IR 05000454/1996301
| ML20135D111 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/03/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20135D057 | List: |
| References | |
| 50-454-96-301OL, 50-455-96-301OL, NUDOCS 9612090249 | |
| Download: ML20135D111 (127) | |
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U. S. NUCLEAR REGULATORY COMMISSION .
REGION lli l
Docket Nos: 50-454; 50-455 l Licenses No: NPF-37; NPF-66 Reports No: 50-454/96301(OL); 50-455/96301(OL)
Licensee: Commonwealth Edison Company (Comed)
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I Facility: Byron Generating Station, Units 1 & 2 )
Location: 4448 North German Church Road Byron, IL 61010-9750 Dates: September 30 - October 4,1996 &
October 14,1996 Examiners: R. M. Bailey, Examiner Rill M. E. Bielby, Examiner Rlli C. C. Osterholtz, Examiner Rlli H. Peterson, Examiner Rlli B. Hughes, Examiner NRR Approved by: Melvyn Leach, Chief Operator Licensing Branch !
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4 1 9612090249 961203 ,
PDR ADOCK 05000454
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EXECUTIVE SUMMARY Byron Generating Station, Units 1 & 2 NRC Examination Reports No. 50-454/96301; 50-455/96301 A licensee developed and NRC approved initial operator licensing examination at the senior l reactor operator (SRO) level was administered to ten applicants. This examination process '
incorporated a one week on-site period for administration of the operating examination and ;
the supporting NRC review and approval effort. In addition, the written examination was l reviewed by and received NRC approval for administration by the license l i
During this examination, a loss of examination material control was identified by the '
licensee and was in violation of NRC requirements as stated in 10 CFR 55.49, " Integrity of Examinations and Tests." However, as described in the enclosed examination report l (section 05.1), the violation is not being cited because the criteria specified in Section '
Vll.B.1 of the NRC Enforcement Policy was satisfie Results: Eight applicants passed all portions of their respective examinations and were issued senior reactor operator licenses. Two applicants failed one or more portions of their ,
respective examinations and were denied an operating licens l l
Summarv: The following is a summary of performance:
e A loss of examination material control required the generation of replacement examination material and resulted in a substantial delay in examination administration. (Section 05.1)
e Licensee validation of examination materiallacked comprehensive review as evident by errors detected during examination administration. (Section 05.2)
e A generic knowledge deficiency regarding independent verification requirements was recorded while performing a clearance tagout review. (Section 05.3)
e The applicants' effective use of communications during dynamic scenarios enhanced good teamwork. (Section 05.5)
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Reoort Details i i
1. Ooerations 05 Operator Training and Qualification j 0 General Comments j The licensee volunteered to participate in a pilot process in which the NRC initial ;
operator license examination consisting of operating and written test sections were '
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prepared by the licensee, then reviewed and approved by the NRC. The operating i test was administered by the NRC and consisted of two areas: i e A walkthrough section covering two categories which focused on administrative topics and job performance tasks related to control room and inplant system e A performance-based section focused on dynamic, integrated plant !
conditions evaluated during dynamic simulator scenario sets.
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The written test was administered by the licensee and consisted of 100 multiple- i choice questions focusing on a broad spectrum of plant systems' design and
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operatio ;
J 05.2 Licensee Developed Examination Material and Security 1 Examination Scone (NUREG-1021 & GL-95-06)
Examination development guidance and security measures were prescribed by NUREG 1021, " Operator Licensing Examiner Standards," Revision 7, and
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superseded in part by Interim Pilot Examination Guidance provided in Generic Letter (GL) 95 06, " Changes in the Operator Licensing Program."
" Observations and Findinas l
l Security Measures (1)
During the initial development phase, the licensee became aware on Monday, July 29,1996, that the examination development room's secure door had been left unlocked for some 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> between July 26,1996 and July 29,1996. The loss of examination material control was reported to the
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NRC on Tuesday, July 30,1996. Also, the NRC was informed of a similar but minor occurrence on Thursday, July 25,1996. The examination development room had been left unattended for approximately 2 minutes with the secure door locked but aja i
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The examination development room contained stand-alone computers which were password protected and a lockable file cabinet used for storage of examination material. This second level of security was deemed appropriate by the licensee until the NRC examiners questioned the security risk of routinely storing the lockable file cabinet's key within the examination development room. This key had been located in the examination development room during both occurrence The licensee took prompt action to submit a Problem Identification Form (PIF) addressing both occurrences and initiate short-term corrective actions (such as: replacement of the secure door key-lock with an improved security mechanism, and permanent removal of the lockable file cabinet's i key from the room). Additional long term corrective actions were proposed l to strengthen existing security measures. Based upon NRC concerns, the j licensee proposed replacing all of the potentially compromised examination material with newly developed item Following a review of this occurrence and the licensee's corrective actio the NRC concluded that the examination material compromise could have been prevented or minimized based upon corrective actions take ;
Additionally, the replacement of all compromised examination material was j deemed appropriate. Because of the added burden to develop replacement examination material, the NRC agreed to delay the operating test by three weeks and the written test by six week ,
l Failure to maintain control of examination material was a violation of 10 CFR I 55.49, Integrity of Examinations and Tests, which states in part that licensees shall not engage in any activity that compromises the integrity of 1 any test required by this part. This licensee-identified and corrected violation l I
is being treated a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev (50-454/96301-01(DRS)).
(2) Develonment of Examination Material The licensee provided an integrated examination outline for NRC review and approval some 60 days in advance as required. The outline content was !
acceptable and met the requirements specified in NUREG-1021 to begin l examination developmen The replacement examination material was delivered to the NRC some 30 days prior to examination administration as required. In general, the replacement material was consistent with the guidance in NUREG 1021 and GL-95-06. Some twenty percent of the written examination questions and JPM tasks / questions were revised or replaced to enhance the examination validity and/or level of difficult The following information is provided for evaluation by the licensee via their SAT based training program. No response is require !
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(a) Written Examination The written test was considered technically accurate with two exceptions as evident by the post examination comments. In order to clarify technically incorrect information, the licensee required a change to one question's stem while another question required a change to the wording in one of its distractors. These changes were considered appropriate but should have been identified during the validation revie The following are examples of selected test items which contributed to a reduced level of difficulty requiring NRC enhancement:
- Selected distractors were implausible and could be easily discarde e Selected distractors were not related to the intent of the question or could be easily eliminated (such as exact opposite responses " decrease versus increase").
e r; elected distractors referenced specific information in the stem of the question and reflected the most correct choic e Use of selected plant conditions or events in the stem of some questions reduced the evaluation focus from complex analysis to simple comprehensio The licensee incorporated the NRC review comments into a revised version of the examination which was approved for administratio (b) Walkthrouah The walkthrough test was considered technically accurate requiring a few minor changes within selected tasks and the replacement of one task to enhance the examination qualit (c) Dynamic Simulator Scenarios The integrated operating simulator test was considered technically accurate required only a few minor changes to enhance selected scenario event .3 Ooerational Examination Validation Examination Scoce (NUREG-1021)
Using NUREG-1021, Operator Licensing Examiner Standards, Revision 7, the examination team revised and validated the operational examination material
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consisting of walkthrough and dynamic scenario portions. This involved extensive use of the plant specific simulator and an inplant tou Observations and Findinos The replacement operational examination material was content validated in accordance with guidance contained in NUREG-1021. The NRC examiners verified that each portion of the operational examination was valid and ready for administration.
1 During examination administration, the NRC examiners identified a number of minor
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validation deficiencies. However, one of these was deemed significant enough to potentially impact the applicants' performance requiring prompt NRC intervention to ensure exam validity was maintained. The following are selected examples of improper attention-to-detail during validation:
i e The lack of a Master Out-Of-Service Card for the perfoimence of an i
administrative tagout review was questioned by the apolicants. Tiiis card
- should have been included with a normal package review bui was not identified during the validation process as being necessar e JPM NRC-9 (Locally Close RWST to RH Suction Valve) contained an evaluator cue to be given to the applicants which directed them to perform an operator action on valve "2RH8812." A discrepancy was identified by i the applicants during the examination administration. Valve 2RH8812 does
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not exist. The valve nomenclature should have been "2Sl8812."
e Dynamic Scenario 2-3, Event E (RCP Seal Failure following a loss of I
Instrument Bus 114), resulted in a loss of power to selected meters and chart recorders on the main control board. During the validation process, the licensee failed to identify that a loss of Bus 114 would result in a complete loss of meter / recorder indication for the failed RCP pump seal. The required main control board indication necessary to identify the event in progress was not available as expected. The NRC examiners were required to provide appropriate indication by communicating the necessary data through the simulator operato .4 Ooerational Examination Administration Examination Scooe (NUREG-1021)
Using NUREG-1021, Operator Licensing Examiner Standards, Revision 7, the examiners administered an operational examination consisting of a job performance (JPM) walkthrough section and a dynamic scenario set section to each applican This involved extensive use of the plant specific simulator to perform dynamic scenarios and a section of the JPM walkthrough. A separate section of the JPM walkthrough was simulated during an inplant tou .
