ML20211J941
| ML20211J941 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 10/06/1997 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9710090064 | |
| Download: ML20211J941 (109) | |
Text
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%,I gs g NUCLEAR REGULATORY COMMISSION
,I RE0lON IV 34
- f -- sii nvAN etAzA on vE. suite 4oo 9- AnLINGTON TEXAS 76o11 So64 l
October 6, 1997 i
NOTE T0: NRC Dbcuitent Control' Desk Mail Stop 0 5 D 24 FROM: Laura Hurley, Licensing Assistant -
. Operations Branch, Region IV
SUBJECT:
OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON AUGUST 25 29, l 1997. AT WOLF CREEK GENERATIrlG STATION, UNIT 1
(
DOCKET #50 482 On August 25 29, 1997, Operator Licensing Examinations were, administered-at the referenced facility. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070, b) As given operating examination, designated for distribution under RIDS Code A070.
Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.
If you have any questions, please contact Laura Hurley, Licensing Assistant.
Operations Branch, Region IV at (817) 860 8253.
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- Wolf Creek Nuclear Operating Corporation i
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E Mail report to Document C ntrol Desk (DOCDESK) bec to DCD (IE01)
L bec distrib. by EP1 w/ Enclosure and Attachments 13:
l Regional Administrator Resident inspector l l DRP Director SRI (Callaway, RIV)
Branch Chief (DRP/B) DRS-PSB
' Project Engineer (DRP/B) MIS System Branch Chief (DRP/TSS) RIV File L. Vick (HOLB/NRR) bec distrib. by RIV w/ Enclosure and Attachments and Attachmonts 14:
R. Gallo (HOLB/NRR)
L. Hurley DOCUMENT NAME: R:\_WC\WC7301 RP.HFB .
I/i To receive copy of document. indicate in box: ac" = Ccpy weout enclosures "E" = Copy we enclosures *N" = No - /
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\ NULLEAR REGULATORY COMMISSION R EGioN IV 611 RY AN PLAZA DRIVE, sulTE 400
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Otto L. Maynard, President and Chief Executive Officer
!- Wolf Creek Nuclear Operatir.g Corporation L
P.O. Box 411 Burlington, Kansas 66839
SUBJECT:
NRC INSPECTION REPORT S0-482/97 301 l
Dear Mr. Maynard:
i An NRC inspection was condveted August 25 29,1997, at your Wolf Creek Generating Station reactor facility. The enclosed report presents the scope and results of that inspection.
The inspection included an evaluation of five applicants for reactor operator licenses and three epplicants for_ senior reactor operator licenses. - We determined that three applicants for reactor operator licenses and one applicant for a senior reactor operator licence satisfied the requirements and the appropriate licenses have been issued.
Following the exit' meeting, the results of this inspection were discussed on September 4, 1997, by Mr. McKernon with Mossrs. Pippin and Guyer of your staff with regard to the number of technical accuracy issues with the written examination developed during the post examination analysis. Over 5 percent of the written examinatian answers were changed based on technical accuracy issues identified in your post-examination review.
We are in receipt of your letter dated September 5,1997, requesting a meeting to discuss the results of the examination, your training prog am, and the evamination process. We agree that a meeting is appropriate. At this meeting, we request that you: address the potential knowledge weaknesses diset.ssed in Section 04.1 of the enclosed inspection report; provide the NRC with information explaining your staff's failure to identify the high
. number of examination technical accuracy problems before examination administration; and describe corrective actions your staff plans to take to assure that the next licensed operator, written examination is of higher quality. Please contact Mr. J. Pellet at 817/860 8159 to schedule this meeting, in accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will De placed in the NRC Public Document Room (PDR).
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- Wolf Creek Nuclear Operating Corporation 2-1 J l
Should you have any questions concerning this inspection, we will be pleased to discuss 1 them with you,
)
i Sincerely, ,A in /
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. Arthur T. Ho sirIll, Director Division of Reactor Safety
- Docket No.: 50 482 '
License No.: NPF 42 i-
Enclosure:
NRC Inspection Report -
60 482/97 301 ,
l .cc w/ enclosure and Attachments 13: ,
. Chief Operating Officer Wolf Creek Nuclear Operating Corp.
P.O. Box 411 4 Burlington, Kansas 66839 j Jay Silberg, Esq.
Shaw, Pittman, Potts & Trowbridge 2300 N Street, NW-Washington, D.C. P.0037 l
- - Supervisor Licensing i Wolf Creek Nuclear Operat!ng Corp.
P.O. Box 411
.. Burlington, Kansaa 06839 Chief Engineer Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Rd.'
- Topeka, Kansas 66604 4027
)- Office of the Governor '
State of Kansas Topeka, Kansas' 66612
Wolf Creek Nuclear Operating Corporation 3-l l Attorney General l Judicial Center l 301 S.W.10th
! 2nd Floor
!- Topeka, Kansas - 66612 1597 County Clerk _
Coffey County Courthouse
-)
l Burlington, Kansan 668391798 I Vick L. Cooper, Chief -
Radiation Control Program
- Kansas Department of Health -
and Environment Sureau of Air and Radiation Forbes Field Building 283' Topeka, Kansas 66620 Mr. Frank Moussa Division of Emergency Preparedness 2800 SW Topeka Blvd Toprika, Kansas 66611 1287
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Wolf Creek Nuclear Operating Corporation !
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! E Mail report to T. Frye (TJF)
E Mail report to T. Hiltz (TFH) i E Mail report to NRR Event Tracking System (IPAS)
E Mail report to Document Control Desk (DDCDESK) bec to DCD (IE01) bec distrib, by RIV w/ Enclosure and Attachments 13 1
Regional Administrator Resident inspector DRP Director SRI (Callaway, RIV)
Branch Chief (DRP/B) DRS PSB 4
Project Engineer (DRP/B) MIS System
, Branch Chief (DRP/TSS) RIV File L. Vick (HOLB/NRR) bec distrib. by RIV w/ Enclosure and Attachments and Attachments 14:
- R. Gallo (HOLB/NRR)
L. Hurley 1
DOCUMENT NAME: R:\_WC\WC7301P.P.HFB ).
To receive copy of document Indicate in boa: "C" = Copy wthout enclosures "E" = Copy with enclosures "N"
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[SQLQSURE U.S. NUCLEAR REGULATORY COMMISSION REGION IV ,
Docket No.: 50 482 License No.: NPF 42 - t Report No.t 50 482/97 301 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane, NE ,
Burlington, Kansas Dates: August 25 29,1997 Inspectors: H. Bundy, Chief Examiner, Operations Branch T. McKernon, Examiner, Operations Branch T, Meadows, Examiner, Operations Branch M. Murphy, Examiner, Operations Branch L. Vick, Examiner, Operator Licensing, Office of Nuclear Reactor Regulation Approved By: - J. L. Pellet, Chlef, Operations Branch Division of Reactor Safety ATTACHMENTS:
Attachment 1: Supplemental Information Attachment 2: Simulation Facility Report
. Attachment 3: Facility initial License Written Examination Comments Attachment 4: Final Written Examination and Answer Key t
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SUMMARY
Wolf Creek Generating Station NRC Inspection Report 50 482/97 301 NRC examiners evaluated the competency of five reactor operator and three senior reactor operator applicants for issuance of operating licenses at the Wolf Creek Station facility.
The licensee developed the initial license examinations using NUREG 1021, " Operator Licensing Standards for Power Heactors," Interim Revision 8. NRC examiners reviewed, approved, and administered the examinations. The initial written examinations were administered to all eight applicants on August 25,1997, by f acility proctors in accordance with instructions provided by the chief examiner. The NRC examiners administered the operating tests on August 26 28,1997. Three applicants for reactor operator licenses and one applicant for a senior reactor operator license displayed the requisite knowledge and skills to satisfy tha requirements of 10 CFR Part 55 and were issued the appropriate
- licenses, pocrations
- One of three applicants passed the senior reactor operator written examination and three of five applicants passed the reactor operator written examination. Both groups of applicants demonstrated a potential generic knowledge weakness of nuclear instrumentation. The reactor operator applicants demonstrated potential knowlecige weakness in the administrative requirements area (Section 04.1).
- All eight applicants passed the operating test. Overall, strong applicant performance was demonstrated during the dynamic simulator scenarios (Section 04.2).
- The licensee submitted an examination outline which was adequate for examination development. Several enhancement suggestions provided by the examiners were incorporated in the final submittal (Section 05.1.11.
- The examination materials were acceptable for administration as submitted, but the reactor operator walkthrough tasks were minimally challenging. The licensee made several changes to the reactor operator walkthrough tasks pursuant to Region IV comments to raise the discriminatory value of the examination to an average level.
As a result of the post examination analysis, it was concluded that the written examination contained an excessive number of technicalinaccuracies (Section 05.1.2).
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[kslt Details Symmplv of Plant Status The plant operated at essentially 100 percent power for the duration of this inspection.
L QM1ptions 04 Operator Knowledge and Performance 04.1 initial Written Exornination
- a. lnlipaction Scong On August 25,1997, the license 9 proctored the administration of the written examination approved by the chief examiner and NRC Region IV supervision to five individuals, who had applied for initial reactor operator licenses, and three individuals, who had applied for initialinstant senior reactor operator licenses, The licensee graded the written examinations and the staff reviewed the results. The licensee also performed a post examination question analysis, which was reviewed by the examiners,
- b. Observations and Findinas The minimum passing score was 80 percent. Scores of applicants for reactor operator licenses ranged from 71.7 to 88.9 percent. Three reactor operator license applicants passed with an average score of 88.2 percent. Two reactor operator applicants failed with an average f ailing score of 74.3 percent. Scores of applicants for senior reactor operator licenses ranged from 75.8 to 90.9 percent. One senior reactor operator applicant passed with a score of 90.9 percent. The two senior reactor operator applicants who failed had an average score of 76.8 percent.
The above grades reflect the results after examination changes recommended by the licensee as a result of its post examination question analysis were incorporated.
The Region IV staff reviewed and approved these recommendations based on the technical merits of each recommendation. The licensee initiated a performance improvement request, which willinclude a root cause analysis, to investigate "a pass rate that is unacceptable to Wolf Creek Generating Station." It also requested a meeting with Region IV to discuss the results of the performance improvement request.
Questions 1 to 75 were the same on both examinations. Questions 76 to 100 were unique to the specific examination. As a result of the post. examination analysis, changes were made to the answer keys as follows: For both examinations, j Ouestion 6 was deleted, two answers were accepted on Questions 6,9,12,68, and 82, and the answer was changed for Question 48. Also, two answers were
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Gowed for Questions 88 and 94 on the reactor operator examination and the answer was changed for Question 91 on the senior teactor operator examination. ;
This resulted in nine changes to the answer key for the reactor operator exarnination ,
and eight changes to the answer key for the senior reactor operator examination.
Further, the Chief Examinor identified a potential weakness of reactor operator applicants in the administrative area in that Ouestions 92,94,95, and 97, dealing with administrative requirements, were missed by more than half the applicants, i Also, a potential applicant knowledge weakness relating to the characteristics of nuclear instrumentation was identified in that more than half of all applicants missed Question 41, which related to intermediate range nuclear instrumentation, and half the applicants missed Question 42, which related to source range nuclear instrumentation, c.. Conclusions One of three app licants passed the senior reactor operator written examination and three of five applicants passed the reactor operator written ex6mination. Both ,
groups of applicants demonstrated a potential generic knowledge weakness of nuclear. instrumentation. The reactor operator applicants demonstrated potential knowledge weakness in the administrative requirements area.
04.2 initial Ooeratino Test a.- Insoection Seggg The examination team administered the various portions of the operating examination to the eight applicants on August 26 28,1997. Each applicant participated in two or three dynamic simulator scenarios. Each also received a walkthrough test which consisted of ten system tasks together with two followup
! questions for each system. Four of five subjects in four administrative areas were covered by administrative tasks. The rernaining administrative subject was covered by two questions,
- b. Observations and Findinas All applicants passed all portions of the operating test. With minor exceptions, applicant performance in the dynamic simulator scenarios was strong.
-. Communications were good and directions provided to the crews by the senior reactor operator applicants were timely and appropriate. Performance of all applicants in the administrative area was good. Some applicants demonstrated performance problems on specific system tasks duiing the walkthrough, but overall performance was satisf actory. No broad performance weaknesses were identified.
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- c. Cp_rX!usions All eight applicants passed the operating test. Overall, strong applicant performance was demonstrated during the dynamic simulator scenarios.
05 Operator Training and Qualification 05.1 Initial Licensina Examination Develonment l The licensee developed the initiallicensing examination in accordance with guidance
! pruvided in NUREG 1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8.
05.1.1 Examination Outline
- a. Insnection Sgsp_2 The licensee submitted the initial examination outline on June 25,1997. The examiners reviewed the submittal against the requirements cf NUREG 1021.
- b. Observations and Findinag The initial examination outline was adequate as a guide for development of the examination. However, the reviewing examiner provided several enhancement suggestions to achieve the expected level of difficulty for the examinations and provide additional opportunities to evaluate the competency of the applicants. As submitted, the dynamic scenarios provided only the minimum quantitative criteria for evaluating the response of individual applicants to component and instrument f ailures. This could necessitate running additional scenarios because the expected applicant does not always respond to a given event. The examiner also observed that the scenarios were mostly a collection of unrelated events, with little impact on the major transient mitigation strategy and that there was minimal challenge to the senior reactor operator applicants in prioritizing actions and making decisions.
Specific comments were provided on each scenario for licensee consideration. The written and walkthrough examinations were consideied satisfactory with minor enhancement suggestions offered by the examiner. The licensee discussed proposed responses to the examiner comments with the chief exa'niner on July 8, 1997, and submitted a revised outline on July 11,1997. The revised outline satisfactorily resolved examiner comunts. Extra events with clearer linkage to the major transients had been added to make mitigation strategy more challenging. The outline was condensed in the examination package received on July 25,1997, to reflect fewer examination materials because of the withdrawal of two potential applicants, j
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- c. Conclusions The licensee submitted an examination outline which was adequate for examination development. Several enhancement suggestions provided by the examiners were incorporated in the final submittal.
05.1.2 Examination Packane
- a. Lrnp_gstion Scope i
The licensee submitted the completed examination package on July 25,1997. The chief examiner reviewed the submittal against the requirements of NUREG 1021.
I b. Observa+ ions and Findinos The licensee submitted 125 draft written examination questions of which 75 were designated to be common to both the reactor operator and senior reactor operator examinations. Of the 125 questions, 25 were uniqua to the reactor operator examination and 25 were unique to the senior reactor examination. The chief examiner provided comments or questions on a substantial number of questions. In resolving these comments and questions, the licensee modified or replaced five questions, which were common to both examinations, and six questions, which appeared only on the senior reactor operator examination. Many of these changes were minor editorial or enhancement changes. The chief examiner concurred with the resolution of the comments and the final product. As discussed in Section 04.1, deletion of one question common to both examinations and answer modifications for five questions common to both examinations, three questions appearing only on the reactor operator written examination, and two questions appearing only on the senior reactor operator examination were required as a result of post examination reviews to make the examinations technically accurate. The quantity of changes is considered excessive in accordance with the 5 percent threshold guidance provided in NUREG 1021, ES 501.
The licenseo submitted seven dynamic scenarios, including one backup scenario, which was not used during the examF tion. Also, because of a reduction in the number of applicants, one of the primary scenarios was not administered to the applicants. The submitted scenarios were adequate for administration. The licensee subsequently iricorporated several enhancement suggestions provided by the NRC examiners as a result of a table top review and onsite evaluation. Further scenario changes having minimal effect on examination administration were made as follows: in Scenarios Two and Two A, Pressurizer level 460 channel f ailure was substituted for Volume Control Tank Level Channel BG LT 149 f ailure because of a i
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7 problem weh the plant procederes for respol. ding to the volume coritrol tank level channel failure, which was discovered during simulator setup; in Scenarlor, Three and Three A, Event 3, the failing steam generator level channel was changed from 517 to 551 to implement a correction that was identified during preparation week, but inadvertently not entered; and in Sct.nario 3, Event 3, Nuclear Instrt rnent Power Range Channel 44 was inadvertently failed instead of 42.
To support the systems walkthrough section of the operating test, the licensee provided two sets of job performance measures developed to evaluate selected operator tasks. One set was designed for reactor operator applicants and the other set was designed for senior reactor operator applicants. Both sets contained v,e?l-written task elements, performance standards, and comprehenshte evaluator cues.
The tasks designed for the senior reactor operator applicants were high quality and acceptable for administration as submitted. However, based on an '.n-office review, the tasks designed for the reactor operator applicants discriminated at too low a levelin contrast to the other set reviewed. Subsequently, as a result of onsite validation, the chief examiner concluded that the reactor operator tasks were marginally adequate.
