit nitrogen pressure to the normal band with the associated
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| control rod withdrawn.
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| 2. Given the following conditions:
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| l
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| A fully qualified Nuclear Control Operator (NCO) with an active license has just
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| returned from 10 days vacation
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| . On the first day back on shift, this NCO worked a normal 12 hour s'hift and then
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| .
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| accepted and worked 4 hours of overtime
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| Which of the following is the maximum number of hours this NCO may work on the second
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| day back on shift? (Assume no additional authorizations have been made.)
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| 1
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| a. 8 hours j
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| b.12 hours !
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| c. 14 hours
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| d. 16 hours
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| 1
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| 3. Which of the following conditions require the Operations Superintendent to perform a formal
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| turnover when delegating his Control Room Command Authority to another individual?
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| a. Command Authority is being delegated to the current on-shift Nuclear Control Operator
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| (RO) and the plant is in Op Con 4.
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| b. Command Authority is being delegated to the current on-shift Control Room Supervisor. i
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| c. Command Authority is being delegated to a current on-shift Nuclear Control Operator
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| -(RO) and the plant is in Op Con 3..
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| d. Command Authority is being delegated to a Senior Reactor Operator with an active
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| license who is not a member of the current on-shift crew.
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| Page 1.of 46
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| S nler R:act r Operater Examinatisn
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| 4. A t;gging request with switching ord:r has been receiv:d from th3 Syst:m Operctor. Tha ,.
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| Switching Order has been confirmed and the tags prepared. The System Operator has
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| contacted Hope Creek and directed the performance of the tagging request and switching
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| order. . .
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| Which of the following personnel are required to be present in the 500KV switdiyard
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| blockhouse for completion of the tagging request and switching order?
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| a. A Nuclear Equipment Operator and a Nuclear Control Operator.
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| b. Two Nuclear Equipment Operators.
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| c. A Nuclear Equipment Operator and a Control Room Supervisor.
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| d. A Nuclear Equipment Operator and a member of the Syste.ms Operation Department..
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| ,
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| 5. Followirig shift turnover the Nuclear Control Operator (RO) notes that data entere the
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| narrative log by the previous shift is incorrect. -
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| The RO draws a single line through the incorrect entry, makes the rect entry and initials
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| and dates the change. Which of the following describes how RO should highlight and
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| explain the change?
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| a. The correct entry should be circled in red wi n explanation placed in the comments
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| section.
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| b. The correct entry should be cire in red with an explanation made next to the corrected
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| entry.
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| c. The incorrect entry uld be circled in red with an explanation placed in the comments
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| section.
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| d. The in ect entry should be circled in red with an explanation made next to the
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| cted entry.
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| DeItTC ^ Se e A TTML e a f ym a d l dy .1
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| Aft 3-S-% 1) r!_ c m t.c5 } 3lat )$
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| Page 2 of 46
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| S:nior R:cct:r Op rator Excminction
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| '~
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| -
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| 6. During a valid high rarctor prcssura condition, th) R circulation Pumps did NOT
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| automatically trip as designed.
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| Which of the following actions must be taken by the Control Room to open the Recirculation
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| ~
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| "
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| ' Pump Trip (RPT) Breakers.'
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| . .
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| I
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| '
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| a. Manually initiate both channels of the Redundant Reactivity' Control System (RR.CS).
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| b. Verify the RPT Breakers trip when the Recirculation Pump MG Set Drive Motor Breakers
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| are opened.
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| c. Direct the local tripping of the RPT Breakers. -
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| .d. Depress the RPT Breaker " Trip" pushbuttons.
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| ' '
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| 7 'Which of the following are the MINIMUM guidelines f'or' Operations Superinte'ndent (OS)
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| review of critical plant parameters (reactor power, level, ' pressure and turbine load) during
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| normal, steady-state plant operations?
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| The OS shouId:
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| a. receive a verbal report from the. Control Room Supervisor (CRS) every hour..
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| l
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| .
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| b. review the current operating logs, review CRIDS, or perform a panel walkdown at least
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| I
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| twice during the 12-hour shift.
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| c. view current plant conditions on the Control Room information Display System (CRIDS) i
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| every hour,
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| i
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| d. walk-down the control room panels at least four times during the 12-hour shift.
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| 4
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| Page 3 of 46
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| *t
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| Sanior Reactor Op::rator Examination
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| ?
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| 8. Given the following conditions: .
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| .. A plant shutdown with control rod insertions occurring is in progress
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| * Reactor power is 22% with generator output at 242 MWe
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| '
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| '- The sec6nd NCO (PO) begiris deinerting the'drywell' '
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| ' '
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| * The CRS is reviewing procedures at the CRS desk
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| -
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| * No other personnel are in the Control Room
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| Which of the following additi,onal requirements, if met, would allow a License Class instant
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| ,
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| .SRO trainee under direction of the Nuclear Control Operator (RO) to continue rod. motion for.
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| the given conditions? -
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| a. Operations Manager written permission to allow a. License''Class trainee to insert control
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| ~
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| rods.
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| lb. Another technically qualified member of the unit technical staff,to observe rod movement.
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| c. Verification that the Rod Worth Minimizer is operating properly before reducing power
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| below 20%.
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| d.' A Reactor Engineer's presence to satisfy Technical Specification requirements.
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| .
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| 9. Given the following conditions:
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| The plant is shutdown for a maintenance outage
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| A Red Blocking Tag (RBT) is hung on 4160 VAC breaker
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| + The breaker is tagged in the " Test Disconnect" position -
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| + Later in the outage, the breaker is being removed from its cubicle for maintenance
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| Which of the following describes the required tagging actions for the given conditions?
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| a. The RBT shall remain on the breaker, the breaker removed from the cubicle and an
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| additional RBT installed on the ropettape placed across the opening.
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| b. The RBT shall be removed from the breaker but kept active and maintained in the
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| physical possession Gf Operations while the breaker is out of the cubicle.
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| c. The RBT shall be removed from the breaker, the breaker removed from the cubicle and
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| the same RBT installed on the safety rope / tape placed across the cubicle opening.
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| d. ' The RBT shall remain on the breaker, the breaker removed from the cubicle and a White
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| Caution Tag installed on the safety rope / tape placed across the cubicle opening.
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| Page 4 of 46
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| S:;nier Reactor Op rator Examination
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| '' 10. Which of the following describes how the Operations end Chemistry D:ptrtm:nts coordinita
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| installing Red Blocking Tags on the Hydrogen injection System?.
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| a. - Operations positions all system components
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| . Chemistry. monitors the system component positioning
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| - Operations installs the tags
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| - Chemistry performs the independent verification
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| b. - Chemistry positions all system components
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| - Operations monitors the system component positioning
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| - Chemistry installs the tags
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| - Operations performs the independent verification
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| c. -- Operations positions all system components
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| - Chemistry monitors the system component positioning
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| ~
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| .
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| - Chemistry installs the tags -
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| ~
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| - Chemistry performs the independent verification - -
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| ' d. - Chemistry positions all system components . ,
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| -
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| l
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| - Operations monitors the system component positioning
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| - Operations installs the tags
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| - Operations performs the independent verification ,
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| 11. Given the following conditions:
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| Power is 89%
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| At 1200 on 2/16/98 is discovered that, due to a recent procedure change, part of a TS
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| required surveillance was not performed.
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| The last complete satisfactory surveillance was completed at 1200 on 1/15/98
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| The incomplete surveillance was performed on 2/13/98 l
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| The surveillance is required to be performed at least once per 31 days
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| The action statement requires that inoperable equipment must be restored within 72 hrs,
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| or be in Hot Shutdown within 12 hrs and in Cold Shutdown within next 24 hours.
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| If the surveillance is not satisfactorily performed, which of the following identifies the date
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| when the unit must be in Hot Shutdown?
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| a. 2/18/98
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| b. ~ 2/19/98
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| c. 2/23/98
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| d. 2/26/98
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| Page 5 of 46
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| ''
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| S:nier Reactar Op:: rat:r Examinati:n
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| 12. Given the following conditions: .-
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| A General Emergency has been declared
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| All Emergency Response Organization facilities have been activated '
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| *
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| Planned emergency exposures 'are necessary to evacuate injured plant persorinel
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| The Radiation Protection Supervisor - Exposure Control's ALARA Analysis shows
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| expected rescue team individual exposures of 6500 mrem-
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| The Operations Support Center Coordinator, Operations Superintendent and
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| Radiological Assessment Coordinator have determined that emergency exposure
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| ~
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| .must be' received
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| Which of the following individuals must authorize the emergency exposure for the given
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| conditions? -
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| ,
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| '
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| a. Emergency Duty Officer
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| b. Emergency Coordinator
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| c. Radiological Assessment Coordinator
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| d. Operations Support Center Coordinator ,
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| 13. The estimated time to independently verify a valve position'is 15 minutes.
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| Of the listed dose rates, which is the MINIMUM dose rate that would allow waiving the " Hands
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| On" independent verification requirement for the conditions given?
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| a.10 mrem /hr
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| b. 30 mrem /hr
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| c. 45 mrem /hr
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| d. 60 mrem /hr
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| Page 6 of 46
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| e
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| Ssnisr Reactor Op:rator Examination
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| I
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| ** 14. An em:rg:ncy his occurred immidiattly r; quiring rcasonablo cctions to be taken that d:part
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| -
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| from Technical Specifications. No actions consistent with Technical Specifications that can
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| provide adequate equivalent protection are immediately apparent.
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| ' '
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| Which'of the following' identifies who is required to approve the action and under what' -
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| conditions the action can be performed?
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| a. The Control Room Supervisor approves actions to be taken to protect the health and
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| I
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| safety of facility personnel.
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| bJ The Control Room Supervisor approves actions to be taken to protect the health and
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| safety of the public.-
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| c. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to be
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| .taken to protect the health and safety of facility personnel.
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| ~
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| '
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| d.' The Emergency Coordinator, in the Emergency Ope,ra. ting Facil'ity, approves actions to'be
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| taken to protect the health and safety of the public.
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| . .
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| ~
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| ~
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| 15. V hich of the following is the first no'ification
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| t requirement and when must that notification be
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| made when a plant event requires declaration of an Alert? I
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| ~
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| a. To the N'RC - within 15 minutes of the everit occurring.
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| l
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| b. To the State and Local agencies - within 15 minutes of the event occurring.
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| c. To the NRC - within 15 minutes of the Alert declaration.
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| d. To the State and Local agencies - within 15 minutes of the Alert declaration.
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| Page 7 of 46
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| S:nlar R: actor Op;ratsr Excmination
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| '-
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| .
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| 16. Given the following conditions:
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| A major plant transient has occurred
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| '
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| '
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| 'The plant is now in a stable condition
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| * Post transieilt reviewindicates an' Alert should have'been" declared ~30 ' minutes *
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| ago but the conditions do not currently exist
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| Which of the following describes the requirements for event declaration and notification by the
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| Operations. Supervisor (OS)?
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| ?
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| 'a. The OS should declare the Alert, make the appropriate St' ate, Local and NRC
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| notifications and immediately downgrade or terminate the classification as appropriate for
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| current plant conditions.
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| b. The OS neeci not.declaie the' Alert 'but should make a non-emergency one hour report to '
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| '
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| the NRC Operations Center. .
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| c. The OS should declare the Alert, make the State, Local and NRC notifications and hold
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| at this classification until the Emergency Duty Officer (EDO) terminates the event.
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| d. The OS need not declare the Alert but should make a non-emergency four hour report to
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| the NRC Operations Center.
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| 17. Given the following conditions:
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| The plant is performing a shutdown in accordance with 10-0004, " Shutdown
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| From Rated Power To Cold Shutdown"
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| At 20% power the shutdown is completed by placing the Reactor Mode Switch
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| to " Shutdown"
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| All plant systems responded as designed during the scram
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| Post Scram Actuation Review is in progress in accordance with HC.OP-AP.ZZ-0101,
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| Post Reactor Scram /ECCS Actuation Review and Approval Requirements
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| Which of the following should be the FIRST reactor scram signal identified when reviewing
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| the Sequence Of Events printout?
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| a. Reactor Mode Switch in " Shutdown"
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| b. IRM Neutron Flux - High
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| c. Scram Discharge Volume Water Level- High
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| d. APRM Neutron Flux - Upscale, Setdown
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| Page 8 of 46
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| J
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| ic
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| L Stnior Rocctcr Op::rator Examination
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| ' 18. Giv:n ths following conditions:
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| l
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| The plant is at normal operating pressure and temperatures
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| j
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| ~
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| '
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| All' plant systems are operating as designed '
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| '
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| '
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| The "A" arid "B" scrarn to00le' switches at the hydraulic control unit for
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| '
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| control rod 42-03 have been placed in " Test" -
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| Which of the following is the expected response of the Scram Pilot Valves for Control Rod 42-
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| 03 and the Scram Dump Valves for the given conditions?
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| a. -- The Scram Pilot Valves reposition to vent the Scram inlet and Outlet Valves
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| - The Scram Dump Valves remain in their initial positions -
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| I
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| , .b. - The Scram Pilot Valves remain in their initial po'sitions. '
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| - The Scram Dump Valves remain in their initial positions
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| ~
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| '
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| c. '- The Scram Pilot Valves reposition to vent the Scram inlet and Outlet Valves .
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| - The Scram. Dump Valves reposition to vent the Scram Discharge Vent and Drain
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| '
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| Valves . .
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| d. - The Scram Pilot Valves remain in their initial positions. .
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| L - The Scram Dump Valves repcsition to vent the Scram Discharge Vent and Drain
| |
| '
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| i Valves .
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| .
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| 19. Given the following conditions:
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| The plant is performing the control rod exercise surveillance
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| i
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| The Nuclear Control Operator (RO) selects control rod 34-19 on the rod select module
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| ! Only one half of the selected rod pushbutton illuminates
| |
| Which of the following describes what has failed and how that affects the ability to move
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| control rods? ;
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| i
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| a. The selected control rod activity control card is in the scan mode and rod motion is
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| '
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| I
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| i
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| allowed.
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| b. The selected control rod activity control card is in the scan mode and rod motion is not
| |
| allowed. !
| |
| c. Only one of the two RMCS transmitter cards has successfully selected the control rod
| |
| .and rod motion is not allowed.
| |
| d. Only one of the two RMCS transmitter cards has successfully selected the control rod ,
| |
| and rod motion is allowed. j
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| Page 9 of 46
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| ''
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| Soniar Reactor Op:rator Examination
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| 20. Given the following conditions: .-
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| The plant is operating at 25% power performing a startup
| |
| Control rod 18-23 has been determined to be stuck *
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| While attem ting to withdraw the controi rod, indicated drive water flow is' reading
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| "0" gpm
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| + .
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| Which of the following is the cause of this indication?
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| a. Hydraulic Control Unit Directional Control Valve (122) has failed to reposition.
| |
| b. The 2 gpm Stabilizing Valve has failed to reposition.
| |
| c. Both Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed
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| '
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| '
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| open. .-
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| ~
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| .d. The Drive Water Header Pressure . Control Valve has failed closed.
| |
| 21. Given the following conditions:
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| - Control rod insertions are in progress for, a plant shutdown
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| The last control rod in Group 35 was inserted to Notch "02"
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| The first three control rods in Group 34 were then fully inserted
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| Insert and withdraw limits for these two Groups are Notch "00" and Notch "12"
| |
| respectively-
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| Which of the following describes what the Rod Worth Minimizer (RWM) will be displaying for
| |
| the given conditions?
| |
| a. The RWM will be displaying normal operation parameters wi'.h no alarms or errors in
| |
| .
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| effect.
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| b. The RWM will be displaying a select error with no other alarms or errors in effect.
| |
| c. The RWM will be displaying a select error with the Group 35 control rod at Notch "02" in
| |
| the withdraw error box. A rod withdrawal block is in effect.
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| d. The RWM will be displaying a select error and three insert errors. A rod insert block is in
| |
| effect.
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| Page 10 of 46
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| <
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| S:nier Reactor Operatur Examination
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| " 22. Given the following conditions:
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| A reactor startup is in progress
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| ' Reactor power is,30%
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| 'The total steam flow sisinal output from the Feedwster l'evel Control Spstem fails to the ' '
| |
| equivalent of 16% power.
| |
| Which of the following describes how the Rod Worth Minimizer will enforce control rod 3
| |
| movement for the given conditions?
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| a. The Rod Worth Minimizer will allow continued control rod movement but only in single
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| notchincrements.
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| *
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| _ b. .The Rod Worth. Minimizer will allow all normal control rod motion until actual reactor .
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| power is less than the Low Power Setpoint-
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| ,
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| .
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| - c. The Rod Worth Minimizer will immediately prevent all control rod insertions and
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| withdrawals.
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| -
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| - - .
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| '
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| id. The Rod Worth Minimizer'will' prevent co'ntrol rod withdrawals if anp control rod is
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| , withdrawn past its withdraw limit. ,
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| .
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| ,
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| ,
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| ^ 23. Given the following conditions:
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| The plant is operating at 75% power
| |
| Confirmed seal failures have occurred on the "B" Recirculation Pump
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| The pump has just been tripped
| |
| Which of the following describes the preferred order for isolation of the "B" Recirculation
| |
| Pump and the reason for that order?
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| a. Close the Suction Valve', isolate seal purge and close the Discharge valve - This order
| |
| ensures further damage is not done to the seal package from overpressure.
| |
| b. Close the Discharge Valve, isolate seal purge and close the Suction Valve - This order ,
| |
| '
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| ensures the Discharge Valve is stroked against a minimal differential pressure.
| |
| 1
| |
| c. Close the Suction Valve, isolate seal purge and close the discharge valve - This order
| |
| ansures the Suction Valve is stroked against a minimal differential pressure.
| |
| ;
| |
| d. " Close the Discharge Valve, isolate seal purge and close the Suction Valve - This order
| |
| ensures further damage is not done to the seal package from overpressure.
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| .
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| Page 11 of 46
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| ,
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| S;nior R:acter Operatar Examin tion
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| "
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| 24. Given the following conditions:
| |
| Preparations are complete to start the "A" Recirculation Pump
| |
| .
| |
| The Pump Discharge Valve (F031 A) is closed
| |
| - .. .
| |
| .,.
| |
| .
| |
| .
| |
| Which of the following describes how the "A" Recirculation Pump trip on t'he discharge. valve
| |
| ~ ~
| |
| closure is bypassed to allow the pump to.be started?
| |
| a. This trip is bypassed until the pump start sequence is complete within prescribed time
| |
| , limits. -
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| ~
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| ~
| |
| b. This trip is bypassed until the discharge valve has reached the 10d% open position.
| |
| c. This trip is bypassed until the pump has been running for 9 seconds.
| |
| d.' This trip is bypassed until'the discharge valve Jog (open) circuit has timed out.
| |
| .
| |
| 25. Given the following conditions: -
| |
| The plant is operating at 60% power with the "B" Recirculation Pump scoop tube locked
| |
| The operator is preparing to reset the scoop tube .
| |
| , ,
| |
| Speed demand on the "B" Recircybtion Pump is slightly LESS than indicated speed
| |
| Which of the following actions is the operator directed to perform if pump speed begins to
| |
| slowly rise at approximately 2% per minute after the Scoop Tube Trip / Reset pushbutton is 4
| |
| I
| |
| pressed?
| |
| a. Immediately press the "A" Recirc MG Set Drive Motor Breaker Trip pushbutton.
| |
| b. Attempt to control speed with the Increase / Decrease arrows on the Pump Speed Control
| |
| Station for the "B" Recirc pump.
| |
| c. Press the Scoop Tube Trip pushbutton for the "B" Recirc pump,
| |
| d. Press the Recirc Pump Trip (RPT) Breaker Trip pushbutton for the "B" Recirc pump.
| |
| .
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| '
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| 1
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| Page 12 of 46
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| ;
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| _ - - - - - - _ _
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| ,
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| s-
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| S ni:r R: actor Operc.tur Examinnti:n
| |
| ~~~ 26. Given tha following conditions:
| |
| The plant is operating at 75% power .
| |
| Valve. stroke tim.e testing is in pr, ogress on the "A" RHR Pump Torus Suction
| |
| ' '
| |
| Valve (F004A)
| |
| The valve is currently closed l
| |
| All other RHR system components are in their normal standby lineup
| |
| A steam break causes drywell pressure to reach 2.0 psig.
| |
| Which of the following' describes the response'of the F004A vafve and the "A" RHR pump?
| |
| a. The F004A valve automatically ~ opens and the "A" RHR Pump automatically starts after
| |
| F004A is fully open.
| |
| b. 'The F004A valve must be manually opened and the "A" RHR Pump automaticatiy starts
| |
| after F004A is fully open. ,
| |
| c. The F004A valve automatically opens but the "A" RHR Pump must be started by the
| |
| operator after F004Ais fully open.
| |
| d. The F004A valve must be manually opened and the "A" RHR Pump manually started
| |
| after F004A is fully open.
| |
| 27. Given the following conditions:
| |
| The plant is operating at 90% power
| |
| The Reactor Water Cleanup (RWCU) Inboard Isolation Valve (F001) has just
| |
| stroked closed
| |
| No other RWCU valve repositioned
| |
| RWCU responded as designed
| |
| Which of the following initiated the RWCU isolation?
| |
| a. RWCU system differential flow is excessive. >
| |
| b. The RWCU Filter /Demineralizer inlet temperatures are excessive.
| |
| c. The "A" Reactor Protection System MG set tripped.
| |
| d. The "A" and "D" NSSSS Manual Isolation pushbuttons have been armed and depressed l
| |
| simultaneously.
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| Page 13 of 46
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| _ _ _ _ _ _ _ _
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| '
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| S:niar Rrct:r Operatnr Examin:. tion
| |
| 28. Which of the following describes the rcison for hcving th3 capability to byp;ss ths Residuni ..
| |
| .
| |
| Heat Removal (RHR) Pump suction path interlocks?
| |
| a. Allows operation'of the RHR Pumps for shutdown cooling from the Remote Shutdown
| |
| -
| |
| Panel. -
| |
| >
| |
| - .
| |
| ..
| |
| b. Allows the "C" and "D " RHR Pumps to be utilized for an alternate path of suppression
| |
| pool heat removal.
| |
| c. Allows operation of the RHR Pumps for cooling the containment Hydrogen Recombiners
| |
| post-LOCA. .
| |
| d. Allows the "C" and "D" RHR Pumps to be utilized for an alternate path of core decay heat
| |
| removal.
| |
| .
| |
| 29. The plant is'in Mode 4 with' Shutdown Cooling in service on the "A" Residual' Heat Removal
| |
| (RHR) loop with the "A" RHR Pump running.
| |
| Which of the following describes how a loss of the "B" Rea'ctor Protection System (RPS) bus
| |
| will affect the inboard and Outboard Sh'utdown Cooling Iso'lation Valves (F008 & F009)?
| |
| . a. The F008 arid F009 valves' b'oth close.
| |
| b. The F008 valve closes and the F009 valve remains open. ~
| |
| c. The F008 and F009 valves both remain open.
| |
| d. The F008 valve remains open and the F009 valve closes.
| |
| 30. Given the following conditions:
| |
| . The plant is shutdown
| |
| . The reactor head is removed but no fuel has been removed from the vessel
| |
| . Shutdown Cooling is in service on the "B" Residual Heat Removal loop
| |
| Reactor coolant temperature decreases to 65 *F
| |
| Which of the following would be the expected result of the low reactor coolant temperature?
| |
| a. The reactor vessel flange thermal stress limits will be exceeded.
| |
| b. The Technical Specification reactor coolant chemistry condt::tivity limit will be exceeded.
| |
| c. The reactor temperatures can no longer be monitored.
| |
| d. The calculated shutdown margin would be invalid.
| |
| Page 14 of 46
| |
| | |
| [ .-
| |
| S:ni:r R: actor Op; rater Examinttion
| |
| ,
| |
| '' 31. During a loss of feedwater, a manual start of the High Pressure Coolant injection (HPCI)
| |
| i system was done at a water level of -20 inches by operator manipulation of the system
| |
| components.
| |
| iWhich of the folloWing describes'ths HPCI system response as reactor water level' continues
| |
| t to change?
| |
| a. It will automatically trip at +54 inches and will automatically restart at -38 inches.
| |
| b. It requires operator action to secure injection when level is greater than +54 inches and
| |
| automatically restarts at -38 inches.
| |
| c. It requires operator actions to secure injection when level is greater than +54 inches and
| |
| to restart when level is less than -38 inches.
| |
| ~
| |
| '
| |
| l
| |
| d. It wili automatically trip at +54 in'ches and Will require operator action to restart when levsl l
| |
| '
| |
| is less than -38 inches.
| |
| \, .
| |
| 32. Given the following coriditions:
| |
| A loss of coolant accident has occurred
| |
| Reactor water level reached -140 inches and is currently -50 inches and rising
| |
| Drywell pressure is 6 psig
| |
| All plant systems responded as designed
| |
| For the given conditions, which of the following describes the system isolation capabilities for
| |
| the Core Spray System (CSS) Downstream Loop Injection Valve (F0058) and the CSS
| |
| Upstream Loop Injection Valve (F004B), should Core Spray Loop "B" isolation be required?
| |
| a. Only F005B valve may be closed.
| |
| b. Neither the F0048 or F005B valves may be closed.
| |
| c. Only the F004B valve may be closed.
| |
| d. Both the F004B and F005B valves may be closed.
| |
| j
| |
| l
| |
| l
| |
| l
| |
| l
| |
| Page 15 of 46
| |
| i
| |
| l
| |
| | |
| '
| |
| S:nior R:acter Op:rator Examination
| |
| e
| |
| 33. Given the following conditions: ,
| |
| A failure-to-scram with Main Steam isolation Valve (MSIV) closure has occurred
| |
| .
| |
| The pressure spike.on the MSIV closure was 1120 psig
| |
| '
| |
| '
| |
| Reactor power is 16% and water level is -25 inches' as the 3.9 minute' timer times out
| |
| * Only Division ll of the Redundant Reactivity, Control System automatically initiates
| |
| ~
| |
| . No operator actions are taken
| |
| Which of the following is the expected plant response for the given conditions.
| |
| a. Both SLC Pumps start, both Squib Valves fire and the RWCU lsolation Valves (Inboard -
| |
| 1
| |
| F001 & Outboard - F004) close.
| |
| b. The "B" SLC Pump starts,.the "B" Squib Valve fires and only the RWCU inboard Isolation
| |
| Valve (F001) closes.
| |
| -
| |
| c. Both SLC Pumps start, both Squib Valves fire and only the RWCU ' Inboard Isolation
| |
| Valve (F001) closes.
| |
| d. The "B" SLC Pump starts,'the "B" Squib Valve fires and only the~RWCU Outboard -
| |
| Isolation Valve (F004) closes.
| |
| 34. Given the following conditions:
| |
| The plant is in a failure-to-scram condition
| |
| . Standby Liquid Control (SLC) has been initiated by the operator
| |
| * Approximately 13 minutes later the operator noted SLC Storage Tank level analog
| |
| indication on Panel 10C651 is "0" gallons
| |
| * No additional SLC system abnormalities were noted
| |
| Which of the following describes how boron injection would be continued for the given
| |
| > conditions?
| |
| a. Boron injection would continue with two SLC Pumps running.