. Observations and Findinas in general, the applicants demonstrated an above average level of skill while operation the licensee's plant specific simulator facility. The applicants' effective useci eway communication and crew mini-briefings during the dynamic sce' . contributed to good teamwor Onu administrative section required the applicants to perform a review of an out-of-service tagout prior to return to service. Six of the ten applicants failed to determine that one of the eight components had been improperly verified to be in its required position. Each component was required to be acknowledged as being in its correct position by a second, independent verifier. The NRC examiners informed the licensee of this knowledge deficiency and were informed by the licensee that prompt corrective actions were being formulate Simulator operator support during the dynamic scenarios and simulator JPMs was mixed. The following concerns regarding simulator operator support were noted:
e During one scenario, the simulator operator failed to place the simulator in RUN following a turnover until prompted by one of the board operators. This caused a short delay in scenario implementation but did not impact the applicants' performanc e During one scenario, the simulator operator caused a premature reactor trip during an attempt to remove a previously set malfunction per NRC's reques This resulted in a small delay to allow a reset of the simulator but did not impact the applicants' performanc e During JPM setup on the RM-11 console, the simulator operator reversed the HIGH and ALERT setpoints on one JPM. However, this error had no impact on the applicant's performance and was corrected prior to performance by another applican The licensee's operational and support staff provided quick, responsive assistance to the NRC examiners' needs ensuring minimal delays in the examination process during simulator operations and inplant tours. A licensed reactor operator's performance in the surrogate role was appropriate and met NRC expectation .5 Simulator Fidelity Examination Scone (NUREG-1021)
Using NUREG-1021, Operator Licensing Examiner Standards, Revision 7, the examiners observed the performance of the plant specific simulator during job performance (JPM) walkthrough and dynamic scenario sets. This fidelity review ]
was conducted in relation to simulation tasks performed on the plant specific i simulator during the examination proces l
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. Observations and Findinas The plant specific simulator was able to accurately mimic plant parameters during a variety of malfunctions and plant conditions. However, during the performance of selected dynamic scenarios and JPM tasks, the examiners observed three separate equipment fidelity occurrences which are described in Enclosure 2, Simulation Facility Repor .6 Conclusions on Ooerator Trainina and Qualification i The licensee developed examination material met the examiners' standards as j outlined in NUREG-1021 as superseded by GL-95-06. The written examination portion was evaluated as having a low discrimination value requiring more than average NRC effort to enhance its quality. The walkthrough and dynamic scenario )
examination portions were evaluated as having a medium discrimination value !
requiring less than average NRC effort to modify the examination materials to meet I
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the standard The applicants appeared well prepared for the examination and their overall performance was good. This was most evident during the dynamic simulator evaluations as demonstrated through effective use of three-way communications and crew mini-briefs during routine as well as abnormal plant condition V. Manaaement Meetinas X1 Exit Meetino Summarv ;
The chief examiner presented the examination team's observations and findings to members of the licensee's management on October 17,1996. The licensee acknowledged the findings presented. No proprietary information was identified during the examination or at the exit meetin .
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PARTIAL LIST OF PERSONS CONTACTED i Licensee
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K. L. Kofron, Byron Station Manager
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R. A. Colglazier, NRC Coordinator - Regulatory Assurance P. DiGiovanna. PWR Operations Training Supervisor - PTC T. E. Gierich, Operations Manager T. K. Higgins, Support Services Director S. W. Pettinger, Operating Training Supervisor j
, R. Wagner, Shift Operations Supervisor
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- S. Burgess, Senior Resident inspector
- N. Hilton, Resident inspector (*) Personnel not in attendance at the exit meeting on October 17,199 ITEMS OPENED, CLOSED, AND DISCUSSED
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NONE Closed
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! 50-454:455/96301-01 VIO failure to maintain control of examination material l Discussed
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t Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: Byron Generating Station, Units 1 & 2 Facility Licensee Docket No: 50-454, 50-455 Oparating Tests Administered: September 30 - October 3,1996 The following documents observations made by the NRC examination team during the October 1996, initiallicense examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b). 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 observation During the conduct of the simulator portion of the operating tests, the following item was observed-ITEM DESCRIPTION (1) Boric Acid Transfer The boric acid transfer pump's switch failed to Pump function upon demand during normal boration on two occasions. These were isolated cases and could not be I repeated by the simulator operato l (2) Condensate / Condensate The "A" CD/CB Lube Oil pump switch failed to Booster Pump 1 A function upon demand during one JPM. This was an isolated case and could not be repeated by the simulator operato (3) Loop 1D Steam Flow The steam flow recorder for loop D failed to Recorder properly ink at the start of one scenario. This problem
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was not readily repairable and required the detector to I be declared Out-of-Service during the scenari '
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- " "'4 t, UNITED STATES NUCLEAR REGULATORY COMMISSION
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5 [ 801 WARRENVILLE ROAD LISLE, ILLINOIS 60532-4351
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....+ l December 4, 1996 l l
NOTE T0: NRC Document Control Desk Mail Stop 0-5-D-24 l
FROM: Mary Ann Bies, Licensing Assistant t Operating Licensing Branch, RIII SUBJECT: OPERATOR LICENSING EXAMINATION ADMINISTERED THE WEEK 0F SEPTEMilER 30, 1996, AND OCTOBER 14, 1996, AT BYRON, DOCKET NOS. 50-454 AND 50-455 On September 30 to October 4, and October 14, 1996, 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 POR:
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[esignated t b) ae ' -
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Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE4 ,
l Attachments: As stated
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I U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR REACTOR OPERATOR LICENSE $. : REGION 3 i
CANDIDATE'S NAME: j MASTER EXAMINATION ;
I FACILITY: Byron 1 & 2 ]
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. REACTOR TYPE: PWR-WEC4 l
DATE ADMINISTERED: 96/10/14 INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of- ,
at least 80%. Examination papers will be picked up four (4) hours after the !
examination start ]
CANDIDATE'S TEST VALUE SCORE %
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100.00 % TOTALS I l
FINAL GRADE All work done on this examination is my ow I have neither given nor received ai Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS f
During the administration of this examination the following rules apply:
i 1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie 't A'fter the examination has been completed, you must sign the statement on
, the cover sheet indicating that the work is your own and you have not
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received or given assistance in completing the examination. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may
< leave. You must avoid all contacts with anyone outside the examination
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room to avoid even the appearance or possibility of cheating.
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4. lise black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and 3ach answer shee . Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG .
l 7. Before you turn in your examination, consecutively number each answer sheet,
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including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it out.
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9. The point value for each question is indicated in parentheses after the
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! 10. Show all calculations, methods, or assumptions used to obtain an answer to any short answer question . P.rtial credit may be given except on multiple choice questions. Therefore,
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ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25
. points, and you give five responses, each of your responses will be worth 0.20 points. If one of your five responses is incorrect, 0.20 will be i deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner onl ,
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i 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap pape . Ensure all information you wish to have evaluated as part of your answer is
, on your answer sheet. Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoke ,
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HIST N Used SRO Q# 1 DIFF M Answer !
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l - During Critical Safety Function Status Tree monitoring two (2)
l functions have Orange paths.
- One (1)bf the Orange functions is Heat Sin Which ONE of the following functions would take precedence over Heat Sink?
c. Inventor :
b. Containmen c. Integrity, d. Core Cooling.
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l K-A # 194001 A1.03 2.5 \ Sample Plan PWG A1.03 l Reference BAP 340-1 pp.15,17 l
COMMENTS suggested changing K/A to 194001 A1.08 or 15, A1.03 appears to be more correct,
- see attached K/A description page j ok per Max Bailey 10/10/96 i
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l HIST N Used l SRO Q# 2 DIFF c Answer Unit 1 has experienced a spurious reactor trip and safety injection 15 minutes ago. The operators just 1
l - Pressurizer level ......... 65% slowly increasing l - VCT levbl .:.................. 70% stable
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- S/G levels (all) ............ 40% stable
- RMCS M/U setting ...... 750 ppm l The Reactor Operator requests permission to adjust the RMCS M/U setting to a concentration greater than the RCS boron concentration. (note this action is addressed in step 12 of the procedure)
! The SRO should ... (Select ONE of the following)
a. ... permit the execution of the step since the early performance of the step WILL significantly mitigate the effects of the even b. ... permit the execution of the step since the early performance of the L step WILL NOT adversely impact other steps in the procedur c. ... NOT permit the execution of the step since the early performance of the step WILL NOT significantly mitigate the effects of the even d. ... NOT permit the execution of the step since the early performance of the step WILL adversely impact other steps in the procedure.
K-A # 194001 A1.11 2.8 \ Sample Plan PWG A1.11 Reference BAP 340-1 p.12 COMMENTS stem and distractors revised per Max Bailey
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I HIST M Used SIMILAR TO BRWD NRC SRO11 at SRO Q# 3 DIFF M Answer i
' The OPEN BULLETS in the following RNO step taken from BEP ES-1.1 "Si Termination". indicates that l
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b. CC system flow to RCPs - b. Establish normal CC NORMAL WITH THE FOLLOWING cooling to RCPs: ,
j ANNUNCIATORS NOT LIT: l l 1) Reset Cnmt isol Phase B,
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FLOW HIGH LOW i
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o 1CC685 c. ... ALL the valves must be opened sequentiall b. ... ALL the valves must be opened but may be opened in any orde c. ... only the valves which apply must be opened sequentiall d. ... only the valves which apply must be opened in any order.
! K-A # 194001A1.02 4.1 \ Sample Plan PWG A1.02 ;
R:ference BAP 340-1 p.12 l
COMMENTS revised to include BEP ES-1.1 step as example per Max Bailey
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_ HIST N Us:d SRO Q# 4 DIFF c Answer b.
With U-1 defueled and U-2 in Mode 4, the Shift Engineer slipped and fell in the Turbine Building.