For example, although Job Performance Measure 1 contained only one critical step, the task was of high importance and f amiliarity with vital electrical bus instrumentation and controls together with proper sequencing of preliminary steps was required to successfully complete the task, in response to the comment concerning task difficulty, the licensee increased task complexity by requiring the applicant to reenergize a dead bus rather than switching power supplies on a live bus. Another example involved Job Performance Measure 3, which was essentially a two critical step task: recognize which valves did not automatically close as expected on a Phase A isolation signa and manually close the valves. This task was of high importance and a significant number of manipulations on several control panels were required to successfully complete the task. The licensee enhanced the difficulty of this task by inserting a valve position indication failure which required the applicant to use alternate indications to confirm task completion. Also, the licensee upgraded another reactor operator task and replaced one task pursuant to the chief examiner's comments. A second task on the reactor operator walkthrough cection was replaced when it was potentially compromised as a result of an inadvertent encounter, at the task location during NRC validation, with individuals not on the security agreement, who were involved in an operator requalification examination. This was considered to be a scheduling problem on the part of the licensee.
The final reactor operator walkthrough tasks were considered to be of high quality and of overage difficulty. Two followup questions associated with each walkthrough task were also submitted by the licensee. While the questions were considered acceptable for administration as submitted, the chief examiner provided the licensee with a few enhancement suggestions, which were incorporated by the licensee. During administration of the test it was discuvered that Question 14,
8-associated with Reactor Operator Simulator Job Performance Measure Three was technically flawed and the wrong answer was given in the answer key. The chief examiner replaced it with another question for four of the five applicants. The one applicant who correctly responded to the original question was given credit and did not have to respond to the replacement question.
The licenseo submitted two sets of job performance measures and questions to cover the administr6tive section of the examination. One set was designed for reactor operator applicants and the other set was designed for senior reactor operator applicants. The job performance measures and questions were of high quality with one exception. The submitted job performance measure for the reactor operator applicant on the emergency plan topic was non discriminatory. The licensee replaced this job performance measure with two discriminatory questions.
The final product was of high quality,
- c. Conclusions The examination materials were acceptable for administration as submitted, but the reactor operator walkthrough tasks were minimally challenging. The licensee made several changes to the reactor operator walkthrough tasks pursuant to Region IV comments to raise the discriminatory value of the examination to an average level.
As a result of the post examination analysis, it was concluded that the written examination contained an excessive number of technicalinaccuracles.
05.2 Simulatign Facility Performance a inspection Scep_q The examiners observed simulator performance with regard to fidelity during the examination validation and administration,
- b. Qb.servations and Findinng The simuletion f acility supported examination administration well. Four minor deficiencies affect;ng examination administration are discussed in Attachment 2.
Three deficiencies, which did not affect examination administration were identified during preparruon week and are also discussed in Attachment 2. They had a minor effect on examination validation, but did not require any changes to the scenarios aff ected. Other than the issue involving the lamps during loss of power conditions, none of these deficiencies were repeated during the examination process. The licensee made interim simulatb. adjustments to achieve proper readings for reactor coolant pump seallbearing water temperatures and allow loading of the emergency diesel generator. The licensee initiated simulator modification requests to addrmss these deficiencies.
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- c. Conclusions The simulation f acility supported the examination administration well.
V. Management Meetings X1 Exit Meeting Summary The examiners presented the inspection results to members of the licensee-management at the conclusion of the inspection on August 29,1997, The licensee acknowledged the findings presented. Although individual examination scores were i- not discussed, the licensee indicated that it was not satisfied with the preliminary
. results from the written examination and had formally requested a meeting with Region IV to discuss the root causes of the poor results, any operational implications, and planned corrective actions.
On Septernber 4,1997, Mr. McKernon discussed the excessive number of technical accaracy issues with the written examination discovered as a result of the post-examination analysis, as well as, potential agenda items for the licensee requested meeting with Messrs. Pippin and Guyer.
The licensee ded not identify as proprietary any information or materials examined during the inspection.
ATTACHMENT 1 SUPPLEMENTAL INFORMATION l
PARTIAL LIST OF PERSONS CONTACTED Lkensee T. Damashek, Superviser, Licensing R. Flannigan, Manager, Nuclear Engineering, Safety, and Licensing l R. Guyer, Superintendent, Operations Training i S. Hatch, Instructor, Training O. Maynard, President and Chief Executive Officer l B. McKinney, Plant Manager A. Palmer, Lead Senior instructor, Initial Licensed Operator Training J. Pippin, Manager,- Training 4 G. Smith, Senior Instructor, Training C. Warren, Vice President Operations, Chief Operating Officer C. Younie, Manager, Operations NBC F. Ringwald, Senior Resident inspector
ATTACHMENL2 SIMULATION FACILITY REPORT Facility Licensee: Wolf Creek Nuclear Operating Corporation Facility Docket: 50 482 Operating Examinations Administered at: Wolf Creek Generating Station Operating Examinations Administered on: April 15 10,1997 These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(bh These observations do not affect NRC certification or approval of the simulation facility, other than to provide information, which may be used in future evaluations. No licensee action is required in response to these observations. Only the first observation was repeatable during the examinations.
Deficiencies identified Durina Examination Prenaration
- The push to test feature lighted lamps on pump and breaker control panels under loss of power conditions that should have precluded this feature.
- A power decrease was in progress and the drain valves on the extraction steam lines started cycling for no apparent reason.
- After a Charging Pump B trip with a:t other charging secured, the following anomaties were observed: Although the seal flow meters on Panel RLOO1 went to zero, other indicators showed approximately 7 gpm per pump flow to the reactor coolant pumps and Charging Pump A boron injection flow increased to 600 gpm.
Also, the seal / bearing water temperature meter was pegged low and the NPIS point indicated 3.5 degrees F while the reactor coolant temperature was 535 degrees F.
Deficiencies identified Durino Examination Adminigration
- While performing a task involving starting and loading the emergency diesel generator, it was noted that the synchroscope check light did not illuminate when Bus NE02 voltage was adjusted 50100 volts higher than Bus NB02 voltage in accordance with procedure. The instructor adjusted NE02 volts until the light was illuminated and then set Meter NB El 29 to read 50100 volts higher than Bus NB02 l voltage. No further problems were noted during subsequent performances of this task.
- While performing Procedure STS SF-001 during a dynamic scenario, when the operator returned the rod control bank selector switch to AUTO it appeared that the switch did not select AUTO. Rod speed indicated 48 steps per minute instead of 0.
The applicant cycled the switch to manual and then back to AUTO to achieve the desired results.
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- During a dynamic scenario after a trip of Charging Pump B followed by the operator starting the normal charging pump and adjusting flow to restore pressurizer level, the charging flow indication went to zero, s
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4 ATTACHMENI_2 FACILITY INITIAL LICENSE WRITTEN EXAMINATION COMMENTS 8
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INITI AL 1.lCENSED WillTTEN EXAM COMMENTS Question O Question states immediate Actions of EhiO E 0 are being performed implies very early after a Reactor Trip. Tavg has dropped 10*F below no load Tavg.and steam pressure is -100 lbs. tow. It is not possible to differentiate between a steam line break and feed line break without knowing the trends for S/O level and pressure since ' A' S/O level is below the feed ring . break is blowing primarily steam.
Even on a steam line break the affected S/0 level will decrease more rapidly due to now restriction in the cross over pipe. Since FWlV's would be closed, break flow would no longer be present if key answer (C) is correct, but without trends it cannot be detennined if break now is still present.
e insuf0cient information is given to determine the correct response.
- Recommend deleting this item from the esamination.
Question 9 (Reference BD-EMG C 0) tiasis for depressurizing the RCS 260 lbs. in EhtG C-0.
Answer (C), key answer, is to reduce leakage through RCp seals. Dis data appears in DD-Eh10 C 0 on page 61 under the BASIS discussion for step 28.
STATEMENT FROh! DACKGROUND: "His step depressurizes intact S/Os, thereby reducing RCS temperature and pressure to reduce RCp seal leakage and minimize RCS inventory loss."
in the third paragraph of the DASIS on page 61, the following also appears:
STATEhiENT FROhi DACKGROUND "It is important that the depressurization not reduce S/0 pressures in an uncontrolled manner that under shoots the ptessure limit, thus permitting potential introduction of nitrogen from the accumulators into the RCS."
Answer (A) is also conect in that the basis for the 260 psig limit is to prevent accumulator nitrogen from injecting into the RCS.
. Recommend accepting Answers A and C, Question 12t (Reference LO 14 036 00) Key answer is 'D' For this answer to be conect, the diesel fire pump would have to start on low pressure along with the electric driven fire pump. After 30 minutes, the electric fire pump would shut off, and the diesel Ore pump would supply suf0cient now.
He electric fire pump starts at a higher pressure than the diesel Dre pump and is capable of supplying 3300 gpm (only 2000 is being used). Depending on how fast pressure dropped, the diesel fire pump may or may not have started -if the pressure drop were slow enough, the electric driven pump would start and supply suf0cient now to prevent the diesel pump from starting, if the pressure drop is rapid, the pressure may decrease below the start pressure for the diesel pump before the electric pump comee up to speed. After 30 minutes, the electric pump would shut off, then re start on low pressure (if the diesel pump is not running) again before the dicscl Orc pump would receive a statt signal.
- Recommend accepting Answers A and il Question 16: His question is conect as written. Suf0cient information is presented to allow the student to determine the conect response. Training occurred under TIN LO 17 32317 on the effects of natural circulation.
e itecommend accepting only key Answer D.
l Question 21: This question is cortect as written. Sumcient infonnation is presented to allow the student l
to determine the correct response. Training occurred under llN LO 13 002 00.
- Recommend accepting only key Answer B.
Question 22: nis question is correct as wntten. Sumclent information is presented to allow the student to detennine the correct response. Training occuned under TIN LO 17 323 21.
- Recommeed accepting only key Answer D.
Question 4it his question is conect as written. Sumcient information is presented to allow the student to determine the conect response. Training occurred under TIN LO 13 015 01.
f e Recommend accepting only key Answer A.
Question 42: Dis question is correct as writ'en. Sumelent infonnatlon is presented to allow the student to determine the conect response. Training occuned under TIN LO 13 015 01, e Recommend accepting only key Answer D.
Question 48: (Reference LO O 003 00. Reactor Coolant fumps) #3 seal is a split seal. 400 cc goes to the Conta!nment Sump (from the upper seal) and 400 cc mixes with #2 seal retum and goes to the Reactor Coolant Drain Tank.
. . Key Answer 'D' . only #2 seal lenkoff goes to Reactor Coolant Drain Tank.
- More Correct Answer: 'B' lenko(T from #2 and #3 seals drain the Reactor Coolant Drain Tank, e Recommend changing key to accept only Answer B.
Question 58: This question is conect as written. Sumclent information is presented to allow the student to determine the conect response. Training occurred under TIN LO 13 026 00, e Recommend accepting only key Answer A. .
Question 59: nis question is correct as written. Sumelent information is presented te sitow the student to determine the correct response. Training occurred under TIN LO 13 028 00.
- Recommend accepting only key Answer B.
Question 63: nis question is correct as written. Sumcient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 15 062 05.
- Recommend accepting only key Answer B.
J
k i
1 t ,
i 7
' Question 68: (Rcfcrences EAIG /3 ll or OFN BB 3/> EMG E511 and Ol'N Dil 031 both have
- Refueling Water Storage Tank swapover reset prior to aligning a train for shutdown cooling. Only in GEN 00 006 could you end up with one train in S/D cocling mode and one train injectmg. Ol'N B0 031 would
! have you initially secure the S/D cooling train and isolate it from the RCS. then realign later if S/D cooling
' _ is required.. Initial conditions do not tell what mode the LOCA occuned in. Key answer 'C' would be cortect for a Mode 4 LOCA during the very early stages of Ol'N DB 031. Answer 'O' is correct for EMG l
4
- Recommend accepting Answers B and C.
Question 82: (Reference P/R 941148 # 71194). securing a condensate pump at power is a non-significant event with little consequences. WCGS had a condensate pump secured at full power on Sept.
l- 22,1994 (after rernte). The plant continued to operate and S/G levels recovered (the pump was off for a total of 28 minutes). TIN LO 15 059 00 does' state that it is doubtful that adequate flow can be maintained.
I llowever, actual plant experience is that it can be maintained.
- Reco 6 mend accepting Answers A and D.
a Question 86: This question is correct as written. Sufficient information is presented to allow the student to determine the correct response. Training occurred under TIN LO 14 078 00.
' e Recoermend acespting only key Answer C.
Question 88t Key answer 'C' is one possible answer, if operating " valves wide open"(control valves fully open) or on the limiter such that control valves will not open further to maintain Mwe, and an MSR relicf valve opened. Answer D is also correct if not operating " valves wide open", control valves will open to try i
to maintain Mwe. Decreased efliciency due to lower vacuum will cause Mwe to decrease whh " essentially constant" steam flow and reactor power".
J
- Recommend accepting Answers C and D..
4 Question 89: 'Ihis question is correct as written. SufTicient information is presented to allow the student to detennine the correct response. Training occurred under TIN LO 14 08900.
4
- Recommend accepting only key Answer C.
Question 92t This question is correct as written. Suflicient information is presented to allow the student to
- determine the conect response. Training occurred under TIN LO 17 332 01.
- Recommenu accepting only key Answer B.
Question 941 (Reference AP 21E 00I)
Key answer 'D' is correct. AP 2 tE 001 allows indep:ndent verification to be done by remote indication prior to deenergizing a valve, A iswer 'D' is also correct if a blocking device is used, which is allowed by AP 2 tE 001 (also on page 14). NOTE; Although the procedure does not specifically require the valve to be de energized prior to verifying" blocking device," th!s is a standard operating practice.
- Recommend accepting Answers B and D.
. _ . - - .. ~ - - - . . ~ _ _ - . - -. - - - - . _ - - - . - -
I l
A l ,. i
=
i 6 i
l
! Questice sr 95: t his question inorrect as w ritten Sulficient information is presented to allow the student i
!- to deterwe the cortect reso nse. Tr6ning 1.ccurred under llN LO 17 332 0$. l f
l
- Decsmmend accepting only arv arswer C.
?
Questloc 97t TUs question is corrett as written. Suf0cient information is presented to allow the student to '
r i determine the ccurc remoas; Tr4ini.4 cccurred under llN LO 17 332 01.
i e Recommend acuptics M .y key Answer 14. ,
l f Question 107t (Rcferenect EMG E Ofoldm!page) This question asks if Adverse Containment values l - should be used. EMO foldout pages list 2 criteria to determine if Adverse Containment values should be used.
- 1) is containtnent pressure greater than 5 psig. If pressure has been above 5 psig. adverse containment ,
values need not be used once pressure decreases below $ psig.
- 2) is cont.inment radiation greater than 10'R/hr. or integrated dose has exceeded 10' Rad if radiation levels in containment have exceede ! the 10' R/hr les el. adverse containment values must be used until an engineering analysis is performed ej the TSC to vetify integrated exposure has remained acceptable.
I
! Answer C is correct since the integrated dose level has been exceeded.
Answer D is conect since the instantaneous dose rate level has been exceeded and no engineering analysis has been performed. ,
e Recommend accepting Answers C and D.
Question 114: This question is conect as written. Sufficient information is presented to allow the student to determine the conect response. Training occuned under TIN LO 13 00100.
l
- Recommend accepting only key Answer C.
Question iI6t (Reference AP 211001. Temporary Modylcation (Rev 2) *the Temporary Modification =
procedure does not exclude non safety related equipment, it allows this type of replacement on non safety related equipment only, but contains no relaxation from the requirement to issue a Temp. Mod.
(-
- Key answer is not correct 'D' in that a Temp. Mod should be issued.
- Answer 'B' is conect.
- Recommend changing the key and accepting only answer B.
i Question 119: This question is conect as written. Sufficient information is presented to allow the student to deterinine the conect response. Training occurred under TIN LO 13 032 00.
Y e - Recommend acceptingonly key Answer A.
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1 ATTACHMENT 4 FINAL WRITTEN EXAMINATION AND ANSWER KEY
SRO ANSWER KEY
- 1) A 26 ) C 51 ) A 76 ) D
- 2) D 27 ) A 52 ) A 77 ) A
- 3) C 28 ) A 53 ) B 78 ) D -
- 4) B 29 ) A _ . 54 ) A 79 ) B i
- 5) D . 30 ) D 55 ) A 80 ) D D=M M ; C r# 31 ) B 56 ) B 81 ) C
- 7) B 32 ) D 57 ) B 82 ) C e.D 'WJE
- 8) D 33 ) D 58 ) A 83 ) D
- 9) C -A $ 34 ) D 59 ) B 84 ) D 10 ) A 35 ) D 60 ) C 85 ) A 11 ) D 36 ) D 61 ) C 86 ) A - '
12 ) B -A sWS 37 ) B 62 ) C 87 ) B 13)C 38 ) A 63 ) B 88 ) C '
14 ) B 39 ) A 64 ) B 89 ) C '
15 ) B 40 ) D 65% A. 90 ) A 16 ) D 41 ) A 66 ) D 91 ) P & M s7) 3 42 ) D 67 ) D 92 ) C 18 ) B 43) A 68 ) C e5 ei:8s 93 ) 8 19 ) A 44 ) B 69 ) C 94 ) A 20 ) C 45 ) B 70) D 95 ) B 21 ) B 46 ) C 71 ) A 96 ) D 22 ) D 47 ) D 72 ) D 97 ) A 23 ) D 48 ) PS 1t95 73 ) D 98 ) B 24 ) B 40 ) A 74 ) D 99 ) C ,
25 ) D 50 ) C 75 ) B 100 ) C
&;-hmMb$f Yt'h$X4 @'&
n 7/s/97. 9gq;q$'n.~
i CHIEF EXAMINER SRO WRITTEN EXAM COMMENTS - WOLF _ CREEK 8/25/97 l
Question No, Comment 2 Answer is not supported by OFN 88-005 in that loss of CCW flow is not listed as a RCP immediate shutdown criterion. Also, the catalog number should be 015AA2.10. Rasolution: Deleted the word immediate from the stem and corrected catalog number.