| |
| b. Boron injection would continue with the "A" SLC Pump running.
| |
| c. Boron injection would continue with the "B" SLC Pump running,
| |
| d. Boron injection would have to be transferred to RWCU as directed by EOP-0304.
| |
| (
| |
| Page 16 of 46
| |
| _
| |
| | |
| o ,.
| |
| Sanier R:acter Op;rator Examination
| |
| # 35. Giv:n th3 following conditions:
| |
| ( * The reactor scrammed and HPCI and RCIC initiated on low reactor water level
| |
| following a loss of feedwater
| |
| . Water' level has bee'n restored to'the normal band '
| |
| , All required operator actions were taken on the scram
| |
| . All Scram Roset switches have been placed in RESET and released
| |
| ,
| |
| ! Which of the following would prevent the scram air header from repressurizing for the
| |
| l conditions given?
| |
| ' a. The Scrarn Discharge Volume High Level Scram Bypass Switch was not returned to
| |
| NORMAL.
| |
| '
| |
| b. The RPS trip logic channels'B1 and 82 fail to reenergize when RPS is reset.
| |
| .
| |
| l c. 125 VDC power is lost to one Backup Scram valve.
| |
| 1
| |
| d. The Redundant Reactivity Control System Alternate Rod insertion logic is not reset. l
| |
| .
| |
| i
| |
| 36. Given the following conditions:
| |
| The plant was performing a stdrtup following a refueling outage when a reactor
| |
| scram occurred (all rods inserted)
| |
| * The sequence of events printout shows that just prior to the scram, Average
| |
| Power Range Monitoring (APRM) channels "B" and "D" were upscale Hi-HI
| |
| Which of the following additional conditions, by itself, could have caused the full reactor
| |
| scram signal?
| |
| a. Rod Block Monitor Channel "A" has failed.
| |
| b. RPS Bus "B" has deenergized.
| |
| c. SRM Channels "A" and "C" are reading 1.5 E5 counts per second.
| |
| d. The Reactor Protection System shorting links are removed.
| |
| i
| |
| .
| |
| !
| |
| !
| |
| !
| |
| Page 17 of 46
| |
| i
| |
| | |
| '
| |
| S:nler Reactor Operatar Examination
| |
| t-
| |
| 37. Giv n th)following conditions:
| |
| The plant is operating at.100% power
| |
| * APRM Channel"D."is bypassed with the joystick
| |
| '
| |
| ;
| |
| j
| |
| ~~
| |
| '' *
| |
| ' Control rod 30-31 is selected ~
| |
| All other plant systems are operating as designed
| |
| Which of the following occurs if APRM Channel"F" fails full"downscale" for the given ;
| |
| '
| |
| conditions?
| |
| a. R~od Block Monitor Charinel "B" automatically shifts'to the "B" APRM as'its reference.
| |
| b. Rod Block Monitor Channel"B" generates a rod withdrawal block on a failure to null.
| |
| '
| |
| c. ' Rod Block Monitor Channel"B"is indicating 0%. - . ,
| |
| <
| |
| -
| |
| .
| |
| d.c: Rod Block Monitor Channel "B" is bypassed on the reference. AP.RM downscale.
| |
| < -
| |
| .;
| |
| 38. Given the following conditions:
| |
| ,
| |
| The plant is performing control rod withdrawals for a reactor startup
| |
| The reactor is subcritical-
| |
| Reactor power is 75 counts per second (CPS) in the source range
| |
| The Nuclear Control Operator (RO) selects the "B" Source Range Monitoring (SRM)
| |
| detector then holds its " Drive Out" pushbutton in the depressed position
| |
| Which of the following describes the plant response?
| |
| a. The "B" SRM detector will not withdraw due to the current power level.
| |
| b. The Retract Permit light will extinguish and the SRM Det Removal Not Permitted alarm
| |
| will be received.
| |
| c. The "B" SRM detector will retract until source range indicates less than 3 cps.
| |
| d. A Control Rod Withdrawal Block will be generated.
| |
| Page 18 of 46
| |
| | |
| <
| |
| Sani:r R:act:r Op:rctor Examination
| |
| *#
| |
| ; 39. Given the following conditions:
| |
| l
| |
| l The plant is operating at 55% power
| |
| '
| |
| * Average Power Range Monitoring (APRM) . Channel"C" currently has 14 " good"
| |
| ' '
| |
| ~ '
| |
| '' - ' ~
| |
| LPRM input signals
| |
| ^
| |
| Which of the following will result in receipt of the APRM Sys A Upscale Trip /inop alarm (C4 on
| |
| l Section C3)?
| |
| a. APRM "C" meter function switch is placed in " Flow".
| |
| b. .One of the " good" LPRMs mode switch is placed in "C" (Calibrate).
| |
| c. APRM "C" meter function switch is placed in " Average".
| |
| -
| |
| d. 'One of the " good"i.PRMs fails "downscale".
| |
| . .
| |
| 40. Which of the following describes the difference in actual reactor water level versus indicated
| |
| . wide range reactor water level and the expected change in that difference during a power
| |
| ' *
| |
| reduction from 100% to 65%?
| |
| .
| |
| a. ' Actual water level is lower than indicated level and the difference will get larger during
| |
| ,
| |
| the power re' duction.
| |
| b. Actual water level is higher than indicated level and the difference will get larger during
| |
| the power reduction.
| |
| c. Actual water level is lower than indicated level and the difference.will get smaller during
| |
| the power reduction.
| |
| d. Actual water level is higher than indicated level and the difference will get smaller during
| |
| the power reduction.
| |
| 41. The Reactor Core isolation Cooing (RCIC) system flow controller has failed full downscale
| |
| demanding a "0" gpm flowrate. The controller is in " Automatic".
| |
| Which of the following is the expected RCIC turbine response upto receipt of a valid initiation
| |
| signal for the given conditions?
| |
| a. RCIC will start, accelerate and trip on mechanical overspee'd.
| |
| b. RCIC will start, accelerate then slow to a stop.
| |
| c. RCIC will start, accelerate then will slow to and run at a low speed.
| |
| d. RCIC will start, accelerate to and run continuously at approximately 4000 rpm.
| |
| i
| |
| Page 19 of 46
| |
| i
| |
| !
| |
| | |
| s
| |
| S3nior R:actsr Operator Examinction
| |
| ..
| |
| .
| |
| 42. Given the following conditions:
| |
| * Aloss of all AC power has occurred
| |
| No, Diesel Generators are running .
| |
| The Reactor Core isolation Cooling (RCIC) systein has initiated and is injecting
| |
| A valid RCIC steam line high flow signal is received
| |
| 4
| |
| Which of the following describes the RCIC inboard and Outboard Steam Supply isolation
| |
| kMves (F007 & F008) isolation capabilities and the response of the RCIC turbine for the
| |
| given conditions?.
| |
| a. The F007 and F008 valves remain open but can be closed from the Control Room.
| |
| b. .The F007 and F008 valves remain. open and cannot be closed.from .the Con. trol Room.
| |
| -
| |
| c. Only the F007. valve closes. _ .
| |
| '
| |
| '
| |
| 'd.. Only the F008 valve closes.
| |
| .
| |
| 43.' Given the following conditions:
| |
| ~
| |
| The Automatic Depressurization System (ADS) Manual Initiatiori Channel "B"
| |
| and "F" pushbuttons (S6B and S6F) have been armed and depressed
| |
| + There is no Safety Relief Valve response
| |
| Which of the following "B" Division electrical bus failures caused this system response?
| |
| a. A loss of 120 VAC Bus 1BJ481
| |
| b. A loss of 250 VDC Bus 10D261
| |
| c. A loss of 125 VDC Bus 1BD417
| |
| d. A loss of 480 VAC Bus 10'B420
| |
| l
| |
| L
| |
| Page 20 of 46
| |
| L___________-____-_-_____-__-___-_-_ _ _ - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
| |
| | |
| ,
| |
| S nior R:act:r Op;ratar Extminttion
| |
| '' 44. Which of the following is the MINIMUM number of Stftty R:li:f Vcivas (SRV) th;t must be
| |
| opened during an Emergency Depressurization and the reason for that minimum number?
| |
| a. 4 SRVs provide the minimum depressurization rate required to ensure the low pressure
| |
| ECCS systems inject soon enough to minimize the amount of time water level is below
| |
| l the top of active fuel.
| |
| ! b'. 5 SRVs provide the minimum depressurization rate required to ensure the low pressure
| |
| ECCS systems inject soon enough to minimize the amount of time water level is below
| |
| the top of active fuel.
| |
| i
| |
| c. 4 SRVs provide the minimum steam flow through the core required to assure adequate
| |
| core cooling.
| |
| d. S SRVs provide the minimum steam flow through the core required to assure adequato
| |
| core cooling.-
| |
| '
| |
| .
| |
| 45. Given the following conditions: )
| |
| )
| |
| The plant has been operating ~at 100% power for several weeks !
| |
| All systems are operating as designed
| |
| Which of the following is the reason why periodic nitrogen makeup to the drywell is required
| |
| for the given conditions?
| |
| a. Due to leaks from drywell air operated equipment.
| |
| b. Due to PCIG normal system leakage.
| |
| c. Due to normal periodic cycling of the Torus - Drywell Vacuum Breakers.
| |
| d. Due to normal drywell air inleakage.
| |
| l
| |
| t
| |
| Page 21 of 46
| |
| l
| |
| | |
| *
| |
| S;nier Re::ctor Operator Exeminatisn
| |
| 46. Given the following conditions: 1
| |
| The plant had been operating at 75% power
| |
| A loss.of main condenser vacuum caused a complete Main Steam isolation ' -
| |
| ' Velve'(MSIV)' closure '
| |
| .
| |
| . .. The Main Condenser Vacuum Breakers have been opened
| |
| The main turbine did NOT trip and was NOT manually tripped o'n the scram ,
| |
| The MSIV switches have been placed in "Close"
| |
| . Which of the following conditions are required to allow resetting the NSSSS MSIV isolation
| |
| logic for the given conditions?
| |
| a. The Mai.n Condenser Low Vacuum Bypass switches must be in " Bypass" and the Turbine
| |
| .
| |
| '
| |
| Control Valves must be closed.
| |
| b. - The Reactor Mode Switch must be out of "Run".a.nd the Turbine Control Valves must be
| |
| closed.
| |
| ' c. The Main Condenser Low Vacuum Bypass switches must be in " Bypass" and the Turbine
| |
| Stop Valves.must be closed to less than 90% open.
| |
| d. The Reactor Mode Switch must be out of "Run" and the Turbine Stop Valves must be
| |
| closed to less than 90% open.
| |
| 47. Which of the following conditions would prevent opening the RHR "B" Loop inboard and
| |
| Outboard Drywell Spray Valves (F0218 and F0168) following a LOCA?
| |
| a. The LPCI Injection Valve (F0178) is not fully closed.
| |
| b. Less than 5 minutes have elapsed since the "B" RHR initiation occurred.
| |
| c. The RHR Full Flow Test Valve (F0248) is not fully closed.
| |
| d. Reactor water level is above -129 inches.
| |
| 1
| |
| Page 22 of 46
| |
| ,
| |
| | |
| .
| |
| S$nior R: actor Opsrator Examination >
| |
| " 48. Giv:n ths following conditions:
| |
| The Fuel Pool Cooling system is operating with one pump and heat exchanger
| |
| in service
| |
| '
| |
| The Fuel Pool Gates areinstalled'
| |
| No makeup water sources are available
| |
| Which of the following is the expected effect on Spent Fuel Pool water level and cooling
| |
| capability if a leak develops on the common FPCC Pump Suction?
| |
| .
| |
| a. Cooling capability and water level will be unchanged.
| |
| b. Cooling capability will be lost and water level will lower slightly and stabilize.
| |
| c. Cooling capability will be unchanged and water level will lower-slightly~and stabilize.
| |
| d. Cooling capability will be lost and water level will continuously lower.
| |
| ,
| |
| 49. Which of the following describes how the main steam line flow restrictors assist in maintaining
| |
| adequate core cooling for steam line break between the flow restrictors and the Main Steam
| |
| isolation Vawes?
| |
| a. They ensure the ' total inventory loss from the reactor vessel maintains level above the top
| |
| of active fuel until one division of low pressure ECCS is injecting.
| |
| b. They limit the total inventory loss from the reactor vessel to maintain water level above
| |
| the top of active fuel for a minimum of 5 seconds.
| |
| c. They ensure the total energy release rate to the Primary Containment does not result in
| |
| exceeding suppression chamber design pressure.
| |
| d. They limit the total inventory loss from the reactor vessel to maintain level above the top
| |
| of active fuel until HPCI is at rated flow.
| |
| 50. Which of the following describes the expected indicated steam flow response with an open
| |
| Safety Relief Valve (SRV) and the reason for that response?
| |
| a. Indicated steam flow goes up, because SRV steam flow is seen as additional steam flow i
| |
| over what is going to the main turbine. l
| |
| b. ' Indicated steam flow goes down, because the SRV steam flow is not monitored by the j
| |
| main steam system flow detectors. l
| |
| c. Indicated steam flow remains constant, because the Turbine Control Valves and intercept , i
| |
| Valves throttle open to maintain a steady MWe output. I
| |
| d. Indicated steam flow remains constant, because the Turbine Control Valves throttle
| |
| closed to maintain constant reactor pressure.
| |
| Page 23 of 46
| |
| | |
| '
| |
| S:ni:r R:act::r Op:: rat:r Examination
| |
| "
| |
| 51. Given the following conditions:
| |
| The plant is operating at 70% power
| |
| + The "B" EHC Pressure Regulator is tagged out of service
| |
| ' Unknown to the operator, the "A" EHC Pressure Reg'ulator output signal is
| |
| failed "as is"
| |
| Which of the following would be the expected response of the Turbins Control Valves and
| |
| Turbine Bypass Valves as the Nuclear Control Operator (RO) begins to raise power using
| |
| recirculation flow for the given conditions? (Figure attached)
| |
| a. -- The Turbine Control Valves will close
| |
| - The Turbine Bypass Valves will open
| |
| b. IThe Turbine Control Valves will close
| |
| .The Turbine Bypass Valves will not. move
| |
| ~
| |
| c. - The Turbine Control Valves will.not move
| |
| j
| |
| .The Turbine Bypass. valve will not' move
| |
| I
| |
| d. - The Turbine Control Valves will not move.-
| |
| - The Turbine Bypass Valves will open
| |
| l
| |
| l 52. Given the following conditions:
| |
| . A loss of off-site power (LOP) has occurred from 75% power
| |
| . Within 10 seconds a loss of coolant accident (LOCA) occurs
| |
| Which of the following is the expected response of the LOP and LOCA sequencers?
| |
| !
| |
| l a. As soon as power is restored to the buses, the LOCA sequencer will control the
| |
| l restoration of allloads.
| |
| b. The LOCA sequencer will begin to sequence until the diesel generator output breakers
| |
| close, then the LOP sequencer will complete load restoration.
| |
| c. As soon as power is restored the buses, the LOP sequencer will control the restoration of
| |
| all loads.
| |
| d. The LOP sequencer will begin to sequence until the diesel generator output breakers
| |
| close, then the LOCA sequencer will complete load restoration.
| |
| l
| |
| l
| |
| l
| |
| l
| |
| l Page 24 of 46
| |
| l
| |
| | |
| ~
| |
| S::nier Reacter Op::rator Examination
| |
| '' 53. Giv:n the following conditions:
| |
| The "B" Emergency Diesel Generator (EDG) had started following a valid
| |
| LOCA signal . .
| |
| Some time fater the EDG was shutdown ~using~the local Emergency Stop pushbuttons -
| |
| due to fluctuating oil pressure . ~
| |
| Concurrent with stopping the EDG, the 10A402 bus lost power
| |
| Which of the following describes the actions, if any, regarding resetting the Engine Shutdown
| |
| Relay (ESR) and th.e (86R) Lockout Relay to restart the "B" EDG and reenergize the 10A402
| |
| bus?
| |
| a. ESR must be reset
| |
| (86R). Lockout Relay reset is not re'quired
| |
| ! b. ESR must be reset
| |
| (86R) Lockout Relay must be reset
| |
| c. - ESR reset is not required
| |
| (86R) Lockout Relay reset is not required
| |
| d. ESR reset is not required
| |
| . (86R) Lockout Relay must be reset
| |
| ! 54. Which of the following parameter changes indicate the moisture content of charcoal adsorber
| |
| l bed of the Gaseous Radwaste System (GRW)is rising?
| |
| a. GRW post-treatment radiation level due to Krypton is rising.
| |
| b. GRW charcoal adsorber bed temperature is lowering.
| |
| c. GRW post-treatment radiation level due to lodine is rising.
| |
| d. GRW charcoal adsorber bed hydrogen concentration is lowering. l
| |
| l
| |
| I
| |
| {
| |
| Page 25 of 46
| |
| 4
| |
| | |
| *
| |
| Sanier Rgactcr Op;ratar Excminction
| |
| s-
| |
| 55. Giv:n the following conditions:
| |
| The plant has been operating at 100% power for several weeks
| |
| ,
| |
| Mairi Steam. Line (MSL) radiation levels have been averaging 80 mrem but are now
| |
| slowly trending upwards
| |
| Chemistry has' verified the. higher radiation readings are due to failed fue!
| |
| What are the immediate Operator Actions required for the given conditions?
| |
| a. Place additional Condensate Domineralizers in service if possible,
| |
| b. Scram the reactor and close the Main Steam Isolation Valves when MSL levels are
| |
| greater than 120 mrem.
| |
| c. Direct Reactor Water Cleanup flow to the main condenser to reduce coolant activity.
| |
| '
| |
| d . Reduce reactor power to maintain MSL radia! ion levels less than 120 mrem.
| |
| . .
| |
| 56. Given the following conditions: 4
| |
| The plant is operating at 50% power
| |
| All systems are operating normally .
| |
| One Reactor Building Ventilation System (RBVS) Exhaust Fan discharge damper
| |
| has failed to the full "open" position with the fan running
| |
| No other RBVS components have changed
| |
| Which of the following describes how this will affect the initiation of the Emergency Core
| |
| Cooling Systems (ECCS) and the reason for this?
| |
| a. ECCS will initiate after it is required because the failed damper lowers Reactor Building
| |
| pressure resulting in a lower indicated drywell pressure.
| |
| b. ECCS will initiate before it is required because the failed damper raises Reactor Building
| |
| pressure resulting in a higher indicated drywell pressure.
| |
| c. ECCS will initiate after it is required because the failed damper raises Reactor Building
| |
| pressure resulting in a lower indicated drywell pressure.
| |
| d. ECCS will initiate before it is required because the failed damper lowers Reactor Building
| |
| pressure resulting in a higher indicated drywell pressure.
| |
| l
| |
| Page 26 of 46
| |
| t
| |
| | |
| -
| |
| S :nisr R actar Operator Excminatisn
| |
| ..
| |
| 57. Given the following conditions:
| |
| The plant is operating at 40% power
| |
| . The Jet Pump operability surveillance indicates that one jet pump has failed
| |
| Technical Specifications ~ require the' plant to' be in hot shutdown within 12 hours
| |
| Which of the following describes why such a severe' restriction placed on continued operation
| |
| for the given conditions?
| |
| a. A jet pump failure at this low power level will significantly affect the core flows and result
| |
| in unacceptable thermal limits (MCPR).
| |
| b. A jet pump failure may limit reactor water level restoration capability during the reflood
| |
| portion of a Loss Of Coolant Accident.
| |
| l
| |
| c. A jet pump failure combined with the flow restricting orifices may adversely affect core
| |
| flow to the higher power fuel bundles.
| |
| '
| |
| 'd. ' A jet pump failure results in'less conservative protective ~ action setpoints for
| |
| ~ instrumentation using recirculation loop flow as an input signal.
| |
| ..
| |
| 58. Given the following conditions:
| |
| The "A" Recirculation Pump has tripped
| |
| The "A" Recirculation Pump discharge valve is open
| |
| RECIRC LOOP A JET PUMP FLOW (TOTAL) indicates 2 mlbm/hr
| |
| RECIRC LOOP B JET PUMP FLOW (TOTAL) indicates 35 mlbm/hr
| |
| RECIRC PMP B FLOW indicates 24,000 gpm
| |
| Recirc pump "B" speed is 49%
| |
| Which of the following would be expected values for total JET UMP FLOW (the flow
| |
| recorder) and actual core flow?
| |
| a. Flow recorder - 33 mlbm/hr, Actual core flow - 33 mlbm/hr
| |
| b. Flow recorder - 33 mlbm/hr, Actual core flow - 37 mlbm/hr
| |
| c. Flow recorder - 37 mlbm/hr, Actual core flow - 33 mlbm/hr
| |
| d. Flow recorder - 37 mlbm/hr, Ac.ual core flow - 37 mlbm/hr
| |
| !
| |
| !
| |
| l .
| |
| L l
| |
| l
| |
| ^
| |
| i
| |
| Page 27 of 46
| |
| | |
| ,
| |
| Sanisr R actgr Operater Examination
| |
| "
| |
| 59. Given the following conditions: ,
| |
| l
| |
| * The plant is operating at 90% power ,
| |
| '
| |
| All main turbine' sealing steam normal and backup supplies have been lost
| |
| "
| |
| '
| |
| There is no time estimate for repair / restoration
| |
| Which of the following are the immediate operator act' ions for the given conditions? i
| |
| l a. Reduce power as necessary to maintain condenser pressure less than 5.0" HGA.
| |
| .
| |
| b. Reduce recirculation flow to minimum, unload 'and trip the main turbine.
| |
| :
| |
| c.~ Reduce power as necessary to maintain adequate self-sealing steam to the main turbine
| |
| l
| |
| seals.
| |
| d. ' Reduce recirculation flow t'o maintain power less than 25% (Bypass Valve capacity).
| |
| . .
| |
| ;
| |
| *
| |
| ! ,
| |
| I
| |
| '
| |
| 60. . During a loss'of off-site power the operator is cautioned not to acknowledge the flashing '
| |
| " Trip" pushbuttons for the 4.16 KV Vital 1 E Bus infeed breakers.
| |
| 8
| |
| Which of the following will occur if these pushbuttons are pressed?
| |
| a. 'That' bus' feeder breaker will attempt to close until the anti-pump feature causes it to trip
| |
| open and remain open.
| |
| b. The Diesel Generator associated with that bus, if running, will trip and its output breaker
| |
| will open.
| |
| c. That bus' alternate feeder breaker will trip open and then immediately reclose when the
| |
| l pushbutton is released
| |
| I
| |
| d. The Diesel Generator associated with that bus will not load.
| |
| 1
| |
| !
| |
| Page 28 of 46
| |
| u
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| | |
| .
| |
| S:nier Rrctor Optrator Examination
| |
| " 61. Giv:n the following conditions:
| |
| The plant is at 45% with power ascension to 100% in prpgress
| |
| * One of the Electrical Protection Assembly (EPA) breakers on the "B" Reactor
| |
| .
| |
| ~
| |
| ' Protection Systerri(RPS) MG' set has just tripp'ed -
| |
| Breaker investigation.shows a trip on "overvoltage"
| |
| Which of the following describes the response of the Recirculation Pumps if a main turbine
| |
| trip occurs before the "B" RPS Bus is reenergized for the given conditions?
| |
| a. Both Recirculation Pumps runback to " minimum" speed.
| |
| '
| |
| ,
| |
| b. The "A" Recirculation Pump trips, the "B" Recirculation Pump runs back to " minimum"
| |
| speed. ,
| |
| ,
| |
| l
| |
| c. Both Recirculation Pumps trip.
| |
| d. 'The "B" Recirculation' Pump' trips, the "A" Recirculation Pump runs b'ack to " minimum"
| |
| . speed. .
| |
| s
| |
| 62. Given the following conditions: ;
| |
| A plant startup is in progress with the Reactor Mode Switch in "Run"
| |
| The Main Turbine is reset and is at 950 rpm accelerating to 1800 rpm
| |
| . A loss of 125 VDC power from distribution panel 1CD318 to the EHC control
| |
| logic occurs
| |
| Which of the following is the expected plant response?
| |
| a. Main turbine trips.
| |
| b. Main turbine startup would continue at the selected acceleration rate.
| |
| c. Main turbine speed will remain constant at 950 rpm. !
| |
| d. Main turbine control valves throttle closed due to a loss of the speed reference signal.
| |
| Page 29 of 46
| |
| | |
| ,
| |
| S::nicr R cter Op: rat:r Examinati n
| |
| "
| |
| 63. Givrn the following conditions:
| |
| The plantis< operating at 20% power .
| |
| A main generator load reject has just occurred
| |
| The powerhoad unbalance circ 6it tripped unexpectedly during the load reject
| |
| Which of the Ibmowing is the expected response of the Turbine Control Valves and the
| |
| Reactor Protedhon System (RPS) for the given conditions?
| |
| a. - The Twtbine Control Valves throttle closed ,
| |
| - RPS dzes not trip
| |
| b - The Turtbine Control Valves fast close
| |
| .RPS trips
| |
| c. - The Tudbine Control Valves throttle closed -
| |
| - RPS Mps
| |
| d. - The Tur'bine Control Valves fast close , ,
| |
| - RPS daes not trip
| |
| .
| |
| 64. Which of the tiillowing describes when the Main Turbine is . required to be tripp'e d'following a .
| |
| reactor scram?
| |
| a. At 50 MWe lowering
| |
| ,
| |
| b. At 25 NMAe lowering
| |
| c. At 0 MWe
| |
| d. At 50 MWe rising (reverse power)
| |
| 65. During a failure 4o-scram condition, which of the following is the criteria used to determine if
| |
| HC.OP-EO.ZZ4100(Q), " Reactor Scram", should be exited and HC.OP-EO.ZZ-0101(Q),
| |
| " Reactor / Pressure Vessel (RPV) Control", entered?
| |
| a. Reactor period on SRM Period meters is stable at -80 seconds
| |
| I
| |
| b. All APRB4*downscale" lights are not illuminated
| |
| c. . All four RPS logic channels are deenergized
| |
| l
| |
| d. All controE tods are inserted to or beyond Notch "02"
| |
| i
| |
| Page 30 of 46 l
| |
| !
| |
| <
| |
| i
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| -
| |
| .
| |
| . . .
| |
| . .
| |
| .
| |
| ___._______U
| |
| | |
| .
| |
| S:nier Recctor Optrator Excmination
| |
| .a
| |
| 66, Following a reactor scram and Main Steam Isolation Valve closure, reactor pressure reaches
| |
| 1050 psig causing the "H" and "P" Safety Relief Valves (SRV) to open.
| |
| ,,
| |
| Which of the following lists the operating setpoints..for subsequent openings of the ",P" SRV7
| |
| , ,
| |
| a. SRV "P" opens at 1047 psig and closes at 935 psig.
| |
| b. SRV"P" opens at 1047 psig and closes at 905 psig.
| |
| c. SRV "P" opens at 1017 psig and closes at 935 psig.
| |
| d. SRV "P" opens at 1017 psig and closes at 905 psig.
| |
| 67. With the plant at 100% power a severe overfeeding transient is' occurring., Water level is +50 :
| |
| inches and rising rapidly.
| |
| .
| |
| ..