There b one SRO in the control room. Radiation Technicians dispatched to administer first aid have reported to the control room that the Shift Engineer is unconscious and an ambulance is requested.
Which ONE of the following is required by Technical Specifications?
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c. Minimum manning is met, no other actions are require b. Prompt actions must be taken to replace the Shift Engineer within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> c. The control room SRO must assume the duties of the Shift Engineer within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ;
i d. Initiate actions within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to replace the Shift Engineer and replace the Shift Engineer within the following 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> !
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K-A # 194001A1.03 2.5 \ Sample Plan PWG A1.03 Reference Tech Spec 6.2.2 p. 6-2, Table 6.2-1 note (a) p. 6-5, and BAP 320-1 p. 2 COMMENTS distractor a and answer revised per Max Bailey
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HIST N Used SRO Q# 5 DIFF c Answer During which ONE of the following conditions may 1BEP ES-0.0. "Rediaanosis". be used?
a. If a twenty (20) gpm Steam Generator tube leak is detected while performing 1BEP ES-0.2. " Natural Circulation Cooldown ".
b. If a RED path is detected in Heat Sink while performing 1BEP-1. " Loss of Reactor Or Secondary Coolant".
c. If Sl is actuated erroneously while performing the actions of1BE ES-0.1. " Reactor Trio Recovery".
d. If the Main Steam line radiation monitors alarm while performing the actions of 1BEP-1. " Loss of Reactor Or Secondarv Coolant".
K-A # 194001 A1.02 4.1 \ Sample Plan PWG A1.02
' Reference BEP ES-0.0 Symptoms or Entry Conditions statement COMMENTS distractors and answer revised per Max Bailey l
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! HIST N Used l SRO Q# 6 DIFF M Answer c.
l l An operator is receiving a turnover on Unit 1 after retuming from a week of vacation. Which ONE of the
, following statements correctly reflects the requirements of BAP 335-1, Operating Shift Turnover and l Relie e. The on' coming operator shall review the Unit log through the last previous day on shift as time permits after assuming the shif b. Oncoming individuals shall walkdown the Unit control board immediately after assuming the shif c. The oncoming operator shall personally verify important Unit operating parameters prior to assuming the shif d. It b the responsibility of the offgoing operator to ensure their r: lief has met the requirements to maintain an active license prior to cssuming the duties of the shift.
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K-A # 194001A1.03 2.5 \ Sample Plan PWG A1.03 l Reference BAP 335-1 step C.1.c. page 2 COMMENTS distractor d revised per Max Bailey i
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l HIST B Used no l SRO Q# 7 DIFF M Answer c.
l l While reviewing the Unit 1 Aux Bldg Equipment Attendant Daily Logs, the operator notices a triple
, asterisk ("*) following a parameter.
Thb symbol denotes
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those parameters that are recorded ... (Select ONE of the following)
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c. ... at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> b. ... as part of NPDES requirements.
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c. ... to meet Technical Specifications requirements.
d. ... to meet requirements of a station commitment.
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! K-A # 194001A1.06 3.4 \ Sample Plan PWG A1.06 Reference BAP 350-5 item C.7. page 2, BAP 340-1 item C.1.d.4. page 4 COMMENTS distractor a [ originally d) and answer c revised per Max Bailey, stem wording altered ok per Max Bailey 10/10/96
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HIST N Used
- .SRO Q# a DIFF M Answer b.
l A General Emargency has been declared. All emergency response facilities are fully manned. Where l would the Operations Director be found? (Select ONE of the following)
l a. Control Roo '
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- b. Technical Support Cente c. Emergency Offsite Facilit d. Operations Support Cente .
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l K-A # 194001A1.16 3.1 \ 3.4 Sample Plan PWG A1.16 l Reference BZP 400-1 item C.3. page 1, BZP 100-T2 item 8. page 1
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l COMMENTS no change
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I HIST N 'ssed l SRO Q# 9 DIFF C Answer !
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An operator has been directed to Retum to Service the DC control power to the i 1 A Diesel Generator. When the operator requested another operator to accompany him for l
independent verification, he was told that the second verification will be performed by the next shift. l Thb b ... (Select ONE of the following) l l
c. ... the preferred method to perform an independent verification and l meets the definition of " APART IN ACTION" method of Independent Verificatio b. ... the preferred method to perform an independent verification and meets the definition of " APART IN TIME" method of Independent Verificatio c. ... an acceptable alterna~ive method to perform an independent verification if approved by an SR d. ... NOT an acceptable attemative method to perform an independent verificatio .
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l l l K-A # 194001K1.01 3.6 \ 3.7 Sample Plan PWG K1.01 Reference BAP 100-25 steps C.2 and 3. pages 1 and 2 COMMENTS stem and distractors a and b revised per Max Bailey
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revised distractor d. per Max Bailey 10/10/96
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HIST a Used no l SRO Q# 10 DIFF M Answer d.
! An OOS is being generated on Unit 1 involving an air operated valve that fails to a position other than l the OOS position. Which ONE of the following requirements applies to this valve for the OOS? I a. The control power MUST be removed and taken OOS to ensure properbrotection of personne b. The valve MUST be isolated by use of a freeze sea c. The instrument air should be isolated to the valve to provide more positive isolatio d. The valve should be held in its position by a mechanical block or other appropriate means, i
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- K-A # 194001K1.02 3.7 \ Sample Plan PWG K1.02 R:ference BAP 330-1, step C.4.c.7), p. 22 l COMMENTS stem revised per Max Bailey j
. distractor b. revised per J. Heaton, ok per Max Bailey 10/10/96 1
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HIST N Used SRO Q# 11 DIFF c Answer b.
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l Whenever a LOCKED valve is UNLOCKED ... (Select ONE of the following)
l C. ... and is manipulated and restored in accordance with a surveillance I test ;d shall be logged in the Abnormal Valve Lineup Log.
( b. ... and is closed at the request of the shift supervisor to secure a
! system leak it shall be logged in the Abnormal Valve Lineup Lo c. ... and in a position contrary to its procedural locked position, it shall be logged in the Locked Valve Lo d. ... and in any position, it shall be logged in the Locked Valve Log.
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K-A # 194001A1.06 3.4 \ Sample Plan PWG A1.06 Reference BAP 330-3 item C.1.g page 3 COMMENTS no change
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l HIST N Used
, SRO Q# 12 DIFF A Answer :
l A severe accident with core damage has occurred on Unit 1. Upon completion of ECCS switchover to i cold leg recirculation, a fault occurred on Bus 142. An operator is being dispatched to isolate a leak which b spraying onto the running 1 A RH pump. The RP Director has estimated a Dose of 18 rem TEDE to accomplish this task. Which ONE of the following correctly adheres to the restrictions for this
' - *
entry?
s. The entry SHOULD be allowed as long as the operator would NOT be cllowed to receive occupational exposure for the remainder of his/her career.
!
l b. The entry SHOULD be allowed since the estimated exposure is less than
- the limits for planned special exposure *
l l c. The entry SHOULD NOT be allowed since it would result in exceeding NRC l cxposure limit d. The entry SHOULD NOT be allowed since the estimated exposure exceeds l the limits for protecting safety related equipment.
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K-A # 194001K1.03 2.8 \ 3.4 Sample Plan PWG K1.03 R fsrence BRP 5300-2 item F.7.a. page 15 i COMMENTS distractors and answer CAPS revised per Max Bailey revised answer and distractor d. per Max Bailey 10/10/96
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l HIST N Used SRO Q# 13 DIFF c Answer c.
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r l A weld repair on a valve leak inside the missile barrier on Unit 1 is being planned. The unit will be held stable at 8% power. This is the only work planned for this containment entry and the estimated i dose rate will be 200 mrem /hr. The workers will radiograph the weld at the completion of the wor Which ONE of the following statements is correct conceming this work? '
t :
' An ALARA Action review is ...
I e. ... REQUIRED since unit 1 is at greater than 5% powe i l b. ... NOT REQUIRED since the dose rate is less than 500 mrem /hr.
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c. ... REQUIRED since the work includes radiograph ,
d. ... NOT REQUIRED since only a single job is being planned for this entry.
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K-A # 194001K1.04 3.3\ Sample Plan PWG K1.04 R ference BAP 700-2 item C.1.m page 3 COMMENTS ok, no change or comment, revised stem format GLW 10/11/96
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HIST N Used j SRO Q# 14 DIFF M Answer l
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! Which ONE of the following describes ALL the positions included in meeting the Fire Brigade
- requirements?
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l a. Shift Supervisor, Radwaste Supervisor, Equipment Attendants not requireti for safe shutdown of the unit.
, b. Shift Supervisor, Equipment Operators not required for safe L shutdown of the uni c. Redwaste Supervisor, NSOs and Equipment Attendants not required for safe shutdown of the uni d. Radwaste' Supervisor, Equipment Operators and Equipment Attendants not required for safe shutdown of the unit.
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l K-A # 194001K1.16 3.5 \ 4.2 Sample Plan PWG K1.16 Reference BAP 320-1 items C.1.c. and C.10 pages 2 and 4 i COMMENTS no change <
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l HIST N Used i SRO Q# 15 DIFF M Answer c.
l we isor i A valve four feet inside a valve room is producing a 1500 mrem /hr field at two4eet (30 centimeters)
l from the valv W I Which ONE of the following is the proper posting / method of control for this room?
i 0. Radiation Area b. High Radiation Area
- c. Locked High Radiation Area d. V;ry High Radiation Area
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K-A # 194001K1.03 2.8 \ Sample Plan PWG K1.03 R;ference BRP 5010-1 items F.1.d and F.3.d. , BAP 1450-2 item F. COMMENTS no problems, no change
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l HIST N Used SRO Q# 16 DIFF c Answer j l
Which ONE of the following operations could cause a Technical Specification Operability concem?