3 Reference does not support the answer. Resolution: IJsed attached l references together with the applicable drawing to determine that answer is correct.
5 The correct catalog number should be 000027AK2,03. Resolution:
Corrected catalog number.
7 Catalog number should be 051 AA2.02. Reference does not support ariswer.
Resolution: Corrected catalog number and reference.
10 Catalog number does not match outline. Resolution: Changed outline.
Meets standatds.
14 Suggest rewording the stem to indicate the reason for the time limit on running an RHR pump on circulation without CCW cooling is to prevent pump damage. This is the purpose given in the reference. Resolution:
Reworded stem.
22 This question only requires the applicant to know RCP trip criter'a to chose the correct answer. Thir knowledge is covered in the dynamic scenarios.
Develop a more in-depth question. Resolution: Question replaced.
28 Reference does not support answer. Resolution: Provided correct reference.
43 Whereas the question is soliciting one set of conditions, most applicants would reject distractors B and C because they are subsets of A and D. Are the applicants expected to have the CSF status tree for core cooling memorized? If not, this does not oppear to be a fair question. Resolution:
Resolution: Appil:, ant is expected to know approximate temperature range.
Choices B and C are not subsets because of different f!ow values.
53 The answer should be C vice B. Resolution: Provided reference to prove B.
54 fieference does not support the answer. Rosolution: Reference does support answer.
61 Choice D does not appear to be a plausible distractor. Resolution: replaced distractor.
73 Reference does not support answer. Resolution: Provided an additional referonce.
- v .. . -
, ,. -----,_r . - - ,
_ . , _ . . _ ... _. _ - _ _ . _ . . _ _ _ - _ _ _m._.__.___ . . ._ . _ . _ -
t
- i. ,
4 74f Why is Choice B not correct? Resolution: Qualified Choice B to m;ke it
- incorrect?
4
- 76_. KA number does not match outline. Resolution: Licensee identified KA on '
- outline. -
a p 90: The ar.swer is so obvious that this question is non-discriminatory. Develop i .another question._ Resolution: Renrote question to make it discriminatory.
}91 The reason given in the answer for allowing the substitution draws one to l -the correct answer. Reword the answer. Resolution: Incorporated.
p L 92 Any applicant will know that violation _of a safety limit requires a 50.72 i report. The question remaining is whether is requires an immediate, one I hour, or four hour report. Distractors A and B will be eliminated by
[ - everyone. . Rewrite the distractors, liesolution: Incorporated. _ ,
i
- 94- It could be argued that the answer is incomplete. Suggest that it state j1 terminate core alta.ations and movement of irradiated fuel. Resolution
Incorporated.-
~
99 Reference does not support answer. Further, if choice C is correct, one would assume that choice D is also correct. Rework this question.
Resolution: Rewrot'e question.
4 I 100 Under certain conditions, it appears that choices A and D could also be -
- - correct. Ensure that the proper logic has been used in designating the e answer. Resolution
- Restructured question.
t 4
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- g. +w-, - - , , .,v-. - - -,,e.,w-r w
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+
1 ES-401 Site Specific Written Form ES-401-7 Examination Cover Sheet 1
U.S. Nuclear Regulatory Commision Site-Specific Written Examination-I Applicant Information Name:-. Region: ,
Date: 8-25-97 Facility / Unit WCGS
] License Level RO/SRO Reactor Type - W ,
Start Time: Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this I- 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.
4 Applicant Certification All work done on this examination is my own.- I have neither given nor received aid.
- Applicant's Signature -
i Results Examination Value Points Applicant's Score Points
. Applienat's Grade Percent NUREG-1021 Interim Rev. 8, January 1997 y w,,,,,- y
EXAMINATION ANSWER SHEET Wolf Creek Generating Station Written Examination
~ PRINT NAME: DATE: 08/25/97 HRST M LAST
- 1. A B C- D 26. A B C D $ 1. A B C D 76. A B C D
- 2. A B C D 27. A B C D 52. A B C D 77. A B C D l
- 3. A B C D 28. A B C D 53. A B C- D 78. A B C D 4.- A B C D 29. A B C D $4. A B C D 79. A B C D
- 5. A B C D 30. A- B C D $5. A B C D 80. B C D
- 6. -A B C D 31. A B C D 56. A B C D 81. A B C D
- 7. A B C D 32. A B C D 57. A B C D 82. A B C D
- 8. A B C D 33. A D C D 58. A B C D 83. A B C D
- 9. A B C D 34. A B C D 59. A B C D 84. A B C D
- 10. A B C D 35. A B C D 60. A B C D 85. A B C D
- 11. A B C D 36. A B C D 61. A B C D 86. A B C D
- 12. A B C D 37. A B C D 62. A B C D 87. A B C D
- 13. A B C D 38. A B C D 63. A B C D 88. A B C. D
- 14. A B C D 39. A B C D 64. A B C D 89. A B C D
- 15. A B C D 40. A B C D 65. A B C D 90. A B C D
- 16. A B C D 41. A B C D 66. A B C D 91. A B- -C D
- 17. A B C D 42. A B C D 67. A B C D 92. A B C D
- 18. A B C D 43. A B C D 68. A B C D 93. A B C D 19, A B C D 44. A B C D 69. A B C D 94. A B C D
- 20. A B C D 45. A B C D 70. A B C D 95. A B C D
- 21. A B C D 46. A B C D 71. A B C D 96. A B C D
- 22. A B -C D 47. A B C D 72. A B C D 97. A B C D 23, A B C D 48. A B C D- 73. A B C D 98. A B C D
- 24. A B C D 49. A B C D 74. A B C D 99. A B C D
- 25. A B C D 50. A B C D 75. A B C D 100. A B C D
Wolf Creek Generating Station License Examination
, 001 A Rod Control Urgent Failure Annunciator is in alarm. The operators find that it is still pc.isible to move control rods y selecting the individual banks and moving the selected bank manually, Which one of the following can be concluded E from these symptoms?
a.: The problem is in a Power Cabinet.
- b. The problem is in a Logic Cabinet.
- c. There is an additional failure in that no rod motion should be possible. l
- d. No conclusions may be drawn from the stated symptoms.
i 002
! Which one of the following conditions requires the affected RCP to be tripped.?
j a. Seal outlet temperature reaches 150*F
- b. RCP motor winding temperature reaches 225'F
- c. Seal injection flow is lost.
4
- d. CCW flow is lost for over 5 minutes.
003 The Nomtal Charging Pump is the only charging pump available and Emergency Boration via the BIT becomes necessary, Which one of the following is necessary to accomplish this?
- a. This cannot be donc until at least one Centrifugal Charging pump can be placed in senice, b.- Start at hast one Safety injection pump.
- c. Open the Centnfugal Charging pump discharge flow control valve (FCV-121)
. d. - Close RCP Seal Injection flow control valve (BG HC-182).
Wolf Creek Generating Station License Examination 004 A leak in the CCW system is larger than the capacity of makeup from the Demineralized Water system. Which one of ic following is the preferred source of additional makeup?
- a. Service Water.
- b. ESW.
- c. Circ Water,
- d. Isolate the Senice Loop from the Safety Train and make up to the Safety Train from the Senice Loop.
005 A Pressurizer Pressure instrument has failed low and the altemate instruments have been selected. Which one of the following remains inaccurate despite the fact that the attemate instruments have been selected?
- a. PZR Pressure Control.
- b. PZR Pressure Recorder.
- c. OPAT/OTATTemperature Recorder,
- d. Subcooling Margin Monitor.
006 While performing immediate action steps in E-0, the following conditions are obsen ed:
-+ RCS Tm 547 F SG A Press /WR Level = 1005 psig/20%
-+ RCS Press - 2100 psig SG B Press /WR Level = 1005 psig/55%
-+ PZR LVL - 21% SG C Press /WR Level = 1005 psig/50%
Which one of the following is the most likely cause of this event?
- a. 'A' Steam line break between the S/G and the MSIV.
- b. 'A' Steam line break between the MSIV and Main Turbine,
- c. 'A' Feed line break between Feed Reg Valve and Main Feed Line Check Valve,
- d. 'A' Feed line break between Main Feed Line Check Valve and the S/G.
.. Wolf Creek Generatin~g Station License Examination -
007
- Which one of the following conditions requires an immediate manual turbine trip?
L a. Turbine Exhaust Hood Temperature 210*F.
b; Condenser Pressure is 9 inches Hg.
- c. Turbine Shaft Pump discharge pressure 120 psig.
- d. Condenser Vacuum Pump trip; 008 He operators are perfonning EMO C-0 (Loss of all AC Power). If an SI signal occurs at this time, which one of the
. ' following describes the proper response to this Sl?
. a; Do not reset it to ensure rapid injection of coir cooling water when power is restored.
I
- b. Do not reset it to ensure sufficient load exists to prevent a diesel generator overspeed trip.
- c. Reset it to allow the manual initiation of D/G cooling when AC becomes available.
T d. Reset it to permit manual loading of needed equipment on AC Emergency buses when they become -
l available, t
009 Which one of the following is the basis for the direction in EMG C-0, (Loss of All AC Power), to depressurize the intact steam generators to 260 psig?
- a. Prevent accumulator nitrogen from injecting into the RCS.
A b.. Minimize CVCS letdown flow.-
- d. Reduce the time the AFW System must be in operation.
i
Wolf Creek Generating Station License Examination 010 Power has been lost to 120 VAC Instrument Bus NN01 and the following conditions exist:
F As expected, Charging pump suction has swapped to the RWST.
- All failed instmments have been selected out.
4 _'Ihe operators have established Excess Letdown per OFN NN.021.
Which one of the following describes why Normal Letdown is not used?
- a. Charging flow will have been minimized down to RCP seal injection only,
- b. A locked in Letdown Isolation signal exists.
i c. Power to the solenoid for Letdown isolation valve BG HIS459 has been lost closing the valve,
- d. To prevent flashing in the Normal Letdown line.
l 011 If ESW is lost while ofTsite power is still available, the opemtors are directed to trip the RCPs. Which one of the following describes why this action is necessuy?
- a. To minimize containment heating dt.e to loss of RCP motor air coolers,
j c. To climinate the possibility of RCP motor damage due to loss of motor cooling,
- d. To reduce the heat load on the CCW system to extend cooling of safety related components.
012
' For the past 45 minutes, the Fire Brigade has been fighting a fire requiring 2000 gpm from the Fire Main. No operator actions have been taken other than those directly involved with fire fighting. Which one of the following specifies which Fire pumps can be expected to be mnning at this time?
- a. ' Jockey Pump and Electric Fire Pump,
- b. Jockey Pump and Diesel Fire Pump.
f
! c. Electric Fire Pump and Diesel Fire Pump.
- d. Diesel Fire Pump only.
1 D
(
I Wolf Creek Generating Station License Examination
-013:-
- Tech Specs limit containment pressure to 1.5 psig during operations. Which one of the following is the reason for this j, :striction?
a._ To maintain accuracy ofinstruments with detectors inside containment. ,
- b. To assure any leakage from containment remains within 10CFR100 limits.
$ c. To assure that peak pressure during an accident will remain within design limits,
- d. To prevent long term degradation of containment pressure boundary capability. .
J
- 014 To prevent RHR pumps from damage the EMGs have a time limit on how long an RHP. pump can be run on recirculation without CCW flow to tl.
- RHR heat exchangers. Which one of the following is that time limit?
- a. 2.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />,
- b. ' 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
c .-- 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,
- d. 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
015 Which one of the following alarms requires entry into 0FN-BB-006 (High Reactor Coolant Activity)?
- a. Steam Generator Blowdown RAD High on RM-II.
- b. - Letdown RAD High on RM 11.
- c. Process RAD HI on MCB.
- d. Condenser Air Removal RAD High on RM-II.
' Wolf Creek Generating Station License Examination
'016; A LOCA has occurred and the following conditions exist:
.+ ; %c plant is being cooled by natural circulation.
- = All MSIVs are closed. .
- - Nitrogen from an accumulator injection has collected in the U tubes of the 'C' steam generator inhibiting natural circ in the 'C' loop.
-+- Natural cire flow in all other loops is normal.
- All ARVs are set to 20% open.
Which one of the following will be the best indication of the failure of natural cire in the 'C' loop?
- a. ~ 'C' Loop AT will rapidly decrease to near zero.
b 'C' Loop T 6will increase relative to the unaffected loops. ,
_ c. 'C' S/G pressure will increase relative to the unaffected loops.
- d. 'C' S/G will depressurize.
017.
- A plant startup is in progress and the following conditions exist:
.*; : The reactor is critical at 5 X 10 Amps on the Intermediate Range. i 4- Tm is at its no load value. i i
-+ ' ~ A rod control failure causes the selected control bank of rods to step out 10 steps before being stopped by the operators.
- No reactortrip occurs.
- Which one of the following describes the effect that this event will have on reactor parameters? -
I
- a. Reactor power and Tm will both rise.
- b. Reactor power will rise. Tm will be unaffected. ;
= c. Reactor power will be unaffected; Tm will rise.
- d. Reactor power will increase. He response ofTm will depend upon core life.-
018-a nc plant is operating at full power when a fully withdrawn control rod drops into its fully inserted position. Which one
. of the following describes the effect that a delay in recovering the rod will have on the potential for core damage as a result of the recovery of the rod?
. a. . Any delay will not appreciably affect the potential for damage.
b.- Changes in xenon concentration will make rod recovery more hazardous as time goes on.
i c. Changes in xenon concentration will make rod recovery less hazardous as time goes on. >
- d. Changes in fuel temperature and fueJ bumup will make rod recovery more hazardous as time goes on.
Wolf Creek Generating Station License Examination 019 A LOCA has caused containment pressure to increase to 27 psig. Which one of the following describes when and why 11 RCPs must be tripped?
- a. Immediately due to loss of cooling to the seals and motors,
- b. Immediately due to potential damage to any nmning RCPs caused by containment spray,
020 The plant is operating at full power when a pressurizer safety valve fails open and can not be reclosed. As the RCS blows down from opemting pressure to 1000 psia, which one of the following trends do you expect to see on safety valve tail pipe temperature?
- a. Temperature will decrease from 650 F to 545'F
- b. Temperature will decrease from 350 F to 220 F
- c. Temperature will increase from 220*F to 350'F
- d. Temperature will increase from 350 F to 545'F 021 Which one of the following describes the Tech Spec IDENTIFIED LEAKAGE limit of 10 gpm?
- a. Does not include S/G leakage and is kept low enough so that it will not interfere with detection of UNIDENTIFIED LEAKAGE.
- b. Includes S/G leakage and is kept low enough so that it will not interfere with detection of UNIDENTIFIED LEAKAGE.
- c. Does not include S/G leakage and is kept low enough to be a small fraction of 10 CFR part 100 dose guidelines.
- d. Includes S/G leakage and is kept low enough to be a small fraction of 10 CFR part 100 guidelines.
Wolf Creek Generating Station License Examination 022 A LOCA has occurred. While isolating the accumulators in EMG ES 11, one accumulator isolation valve did not go losed. He operators are depressurizing the unisolated accumulator. During the subsequent recevery actions the operator is told not to depressurize RCS pressure below any unisolated accumulator pressure.
Which of the following recovery actions could result in a pressure decrease of the RCS during the recovery?
- a. starting a RCP on a loop with an intact steam generator,
- c. securing the RCPs.
d.- placing RilR in scavice with CCW lined up to the heat exchangers.
023 Loss of seal injection flow will have which one of the following effects on CCW AT in the Hermal Barrier Heat Exchanger?
- a. It will decrease since there is no longer any flow to cool,
- b. It will remain almost constant since most of the heat load comes from RCP internals.
- c. It will remain almost constant since no injection flow flows past this heat exchanger.
2
- d. It will increase since hot primary coolant will replace the cooler injection flowing past the thermal barrier.
024 he following plant conditions exist:
-+ Re plant has been shutdown for 10 dqs after setting the world record for the longest full power run by a PWR.
- No fuel has been replaced as yet.
- ne RCS is drained to mid loop.
-+ Re RCS is vented to atmosphere.
-+ Heat removal is being provided by RHR.
Which one of the following is closest to the time it will take for the caset of boiling if PHR is lost,
- a. 12 minutes b, 14 minutes
- c. 32 minutes 1 d. 34 minutes
l t 4
Wolf Creek Generating Station License Examination 1
025 2 A LOCA has resulted in Contaimnent pressure increasing to 7 psig. The reactor has tripped and the crew has critered
.MG E-0. Which one of the following instruments should be used to make the determination that neutron flux is
- decreasing?
- a. Source Range
- b. Intennediate Range
- c. Power Range
- d. Gamma-Metrics 026 A Steam Generator Blowdown isolation has occurred and the operators note the following valve positions:
-o Blowdown Isolation Valves (HV l-4) are closed.