| |
| ,
| |
| ,
| |
| ; Which of the following reactor water levels require termination of all feed to the reactor,
| |
| closing'the MSIVs and a reactor scram assuming none of these actions have occurred? -
| |
| l a. +54 inches
| |
| b. +65 inches
| |
| '
| |
| c. +90 inch'es
| |
| d. +118 inches
| |
| 68. Given the following conditions:
| |
| The plant is operating at 80% power
| |
| All three Feedwater Pumps are in service
| |
| Feedwater Level Control is in " Automatic - Three Element" control
| |
| . Narrow Range level "A" is reading 34 inches
| |
| Narrow Range level "B" is reading 36.5 inches
| |
| * Narrow Range level "C" is reading 35.0 inches
| |
| Which of the following would be the expected response of the Feed Water Level Control
| |
| System and reactor water level if Narrow Range level "B" failed to the low end of the rangel
| |
| a. It would transfer to Single Element Control and level would remain unchanged.
| |
| '
| |
| b. It would remain in Three Element Control and level would remain unchanged.
| |
| c. It would transfer to Single Element Control and would raise level by approximately 1.5
| |
| inches.
| |
| d It would remain in Three Element Control and would raise level by approximately 1.0
| |
| inches.
| |
| I
| |
| ;
| |
| Page 31 of 46
| |
| 1
| |
| | |
| '
| |
| S niar Reacter Op rator Excminati:n
| |
| "
| |
| 69. Which of the following is the b sis of the 65 psig Suppression Ch:mber Pressura limit?
| |
| a. 65 psig is the primary containment maximum expected post-LOCA pressure.
| |
| b. Above 65 psig, the system lineup required for containment venting may not be able to be
| |
| .
| |
| completed.
| |
| c.. Above 65 psig, the Safety Relief Valves'may not be available when required for an-
| |
| Emergency Depressurization.
| |
| d. 65 psig is the operational limit of the Torus to Drywell vacuum breakers.
| |
| 70. Given the following conditions:
| |
| The plant is operating at 95% power
| |
| * All Drywell Cooling Chilled Water pumps have tripped
| |
| Drywell pressure is rising
| |
| HC.OP-AB.ZZ-0201,"Drywell High Pressure / Loss Of Drywell Cooling", has been ,
| |
| entered and the Reactor Auxiliary Cooling System (RACS) is being aligned to supply
| |
| backup cooling to the Chilled Water System
| |
| Which of the foll'owing describes the effect of failing'to close the Chilled Water isolation
| |
| Supply and Return Valves (HV9532-2 and HV9532-1) before the transfer to RACS?
| |
| a. The RACS Pump automatic start permissives will be bypassed until the valves are closed.
| |
| b. The RACS. valves will not automatically sequence open to supply Chilled Water should a
| |
| loss of off-site power occur.
| |
| c. Chilled Water system flow will divert back into the RACS system overflowing the RACS
| |
| head tank.
| |
| d. RACS system flow will divert back into the Chilled Water system overflowing the Chilled
| |
| Water head tank.
| |
| Page 32 of 46
| |
| | |
| .,
| |
| S:ni::r R actor Operator Excmin*_tirn
| |
| " 71. During a loss of cool:nt eccid:nt the following conditions exist.
| |
| '
| |
| S'
| |
| ( '' -
| |
| j Reactor pressure is 125 psig
| |
| '
| |
| D_rywell temperature is 325 'F p b .b
| |
| Which of the following describes the accu acy and triending capabilities of wide range reactor
| |
| water level indication for the given conditi.ons?
| |
| ~-
| |
| a. They are not providing accurate reactor water level or level trend information.
| |
| ! b. They are providing acc6 rate reactor water level but level trend is not reliable.
| |
| - c. They are providing accurate reactor water level and level trend information.
| |
| ,
| |
| d, The tiot providing accurate reactor water level but level trend is reliable.
| |
| 72. Given the following conditions:
| |
| The piant is operating at 95% power
| |
| * Suppression pool temperature is 92 'F
| |
| At 0915, Safety Relief Valve (SRV)"G" opened ~
| |
| After several cycles of the SRV Open and Close pushbuttons, the operator notes
| |
| that tailpipe temperature for the SRV is stable at 305 'F and NO other plant parameters
| |
| have changed
| |
| Which of the following describes the limitations on continued reactor operation for the given
| |
| conditions? *
| |
| a. Reactor operation may continue until pressure set is reduced to less than 850 psig.
| |
| b. Reactor operation may continue until suppression pool temperature reaches 120 'F.
| |
| c. Reactor operation may continue indefinitely.
| |
| d. Reactor operation may continue until 0917.
| |
| ,
| |
| I
| |
| ,
| |
| l
| |
| l
| |
| l Page 33 of 46
| |
| | |
| ,.
| |
| S:ni r Rrct::r Operator Excmin tien
| |
| "
| |
| 73. Given the following conditions:
| |
| r \
| |
| ''
| |
| Reactor power is 82% 3
| |
| HPCI is in operation for a surv.eillance
| |
| ~ " '
| |
| '
| |
| The "B" loop of RHR is in' Suppressi6n Pool Coolin~g
| |
| Suppression pool temperature is 103 'F when the running ' pump tripped
| |
| ,
| |
| HPCI was secured
| |
| Subsequently, suppression pool temperature incre to 106 *F
| |
| Which of the following lists the suppression poo mperatures requiring entry into HC.OP-
| |
| EO.ZZ-0102, Primary Containment Control ~ entry into the LCO actions for Tech Spec
| |
| 3.6.2.17
| |
| a. EO4102 - 9$ 'F /
| |
| TS 3.6.2.1 - 95 *
| |
| b. EO-0102 - 95 *
| |
| F
| |
| TS 3.6.2.1[ -
| |
| c. EO-0102 e - 105 *F ,
| |
| TS- .1 - 95 *F
| |
| d. -0102 - 105 *F.
| |
| TS 3.6.2.1 - 105 *F ,
| |
| ,,
| |
| rc n a s ine ,, ?- !g5
| |
| '
| |
| , .: _,,
| |
| - t il '
| |
| h f 3(. M r.'s r,G qi im, t,' V"li W '' 6 U l'4 W''I!' 4 U " ,
| |
| 74. Given the following conditions: h,dc'g Wg ljtM(Mj h ''> j NJ /
| |
| ,
| |
| The plant is operating at 100% power
| |
| A feedwater heater trip has resulted in a feedwater temperature of 385 *F
| |
| No operator actions have been taken
| |
| Which of the following is the operational concem for the given conditions?
| |
| a. Entry into the Exit Region of the Power-To-Flow Map.
| |
| b. Violation of the Hope Creek Operatira License.
| |
| l
| |
| c. Immediate thermal hydraulic instabilities.
| |
| d. Recirculation Pump damage.
| |
| Page 34 of 46
| |
| _-
| |
| | |
| .
| |
| Senior Reactor Optrator Examinction
| |
| .,
| |
| 75. Which of the following describes how the operators would know the H ater ~
| |
| Chemistry injection (HWCI) system had NOT been removed from se ' whiie performing a
| |
| shutdown in accordance with HC,OP-lO.ZZ-OOO4(Q), "S, rom Rated Power To Cold
| |
| . Shutdown"?
| |
| * / ~
| |
| . .
| |
| a. Hydrogen explosions in the Mechanica _ "mPump while operating to maintain
| |
| condenser va'cuum.
| |
| b. Post-shutdown (2 hours ine Building radiation levels would be much higher.
| |
| c. Alarms and i ons resulting from a control rod drop accident would not be available
| |
| to the o ors as quickly.
| |
| d e Primary and Secondary Condensate Pumps will cavitate. .
| |
| ,.
| |
| . e Sh5r ?r i n!u l
| |
| 76. Following a reactor scram all rods are at position "00". except one that is at position "24."
| |
| Which of the following describes the capability of the reactor to remain shutdown?
| |
| a. Control rods are inserted to beyond the Maximum Suberitical Banked Withdrawal limit,
| |
| therefore the reactor will remain shutdown under all conditions.
| |
| - b. Control rods are not inserted to beyond the Maximum Suberitical Banked Withdrawal
| |
| limit, therefore it cannot be assured the reactor will remain shutdown under all conditions.
| |
| c. Design basis shutdown margin is met, therefore the reactor will remain shutdown under
| |
| all conditions,
| |
| d. Design basis shutdown margin is not met, therefore it cannot be assured that the reactor
| |
| will remain shutdown under all conditions.
| |
| l
| |
| l
| |
| I
| |
| I
| |
| l
| |
| Page 35 of 46
| |
| | |
| .
| |
| S:ni:r R::ctor Operater Extmination
| |
| "
| |
| 77. Given the following conditions:
| |
| The Control Room has been abandoned in accordance with HC.OP-AB.ZZ-0130(Q),
| |
| " Control Room Evacuation"
| |
| ' Control has been established at the' Remote Shutdown Panel in accordance with'
| |
| .HC.OP-lO.ZZ-0008(Q), " Shutdown From Outside Control Room" ~
| |
| RCIC is operating maintaining reactor water level at +35 inches
| |
| Safety Relief Valves (SRV) are being used to cooldown
| |
| Condensate Storage Tank (CST) level is 135,000 gallons
| |
| The Condensate System is not available
| |
| Which of the following is correct for the given conditions?
| |
| a. RCIC is' operated'without overspeed protection.
| |
| b.'' insufficient CST inventory is available to allow the cooldown to clear the shutdown
| |
| cooling interlocks.
| |
| -
| |
| c. The RCIC Gland Seal Condenser Condensate Pump must be manually operated..
| |
| '
| |
| d. SRVs cannot be operated in a rotation that will evenly distribute heat to the Suppression
| |
| Chamber.
| |
| 78. Which of the following describes the effect of failing to restart the Turbine Building Ventilation
| |
| System if it trips while operating in HC.OP-EO.ZZ-0104(Q)-FC, " Radioactive Release
| |
| Control"?
| |
| a. The Turbine Building will go to a slightly negative pressure.
| |
| b. The total off-site release calculations will not be accurate.
| |
| c. The Turbine Building releases will be monitored but not treated.
| |
| d. The total off-site release will be higher.
| |
| 79. A loss of Reactor Auxiliary Cooling System (RACS) has occurred.
| |
| Which of the following is the MAXIMUM time allowed before a reactor scram is required?
| |
| a. An immediate scram is required
| |
| b. One (1) minute
| |
| c. Ten (10) minutes
| |
| d. Twenty (20) minutes
| |
| Page 36 of 46
| |
| | |
| u
| |
| - 1;
| |
| S:nler React:r Op;ratar Examination
| |
| l
| |
| !
| |
| " 80. Giv:n th3 following conditions:
| |
| I
| |
| A loss of coolant accident has occurred
| |
| l_ The Reactor Auxiliaries Cooling Syste.m (RACS) has been restored
| |
| . .
| |
| ,
| |
| Which of the following describes the availability / response of the Emergency Instrument Air
| |
| '
| |
| Compressor (EIAC) for these conditions should instrument air header pressure begin
| |
| lowering?
| |
| a. The EIAC is not available until the LOCA signal is cleared, reset and the 1E breaker is
| |
| closed.
| |
| I
| |
| b. The EIAC will automatically start on instrument air header pressure less than 85 psig.
| |
| c. The EIAC is not available until the Non-1E breaker is closed and instrument air pressure
| |
| s is less than 85 psig. ,
| |
| d. The EIA' Cwill not automatically start but may be started manually from the Control Room
| |
| ,
| |
| or locally. , ,
| |
| 8.1. Which of the following describes the reason control rods insert during a loss of instrument air?
| |
| a. A flowpath is opened to'the bottom of the drive mechanism operating piston allowing i
| |
| reactor pressure to drift the rod in.
| |
| b. Drive water flow and pressure both rise enough to drift the control rod in, similar to a
| |
| normal insertion.
| |
| c. A flowpath is opened from the top of the drive mechanism operating piston allowing I
| |
| accumulator pressure to drift the rod in.
| |
| d. The normal scram flowpath to and from the drive mechanism operating piston is opened,
| |
| allowing accumulator and reactor pressure to drift the rod in.
| |
| 82. Following a loss of shutdown cooling, decay heat removal is being transferred to the
| |
| Alternate Shutdown Cooling lineup (Injection to the reactor returning to the suppression pool
| |
| via open Safety Relief Valves).
| |
| Which of the following indications is utilized to monitor the margin to Op Con 3 entry in this ;
| |
| lineup?
| |
| a. Safety Relief Valve tailpipe temperatures
| |
| b. Suppression pool temperatures l
| |
| c. Reactor vessel skin temperatures
| |
| d. Local suction temperatures on the running low pressure ECCS pumps
| |
| Page 37 of 46
| |
| _
| |
| | |
| N ,
| |
| Sanior Rsactor Op3 rater Examinstion
| |
| "
| |
| 83. Which of the following describes th3 conditions r: quiring th3 R: ctor Mods Switch to be
| |
| placed in " Shutdown" on a sustained loss of Control Rod Drive charging water header
| |
| pressure (<900 psig) with reactor pressure at 650 psig?
| |
| a. - Within 20 minutes of determining more than one CRD accumulator.is inoperable and at
| |
| least one of.those inoperable accumulators is associated with a withdrawn control rod.
| |
| b. Within 20 minutes of determining any CRD accumulator is inoperable and the inoperable-
| |
| accumulator is associated with a withdrawn control rod.
| |
| c. Immediately upon determining more than one CRD accumulator is inoperable and all the
| |
| inoperable accumulators are associated with fully inserted control rods.
| |
| d. Immediately upon determining any CRD accumulator is inoperable and the inoperable
| |
| accumulator is pssociated with a withdrawn control rod.
| |
| '
| |
| !. .
| |
| 84. Given the following conditions:
| |
| .
| |
| The plant is shutdown for refueling
| |
| The Reactor Protection System shorting links have been removed
| |
| 'A fuel bundle is being moved from the fuel pool to core.
| |
| If SRM "C" fails "downscale", which of the following are the required immediate ections?
| |
| a. Verify a control rod withdrawal block is received. Terminate fuel movement.
| |
| b. Verify a full scram and a control rod withdrawal block is received. Terminate fuel
| |
| movement.
| |
| c. Verify a control rod withdrawal block is received. Fuel movement is required to be
| |
| terminated ONLY if the fuel bundle is to be placed in the quadrant monitored by SRM "C."
| |
| d. Verify a full scram and control rod withdrawal block is received. Fuel movement is
| |
| required to be terminated ONLY if the fuel bundle is to be placed in the quadrant
| |
| ,
| |
| monitored by SRM "C."
| |
| I
| |
| l
| |
| Page 38 of 46
| |
| _
| |
| | |
| ..
| |
| S:nier R:act:r Op;rator Examination
| |
| 85. Given the following conditions:
| |
| A large break loss of coolant accident has occurred
| |
| .
| |
| '
| |
| . Drywell pressure reached a maximum of 22 psig
| |
| Suppression chambe~r sprays have ~NOT been pla'ced in service
| |
| . Drywell sprays are in service .
| |
| Drywell pre'ssure is 4 psig and slowly lowering
| |
| Which of the following is the expected positions of the Torus-to-Drywell Vacuum Breakers and
| |
| the Reactor Building-to-Torus Vacuum Breakers for'the given conditions?
| |
| a. - The Torus-to-Drywell Vacuum Brealiers are open
| |
| . The Reactor Building-to-Torus Vacuum Breakers are open
| |
| b.' - The Torus-to-Drywell Vacuum Breakers are open
| |
| . - The. Reactor Building-to-Torus. Vacuum Breakers .are~ closed ,
| |
| c. - The Torus-to-Drywell Vacuum Breakers are closed
| |
| - The Reactor Building 4o-Torus Vacuum Breakers are closed
| |
| d. - -The Torus-to-Drywell Vacuum Breakers are closed
| |
| - The Reactor Building-to-Torus Vacuum Breakers are open
| |
| .
| |
| 86. Given the following conditions:
| |
| The plant has experienced a loss of coolant accident
| |
| Suppression chamber sprays were placed in service when required
| |
| Drywell sprays were initiated with suppression pool level approximately 145 inches
| |
| Which of the following would be the result of these actions?
| |
| a. The Residual Heat Removal Pumps will be operated outside the NPSH Limit Curves.
| |
| b. Excessive differential pressures between the suppression chamber and the drywell will
| |
| occur.
| |
| c. The suppression chamber venting flowpath will be damaged leading to loss of pressure
| |
| suppression capability.
| |
| d. The suppression chamber spray capacity will be lost. i
| |
| 1
| |
| l
| |
| 4
| |
| I
| |
| Page 39 of 46
| |
| l
| |
| l
| |
| | |
| '
| |
| Senior Reactor Operator Examination
| |
| 87. Following a reccior serrm with e Mein Steam isolation Velva Closure, tha plant is b:ing s- I
| |
| depressurized using the Safety Relief Valves (SRV). !
| |
| Which of the following.is the reason.why the depressurization should be accomplished with
| |
| ~
| |
| ~
| |
| *
| |
| " sustained" SRV opening's 'if the pneumatic supply (PCIG and instrument air) is lost to the
| |
| .
| |
| SRVs?
| |
| a. This prevents exceeding the 100'FIhour cooldown limit during the depressurization while
| |
| conserving the SRV pneumatic supply,
| |
| b. This ensures the available SRV pneumatic supply is sufficient to cooldown to less than
| |
| .
| |
| the shutdown cooling interlocks.
| |
| c. This directs depressurization without regard to the Technical Specification cooldown
| |
| limits before the depleted pneumatic supply results in Ipss of SRV. control. >
| |
| d. This ensures the SRV accumulat.or pneumatic supply is available and adequate for later
| |
| us's if the Emerciency Operating Procedures require Emergency Depressurizatiori.
| |
| .
| |
| 88. The following data was collected following a Group 1 isolation and reactor scram from 100%
| |
| . power:
| |
| The Group 1 isolation was caused by technician error
| |
| The reactor scrammed on high reactor pressure
| |
| Reactor pressure peaked at 1060 psig
| |
| All control rods fully inserted
| |
| The plant was stabilized in Op Con 3
| |
| Which of the following is the basis for a decision not to startup?
| |
| a. A safety limit violation has occurred and the requirements of Technical Specification 6.7,
| |
| " Safety Limit Violation" must met.
| |
| b. The reactor steam dome pressure LCO was violated.
| |
| c. The Reactor Protection System did not respond as expected.
| |
| d. The P.edundant Reactivity Control System did not respond as expected.
| |
| i
| |
| Page 40 of 46 l
| |
| | |
| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _. . _ _ _ _
| |
| ,,
| |
| Senior Reactor Operator Examinaticn
| |
| 89. Which of the following describes the basis for initiating boron injection before exceeding the
| |
| Boron injection initiation Temperature (BilT)? -
| |
| a. This ensures the reactor will be shutdown and in hot-standby conditions before the
| |
| suppression pool reaches the heat capacity level limit.
| |
| b. This ensures the reactor will be shutdown and in hot-standby conditions before the
| |
| suppression pool reaches the heat capacity temperature limit
| |
| c. This ensures the Primary Containment Pressure Limit will not be exceeded before RPV
| |
| pressure is below the Minimum Alternate Flooding Pressure.
| |
| d. This ensures suppression pool temperature will not exceed 150 *F during an Emergency
| |
| Depressurization, if required.
| |
| 90: Given the following condition:
| |
| * The plant is operating in HC.OP-EO.ZZ-0206, " Reactor Flooding"
| |
| Suppression chamber pressure is 22 psig
| |
| Reactor pressure is 105 psig
| |
| ,
| |
| 4 SRVs have been opened and have remained open for 85 minutes
| |
| All reactor water level indicators are off-scale high
| |
| Which of the following would be the MINIMUM expected actual reactor water level for the
| |
| given conditions?
| |
| a. -209 inches
| |
| b. -161 inches
| |
| c. +118 inches
| |
| d. Filled solid
| |
| l
| |
| l
| |
| Page 41 of 46
| |
| _ _ _ _ _
| |
| | |
| .
| |
| Sonier React:r Operatar Examinati:n
| |
| e
| |
| 91. HPCI and RCIC both started and are injecting in response to a valid low reactor water level.
| |
| Current plant conditions are as follows:
| |
| * Reactor water level is +25 inches, steady
| |
| 4 Reactor pressure is'845 psig, rising slowly
| |
| Drywell pressure is 1.1 psig, steady .
| |
| RCIC has been aligned to Full Flow Recirc operation (CST to CST) for pressure control
| |
| HPCI is injecting to the reactor for level control
| |
| After 10 minutes of operation a valid high suppression pool level is received
| |
| Which of the following would be the expected response of RCIC if a valid high suppression
| |
| pool level is received for the given conditions?
| |
| ~
| |
| a. RCIC will remain in Full Flow' Recirculation.
| |
| b. RCIC will trip on high turbine exhaust pressure.
| |
| c. RCIC will trip on low suction pressure.
| |
| '
| |
| d. RCIC will' operate on minimum flow.
| |
| 92. During high primary containment water level condilions, suppression pool water level
| |
| bdications cannot be used.
| |
| Operation of which system will invalidate the alternate method used for determining primary
| |
| containment water level?
| |
| a. RCIC
| |
| b. Core Spray
| |
| c. RHR
| |
| d. HPCI
| |
| I
| |
| I
| |
| l
| |
| l
| |
| Page 42 of 46
| |
| | |
| .,
| |
| S:;nier R: actor Op:ratar Examination
| |
| 93. Given the following conditions:
| |
| A leak has occurred in the suppression pool
| |
| *
| |
| + The reactor is shutdown. ' '
| |
| ' '
| |
| ,
| |
| . A cooldown is being performed using SRVs~
| |
| The Heat Capacity Level Limit (HCLL) curve is being monitored ,
| |
| . The " Action Required' area of the HCLL curve has been entered for several minutes
| |
| .
| |
| Which of the following is a possible effect of initiating an emergency depressurization with the
| |
| given conditions?
| |
| a. The suppression pool may exceed design temperature. ,
| |
| . .b. Failure of the downcomer vent header joints due to " chugging."
| |
| .
| |
| . c. The SRNailpipe Level Limit curve may be exceeded.
| |
| d'. The capacity of the Torus to Drywell vacuum breakers will be' exceeded.
| |
| .
| |
| . . . .
| |
| '94. ' Following the runback of the Recirculation Pumps on a trip of a Primary Condensate Pump,
| |
| the operator may monitor the Source Range Monitoring (SRM) per.iod meters for strong i
| |
| deflections above and below " Infinity".
| |
| Under which of the following conditions may SRM period indications be considered accurate -
| |
| indication of thermal hydraulic instabilities?
| |
| a. Only when the SRM detectors are fully withdrawn from the core,
| |
| .
| |
| b. . Anytime, regardless of detector position, if the detectors are stationary,
| |
| c. Only when the SRM detectors are fully inserted into the core,
| |
| d. Anytime the SRM detectors are moving.
| |
| 1
| |
| l
| |
| l
| |
| I
| |
| 1
| |
| 1
| |
| l
| |
| i
| |
| i
| |
| i
| |
| ;
| |
| i
| |
| l
| |
| Page 43 of 46
| |
| | |
| '
| |
| l
| |
| Seni::r Reactor Operator Examinctisn
| |
| '-
| |
| 95. With the plant et pow;r ths M2in Storm / Rs:ctor Wrtsr Cleanup Arsa Lerk Temperature
| |
| High alarm was received and the RWCU system automatically isolated. The leak has been
| |
| determined to be in the RWCU Pipe Chase Room 4402.
| |
| ~
| |
| Which of the following is NOT a required operator action for the given' conditions?
| |
| ~
| |
| a. Notify Chemistry to close the" Manual' Sample Line Isolation Valves P-RC-V9670 and 1-
| |
| RC-V006.
| |
| b. Observing the RWCU Inboard and Outboard Isolation Valves (F001 & F004) close,
| |
| c. Observing the Recirc Sample Line isolation Valves (BB-SV-4310 and 4311) automatically
| |
| close.
| |
| d. Operate available Reactor Building ventilation fans consistent with plant conditions.
| |
| ,
| |
| -
| |
| ,
| |
| ,
| |
| 96. Given the following conditions:
| |
| ~
| |
| The plant was operating at rated power when a steam line break occurred in the HPCI
| |
| room
| |
| . HPCl isolated due to high room temperatures
| |
| . RBVS exhaust radiation levels reached 1.0 E-2 microcuries/mi
| |
| Which of the following describes the ventilation system response for the given conditions?
| |
| a. RBVS remains in service
| |
| - b. RBVS isolated,6 FRVS Recire and 1 FRVS Vent Fans are in service
| |
| c. RBVS isolated,4 FRVS Recirc and 1 FRVS Vent Fans are in service
| |
| d. RBVS isolated,6 FRVS Recirc and 2 FRVS Vent Fans are in service
| |
| 97. With the Reactor Building Ventilation in a normal lineup, the operator observes that Reactor
| |
| Building pressure is .10 inches of vacuum water gauge.
| |
| Which of the following is an immediate action to restore Reactor Building pressure to the
| |
| required pressure?
| |
| a. Place at least two FRVS units in service.
| |
| b. Secure a reactor building supply fan.
| |
| c. Place an FRVS unit in service and increase FRVS flow rate to maximum.
| |
| d. Place the third Reactor Building Exhaust Fan in service.
| |
| Page 44 of 46
| |
| | |
| ,
| |
| 1 S:nicr ROIctor Operator ExaminLtion
| |
| l
| |
| * 98. Given the following conditions:
| |
| <
| |
| . The reactor has scrammed from power
| |
| . Pilot Scram Valve Group 1 solenoids, Logic "A" and Logic "B" did not doenergize
| |
| ,
| |
| * ' -
| |
| The Screm Discharge Volume is currently full
| |
| Which of the following describes the difference between inserting control rods in accordance
| |
| I with HC.OP-EO.ZZ-0303, " Individual Control Rod Scrams" and HC.OP-EO.ZZ-0302, "De-
| |
| energization Of Scram Solenoids"?
| |
| a. EO-0302 requires resetting RPS and ARI, EO-0303 does not.
| |
| b. EO-0303 requires resetting RPS and ARI, EO-0302 does not.
| |
| c. EO 0303 does not isolate the Scram Discharge Volume, E04302 'does.-
| |
| l d. EO-0302 results in a flowpath from the RPV to the Reactor Building Sump, EO-0303.does
| |
| '
| |
| .not.
| |
| .
| |
| .
| |
| 99. Which 'of the following are the appropriate hydrogen concentration values to complete the
| |
| . following statement following a loss of coolant accident with hydrogen generation occurring?
| |
| Rising containment hydrogen concentrations require corrective actions be taken at
| |
| l and reentry into HC.OP-EO.ZZ-0102, " Primary Containment Control", at
| |
| '
| |
| !
| |
| a. 2.0%, - 0.5%
| |
| b. 0.5%, 2.0%
| |
| c. 2.0%, 2.0%
| |
| d. 0.5%, 0.5%
| |
| I
| |
| L
| |
| Page 45 of 46
| |
| ,
| |
| | |
| .
| |
| S:nier React:r Op:ratcr Extminatian ,,
| |
| 100 Givon '.he following conditions:
| |
| A loss of coolant accident has occurred
| |
| Hydrogen is present in the primary containment
| |
| ~The Hydrogen Recombiners have been started
| |
| Which of the following is the hydrogen concentration that requires termination of Hydrogen
| |
| Recombiner operation and the reason why that value is selected?