Closed tid c &
c. Fully gr. r.g ;r.d _y!^9gv Ive 1Sl8814,1A SI Pp Dsch Recirc isol V;lve, two times in one minut !'
- ie b. Adjusting the packing on 1FW006A,1 A FW Reg Upst Isol Valv c. Manually backseating valve 1Sl8812A,1 A RH Pp RWST Suct Isol Valve.
l d. Manually seating valve 1MS009D,1B MSR MS Shutoff Valv ,
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K-A # 194001K1.07 3.6 \ 3.7 Sample Plan PWG K1.07 R ference BAP 300-1 item C.3.bb. page 25
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COMMENTS added noun names for equipment, revised per Max Bailey
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i l HIST N Used SRO Q# 17 DIFF M Answer c.
l l With the unit in refueling, work is being performed in the Main Condenser waterbox which has been l designated as an Altemate Procedure Confined Space. Which ONE of the following is correct ,
! concoming controls or requirements for this work? '
- o. Flammhble gas concentration must be less than 2% of the lower explosive
! limit i
b. Personal oxygen meters are required while performing this wor )
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l c. Continuous forced air ventilation is required while performing the j wor !
- d. The entrahce ettendant must be respirator qualifie l l
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K-A # 194001K1.14 3.3 \ 3.6 Sample Plan PWG K1.14 Reference BAP 3000-7, step C.1.a., p.1 and step F.8.a.3), pp.11 - 13
- COMMENTS no problem, no change t
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l l HIST B Used no SRO Q# 18 DIFF C Answer c.
! During a surveillance test with the plant at 90% power, an instrument technician removes the iinstrument power fuses from N42 prior to bypassing channel N42 inputs. Assuming no operator l actions the control rods will ... (Select ONE of the following)
c. ... drive'IN'at 72 spm then drive OUT as temperature DECREASE b. ... drive OUT at 72 spm then drive IN as temperature INCREASE c. ... be blocked from moving OUT in AUTO and MANUA d. ... be blocked from moving OUT in AUTO onl .
l l K-A # 001000K1.05 4.5 \ Sample Plan PS GRP 1001 R:fsrence System Ip chapt. 31, pp.18,20, and figure 31-24 l
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l COMMENTS stem and selections formatting revised per Max Bailey
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, SRO Q# 19 DIFF C Answer The following conditions exist:
- A reactor startup is in progres With control bank C at 20 steps, the NSO stops rod withdrawa Two miriutes later, the NSO notes the following:
- SUR = +0.3 DPM and STABLE
- SR Count rate is 800 cps and INCREASING Which ONE of the following reflects required operator response to the above situation?
e. Perform a manual Reactor Tri b. Reinsert all rods and verify shutdown margi c. Continue the startup when SR counts stabiliz d. Initiate an Emergency Boration of 100 ppm and reinsert all control banks.
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l K-A # 001000G.13 3.7 \ Sample Plan PS GRP 1001 Reference BGP 100-2A1 CAUTION prior to step COMMENTS Rewrote stem M-C, changed 2 dists
- distractor a and answer wording revised per Max Bailey, psychometric arrangement
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i HIST N Used SRO Q# 20 DIFF c Answer d.
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l Compare the output of the reactor control unit, with rod control in auto, for the two conditions listed below:
- A 2% step change in turbine load at 90%.
l - A 2% stbp change in turbine load at 40%.
! Which ONE of the following describes the comparison of the signal output and the reason for the i difference?
c. Larger at 90% due to the response of the Variable gain unit b. Larger at 90% due to the response of the Non-Linear gain uni c. Smaller at 90% due to the response of the Non-Lineu gain uni d. Smaller at 90% due to the response of the Variable gain uni !
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K-A # 001000A1.06 4.1 \ Sample Plan PS GRP 1001 Reference System Ip chapt. 33 Power Range Nis pp. 24,25 COMMENTS no change f
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l HIST BM Used no l SRO Q# 21 DIFF C Answer a.
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l l The following conditions exist:
- RCS temperatuia - 340*F
- Steam Generator pressure - 50 psig
- A bubble' exists in the Pressurizer i Which ONE of the following statements would describe the initial primary plant response if a Reactor
- Coolant Pump were started?
RCS RCS temoerature pressure DECREASE DECREASE INCREASE DECREASE j l DECREASE INCREASE i l
l INCREASE INCREASE l
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K-A # 003000K1.10 3.0 \ 3.2 Sample Plan PS GRP 1003 Reference BOP RC-1 item D.3, a Tech Spec Bases for LCO 3.4.1
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COMMENTS Modified Stem, Changed answer, made into a table upper case decrease and increase, ok per Max Bailey i
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l HIST N Us@d !
i SRO Q# 22 DIFF C Answer .
Unit 2 was operating at 100% with Excess letdown in service and Normal Letdown secured. A :
l Reactor trip and Safety injection were manually actuated due to decreasing PZR pressure and :
l increasing CNMT pressure. Prior to SI Reset, what is the status of Excess Letdown Flow? (Select ONE !
l of the following) , ,
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G Secured due to closure of CNMT Instrument Air isolation valves.
i b. Secured due to closure of CNMT seal retum isolation valve c. Diverted to the RCDT due to 2CV8143 failing to the RCDT positio >
d. Diverted to the PRT via relief valve due to closure of CNMT isolation
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l K-A # 004020A1.08 3.0 \ Sample Plan PS GRP 1004 Reference System Ip chapt.15A Chemical and Volume Control pp. 54, 56 t
COMMENTS stem and selections revised per Max Bailey, need clarification from Max on intent of
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wording for a. see printout, corrected per Max Bailey
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HIST N Used SRO Q# 23 DIFF A Answer i With unit 2 in Mode 3 at 557 degrees a PZR level control malfunction has resulted in a PZR level
' DECREASE from 25% to 15%.
I
' Which one of the following describes the response of the CC system temperature and pressure to these conditions?
CC system CC system temoerature pressure INCREASE DECREASE INCREASE
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INCREASE i DECREASE DECREASE l DECREASE INCREASE l
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! K-A # 004000K1.18 2.9 \ 3.2 Sample Plan PS GRP 1004 l
Reference System Ip chapt.19," Component Cooling", pp.10,12,38,39
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COMMENTS stem corrected / revised per Max Bailey
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i HIST N Used SRO Q# 24 DIFF C Answer a.
' Given the following plant conditions:
- The RCFC's are in their normal lineu A loss of offsite power has occurred, resulting in a reactor tri Both EDGs' start and sequence on the safe shutdown load A Safety injection (SI) signal is subsequently received.
! Which ONE of the following describes the response of the Reactor Containment Fan Coolers (RCFCs)
- from the time the Si signal is received?
0. RCFCs trip on the Si signal, and restart on the vital bus, b. RCFCs trip on the Si signal, and must be restarted manually by the operator, c. RCFCs do not trip on the SI signal, and remain energized and running throughout the transien I d. RCFCs do not trip on the Si signal, and must be secured by the operator to prevent overloading the ED :
l K-A # 022000A3.01 4.1 \ Sample Plan PS GRP 1022 Reference System Ip chapt. 42, "CNMT Vent", p. 38; System Ip chapt. 61 ESFAS p. 52; VP 4030s l
COMMENTS no change ,
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i HIST N Used SRO Q# 25 DIFF C Answer Assuming 100% power and 75 gpm Letdown. Which ONE of the following failures will result in a loss i of makeup capability AND cause a loss of RCS charging if the Reactor Makeup Control system is in AUTO and NO operator action is taken?
VCT level trhnsmitter ... (Select ONE of the following)
c. ... LT-112 fails LO b. ... LT-112 fails HIG c. ... LT-185 fails LO d. ... LT-185 fails HIG !
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K-A # 004000K1.06 3.1 \ Sample Plan PS GRP 1004 R;ference System Ip chapt.15A, "CVCS", figure 15a-12, 'VCT level tree diagram" COMMENTS stem corrected / revised per Max Bailey corrected stem format GLW 10/11/96 l
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HIST N Used SRO Q# 26 DIFF M Answer l l
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Which ONE of the following isolations will initiate upon MANUAL Containment Spray actuation?
a. Feedwater
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b. Main Stearn Line c. Containment Phase A ,
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d. Containment Ventilation l
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K-A # 013000A4.03 4.5 \ Sample Plan PS GRP 1013 Reference System Ip chapt. 61," Engineered Safety Features", p. 36 COMMENTS stem and selections revised per Max Bailey, psychometric arrangement changed i answer to e revised distractor d. and psychemtrically swapped c and d per Max Bailey 10/10/96
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HIST N Used i SRO Q# 27 DIFF c Answer !
l The group demand counters for Control Bank D are at 94 step l l In order to satisfy Technical Specification limits, the digital rod position indications for rods in Control l B:nk D must indicate
, ,
from: (Select ONE of the following) ;
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c. 84 to 102 steps ,
b. 90 to 102 steps c. 84 to 108 steps d. 90 to 108 steps
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K-A # 014000G.05 3.1 \ Sample Plan PS GRP 1014 R;ference TecS Spec 3.1.3.1 and 3.1.3.2, BOS 1.3.1.1-2 data sheet D-5 COMMENTS no change
.. . . - _ - - - - _. - - . _. . - -.- _ - ._-..._ - -..~._-. -
i j HIST N Used 3 SRO Q# 28 DIFF c Answer a.
i j Which ONE of the following is the maximum allowable ADMINISTRATIVE deviation in power range N1 l channelindications following proper completion of the required actions of a calorimetric surveillance l st 90% power?