-4 Upper Sampic Valves (HV 19 22) are open.
-o Lower Sampic Valves (HV 35 38) are open.
Which one of the following was the cause of the blowdown isolation?
- a. Auxiliary Feedwater Actuation Signal,
- b. Under-voltage on ND-01,
- c. Condenser Air Removal high mdiation.
- d. Safety injection Signal.
027 EMG ES-31, Post SGTR Cooldown Using Backfill, has a CAUTION that informs the operator that if all RCPs are
- stopped, the first RCP that is restarted should be in a non-ruptured loop. Which one of the following could result from starting a RCP in a loop with a ruptured SG7
. a. An inadvertent criticality.
- b. Enlargement of the mpture.
- d. Cold shock of the RCS.
4
- Wolf Creek Generating Station License Examination 028
- Which one of the following conditions will prevent termination of Safety injection?
- a. - Pressurizer level 3%
- b. _ AFW ilow 350,000 lbm/hr.
- c. RCS pressure stable at 1100 psig.
, d. RCS subcooling is 35'F.
029 A LOCA has occurred and the operators are conducting EMG ES-11 (Post LOCA Cooldown and Depressurization). No RCPs are running, Step 17 instmets them to 'Depressurize the RCS to Refill the Pressurizer'. Which one of tbc f:ll: wing describes the preferred method for accomplishing this task?
- a. - Open a Pressurizer PORV to avoid stresses on the Pressurizer Spmy nozzle.
, b. Open a Pressurizer PORV because this will depressurize the RCS most rapidly.
- c. Initiate Auxiliary Spray to preserve RCS inventory.
- d. Initiate Auxiliary Spray because this will raise Pressurizerlevel most rapidly.
030 Entry into EMG C-11 (Loss of Emergency Coolant Recire) is most likely to be required under which one of the following conditions?
4
- a. Steam Leak inside containment.
- b. Steam Leak outside containment.
- c. - LOCA inside containment.
=d. LOCA outside contaimnent.-
4 I
i
- . - - ..- ..-.- - - ..=.--..- =- - ---..- - -
i 1
.I j
i Wolf Creek Generating Station License Examination -
F During a LOCA, which one of the following conditions is an indication that adequate core cooling will depend upon reak flow?
[
- a. RC system equilibrium pressure is above S/G pressure.
b, RC system equilibrium pressure is below S/O pressure,
- c. Injection flow is greater than break flow.
f i: d. Injection flow is less than break flow,
. 032 I
'Ihe operators are conducting EMG C-11 (Loss of Emergency Coolant Recirc) and find that RWST level cannot be j' maintained, Which one of the following describes the preferred attemate source ofinjection to the RCS under these i_ conditions?-
- a. One RilR pump taking a suction on the Spent Fuel Pool.
b, One Si pump aligned to the RHR pump discharge which are, in turn, taking a suction from the Spent
[
- i. Fuct Pool.
I a
j 033 Electrical Power has been restored follcwing a loss of offsite power. If an RCP is to be started, which one of the following is the first RCP that should be started?
q a. A
! b. B
- c. C Ld. D.-
Wolf Creek Generating Station License Examination
. 034 --
t A reactor trip has occurred due to a loss ofinstmment Air, Under these conditions, the ability to control the feeding and L .eaming of the S/Gs can be maintained for which one of the following times?
- a. I 150ur: ,
b, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ,
- c. _ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> -
. d.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> .
- 035-While performing a reactor start-up, criticality is achieved with the control bank below its insertion limit. Which one of the following actions must be taken as a result of this? .
- a. Fully insert the control and shutdown banks and recalculate the ECP, b.- Stop all further outward rod motion, begin a normal boration and recalculate the ECP.
- c. Stop all further outward rod motion and contact Reactor Engineering for support.
- d. - Insert all control banks and Emergency borate until adequate shutdown margin is demonstrated
-036 A Low RC Pump Standpipe level alarm is an indication of which one of the following problems?
ni Low seal injection flow,
- b. Failure of #1 seal,
- c. Failure of #2 seal,
- d. Failure of #3 seal.
4
Wolf Creek Generating Station License Examination 037-
- A failure of Containment cooling causes equilibrium Contr.8ement temperature to increase from 80*F to 105'F.
.ssuming no change in Tm or Letdown flow rate, which one of the following describes the effect that this will have on
- a. - - it will throttle open slightly during the course of the temperaue change and then retum to its original position.
l b. It will throttle closed slightly during the course of the temperature change and then retum to its original position.
- c. It will throttle open slightly during the course of the temperature change and remain in that positian.
- d. It will throttle closed slightly during the course of the temperature change and remain in that position.
- - 038 i.
He plant is at 75% power when a new Letdown Demineralizer is placed in senice. Which one of the following will occur if the boric acid saturation of this demineralizer is incomplete?
I a. Coatrol rods will step inward.
- b. Control rods will step outward.
Primary coolant pH will decrease.
c.
- d. Letdown flowrate will decrease.
l 039 Which one of the following safeguards features will be disabled by a loss of electrical power to its bistable?
- a. Containment Spray initiation.
. b. Safety lejection initiation.
l c. Main Steam Line Isolation.
Feedwater Isolation.
. d.
y-- , _ -
o 4 - Wolf Creek Generating Station License Examination 040 '
Blocking the Low Steamline Pressure Signal below P 11 has which one of the fol'owing effects?
- a. Si and Steamline Isolation due to steam line pressure are inhibited,
- b. Si and Steamline isolation initiations change from low steamline pressure to high rate of steamline pressure decrease.
- c. Si initiation changes from low steamline pressure to high rate of steamline pressure decrease, Steamline Isolation initiation is unafTected, 2 ~
- d. Si initiation due to steamline pressure is inhibited. Steamline Isolation initiation changes from low steamline pressure to high rate of steamline pressure decrease.
l l 041
%c plant had been operating at 100% power when the following events occurred.
j 4 A Cire Water problem forced the reduction of reactor power to 18% at the maximum allowable rate.
- l. -o Excess boration resulted in a significant decrease in Tave'.
Which one of the following describes the effect that this will have on the Nuclear Instrumentation High Power trip points? -
- a. Power Range will be less conservative. Intemiediate Range will not be active.
b, Power Range and Intermediate Range will be less conservative,
- c. Power Range and Intennediate Range will be more conservative,
- d. Power Range will be more conservative. Intermediate Range will not be active.
042 ,
During a plant startup the n: actor is critical at 5 X 10' cps. I&C is working on Source Range Channel N-31. Which one of the following actions can be taken on the N-31 drawer without causing a reactor trip?
- a. Remove the control power fuses,
- b. Remove the instrument power fuses.
- c. Going to trip bypass and then removing the control power fuses,
- d. Going to trip bypass and then removing the instrument power fuses.
.. - - _ - - - ~ - - - - - - . . _ _ . - , . .. . -- . - . .- .. .- .
3,m,--n
- Wolf Creek Generating Station License Examination 043.
The.RCS Subcooling Margin is 10*F. Under which one of the following sets of conditions requires entry into FR-C.1 3tesponse to inadequate Core Cooling)?
- a. No RCPs are running, Natural Circ RVLIS reads 40%, the 5 highest reading Core Exit TCs are between 750'F and 800'F.
- b. No RCPs are running, Natual Cire RVLIS reads 30%, the 5 highest reading Core Exit TCs are between 650*F and 700'F.
- c. One RCP is running, Forced Flow RVLIS reads 45%, the 5 highest reading Core Exit TCs are between 850'F and 900'F.
- d. One RCP is running, Forced Flow RVLIS reads 95%, the 5 highest reading Core Exit TCs are between 1100*F and 1150'F.
044 Which one of the following describes the operation of the CRDM Cooling Fans during Loss of Offsite Power conditions?
- a. All fans trip and cannot be restarted until normal pow er is restored.
- b. All fans trip but two can be restarted.
- c. Two fans trip and two fans remain running,
- d. All fans remain running.
045 The plant is operating at 50% power when the 'A' S/G level rises to 60%. All other steam generator levels remain at 50%
level. Which one of the following is the most likely cause of this?
- a. Either steam flow transmitter FT 512 or steam pressure transmitter PT-516 has failed low, b, Either steam flow transmitter FT-513 or steam pressure transmitter PT-514 has failed high.
- c. Feedwater pressure transmitter PT-508 has failed low,
- d. . Either feed flow transmitter FT-510 or FT-511 has failed high
Wolf Creek Generating Station License Examination 046'
- A plant fire has forced the evacuation of the control room and caused a loss of all Train 'A' Vital AC. The plant is being -
antrolled from the ASP. Which one of the following describes how these conditions will affect the water sources to the Auxiliary Feedwater system?
- b. ESW is no longer available as an attemate source of AlW.
- c. LSP will no longer automatically swap ARV suction from the CST to ESW but it can be done manually.
- d. We functionality of the AFW water sources will not be affected by these conditions.
i 047-Which one of the following conditions will cause the automatic, at of both motor driven AFW pumps, BUT will NOT start the turbine driven AFW pump?
- a. AMSAC initiation.
- b. Lo-Lo level on any two S/Gs.
- c. Site Blackout.
- d. Safety injection.
048 Which one of the following describes the relationship between the RCP seals and the RC Drain Tank?
i a. He leakoffs from all three seals normally drain to the RCDT.
t
- b. The Icakoffs from the #2 and #3 seals drain to the RCDT.
- c. He leakoff from the #2 seal and, during a Phase A isolation only, the #1 seal leakoff, drain to the RCDT.
- d. _ Ec leakofT from the #2 seal only drains to the RCDT.
T 5
i Wolf Creek Generating Station License Exarninatlca "0491 - i ne Waste Gas Decay system is monitored by the following instraments:
1 LGH RE 10A -_ Radwaste Building Emuent Particulate and lodine
-i - GH RE 10B) Radwaste Buil; ling Emuent Wide Range Gas
-+' GH RE 23 -Waste Gas Ventilation Exhaust '
A HiHi Alarm (s) on which one of t!e following will result in the autcmatic isolation of the Waste Gas Decay tek discharge fmm the Radwaste Building Ventilation Exhaust?
- a. Either Gil RE 10A or GH RE 10B but not GH RE 23. [ ,
b.~ Both GH RE 10A and GH RE 10B simultaneously but not GH RE 23.
' c, Either GH RE 10A, GH RE 10B, or GH RE 23,
- d. GH RE 23 only.
1050
- With refueling in progress, which one of the following alarms on the Main Control Board will require entry into OFN-KE-018 (Fuel Handling Accident)? ; ,
- a. HI alarm on Spent Fuel Pool Area Radiation
- b. HI alann on Containment Building Area Radiation.
- c. HI H1 alarm on Containment Purge,
- d. Hi HI alarm on Plant Vent 051~
Which one of the following CANNOT be caused by a single RTD failure?
a.' T,v. recorder indication decreases by 5'F.
- b. T,v. recorder indication increeses by 5'F.
c .- Loop 'N T,y, increases by 5'F.
- d. Loop 'D' T,y, decreases by 5'F.
+> t i
Wolf Creek Generating Station License Examination
%c following plant conditions exist: -
j V % i.( 618.2'Fm
- 9- 1T. is 554.8'F.
-+J Power Range Channels N41 aid N44 read 100% ,
t
-+L Power Range Channels N42 and N43 read 101%
! . Undei these conditions, a High ReaMor Power trip, ifit occurs, will occur when plant themial power is at which one of the following values? -
u
< a. 33850 MWt. ;
- b. 13886 MWt.
. c. 3921 MWt.
- d. Here is no predictable correlation between thennat and Ni power.
-053-The following sequence of events occurs::
<-+ :- A Safety injection Signal has started all ECCS pumps.
-+; . De diesels are running normally, L -+ - 5 minutes after the SIS, offsite power is lost.
l'hich one of the following describes the expected response of the ECCS pumps to the loss of offsite power?
. a. - J All ECCS pumps restart when their train's D/G breaker closes.
b.: _ he LOCA Sequencer will restart the ECCS pumps.
- c. he Shutdown Sequencer will restart the ECCS pumps.
- d. He SIS will have to be reset and the pumps started by operator action.-
054 he plant is operating at full power with pressure control in 455/458 mode. Which one of the following describes the
. expected result of Pressurizer Pressure Instrument BB PT-458 failing fully downscale?? '
a.- No immediate efrect but PORV PCV-456A will not open on a high pressure condition.
b.' %e pressuria.er spray valve will close and RCS pressure will increase to the trip setpoint.
L c. Pressurizer heaters will come on and RCS pressure will be controlled by PORV PCV-455A cycling.
- d. Pmssurizer heaters will come on and RCS pressure will incmase to the reactor trip setpoint.
- Wolf Creek Generating Station License Examination
-055.
i: A maintenarce worker inadvertently shears the air line off Charging Flow Control valve FCV-121.- Which one of the Alowing wii! occur if no operator action is taken?
- a. - Reactor trip on high pressurizer level.
- b. Reactor trip on low pressure due to low pressurizer level.
- c. RCP seal damage due to inadequate seal injection flow.
- 1. - RCP seal flow isolation due to high seal flow.
056 he following conditions exist:
-+. Reactor poweris 5%.
+ RCS pressure is 1900 psig.
Which one of the following events will cause a reactor tdp? .
- a. He reactor should already be tdpped.
- b. One Turbine Impulse Pressure Channel fails high.
i
- c. Loss ofpower to one RC pump bus.
- d. One Power Range Channel fails high.
i l
1 Wolf Creek Generating Station License Examination j 057-During the course of a startup Bank D is at 8 steps, He bottom lights of several rods in Bank D are then seen to come i
- n. Which one of the following DRPI failures could explain this observation?
a.. Data A ..
- b. Data B
- c. Either Data A or Data B.
- d. Neither, this is nonnal for this rod height.
i -058
- A LOCA has occurred and Containment Spray from the RWST is in progress, ne Recirc Sump Isolation valves (EN HV 1 & 7) have received a close signal as a result of the Phase A Isolation. Assuming all ESFAS signals an still active,
. which one of the following must be done to allow the manual opening of the Sump Recirc valves from the MCB?
- a. Nothing, the valves may be opened manually at anytime,
- b. He RWST must have reached the LoLo Level alarm point.
- c. Reset the Phase A Isolation (CIS A).
- d. - Reset the Containment Spray Signal (CSAS) 059
- Which one of the following radiation monitor alarms will result in the automatic isolation of both the Shutdown Purge and Mini Purge isolation valves?
- a. High Containment Purge gaseous radiation only.
- b. Either High Containment Atmosphere gaseous radiation or High Containment Purge gaseous radiation.
- c. Either High Containment Purge gaseous or particulate radiation.
- d. Either High Containment Purge gaseous or particulate radiation or High Containment Atmosphere gaseous or particulate radiation.
%it Creek Generating Station License Examination 060 Spent Fuel Pool level is dec easing due to a non-isolable leak. Un-borated water may be used to maintain level under thich one of the following conditions?
- a. With SRO approval, there are no other restrictions.
- b. Only if radiation levels will exceed 10 CFR 100 limits without it,
- c. Only if boiling will occur without it,
- d. It may never be used.
061 He following conditions exist:
-+ Re unit is in Mode 5 with the RCS filled.
l
-+ Heat removal is being provided by Train 'A' of the RHR sfc .em.
-+ RHR Train 'B' is operable but not in operation.
-+ S/G WR levels are: 'A' 5%, 'B' 8%, 'C' 22%, 'D' 0%.
It is desired to remove Train 'B' of the RHR system from operable status for a system upgrade. He S/Gs will act as the required second heat removal method while the work is in progress. Which one of the following describes the j acceptability of removing RHR Train 'B' from senice under these conditions?
l l a. It may be done as long as the RCS remains filled,
- b. It may be done as long as RCS temperature remains below 200*F.
- c. It may not be done because tir., S/Gs are not an adequate heat sink under these conditions,
- d. It may not be done because two RHR trains are required for Mode 5.
062 Which one of the following describes the operation of the MSIVs?
- a. An MSLIS can only fast close the MSIVs, the operators can only slow close them,
- b. An MSLIS can either fast or slow close the MSIVs, the operators can only fast close them.
- c. An MSLIS can only fast close the MSIVs, the operators can either fast or slow close them,
- d. Both an MSLIS and the operators can either fast or slow close the MSIVs.
1
W Wolf Creek Generating Station License Examination i- 063-A Loss of Offsite Power has deenergized the Startup Transfonner and resulted in a reactor trip from full power. Which
- ne of the following describes the expected response of the Unit Auxiliary Transformer breaker (PA0101) and Startup fransfonner breaker (PA0110)? (Only one train is considered to improve clarity.)
i
- . a. - PA0101 will trip when when the reactor trips. PA0110 will close when PA0101 trips.
4
- b. - PA0101 will trip when the Generator Breaker opens, PA0110 will close when PA0101 trips.
I c. PA0101 will trip when the Generator Breaker opens. PA0110 will not close until power is restored to the Startup Transformer.
- d. PA0101 will remain closed until power is restored to the Startup Transfonner. PA0110 will close wheu l PA0101 opens.
i 064
! D/G NE01 is running unloaded when Vital DC bus NK01 is lost. Which one of the following describes the effect that
- this event will have on the D/G7 l
- a. - D/G capability will be unaffected.