| |
| a. The Hydrogen Recombiners are secured at 4% hydrogen concentration because there is
| |
| insufficient oxygen available to support the recombination reaction.
| |
| b. The Hydrogen Recombiners are secured at 6% hydrogen concentration because there is
| |
| . insufficient. oxygen available to support the recombination r.eaction.
| |
| c. The Hydrogen.Recombiners are secured at 4% hydrogen concentration in order to
| |
| * '
| |
| prevent their becoming an ignition source.
| |
| .d. The Hydrogen Recombiners are secured at 6% hydrogen concentration in order to
| |
| prevent tiieir becoming'an igniti6n sou'rce'.
| |
| .
| |
| .
| |
| .
| |
| I
| |
| Page 46 of 46
| |
| | |
| Seni:r R :ct:r Operator Answ:r K;y
| |
| .,
| |
| i
| |
| 1. b 294001G101 26. d 203000K406
| |
| 2a 294001G102 . 27. c 204000K115 .
| |
| 3. d 294001G104 28. d- 205000A104-
| |
| 4. c 294001G108 29. gn 205000A203
| |
| n c ~ r r~ ~ e n -a s r> tw 3 s 'n
| |
| '+/ .
| |
| 5. e 204001OM8~
| |
| ,seu res we en ,nro.s riot 1.>.,b
| |
| .
| |
| 30. d 205000G421 N ''1"lI l
| |
| 6. c 294001G128 31 a 206000K102
| |
| 7.' b 294001G131 32. a 209001A403 ,
| |
| 8. b 294001G202 - 33. a. 211000A208
| |
| 9 '. . c 294001G213. 34. a 211.000K506 .
| |
| 10. d 294001G217 35. d .212000A414
| |
| 11. d 294001G222 36 d 212000K411
| |
| 12. a 294001G304 37. d 215002K604
| |
| 13. c 294001G310 38, d 215004A104
| |
| 14. b 294001G412 39. b -215005K104
| |
| 15. d 294001G440 40. d 216000A301
| |
| 16. b 294001G441 41. c 217000A210
| |
| 17. d 294001G448 42. b 217000K201
| |
| 18. a 201001K405 43. c 218000K201
| |
| i
| |
| ! 19. c 201002A405 44. c' 218000K302
| |
| 20. a 201003A207 45. b 223001K103
| |
| 21. a' 201006K514 46. c 223002A403
| |
| 22. d 201006K602 47. a 226001K403
| |
| .
| |
| 23 c 202001A210 48. b 233000K302
| |
| 24. aoed 202001A302 49. b 239001G128
| |
| s e < nrr~ ke a h- A& v Gs ifd Ff*
| |
| 25. b 202002A101 d' z'' . ,! I2 50. b 239002A109
| |
| Page 1
| |
| | |
| .
| |
| .
| |
| S:ni:r Rrct:r Operator An w:r KGy ..
| |
| 51. c 241000K302 76. c 295015A202
| |
| 52; a 262001A304 ;77. c 295016A108
| |
| 53'. b '264000K603 78. b 295017K302
| |
| 54. a 271000A408 79. Cc Y' 295018K202wt M.
| |
| sn . rteejgg
| |
| '2d$d1NAT0I ' yp'###a%4'N3+W 7 "I' '
| |
| 55. d 272000A201 80. a
| |
| 56. d 290001K601 81. d 295019K201
| |
| 57. b 290002K401 82. a 295021A104- .
| |
| 58. a 295001A203 83..d 295022K207.
| |
| 59..a 295002A105 84. a 295023G23.2
| |
| 60. d 295003A101 85. b 295024A116
| |
| 61 c 295003K204 86.'b ~295024K101
| |
| 62. a 295004K203 87. d 295025K102
| |
| 63 d 295005K201 88. c 295025K201
| |
| 64. c 295006G449 89. b 295026K304
| |
| 65. b 295006K103 90. b 295028K302
| |
| 66. a 295007K304 91. d 295029A104
| |
| 67. c 295008G123 92 d 295029A201
| |
| 68. d 295009K202 93. a 295030K103
| |
| 295031A202
| |
| 69. car b 295010A202see arre ce ugs trorar%g.,pp3lli g). b
| |
| 70. d 295010K302 95. c 295032G448
| |
| e t@ld dSe CXM ,
| |
| '
| |
| ,,, . - . ,. , , i v i I V .P '''* 96. pb 295034K102 '
| |
| *
| |
| " H d 3 ye r M Y' % ' %' '*?
| |
| 72.Sed(y# 235'OT3 Aid 2 ' '"" ''' " k0k'
| |
| ' y~ see wmean e va~~ ~wys dM's W '**hh#I
| |
| 97. d 295035A201
| |
| 295037K205
| |
| ""'L_ -
| |
| -
| |
| 2050iOOiUE _ _
| |
| ' %.;.n.u w-h 6n.gr{nn
| |
| , :< fu
| |
| -
| |
| 6%
| |
| 98. c
| |
| 74. b 295014G110
| |
| "
| |
| 99. b 500000G404
| |
| r>. ( 500000K303
| |
| 75. c 23501 iKivo 00 c
| |
| 4 l' V TC d ,~Sf* , , , p,,
| |
| WMM&f*"I f g #*
| |
| Y
| |
| 3-5-11
| |
| , ,. 7
| |
| '' F , .7 -b :m - f Page 2
| |
| | |
| . .
| |
| .
| |
| - - . . ,
| |
| ,
| |
| Y
| |
| 3/4.0 aPPLI M afLITY ~
| |
| 4
| |
| ^ LIMITING CONDITION FOR OPERATION - . ... .. .
| |
| .... .. .. ........................... ....... .-
| |
| 3.0.1 Compliance with the Limiting conditions .for Operation contained in th's >
| |
| succeeding Specifications is required during the OPERATIONAL CONDITIONS or
| |
| other conditions specified thereins except that upon failure to most the
| |
| Limiting Conditions for Operation, the associated ACTION requirements.shall be
| |
| met.
| |
| 3.0.2 Noncompliance with a Specification shall exist when the requirements of
| |
| the Limiting Condition for Operation and associated
| |
| ~
| |
| ACTION requirements are
| |
| If the Limiting condition for
| |
| not met within the specified time intervals.
| |
| Operation is restored prior to expiration of the specified time intervals,
| |
| completion of the Action requirements is not required.
| |
| 3'.0.3 When a Limiting condition for Operation is. net not, ascept as provided
| |
| in the associated ACTION requirement's,'within one hour action shall be
| |
| initiated to place the unit in an OPERATIONAL CONDITION in which the
| |
| ,
| |
| '
| |
| ' Specification does not apply 'by placing it,- as applicable, in
| |
| ,
| |
| 1. At least'STARTUF within the nest 6 hours,
| |
| 2. At.least NOT SEUTDONN within the following 6 hours, and
| |
| 3. At least 00LD SNUTDONN within the subsequent 24 hours.
| |
| ~
| |
| Where corrective measures are completed that permit operation under the ACTION
| |
| - requirements, the ACTION may be taken in accordance with the specified time
| |
| limits as measured from the time of failure to meet the Limiting condition for
| |
| Operation. Raceptions to these requirements are stated in the individual
| |
| Specifications.
| |
| This Specification is not applicable in OPERATIONAL CONDITIONS 4 or 5.
| |
| 3.0.4 Entry into an OPERhTIONAL CONDITION or other specified condition shall
| |
| not be made when the conditions for the Limiting condition for Operation are *
| |
| not met and the associated ACTION requires a shutdown if they are not met
| |
| within a specifLed time interval. Entry into an OPERATIONAL CONDITION or
| |
| other specified condition may be made in accordance with the ACTION
| |
| requirements when conformance to them permits continued operation of the
| |
| facility for an unlimited period of time. This provision shall not prevent
| |
| passage through or to OPERATIONAL CONDITIONS as required to. comply wit
| |
| requirements. Exceptions to these requirements are stated in the individual
| |
| SpecLtLeatLons. '
| |
| 3.0.5 Equipment removed from service or declared ' inoperable to comply with
| |
| ACTIONS may be returned to service under administrative control solely to
| |
| perform testing required to demonstrate its OPERASILITY
| |
| other equipment.
| |
| service under administrative control to perform the testing required to
| |
| demonstrate OPERABILITY.
| |
| l
| |
| !
| |
| Amendment No. 63 l
| |
| ROFE CREEK
| |
| 3/4 0-1
| |
| | |
| _
| |
| f
| |
| j[id$i$hh!$2!p,x ''/ea
| |
| y ** 'oi
| |
| 4
| |
| e,1
| |
| % ;, p*a g g; g,.
| |
| .
| |
| se gx .-
| |
| .
| |
| .34i
| |
| .
| |
| $
| |
| 1
| |
| 4 .,
| |
| o
| |
| vn.. . n \
| |
| r ,_
| |
| | |
| ' .
| |
| :
| |
| ..
| |
| {
| |
| APPLICABILITY
| |
| ,
| |
| l
| |
| SURVEILLANCE REQUIREMENTS (Continued) l
| |
| . . \
| |
| Pressure Vessel Code and applicable Addenda shall be applicable as
| |
| follows in these Technical Specifications: 3
| |
| ASNE Boiler and Pressure Vessel Required frequencies
| |
| Code and applicable Addenda for performing inservice
| |
| terminology for inservice inspection and testing
| |
| inspection and testing activities activities
| |
| )
| |
| Weekly At least once per 7 days
| |
| Monthly At least once per 31 days
| |
| Quarterly or every 3 months At least once per 92 days
| |
| Semiannually or every 6 months At least once per 184 days
| |
| Every 9 months At least once per 276 days i
| |
| Yearly or' annually At least once per 366 days
| |
| c. The provisions of Specification 4.0.2 are applicable to the above
| |
| required frequencies for performing inservice inspection and testing
| |
| '
| |
| activities.
| |
| -
| |
| d. Performance of the above inservice inspection and testing activities
| |
| shall be in addition to'other specified Surveillance Requirements.
| |
| e. Nothing in the ASME Boiler and Pressure Vessel Code shall be con-
| |
| ',- strued to supersede the requirements of any Technical Specification.
| |
| f. The Inservice Inspection Program for piping identified in NRC
| |
| Generic Letter 88-01 shall confom to the staff positions on schedule,
| |
| methods, and personnel, and sample expansion included in that generic
| |
| letter, or as otherwise approved by the NRC.
| |
| l
| |
| i
| |
| !
| |
| l
| |
| 1
| |
| 3/4 0-3 Amendment No. 51
| |
| HOPE CREEK
| |
| | |
| .,
| |
| .
| |
| i
| |
| HC.OP-SO.CH-0001(Z) .-
| |
| ATTACHMENT 4
| |
| (Page1of1)
| |
| .
| |
| MAIN TURBINE CONTROL OIL (EHC) SYSTEM OPERATION - .
| |
| 'EHC CONTROL LOGIC DIAGRAM
| |
| - _,e --
| |
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| l
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| Hope Creek Page x2 or 84 Rev.19
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| I
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| #
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| ed
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| ATTACHMENT 2
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| RO EXAM AND ANSWER KEY
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| t
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| ,
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| ..
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| U.S. Nuclear Regulatory Commission
| |
| . Site-Specific .
| |
| Written Examination
| |
| Applicant Information
| |
| Name: Region: 1
| |
| Date: Date:. 2/23/98 - Facility: Hope Creek
| |
| License Level: RO Reactor Type: GE
| |
| Start Time: Finish Time:
| |
| Instructions
| |
| Use the anr,wer sheets provided to document your answers. Staple this cover sheet
| |
| on top of the answer sheets. The passing grade requires a final grade of at least
| |
| 80.00 percent. Examination papers will be collected four hours after the examination
| |
| starts.
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| Applicant Certification
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| l
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| All work done on this examination is my own. I have neither given nor received aid.
| |
| Applicant's Signature
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| I
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| Results
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| Examination Value Points
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| Applicant's Score Points
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| Applicant's Grade Percent
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| .
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| .
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| R: actor Oper_"ttr An:wer Sheeta
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| =s
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| Circle the correct answer, if an answeris changed write it in the blank.
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| 1. a b c d 26. a b c d
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| 2. a b c d - ~27..a bec d .
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| 3. a b' c d 28. a b c d
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| 4. a b c d 29. a b c d
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| 5. a b c d 30. a b c d
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| 6. a b c d 31. a b c d
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| * -
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| abcd
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| *
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| 7'. 32. a b'c a
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| 8. a b c d ' 33, a b c d- -
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| 9. a b c d .
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| 34. a..b c.d .
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| 10. a b c d 35. a b c d
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| 11. a b c d 36. a b c d
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| 12. a b c d 37. a b c d I
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| i
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| 13. a b c d 38. a b c d
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| 14. a b c d 39. a b c d
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| 15. a b c d 40. a b c d !
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| 16. a b c d 41. a b c d
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| ^
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| 17. a b c d 42, a b c d
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| 18. a b c d 43, a b c d
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| 19. a b c d 44. a b c d
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| 20. a b c d 45. a b c d
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| 21. a b c d 46. a b c d
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| 22. a b c d 47 abcd
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| 23. a b c d 48. a b c d
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| l
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| -24. a b c d 49. a b c d
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| [
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| 25. a b c d 50. a b c.d
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| Page.1
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| 1
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| R: actor Operator An:wcr Shscts
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| ,.
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| Circle the correct answer. If an answer is changed write it in the blank.
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| .
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| 51. s_b c d' 76. a b c d
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| 52- a.b c d '
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| .
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| -
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| 77. a b c d .
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| -
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| '
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| 53.'s b_c d 78. a b c d .
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| 54. a b c d 79. a b c d
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| 55. a b c d 80. a b c d
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| ~ 56. a b c d 81. a b c d
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| '57, a b c d '82. a b c d
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| 58. a b c.d 83. a b c d
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| 59. a b c d , .84. a-b c d ,
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| 60. a b c d- 85. a b c d
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| 61. a b c d -86. a b c d
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| 62, a b c d 87. a b c d
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| 63, a b c d 88. a b c'd
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| 64, a b c d 89. a b c d
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| 65.'a b c d 90 a b c d
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| 66. a b c d 91, a b c d
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| 67, a b c d 92. a b c d
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| 68. a b c d 93. a b c d
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| 69. a b c d 94 abcd
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| 70. a b c d 95. a b c d
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| 71. a b c d 96. a b c d
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| 72. a b c d 97. a b c d
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| 73. a b c d 98. a b c d
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| 74. a b c d 99, a b c d
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| 75. a b c d 00. a b c d
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| Page 2
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| i
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| ,, Reactor Operatar Examination
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| 1. Which of the following evolutions is NOT allowed to be performed by the Reactor Building
| |
| Equipment Operator?
| |
| a. Transferring an RPS bus to its alternate power supply with the reactor at power.
| |
| ti. ' Test scramming a control rod'from the' individual test switch'es on the hydraulic control
| |
| unit.
| |
| c. Operating the Standby Liquid Control system in the Test Tank to Test Tank mode.
| |
| d. Reducing hydraulic control unit nitrogen pressure to the normal band with the
| |
| associated control rod withdrawn.
| |
| 2. Given the following conditions:
| |
| * A fully qualified Nuclear Control Operator (NCO) with an active license has just
| |
| returned from 10 days vacation
| |
| ' On the first day back on shift, this NCO wo*ed a normal 12 hour shift and then
| |
| accepted and worked.4 hours of overtime
| |
| Which of the following is the maximum number of hours this NCO may work on the second
| |
| day back on shift? (Assume no addition'ai authorizations have been made.)
| |
| a. 8 hours
| |
| b. 12 hours
| |
| c. 14 hours- -
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| l
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| 1
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| d. 16 hours
| |
| 3. A tagging request with switching order has been received from the System Operator. The
| |
| Switching Order has been confirmed and the tags prepared. The System Operator has
| |
| contacted Hope Creek and directed the performance of the tagging request and switching
| |
| order.
| |
| Which of the following personnel are required to be present in the 500KV switchyard
| |
| blockhouse for completion of the tagging request and switching order?
| |
| l a. A Nuclear Equipment Operator and a Nuclear Control Operator.
| |
| b. Two Nuclear Equipment Operators.
| |
| c. A Nuclear Equipment Operator and a Control Room Supervisor.
| |
| d. A Nuclear Equipment Operator and a member of the Systems Operation Department.
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| !
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| !
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| I
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| 1
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| l '
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| Page 1 of 45
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| l
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| ,
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| R actor Op rator Examination
| |
| -
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| 4. Following shift turnover the Nuclear Control Operator (RO) notes that data entered in t
| |
| narrative log by the previous shift is incorrect.
| |
| The RO draws a single line through the incorrect entry, makes the corr entry and initials
| |
| ,
| |
| and dates the change. Which of the following describes how the should highlight and
| |
| explain the change?
| |
| a. The correct entry should be circled in red wit explanation placed in the comments
| |
| section.
| |
| b. The correct entry should be cir in red with an explanation made next to the
| |
| corrected entry.
| |
| c. The incorrect ent ould be circled in red with an explanation placed in the comments
| |
| section.
| |
| d. The ' rrect entry should be circled in red with an explanation made next to the
| |
| rrected entry.
| |
| Deterea see cn m ros:s srueue f(sc 3-s-W
| |
| 5. Which of the following will identify when Op Co'n 2 is entered during a reactor startup and
| |
| heatup?
| |
| a. When the reactor is declared critical.
| |
| b. When the first control rod is withdrawn.
| |
| c. When the MODE switch is placed in Startup/ Hot Standby.
| |
| ~
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| d. When enough control rods are withdrawn to increase keff to greater than or equal to .99.
| |
| 6. During a valid high reactor pressure condition, the Recirculation Pumps did NOT
| |
| automatically trip as designed.
| |
| l Which of the following actions must be taken by the Control Room to open the Recirculation
| |
| Pump Trip (RPT) Breakers,
| |
| s. Manually initiate both channels of the Redundant Reactivity Control System (RRCS).
| |
| b. Verify the RPT Breakers trip when the Recirculation Pump MG Set Drive Motor Breakers
| |
| are opened.
| |
| c. Direct the local tripping of the RPT Breakers.
| |
| d. Depress the RPT Breaker " Trip" pushbuttons.
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| Page 2 of 45
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| l
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| ~ Reacter Operator Excmination
| |
| 7. Which of the following are the minimum requirements for the " Board" Nuclear Control
| |
| Operator (RO) to review critical plant parameters (reactor power, level, pressure and turbine
| |
| load) and walk down the control boards during normal, steady-state plant operations?
| |
| The RO should:
| |
| a. continuously monitor critical plant parameters and perform a complete control board
| |
| walk down every hour.
| |
| b. monitor critical plant parameters every five (5) minutes and perform a complete control
| |
| board walk down every two (2) hours.
| |
| c. continuously monitor critical plant parameters and perform a complete control board
| |
| walk down every two (2) hours.
| |
| d. monitor critical plant parameters every five (5) minutes and perform a complete control
| |
| board walk down every hour.
| |
| 8. Given the following conditions:
| |
| A plant shutdown with control rod insertions occurring is in progress
| |
| Reactor power is 22% with generator output at 242 MWe
| |
| The second NCO (PO) begins deinerting the drywell
| |
| The CRS is reviewing procedures at the CRS desk
| |
| No other personnel are in the Control Room
| |
| Which of the following additional requirements, if met, would allow a License Class Instant
| |
| SRO trainee under direction of the Nuclear Control Operator (RO) to continue rod motion for j
| |
| the given conditions?
| |
| '
| |
| a. Operations Manager written permission to allow a License Class trainee to insert control
| |
| rods.
| |
| b. Another technically qualified member of the unit technical staff to observe rod movement.
| |
| c. Verification that the Rod Worth Minimizer is operating properly before reducing power
| |
| below 20%.
| |
| d. A Reactor Engineer's presence to satisfy Technical Specification requirements.
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| 4
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| Page 3 of 45
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| ~ i
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| R:actar Op rct:r Ex minatian l
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| -
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| 9. Given the following conditions:
| |
| The plant is shutdown for a maintenance outage j
| |
| '
| |
| A Red Blocking Tag (RBT) i,s hung on 4160 VAC breaker
| |
| The breaker is tagged in the " Test Disconnect" position I
| |
| -
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| Later in the outage, the breaker is being removed from its cubicle for maintenance
| |
| Which of the following describes the required tagging actions for the given conditions?
| |
| a. The RBT shall remain on the breaker, the breaker removed from the cubicle and an
| |
| additional RBT installed on the rope / tape placed across the opening.
| |
| b. The RBT shall be removed from the breaker but kept active and maintained in the
| |
| physical possession of Operations while the breaker is out of the cubicle.
| |
| c. The RB,T shall be removed from the breaker, the breaker removed from the cubicle and
| |
| .
| |
| l the same RBT installed on the ~ safety rope / tape placed across the cubicle opening.
| |
| . d. The RBT shall remain on the breaker, the breaker removed from the cubicle and a
| |
| White Caution Tag installed on the safety rope / tape placed across the cubicle open;ng.
| |
| ~
| |
| \
| |
| 10. Given the following conditions:
| |
| A Hope Creek radiation worker is fully qualified with current lifetime exposure
| |
| records on file
| |
| I
| |
| This individual's current yearly exposure (TEDE) is 355 mrem
| |
| A Site Area Emergency has just been declared
| |
| .
| |
| Which of the following is the MAXIMUM additional exposure that can be received by this
| |
| ! individual without exceeding any administrative or procedurally based limits? (Assume no
| |
| I additional approvals have been received.)
| |
| a. 1645 mrem
| |
| b. 4145 mrem
| |
| c. 4395 mrem
| |
| i d. 4645 mrem
| |
| i
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| Page 4 of 45
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| ..
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| l .
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| l Rrct:r Oper; tor Excmin tien
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| l~
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| l
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| 11. The estimated time to independently verify a valve position is 15 minutes.
| |
| !
| |
| Of the listed dose rates, which is the MINIMUM dose rate that would allow waiving the " Hands
| |
| ~
| |
| On" independent verification requirement for the conditions given?
| |
| .
| |
| a. 10 mrem /hr
| |
| b. 30 mrem /hr
| |
| c. 45 mrem /hr
| |
| d. 60 mrem /hr
| |
| 12. An emergency has occurred immediately requiring reasonable actions to be taken that depart
| |
| from Technical Specifications. No actions consistent with Technical Specifications that can
| |
| provide adequate equivalent protection are immediately apparent.
| |
| I
| |
| Which of the following identifies who is required to approve the action and under what
| |
| conditions the action can be performed?
| |
| a. The Control Room Supervisor approves actions to be taken to protect the health and )
| |
| safety of facility personnel,
| |
| b. The Control Room Supervisor approves actions to be taken to protect the health and
| |
| safety of the public. ,
| |
| 1
| |
| c. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to
| |
| be taken to protect the health and safety of facility personnel.
| |
| d. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to
| |
| be taken to protect the health and safety of the public.
| |
| 1
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| !
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| Page 5 of 45
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| | |
| Rxctor Operater Examination
| |
| 13. Given the following conditions:
| |
| The plant is performing a shutdown in accordance with 10-0004, "Shu,down
| |
| . From Rated Power To Cold Shutdown" . _
| |
| .
| |
| At 20% power the shutdown is completed by pla'cing the Reactor Mod..i Switch
| |
| to " Shutdown"
| |
| All plant systems responded as designed during the scram
| |
| . Post Scram Actuation Review is in progress in accordance with HC.OP-AP.ZZ-0101,
| |
| Post Reactor Scram /ECCS Actuation Review and Approval Requirements
| |
| Which of the following should be the FIRST reactor scram signal identified when reviewing
| |
| the Sequence Of Events printout?
| |
| a. Reactor Mode Switch in '' Shutdown"
| |
| b.' IRM Neutron Flux - High ,
| |
| c. Scram Discharge Volume Water Level- High
| |
| d. APRM Neutron Flux- Upscale, Setdown
| |
| *
| |
| ,
| |
| l
| |
| l 14. Given the following conditions:
| |
| The plant is operating at 55% power
| |
| All systems are operating normally in automatic
| |
| Which of the following is the expected response of the Scram Discharge Volume (SDV) vent
| |
| and drain system if APRM Channel"A" fails full" upscale"?
| |
| a. One Scram Dump Valve repositions, all SDV Vent and Drain Valves close.
| |
| b. One Scram Dump Valve repositions, all SDV Vent and Drain Valves remain open.
| |
| c. The Scram Dump Valves do not change position, all SDV Vent and Drain Valves remain
| |
| open.
| |
| d. One Scram Dump Valve repositions, one set of SDV Vent and Drain Valves close.
| |
| l
| |
| l
| |
| l
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| l
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| Page 6 of 45
| |
| | |
| -
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| a
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| ..
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| .