4 c. 0.5% diffe'rence between the highest channel and the lowest channe b. 0.5% difference between an individual channel and the average of all 4 channel c. 2% difference between the highest channel and the lowest channe d. 3% difference between an individual channel and the average of all 4 channels.'
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K-A # 015000A3.04 3.3 \ 3.5 Sample Plan PS GRP 1015 ,
i Reference 1BOS 3.1.1-2 step F.27 page 9 I l
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l HIST N Used SRO Q# 29 DIFF c Answer d.
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l A LOCA has resulted in steam voiding in the Reactor Vessel downcomer and fuel region I
!
Will Source Range indication INCREASE or DECREASE and what is the cause for the indication ch!ngs? (Se, lect ONE of the fo!!owing)
c. DECREASE due to lower thermal neutron leakag b. DECREASE due to lower fast neutron productio c. INCREASE due to higher fast neutron productio d. INCREASE due to higher fast neutron leakag K-A # 015000K6.01 2.9 \ Sample Plan PS GRP 1015 Reference Mitigating Core Damage Ip, MCD -9, " Accident Repsonse Excores", pp.14,16 l
COMMENTS no change i
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HIST N Used SRO Q#' 30 DIFF c Answer b.
The following conditions exist on Unit 1:
- Containment pressure transmitter PT-937 declared inoperable
- Required Technical Specification Actions have been taken for channel 937
,
Which ONE of the following statements describes the coincidence for a Containment Spray Actuation to occur and the actions that will result in this coincidence?
e. 2/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A) in the BYPASS positio i
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b. 2/3 coincidence after the channel is placed in the BYPASS condition, by placing' bistable (PB-937A) in the TRIP positio c.1/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A) in the BYPASS positio d.1/3 coincidence after the channel is placed in the BYPASS condition, by placing bistable (PB-937A) in the TRIP positio K-A # 026000A4.01 4.5 \ Sample Plan PS GRP 1026 i Reference Technical Specification 3.3.2 ACTION c, Table 3.3-1 FU 2.c ACTION 16, pages 3/4 3-15 & 3/4 3-21; BOA INST-2 Att J step 2 p. 56; System Ip chapt. 61 ESFAS pp. 38,39 - ;
COMMENTS distractors c and d; K/A, revised per Max Bailey
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HIST BM Ussd N-96-01-05 SRO Q# _31 DIFF C Answer i l
Which ONE of the following actions occurs automatically when pressure at the feed pump suction ;
decreases 6.3% below the NPSH Low alarm setpoint? :
,
12. The Low Press Heater bypass valve (CB025) modulate OPE .
. b. The Gland Steam Condenser bypass valves (CB157A&B) OPE ;
c. The Heater Drain pump discharge valves (HD046A&B) modulate l CLOSE :
d. The Main Feedwater Pump recirc valves (FWO12B&C) CLOS ii i
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K-A # 056000K1.03 2.6 \ Sample Plan PS GRP 1056 Reference BAR 1-16-E1, System Ip chapt. 25 Condensate and FW pages 92,93 l COMMENTS ok, no change, rearranged selections to make answer b instead of c for correct answer distribution '
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l HIST BM Used 96-01-FIN syst::m fin SRO Q# 32 DIFF c Answer c.
I 1B steam generator level INCREASES to 84% causing a feedwater isolation. Following the reactor trip, steam generator level retums to normal level and Tave DECREASES to 557'F.
] Which ONE of the following describes when the Feedwater isolation will/can be reset?
,
, c. As soon as 1B steam generator level decreases below 81.4%.
,
b. As soon as 1B steam generator level decreases below 81.4% and the
- feedwater isolation resets are depressed.
j c. After the reactor trip breakers are closed and the feedwater isolation
- resets are depresse ,
d. After Tave increases to above 564*F and the feedwater isolation resets cre depressed.
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K-A # 059000A4.11 3.1 \ Sample Plan PS GRP 1059 Reference System Ip chapt. 25 Condensate and FW pages 124,126,128,130 COMMENTS Upper case incr/decr, ok per Max Bailey
-. .. . . . - . - - -. - -__ .._-_- -- . . - - - . . . - - - ...
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] HIST s Used no 4 SRO Q# 33 DIFF C Answer b.
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The 18 Auxiliary Feedwater Pump started on a safety injection signal. Ten (10) seconds after it started,
- it tripped and a Main Control Board alarm " Aux FW Pump Diesel Trouble" was received. Which ONE of
, the following is the cause of the diesel trip?
,
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s. Overcrhnk i
! b. Low oil pressure
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c. High Fuel Oil Filter AP
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d. High Crankcase Pressure i ,
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I K-A # 061000A2.04 3.4 \ Sample Plan PS GRP 1061 Reference BAR 1-3-A6 COMMENTS no change
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HIST s Used no SRO Q# 34 DIFF M Answer The plant had received annunciator 125VDC PANEL 111/113 VOLT LOW. Which ONE of the following is an indication that ONLY Bus 113 is deenergized?
a. No DC volts on Main Control Boar '
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l b. Normal DC volts on main control boar l c. Fuses inside ground detection cabinet blow I
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d. Annunciator 125 VDC BUS TIE BREAKER TO BUS 211/111 CLOSE/ TRIP LI i I
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K-A # 063000A3.01 2.7 \ Sample Plan PS GRP 1063 Ref.:rence BOA ELEC-1 step 1
' COMMENTS no change
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i HIST B Used no
- SRO Q# 35 DIFF M Answer c.
Which ONE of the following is the HIGHEST pressure that would be seen in a gas decay tank BEFORE i the system autometically transferred to a standby tank?
i j c. 60 psig y s.
! b. 75 psig i
c. 95 psig d.110 psig i
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K-A # 071000K4.01 2.6 \ Sample Plan PS GRP 1071 Reference System Ip chapt. 46, " Gaseous Radwaste", p. 24 COMMENTS changed K&A per Max Bailey
-. .-. - . . -. . - -_ .
- - . . - -. --- __
HIST BM Us d no
- SRO Q# 36 DIFF M Answer Which ONE of the following actions will occur as a result of a High Radiation alarm on ORE-PR032 (Control Room Outside Air intake Monitor)?
-
Makeu Control Room turbine Recirculation Air' building air Charcoal Absorber
. Fan intake inlet / Outlet Dampers Starts Opens Open Stops Shuts Open 1 Starts Shuts Shut i
4 Stops Opens Open
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- K-A # 072000K1.04 3.3 \ Sample Plan PS GRP 1072
.
Reference BAR RM11-2-0PR32J; System Ip chapt. 43b, " Control Room HVAC", pp. 28,29,38
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COMMENTS Max requested changing to K4.01 - Knowledge of ARM system design feature (s)
and/or interlock (s) which provide for the following: Containment ventilation isolation, K1.04 is more applicable, ok per Max Bailey 10/10/96
HIST N Used SRO Q# 37 DIFF C Answer c.
Given the following plant conditions:
- Unit 1 is in solid plant operatio Train A RHR loop in servic RH letdown is in service j
- Charging is in servic l
- Letdown Pressure Control Valve, PCV-131, is inadvertently close l Which ONE of the following describes the plant response as a result of the closure of PCV-1317 l
m. Train A RHR flow will DECREASE due to valve closure, RCS pressure will l INCREASE due to resulting heatu !
b. The RCS Pressure will DECREASE due to letdown line relief flow being greater than charging flo c. RCS pressure will INCREASE due to continued charging flow until the Cold Overpressure Protection system actuate d. No effect on the RCS because auto control of charging flow will- i mrintain balanced conditions between letdown and chargin ,
K-A # 002000K4.10 4.2 \ 4.4 Sample Plan PS GRP 2 002 Reference System Ip chapt.18, " Residual Heat Removal", p. 24; Byron Curve Book, figure 29; BOP RC-1a, step F.20., p.16 COMMENTS distractor b revised per Max Bailey
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HIST M Used cert. at SRO Q# 38 DIFF c Answer The plant is operating at 75% power and the latest leak rate data shows:
- Total RCS leakage rate .................................... 4.5 GPM
- Leakage into the Pressurizer Relief Tank ........ 1.5 GPM
- Leakage.into the Reactor Coolant Drain Tank . 1.2 GPM
- Total primary to secondary leakage ................. 0.05 GPM-Which ONE of the following Technical Specification leakage limits has been exceeded?
c. Identified b. Unidentified
.
c. Pressure Boundary d. Primary to Secondary K-A # 002000G.05 3.6 \ Sample Plan PS GRP 2 002 Reference Tech Spec LCO 3.4. !
COMMENTS 0.05 * 60 = 3 gph * 24 = 72 gpd 4.5 -(1.2 + 1.5 + 0.05) = 1.75 gpd which is > 1.0 gpm unidentified, ok ;
revised stem per Max Bailey 10/10/96 l l
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HIST B Used no
- OrtO Q# 3s DIFF c Answer a.
a J
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RCS leakage is occurring on Unit 1. The NSO has reported that he suspects the leak to be from a PZR
'
code safety valve. Which ONE of the following reported conditions is NOT an indication for the diagnosis of a leaking PZR code safety valve? .
f c. PZR Level DECREASING.