- b. It will continue to run but cannot be shutdown except by using the fuel racks.
! c. Speed control will be lost and the D/G will trip on mechanical overspeed, i d. He engine will trip immediately.
L
{- 065 Which one of the following describes how the D/G Fuel Oil Transfer pump operates to maintain Day Tank level?
f.
- a. - With the D/G running, the pump will start on low level and then run continuously. With the D/G in
- Standby, the pump cycles to maintain level between its high and low setpoints.
- b.- ne pump always cycles to maintain level between the high and low setpoints.
i
- c. With the D/G running, the Transfer Pump is stopped and level is maintained by the shaft driven pump.
With the D/G in Stan.'by, the Transfer Pump cycles to maintain Day Tank level between its high and low -
setpoints.
! d. The pump always runs continuously to keep the Day Tank full. Any overflow is retumed back to the i Fuel Oil Storage Tank.
i l-i
Wolf Creek Generating Station License Examination 066 Which one of the following will occur in the event of a high radiation alarm on the Unit Vent Monitor (GT RE 21B)?
- a. Ecre are no automatic actions associated with this instrument.
- b. He Unit Vent will be isolated.
- c. The sample path of the monitor is isolated,
- d. Lo sample path of the monitor swaps to the room in which it is located.
067 he following conditions exist:
- Lake temperature is 65 F 4 hree Cire Water pumps are running.
-e Condenser pressure is 3 inches Hg.
4 he plant is at 100% power.
- One Cire Water pump then trips.
-o Condenser pressure subsequently degrades to 5 inches.
Which one of the following describes how turbine load will respond to these events?
- a. Here will be no response unless vacuum degrades further.
- b. A turbine runback will occur until vacuum is restored to its original value,
- c. A turbine runback to minimum load will occur because the initiating condition (pump trip) will not clear,
- d. A Turbine Load Limit of approximately 80% will be imposed and turbine load will set back to that value.
4 068 Following a LOCA the following conditions exist:
-o RCS pressure is 180 psig.
- RHR Train 'A'is Shutdown Cooling Mode.
-+ RHR Train 'B' is injecting from the RWST.
Which one of the following will occur when the RWST reaches its LoLo Level setpoint?
- a. Both RHR Trains swap to Cold Leg Recire,
- b. RHR Train 'B' swaps to Cold Leg Recire. No change in Train 'A'.
- d. RHR Train 'A' will trip on low suction pressure.
[ _
Wolf Creek Generating Station License Examination 1 069
' ~
E 'Ihc following conditions existi l 4L Both trains of RHR cooling are in' service.- l
-+ ( . RCS te.npemture is 180*F.
. In krder to slowly heat up to 290'F, the operator secures CCW to one of the RHR heat exchangers. Which one of the following is a likely consequence of these events?
4 1 a. Overheating RHR Pump seal,
- b. Runout of a CCW pump.
- c. Waterhammer in the CCW System.
. d. Waterhammerin the RHR System.
070-
_ A large LOCA has resulted in a CISA and a CISB. Which one of the following must be done to open the flowpath for the Hydrogen Analyzer?
?
- a. . Reset the CISA and manually open the sample isolation valves.
- b. Reset the CISB ar' manually open the sample isolation valves.
- c. Placing the Hydrogen Analyzer in service will cause the valves to automatically open.
l d. ~ 'Ihe sr.mple isolation valves may be manually opened at the MCB. No reset is necessary.
071 When using the RHR system to drain water from the Refueling Pool, prolonged operation at low KHR pump flow must be avoided. One solution to this is the use of BN V-8717 (RHR Pump Retum to the RWST). Operation of this valve is
. under very close administrative control.1What is the principal danger of using this flow path?
a.- Pumping the RCS to the RWST.
bi hiaccurate RHR flow indication.
'd. RHR pump damage due to vortexing.
Wolf Creek Generating Station License Examination -
. 073 A plant shutdown is in progress and the steam dumps havejust been switched to the Steam Pressure Mode. Which one f the following describes steam dump system response if the steam header pressure instrument then fails high?
- a. All steam dumps will close.
- b. He condenser dump valves will open and remain open until closed by operator action,
- c. All condenser and atmospheric dump valves will open until T m reaches 550*F where they will close.
- d. The condenser dump valves will open until T,v, reaches 550 F where they will close.
l 073 Which one of the following describes the effect that a loss ofInstrument Air pressure will have on hotwell level ccatrol?
- a. Hotwell level will rise because the makeup valves will fail open.
- b. Hotwell level will decrease because the reject valves will fail open.
- c. Hotwell level control is uncertain because both the makeup and reject valves fail open,
- d. Hotwell level control is uncertain because both the makeup and reject valves fail closed.
74 During an emergency it becomes necessary to close a valve in a Very High Radiation Area to stop a leak. Which one of the following is the minimum RWP requirement needed in order to make the entry?
- a. The entry can be made under an existing General RWP.
- b. A Specific RWP is required to be written prior to entry.
3 % a
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- Wolf Creek Generatlag Station License Examination - _
07$
, Which one of the following is the HASIS for the following NOTE in C 0, less of All AC Power?
CSF status trees shall be monitored for information only Function Restoration Procedures shall not be implemented i during this procedure.
i 1
- a. Functica Restoration Procedures should not be entered until EOPs are exited.
4 :
- b. Function Restomtion Procedures assume that at least one AC emergency bus is energired. ;
! c. Implementing C 0 and Function Restoration Procedures conewontly could result in a Technical- ,
i Specification violation.
! . d. C-0 has priority over all other procedures.
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Wolf Creek Generating Station License Examination SRO question 76 101
'ollowing a LOCA the following conditions are observed:
- Average of 5 highest Core Exit T/Cs 618'F and stable
-c flot leg Temp 550'F
-c RCS pressure 1100 psig a Cold leg temp 300'F Which one of the following describer, the present status core cooling and inventory?
- a. The core is covered and being cooled by natural circulation.
- b. The core is un-covered and being cooled by natural circulation.
- c. The core is covered and being cooled by reflux boiling,
- d. 1he core is un covered and being cooled by reflux boiling.
SRO question 77 102 The Release Iso!ation valves for both the Liquid Radwaste (llD RV 18) and Secondary Liquid Radwaste (liF RV-45) systems have received automatic close signals. Which one of the following could be the cause of this?
- a. Trip of all Cire Water pumps,
- b. Trip of all Service Water pumps.
- c. lli radiation on Liquid Radwaste Monitor (lid RE-18).
- d. 111 radiation on Secondary Liquid Radwaste Monitor (11D RE-45).
SRO question 78 103 1he following conditions exist:
-+ A large steam break has caused a reactor trip.
+ One Reactor Trip Dreaker failed to open.
+ All control rods are fully inserted.
-o A significant cooldown and depressurization of the RCS occurs.
l Which one of the following describes how recovery from this event will be affected by the conditions described?
l
- a. Both trains of Safety injection will fail to automatically initiate,
- b. One train of Safety injection will fail to automatically initiate.
- c. Both trains of Safety injection cannot be reset.
- d. One train of Safety injection cannot be reset.
l
f Creek Generating Station License Examination SRO question 79-
-104 lie following conditions exist:
4 - 1he Shutdown Banks are tully withdrawn.
-o The Control Banks are fully insented.
l --+ Source Range 'A'is stable at I X 10' cps.
i e A reactor trip then occurs int.crting the Shutdown Hanks.
-*- One hour after the trip it is noted that Source Range 'A' reads 8 X 10' cps.
Which one of the following conclusions can be drawn about the status of Source Range 'A' from these events?
- a. it is functioning as expected,
- b. Its discriminator is misadjusted and is passing too many pulses,
- c. Its discriminator is misadjusted and is passing too few pulses.
d.- The change in rod shadowing caused by insertion of the Shutdown Banks is affecting its accuracy.
I
- SRO question 80 105 The following plant conditions exist:
4 A plant shutdown is in progress.
4 1he compensating voltage power supply in the NI 35 Intennediate Range Ni drawer has failed to zero.
This will have which one of the following effects on the plant as the shutdown continues?
- a. 1he plant will trip when power goes below 10%.
b.- Source Range NI 31 will energize at a higher than nonnal power level,
- c. Source Range NI 31 will remain deenergized until its Block / Reset switch is taken to RESET,
- d. Both Source Range channels will remain dcenergized until their Block / Reset switches are taken to RESET.
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Wolf Creek Generating Station License Examination SRO question 81 106
'MG ES-04 (Natural Circulation Cooldown), contains the following CAUTION:
If RCP seal cooling had previously been lost, the affected RCP(s) shall not be started prior to a status evaluation.
Which one of the following is a condition where this CAUTION can be ignored and the RCP started without a status evaluation?
- a. He PZR is empty and a bubble has fonned in the vessel head,
- b. RCS subcooling is s10*F.
- c. A CSF Red or Orange path on Core Cooling has been diagnosed.
- d. Any CSF Red path has been diagnosed.
SRO question 82 107 Under vehich one of the following conditions are Adverse Containment setpoints in effect?
- a. Containment pressure is now 2 psig. It had reached a peak of 35 psig and decayed to 5 psig over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> it took an additional 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to reach its present value,
- b. Containment pressure is now 4 psig and has reached this value slowly building up over the past 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 4
- c. Containment dose rate is presently 5 X 10 R/hr. It has been at this level for the past week. ,
- d. Containment dose rate is presently 3 X 10' R/hr. It rapidly spiked to a peak of 5 X 10' R/hr and then decayed to this level in 5 minutes, it has been at tids level for the past 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
SRO question 83 108
%e plant is in Mode 3. He following is a summary of RCS leakrates.
- Primary to secondary S/G 'A' .63 GPM
-o Primary to secondary S/G 'B' .30 GPM
- Primary to secondary S/G 'C' 0 GPM 4 Primary to secondary S/G 'D' 0 GPM a Weld crack in T6 Loop 'A' RTD thennowell .25 GPM Which one of the following describes the action required by Tech Specs as a result of these conditions?
- a. No action, leak rates are within limits for this mode. Change to Mode 2 is not permitted.
- b. Reduce leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SilUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
- c. Reduce leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD S11UTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- d. De in C01.D SilUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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Wolf Creek Generating Station License Examination .
SRO question 84 109 tuning refueling operations, a fuel assembly is dropped in the reactor vessel. No damage is apparent. Which one of tha Ibilowing describes the appropriate use of 0FN KE 018 (Fuel liandling Accident)?
- a. It is entered at the discretion of the Refucilng SRO. ;
- b. It is entered at the discretion of the Control Room SRO. .
- c. It must be entered and is conducted by the Refueling SRO.
- d. It must be entered and is conducted by the Control Room SRO.
SRO question 85 110 De following conditions exist:
-o The plant is in Mode 3. l 9 Letdown valves LCV-459 and LCV-460 are open. l
-+ Letdown valves IIV 8160 and liv 81$2 are closed. -
Which one of the followin is the most likely cause of these conditions? ,
- n. Phase A isolation,
- b. Phase B isolation,
- c. Pressurizer levelis 5%.
- d. Letdown temperature 150'F SRO question 86 111 He plant is at full power and the compliment oflicensed personnel in the Control Room is one less than the minimum required by Tech Specs. Which one of the following actions must be taken if this condition lasts for more than two hours?
- a. No action other than continue to try to obtain the necessary person.
- b. Initiate a one hour report to the NRC via the FTS Phone, c, Be in 110T S'I ANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- d. Summon a Nuclear Station Operator to the Control Room to assist with operations not involving
. reactivity changes.
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Wolf Creek Generating Station License Examination 112 Yhich one of the following describes a violation oflicensed power limits?
- a. liighest Power Range Channelis 102% Instantaneous thermal power (computer point SCUIi1B)is -
3560 MWt.
- b. Instantaneous thennal power (SCU1118) is 3640 MWt. Moving 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average (RJU159MA) is 3560 i MWt.
! c. - Instantaneous thermal power (SCUlll8) is 3600. Moving hour average (RJUl58MA) is 3565 MWt? *
- d. Moving hout average thermal power (RJUl58MA) is 3575 MWt. Moving 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average (RJul59MA) '
is 3560 MWt.
- SRO question 88 ,
113 L <
During an outage with the plant in Shutdown Cooling the following plant conditions exist:
- Reactor Vessel temperature 210*F 4 Reactor Vessel pressure 250 psig 4 Steam Generator temperature 65'F ;
4 Steam Generator pressure atmospheric '
USAR Chapter 16 require which one of the following actions?
f
- a. Increase S/G pressure to 200 psig within 30 minutes.
- b. Increase S/G pressure to 200 psig within I hour.
- c. _ Decrease Reactor Vessel pressure to 200 psig within 30 minutes,
! d. Decrease Reactor Vessel pressure to 200 psig within I hour.
- SRO question 89 i 114 l - ,- -
De plant is operating at 50% power when the operators discover that Control Bank D cannot be moved. All rods in the bank are presently at 80 steps. I AC reports that the problem is a fault in the rod control system and not the rod drives
, themselves, Which one of the the following describes the restrictions on continued operation which will be in force for the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />? :
- a. None, any restrictions come into force after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4
- b. power is limited to 75% -
. c. ~ Power is limited to 65%
! d. De in 110T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Wolf Creek Generating Station License Examination SRO question 90 115
'ntry into Containment in Mode 1. 2, or 3 is controlled. Two teams must enter Containment in Mode 1 for leak mspection. De requirements for IIP coverage during the Containment entry includes:
- a. Each team must have an llP technician escort,
- b. One llP technician will enter with the teams and provide monitoring as required.
- c. Each team is responsible for radiological monitoring with IIP controlling Containment access.
- d. Each individual is responsible for radiological monitoring with IIP controlling Containment access.
SRO question 91 116 he 'A' lleater Drain pump has a failed shan scal. An approved replacement seal is not in stock, but there is a seal in stock that will do the job. Which one of the following is the proper course of action?
- a. The job may not be donc unless an approved replacement shaf'. seal is obtained,
- b. Log and assign a Temporary Modification to use the shan seal in stock until an approved shan seal can be obtained.
- c. Issue an Emergency Temporary Modification, that is approved by the SS, to use the shan seal in stock until an approved shan seal can be obtained.
. d. The in stock shan seal may be used without any required Temporary Modification paperwork.
SRO question 92 117 Which one of the following events would require the plant be in 110T STANDBY within one (1) hour and report to the NRC Operations Center within one (1) hour?
a, Failing to restore an INOPERABLE CCP to OPERABLE status in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in mode one, b; llaving both ESW trains freeze up with the potential of an accompanying loss of service water due to fn:czing while in mode one.
- c. Reactor Coolant Pressure exceeds 2800 psig while in mode one.
- d. Determining that the Power Range Neutron Flux liigh Setpoint exceeded 115% rated thennal power.
i
l Wolf Creek Generating Station License Examination SRO question 93 118 ISAR Chapter 16 requires the existence of a Reactor Vessel llead Vent Path. Which one of the following is the basis for this requirement?
- a. A vent during reactor vessel filling,
- b. A vent for gases that could inhibit natural circulation core cooling.
- c. Prevent a buildup of explosive gases in the reactor vessel head.
1
- d. Provide gas samples for chemical analysis.
SRO question 94 119 ..
The plant is in Mode 6 when the following occur; e Fuel handling operations are in progress.
-o There is a spent fuel assembly on the cart in the Refueling Pool.
- S/O 'A' has a secondary side manway open and S/O maintenance is in progress, a The Aux Duilding Watch reports that the blank flange is missing from a safety valve on S/O 'A'.
Which one of the following actions is required as a result of these conditions?
- a. Immediately temiinate all movement ofirradiated fuel or core alterations..
- b. Immediately terminate any core alterations and transport the spent fuel assembly presently on the cart to the Fuel Building.
- c. Replace the missing blank flange within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- d. Immediately tenninate all work in containment until the flange is replaced.
SRO question 95 120 You are the Sh!ft Supervisor and one of the Station Operators on your shift will soon exceed his administrative limit on Whole Body Dose, You deta mine that an extension of 1000 mR will be suflicient for the rest of the year. Whose approval is required for such an extension?
- a. Your approval as Shift Supervisor is sufficient and must be entered into your log.
- b. Health Physics Supenisor Support (HPSS).
- c. The PSRC and the Plant Manager,
- d. Extensions are not permitted in non-cmergency situations, l
Wolf Creek Generating Station License Examination SRO question 96 121 ;
inder which one of the following circumstances are you required to sign the RWP Sign in Sheet when making an entry mto the RCA?
- a. Whenever an RCA entry is made,
- b. The first time you enter using that RWP. 1
- c. The first entry after any change to the RWP.
- d. Only if the Automated Radiological Access Control System is out of senice, i
SRO question 97 i 122 i You are the Shift Supmisor and havejust approved a Release Permit. Which one of the following groups is responsible j for perfonning the actual release?
) a. Operations
- b. Chemistry
]
- c. licalth Physics ;
i d. Operations if a liquid release, Chemistry if gaseous.
a 1 SRO question 98
! 123
, ' Which one of the following describes a principal advantage and a principal- disadvantage of the Steam Dump method of Post SGTR Cooldown compared to the other methods pmvided by the EMGs?