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| R ; actor Op:: rater Examination
| |
| l- 15. Given the following conditions:
| |
| -
| |
| * The plant is at normal operating pressure and temperatures ,
| |
| l. . . All plant systems are ope,ating
| |
| r as designed . . ,. , ,.
| |
| The "A" and "B" scram toggle switches at the hydraulic control unit for
| |
| ,
| |
| control rod 42 03 have been placed in " Test"
| |
| Which of the following is the expected response of the Scram Pilot Valves for Control Rod 42-
| |
| 03 and the Scram Dump Valves for the given conditions?
| |
| a. - The Scram Pilot Valves reposition to vent the Scram inlet and Outlet Valves ,
| |
| -- The Scram Dump Valves remain in their initial positions
| |
| . b. - The Scram Pilot Valves remain ~in their initial positions
| |
| . The Scram Dump Va.lves remain in their initial positions j
| |
| c. -- The Scram Pilot Valves reposition to vent the. Scram inlet and Outlet Valves
| |
| -- The Scram Dump Valves reposition to vent the Scram Discharge Vent and Drain
| |
| '
| |
| ' '- '
| |
| Valves -
| |
| '
| |
| d. -- The Scram Pilot Valves remain in their initial positions
| |
| - The Scram Dump Valves reposition to vent the Scram Discharge Vent and Drain
| |
| Valves. ,
| |
| 16. Given the following conditions:
| |
| The plant is performing the control rod inxercise's'urveillance
| |
| The Nuclear Control Operator (RO) selects control rod 34-19 on the rod select module
| |
| Only one half of the selected rod pushbutton illuminates
| |
| Which of the following describes what has failed and how that affects the ability to move
| |
| control rods?
| |
| a. The selected control rod activity control card is in the scan mode and rod motion is
| |
| allowed,
| |
| b. The selected control rod activity control card is in the scan mode and rod motion is not !
| |
| allowed.
| |
| c. Only one of the two RMCS transmitter cards has successfully selected the control rod
| |
| and rod motion is not allowed.
| |
| d. Only one of the two RMCS transmitter cards has successfully selected the control rod
| |
| ,
| |
| and rod motion is allowed.
| |
| I
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| l
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| Page 7 of 45
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| :
| |
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| , .
| |
| Reactor Operator Examination
| |
| -
| |
| -
| |
| 17. Given the following conditions:
| |
| * The plant is operating at 25% power performing a startup
| |
| . Control rod 18-23 has been determined to be stuck
| |
| . While attempting to' withdraw the control rod, indicated drive water flow is reading
| |
| "0" gpm
| |
| Which of the following is the cause of this indication?
| |
| a. Hydraulic Control Unit Directional Control Valve (122) has failed to reposition.
| |
| b. The 2 gpm Stabilizing Valve has failed to reposition.
| |
| c. Both Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed
| |
| open.
| |
| d. The Drive Water Header Pressure Control Valve ha's failed closed.
| |
| 18. The current Rod Worth Minimizer (RWM) group has insert and withdraw limits of Notch 24
| |
| and Notch 36 respectively.
| |
| Which of the following are the control rod attemate limits allowed by the RWM for this group?
| |
| a. Notch 22 and Notch 34
| |
| b. Notch 22 and Notch 38
| |
| c. Notch 26 and Notch 34
| |
| d. Notch 26 and Notch 38
| |
| Page 8 of 45
| |
| | |
| Recctor Op: rater Examinati::n
| |
| ..
| |
| 19. Given the following conditions:
| |
| The p is operating at 75% power
| |
| . Confirmed . failures have occurred on the "B" Recirculation Pump
| |
| The pump has ju en tripped
| |
| '
| |
| Which of the following descri the order for "B" Recirculation Pump valve manipulation that
| |
| must be followed in order to ensu e pump will be completely isolated?
| |
| a. Close the Discharge Valve, isolate al purge, isolate RWCU flow from the loop and
| |
| close the Suction Valve.
| |
| b. Isolate the seal purge, close the Suction Val isolate RWCU flow from the loop and
| |
| close the Discharge Valve. .
| |
| c. Close the Suction Valve, close the Distarge Valve, i te seal purge, and isolate
| |
| RWCU flow from the loop.
| |
| .d. Isolate the seal, purge, close the ,Dischar
| |
| s e Vs Ive iso} ate RW ow frope loop and
| |
| .
| |
| close the Suction Valve. p
| |
| .
| |
| 20. Given the following conditions:
| |
| Preparations are complete to start the "A" Recirculation Pump
| |
| The Pump Discharge Valva (F031 A) is closed
| |
| Which of the following describes how the "A" Recirculation Pump trip on the discharge valve
| |
| closure is bypassed to allow the pump to be started?
| |
| a. This trip is bypassed until the pump start sequence is complete within prescribed time
| |
| -
| |
| limits.
| |
| b. This trip is bypassed until the discharge valve has reached the 100% open position,
| |
| c. This trip is bypassed until the pump has been running for 9 seconds.
| |
| d. This trip is bypassed until the discharge valve jog (open) circuit has timed out.
| |
| 21. With the plant at 100% power, which of the following would cause a drop in reactor power and
| |
| a rise in the "A" Recirculation Loop drive flow?
| |
| a. A jet pump has failed in the "B" Recirculation loop.
| |
| b. The "B" Recirculation Pump speed has risen.
| |
| c. A jet pump has failed in the "A" recirculation loop.
| |
| d. The "A" Recirculation Pump speed has risen.
| |
| Page 9 of 45
| |
| | |
| ,
| |
| R:: actor Op ratcr Examination j
| |
| ..
| |
| 22. Given the following conditions:
| |
| The plant is operating at 60% power with the "B" Recirculation Pump scoop tube locked
| |
| '
| |
| .
| |
| The operator is preparing to reset th.e scoop tube
| |
| Speed demand on the "B" Recirculation Pump is slightly LESS than indicated speed
| |
| Which of the following actions is the operator directed 'to perform if pump speed begins to
| |
| slowly rise at approximately 2% per minute after the Scoop Tube Trip / Reset pushbutton is
| |
| pressed?'
| |
| a. Immediately press the "A" Recirc MG Set Drive Motor Breaker Trip pushbutton.
| |
| b. Attempt to control speed with the increase / Decrease arrows on the Pump Speed Control
| |
| Station for the "B" Recirc ~ pump.
| |
| c. Press the Scoop Tube Trip pushbutton for the "B" Recirc pump.
| |
| d. Press the Recirc Pump Trip (RPT) Breaker Trip pushbutton for th'e "B" Recirc pump.
| |
| . . .
| |
| 23' Which of the following is the MAXIMUM speed at which the Recirculation Pumps can operate
| |
| .with NO Reactor Feedwater Pumps operating?
| |
| a. 20%
| |
| b. 30%
| |
| c. 45%
| |
| d. 50%
| |
| Page 10 of 45
| |
| | |
| R actor Operat:r Examin2tian
| |
| ..
| |
| 24. Given the following conditions:
| |
| * The plant is operating at 75% power
| |
| ,
| |
| Valve stroke time testing is in progress on the "A" RHR Pump Torus Suction
| |
| Valve (F004A)
| |
| The valve is currently closed .
| |
| * All other RHR ~ system components are in their normal standby lineup
| |
| * A steam break causes drywell pressure to reach 2.0 psig.
| |
| Which of the following describes the response of the F004A valve and the "A" RHR pump?
| |
| a. The F004A valve automatically opens and the "A" RHR Pump automatically starts after
| |
| , F004A is fully open. , ,
| |
| I
| |
| .b. The F004A valve must be manually opened and the "A' RHR Pump automatically starts
| |
| 'after F004A is fully open. .
| |
| ,
| |
| c. The F004A valve automatically opens but the "A" RHR Pump must be started by the
| |
| -
| |
| -
| |
| ' operator after F004A l's fully open.
| |
| -
| |
| d.' The F004A valve must be manually opened 'and the "A" RHR Pump manually started -
| |
| after F004A is fully open.
| |
| 25. Given the following conditions:
| |
| The plant is operating at 90% power
| |
| The Reactor Water Cleanup (RWCU) Inboard Isolation Valve (F001) has just
| |
| stroked closed I
| |
| No other RWCU valve repositioned
| |
| RWCU responded as designed
| |
| Which of the following initiated the RWCU isolation?
| |
| a. RWCU system differential flow is excessive.
| |
| b. The RWCU Filter /Demineralizer inlet temperatures are excessive,
| |
| c. The "A" Reactor Protection System MG set tripped.
| |
| d. The "A" and "D" NSSSS Manual isolation pushbuttons have been armed and depressed
| |
| simultaneously.
| |
| !
| |
| !
| |
| !
| |
| !
| |
| '
| |
| Page 1.1 of 45
| |
| ;
| |
| | |
| -
| |
| Reactcr Op;rator Examinttion
| |
| "
| |
| 26. Which of the following describes the reason for having the capability to bypass the Residual
| |
| Heat Removal (RHR) Pump suction path interlocks?
| |
| a. Allows operation of the RHR Pumps for shutdown cooling from the Remote Shutdown
| |
| -
| |
| Panel.
| |
| b. Allows the "C" and "D " RHR Pumps to be utilized for an alternate path of suppression
| |
| pool heat removal.
| |
| c. Allows operation of the RHR Pumps for cooling the containment Hydrogen Recombiners
| |
| post-LOCA.
| |
| d. Allows the "C" and "D" RHR Pumps to be utilized for an alternate path of core decay
| |
| heat removal.
| |
| 27. The plant is in Mode 4 with Shutdown Cooling in servics on the "A" Residual Heat Removal
| |
| (RHR) loop with the "A" RHR Pump running.
| |
| Which of the following describes how a loss of the "B" Reactor Protection System (RPS) bus
| |
| will affect the Inboard and Outboard Shutdown Cooling isolation Valves (F008 & F009)?
| |
| a. The F008 and F009 valves both close.
| |
| b. The F008 valve closes and the F009 valve remains open.
| |
| c. The F008 and F009 valves both remain open.
| |
| d. The F008 valve remains open and the F009 valve closes.
| |
| 28. During a loss of feedwater, a manual start of the High Pressure Coolant injection (HPCI)
| |
| system was done at a water level of -20 inches by operator manipulation of the system
| |
| components.
| |
| Which of the following describes the HPCI system response as reactor water level continues
| |
| to change?
| |
| a. It will automatically trip at +54 inches and will automatically restart at -38 inches.
| |
| b. It requires operator action to secure injection when level is greater than +54 inches and
| |
| automatically restarts at -38 inches.
| |
| c. It requires operator actions to secure injection when level is greater than +54 inches and
| |
| to restart when level is less than -38 inches.
| |
| d. It will automatically trip at +54 inches and will require operator action to restart when
| |
| level is less than -38 inches.
| |
| Page 12 of 45
| |
| -__ _____ _ -____--______- _ _ _ _____ - - -
| |
| | |
| ,,
| |
| Reactor Operator Examination
| |
| 29. Given the following conditions:
| |
| The plant is operating at 70% power
| |
| An inadvertent initiation of HPCI has occurred *
| |
| . HPCI injection to the vessel is' occurring
| |
| Which of the following is the required IMMEDIATE action for the given conditions?
| |
| a. Close the HPCI Main Pump Discharge Valve (F007) and depress the Turbine Trip
| |
| pushbutton.
| |
| b. Depress the Turbine Trip pushbutton and stop the Auxiliary Oil Pump.
| |
| c. Control. reactor water level manually to maintain level between Level 4 and Level 7.
| |
| d. Reduce reactor power as necessary by running bacii Recirculation flow and inserting
| |
| -
| |
| control rods. .
| |
| .
| |
| . . .
| |
| '30. Given the following conditions:
| |
| A loss of coolant accident has occurred
| |
| Reactor water level ~is -110 inches and lowering
| |
| Reactor pressure is 290 psig and lowering
| |
| Which of the following is the minimum combination of the CSS Manual Initiation pushbuttons
| |
| that must be armed and depressed to place four Core Spray Pumps in service and injecting?
| |
| (Assume the manual initiation pushbuttons are operable.)
| |
| a. "A" and "B"
| |
| b. "A" and "C"
| |
| c. "C" and "D"
| |
| d. "A", "B", "C" and "D"
| |
| Page 13 of 45
| |
| - .
| |
| _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| ~
| |
| R cctor Opercter Examinatian
| |
| "
| |
| 31. Given the following conditions: ,
| |
| * A loss of coolant accident has occurred
| |
| . Reactor water level reached -140 inches and is currently -50 inches and rising
| |
| ,
| |
| * Drywell' pressure is 6 psig ,
| |
| All plant systems. responded as designed
| |
| For the given conditions, which of the following describes the system isolation capabilities for
| |
| the Core Spray System (CSS) Downstream Loop injection Valve (F0058) and the CSS
| |
| Upstream Loop Injection Valve (F004B), should Core Spray Loop "B" isolation be required?
| |
| a. Only F005B valve may be closed.
| |
| . b. Neither the F004B or F0058 valves may be closed.
| |
| c. Only the F004.B valve may be closed.
| |
| d. Both the F004B and F0058 valves may be closed.
| |
| . . .
| |
| .
| |
| ,
| |
| 32. Given the following conditions:
| |
| A failu're-to-scram with Main Steam Isolation Valve (MSIV) closure has occurred
| |
| . The pressure spike on the MSIV closure was 1120 psig
| |
| . Reactor power is 16% and water level is -25 inches as the 3.9 minute timer times out
| |
| Only Division 11 of the Redundant Reactivity Control System automatically initiates
| |
| No operator actions are taken
| |
| Which of the following is the expected plant response for the given conditions.
| |
| a. Both SLC Pumps start, both Squib Valves fire and the RWCU isolation Valves (Inboard -
| |
| F001 & Outboard - F004) close.
| |
| b. The "B" SLC Pump starts, the "B" Squib Valve fires and only the RWCU inboard
| |
| Isolation Valve (F001) closes.
| |
| L c. Both SLC Pumps start, both Squib Valves fire and only the RWCU Inboard Isolation
| |
| Valve (F001) closes.
| |
| d. The "B" SLC Pump starts, the "B" Squib Valve fires and only the RWCU Outboard
| |
| Isolation Valve (F004) closes.
| |
| Page 14 of 45
| |
| | |
| .
| |
| V
| |
| )
| |
| i
| |
| .. Rxcter Operatcr Examinstion
| |
| 33. Given the following conditions:
| |
| The plant is in a failure-to-scram condition
| |
| Standby Liquid Control.(S,LC) has been initiated by the operator.
| |
| L
| |
| . Approximately 13 minutes later the operator noted SLC Storage' Tank level analog
| |
| '
| |
| indication on. Panel 10C651 is "0" gallons'
| |
| No additional SLC system ' abnormalities were noted
| |
| Which of the following describes how boron injection would be continued for the given
| |
| j conditions? -
| |
| a. Boron injection would continue with two SLC Pumps running.
| |
| L b. Boron injection would continue with the "A" SLC Pump running.
| |
| c. Boron injection would continue with the "B" SLC Pump running. ,
| |
| d. Boron injection would have to be transferred to RWCU as directed by EOP-0304.
| |
| < .
| |
| . .. . ,
| |
| ,
| |
| ,
| |
| ^ '
| |
| ! ' 34. Which of the following is the raison why the Reactor Protection System (RPS) power supplies
| |
| l contain Electrical Protection Assembly (EPA) broakers for specific protection against
| |
| i undervoltage, overvoltage and underfrequency conditions? ,
| |
| j a. To maintain bus parameters during short duration power interruptions (less than 2
| |
| '
| |
| seconds).
| |
| b. To provide a highly reliable, stable power supply to the RPS supplied loads, specifically
| |
| l instrumentation. ,
| |
| l c. To maintain a close tolerance power supply for the Scram Pilot Valve solenoids I
| |
| I
| |
| l preventing spurious deenergization.
| |
| '
| |
| d. To provide a highly reliable, stable power supply to ensure the Scram Pilot Valve
| |
| ;
| |
| . solenoids will reposition during a reactor scram. j
| |
| l :
| |
| l-
| |
| l-
| |
| .
| |
| Page 15 of 45
| |
| | |
| L
| |
| Renctsr Operatsr Examinati::n
| |
| ..
| |
| 35. Given the following conditions:
| |
| The plant was performing a startup following a refueling outage when a reactor
| |
| . , , scram occurred (all rods inserted)
| |
| The sequence of events printout shows that just prior to the scram,' Average
| |
| Power Range Monitoring (APRM) channels "B" and "D" were upscale Hi-HI -
| |
| Which of the following additional conditions, by itself, could have caused the full reactor
| |
| scram signal?
| |
| a. Rod Block Monitor Channel "A" has failed.
| |
| b. RPS Bus "B" has deenergized.
| |
| c. SRM Channels "A" and "C" are reading 1.5 E6 iounts per second.
| |
| d. The Reactor Protection System shorting linktare removed.
| |
| 36. The Nuclear Control Operator (PO) is performing backpanel checks and reports the following
| |
| _
| |
| . indications on the Traversing incore Probe (TIP) "A" and "B" subsystem panel (Refer to '
| |
| attached figure):
| |
| Squib Monitor lights - both illuminated
| |
| Shear Valve Monitor lights . - both extinguished
| |
| Ball Valve "Open" lights - both extinguished
| |
| Ball Valve " Closed" lights - both illuminated
| |
| Which of the following is the status of the "A" and "B" TIP shear valves and primary
| |
| containment integrity?
| |
| a. The TIP Shear Valves are operable and primary containment integrity is met.
| |
| b. The TIP Shear Valves are inoperable and primary containment integrity is met.
| |
| c. The TIP Shear Valves are inoperable and primary containment integrity is not met.
| |
| d. The TIP Shear Valves are operable and primary containment integrity is not met.
| |
| i
| |
| l
| |
| Page 16 of 45
| |
| .. .. . _ _ . ..
| |
| | |
| .,
| |
| R:act:r Operat:r Extminati:n
| |
| '
| |
| 37. Given the following conditions:
| |
| l
| |
| The plant is operating at 100% power
| |
| ;
| |
| APRM Ch,annel "Q" is bypassed with the joystick ,,
| |
| * Control rod 30-31 is selected - ~
| |
| All other plant systems are operating as designed
| |
| Which of the following occurs if APRM Channel "F" fails full "dow.ucale" for the given
| |
| conditions?
| |
| a. Rod Block Monitor Channel"B" automatically shifts to the "B" APRM as its reference,
| |
| b. Rod Block Monitor Channel "B" generates a rod withdrawal block on a failure to null.
| |
| c. Rod Block Monitor Channel"B"is indicating 0%.
| |
| d. Rod Block Monitor Channel"B"is bypassed on the reference APRM downscale.
| |
| - . . ..
| |
| .
| |
| 38. Given the following conditions.:
| |
| Control rod insertions are in progress for scheduled plant shutdown
| |
| ' Current reactor power is 17%
| |
| Intermediate Range Monitoring (IRM) Channel "A" has failed full" upscale" and
| |
| has NOT been bypassed with the joystick
| |
| Whico of the following describes what will occur as the power reduction continues in
| |
| accordance with HC.OP-lO.ZZ-0004(Q), " Shutdown From Rated Power To Cold Shutdown"
| |
| and when it will occur?
| |
| a. A half scram will occur when the IRM detectors are fully inserted.
| |
| b. A control rod block will occur when IRM "A" is ranged down from Range 8 to Range 7.
| |
| c. A half scram will occur when the Mode Switch is placed in Startup.
| |
| d. A control rod block will occur when the IRM detectors are fully inserted.
| |
| !
| |
| !
| |
| Page 17 of 45
| |
| | |
| .
| |
| R:actsr Operater Examinati:n
| |
| -
| |
| 39. Given the following conditions:
| |
| The plant is performing control rod withdrawals for a reactor startup
| |
| '
| |
| The reactor is suberitical
| |
| Rea'ctor power is 75 cou'nts per second (CPS) irithe so'urce rafige
| |
| '
| |
| The Nuclear Control Operator (RO) selects the "B" Source Range Monitoring (SRM) ,
| |
| '
| |
| detector then holds its " Drive Out" pushbutton in the depressed position
| |
| t
| |
| Which of the following describes the plant response?
| |
| a. The "B" SRM detector will not withdraw due to the current power level.
| |
| b. The Retract Permit light will extinguish and the SRM Det Removal Not Permitted alarm
| |
| will be received.
| |
| c. The "B" SRM detector win retract until source range indicates less than 3 cps.
| |
| d. A Control Rod Withdrawal Block will be generated.
| |
| 40. Given the following conditions:
| |
| The plant is operating at 55% power
| |
| Average Power Range Monitoring (APRM) Channel "C" currently has 14 " good"
| |
| LPRM input signals
| |
| Which of the following will result in receipt of the APRM Sys A Upscale Trip /Inop alarm (C4 on
| |
| Section C3)?
| |
| a. APRM "C" meter function switch is placed in " Flow".
| |
| b. One of the " good" LPRMs mode switch is placed in "C"(Calibrate).
| |
| c. APRM "C" meter function switch is placed in " Average".
| |
| 'd. One of the " good" LPRMs fails "downscale".
| |
| '
| |
| Page 18 of 45
| |
| | |
| .
| |
| - Reacter Op:;rator Examination
| |
| 41. With the plant operating at 85% power, steady state conditions, a narrow range water level is
| |
| reading 35".
| |
| Which of the following will be the indicated " level." from this instrument if the differential
| |
| ~
| |
| .
| |
| ~
| |
| pressure acros's the detector fails to "O" psid for the given conditions?
| |
| a. O inches
| |
| b. 30 inches
| |
| c. 35 inches
| |
| d. 60 inches
| |
| 42. Which of the following describes the difference in actual reactor water level versus indicated
| |
| wide range reactor water level and the expected change in that difference during a power
| |
| reduction from 100% to 65%7
| |
| a. Actual water leDel is iower than indicated level and the difference will get larger during
| |
| the power reduction.
| |
| b. Actual water level is higher than indicated level and the difference will get larger during
| |
| the power reduction.
| |
| c. Actual water level is lower than indicated level and the difference will get smaller during
| |
| the power reduction.
| |
| d. Actual water level is higher than indicated level and the difference will get smaller during
| |
| the power reduction.
| |
| ,
| |
| '
| |
| ?
| |
| l
| |
| Page 19 of 45
| |
| l
| |
| | |
| '
| |
| R: actor Operat r Examinatl2n
| |
| -
| |
| 43. Given the following conditions:
| |
| The Reactor Core Isolation Cooling (RCIC) is oper.ating in Full Flow Recirc
| |
| The RCIC flow controller is in " Automatic" ,
| |
| RCIC turbine speed is 2450 rpm
| |
| Which of the following describes the expected res~ponse of RCIC turbine speed and system
| |
| flow if the operator throttles the RCIC Test Bypass To CST isolation Valve (F022) in the
| |
| "open" direction for the given conditions?
| |
| (Compare the conditions after they stabilize to before the valve was throttled.)
| |
| a. - RCIC turbine speed lowers
| |
| - System flow remains unchanged
| |
| b. - RCIC turbine speed lowers
| |
| - System flow goes down
| |
| c. - RCIC' turbine speed raises'
| |
| - System flow remains unchanged
| |
| d. - RCIC turbine speed raises
| |
| - System flow goes up-
| |
| 44. Given the following conditions:
| |
| A loss of all AC power has occurred
| |
| ,
| |
| No Diesel Generators are running
| |
| ! The Reactor Core Isolation Cooling (RCIC) system has initiated and is injecting
| |
| A valid RCIC steam line high flow signal is received
| |
| Which of the following describes the RCIC Inboard and Outboard Steam Supply isolation
| |
| Valves (F007 & F008) isolation capabilities and the response of the RCIC turbine for the
| |
| given conditions?
| |
| j a. The F007 and F008 valves remain open but can be closed from the Control Room.
| |
| l b. The F007 and F008 valves remain open and cannot be closed from the Control Room.
| |
| !
| |
| c. Only the F007 valve closes.
| |
| d. Only the F008 valve closes.
| |
| l
| |
| l
| |
| Page 20 of 45
| |
| l
| |
| t -- . .
| |
| . . . .
| |
| .. . . . .
| |
| ..
| |
| ,. .
| |
| | |
| ,
| |
| R:actsr Operc.tcr Excmination
| |
| l
| |
| *
| |
| l 45. Giv:n the following conditions:
| |
| The Automatic Depressurization System (ADS) Manual initiation Channel "B"
| |
| and "F".pushb.uttons (S6B and S6F) have been armed.and depressed
| |
| l .
| |
| *
| |
| There is no Safety Relief Valve response ,
| |
| ~
| |
| ;
| |
| L Which of the following "B" Division electrical bus failures caused this system response?
| |
| l a. A loss of 120 VAC Bus 1BJ481
| |
| b. Aloss of 250 VDC Bus 10D261
| |
| c. A loss of 125 VDC Bus 1BD417
| |
| d. A loss of 480 VAC Bus 108420
| |
| l
| |
| 46. Given the following conditions:
| |
| . .
| |
| . .
| |
| L
| |
| The plant has been operating at 100% power for several weeks
| |
| '
| |
| All systems are operating 'as designed
| |
| Which of the following is the reason'why periodic riitrogen makeup to the drywell is required
| |
| for the given conditions?
| |
| ) a. Due to leaks from drywell air operated equipment.
| |
| ! b. Due to PCIG normal system leakage.
| |
| c. Due to normal periodic cycling of the Torus - Drywell Vacuum Breakers.
| |
| d. Due to normal drywell air inleakage.
| |
| l
| |
| :
| |
| l
| |
| l. )
| |
| !
| |
| I
| |
| .
| |
| l
| |
| Page 21 of 45
| |
| l
| |
| | |
| .
| |
| R0 actor Operatcr Examinatien
| |
| ~
| |
| 47. - Given the following conditions:
| |
| The plant had been operating at 75% power i
| |
| .
| |
| A loss of main condenser vacuum caused a complete Main Steam isolation '
| |
| Valve (MSIV) closure l
| |
| Vacuum has been reestablished and is currently 15" Hg absolute
| |
| .
| |
| '
| |
| Which of the following conditions is REQUIRED in order to reset the NSSSS MSIV isolation
| |
| logic?
| |
| a. The Reactor Mode Switch must be in " Shutdown".
| |
| b. : The Main Condenser Low Vacuum Bypass Switches must be in " Bypass".
| |
| c. The MSIV control switches must be in "Close"
| |
| d. The Turbine Stop Valves must be closed.
| |
| - -
| |
| .
| |
| 48. Which of the following conditions would preven.t.. opening the RHR "B" Loop Inboard and
| |
| .
| |
| '
| |
| Outboard Drywell Spray Valves (F021B and F016B) following a LOCA?
| |
| a. The LPCI Injection Valve (F0178) is not fully close'd.
| |
| b.- Less than 5 minutes have elapsed since the "B" RHR initiation occurred.
| |
| c. The RHR Full Flow Test Valve (F024B) is not fully closed.
| |
| d. Reactor water level is above -129 inches.
| |
| 49. Given the following conditions:
| |
| The Fuel Pool Cooling system is operating with one pump and heat exchanger
| |
| in service
| |
| The Fuel Pool Gates are installed
| |
| No makeup water sources are available
| |
| Which of the following is the expected effect on Spent Fuel Pool water level and cooling
| |
| capability if a leak develops on the common FPCC Pump Suction?
| |
| a'. Cooling capability and water level will be unchanged.
| |
| b. Cooling capability will be lost and water level will lower slightly and stabilize.
| |
| c. Cooling capability will be unchanged and water level will lower slightly and stabilize.
| |
| d. Cooling capability will be lost and water level will continuously lower.
| |
| Page 22 of 45
| |
| | |
| '
| |
| -
| |
| React:r Op:;rator Excmination
| |
| "
| |
| 50. Which of the following de:cribes how the main sterm line flow restrictors essist in maintaining
| |
| adequate core cooling for steam line break between the flow restrictors and the Main Steam
| |
| Isolation Valves?
| |
| a. They ensure'the total ~ inventory loss from the reactor. vessel maintains level above. the
| |
| top of active fuel until one division of low pressure ECCS is injecting.
| |
| b. They limit the' total inventory loss from the reactor vessel to maintain water level above
| |
| the top of active fuel for a minimum of 5 seconds. l
| |
| c. They ensure the total energy release rate to the Primary Containment does not result in
| |
| exceeding suppression chamber design pressure.
| |
| d. They limit the total inventory loss from the reactor vessel to maintain level above the top i
| |
| of active fuel until HPCI is at rated flow.
| |
| 51. Given the following conditions:
| |
| A reactor scram and Main Steam isolation Valve (MSIV) closure from 90% power
| |
| has occurred
| |
| The Safety Relief Valves (SRVs) are cycling to control pressure
| |
| Which of the following primary containment parameters indicates that one of the SRV tailpipe
| |
| vacuum breakers has failed open?
| |
| a. Suppression chamber pressure will go up each time the SRV cycles.
| |
| b. Suppression pool water temperatures will show rapid localized rises from the SRV
| |
| discharge flow bypassing the T-quenchers.
| |
| c. Drywell pressyre will go up each time the SRV cycles.
| |
| d. The Torus to Liywell ditarential pressure will rise each time the SRV opens.
| |
| 52. Which of'the following plant systems must be in operation to support the Main Steam
| |
| Isolation Valve (MSIV) Seal System.
| |
| a. Primary Containment Instrument Gas (PClG)
| |
| b.125 VDC Electrical Distribution
| |
| c. NUMAC Leak Detection System
| |
| d. Process Radiation Monitoring System
| |
| I
| |
| .