! b. PZR safety valve OPEN indicating light LI c. PRT temperature and pressure INCREASING.
j d. PZR SAFETY REllEF DISCHARGE TEMPERATURE HIGH annunciator LI ,
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K-A # 002000K4.05 3.8 \ Sample Plan PS GRP 2 002
l Reference System Ip chapt.14, Pressurizer, simulator demonstration p. 34
, COMMENTS revised dist c. per Max Bailey l
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HIST N Used SRO Q# 40 DIFF C Answer b.
Given the following plaint conditions:
- A Unit 1 heatup is in progres RCS temperature is 150 degrees RCS pressure is 350 psi RHR system is in service.
Which ONE of the following conditions must be corrected before RCS temperature is increased to 210 degrees F7 c.1B RH Pump Out of Servic b.1B CS Pu'mp is Out of Servic c. Loop A ucumulator boron concentration is 2170 pp ,
d. 0A and OE SX Tower High Speed fans are out of servic I l
l K-A # 006000G.05 3.5 \ 4.2 Sample Plan PS GRP 2 006 Reference Tech Spec 3.6. COMMENTS no change
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i HIST N Used i SRO Q# 41 DIFF C Answer a.
The following conditions exist at Unit 1:
- Mode 4 - RCS Temp / Press is 210 Deg. F1450 psig
- Maintenance is being performed on the Containment Pressure Detectors
- An Si actuation occurs Which ONE of the following statements describes the response of the ECCS Accumulators; and the reason for that response?
c. The Accumulators will not discharge into the RCS; because the outlet valves are shut with their power supply locked-ou b. The Accurnulators will not discharge into the RCS; because RCS l pressure is less than P-11 (1950 psig). I c. The Accumulators will discharge into the RCS; because the outlet l vIlves are interlocked to open on an SI Signa l d. The Accumulators will discharge into the RCS; because the outlet v lves are open with their power supply locked-ou '
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K-A # 006000K6.02 3.4 \ Sample Plan PS GRP 2 006 Reference 1BGP 100-5 step 35 page 23 COMMENTS replacement for 41, need approval from Max Bailey ok per Max Bailey 10/10/96, corrected spelling in answer and d. per Pete Peterson 10/11/96
HIST N Used SRO Q# 42 DIFF c Answer b.
The following conditions exist: i
- Unit is at 100% powe Pressurizer backup heater groups A and B are positioned to O RCS wide range pressure is 2235 psig.
The selected controlling pressurizer pressure channel input fails to 1900 psig over 5 seconds.
Assuming NO operator action which ONE of the following describes the plant response to this event?
Actual Pressurizer pressure ...
c. ... INCREASES to 2385 psig resulting in a reactor trip signa b. ... INCREASES to 2335 psig and one PORV cycles to control pressur !
c. ... DECREASES to 2210 psig and bacr.up heater groups cycle to maintain a reduced pressur j d. ... INCREASES to 2260 psig and spray valves open to stabilize pressur !
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K-A # 010000K3.01 3.8 \ Sample Plan PS GRP 2 010 Reference System Ip chapt.14, " Pressurizer", p. 70 I i
COMMENTS stem and distractor d revised per Max Bailey initial review 4
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3 HIST N Ussd j SRO Q# 43 DIFF M Answer a.
- A fire b occurring in the Upper Cable Spreading Room directly above the unit 1 desk that has resulted
- in Fire Detection alarms on 1PM09J from ALL the detectors in the area. Which ONE of the following is l, the initial operator response regarding Fire Suppression to this even c. Verify Automatic Halon Suppression Actuation.
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b. Verify Automatic Carbon Dioxide Suppression Actuatio I c. Dispatch an operator to locally actuate Halon Suppression.
i l d. Dispatch an operator to locally actuate Carbon Dioxide Suppression.
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l K-A # 086000K4.05 3.0 \ 3.4 Sample Plan PS GRP 2 086 R:fsrence System Ip chapt. 57, " Fire Protection", p. 24,25,28,29 COMMENTS suggested replacing submitted question, new question to replace 43, need review by Max ok per Max Bailey 10/10/96
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I HIST N Used SRO Q# 44 DIFF c Answer c.
J d
d Given the following plant conditions:
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- - Reactor Power at 60%.
! - Pressurizer pressure is 2240 psig.
i - Charging flow is being controlled in MANUAL.
- All other contro!a are in AUTOMATIC.
l -The backup heaters just ENERGlZED.
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[Which ONE of the following is the pressurizer level the operators should see given these conditions?
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! O.41 % -
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b. 46%
c. 51%
d. 56%
K-A # 011000K6.03 2.9 \ Sample Plan PS GRP 2 011 Reference System Ip chapt.14, " Pressurizer", p. 56 COMMENTS heaters energizing automatically indicates a +5% level deviation ok, no change
HIST BM Used no SRO Q# 45 DIFF c Answer a.
The plant is at 70% power. All control systems are in AUTO. PZR level control selector switch is in the 461/460 position while IMs are troubleshooting erraHtcperation of level channel 459.
While in this lineup, level channel 461 fails AS IS s. :
Load b subsequently INCREASED to 90% power. Assuming no operator action, which ONE of the following represents the plant response when power is INCREASED to 90%?
Cha flow Ltdwn flow Hi Ivl Rx trio c. INCREASE CONSTANT YES b. DECREASE ISOLATE YES c. DECREASE
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CONSTANT NO l d. INCREASE ISOLATE YES l l
l K-A # 011000A1.01 3.5 \ Sample Plan PS GRP 2 011 Reference System Ip chapt.14, " Pressurizer", pp. 74,76,78 COMMENTS changed stem to a power increase, change to answer a, need review by Max Bailey ok per Max Bailey 10/10/96
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HIST N Used SRO Q# 46 DIFF c Answer I I
Given the following plant conditions:
- Reactor startup in progress with Power Range N43 out of service
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- Source range indication is at 12,000 cps 1
- RO notices'the " POWER ABOVE PERMISSIVE P-10" lamp deenergizes and l
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at the same time BOTH Source Range instruments de-energize
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' Which ONE of the following is correct regarding this condition?
c. Reactor startup may continue after both source range block switches are placed to RESET.
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! b. Reactor startup must be terminated, since there is no longer an indication of reactor power leve c. R actor startup may continue, as long as both Intermediate Ranges indications are on scal d. R actor startup must be terminated, since no SR channels are operabl K-A # 015000G.05 3.3 \ 3.8 Sample Plan PS GRP 1015 R;ference BOA INST-1, " Nuclear Instrument Malfunction", Att. C, "SR Channel Failure", step 4 page 11 COMMENTS K/A number revised per Max Bailey, the revision changed the count for PS GRP 1 and 2,015 is GRP 1 ok per Max Bailey 10/10/96
HIST BM Used no SRO Q# 47 DIFF C Answer The following plant conditions exist:
- 80% power, steady state
- Rod controlis in AUTO
- PZR Lekt tontrol is in MANUAL while troubleshooting erratic operation of the level controlle Which ONE of the following statements describes the plant response t7 an RCS narrow range cold leg temperature transmitter failure?
c. A HIGH failure will cause rods to WITHDRAW and pressurizer level to ;
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INCREAS '
i b. A HIGH failure will cause the OPAT setpoint to INCREASE and control I rods to INSER ,
c. A LOW failure will cause the OTAT setpoint to INCREASE and loop AT to INCREAS !
d. A LOW failure will cause rods to WITHDRAW and pressurizer level to !
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INCREAS l
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K-A # 016000A2.01 3.0 \ 3.1 Sample Plan PS GRP 2 016 Ref;rence System Ip chapt.14, " Pressurizer", p. 80, 84 COMMENTS leave as is, no change
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I HIST M Ussd 1996 BRWD NRC at ,
SRO Q# 4a ' DIFF c Answer :
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The unit is operating at 100% power and all systems are in automatic. Steam pressure channel (PT-535) fails HIGH. (PT-535 inouts to the CONTROL i ING stamm flow channel 3 Which ONE of the following describes the system response AND the operator action required?
Feed flow will ...
o. ... INCREASE, the operator must select the attemate steam FLOW channe b. ... INCREASE, the operator must select the attemate steam PRESSURE channe c. ... DECREASE, the operator must select the attemate steam PRESSURE channe ,
d. ... DECREASE, the operator must select the attemate steam FLOW channe i
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j K-A # 035010A2.03 3.4 \ 3.6 Sample Plan PS GRP 2 035 Reference System Ip chapt. 27, "S/G Water Level Control", pp. 38, 39; BOA INST-2, " Operation with a Failed Instrument Channel", Att. F, p.39 COMMENTS changed to different steam pressure channel, revised per Max Bailey m - , -+%,--
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- SRO Q# 49 DIFF M Answer l l
, The Fuel Handling Supervisor has just notified the Control Room that audible counts in CNMT ceased
' as they were lifting a fuel assembly from the upender. The NSO finds that audible count rate in the control room has also been lost. The Fuel Handlers have suspended core alterations.
l Which ONE bf the following is a requirement prior to resuming core alterations?
c. Set the audible count rate selector switch to the other channel, verify cudible count rate to Control Room and CNMT with one SR channel operabl b. Initiate emergency boration until it can be verified that all filled portions of the RCS are at least 2300 ppm boron concentratio c. Set the audible count rate selector switch to the other channel, verify cudible count rate to Control Room and CNMT with both SR channels operabl '
d. Set the audible count rate selector switch to the other channel, verify cudible count rate to Control Room with one SR channel operable.