- a. Fastest but has biggest reactivity risks.
j
[ b. Fastest but has the biggest radiological risks.
- c. Limits spread of contamination but is slowest.
- d. Limits spread of contamination but has biggest reactivity risks.
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Wolf Creek Generating Station License Examination SRO question 99 124
.t which one of the following Emergency Action levels is a Protective Actic a Recommendation first mandatory?
- a. Notification of Unusual Event
- b. Alert
- c. Site Area Emergency
- d. General Emergency l
l SRO question 100 125 noting tim enndnce nrvli,9Mn. viw nperators encounter a Continuous Action Step. Wluch one of the following is the mle that defines how long the Continuous Action Step is valid?
- a. Applies only to the procedures main flowpath once it is encountered but not to include the attachments, tables, or figures.
- b. Applies to the rest of the EMO procedures used during this entty into the EMOs.
- c. Applies to the rest of the procedure once it is encountered and new procedures unless superseded by steps in the new procedure.
- d. Applies only to the rest of the procedure it is encountered in, j
RO ANSWER KEY
- 1) A 26 ) C 51 ) A 76 ) C
- 2) D 27 ) A 52 ) A 77 ) A
- 3) C 28 ) A 53 ) B 78 ) A
- 4) B 29 ) A M)A 79 ) D 5 D 30 ) D 55 ) A 80 ) D AMd") )C = # 31 ) B 56 ) B 81 ) C
- 7) B 32 ) D 57 ) B 82 ) D wA h
- 8) D 33)D 58 ) A 83 ) A 9 ) C *s4 @ 34 ) D 59 ) B 84 ) D 10 ) A 35) D 60 ) C 85)B 11 ) D 36 ) D 61 ) C 86 ) C 12 ) B -A M 37 ) B 62 ) C 87 ) A 13 ) C 38 ) A 63)B 88) C wDb 14 ) B 39 ) A 64 ) B 89 ) C 15 ) B 40 ) D 65 ) A 90 ) C 16 ) D _ _41 ) A 66 ) D 91 ) D 17 ) B 42) D 67 ) D 92 ) B 18 ) B 43) A 68 ) C -vB d45 93 ) A 19 ) A 44 ) B 69 ) C 94 ) B *Dqfsp 20 ) C 45) B 70 ) D 95) C 21 ) B 46 ) C 71 ) A 9 6 ) 83 22 ) D 47 ) D 72 ) D 97 ) B 23 ) D 48 ) e1R 9C!b 73 ) D 98 ) C 24 ) B 49) A 74 ) D - 99 ) B
- 25) D 50 ) C 75)B 100 ) D d e w - 4 m .c Wd $f Ck t'*S [ya,nt'w
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4 CHIEF EXAMINER RO WRITIEN. EXAM COMMENTS WOLF CREEK S/25/97 Question No. Comment 2 Answer is not supported by OFN BB-005 In that loss of CCW flow la not listed as a RCP immediate shutdown criterion. Also, the catalog numbe-should be 015AA2.10. Resolution: Deleted the word immediate from the stem and corrected catalog number.
3 Reference does not support the answer. Resolution: Used attached references together with the applicable drawing to determine thbt answer is Correct.
5 The correct catalog number should be 000027AK2.03. Resolution:
Corrected catalog number.
7 Catalog number should be 051 AA2.02. Reference does not support answer.
Resolution:- Corrected catalog raumber and reference.
10 Catalog number does not match outline. Resolution: Changed outline.
Meets standards, 14 Suggest rewording the stem to Indicate the reason for the time limit on running an RHR pump on circulation without CCW cooling is to prevent pump damage. This is the purpose given in the reference. Resolution:
Reworded stem.
22 This question only requires the applicant to know RCP trip criteria to chose the correct answer. This knowledge is covered in the dynamic scenarios.
Develop a more in-depth question. Resolution: Question replaced.
28 Reference does not support answer. Resolution: Provided correct reference.
43 Whereas the question is soliciting one set of conditions, most applicants would reject distractors B and C because they are subsets of A and D. Are the applicants expected to have the CSF status tree for core cooling memorized? If not, this does not appear to be a fair question. Resolution:
Resolution: Applicant is expected to know approximate temperature range.
Choices B and C are not subsets because of different flow values.
53 The answer should be C vice B. Resolution: Provided reference to prove B, 54 Reference does not support the answer. Resolutlom Reference does support answer.
61 Choice D does not appear to be a plausible distractor. Resolution: replaced distractor.
73 Reference does not support answer. Resolution: Provided an additional reference, s
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l 74 Why is Choice B not correct?_ Resolution: Qualified Choice B to make it incorrect, i
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! 90 Reference does not support answer. Resolution: Provided correct reference.
! 92 Reference does not support answer. Resolution: Provided correct reference. !
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- 95 Reference does not support answer. Resolution
- Provided correct reference. !
[ 96 Reference does not support answer. Resolution: Provided correct referenco.
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ES 401 Site Specinc Writwn l'orm ES 4017 Examination Cover Sheet U.S. Nuclear Regulatory Commision ,
Site-Specific l Written Examination Applicant Information Name: Region: IV Date: 8-25-97 Facility / Unit WCGS License Level (ljG/ SRO Reactor Type W Start Time: _
Finish Time:
Instructions Use the answer sheets provided to document your answers. Staple this
- cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination starts.
4 Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
- Applicant's Signature -
Results _
Examination Value Points Applicant's Score Points Applienat's Grade ,
Percent NUREG 1021 Intuim Rev 8. January 1997
EXAMINATION ANSWER SHEET Wolf Creek Generating Station Written Examination PRINT NAME: DATE: 08/25/97 61RSI M LASI
- 1. A B C D 26. A 11 C D $1. A B C D 76. A 11 C D
- 2. A 11 C D 27. A B C D $2. A 11 C D 77. A B C D
- 3. A B C D 28. A B C D $3. .4 Il C D 78. A 11 C D
- 4. A B C D 29. A B C D $4. A 11 C D 79. A B C D
- 5. A Il C D 30. A 11 C D $5. A Il C D 80. A B C D
- 6. A !! C D 31. A B C D 56. A !! C D 81. A B C D
- 7. A B C D 32. A B C D 57. A B C D 82. A Il C D
- 8. A B C D 33. A Il C D $8. A B C D 83. A B C D
- 9. A B C D 34. A B C D 59. A B C D 84. A Il C D
- 10. A B C D 35. A Il C D 60. A B C D 85. A B C D I1. A B C D 36. A B C D 61. A B C D 86, A 11 C D
- 12. A B C D 37. A B C D 62. A 11 C D 87. A B C D
- 13. A B C D 38. A B C D 63. A B C D 88. A B C D
- 14. A 11 C D 39. A B C D 64. A B C D 89. A B C D 1
- 15. A B C D 40. A B C D 65. A B C D 90. A B C D
- 16. A B C D 41. A B C D 66, A B C D 91. A B C D
- 17. A 11 C D 42. A B C D 67. A B C D 92. A B C D
- 18. A B C D 43. A b C D 68. A B C D 91. A B C D
- 19. /. B C D 44. A B C D 69. A B C D 94. A B C D
- 20. A B C D 45. A B C D 70. A B C D 95. A B C D 2 *. . A B C D 46. A- B C D- 71. A B C D 96. A B C D
- 22. A B C D 47. A B C D 72. A B C D 97. A B C D
- 23. A B C D 48. A B C D 73. A B C D 98. A D C D
- 24. A B C D 49. A B C D 74. A B C D 99. A B C D
- 25. A B C D 50. A Il C D 75. A B C D 100. A B C D i-
i 2
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Wolf Creek Generating Station License Examination
- 001
- A Rod Control Urgent Failure Annunciator is in alami. The operators find that it is still possible to move control rods y selecting the individual banks and moving the selected bank manually. Which one of the following can be concluded mm these symptoms?
- a. He problem is in a Power Cabinet.
j b. He problem is in a Logic Cabinet.
i i c. Acre is an additional failure in that no rod motion should be possible.
- d. No conclusions may be draw from the stated symptoms, i I
1 002 Which one of the following conditions requires the affected RCP to be tripped.? i
- a. Seal outlet temperature reaches 150'F
- b. RCP motor winding temperature reaches 225'F
- c. Sealinjection flow is lost.
- d. CCW flow is lost for over 5 minutes.
,03 he Nonnal Charging Pump is the caly charging pump available and Emergency Boration via the BIT becomes necessary. Which one of the following is necessary to accomplish this?
- a. His cannot be donc until at least one Centrifugal Charging pump can be placed in senice.
- b. Start at least one Safety injection pump.
- c. Open the Cen+rifugal Charging pump discharge flow control valve (FCW121)
- d. Close RCP Seal injection flow contro! valve (30 HC-182).
9
Wolf Creek Generating Station License Examination
- 004 A leak in the CCW system is larger than the capacity of makeup from the Dominerallaed Water system. Which one 6f te following is the preferred source of additional makeup?
- a. Senice Water, i
- b. ESW.
- c. Cite Water,
- d. Isolate the Service Loop from the Safety Train and make up to the Safety Train from the Senlee Loop.
005 l A Pressurizer Pressure instmment has failed low and the altemate instruments have been selected. Which one of the following remains inaccurate despite the fact that the attemate instruments have been selected?
- a. I'ZR Pressure Control.
- b. PZR Pressure Recorder,
- c. OPAT/0TATTemperature Recorder.
- d. Subcooling Margin Monitor.
.iO6 While perfonning immediate action steps in E-0, the following conditions are obsened:
e RCS T,s., 547'F SO A Press /WR Level = 1005 psig/20%
- PZR LVL + 21% SG C Press /WR level = 1005 pig /50%-
Which one of the followirr, is the most likely cause of this event?
a, 'A' Steam line break between the S/0 and tbc MSIV,
- b. 'A' Steam line break between the MSIV and Main Turbine,
- c. 'A' Feed line break between Feed Reg Valve and Main Feed Line Check Valve.
- d. 'A' Feed line break between Main Feed Line Check Valve and the S/G.
4 Wolf Creek Generating Station License Examination Which one of the following conditions requires an immediate manual turbine trip? - *
- a. Turbine Exhaust flood Temperature 210'F.
- b. Condenser Pressure is 9 inches Hg.
)
- c. Turbine Shar1 Pump discharge pressure 120 psig.
5 d. . Condenser Vacuum Pump trip.
~
' 008 he operators are perfonning EMO C 0 (less of all AC Power). If un SI signal occurs at this time, which one of the following describes the proper response to this Sl?
1 ;
l a. Do not reset it to ensure rapid injection of core coolhag water when power is restored. !
- b. Do not reset it to ensure sufHelent load exists to prevent a diesel generator overspeed trip.
- c. Reset it to allow the manual initiation of D/O cooling when AC becomes available. !
d, - Reset it to pennit manual loading of needed equipment on AC Emergency buses when they become available s09 Which one of the fc,llowing is the tasis for the direction in EMO C-0, (Loss of All AC Power), to depressurize the intact steam generators to 260 psig? '
i
- a. Prevent accumulator nitrogen from injecting into the RCS.
E Minimize CVCS letdown flow, b.
l d. Reduce the time the AFW System must be in operation.
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Wolf Creek Generating Station License Examination 010 P:wer has been lost to 120 VAC Instn' ment 13us NN01 and the following conditions exist: -
- As expected. Charging pemp suction has swapped to the RWST. !
+ All failed instruments have been selected out.
- *Ihe operators have estabilshed Excess Letdown per OlH NN 021.
Which one of the following describes why Nonnal Letdown is not used?
- a. Charging flow will have been minimized doen to RCP seal injection only,
- b. A locked in Letdown isolation signal exists.
- d. To prevent flashing in the Normal Letdown line.
011 If ESW is lost while offsite power is still available, the operators are directed to trip the RCPs. Which one of the following describes why this action is necessary?
- a. To minimize containment heating due to loss of RCP motor air coolers.
- c. To eliminate the possibility of RCP motor damage due to loss of motor cooling, e
- d. To rcduce the hem load on the CCW system to extend cooling of safety related components.
012 For the past 45 minutes, the Fire Brigade has been fighting a tire requiring 2000 gpm from the Fire Main. No operator actions have been taken other ;han those directly involved with fire fighting. Which one of the following specifies which Fire pumps can be expected to be running at this time?
- a. Jockey Pump and Electric Fire Pump,
- b. Jockey Pump and Diesel Fire Pump.
- c. Electric Fire Pamp and Diesel Fire Pump.
- d. Diesel Fire Pump nnly.
1 l
Wolf Creek Generating Station License Examination 013 Tech Specs limit containment pressure to 1.5 psig during operations. Which one of the fo!!owing is the reason for this astriction?
- a. To maintain accuracy ofinstruments with detectors inside containment,
- b. To assure any leakage from containment remains within 10CFR100 lirnits.
- c. To assure that peak pressure during an accident will remain witida design limits,
- d. To prevent long tenu degradation of containment pn:ssure beundary capability.
014 To prevent RilR pumps from damage the EMGs have a time limit on how long an RHR pump can be run on recirculation without CCW ' low to the RHR heat exchangers. Which one of the following is that time limit?
- a. 2.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />,
- b. 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />,
- c. 3,5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
- d. 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
015 Which one of the following alarms requires entry into 0FN BB-006 (High Reactor Coolant Activity)?
- a. Steam Generator Blowdown RAD High on RM ll,
- b. Letdown RAD High on RM ll.
- c. Proces RAD HI on MCB.
- d. Condenser Air Removal RAD High on RM 11.
1 Wolf Creek Generating Station License Examination 016
, A LOCA has occurred and the following conditions exist: -
o The plant is being cooled by natural circulation.
4 All MSIVs are closed.
- Nitrogen from an accumulator injection has collected in the U tubes of the 'C' steam generator inhibiting natural cire in the 'C' loop. i
, 9 Natural cite flow in all other loops is normal. l e All ARVs are set to 20% open.
! Which one of the following , 'll be the best indication of the failure of natural cire in the 'C' loop?
- a. 'C' loop AT will rapMly decrease to near uro.
- b. 'C' Loop T6 will ine ;ase relative to the unaffected loops,
- c. 'C' S/O pressure wLl increase rluive to the unaffected loops, l
- d. 'C' S/U will depressurize. l l
l 017
. A plant startup is in pmgress and the following conditions exist:
4 j * 'The reactor is caitical at 5 X 10 Amps on the Intennediate Range.
- T,y,is at its no load value.
) -+ A rod control failure causes the selected control bank of rods to step out 'O steps before being stopped by the i operators.
- No reactor trip occurs.
Which one of the following descrhes the effect that this event will have on stactor pwameters?
- a. Reactor pow, r c:.d T,v, will both rise, i b. Reactor power will rise. T,y. will be unaffected,
- c. Reactor power will be unaffected. T,v. will rise.
- d. Reactor power will increase. 'Ihe response of Tm will depend upon core life.
T.
i 018
- Ihe plant is operating at full power when a fully withdrawn control rod drops into its fully inserted position. Which one of the following describes the effect that a da. lay in recovering the rod will have on the potential for core damage as a resuh of the recovery of the rod?
- a. Any delay will not appreciably affect the potential for damage.
- b. . Changes in xenon concentration will make rod recovery more hazardous as time goes on.
- c. Changes in xenon oncentration will make rod recovery less hazardous as time goes on.
' d. Changes in fuel temperature and fuel bumup will make rod recove:y more hazardous as time goes on.
-,e ~_ _ - , .,, . .. ,,. _, , . - . . _ , _ ._- ...,.~~,,3r- , - , . . , .9 -
Wolf Creek Generating Station License Examination 019 A LOCA has caused containrpent pressure to increase to 27 psig. Which one of the following describes when and why I 11 RCPs must be tripped? i
- a. Immediately due to loss of cooling to the seals and moton,
- b. Immediately due to potential damage to any rurming RCPs caused by containment spray,
- c. Only if subcooling decreases to O'F and at least one CCP or Si pump is on.
030 l
He plant is operating at full power when a pressurizer safety valve fails open and can not be reclosed. As the RCS blows down from operating pressure to 1000 psia, which one of the following trends do you expect to see on safety valve tall pipe temperature?
- a. Temperature will decrease from 650'F to 545'F
- b. Temperature will decrease from 350'F to 220'F
- c. Temperature willincrease from 220'F to 350'F
- d. Temperature will increase from 350*F to 545'F 021-Which one of the following describes the Tech Spec IDENTIFIED LEAKAGE limit of 10 gpm?
- a. Does not include S/G leakage and is kept low enough so that it will not interfere with detection of UNIDENTIFIED LEAKAGE.
- b. Includes S/G leakage and is kept low enough so that it will not interfere with detection of UNIDENTlHED LEAKAGE.
- c. Does not include S/G Icakage and is kept low enough to be a small fraction of 10 CFR part 100 dose guidelines.
~ d. Includes S/G leakage and is kept low enough to be a small fraction of 10 CFR part 100 guidelines.
Wolf Creek Generating Station License Examination 022 A LOCA has occurred. While isolating the accumulators in EMO ES 11, one accrmulator isolation valve did not go.
closed. He operators are depressurizing the unisolated accumulator. During the subsequent recovery actions the
.perator is told not to depressurize RCS pressure below any unisolated accumulator pressure.