| |
| Page 23 of 45
| |
| | |
| _-
| |
| ,
| |
| R:actsr Operat:r Examinatien
| |
| "
| |
| 53. Giv;n the following conditions: >
| |
| The plant is operating at 70% power
| |
| The "B" EHC Pressure Regulator is tagged out of service
| |
| '
| |
| . Unknown to the' operator, the "A" EHC Pressure Regulator out'put signal is
| |
| '
| |
| failed "as is"
| |
| Which of the following would be the expected response of the Turbine Control Valves and
| |
| Turbine Bypass Valves as the Nuclear Control Operator (RO) begins to raise power using
| |
| recirculation flow for the given conditions? (Figure attached)
| |
| a. - The Turbine Control Valves will close
| |
| - The Turbine Bypass Valves will open
| |
| b. - The Turbine Control Valves will close
| |
| - The Turbine Bypass Valves will not move
| |
| c. - The Turbine Control Valves will not move
| |
| - The Turbine Bypass valve will not move -
| |
| '
| |
| -
| |
| d. - The Turbine Control Valves will not move
| |
| - The Turbine Bypass Valves will open
| |
| 54. Due to a main turbine vibration problem with a generator load of 110 MWe, a successful
| |
| manual turbine trip is performed.
| |
| _.
| |
| Which of the following describes when the operator is REQUIRED to open the generator
| |
| ;
| |
| Output Breakers for the given conditions? (Assume they have not already tripped on reverse
| |
| power.)
| |
| a. Immediately
| |
| 'b. Within 15 seconds of the turbine trip
| |
| c. Within 60 seconds of the turbine trip
| |
| d. Within 90 seconds of the turbine trip
| |
| l
| |
| l
| |
| Page 24 of 45
| |
| | |
| ..
| |
| ,
| |
| .,
| |
| Rcactor Optrator Examination
| |
| i 55. Given the following initial conditions:
| |
| The plant is operating at 25% power performing a plant startup
| |
| All plant systems are operating as designed
| |
| The "A" Reactor Feedwater Pump is in service in auto at approximateiy 3850 rpm
| |
| Following a plant transient the following conditions exist:
| |
| The reactor failed to scram when required
| |
| Reactor power is 14% and reactor pressure is 1105 psig
| |
| i
| |
| L. The Nuclear Control Operator (RO) notes that the "A" RFP speed has slowed
| |
| - to less than.1000 rpm
| |
| The RFP TURBINE AUTO XFR TO MANUAL (B3-F3) annunciator is in alarm
| |
| L Which of the following describes the reason for the "A" RFP speed reduction?
| |
| ,
| |
| a. The "A" RFP is responding properly to a Redundant Reactivity Control System runback.
| |
| ,
| |
| ,
| |
| b. The "A" RFP is responding to the S'etpoint Setdown feature of Digital Feedwater Control
| |
| l calling for a lower level,
| |
| c. The "A" RFP is responding to a' Control Signal Failure..
| |
| d. The "A" RFP is responding to a loss of one Primary Condensate Pump and one
| |
| Secondary Condensate Pump.
| |
| 56. Given the following conditions:
| |
| ,
| |
| '- A loss of off-site power (LOP) has occurred from 75% power
| |
| Within 10 seconds a loss of coolant accident (LOCA) occurs
| |
| l
| |
| Which of the following is the expected response of the LOP and LOCA sequencers?
| |
| L a. As soon as power is restored to the buses, the LOCA sequencer will control the
| |
| restoration of allloads.
| |
| b. The LOCA sequencer will begin to sequence until the diesel generator output breakers
| |
| ' close, then the LOP sequencer will complete load restoration.
| |
| l c. As soon as power is restored the buses, the LOP sequencer will control the restoration
| |
| '
| |
| of allloads.
| |
| d. The LOP sequencer will begin to sequence until the diesel generator output breakers
| |
| ' close, then the LOCA sequencer will complete load restoration.
| |
| ' Page 25 of 45
| |
| [
| |
| l
| |
| | |
| %
| |
| R:act:r Op ratar Examinatien
| |
| ..
| |
| 57. Given the following conditions:
| |
| The "B" Emergency Diesel Generator (EDG) had started following a valid
| |
| LQCA signal .
| |
| Some time later the' EDG was shutdown using the local Emergency Stop pushbuttons
| |
| due to fluctuating oil pressure
| |
| a Concurrent with stopping the EDG, the 10A402 bus lost power
| |
| Which of the following describes the actions, if any, regarding resetting the Engine Shutdown
| |
| Relay (ESR) and the (86R) Lockout Relay to restart the "B" EDG and reenergize the 10A402
| |
| bus?
| |
| a. ESR must be reset
| |
| (86R) Lockout Relay reset is not required
| |
| b. ESR mest be reset
| |
| (86R) Loc;*out Relay mus'. be reset
| |
| '' c. ESR reset is i ?t required
| |
| (86R) Lockout Relay .%et is not required
| |
| d. ESR reset is not required
| |
| .(86R) Lockout Relay must be reset
| |
| 58. Which of the following parameter changes indicate the moisture content of charcoal adsorber
| |
| bed of the Gaseous Radwaste System (GRW)is rising?
| |
| a. GRW post-treatment radiation level due to Krypton is rising.
| |
| b. GRW charcoal adsorber bed temperature is lowering.
| |
| c. GRW post-treatment radiation level due to lodine is rising.
| |
| d. GRW charcoal adsorber bed hydrogen concentration is lowering.
| |
| l
| |
| l
| |
| Page 26 of 45
| |
| .
| |
| | |
| .,
| |
| R: actor Operatsr Excminitlen
| |
| ' 59. Given the following conditions:
| |
| .
| |
| The plant has been operating at 100% power for several weeks
| |
| * Main Steam Line (MSL) radiation levels have been averaging 80 mrem but are now '
| |
| '
| |
| slowly trending upwards .
| |
| Chemistry has verified the higher radiation readings are due to failed fuel
| |
| What are the immediate Operator Actions required for the given conditions?
| |
| a. Place additional Condensate Domineralizers in service if possible.
| |
| b. Scram the reactor and close the Main Steam Isolation Valves when MSL levels are
| |
| greater than 120 mrem.
| |
| c. Direct Reactor Water Cleanup flow to the main condenser to reduce coolant activity.
| |
| d. Reduce reactor power to maintain MSL radiation levels less than 120 mrom.
| |
| 0- *
| |
| 60. Which of the following is the basis for raising the Main Steam Line (MSL) radiation monitor
| |
| setpoints when the Hydrogen Water Chemistry injection (HWCl) system is placed in service?
| |
| a. The setpoint adjustment ensures the higher (approximately two times) background
| |
| radiation does not mask a true fuel element failure.
| |
| b. The setpoint adjustment prevents unnecessary alarms / protective actions from the higher
| |
| (approximately two times) background radiation.
| |
| c. The setpoint adjustment ensures the higher (approximately ten times) background -
| |
| radiation does not mask a true fuel element failure.
| |
| d. The setpoint adjustment prevents unnecessary alarms / protective actions from the higher
| |
| (approximately ten times) background radiation.
| |
| Page 27 of 45
| |
| :
| |
| | |
| R: actor Operater Examination
| |
| ..
| |
| 61. Given the following conditions:
| |
| * A valid EDG room high temperature condition has just occurred
| |
| The Diesel Generator Room Carbon Dioxide Fire protection. system is aligned ~
| |
| ~ fo'r' automatic operation
| |
| Which of the following describes how the Diesel Generator Room Carbon Dioxide Fire
| |
| protection system responds?
| |
| a. A discharge alarm occurs, CO2 with a wintergreen scent is discharged into the room
| |
| immediately.
| |
| b. A pre-discharge alarm is activated and a wintergreen scent is discharged into the room.
| |
| After a time delay, CO2 is discharged into the room.
| |
| c. A pre-discharge alarm is activated. No CO2 is discharged into the room until a valid
| |
| smoke detector alarm is received.
| |
| d. A pre-discharge alarm is activated. After a time delay CO2 with a wintergreen scent is
| |
| -
| |
| discharged into the room.
| |
| 62. Given the following conditions:
| |
| The plant is operating at 50% power
| |
| . All systems are operating normally
| |
| . One Reactor Building Ventilation System (RBVS) Exhaust Fan discharge damper
| |
| has failed to the full "open" position with the fan running
| |
| No other RBVS components have changed
| |
| .
| |
| Which of the following describes how this will affect the initiation of the Emergency Core
| |
| Cooling Systems (ECCS) and the reason for this?
| |
| a. ECCS will initiate after it is required because the failed damper lowers Reactor Building
| |
| pressure resulting in a lower indicated drywell pressure.
| |
| b. ECCS will initiate before it is required because the failed damper raises Reactor
| |
| ~ Building pressure resulting in a higher indicated drywell pressure.
| |
| c. ECCS will initiate after it is required because the failed damper raises Reactor Building
| |
| pressure resulting in a lower indicated drywell pressure.
| |
| d. ECCS will initiate before it is required because the failed damper lowers Reactor
| |
| Building pressure resulting in a higher indicated drywell pressure.
| |
| Page 28 of 45
| |
| ____ -__ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
| |
| | |
| -
| |
| , 1
| |
| !
| |
| ..
| |
| Reactor Op::rator Examination
| |
| 63. Given the following conditions:
| |
| I
| |
| The plant is operating at 40% power
| |
| . .The Jet Pump operability surveillance indicates that one jet pump has fai. led
| |
| ,
| |
| Technical Specifications require the plant to be in hot shutdown within 12 hours
| |
| Which of the following describes why such a severe restriction placed on continued operation
| |
| for the given conditions?
| |
| a. A jet pump failure at this low power level will significantly affect the core flows and result l
| |
| !
| |
| in unacceptable thermal limits (MCPR).
| |
| b. A jet pump failure may limit reactor water level restoration capability during the reflood
| |
| portion of a Loss Of Coolant Accident.
| |
| c. A jet pump failure combined with the flow restricting orifices may adversely affect core j
| |
| flow to the higher power fuel bundles.
| |
| i
| |
| d. A jet pump failure results in less conservative protective action setpoints for
| |
| ~ ~
| |
| instrumentation using recirculation loop flow as an input signalf ~
| |
| l
| |
| 64. Which of the following is the expected status of the Control Area Ventilation after a valid high '
| |
| radiation condition at the Control Area Ventilation air intake occurs?
| |
| The Control Room Emergency Filtration (CREF) units are processing: ,
| |
| a. air entering the control room as well as recirculated air and are maintaining a slight
| |
| negative pressure.-
| |
| b. air entering the control room as well as recirculated air and are maintaining a slight
| |
| positive pressure.
| |
| c. only the current control room atmosphere and are maintaining a slight negative pressure.
| |
| d.' only the current control room atmosphere and are maintaining a slight positive pressure.
| |
| Page 29 of 45
| |
| | |
| -
| |
| R:act:r Operater Examination
| |
| ..
| |
| ' 65. Given the following conditions:
| |
| . The "A" Recirculation Pump has tripped
| |
| . The "A" Recirculation Pump discharge valve is open
| |
| * RECIRC LOOP A JET PUMP FLOW (TOTAL)iridicates 2 mlbm/hr
| |
| RECIRC LOOP B JET PUMP FLOW (TOTAL) indicates 35 mlbm/hr
| |
| . RECIRC PMP B FLOW indicates 24,000 gpm
| |
| . Recire pump "B" speed is 49%
| |
| Which of the following would be expected values for total JET PUMP FLOW (the flow
| |
| recorder) and actual core flow?
| |
| a. Flow recorder - 33 mlbm/hr, Actual core flow - 33 mlbm/hr
| |
| b. Flow recorder - 33 mlbm/hr, Actual core flow - 37 mlbm/hr
| |
| c. Flow recorder - 37 mlbm/hr, Actual core flow - 33 mlbm,hr
| |
| d. Flow recorder - 37 mlbm/hr, Actual . core flow - 37 mlbm/hr
| |
| 66. Given the following conditions:
| |
| . The plant is operating at 90% power
| |
| . All main turbine sealing steam normal and backup supplies have been lost
| |
| . There is no time estimate for repair / restoration
| |
| Which of the following are the immediate operator actions for the given conditions?
| |
| a. Reduce power as necessary to maintain condenser pressure less than 5.0" HGA.
| |
| b. Reduce recirculation flow to minimum, unload and trip the main turbine.
| |
| c. Reduce power as necessary to maintain adequate self-sealing steam to the main turbine
| |
| seals.
| |
| d. Reduce recirculation flow to maintain power less than 25% (Bypass Valve capacity).
| |
| Page 30 of 45
| |
| i
| |
| | |
| -
| |
| 1
| |
| .. Rcacter Operator Examination
| |
| 67. During a loss of off-site power the operator is cautioned not to acknowledge the flashing
| |
| ' Trip" pushbuttons for the 4.16 KV Vital 1E Bus infeed breakers.
| |
| .
| |
| .Which of the following will occur if these pushbuttons are pressed? ,
| |
| a. That bus' feeder breaker will attempt to close until the anti-pump feature causes it to trip
| |
| open and remain open.
| |
| b. The Diesel Generator associated with that bus, if running, will trip and its output breaker
| |
| will open.
| |
| c. That bus' alternate feeder breaker will trip open and then immediately reclose when the
| |
| pushbutton is released
| |
| d. The Diesel Generator associated with that bus will not load.
| |
| 68. Given the following conditions:
| |
| A plant startup is in progress with the Reactor Mode Switch in "Run"
| |
| The Main Turbine is reset and is at 950 rpm accelerating to 1800 rpm
| |
| A loss of 125 VDC power from distribution panel 1CD318 to the EHC control
| |
| logic occurs
| |
| I
| |
| Which of the following is the expected plant response?
| |
| a. Main turbine trips,
| |
| b. Main turbine startup would continue at the selected acceleration rate.
| |
| c. Main turbine speed will remain constant at 950 rpm.
| |
| d. Main turbine control valves throttle closed due to a loss of the speed reference signal.
| |
| ,
| |
| !
| |
| l
| |
| Page 31 of 45
| |
| | |
| '
| |
| Reactcr Operatar Excminttinn
| |
| --
| |
| 69. Giv:n the following conditions:
| |
| . The plant is operating at 20% power
| |
| . A main generator load reject has just occurred .
| |
| . The power / load unbalance circuit tripped unexpectedly during the load reject
| |
| .
| |
| Which of the following is the expected response of the Turbine Control Valves and the
| |
| Reactor Protection System (RPS) for the given conditions?
| |
| a. - The Turbine Control Valves throttle closed
| |
| - RPS does not trip
| |
| b. - The Turbine Control Valves fast close
| |
| - RPS trips
| |
| c. - The Turbine Control Valves throttle closed
| |
| - RPS trips
| |
| d. - The Turbine Control Valves fast close
| |
| - RPS does not trip
| |
| 70. Which of the following describes when the Main Turbine is required to be tripped following a
| |
| reactor scram?
| |
| a. At 50 MWe lowering
| |
| b. At 25 MWe lowering
| |
| c. At 0 MWe
| |
| d. At 50 MWe rising (reverse power)
| |
| 71. During a failure-to-scram condition, which of the following is the criteria used to determine if
| |
| HC.OP-EO.ZZ-0100(Q), " Reactor Scram", should be exited and HC.OP-EO.ZZ-0101(Q),
| |
| " Reactor / Pressure Vessel (RPV) Control", entered?
| |
| L a. Reactor period on SRM Period meters is stable at -80 seconds
| |
| b. All APRM "downscale" lights are not illuminated
| |
| c. All four RPS logic channels are deenergized
| |
| d. All control rods are inserted to or beyond Notch "02"
| |
| f
| |
| L
| |
| Page 32 of 45
| |
| - _ _ _ - - _ - _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ .
| |
| | |
| /
| |
| R:acter Opsrater Examination
| |
| 72. Following a reactor scram and Main Steam isolation Valve closure, reactor pressure reaches
| |
| 1050 psig causing the "H" and "P" Safety Relief Valves (SRV) to open.
| |
| Which of the following lists the operating setpoints for subsequent openings of the "P." SRV7
| |
| a. SRV "P" opens at 1047 psig and closes at 935 psig.
| |
| b. SRV "P" opens at 1047 psig and closes at 905 psig.
| |
| c. SRV "P" opens at 1017 psig and closes at 935 psig.
| |
| d. SRV "P" opens at 1017 psig and closes at 905 psig.
| |
| 73. With the plant at 100% power a severe overfeeding transient is occurring. Water level is +50
| |
| inches and rising rapidly.
| |
| Which of the following reactor water levels require termination of all feed to the reactor,
| |
| closing the MSIVs and a reactor scram assuming none of these actions have occurred?
| |
| a. +54 inches
| |
| b. +65 inches l
| |
| c. +90 inches
| |
| d. +118 inches
| |
| 1
| |
| 74. Given the following conditions:
| |
| . The plant is operating at 80% power
| |
| . All three Feedwater Pumps are in service
| |
| Feedwater Level Control is in " Automatic - Three Element" control
| |
| . Narrow Range level"A"is reading 34 inches
| |
| . Narrow Range level"B"is reading 36.5 inches
| |
| Narrow Range level "C" is reading 35.0 inches
| |
| Which of the following would be the expected response of the Feed Water Level Control
| |
| System and reactor water level if Narrow Range level "B" failed to the low end of the range?
| |
| a. It would transfer to Single Element Control and level would remain unchanged.
| |
| b. It would remain in Three Element Control and level would remain unchanged.
| |
| c. It would transfer to Single Element Control and would raise level by approximately 1.5 l
| |
| inches. i
| |
| d. It would remain in Three Element Control and would raise level by approximately 1.0
| |
| inches. l
| |
| Page 33 of 45
| |
| | |
| s
| |
| R: actor Operator Excminatian
| |
| ''
| |
| 75. Given the following conditions:
| |
| The plant is operating at 95% power
| |
| All Drywell Cooling Chilled Water pumps have tripped
| |
| Drywell pressure is rising
| |
| HC.OP-AB.ZZ-0201,"Drywell High Pressure / Loss Of Drywell Cooling", has been
| |
| entered and the Reactor Auxiliary Cooling System (RACS) is being aligned to supply
| |
| backup cooling to the Chilled Water System
| |
| Which of the following describes the effect of failing to close the Chilled Water Isolation
| |
| Supply and Return Valves (HV9532-2 and HV9532-1) before the transfer to RACS7
| |
| a. The RACS Pump automatic start permissives will be bypassed until the valves are
| |
| closed.
| |
| b. The RACS valves will not automatically sequence open to supply Chilled Water should
| |
| a loss of off-site power occur.
| |
| c. Chilled Water system flow will divert back into the RACS system overflowing the RACS
| |
| head tank.
| |
| d. RACS system flow will divert back into the Chilled Water system overflowing the Chilled
| |
| Water head tank.
| |
| 76. During a loss of coolant accident the following conditions exist:
| |
| Reactor pressure is 125 psig
| |
| Drywell temperature is 325 'F
| |
| Which of the following describes the accuracy and tr ding capabilities of wide range reactor
| |
| water level indication for the given conditions?
| |
| a. They are not providing accurate re or water level or level trend information.
| |
| b. They are providing accurate ctor water level but level trend is not reliable.
| |
| c. They are providing a te reactor water level and level trend information.
| |
| d. They are not prov' ng accurate reactor water evel but level trend is reliable.
| |
| '
| |
| .
| |
| Page 34 of 45
| |
| | |
| a
| |
| Reactor Operator Examinaticn
| |
| 77. Given the following conditions:
| |
| The plant is operating at 95% power
| |
| Suppression pool temperature is 92 'F
| |
| At 0915, Safety Relief Valve (SRV) "G" opened
| |
| After several cycles of the SRV Open and Close pushbuttons, the operator notes
| |
| that talipipe temperature for the SRV is stable at 305 *F and NO other plant parameters
| |
| have changed
| |
| Which of the following describes the limitations on continued reactor operation for the given
| |
| conditions?
| |
| a. Reactor operation may continue until pressure set is reduced to less than 850 psig.
| |
| b. Reactor operation may continue until suppression pool temperature reaches 120 *F.
| |
| c. Reactor operation may continue indefinitely.
| |
| d. Reactor operation may continue until 0917.
| |
| 78. Given the following conditions:
| |
| Reactor power is 82%
| |
| HPCI is in operation for a surveillance
| |
| The "B" loop of RHR is in Suppression Pool Cooling
| |
| Suppression pool temperature is 103 'F when the runni R pump tripped
| |
| HPCI was secured
| |
| Subsequently, suppression pool temperature in sed to 106 'F
| |
| Which of the following lists the suppression temperatures requiring entry into HC.OP-
| |
| EO.ZZ-0102, Primary Containment Cont AND entry into the LCO actions for Tech Spec
| |
| 3.6.2.17
| |
| a. EO-0102 - 95 'F
| |
| TS 3.6.2.1 - 95 *
| |
| b. EO-0102 5 'F l
| |
| - 105 'F
| |
| TS 3.6.2.
| |
| c. EO 02 - 105 'F- I t
| |
| 3
| |
| q@.t .
| |
| '
| |
| 3.6.2.1 - 95 'F g{(O L
| |
| d. EO 0102 - 105 'F
| |
| TS 3.6.2.1 - 105 *F
| |
| # #'
| |
| gg
| |
| Dele 7td 5'ce os Af d FM'" T
| |
| ift 3 -5-1
| |
| Page 35 of 45
| |
| | |
| ,
| |
| Reactar Operater Examination
| |
| ..
| |
| 79. Given the following conditions:
| |
| The plant is at 75% power
| |
| Control rod 22-27 is being withdrawn one notch to Notch "22"
| |
| Which of the following is the required immediate operator action if a control rod drift alarm is
| |
| received and the operator notes control rod 22-27 is continuing to move out and power is
| |
| rising?
| |
| a. Apply a continuous insert signal to control rod 22-27.
| |
| b. Place the Rod Select key lock switch to "Off"(de-select the rod).
| |
| c. Direct the local operator to perform a single rod scram on control rod 22-27.
| |
| d. Runback recirculation flow and insert control rods to reduce power.
| |
| 80. Given the following conditions:
| |
| The plant is operating at 100% power
| |
| A feedwater heater trip has resulted in a feedwater temperature of 385 *F
| |
| No nperator actions have been taken
| |
| Which of the following is the operational concern for the given conditions?
| |
| a. Entry into the Exit Region of the Power-To-Flow Map.
| |
| b. Violation of the Hope Creek Operating License.
| |
| c. Immediate thermal hydraulic instabilities.
| |
| d. Recirculation Pump damage.
| |
| l
| |
| l
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| l- Page 36 of 45
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| l
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| __ - - _ _ _ _ - _ _ _ _ - _ _ _ _ - _ _ _
| |
| | |
| - _ _ ___ _ _ _ _ _ _ .
| |
| React 2r Operatar Examination
| |
| ..
| |
| 81. Following a reactor scram all rods are at position "00" except one that is at position "24."
| |
| Which of the following describes the capability of the reactor to remain shutdown?
| |
| a. Control rods are inserted to beyond the Maximum Suberitical Banked Withdrawal limit,
| |
| therefore the reactor will remain shutdown under all conditions.
| |
| b. Control rods are not inserted to beyond the Maximum Suberitical Banked Withdrawal
| |
| limit, therefore 11 cannot be assured the reactor will remain shutdown under all
| |
| conditions.
| |
| c. Design basis shutdown margin is met, therefore the reactor will remain shutdown under
| |
| all conditions.
| |
| d. Design basis shutdown margin is not met, therefore it cannot be assured that the reactor
| |
| will remain shutdown under all conditions.
| |
| 82. Given the following conditions:
| |
| The Control Room has been abandoned in accordance with HC.OP-AB.ZZ-0130(O),
| |
| " Control Room Evacuation"
| |
| * Control has been established at the Remote Shutdown Panelin accordance with
| |
| HC.OP-lO.ZZ-0008(Q), " Shutdown From Outside Control Room"
| |
| * RCIC is operating maintaining reactor water level at +35 inches
| |
| Safety Relief Valves (SRV) are being used to cooldown
| |
| Condensate Storage Tank (CST) level is 135,000 gallons
| |
| * The Condensate System is not available
| |
| Which of the following is correct for the given conditions?
| |
| a. RCIC is operated without overspeed protection.
| |
| b. Insufficient CST inventory is available to allow the cooldown to clear the shutdown
| |
| cooling interlocks.
| |
| c. The RCIC Gland Seal Condenser Condensate Pump must be manually operated.
| |
| d. SRVs cannot be operated in a rotation that will evenly distribute heat to the Suppression
| |
| Chamber.
| |
| l
| |
| l
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| i
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| Page 37 of 45
| |
| i
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| | |
| %
| |
| R act:r Op: rat 2r Examinatinn
| |
| "
| |
| 83. Which of the following describes the effect of failing to restart the Turbine Building Ventilrtion
| |
| System if it trips while operating in HC.OP-EO.ZZ-0104(Q)-FC, " Radioactive Release
| |
| Control"?
| |
| a. The Turbine Building will go to a slightly negative pressure.
| |
| b. The total off-site release calculations will not be accurate.
| |
| c. The Turbine Building releases will be monitored but not treated.
| |
| d. The total off-site release will be higher.
| |
| I
| |
| 84. A loss of Reactor Auxiliary Cooling System (RACS) has occurred.
| |
| Which of the following is the MAXIMUM time allowed before a reactor scram is required?
| |
| a. An immediate scram is required
| |
| b. One (1) minute
| |
| c. Ten (10) minutes
| |
| d. Twenty (20) minutes
| |
| 85. Given the following conditions:
| |
| * A loss of coolant accident has occurred
| |
| The Reactor Auxiliaries Cooling System (RACS) has been restored
| |
| Which of the following describes the availability / response of the Emergency Instrument Air
| |
| Compressor (EIAC) for these conditions should instrument air header pressure begin
| |
| lowering?
| |
| ! a. The EIAC is not available until the LOCA signal is cleared, reset and the 1E breaker is
| |
| closed.
| |
| b. The EIAC will automatically start on instrument air header pressure less than 85 psig.
| |
| c. The EIAC is not available until the Non-1E breaker is closed and instrument air pressure
| |
| is less than 85 psig.
| |
| d. The EIAC will not automatically start but may be started manually from the Control
| |
| Room or locally.
| |
| Page 38 of 45
| |
| .
| |
| | |
| ,
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| Reactar Operator Examination
| |
| ..
| |
| 86. Which of the following describes the reason control rods insert during a loss of instrument air?
| |
| )
| |
| a. A flowpath is opened to the bottom of the drive mechanism operating piston allowing l
| |
| reactor pressure to drift the rod in. ]
| |
| b. Drive water flow and pressure both rise enough to drift the control rod in, similar to a l
| |
| I
| |
| normal insertion.
| |
| c. A flowpath is opened from the top of the drive mechanism operating piston allowing q
| |
| accumulator pressure to drift the rod in. J
| |
| d. The normal scram flowpath to and from the drive mechanism operating piston is opened,
| |
| allowing accumulator and reactor pressure to drift the rod in.
| |
| 87. Given the following conditions:
| |
| The plant is operating at 20% power following a refueling outage
| |
| An error during a surveillance has resulted in a Group 10 (Drywell Chilled Watar)
| |
| isolation signal
| |
| . The isolation goes to completion (all valves are closed)
| |
| Drywell pressure is 1.25 psig and rising slowly
| |
| Which of the following are the required immediate operator actions for the given conditions?