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K-A # 034000K1.04 2.6 \ Sample Plan PS GRP 2 034 Reference Tech Spec 3.9.2, BOA INST-1, " Nuclear Instrument Malfunction", Att. C, "SR Channel Failure", step 2 p.10 COMMENTS revised stem per Max Bailey, needs review revised distractors a., d., and answer per Max Bailey 10/10/96 l
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4 gr 4 y c -- n- -+.m-
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HIST BM Used C-96-01-02
!' SRO Q# 50 DIFF M Answer a.
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Which ONE of the following describes how power is NORMALLY supplied to 120 VAC instrument bus
.' 1117
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- a. 480 VAC from MCC 131X2; rectified to 125 VDC and inverted to 120 VAC.
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{ b. 480 VAC from MCC 131X2 and transformed to 120 VAC.
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c.125 VDC from battery; supplied to battery bus and inverted to 120 VAC.
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d. 480 VAC from Bus 133X and transformed to 120 VAC.
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K-A # 062000K2.01 3.3 \ Sample Plan PS GRP 2 062 l Reference System Ip chapt. 4, "AC Electrical Power", pp.12,74,76
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COMMENTS inserted actual power supply references, revised per Max Bailey
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HIST BM Used no i SRO Q# 51 DIFF C Answer !
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l Unit 1 b operating at 100% when annunciator 1-20-C4, SAT 142-2 SUDDEN PRESSURE alarms.
] Which ONE of the following correctly describes the plant response to this event?
c. Reactor trips due to loss of 2 RCP' t :
b. Reactor trips due to activation of Main Generatr>r 86 rela c. No Reactor trip occurs but T.S. LCOAR entry is required for offsite power source d. N3 Reactor trip occurs but T.S. LCOAR 3.0.3 entry is required for loss of power to ECCS equipmen ,
K-A # 062000A2.01 3.4 \ Sample Plan PS GRP 2 062 Reference BAR 1-20-C4," SAT 142-2 Sudden Press" COMMENTS Modified stem conditions and two dists SAT referenced revised per Max Bailey
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HIST N Us d I SRO Q# 52 DIFF c Answer l
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Given the following ,
l - Diesel 1 A is carrying bus 141 following an inadvertent trip of ACB1412 SAT feed breake A Safet9 injection actuation has loaded the bu The Operator goes to RAISE on the diesel's speed AND voltage controls for three (3) seconds.
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Which ONE of the followng describes the Diesel Generator response to this action?
VOLTS EEFA Mwn KVAR
' UP SAME SAME UP SAME SAME UP UP - UP UP SAME UP SAME SAME SAME SAME
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K-A # 064000A2.02 2.7 \ Sample Plan PS GRP 2 064 l
- Reference System Ip chapt. 9, "DG and Auxiliaries", p. 34
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COMMENTS no change i i
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l HIST BM Used 1994 BYRON NRC ct SRO Q# 53 DIFF c Answer Unit 2 has been operating at full power for 25 days when the NSO reports that containment average air temperature is indicating 128' In the event of ... (Select ONE of the following)
s-c. ... any accident NO effect will be seen since the accident analysis includes a 10% temperature instrument erro b. ... a high energy line break in containment, ADVERSE containment numbers used in the Emergency Procedures may not be vali c. ... a steam line break inside containment, the peak pressure assumed I in the a'ccident analysis may be exceede d. ... a large break LOCA, ECCS accumulator N2 would be injected into the RC K-A # 103000G.06 2.8 \ Sample Plan PS GRP 2103 Reference Toch Spec Bases 3/4.6.1.5, " Air Temperature", p. B 3/4 6-2 COMMENTS stem and selections revised per Max Bailey t
a
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HIST BR Used C-96-01-01 SRO Q# 54 DIFF A Answer Unit 2 b operating at 100% power when the Unit 2 Component Cooling Water Heat Exchanger l
< develops a tube leak. The Unit 2 CC HX is isolated on October 5,1996 and the Unit 0 CC HX is '
placed in service. Mechanical maintenance has determined that repairs will NOT be completed until October 15,1996 (due to parts availability). What will the status of the unit be when repairs to the heat cxchanger are comoleted ? (Select ONE of the following)
c. Mode 1
- b. Moda 3 i
i c. Mode 4 d. Mode 5 l
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K-A # 008000G.05 3.3 \ Sample Plan PS GRP 3 008 Reference Tech Spec 3.7.3 Action COMMENTS stem revised to include dates per Max Bailey
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HIST BM Used no SRO Q# 55 DIFF A Answer b.
l The following conditions exist on Unit 1:
- P owe r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 5 %
- PT-505 (Mn Turb First Stage Imp Pressure) ... 0%
- PT-506 (Mn Turb First Stage imp Pressure) ... 82% i
- Tave . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 574 * F I
- The actions of1EQ A INST-2. Attachment D. " Turbine Imoulse I Pressure Channel Failure" have been take l I
jWhich ONE of the following failures would result in the AUTOMATIC OPENING of the steam dumps?
c. PT-507 (Mn Steam Header Pressure) fails LOW
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b. PT-507 (Mn Steam Header Pressure) fails HIGH c. 'PT-506 (Mn Turb First Stage Imp Pressure) fails HIGH I d. PT-506 (Mn Turb First Stage imp Pressure) fails LOW l s
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i K-A # 041020A1.02 3.1 \ Sample Plan PS GRP 3 041 f
R;f:rence BOA INST-2 Ip, " Operation with a Failed instrument Channel", pp. 42,43; System Ip chapt. 24, " Main Steam Dumps", pp. 34 - 36 f
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COMMENTS changed dist b, c, modified stem stem and distractor a revised per Max Bailey, changes resulted in new answer b
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l HIST N Used SRO Q# 56 DIFF M Answer l l
Which ONE of the following conditions will DIRECTLY cause a trip of an Essential Service Water (SX) l Pump cfter it has been started manually? '
i c. SX Pump suction pressure 4 psig for 18 second l
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j b. RCFC Inlet valve (SX-016 A/B) goes shu c. SX Pump discharge strainer delta P greater than 6 psi d. SX Pump lube oil pressure 4.5 psig for 15 second .
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K-A # 076000G.10 2.7 \ Sample Plan PS GRP 3 076 Reference BAR 1-2-A1, "SX Pump Trip"; BAR 1-2-C1, "SX Pump Suct Press Low"; System Ip chapt. 20, " Essential Service Water", p. 60 COMMENTS changed K&A per Max Bailey revised distractor b per Max Bailey 10/10/96
_ _ HIST CM Used no SRO Q# 57 DIFF C Answer c.
Assume the following conditions exist:
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-Instrument air pressure ............... 70 psig and decreasing
- Rx Power . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 % ;
- Startup Feedwater pump ............. In service
- A and B CD/CB pumps ................ In service
- 120 gpm letdown in service Assuming instrument Air pressure continues to DECREASE resulting in a Reactor Trip. Failure of which ONE of the following components caused the Reactor Trip?
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c. Letdown Isolation valves. (CV8152, CV8160)
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b. CD/CB Pump Recirc valve. (CD152)
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c. Main Feedwater Reg Bypass valves. (FW 510A, 520A, 530A, 540A) j d. Charging flow control valve. (CV121)
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K-A # 078000K3.02 3.4 \ 3.6 Sample Plan PS GRP 3 078 Reference 1 BOA SEC-4, " Loss of instrument Alf', Table A, p. 7 COMMENTS revised stem and distractors per Max Bailey l
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HIST B Used no
' SRO Q# 58 DIFF c Answer d.
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l The following pertinent plant conditions exist:
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Reactor Power: ....................... 88% - stable Tcve: ....................................... 577 deg F - slowly increasing Turbine load: ........................... 1025 MW - stable
> Control Bank D Rod Position: .180 steps Rod Control: ............................. AUTO
, The operator observes that the Control Bank D rods start stepping out at 72 steps per minute. The operator is required to manually trip the reactor if rods continue to move after the operator has ...
(Select ONE of the following)
l c. ... placed the rods in MANUAL and then cycled the ROD CONTROL IN-HOLD-OUT switch.
i b. ... selected MANUAL from AUTO and then reselected AUT c. ... placed the rods in AUTO from MANUAL with the rods ratcheting at the top of cor ,
i d. ... selected Shutdown Bank D after cycling the ROD CONTROL IN-HOLD-OUT switch while in MANUAL.
K-A # 000001G.12 3.7 \ Sample Plan EPE GRP 1000001 Reference BOA ROD-1," Uncontrolled Rod Motion", step 2, COMMENTS revised question (new question?), need review by Max Bailey ok per Max Bailey 10/10/96
HIST BM Used no SRO Q# se DIFF C Answer The following plant conditions exist:
- Control bank D, rod M-12 bottom light ..........'..... LIT L
- Reactor Power .. . .... . ... . . ... ..... .... . .. . . . .. . .. ... . ... .. .. . .. 65%
- Tave .. ..'. . . . /. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . D EC R EAS I N G i
- ROD CONTROL URGENT FAILURE alarm ...... LIT
- Turbine load is being adjusted to stabilize Tave l While preparing to withdraw rod M-12 per BOA ROD-3. "Drooned or Misahgned Rod". the operator places the Rod Control IN-HOLD-OUT switch to the OUT position AND control bank D rod H-8 drops to the bottom of the core. Which ONE of the following describes the actions required for this event?
c. Trip the reactor and enter B5P- b. Reduce turbine load to stabilize Tav c. Verify SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Hot Standby within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> d. After clearing the Urgent Failure condition, sequentially recover each rod, while adjusting turbine loa i i
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K-A # 000003G.07 3.4 \ Sample Plan EPE GRP 1000003 Reference BOA ROD-3, " Dropped or Misaligned Rod", step 4.b RNO , p. 4 COMMENTS modified stem, changed 3 dists question stem revised per Max Bailey
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l HIST su Used N-96-01-03 !