Which of the following recovery actions could result in a pressure decrease of the RCS during the recovery?
- a. starting a RCP on a loop with an intact steam generator,
- c. securing the RCPs.
- d. placing RilR in service with CCW lined up to tne heat exchangers.
033 Loss of seal injection flow will have which one of the following effects on CCW AT in the hermal Barrier Heat Exchanger?
- a. It will decrease since there is no longer any flow to cool, b.- It will remain almost constant since most of the heat load comes from RCP internals,
- c. It will remain almost constant since no injection flow flows past thl heat exchanger,
- d. It will increase since hot primary coolant will replace the cooler injection flowing past the thermal barrier.
034 he following plant conditions exist:
- ne plant has been shutdown for 10 days after setting the world record for the longest full power run by a PWR.
- No fuel has been replaced as yet.
9 he RCS is drained to mid loop.
- He RCS is vented to atmosphere.
-o llent removal is being provided by RHR.
Which one of the following is closest to the time it will take for the onset of boiling if RHR is lost.
- a. 12 minutes
- b. 14 minutes -
- c. 32 minutes
- d. 34 minutes
Wolf Creek Generating Station License Exaaination 035 A LOCA has resulted in ContJnment pressure increasing to 7 psig. The reactor has tripped and the crew has entered
.MG E-0. Which one of the following instruments should be used to niake the detennination that neutron flux is decreasing?
- a. Source Range
- b. Intermediate Range
- c. Power Range
- d. Gamma Metrics l
l 026 A bteam Generator Blowdown isolation has occurred and the operators note the following valve positions:
-o Blowdown isolation Valves (liv l-4) are closed.
- Upper Sample Valves (liv 19 22) are o;,cn.
- Lower Sampic Valves (liv 35 38) are open.
Which one of the following was the cause of the blowdown isolation?
- a. Auxiliary Feedwater Actuation Signal,
- b. Under voltage on ND 01,
- c. Condenser Air Removal high radiation.
d, Safety injection Signal.
027 EMG ES 31, Post SGTR Cooldown Using Backfill, has a CAUTION that informs the operator that if all RCPs are stopped, the first RCP that is restarted should be in a non ruptured loop. Which one of the following could result from starting a RCP in a loop with a ruptured SG7
- a. An inadvertent criticality.
- b. Enlargement of the rupture,
- d. Cold shock of the RCS.
P Wolf Creek Generating Station License Examination 028 Which one of the following conditions will prevent tennination of Safety injection? - -
- a. Pressurir.cr level 3%. !
- b. AFW flow 350,000 lbm/hr.
- c. RCS pressure stable at 1100 psig.
- d. RCS subcooling is 35'F.
029 A LOCA has occurred and the operators are conducting EMO ES 11 (Post LOCA Cooldown and Depressurization). No RCP: are running. Step 17 instmets them to 'Depressurin the RCS to Refill the Pressuriar'. Which one of the f;11owing describes the preferred method for accomplishing this task?
- a. Open a Pressurizer PORV to avoid stresses on the Pressuriar Spray neule.
- b. Open a Pressurizer PORY because this will depressurin the RCS most rapidly,
- c. Initiate Auxiliary Spray to preserve RCS inventory,
- d. Initiate Auxiliary Spray because this will raise Pressuriar level most rapidly.
J30 Entry into EMG C 11 (Loss of Emergency Coolant RecIT) is most likely to be required under which one of the f:llowing conditions?
- a. Steam Leak inside containment,
- b. ~ Steam Leak outside containment.
- c. LOCA inside contaimaer.t.
- d. LOCA outside contaimnent. '
L i
l l l
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~__ _ _ _ _ _ . . . _ . . _ _ _ _ _ _ _ _ . _. ._ .__ _ _ . _ _ _ _ . _ _._.
1 l
Wolf Creek Generating Station License Examination .
031 During a 1.0CA, which one of the following conditions is an indicM v ' bit adequate core cooling will depend upon- ,
reak flow?
- a. RC system equilibrium pressure is above S/O pressure.
l b. RC system equilibrium pressure is below S/O pressure.
! c. Injection flow is greater than break flow.
! d. Injection flow is less than break flow. ,
032
'Ihe operators are conducting EMO C 11 (loss of Emergency Coolant Recire) and find that RWST level cannot be maintained. Which one of the following describes the preferred alte.nate source ofinjection to the RCS under these conditions?
- c. One R11R pump taking a suction on the Spent Fuel Pool,
- b. One Si pump aligned to the RilR pump discharge which are,in tum, taking a suction from the Spent Fuel Pool.
- c. One Si pump aligned to the VCT.
l
! 033 Electrical Power has been restored following a loss of offsite power. If an RCP is to be started, which one of the following is the first RCP that should be started?
I
- a. A l
- b. D
- c. C l
- d. D l
l 4
i
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l Wolf Creek Generating Station License Examination l A reactor trip has occurred due to a loss ofinstmment Air, Under these conditions, the ability to cor. trol the fooding1md
.eaming of the S/Gs can be maintained fot which one of the following times?
- a. Ihour
- b. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> I
- c. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- d. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> !
03$ !
While perfomilng a reactor start up, criticality is achieved with the cor. trol bak bdow its insertion limit. Which one of the following actions must be taken as a result of this?
j- a. Fully insert the control and shutdown banks and recalculate the ECP, :
bi Stop all further outward rod motion, begin a nonnal boration and rwalculate the ECP,
- c. Stop all finther outward rod motion and contact Reactor Engineering for support,
- d. Insert all control banks and Emergency borate until adequate shutdown margin is demonstrated.
036 A Low RC Pump Standpipe level alarm is an indication of which one of the following problems?
- a. Low seal in,lecticn flow,
- b. Failum of #1 seal,
- c. Failure of #2 seal.
- d. Failure of #3 seal.
l l ^\
Wolf Creek Generating Station License Examination
-037-A failure of Containment cooling causes equilibrium Contaltiment temperature to increu n in 80*F to 105'F. -
.ssuming no change in T v. or Letdown flow rate, which one of the following describes the effect that this will have on Charging Flow Contrel valve B0 FCV 1217
. a. It will throttle open slightly during the course of the temperature change and then retum to its original position.
- b. It will throttle closed slightly during the course of the temperature change and then return to its original position.
- c. It will throttle open slightly during the course of the temperature change and remain in that position.
- d. It will throttle closed slightly during the course of the tempenture change and remain in that position.
038 ne plant is at 75% power when a new Letdo vn Demineralizer is placed in service. 'Vhich one of the following will occur if the boric acid saturation of this domine. 4izer is incomplete?
- n. Control rods will step inward
- b. Control rods will step outward,
- c. Primary coolant pH will decrease,
- d. Letdown flowrate will decrease.
-039 Which one of the (ollowing safeguaals features will be disabled by a loss of electrical power to its bistable?
- a. Containment Spray initiation.
- b. . Safety injection initiation, c, Main Steam Line Isolation.
- d. Feedwater Isolation.
+
Wolf Creek Generating Station License Examination 040 7%Aing the low Steamline Pressure Signal below P-11 has which one of the following effects?
- l a. Si and Steamline Isolation due to steam line pressure are inhibited.
, b. Si and Steamline Isolation initiMions change from low steamline pressure to high rate of steamline
- pressure decrease.-
l c. Si initiation changes from low steamline pressure to high rate of steamline pressure decrease. Steamline
! solation initiation is unaffected.
- d. Si initiation due to steamline pressure is inhibited. Steamline isoladon initiation changes from low steamline pressure to high rate of steamline pressure decrease.
041
- The plant had been operating at 100% power when the following events occurred.
--+ A Cire Water problem forced the reduction of reactor power to 18% at the maximum allowable rate.
-+ Excess boration resulted in a significant decrease in Tave',
Which one of the following describes the effect that this will have on the Nuclear lastnanentation High Power trip points? -
' a. Power Range will be less conservative, Intermediate Range will not be active,
- b. Power Range and Intermediate Ranye will be less conservative,
- c. Power Range and intermediate Range will be most conservative,
- d. Power Range will be ri.ac marvative. Intermediate Range will not be active.
042 Dui;ng a plant ovtup the reactor is critical at 5 X 10' cps. l&C is working on Source Range Channel N-31. Which one of the following actions can be 6aken on the N-31 drawer without causing a reactor trip?
- a. Remove the control power ft :cs
- b. - Remove the instrument power fuses.
- c. Going to trip bypass and then removing the control power fuses.
- d. Going to trip bypass and then removing the instrument power fuses.
c
4 4
- Wolf Creek Generating Station Li 9ense Examinatio's i . 043
. He RCS Subcooling Margin is 10'F. Under which one of the following sets of conditrc w try ires etry into FR-C:1 ~
- l. Response to inadequate Core Cooling)?
4- a. No RCPs are running, Natural Cire RVLIS reads 40%, the 5 highest read'.ag Core Exit TCs a.= between
- 750'F and 800'F.'
1 b. No RCPs are running, Natural Cire RVLIS reads 30%, the 5 highest readmg Core Exit TCs are between 650'F and 700*F.
- c. One RCP is running, Forced Flow RVLIS reads 45%, the 5 highest reading Core Exit TCs are between j 850*F and 900'F.
- d. One RCP is running, Forced Flow RVLIS reads 95%, the 5 highest reading Core Exit TCs are between Il00*F and 1150 F, g
4 044
. - Which one of the following describes the operation of the CRDM Cooling Fans during Loss of Offsite Power conditions?
- a. All fans trip and cannot be restarted until normal power is restored.
- b. A!! Ons trip but two can be restarted,
- c. ino fans trip and two fans remain running.
dc All fans remain running.
~ 045
- De plant is operating at 50% power when the 'A' S/G level rises to 60% All other steam generator levels remain at 50% .
j level. Which one of the following is thw..ost likely cause of this?
- a. . Either steam flow transmitter FT-512 or steam pressure transmitter PT-516 has failed low.
- b. Either steam flow transmitter FT-513 or steam pressure transmitter PT-514 has failed high.
, c.- Feedwater pressure '.ransmitter PT-508 has failed low.
, ' d. Either feed flow transmitter FT-510 or FT-511 has failed high -
6 s
m c- - - - . - ----. .-- - - - - - - _ - - - - - - - - - - _ _ - - ,
Wolf Creek Generating Station License Exarr: ration 046 A plant fire has forced the evacuation of the contml room and caused a loss of all Train 'A' Vital AC. he plant is being ontrolled from the ASP. Which one of the following describes how these conditions will affect the water sources to the Auxiliary Feedwater system?
- c. LSP will no longer automatically swap AFW suction from the CST to ESW but it can be done manually, i ne functionality of the AFW water sources will not be affected by these conditions.
047 Which one of the following conditions will cause the automatic start of both motor driven AFW pumps, BUT will NOT l
start the turbine driven AFW pump?
- a. AMSAC initiation.
- b. Lo-Lo level on any two S/Gs.
- c. Site Blackout,
- d. Safety injection.
048 Which one of the following describes the relationship between the RCP seals and the RC Drain Tank?
- a. The leakotTs from all three seals normally drain to the RCDT.
- b. He leakofTs from the #2 and #3 seals drain to the RCDT.
- c. he leakoff from the #2 seal and, during a Phase A isolation only, the #1 seal leakoff, drain to the RCDT.
- d. He leakoff from the #2 seal only drains to the RCDT.
U
1 Wolf Creek Generating Station License Examination -
049
'Ihe Waste Gas Decay system is monitored by the following instruments: - -
+ GH RE 10A - Radwaste Building Emuent Particulate and lodine
-+. GH RE 10B - Radwaste Building Emuent Wide Range Gas
-+ GH RE 23 -Waste Gas Ventilation Exhaust i
A HiHi Alarm (s) on which one of the following will result in the automatic isol.: tion of the Waste Gas Decay tank discharge from the Radwaste Building Ventilation Exhaust?
- a. - Either GH RE 10A or GH RE 10B but not GH RE 23.
- b. ~ Both GH RE 10A and GH RE 10B simultaneously but not GH RE 23.
- c. Either GH RE 10A, GH RE 10B, or GH RE 23.
- d. GH RE 23 only.
050 i
With refueling in progress, which one of the following alarms on the Main Control Board wil! iequire entry into OFN-KE-018 (Fuel Handling Accident)?
- a. Hi alarm on Spent Fuel Pool Area Radiation
- b. HI alann on Containment Building Area Radiation.
- c. HI HI alarm on Containment Purge.
- d. HI HI alarm on Plant Vent 051 Which one of the following CANNOT be caused by a single RTD failure?
a, T,y, recorder indication decreases by 5'F.
- b. - T,y. recorder indication increases by 5'F.
- c. Loop 'A' T , increases by 5*F.
- d. Loop 'D' T,v. decreases by 5'F.
4
Wolf Creek Generating Station License Examination
- 052 he following plant conditions exist
- -
+. r This 618.2*F -
- oJ T. is 554.8'F
-+ Power Range Channels N41 and N44 read 100%
- -+ Power Range Channels N42 and N43 read 10l%
i
- Under these~ conditions, a High Reactor Power trip, ifit occurs, will occur when plant thermal power is at which one of -
the following values? '
- a. 3850 MWt.~
li. 3886 M Wt.
- c. - 3921 MWt.
- d. Here is no predictable correlation between thennal and NI power.
053 he following sequence of events occurs:
-+- A Safety injection Signal has started all ECCS pumps.
--+ ne diesels are running normally.
-+ 5 minutes after the SIS, offsite power is lost.
' diich one of the following describes the expected response of the ECCS pumps to the loss of offsite power?
a.- All ECCS pumps restart when their train's D/G breaker closes.
- c. He Shutdown Sequencer will restart the ECCS pumps.
- d. He SIS will have to be reset and the pumps started by operator action.
054:
%c plant is operating at full power with pressure control in 455/458 mode. Which one of the following describes the L expected result of Pressmizer Pressure lustrument BB PT-458 failing fully downscale?? -
- a. No immediate effect but PORV PCV-456A will not open on a high pressure condition.
- b. He pressurizer spray valve will close and RCS pressure will increase to the trip setpoint,
- d.
Pressurizer heaters will come on and RCS pressure will increase to the reactor trip setpoint, l
I
. . Wolf Creek Generating Station License Examination l ; 055-
~
- A maintenance worker inadvertently shears the air line off Charging Flow Control valve FCV 121 Which one of the
- >llowing will occur ifre operator action is taken?
. a. Reactor trip on high pressurizer level.-
4.
, b. Reactor trip on low pressure due to low pressurizer level. :
i .. .
- ; c. RCP seal damage due to inadequate seal injection flow.
j : d. RCP seal flow isolation due to high seal flow, i
s-
] ~0 56 p he following conditions exist:
i - -+ Reactor poweris 5%.
-+- RCS pressure is 1900 psig.
'f Which one of the following events will cause a reactor trip?
a.- He reactor should already be tripped, i.
- b. One Turbine Impulse Pressure Channel fails high.
. c. . Loss of power to one RC pump bus.
- d. One Power Range Chanael fails high.
4 5
a 4
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-e+ e -,rm-- , , , . --- - - - n
Wolf Creek Generating Station License Examination
~
0571
- During the course of a startup Bank D is at 8 steps. He bottom lights of several rods in Bank D are then seen to come nE Which one of the following DRPI failures could explain this observation?
- a. Data A -
- b. . Data B
- c. - Either Data A or Data B.
- d. Neither, this is nomnl for this rod height.
J 058 l- _ A LOCA has occurred and Containment Spray from the RWST is in progress. He Recim Sump Isolation valves (EN
- HV 1 & 7) have received a close signal as a result of the Phase A Isolation. Assuming all ESFAS signals are still active, which one of the following must be done to allow the manual opening of the Sump Recirc valves from the MCB?
- a. Nothing, the valves may be opened manually at anytime, b, he RWST raust have reached the LoLo level alann point.
- c. Reset the Phase A Isolation (CIS-A).
- d. Reset the Containment Spray Signal (CSAS)
J d
059 Which one of the following radiation monitor alarms will result in the automatic isolation of both the Shutdown Purge
. and Mini-Purge isolation valves?
j
- a. High Containment Purge gaseous radiation ocly.
- - b. - Either High Containment Atmosphere gaseous radiation or High Containment PurBe gaseous radiation.
- c. Either High Containment Purge gaseous or particulate radiation.
f.
- d. Either High Containment Purge gaseous or particulate radiation or High Containment Atmosphere gaseous or particulate radiation.
. l.
Wolf Creek Generating Station License Examination
- 060 Spent Fuel Pool level is decreasing due to a non-isolable leak. Un-borated water may be used to maintain level under -
ihich one of the following conditions?
a, With SRO approval, there are no other restrictions.
. b. Only if radiation levels will exceed 10 CFR 100 limits without it,
- c. Only if boiling will occur without it.
- d. It may never be used.
061 he following conditions exist:
-+ ne unit is in Mode 5 with the RCS filled.
-+ Heat removal is being provided by Train 'A' of the RHR system.
-+ RHR Train 'B' is operable but not in operation.
-+ = S/G WR levels are: 'A' 5%, 'B' 8%, 'C' 22%, 'D' 0%.