| |
| a. Lineup and commence venting the drywell.
| |
| b. Secure drywell inerting.
| |
| c. Place the Reactor Mode Switch in " Shutdown". i
| |
| d. Align RACS to supply cooling to Drywell Chilled Water.
| |
| 88. Following a loss of shutdown cooling, decay heat removal is being transferred to the
| |
| Alternate Shutdown Cooling lineup (Injection to the reactor returning to the suppression pool
| |
| via open Safety Relief Valves).
| |
| Which of the following indications is utilized to monitor the margin to Op Con 3 entry in this
| |
| lineup?
| |
| a: Safety Relief Valve tailpipe temperatures
| |
| b. Suppression pool temperatures
| |
| c. Reactor vessel skin temperatures -
| |
| d. Local suction temperatures on the running low pressure ECCS pumps
| |
| ,
| |
| Page 39 of 45
| |
| | |
| ,
| |
| R3act r Operator Examination
| |
| ..
| |
| 89. Which of the following describes the conditions requiring the Reactor Mode Switch to be
| |
| placed in " Shutdown" on a sustained loss of Control Rod Drive charging water header
| |
| pressure (<900 psig) with reactor pressure at 650 psig?
| |
| a. Within 20 minutes of determining more than one CRD accumulator is inoperable and at
| |
| least one of those inoperable accumulators is associated with a withdrawn control rod.
| |
| b. Within 20 minutes of determining any CRD accumulator is inoperable and the
| |
| inoperable accumulator is associated with a withdrawn control rod.
| |
| c. Immediately upon determining more than one CRD accumulator is inoperable and all the
| |
| inoperable accumulators are associated with fully inserted control rods,
| |
| d. Immediately upon determining any CRD accumulator is inoperable and the inoperable
| |
| accumulator is associated with a withdrawn control rod.
| |
| 90. Given the following conditions:
| |
| The plant is shutdown for refueling
| |
| The Reactor Protection System shorting links have been removed
| |
| A fuel bundle is being moved from the fuel pool to core.
| |
| If SRM "C" fails "downscale", which of the following are the required immediate actions?
| |
| a. Verify a control rod withdrawal block is received. Terminate fuel movement.
| |
| b. Verify a full scram and a control rod withdrawal block is received. Terminate fuel
| |
| movement.
| |
| c. Verify a control rod withdrawal block is received. Fuel movement is required to be
| |
| terminated ONLY if the fuel bundle is to be placed in the quadrant monitored by SRM
| |
| "C ."
| |
| I d. Verify a full scram and control rod withdrawal block is received. Fuel movement is
| |
| required to be terminated ONLY if the fuel bundle is to be placed in the quadrant
| |
| monitored by SRM "C."
| |
| l
| |
| Page 40 of 45
| |
| l
| |
| l
| |
| | |
| ,
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| R:;cctor Operator Ex;minati:n
| |
| ..
| |
| 91. Given the following conditions:
| |
| * A large break loss of coolant accident has occurred
| |
| * Drywell pressure reached a maximum of 22 psig
| |
| * Suppression chamber sprays have NOT been placed in service
| |
| * Drywell sprays are in service
| |
| * Drywell pressure is 4 psig and slowly lowering
| |
| Which of the following is the expected positions of the Torus-to-Drywell Vacuum Breakers and
| |
| the Reactor Building 4o-Torus Vacuum Breakers for the given conditions?
| |
| a. - The Torus-to-Drywell Vacuum Breakers are open
| |
| - The Reactor Building-to-Torus Vacuum Breakers are open
| |
| b. - The Torus-to-Drywell Vacuum Breakers are open
| |
| - The Reactor Building 4o-Torus Vacuum Breakers are closed
| |
| c. - The Torus-to-Drywell Vacuum Breakers are closed
| |
| - The Reactor Building 4o-Torus Vacuum Breakers are closed
| |
| d. - The Torus-to-Drywell Vacuum Breakers are closed
| |
| - The Reactor Buildiag-to-Torus Vacuum Breakers are open
| |
| 92. Following a reactor scram with a Main Steam isolation Valve Closure, the plant is being
| |
| depressurized using the Safety Relief Valves (SRV).
| |
| Which of the following is the reason why the depressurization should be accomplished with
| |
| " sustained" SRV openings if the pneumatic supply (PClG and instrument air) is lost to the
| |
| SRVs?
| |
| a. This prevents exceeding the 100*F/ hour cooldown limit during the depressurization
| |
| while conserving the SRV pneumatic supply.
| |
| b. This ensures the available SRV pneumatic supply is sufficient to cooldown to less than
| |
| the shutdown cooling interlocks.
| |
| c. This directs depressurization without regard to the Technical Specification cooldown .
| |
| i
| |
| limits before the depleted pneumatic supply results in loss of SRV control.
| |
| d. This ensures the SRV accumulator pneumatic s@ ply is available and adequate for later
| |
| use if the Emergency Operating Procedures require Emergency Depressurization.
| |
| Page 41 of 45
| |
| | |
| s
| |
| Reactor Operator Examinatisn .,
| |
| 93. HPCI and RCIC both started and are injecting in response to a valid low reactor water level. i
| |
| I
| |
| Current plant conditions are as follows:
| |
| + Reactor water level is +25 inches, steady
| |
| Reactor pressure is 845 psig, rising slowly
| |
| * Drywell pressure is 1.1 psig, steady
| |
| * RCIC has been aligned to Full Flow Recire operation (CST to CST) for pressure control
| |
| HPCI is injecting to the reactor for level control
| |
| After 10 minutes of operation a valid high suppression poollevelis received
| |
| Which of the following would be the expected response of RCIC if a valid high suppression
| |
| pool level is received for the given conditions?
| |
| a. RCIC will remain in Full Flow Recirculation.
| |
| b. RCIC will trip on high turbine exhaust pressure.
| |
| c. RCIC will trip on low suction pressure.
| |
| d. RCIC will operate on minimum flow.
| |
| 94. During high primary containment water level conditions, suppression pool water level
| |
| indications cannot be used.
| |
| Operation of which system will invalidate the alternate method used for determining primary
| |
| containment water level?
| |
| a. RCIC
| |
| b. Core Spray
| |
| ,
| |
| c. RHR
| |
| l
| |
| d. HPCI
| |
| 1
| |
| 1
| |
| Page 42 of 45
| |
| | |
| R:act:r Operator Examinati n
| |
| ..
| |
| 95. Given the following conditions:
| |
| A leak has occurred in the suppression pool
| |
| The reactor is shutdown
| |
| A cooldown is being performed using SRVs
| |
| The Heat Capacity Level Limit (HCLL) curve is being monitored
| |
| The " Action Required" area of the HCLL curve has been entered for several minutes
| |
| Which of the following is a possible effect of initiating an emergency depressurization with the
| |
| given conditions?
| |
| a. The suppression pool may exceed design temperature.
| |
| - b. Failure of the downcomer vent header joints due to " chugging."
| |
| c. The SRV Tailpipe Level Limit curve may be exceeded.
| |
| d. The capacity of the Torus to Drywell vacuum breakers will be exceeded.
| |
| I
| |
| 96. Following the runback of the Recirculation Pumps on a trip of a Primary Condensate Pump,
| |
| the operator may monitor the Source Range Monitoring (SRM) period meters for strong
| |
| deflections above and below" Infinity".
| |
| Under which of the following conditions may SRM period indications be considered accurate
| |
| indication of thermal hydraulic instabilities?
| |
| a. Only when the SRM detectors are fully withdrawn from the core,
| |
| b. Anytime, regardless of detector position, if the detectors are stationary. ,
| |
| c. Only when the SRM detectors are fully inserted into the core.
| |
| d. Anytime the SRM detectors are moving.
| |
| l
| |
| :
| |
| Page 43 of 45
| |
| | |
| Reacter Operater Excminction
| |
| "
| |
| 97. With the plant at power tha Main Starm/ R rctor Water Cinnup Aras Lc:k.Temperatura
| |
| High alarm was received and the RWCU system automatically isolated. The leak has been
| |
| determined to be in the RWCU Pipe Chase Room 4402.
| |
| Which of the following is NOT a required operator action for the.given conditions?
| |
| a. Notify Chemistry to close the Manual Sample Line isolation Valves P-RC-V9670 and 1-
| |
| RC-V006.
| |
| b. Observing the RWCU Inboard and Outboard Isolation Valves (F001 & F004) close.
| |
| c. Observing the Recirc Sample Line Isolation Valves (BB-SV-4310 and 4311)
| |
| automatically close.
| |
| d. Operate available Reactor Building ventilation fans consistent with plant conditions.
| |
| 98. Given the following conditions:
| |
| . The plant was operating at rated power when a steam line break occurred in the HPCI
| |
| room
| |
| . HPCI isolated due to high room temperatures
| |
| . RBVS exhaust radiation levels reached 1.0 E-2 microcuries/mi
| |
| Which of the following describes the ventilation system response for the given conditions?
| |
| a. RBVS remains in service
| |
| b. RBVS isolated,6 FRVS Recirc and 1 FRVS Vent Fans are in service
| |
| c. RBVS isolated,4 FRVS Recirc and 1 FRVS Vent Fans are in service
| |
| '
| |
| d. RBVS isolated,6 FRVS Recire and 2 FRVS Vent Fans are in service
| |
| 99. With the Reactor Building Ventilation in a normal lineup, the operator observes that Reactor
| |
| Building pressure is .10 inches of vacuum water gauge.
| |
| Which of the following is an immediate action to restore Reactor Building pressure to the
| |
| required pressure?
| |
| a. Place at least two FRVS units in service.
| |
| b. Secure a reactor building supply fan.
| |
| !
| |
| l c. Place an FRVS unit in service and increase FRVS flow rate to maximum.
| |
| d. Place the third Reactor Building Exhaust Fan in service.
| |
| Page 44 of 45
| |
| . . _
| |
| | |
| , 1
| |
| R: actor Operater Examination
| |
| .,
| |
| 100. Given the following conditions:
| |
| The reactor has scrammed from power
| |
| Pilot Scram Valve Group 1 solenoids, Logic "A" and Logic "B" did not deenergize
| |
| * The Scram Discharge Volume is currently full
| |
| ;
| |
| l Which of the following describes the difference between inserting control rods in accordance
| |
| '
| |
| with HC.OP-EO.ZZ 0303, " Individual Control Rod Scrams" and HC.OP-EO.ZZ-0302, "De-
| |
| energization Of Scram Solenoids"? I
| |
| l
| |
| a. EO-0302 requires resetting RPS and ARI, EO-0303 does not.
| |
| I
| |
| b. EO-0303 requires resetting RPS and ARI, EO-0302 does not. -
| |
| ,
| |
| I
| |
| c. EO-0303 does not isolate the Scram Discharge Volume, EO-0302 does.
| |
| '
| |
| d. EO-0302 results in a flowpath from the RPV to the Reactor Building Sump, EO-0303
| |
| does not.
| |
| !
| |
| ,
| |
| ,
| |
| ! Page 45 of 45
| |
| | |
| .
| |
| R::ct:r Operat:r An:wcr K y
| |
| ..
| |
| 1. b 2seoioiot . 26. d 20soo mio4
| |
| 2. a 2seotato2 27.)(a. -= .
| |
| see nrrm a e = "'Ys's ik 3*'~nf gQ
| |
| 3. c 2s4001010e 28. a 20eoooxto2 3/fgeg
| |
| 1. : : :c;; 29. d 20eooooot
| |
| DeteTed Set AmH6 e n~mi Sr 5 Afst'1-S*NE
| |
| 5. c 2seoici22 % b N!bY 30. d 20eootAso2
| |
| 6. c 2seotot2e 31. a 20eootA40s
| |
| 7. b 2emosaist 32. a 2stoooA20e
| |
| 8. b 2secto202 33, a 2tioooksoe
| |
| 9. c 2seoso213 34. d 212000Atos
| |
| 10. b 2sectoso4 35. d 212000K411
| |
| 11. c 2semic310 36. b 21sootAes
| |
| 12. b 204001 o412 37. d 21soo2xec4
| |
| 13. d 204001044e 38. c 21soasxeos
| |
| '14. c 201mtA204 39. d 21s m4A104
| |
| 15. a 201m1K40s 40. b 21soasxto4
| |
| 16. c 20too2A40s 41. d 21eoooA201
| |
| 17. a 20t m3A207 42. d 21eoooAsoi
| |
| 18. a 201ooskst4 43, a 217oooA4oi
| |
| 19. p b_ _.=_6325, D6 M 44. b 217000x201
| |
| f. _ m ,, a, s ,vf .
| |
| [tt M
| |
| 20. aced ac2001Aso2 45. c 21eom x201
| |
| sre.arrma w~., w sbr (6c s-r-9s
| |
| 21. c' 2020oixios MDb 46. b 223ooixtos
| |
| 22. b 202m2Atoi 47. c 223m2Ae3
| |
| 23. b 202m2xeo4 48. a 22eootxes
| |
| 24. d 203oooKee 49. b 233oooxso2
| |
| 25. c 20eoDK115 50. b 23emiot2e <
| |
| l
| |
| l Page 1 ,
| |
| )
| |
| l
| |
| l
| |
| | |
| '
| |
| l
| |
| Rrctor Operat:r Answ r K y I
| |
| l
| |
| ..
| |
| l
| |
| 51, c 23e m2xsoe 70. :M
| |
| QtsftA
| |
| a
| |
| ,x:
| |
| de
| |
| 4,,
| |
| [ yb hi' _
| |
| . ___
| |
| . .z y r -
| |
| i
| |
| i
| |
| ;- m -
| |
| -
| |
| w _.; -
| |
| 52. a - Aiot 77.-d en 2esoisAto2
| |
| erfr .-o w n
| |
| 3
| |
| ;
| |
| 53. c 241moK3o2 -g,,y m rym) ,6W M D;y ; gg; y y
| |
| -- --
| |
| __
| |
| 54. b 2emooto2 79.
| |
| ', dp 29 sot 4Ato2 _ ,_g,
| |
| ,,, . . . ,
| |
| l
| |
| 15e'ilZwed)
| |
| l g( auefe2 A 3[h
| |
| 55. a 2ssoo2A20e 80. b 29 sot 4cito j
| |
| i
| |
| 56. a 2e2001A3o4 81. c 2esotsA202 {
| |
| J
| |
| 57. b 2s4omxeos 82, c 2esotaAtos
| |
| 58 a 271omA40s 83. b 29 sot 7x302
| |
| 1
| |
| 59. d 272moA201 84. cpd. 29sotex202 , _ , , , , , , , , , - 3'
| |
| s:: .g g-~<
| |
| 60. b 272cooxsoi 85. a 29so19Atot
| |
| 61. d 2esomA4ot 86. d 29sotox201
| |
| 62. d 2somixsot 87, b 2sso20xios
| |
| l 1
| |
| i
| |
| 63. b 2sooo2x4oi 88. a 29so21Ato4
| |
| 64. b 29mosksot 89. d 29so22K2o7
| |
| 65. a 295oo1A203 90. a 295o23o232 !
| |
| 66. a 29soo2A1os 91. b 29so24A1ie
| |
| 67. d 29sm3Aiot 92. d 29so2sK102
| |
| 68, a 29soo4x203 93. d 29so29Ato4
| |
| 69. d 29smsx201 94. d 29so29A20s
| |
| 70. c 2ss008o449 95. a 29smonio3
| |
| ,
| |
| 71. b 29soo6Kto3 96. b 29so31A202
| |
| 29soo7x3o4 97. c 295032G44
| |
| l 72. a
| |
| 73. c 2ssmeat23 98. g6 29so34x102
| |
| 7 S*4
| |
| y
| |
| 74. d 2esooox202
| |
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| |
| 99. d srrnene&
| |
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| |
| 75. d 29sotox3o2 100. c 29so37x20s
| |
| Page.2
| |
| l
| |
| | |
| >
| |
| .
| |
| ..
| |
| l ATTACHMENT 3
| |
| V} PSE&G COMMENTS ON WillTTETJ EXAM
| |
| ,
| |
| - . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| #
| |
| l
| |
| O PSEG
| |
| ~
| |
| .
| |
| Public Service Bectric and Gas Company 244 Chestnut Street Salem. N.J. 08079 Phone 609/935-8560
| |
| l
| |
| ! Nuclear Training Center
| |
| l 1
| |
| l March 6,1998 i
| |
| ! !
| |
| NTC-98-3011
| |
| Mr. Don Florek l
| |
| Chief Examiner
| |
| Division of Reactor Safety
| |
| US Nuclear Regulatory Commission
| |
| 475 Allendale Road
| |
| King of Prussia, PA. 19406-1415
| |
| Dear Mr. Florek
| |
| HOPE CREEK SRO/RO EXAM COMMENTS
| |
| Attached please find our post-examination analysis comments and related backup information on
| |
| the following questions, from our recently conducted Hope Creek RO/SRO examination, Our
| |
| comments are on the page with the applicable question, and are broken into three (3) categories:
| |
| l
| |
| Exam Answer key corrections
| |
| _
| |
| . RO #19
| |
| . RO #27 / SRO #29
| |
| . RO #98 / SRO #96
| |
| Correct Alternate choice answers from oriainal answer key
| |
| . RO #20 / SRO #24
| |
| . SRO #69
| |
| . RO #76 / SRO #71
| |
| . RO #79
| |
| . RO #84 / SRO #79
| |
| Question Deletions
| |
| . RO # 04/ SRO #05
| |
| . RO #78 / SRO #73 '
| |
| *
| |
| . SRO #75
| |
| If you have any questions, comments or require any additional information, please contact Pete
| |
| Doran acting Nuclear Training Supervisor at 609-339-3816 or John Nichols Operations Training
| |
| Manager at 609-339-3769.
| |
| Sincerely,
| |
| , ./tv/L'
| |
| erome F. McMahon
| |
| Director- OA/ Nuclear Training /EP
| |
| c$cN
| |
| I"FOR NUCLE
| |
| TRAINING
| |
| b uwr J is in pur hands.
| |
| M 2169 34EV 4al2
| |
| | |
| $
| |
| EXAM ANSWER KEY CORRECTIONS ,,
| |
| .
| |
| EXAM QUESTION RO #19
| |
| Given the following conditions:
| |
| . The plant is operating at 75%
| |
| . Confirmed seal failures have occurred on the "B" Recirculation Pump
| |
| . The pump hasjust been tripped
| |
| Which of the following describes the order for the "B" Recirculation Pump valve manipulations that
| |
| must be followed in order to ensure the pump will be completely isolated,
| |
| s. Close the Discharge valve, isolate seal purge, isolate RWCU flow from the loop and close the
| |
| suction valve.
| |
| b. Isolate seal purge, close the suction valve, isolate the RWCU flow from the loop and close the
| |
| discharge valve
| |
| c. Close the suction valve, close the discharge valve, isolate seal purge, isolate RWCU flow from the
| |
| 100P-
| |
| d. Isolate seal purge, close the discharge valve, isolate the RWCU flow from the loop and close the
| |
| suction valve.
| |
| Ans: C
| |
| Ref HC.OP-AB.ZZ-0112, " Recirculation pump Trip", rev.13
| |
| LP - 0302-000.00H-000114-rev. 5
| |
| Obj. 3
| |
| 1. Based on pre-examination discussions and referenced procedures, the critical step sequence is
| |
| based on the discussion item 5.7 of HC.OP-AB.ZZ 4112, " Recirculation purnp Trip'(attached) and
| |
| precautions and limitations 3.1.2 of HC.OP-SO.BB-0002 ' Recirculation System Operation"
| |
| (attached)
| |
| 2. The suction valve must be closed before the discharge valve, and the seal purge must be
| |
| closed prior to pump isolation. This makes 'b' the only correct answer.
| |
| Recommendation:
| |
| Change answer key to choice "b" as correct answer
| |
| .
| |
| . 2
| |
| | |
| e
| |
| ,, EXAM ANSWER KEY CORRECTIONS
| |
| EXAM QUESTION RO #27/ SRO #29 !
| |
| The plant is in Mode 4 with Shutdown Cooling in service on the "A" Residual Heat Removal (RHR)
| |
| loop with the "A" RHR Pump running.
| |
| Which of the following describes how a loss of the "B" Reactor Protection System (RPS) but will affect
| |
| the inboard and the Outboard Shutdown Cooling isolation Valves (F008 & F009)?
| |
| a. TheF008 and F009 valves both close. J
| |
| b. The F008 valve closes and the F009 valve remains open.
| |
| c. The F008 and F009 both remain open.
| |
| d. The F008 valve remains open and the F009 valve closes.
| |
| Ans.B
| |
| Ref HC.OP-SO.SM-0001(O), rev 5, page 3, section 3.1.3
| |
| LP 0302-000.00H-000045, rev 12
| |
| Obj. R3.b & R4
| |
| 1. The answer key per the stated reference is incorrect. The correct answer per the stated reference
| |
| is "a".
| |
| ! l
| |
| I
| |
| RECOMMEDATION:
| |
| Change answcr key to choice "a" as the correct answer.
| |
| '
| |
| )
| |
| \
| |
| .
| |
| 3
| |
| | |
| s
| |
| EXAM ANSWER KEY CORRECTIONS
| |
| EXAM CUESTION RO #98 / SR3 #96
| |
| Given the following plant conditions:
| |
| . The plant was operating at rated power when a steam line break occurred in the HPCI room.
| |
| . HPCI isolated due to high room temperatures
| |
| . RBVB exhaust radiation levels reached 1.0 E-2 microcuries/ml
| |
| Which of the following describes the ventilation system response for the given conditions?
| |
| a. RBVS remains running.
| |
| b. RBVS isolated,6 FRVS Recirc and 1 FRVS Vent fans are in service.
| |
| c. RBVS isolated,4 FRVS Recire and 1 FRVS Vent fans are in service.
| |
| d. RBVS isolated,6 FRVS Recirc and 2 FRVS Vent fans are in service
| |
| Ans. A
| |
| Ref. HC.OP-EO.ZZ-0103, rev.10
| |
| LP 0302-000.00H-000127, rev 10, page 8
| |
| Obj. 2 & R6
| |
| 1. The answer key was incorrectly typed, the correct answer should be "b"
| |
| 2. RBVS exhaust radiation levels reached (1.0 E-2 microcuries/ml) is > 1.0 E-3 which is the isolation
| |
| signal for RBVS and an initiation signal for FRVS see HC.OP-SO.GU-0001 " Filtration,
| |
| Recirculation and Ventilation System Operation"
| |
| 3. This is also an entry condition for HC.OP-EO.ZZ-0103, the lesson plan page listed lists the action
| |
| of HC.OP-EO.ZZ-0103 for the retention override that
| |
| if
| |
| . Reactor Bldg. exhaust Rad level exceeds 1 x10'8
| |
| or
| |
| 4
| |
| . Refuel Floo7HVAC Exhaust Rad Level exceeds 1 x 10
| |
| Then
| |
| . Verifyisolation of RBVS
| |
| And
| |
| . Initiation of FRVS
| |
| Recommendation
| |
| !
| |
| Change answer key to choice "b" as the correct answer
| |
| >
| |
| l
| |
| 4
| |
| L_______________________________.__
| |
| | |
| * 1
| |
| CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
| |
| ,,
| |
| EXAM QUESTION RO #20 / SRO #24 .
| |
| Given the following conditions: j
| |
| e Preparations are complete to start the "A" Recirculation Pump
| |
| + The Pump Discharge Valve (F031 A) is closed
| |
| .
| |
| 1
| |
| Which of the following describes how the "A" Recirculation Pump trip on the discharge valve is l
| |
| bypassed to allow the pump to be started?
| |
| a. This trip is bypassed until the pump start sequence is complete within prescribed time limits. 1
| |
| b. This trip is bypassed until the discharge valve has reached the 100% open position.
| |
| c. This trip is bypassed until the pump has been running for 9 seconds.
| |
| d. This trip is bypassed until the discharge valve jog (open) circuit has timed out.
| |
| Ans A
| |
| Ref 0302-000.00H-000020, rev 15, page 39 & 40, section VI.A.2.e.5).c)
| |
| LP 0302-000.00H-000020, rev 15, page 39 & 40, section VI.A.2.e.5).c)
| |
| Obj R10
| |
| 1. The referenced Reciculation Flow Control Lesson Plan does not go into sufficient detail, neither in
| |
| the lesson plan body nor in the learning objectives, to differentiate between the discharge valve
| |
| jog circuit from the pump start sequence as the permissive for pump start process completion. ,
| |
| 1
| |
| '
| |
| 2. Upon review of normal Control Room references (attached) it is shown on marked up sheets 8,
| |
| 14, and 17;
| |
| . That the K51 relay, which is energized during the start sequence, bypasses the 90% open trip
| |
| to the drive motor breaker until 85 seconds after the sequence has been initiated. This makes
| |
| choice "a" a correct answer
| |
| . That the K54 relay, which is denergized by the jog circuit timer, bypasses the full closed trip
| |
| signal to the drive motor breaker for the first three seconds of jog circuit operation. This
| |
| makes choice "d" a correct answer.
| |
| RECOMMENDATION:
| |
| '
| |
| Accept both a and d choices as correct answers.
| |
| 5
| |
| | |
| *
| |
| CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY -
| |
| EXAM QUESTION SRO #69
| |
| Which of the following is the basis of the 65 psig Suppression Chamber pressure Limit? l
| |
| a. 65 psig is the primary containment maximum expected post-LOCA pressure.
| |
| b. Above 65 psig, the system lineup required for containment venting may not be able to be
| |
| completed.
| |
| c. Above 65 psig, the Safety Relief Valves may not be available when required for an Emergency
| |
| Depressurization.
| |
| d. 65 psig is the operationallimit of the Torus to Drywell vacuum breakers.
| |
| Ans. C
| |
| Ref. 0302-000.00H-001268, " Primary Containment Control -Orywell Pressure" , rev
| |
| Obj. R6/R7
| |
| 1. 0302-000.00H-00126B," Primary Containment Control-Drywell Pressure", rev-11 (attached)
| |
| states that 65 psig is the maximum pressure at which SRV's can be opened. This makes "c" the
| |
| correct answer
| |
| 2. 0302-000.00H-00124A, "RPV Water Level Control", rev.10, (attached) states regarding the
| |
| Primary Containment Pressure Limit that above this limit
| |
| . The vent valves in the primary containment vent path above TAF may not open
| |
| . The SRV's may not be able to be manually opened with PCIG at 90 psig.
| |
| 3. This obvious discrepancy was discussed with the Operation Department Emergency Operating
| |
| procedure writers, and the Primary Containment Pressure Limit / Maximum Primary Containment
| |
| Water Level limit worksheet (PSTG WS-9) identifies both the vent valves opening and SRV
| |
| opening as limiting components. This makes "b" also a correct choice
| |
| Recommendation:
| |
| Accept choices "b" and "c" as correct answers
| |
| l
| |
| l
| |
| I
| |
| l
| |
| l
| |
| l
| |
| 6
| |
| | |
| COPIRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
| |
| a>
| |
| EXAM QUESTION RO #76 / SRO #71
| |
| During a loss of coolant accident the following conditions exist: )
| |
| e Reactor pressure is 125 psig
| |
| * Drywell temperature is 325 'F
| |
| Which of the following describes the accuracy and trending capabilities of wide range reactor water
| |
| level indication for the given conditions?