It is desired to remove Train 'B' of the RHR system from operable status for a system upgrade, ne S/Gs will act as the required second heat removal method while the work is in progress. Which one of the following describes the acceptability of removing RHR Train 'B' from service under these conditions?
- a. It may be done as long as the RCS remains filled.
- b. It may be done as long as RCS temperature remains below 200'F.
- c. It may not be done because the S/Gs are not an adequate heat sink under these conditions.
d .- - It may not be done because two RHR trains are required for Mode 5.
062 Leh one of the following describes the operation of the MSIVs?
- a. An MSLIS can only fast close the MSIVs, the oper:: ors can only slow close them.
- b. An MSLIS can either fast or slow close the MSIVs, the operators can only fast close them.
- c. - An MSLIS can only fast close the MSIVs, the operators can either fast or slow close them.
- d. Both an MSLIS and the operators can either fast or slow close the MSIVs.
h 4 n f~
m_ m-Wolf Creek Generating Station Ligense Examination 063 A Loss of Offsite Power has deenergized the Startup Transformer and resulted in a reactor trip from full power. Which ne of the following describes the expected response of the Unit Auxiliary Transformer breaker (PA0101) and Startup Transformer breaker (PA0110)? (Only one train is considered to improve clarity.)
- a. PA0101 will trip when when the reactor trips. PA0110 will close when PA0101 trips.
- b. PA0101 will trip when the Generator Breaker opens. PA0110 will close when PA0101 trips,
- c. PA0101 will trip when the Generator Breaker opens. PA0110 will not close until power is restored to the Startup Transfonner.
- d. PA0101 will remain closed until power is re::tored to the Startup Transformer. PA0110 will close when PA0101 opens.
l 064 D/G NE01 is running unloaded when Vital DC bus NK01 is lost. Which one of the followmg describes the effect that this event will have on the D/G7
- a. D/G capability will be unaffected.
- b. It will continue to mn but cannot be shutdown except by using the fuel racks, c, Speed control will be lost and the D/G will trip on mechanical overspeed.
- d. The engine will tip immediately.
065 Which one of the following describes how the D/G Fuel Oil Transfer pump operates to maintain Day Tank level?
- a. With the D/G running, the pump will start on low level and then run continuously. With the D/G in Standby, the pump cycles to maintain level between its high and low setpoints,
- b. The pump always cycles to maintain level between the high and low setpoints.
- c. With the D/G running, the Transfer Pump is stopped and level is maintained by the shaft driven pump.
With the D/G in Standby, the Transfer Pump cycles to maintain Day Tank level between its high and low setpoints,
- d. The pump always runs continuously to keep ie Day Tank full. Any overflow is retumed back to the Fuel Oil Storage Tank.
i l
I
- ' Wolf Creek Generating Station License Examination 066
.Which one of the following will occur in the event of a high radiation alann on the Unit Vent Monitor (GT RE 21B)?
- n. Here are no automatic actions associated with this instrument.
- b. He Unit Vent will be isolated.
{ 'c. He sample path of the monitor is isolated,
- d. He sample path of the monitor swaps to the room in which it is located.
067
. He following conditions exist:
-+ Lake temperature is 65'F
- -D Hree Cire Water pumps are running.
-+ Condenser pressure is 3 inches Hg.
- The plant is at 100% power.
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-+ One Cire Water pump then trips.
-> Condenser pressure subsequently degrades to 5 inches.
{ Which one of the following describes how turbine load will respond to these events?
! a. Here will be no response unless vacuum degrades further.
- b. A turbine runback will occur until vacuum is restored to its original value,
- c. A turbine runbact *n minimum load will occur because the initiating condition (pump trip) will not clear,
- d. A Turbine Load Limit of al. proximately 80% will be imposed and turbine load will set back to that value.
068 Following a LOCA the followin3 conditiom exist:
-+ RCS pressure is 180 psig.
-+ RHR Tmin 'A' is Shutdown Coolint. Mode.
-+ RHR Train 'B'is injecting from the RWST.
Which one of the following will occur whe i the RWST reacher, he LoLo Level setpoint?
- a. Both RHR Trains swap to Cold Leg Recire.
- b. RHR Train 'B' swaps to Cold Ug Recire. No change in Train 'A'.
- d. RHR Train 'A' will trip on low suction pressure.
Wolf Creek Generating Station License Examination 069
- The following conditions exist: - -
+ Doth trains of RHR cooling are in senice.
4 RCS temperature is 180'F.
In order to slowly heat up to 290*F, the operator secures CCW to one of the RHR heat exchangers. Which one of the f:llowing is a likely consequence of these events?
- a. Overheating RHR Pump seal,
- b. Runout of a CCW pump,
- c. Waterhammerin the CCW System.
- d. Waterhammerin the RHR System.
070 A large LOCA has resulted in a CIS A and a CISB. Which one of the following must be done to open the flowpath for the Hydrogen Analyzer?
- a. Reset the CISA and manually open the sample isolation valves,
- b. Reset the CISD and manually open the sample isolation valves.
- c. Placing the Hydrogen Analyzer in service will cause the valves to automatically open.
- d. The sample isolation valves may be manually opened at the MCB. No reset is necessary.
071 When using the RHR system to drain water from the Refueling Pool, prolonged operation at low RHR pump flow must be avoided. One solution to this is the use of BN V-8717 (RHR Pump Retum to the RWST). Operation of this valve is
-under very close administrative contml. What is the principal danger of using this flow path?
- b. Inaccurate RHR flow indication.
c, Loss of RCS temperature control by bypassing the RHR Heat Exchanger,
- d. RHR pump damage due to vortexing.
s.
Wolf Creek Generating Station License Examination 072 A plant shutdown is in progress and the steam dumps have just been switched to the Steam Pressure Mode Which one i
f the folk wing describes steam dump system response if the steam header pre:sure instrument then fails high?
- a. All steam dumps will close.
- b. The condenser dump valves will open and remain open until closed by operator action.
- c. All condenser and atmospheric dump valves will open until T,,, reaches 550'F where they will close,
- d. The condenser dump valves will open until T,y, reaches 550*F where they will ..ose, 073 Which one of the following describes the effect that a loss ofInstrument Air pressure will have on hotwell level control?
- a. Hotwell level will rise because the makeup valves will fail open.
- b. Hotwell level will decrease because the reject valves will fail open.
- c. Hotwell level control is uncertain because both the makeup and reject valves fail open.
- d. Hotwell level control is uncertain because both the makeup and reject valves fail closed.
74 During an emergency it becomes necessary to close a valve in a Very High Radiation Area to stop a leak. Which one of the following is the minimum RWP requirement needed in order to make the entry?
- a. The entry can be made under an existing General RWP.
- b. A Specific RWP is required to be written prior to entry.
Wolf Creek Generating Station License Examination 075
, Which one of the following is the BASIS for the following NOTE in C-0, Loss ofAll AC Power? -
CSF status trees shall be monitored for information only. Function Restoration Procedures shall not be implemented during this procedure,
- a. Func on Restoration Procedures should not be entered until EOPs are exited.
- b. r' unction Restoration Procedures assume that at leut one AC emergency bus is energized.
- c. Implementing C-0 and Function Restoration Procedures concurrently could result in a Technical Specification violation.
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- d. C-0 has priority over all other procedures, i
'1 k
d
- c r.
Wolf Creek Generating Station License Examination
- 076 Ec plant is in Mode 3 when Vital DC bus NK01 is lost. Which one of the following describes the effect that this will ave on the ability _to control RCS temperature.? -
- a. ' De steam dumps will open and cool the plant down completely due to loss of the function of P-12.
t i b. De steam dumps will open and maintain temperature at the P-12 setpoint.-
l.
- c. De plant'will heat up until the atmospheric dump valves open.
- d. - He plant will heat up until the first S/G safety valve opens.
077 he plant is in refueling mode when the refueling crew manually initiates a Fuel Building Isolation (FBIS). He operator ,
in the control room is in the process of verifying proper system response. Which one of the following would require the operator's intervention?
- a. Control Room Pressurization Fans - 0FF
- b. Control Room' Air Conditioners - ON
- c. Fuel Building Emergency Exhaust Damper - OPEN
- d. Fuel Building Supply Fan - OFF 078 Reactor power is stable at 80% power when the operators find that control rods cannot be withdrawn in either manual or automatic mode. %e problem is traced to the inadvertent actuation of one of the Control Interlocks. Which one of the .
following is most likely to be the scurce of the problem?
- a. C-2 Power Range NIS High Flux.
- b. C-3 Overtemperature AT.
' c. C-4 Overpower AT. .
- d. C-5 Low Power Interlock.
Wolf Creek Generating Station License Examination -
079-With the plant at 25% power, which one of the following will result in a reactor trip on RCP UV7 i
- a. Undervoltage on bus PA01.
^
b, Undervoltage on bus PAC 2; 4
- c. Securing RCPs A & B.
. d. Securing RCPs A & D.
- -080 He following conditions exist
e A Feedwater Isolation has occurred.
-o ne originating signal has been cleared.
! -+ Re reactor trip breakers have been reclosed.
- -+- No other operator actions have been taken.
! -o - ne operators find that they do not yet have control of the Feed system components involved in the isolation.
L ne Feedwater Isolation was caused by which one of the following conditions? .
- a. Safety Injection
- b. 11i111 S/G level (P-14)
- c. LoLo S/G level
- d. Reactor Trip / Low Tm 081.-
Which one of the following describes how a high containment temperature will affect the Containment Cooling system?
- a. Fan Coolers will shift to low speed to prevent overload due to contsponding high pressure.
b, Fan Coolers will shift to high speed to maximize cooling.
- c. A fusible link will relieve the backpressure from the Fan Cooler discharge flow path 'to maximize flow rate through the Fan Coolers.
- d. Essential Service Water valves controlling flow rate to the Fan Coolers go fully open to maximize cooling.
_ _ _ .-- . . _ . _ _ . . _ . - _ . - _ . . _ _ . _ . ~ . . . - . _ . . _ _ . _ . . _ _ _ . _ -
Wolf Creek Generating Station License Examination 082-Loss of a single operating Condensate pump at full power will have which one of the following consequences? ._
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- a. The plant can continue to operate but power should be reduced to 91%
- b. One MFW pump will trip. Fast action on the part of the operators may prevent a plant trip.
j- c; . Both MFW pumps will trip, resulting in a turbine trip and reatr trip.
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- d. Both MFW pumps will speed up but will not be able maintain adequate feed flow and the plant will trip on low S/G level.
i
, .083-
. A plant heatup is in progress and RCS temperature is 530 F. All control rods are fully inserted and the reactor trip breakers are being tested. -Which one of the following Enginected Safety Features is most likely to be inadvertently initiated under these conditions?
- a. Feedwater Isolation.
- b. Main Steam isolation.
- c. AMSAC,
- d. ' Phase A Containment Isolation.
084 An AREA RAD lil alann on the Main Control Boards will result from which one of the following conditions on the Area Rad Monitors themselves? ;
- a. Either a High Alarm or a detector voltage failure,
- b. Either an Alert or a High Alarm.
- c. A Iligh Alarm only,
- d. An Alert only.
A i<
t --_:
Wolf Creek Generating Station License Examination
. 1085 4
i ne plent is at nonnal full power operation with all systems in normal alignment. Steam Header Pressure instrument
[ T 507 fails downscale. Which ONE of the following is the expected plant response?
! . a. Feedwater pump speed increases.
i- -
l b.- Feedwater pump spee? Gecreases.
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- c. Steam dump valves open.
1 j d. Steam dump valves can no longer open automatically, h
i 086 f He Service Air / Instrument Air Isolation valve (KA PV-ll)lias closed due to low air pressure. Which one of the -
i- - following is required _to reopen this valve?
i
- a. When Instrument Air pressure is restored, it will open automatica" j b. When Instrument Air pressure is restored, it can be opened manually.
l- c. When Service Air pressure is restored, it will open automatically.
j d. When Service Air pressure is restored, it can be opened manually, i
I
.087 I he diesel is running unloaded in response to a LOCA signal. He Fire Detection system then detects a fire in the diesel l room. - Which one of the following describes the etTect that this will have on the diesel and us auxiliaries? '
L
- a. There will be no change in the status of+he diesel and its suppcrt system.
. b. %c diesel will continue to run but it cannot be loaded.
i: c. The diesel will trip.
- d. He diesel fuel oil transfer pump will trip.
Wolf Creek Generating Station License Examination
-088?
- Re plant is operating at full power when the following events are noted:__
t - Generated Megawatts deen:ases.
A Reactor Power remains essentially constant.
-o _ Steam flow remains essentially constant.
Which one of the following conditions could have caused these symptoms?
- a. One turbine stop valve has failed closed.
- b. Loss of steam flow to the feedw..ter heaters.
- c. An MSR relief valve has failed open.
- d. Degraded condenser vacuum.
-089 The plant is in Mode 3 when an inadvertent LOCA Sequencer Initiation occurs. No actual SI rignal exists. Which one of
~ the following describes the effect that this event will have on the Essential Senice Water system? -
- b. He ESW pumps will start but ESW will remain cross connected with Senice Water.
- c. ESW will isolate from Service Water but the ESW pumps will not statt.
- d.
ESW will isolate from Senice Water and the ESW pumps will start.
090-When perfomting a procedure, for which ONE of the following reasons can a step or section be marked N/A7
- a. Faulted procedure step logic,
- b. Incorrect guidance,
- c. Not essential to procedure's intent.
df Incorrect step desenption, j
. . . . - .- . . . . . - . . . _ - . - - . - . - . - - - . - . . . - . . - . . - . - - - ~ . - . -
i Wolf Creek Generating Station License Examination 1 091 4 :-
- Materials having NFPA designators have colored labels such that the color designates the type of hazard. Which ONE ofJ :
l- ic following colors are used for materials that are health hazard?
a.' ' Greeni ,
b.' Wilow'
- c. Red i i di Blue-
- - 092- '
The required shin complement of 11 people does not include which ONE of the following?
- a. ShiR Clerk.
f
- b. . Shift Engineer.
I ;c. Health Physics Technician, i .
- d -Chemistry Technician.
[ 93 During a loss of all-AC power which one of the following methods of communicating from the Contml Roc.n to thc Auxiliary Building Watch is lost?
i a. Gaitronics t
j.-. b.-- Plant Phones
- c. - Plant radio syst:m.
- . d. Operator Pters (Beeper) 1 L
,. J
.- -- - _. . .- . ~ . ..~ - - - .- - -- .. . -.-
d Wolf Creek Generating Station License Examination 094 Which one of the following describes the method of Independent Verification to be used on the motor operated valves of tagout?
a, Verifying the position of the valve's breaker is sufficient.
- b. Verify by remote position indication prior to removing power to the valve.
c, lerify by remote position indication ifit is still available after power has been removed from the valve,
- d. Verify the status of the blocking device after the valve has been deenergized.
095 While conducting a surveillance test of the 'A' Motor Driven Auxiliary Feed Pump you find that the pump will not start requiring the termination of the test. Which one of the following describes what must be done with the test package?
- a. It may be discarded after the failure is noted in the Control Room Logbyk.
i b. It should be retained and, if necessary, passed on to your watch relief, until the test can be completed.
i l c. Submit the completed package to the nonnal review process with the test deficiencit,s appropriately noted, 1
- d. Tum the package in to the Surveillance Coordinator with an explanation of the deficiency attached.
096 While in the Auxiliary Building you note that the information on a DNO Tag is not legible. Once the appropriate
- _ investigation has been completed, which one of the following actions must be taken?
I a- Rewrite the information on the tag to make it legible.
4
- b. Prepare a replacement tag and note the action on the Clearance Order,
- c. Write a new Clearance Order to cover that tag.
- d. Re-issue the original clearance order and reverify all tags.
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. Wolf Creek Generating Station License Examination 097
'the plant is in Mode 6. Which one of the following describes the minimum compliment oflicensed personnel required i be present in the control room? :
n
- a. One licensed RO anywhere in the control room.
- b. One licensed RO at the controls.
- c. One licensed SRO caywhere in the control room,
, The maximum dose rate in a room is 1500 mrem /hr one foot from a certain valve. According to the WCGS Radir. ion Protection Manual, this room should be classified as which one of the following?
- a. Radiation ha.
- b. Ifgh Radiation Area.
- c. Locked liigh Radiation Area.
- d. Very liigh Radiation Area.
i 099
- Unless specific authorization is granted, a Station Operator who has been a WCGS employee for the past five years is limited by the site Radiation Protection Manual to which one of the following annual TEDEs?
- a. 500 mrem.
- b. 2000 mrem.
- c. 3000 mrem,
- d. s0 mrem.
i
Wolf Creek Generating Station License Examination 100 A LOCA has resulted in both a Phase A and Phase B Containment isolation. Containment pressure is currently 30 psig.
'.cset of both the Phase A and Phase B Containment Isolations at this time will result in which one of the following?
l
- a. All Phase A and Phase B valves remain closed and locked out,
- b. Phase A valves retum to their pre isolation condition, but Phase B valves remain closed and have their locked-in signal removed.
- c. Phase B valves retum to their pre isolation condition, but Phase A valves remain closed and have their locked in signal removed.
- d. All Phase A and Phasc B valves remain closed and have their locked-in sigr.al removed.
)