| |
| I
| |
| a. They are not providing accurate reactor water level or level trend information.
| |
| .
| |
| b. They are providing accurate reactor water level but level trend is not reliable.
| |
| c. They are providing accurate reactor water level and level trend information. !
| |
| d. They are not providing accurate reactor water level but level trend is reliable.
| |
| Ans. C
| |
| Ref EOP Caution 1, HC.OP-EO.ZZ-0101 RPV Water Level Control Section,
| |
| LP 0302-000.00H-00124A, rev 10
| |
| Obj. 7 l
| |
| 1. The wide range instruments are calibrated for normal operating pressure and temperature, where
| |
| RPV level is significantly below Normal operating range. See attached 0302-000.00H-000002
| |
| " Nuclear Boiler Instrumentation".
| |
| 2. At lower than normal operating pressure the wide range indicators read higher than actual level
| |
| when RPV level is above the mid scale range. See attached temperature compensation curves 3
| |
| from HC.OP-lO.ZZ-0003(O). l
| |
| 3. Since RPV level was not given, the accuracy of the Wide range level instrument is in question,
| |
| )
| |
| depending on the assumption of the candidate. ,
| |
| 4. The conditions given show that the instrument Reference leg should not be affected by potential
| |
| flashing, since we are below the saturation curve, as could be determined by steam tables
| |
| provided to the candidates, this makes the instrument reliable for trending, as stated in EOP
| |
| caution #1
| |
| 5. Based on the assumption of the candidate, either "c" Accurate level and trend, or "d"
| |
| Inaccurate level but reliable trend would be acceptable answers
| |
| RECOMMENDATION:
| |
| Accept "c" or "d" as correct answers
| |
| !
| |
| ,
| |
| l
| |
| | |
| >
| |
| j
| |
| CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
| |
| .
| |
| Exam Question RO #79
| |
| Given the following conditions:
| |
| . The plant is at 75% power
| |
| e _ Control rod 22-27 is being withdrawn on notch to Notch "22"
| |
| Which of the following is the required immediate operator action if a control rod drift alarm is received
| |
| and the operator notes control rod 22-27 is continuing to move out and power is rising?
| |
| a. Apply a continuous insert signal to control rod 22-27.
| |
| b. Place the Rod Select key lock switch to "off"(de-select the rod).
| |
| c. Direct the local operator to perform a single rod scram on control rod 22-27.
| |
| d. Runback recirculation flow and insert control rods to reduce power.
| |
| Ans. D
| |
| Ref HC.OP-AB.ZZ-0204 Positive reactivity addition,
| |
| LP 302H-000.00H-000114, rev 5
| |
| Obj. 1
| |
| 1. Runback recirculation flow and insert control rods to reduce power, is a prescribed method for
| |
| power reduction as stated in HC.OP-AB.ZZ-0204 section 3.1 which makes "d' a correct choice.
| |
| 2. Applying a continuous insert signal to control rod 22-27 is a method of " inserting control rods to
| |
| reduce reactor power" and therefore, makes choice "a" a correct answer IAW HC.OP-AB.ZZ-
| |
| 0204.
| |
| 3. Additionally, since the question states that "the operator notes that control rod 22-27 is continuing
| |
| to move out and power is rising", the operator could enter abnormal procedure HC.OP-AB.ZZ-
| |
| 0102 Dropped Control Rod. IAW with this procedure the immediate actions are to:
| |
| . If necessary then Insert control rods, in sequence, to terminate the power increase.
| |
| . If a scram condition is reached, Then ensure the reactor scrams and implement procedure
| |
| HC.OP-EO.ZZ-0100(O)
| |
| . Ensure that all appropriate automatic actions are complete.
| |
| 4. Inserting control rod 22-27 would be correct for this abnormal procedure since that would be the
| |
| first rod to insert "in sequence".
| |
| RECOMMENDATION:
| |
| Accept choices "a" and "d" as correct answers
| |
| 8
| |
| | |
| e
| |
| ~
| |
| CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
| |
| EXAM QUESTION RO #84 / SRO #79
| |
| A loss of Reactor Auxiliary Cooling System has occurred?
| |
| Which of the following is the MAXIMUM time allowed, before a reactor scram is required?
| |
| a. An immediate scram is required
| |
| b. One{1) minute
| |
| c. Ten (10) minutes
| |
| d. Twenty (20) minutes
| |
| Ans C ,
| |
| Ref HC.OP-AB.ZZ-0123, rev 5, caution 4.8 I
| |
| LP 0302-000.00H-000114, rev 9, page 3
| |
| Obj. 3
| |
| 1. The answer key has "a" as being correct, based on caution (4.8) of HC.OP-AB.ZZ-0123 which
| |
| allows 10 minutes to get RACS restored to the recirc pumps or they must be tripped, the operator
| |
| is cautioned to place the mode swi^.ch in " shutdown" prior to tripping the pumps. This makes "c" a ;
| |
| correct answer.
| |
| l
| |
| 2. Section 4.9 of the same procedure states if a totalloss of RACS has occurred and cannot be
| |
| immediately restored them perform the following:
| |
| . Scram the reactor
| |
| . Trip both Recirc pumps
| |
| . Trip both CRD pumps .
| |
| . Trip both RWCU pumps
| |
| 3. One SRO candidate asked the exam proctor if this loss was a " total loss". His response was yes.
| |
| Using a total loss and following that direction, this would make "a" also a correct answer.
| |
| Recommendation
| |
| Accept cholces "a" or "c" as correct answers
| |
| 9
| |
| | |
| 5
| |
| QUESTION DELETIONS ,,
| |
| Exam Question RO #04 / SRO #05
| |
| Following shift tumover the Nuclear Control Operator (RO) notes that data entered in the narrative log
| |
| by the previous shift incorrect.
| |
| The RO draws a single line through the incorrect entry, makes the correct entry and initials and dates
| |
| the change. Which of the following describes how the RO should highlight and explain the change?
| |
| a. The correct entry should be circled in red with an explanation placed in the comments section.
| |
| b. The correct entry should be circled in red with an explanation made next to the corrected entry,
| |
| c. The incorrect entry should be circled in red with an explanation placed in the comments section,
| |
| d. The incorrect entry should be circled in red with an explanation made next to the corrected entry.
| |
| Ans. A
| |
| Ref HC.OP-AS.ZZ-0002, rev 2, page 20, section - Log Taking
| |
| LP 0302-000.00H-000113, rev 8
| |
| Obj. 125R
| |
| 1. LP-0302-000.00H-000113, rev 8 objective 125 (attached ), specifi:: ally states "Given access to
| |
| control room references, distinguish between proper and improper methods of maintaining
| |
| Operations Department logs IAW HC.OP-AP.ZZ-0110. This procedure was not provided for the
| |
| candidates to review to determine correct choice.
| |
| 2. HC.OP-AP.ZZ-0110 (applicable pages attached) defines the use of the Narrative and Comments
| |
| section logs. It also describes Data logs and requirements of circling abnormal, unusual, or O.O.S.
| |
| data in red ink, additionally it states that any abnormal, unusual, or O.O.S. entries will be
| |
| investigated immediately and recorded on the applicable comments section. HC.OP-AP.ZZ-0110
| |
| further has a description of the Comment Sheets / Sections and states they are the Narrative Log
| |
| for operating stations that do not have a formal Narrative Log ledger.
| |
| 3. HC.OP-AS.ZZ-0002, page 20 (attached) specifically states if an entry is corrected by an individual
| |
| other than the person entering the ggta, the correction must be circled in red with an explanation
| |
| _
| |
| in the comments section.
| |
| 4. The NCO Narrative Log (attached) is a comments logs in itself and not a data log. Data is taken
| |
| on logs such as DL-0002 (attached) which has a comments section. The misapplication of the
| |
| NCO Narrative Log as the Data log vice any DL log supplied with a comments section, prevented
| |
| the candidates from determining the correct selection.
| |
| RECOMMEDATION:
| |
| Delete question from exam
| |
| to
| |
| .. .. .-
| |
| | |
| a }
| |
| QUESTION DELETIONS
| |
| * Exam Cuestion RO #78 / SRO #73 ;
| |
| .
| |
| Given the following conditions:
| |
| . Reactor power is 82%
| |
| . HPCI is in operation for a surveillance
| |
| . The "B" loop of RHR is in Suppression Pool Cooling
| |
| . Suppression Pool temperature is 103*F when the running RHR Pump tripped
| |
| . ' HPCI was secured
| |
| . Subsequently, suppression pool temperature reached 106'F
| |
| i
| |
| Which of the following lists the suppression pool temperatures requiring entry into HC.OP-EO.ZZ- j
| |
| 0102, Primary Containment Control AND entry into the LCO actions for Tech Spec 3.6.27
| |
| a. EO-0102 - 95'F
| |
| TS 3.6.2 -
| |
| 95*F
| |
| b.- EO-0102 - 95'F
| |
| TS 3.6.2 -
| |
| 105 F
| |
| c. EO-0102 - 105 F
| |
| TS 3.6.2 - 95*F
| |
| l
| |
| d. EO-0102 - 105'F
| |
| . TS 3.6.2 -
| |
| 105'F
| |
| i . Ans: D
| |
| '-
| |
| Ref: 0302-000.00H-00125A, " Primary Containment Control - Suppression Pool Temp", rev 10
| |
| ~HC . IS.BJ-0001, "HPCI inservice test", step 5.1.16, rev 29
| |
| Obj. .3
| |
| 1. 0302-000.00H-00125A, " Primary Containment Control - Suppression Pool Temp", Rev 10,
| |
| l objectives do require knowledge of entry conditions to EOP-0102 (attached)
| |
| l
| |
| 2. HC.OP-IS.BJ-0001, rev 29, step 5.1.16 states to implement suppression pool average water
| |
| temperature monitoring of technical specification 3.6.2.1 prior to and during HPCI operations by
| |
| l performing HC.OP-DLZZ-0026(O) (both attached)
| |
| 3. No leaming objective in the Hope Creek Operations Training program requires commitment to
| |
| memory inservice Test cautions and bases behind the cautions.
| |
| l
| |
| 4. No Leaming Objective for Technical Specification evaluation require determination of Technical
| |
| Specification actions without having the applicable section of the procedure available for i
| |
| reference. j
| |
| 4
| |
| , .
| |
| '
| |
| 5. Testing Technical Specification compliance without the materials available for review is not in the !
| |
| best interest of the candidate or in compliance with Hope Creek Operations Training Department
| |
| . objectives, j
| |
| 6.' Nuclear Business Unit Procedural Compliance requirements, and expectations, for use of a j
| |
| Catergory I procedure require step by step compliance. The same level of procedural '. sage .
| |
| should be complied with during examinations, and was not. !
| |
| Recommendation: j
| |
| Delete question 1
| |
| 11
| |
| | |
| +
| |
| QUE3 TION DELETIONS
| |
| EXAM QUESTION SRO #75
| |
| Which of the following describes how the operators would know the Hydrogen Water Chemistry
| |
| injection (HWCl) system had NOT been removed from service while performing a shutdown in
| |
| accordance with HC.OP-lO.ZZ-0004(O), " Shutdown from Rated Power to Cold Shutdown"?
| |
| a. Hydrogen explosions in the Mechanical Vacuum Pump while operating to maintain condenser
| |
| vacuum.
| |
| b. Post-shutdown (2 hours) Turbine Building radiation levels would be much higher,
| |
| c. Alarms and indications resulting from a control rod drop accident would not be available to the
| |
| operators as quickly.
| |
| d. The Primary and Secondary Condensate Pumps will cavitate.
| |
| Ans. C
| |
| Ref HC.OP-AB.ZZ-0102, Dropped Control Rod, rev. 3
| |
| LP 0302-000.00H-000225, rev 05
| |
| Obj. 6 & 7.1
| |
| in order for this situation occur the operators would be required to violate procedure HC.OP-lO.ZZ-
| |
| 0004, " Shutdown from F d Power to Cold Shutdown". If the operators failed to have the Chemistry
| |
| Department remove Hh from service at 35% power, they would also have to miss the next step of -
| |
| the procedure which instructs the operators to have l&C restore the MSL RMS setpoints. If the
| |
| setpoints are restored with HWCl in service then RMS alarms may result which could clue the
| |
| operators in to the problem with HWCl. This scenario requires multiple procedure violations.
| |
| There is'no power level specified for this question and in order for HWCl to remain in service it would
| |
| have failed to trip at 30% power (as it is currently designed).
| |
| Technical Specifications require that with reactor power at 20%, the only control rod motion that is
| |
| allowed is by a scram if MSL Rad Monitor Setpoints have not been restored. HC.OP-AB.ZZ-0102
| |
| " Dropped Control Rod" section 5.3 states "The effects of a rod drop accident above 20% power are
| |
| minimal; therefore, H2 injection system operation is only permitted above 20% power".
| |
| There are numerous protections to prevent the conditions specified in this question from occurring.
| |
| The likelihood of all of these failures and then a rod drop accident are too remote to expect the
| |
| students to select choice "c" as the correct answer.
| |
| RECOMMENDATION:
| |
| Delete question from exam
| |
| 12
| |
| | |
| '
| |
| 3
| |
| , HC.OP-SO.CH-0001(Z)
| |
| ATTACHMENT 4
| |
| -
| |
| (Page1of1)
| |
| MAIN TURBINE CONTROL OIL (EHC) SYSTEM OPERATION
| |
| EHC CONTROL LOGIC DIAGRAM ,
| |
| - ~ ~ um n .a n.,. .,
| |
| 7 **"" "
| |
| 3 ; .H!
| |
| " *
| |
| - + "
| |
| - [-' -e ,". ' !, w === .h3 _,/_,.-.i .cuce =,:
| |
| ;
| |
| ."., 41) .
| |
| .
| |
| " y- ,
| |
| ve 4/s\
| |
| /
| |
| ==.
| |
| m. N,",/- _
| |
| ~-. ,,
| |
| . , . ._
| |
| "- ,L ' *
| |
| u,/ *oum g a
| |
| = '
| |
| + ~
| |
| dh -
| |
| C _,
| |
| _,
| |
| ""
| |
| .* .noama -* uw m, =
| |
| mm.uw& ?. %*' .
| |
| .
| |
| *sf
| |
| <
| |
| l .c
| |
| "m"
| |
| m
| |
| * '
| |
| '
| |
| l f- 7 met
| |
| . lH ".,"2 -
| |
| A ( ;
| |
| ,, l
| |
| E
| |
| . --
| |
| (n==J-IH, !. " ; "
| |
| 7% / f- .
| |
| wr
| |
| *"
| |
| i
| |
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| |
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| |
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| |
| l' 'f \ j
| |
| .I j
| |
| gun l 1
| |
| ir t
| |
| '
| |
| a=_ . T)u~ l
| |
| ==l=,
| |
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| Page M2 of 84 Rev.19
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| Hope Creek
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| ATTACHMENT 4
| |
| l
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| NRC RESOLUTION OF PSE&G COMMENTS ON THE WRITTEN EXAM
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| -
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| . . , ,
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| RO-4 / SRO-5 The facility recommended to delete this question from th,e exam.
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| Based on a review of the references provided, the NRC staff agreed
| |
| with the facility that this question should be deleted from the exam.
| |
| There was no clear reference to clearly support a correct answer to
| |
| this question.
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| RO-19 The facility recommended to change the answer key from "c" to "b".
| |
| Based on a review of the procedure HC.OP-AB.ZZ-0112, Recirculation
| |
| Pump Trip, there was no answer that provided the sequence to isolate
| |
| the recirculation pump as required by the procedure. The NRC staff
| |
| did not totally agree with the facility recommendation. Since there
| |
| were no correct answers to this question, the appropriate action was
| |
| ,
| |
| to, delete the question from the exam.
| |
| RO-20 / SRO-24 The facility recommended to accept both'"a" and "d" as correct
| |
| answers to 'the question. Based on review of the referenc'es, the
| |
| NRC st.aff agreed with the facility. The answer key was revised to ;
| |
| accept "a" and "d" as correct answers.
| |
| 80-27 / SRO-29 The facility recommended to change the answer key from "b" to "a".
| |
| Based on review of the references and prior exam versions, this was
| |
| clearly a typographical error, the NRC staff agreed with the facility.
| |
| The answer key was revised from "b" to "a".
| |
| RO-76 / SRO-71 The facility recommended to accept both "c" and "d" as correct i
| |
| answers to the question. Caution 1 of the emergency operating !
| |
| procedures indicated that if drywell temperature and reactor pressure l
| |
| were below the saturation curve then wide range level indication was
| |
| a reliable instrument. Since the given conditions were below the
| |
| saturation curve and steam tables were available to @e applicants,
| |
| they had sufficient information to concluded that the 4 vide range level
| |
| instrument was useable for the entire range and thus "c" was could l
| |
| be a correct answer since no accuracy range was delineated.
| |
| The examiner further reviewed the licensee provided curve showing
| |
| inaccuracy of the water level instruments over a range of levels and
| |
| of reactor coolant system pressures and temperatures. Answer "d"is
| |
| also correct in that the level instrument would not be providing
| |
| accurate water level indication but the trend would be reliable. After
| |
| further review answer "a" is also correct with the "or" condition that
| |
| the instruments "are not providing accurate reactor water level or
| |
| level trend information.
| |
| Accordingly, since the question has three correct answers, it was
| |
| deleted from the exam for the reasons noted by the examiner above.
| |
| | |
| *
| |
| a
| |
| RO-78 / SRO-73 The facility recommended that this question be deleted from the
| |
| exam. The licensee indicated that the question was testing the
| |
| applicant's memory of specific technical specification limiting -
| |
| - condition for operations (LC) actions or emergency operating .
| |
| procedure actions as suggested. The examiner viewed the question
| |
| as testing the applicant's knowledge of the entry conditions into these
| |
| documents at the analysis level, which is a more challenging question.
| |
| This was a acceptable testing area as identified by the KA assigned to'
| |
| this question and because of the importance of this LC. Since there
| |
| was a single correct answer to the question, there was no basis to {
| |
| delete the question from the exam. An acceptable basis would have
| |
| been no correct answer or more than two correct answers. The
| |
| facility comment was not accepted.
| |
| RO-79 The facility recommended to accept both "a" and "d" as correct
| |
| answers to the question. The question required the applicant to
| |
| ,
| |
| '
| |
| identify the required immediate operator actions. Answer "a" was not
| |
| a required immediate operator action identified in HC.OP-AB.ZZ-0204,
| |
| Positive Reactivity Addition. The facility recommendation was not
| |
| . .
| |
| accepted.
| |
| . RO-84 / SRO-79 The facility recommended to accept both "a" and "c" as correct
| |
| answers since one applicant was told by the proctor, in response to a
| |
| question, that this was a total loss of RACS. The proctor's response
| |
| did not alter the question since ten minutes is still the maximum time
| |
| allowed before a reactor scram is required and answer "c" is the only
| |
| correct answer. There was no change to the answer key.
| |
| RO 98 / SRO-96 The facility recommended to change the answer key from "a" to "b".
| |
| Based on review of the references and prior exam versions this was
| |
| clearly a typographical error and the NRC staff agreed with the
| |
| facility. The answer key was revised from "a" to "b".
| |
| SRO-69 The facility recommended to accept both "b" and "c" as correct
| |
| answers to this Destion. Based on review of the references, the
| |
| NRC staff agreed with the facility. The answer key was revised from
| |
| to accept both "b" and "c" as correct answers.
| |
| SRO-75 The facility recommended to delete the question from the exam
| |
| without sufficient supporting justification as to why it should be
| |
| deleted. The question was based on the discussion section of HC.OP-
| |
| AB.ZZ-0102, Dropped Control Rod, on why hydrogen injection is
| |
| secured at low power. This was a legitimate testing area as identified
| |
| by the KA assigned to the question. Since the question was valid
| |
| with the one correct answer to the question, there was no basis to
| |
| delete the question from the exam. There was no change to the
| |
| answer key.
| |
| ..
| |
| | |
| O
| |
| e
| |
| ATTACHMENT 5
| |
| SIGNIFICANT CONTROL MANIPULATION DETAILS
| |
| APPLICANT DATE IYPE ASSESSMENT- ,
| |
| 55-62176 4/6/97 Flow Acceptable.
| |
| 4/6/97 Flow Unacceptable - No documentation available to support
| |
| that this was not part of a continuous power change.
| |
| 4/6/97 Flow Unacceptable - No documentation available to support
| |
| that this was not part of a continuous power change.
| |
| 4/6/97 Rods Acceptable.
| |
| 4/6/97 Rods Unacceptable - Documentation indicated that this was ,
| |
| part of'a continuous power change.
| |
| 4/6/97 Rods Unacceptable - Documentation indicated that this was
| |
| part of a continuous power change.
| |
| ~
| |
| l
| |
| 4/6/97 Rods Unacceptable - Documentation indicated that this was
| |
| part of a continuous power change.
| |
| l
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| l
| |
| | |
| o
| |
| se
| |
| APPLICANT DAIE TYPE ASSESSMENT
| |
| 55-62178 2/1/97 Flow Acceptable
| |
| 3/1/97 Flow Acceptab.le
| |
| 1
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of inserting 5 i
| |
| rods from 16-12 to reduce power and then three
| |
| examples of partially withdrawing a control rod
| |
| individually scrammed by a licensed operator as part of
| |
| individual control rod scram testing, another applicant
| |
| also completed the withdrawal. No documentation was
| |
| available to support that the control rod movement
| |
| resulted in an observable effect on power. Rod
| |
| withdrawal to recover from an individual rod scram test
| |
| was not considered to be significant.
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of eight
| |
| examples of partially withdrawing a control rod
| |
| individual'ly scrammed by a licensed operator is part of
| |
| individual control rod scram testing. Another applicant
| |
| also completed the withdrawal. Rod withdrawal to
| |
| recover from an individual rod scram test was not
| |
| considered to be significant.
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of three
| |
| examples of partially withdrawing a control rod
| |
| individually scrammed by a licensed operator as part of
| |
| individual control rod scram testing, another applicant
| |
| also completed the withdrawal, and withdrawing 5 rods
| |
| from 12-16. No documentation was available to
| |
| support that the control rod movement resulted in an
| |
| observable effect on power. Rod withdrawal to recover
| |
| from an individual rod scram test was not considered to
| |
| be significant.
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| | |
| o
| |
| S
| |
| APPLICANT DATE IYEE ASSESSMENT
| |
| 55-62183 2/2/97 Flow Acceptable
| |
| 1
| |
| 3/1/97 Flow Acceptable
| |
| I
| |
| '
| |
| 2/2/97 Rods Unacceptable - Rod movement consisted of inserting
| |
| four rods from 10-06 and then withdrawing the same
| |
| four rods from 06-10. This did not meet the PSE&G
| |
| acceptance criteria of at lease one notch for a minimum
| |
| of eight rods.
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of inserting 3
| |
| rods from 08-00, four rods from 14-12 and three rods
| |
| from 16-12. There was no documentation to support
| |
| l
| |
| that this resulted in an observable power affect.
| |
| i
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of eight
| |
| l
| |
| examples of partially withdrawing a control rod j
| |
| ' individually scrammed by a licensed operator as part of j
| |
| individual control rod scram testing. Another applicant )
| |
| also completed the withdrawal. Rod withdrawal to
| |
| recover from an individual rod scram test was not
| |
| considered to be significant, ,
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| l
| |
| l'
| |
| l
| |
| l
| |
| I
| |
| l
| |
| | |
| 0
| |
| m
| |
| APPLICANT DATE IyfE ASSESSMENT
| |
| 55-62187 2/2/97 Flow Acceptable
| |
| ,
| |
| - -
| |
| ,
| |
| 3/1/97 Flow Acceptable
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of eight
| |
| examples of partially withdrawing a control rod
| |
| individually scrammed by a licensed operator as part of
| |
| individual control rod scram testing. Another applicant
| |
| also completed the withdrawal. Rod withdrawal to
| |
| recover from an individual rod scram test was not
| |
| considered to be significant.
| |
| 3/1/97 Rods Unacceptable - Rod movement consisted of eight
| |
| examples of partially withdrawing a control rod
| |
| individually. scrammed by a licensed operator as part of
| |
| individual control rod scram testing.~ A'nother applicant
| |
| also completed the withdrawal. Rod withdrawal to
| |
| recover'from'an individual rod scram test was not
| |
| considered to be significant.
| |
| 3/1/97 Rods Unacceptette - Rod movement consisted of
| |
| withdrawing 7 rods from 12-16 and one rod from 00-
| |
| 08. There was no documentation to support that this
| |
| resulted in an observable power affect.
| |
| 2/21/98 Flow Acceptable
| |
| l
| |
| 2/21/98 Flow Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| !
| |
| !
| |
| I
| |
| f
| |
| | |
| _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ - - _ - _ _ _ _ - _ - _ - _ _
| |
| _ . . . . . . - - . . . . . . . .-
| |
| 4
| |
| m, .
| |
| APPLICANT DATE Iyfg ASSESSMENT
| |
| 55-62175 4/6/97 Rods Acceptable
| |
| 4/6/97 Rods Acceptable
| |
| 4/6/97 Rods Acceptable
| |
| b
| |
| 4/6/97 Rods Unacceptable - Rod movement consisted of
| |
| withdrawing four control rods from notch 00-06 and
| |
| then withdrawing the same four rods from notch 06-12.
| |
| This did not meet PSE&G acceptance criteria of at least
| |
| one notch for a minimum of eight rods.
| |
| 6/20/97 Flow Acce,ptable
| |
| 2/21./98 Flow Acceptable
| |
| \
| |
| F
| |
| F
| |
| l
| |
| .
| |
| r.
| |
| f
| |
| .
| |
| 1
| |
| _ _ _ . . .
| |
| | |
| F
| |
| A
| |
| APPLICANT DATE TYPE ASSESSMENT
| |
| 55-62174 4/6/97 Flow Acceptable
| |
| 6/3/97 Flow Acceptable
| |
| 7/10/97 Flow Acceptable
| |
| 9/4/97 Flow Acceptable
| |
| 4/6/97 Rods Acceptable
| |
| 4/6/97 Rods Acceptable
| |
| 4/6/97 Rods Acceptable
| |
| 4/6/97 Rods Unacceptable - Rod movement consisted of
| |
| withdrawing four control rods from notch 12-14, then
| |
| these same four rods from 14-16, and these same four i
| |
| rods again from 16-18. This did not meet the PSE&G l
| |
| acceptance criteria of at least one notch for a minimum
| |
| of eight rods.
| |
| l
| |
| 5/9/97 Rods Did not assess since applicant had the required number.
| |
| l
| |
| l
| |
| l
| |
| l
| |
| l
| |
| l
| |
| f
| |
| | |
| k 1
| |
| d,
| |
| APPLICANT DATE TX25 ASSESSMENT
| |
| 55 60013 12/13/97 Rods Acceptable
| |
| 12/13/97 Rods Unacceptable - Documentation indicated that this was
| |
| part of a continuous power change.
| |
| 12/13/97 Rods Unacceptable - Documentation indicated that this was
| |
| part of a continuous power change.
| |
| 12/13/97 Rods Acceptable
| |
| 12/14/97 Flow Acceptable
| |
| 12/14/97 Rods Acceptable
| |
| 2/21/98 Flow Acceptable
| |
| i
| |
| ;
| |
| }}
| |