SRO Q# so DIFF M- Answer ;
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[ Which ONE of the following is the reason for maintaining the control banks above the setpoint for the l
! " ROD BANK LO-2 INSERTION LIMIT" alarm? -
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I l i c. Ensures the maintenance of acceptable power distribution limits
- maintains. minimum shutdown margin; and limits the potential effects of l j rod misalignment on the associated accident analysis.
l b. Ensures adequate shutdown margin; maintains acceptable core thermal
! limits and limits the potential effects of rod misalignment on power operations.
l c. Ensures adequate DNBR; limit fission gas release; and maintains fuel
- pellet iemperature and cladding mechanical properties within design j criteria of the associated accident analysis.
j j d. Ensures that additional restrictions on thermal power and increased
! frequency of peaking factor measurements are not required.
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i K-A # 000005EK3.02 3.6 \ Sample Plan EPE GRP 1000005 Reference Tech Spec Bases 3/4.1.3, Movable Control Assemblies, p. B 3/41-4 COMMENTS made into multiple choice question ok, no change revised distractor c. per Max Bailey 10/10/96
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HIST BM Used no DIFF A
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SRO Q# 61 Answer c.
b j i
i A reactor trip with Sl occurred at 0100. At 0630, the control room operators were directed to enter l
BCA-1.1. " Loss of Emergency Coolant Recirculation". The operators are now determining if Si flow l can be terminated. The following conditions currently exist at 0700: '
.
I - Total Slhow rate ..... 600 gpm j - CNMT Press ............ 3.8 psig
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- CETC temperature .. 500 *F All RCPs .................. OFF
] - RVLIS Plenum ........ 55%
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Based on these, indications and the attached Step 15 and associated figures, the control room
, operators should: (Select ONE of the following)
i c. Terminate Si and establish 70 gpm charging flow, b. Reduce total Si flow rate to approximately 150 gpm.
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c. Reduce total Si flow rate to approximately 220 gp :
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d. Reduce total Si flow rate to approximately 360 gpm.
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K-A # 000011G.12 4.0 \ 4.1 Sample Plan EPE GRP 1000011 i Reference BCA-1.1, Loss of Emergency Coolant Recirculation", step 15, pp.12,30,33,34 l
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COMMENTS Modified stem and all responses Attach step 15, Figure 1BCA 1.1-4, and 1BCA 1.1-1 ok, no change revised stem per Max Bailey 10/10/96
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l HIST BM Used 1994 Byron NRC ct SRO Q# 62 DIFF c Answer a.
, Unit 1 b at 25% power during a power ascension when the following annunciators come in:
j - RCP 1A Brkr Open or Low Flow Alert
- - RCP Trip ,
The control room operators should ... (Select ONE of the following)
c. ... perform a manual Reactor Tri b. ... verify an automatic Reactor Tri c. ... reduce, reactor power to less than 10%.
d. ... place the unit in Hot Standby within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ,
l K-A # 000015G.07 3.1 \ Sample Plan EPE GRP 1000015 Reference BAR 1-13-A3, "RCP 1 A BRKR Open or Flow Low Alert"; BAR 1-13-E3, "RCP Trip" COMMENTS Modified stem conditions and 2 dists answer and distractor b revised per Max Bailey
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HIST N Used SRO Q# 63 DIFF M Answer a.
Which ONE of the following is the reason for promptly closing the seal leakoff isolation valve for a RCP with a high number 1 seal leakoff once the RCP has stopped rotating?
c. Protect number 2 seal from possible debris from the number 1 sea b. Prevention of damage to the thermal barrier due to high flo c. Minimize the amount of RCS water that is routed to containment sum ;
d. Assure a minimum back pressure is maintained on the number 3 )
sea l l
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K-A # 000015EK2.07 2.9 \ Sample Plan EPE GRP 1000015 R:ference BOA RCP-1 Ip, " Reactor Coolant Pump Seal Failure", steps 7,8, pp.10,11 COMMENTS ok, no change
, HIST N Used SRO Q# 64 DIFF c Answer i
! On o reactor trip four (4) control rods did NOT fully insert. The SCRE directs you to Borate using t; l RWST. Charging flow is currently at 132 gpm.
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! Which ONE of the following is the MINIMUM boration time to ensure adequate shutdown margin ,
i requirements are satisfied. (assuming constant charging flow) '
e.1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> )!
b. 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> I c. 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> d. 3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />'
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K-A # 000024EK1.03 2.4 \ Sample Plan EPE GRP 1000024 )
Rcference BEP ES-0.1, " Reactor Trip Response", step 5 RNO, p. 5 COMMENTS stem and all selections revised per Max Bailey, increased charging flow from suggested (110) to 132 due to normal plant expected flowrates for 120 gpm letdown I flow. 4*3600/132 gpm = 109.1 min vs. 4*3600/110 gpm = 130.9 min, ok 10/10/96 f
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HIST N i SRO Q# as DIFF M Answerg H
Which ONE of the following is an acceptable neans of executing a Technical Specification required i emergency boration?
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c. Using rpanual boration valve 1CV8439 full ope ..
i b. Using normal boration system set to BORATE; with the flowpath verified ,
! through the 1CV110B BA Blender to VCT Outlet at a rate of 40 gp l
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l c. From the RWST via 1CV112D RWST to CENT CHG pump suction valve at a l
. rate of 100 gp I I l j d. From the BAT via 1CV8104 Emergency Boration valve and a Boric Acid l Transfer pump at a rate of 28 gpm.
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! K-A # 000024EK3.02 4.2 \ 4.4 Sample Plan EPE GRP 1000024
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Reference PRI-2, " Emergency Boration", step 1, pages 2-3; ILT Simulator Phase Lesson Plan
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PRI-2, pages 4-6; System Ip chapt.15b, " Reactor Makeup Control", p. 66 ;
COMMENTS ok, no change revised distractor b. per Max Bailey 10/10/96
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HIST N Usad SRO Q# 66 DIFF M Answer b.
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Which ONE of the following is the NEXT action the operator is required to take if the main turbine 1oes
. NOT trip automatically and CANNOT be tripped from the MCB; per BFR-S.1 "Resoonse to Nuclear
! Power Generation /ATWS"?
l c. Shut tO MSIVs.
l b. Manually RUNBACK the turbin c. Trip the turbine locally at the pedesta d. Place both EH Pump control switches in PULL OUT.
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l K-A # 000029G.10 4.5 \ Sample Plan EPE GRP 1000029 Reference BFR-S.1, " Response to Nuclear Power Generation /ATWS", step 2 RNO, page 3 COMMENTS stem and distractors c and d revised per Max Bailey
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HIST B Used no SRO Q# 67 DIFF M Answer BFR-S.1. "Resoonse to Nuclear Power Generation /ATWS". requires that pressure be verifed to be less than 2335 PSIG and if not reduce pressure to less than 2135. The reason for this action is to reduce l pressure to ... (Select ONE of the following)
i c. ... maximiie boration flo ,
b. ... limit cycling of PZR PORVs.
l c. ... avoid opening of the PZR Safety Valve ,
i d. ... limit the amount of injection required from charging.
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K-A # 000029EK3.12 4.4 \ 4.7 Sample Plan EPE GRP 1000029 R;ference BFR-S.1," Response to Nuclear Power Generation /ATWS" COMMENTS stem and all selections changed for formatting, revised distractor c per Max Bailey
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HIST N Used l
. SRO Q# 68 DIFF M Answer c.
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I Given the following conditions:
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- The plant was operating normally at 100% power l l - S/G A r{ arrow range level rapidly decreases to 2%
- - S/Gs B, C; and D narrow range levels are at their normal operating level ,
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-The reactor and turbine do NOT trip !
- Auxiliary Feedwater pumps do NOT start
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- An Anticipated Transient Without Trip [ATWS] condition is announced
- Control rods are manually inserted
- Power Range instrumentation is decreasing at 10% per minute due to rod insertion i
Which ONE of the following is the expected response of the ATWS Mitigating System (AMS)? l
)
c. The AMS system will automatically trip the reactor which then causes a turbine tri ;
b. The AMS system will trip the turbine and automatically start all AFW pumps, j l
c. The AMS system will not actuate since the required S/G low level logic has not been satisfie d. The AMS system is blocked from actuation since power level will be less than 30% power before the AMS time delay expire ,
K-A # 000029EA2.09 4.4 \ Sample Plan EPE GRP 1000029 Reference Figure 60c-1, ATWS Mitigation System logic COMMENTS replacement question for 68, need approval by Max Bailey spelling correction ok per Max Bailey 10/10/96
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HIST N Used
- SRO Q# 69 DIFF C Answer !
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- With the plant operating at 100% power and all controls in AUTO, which ONE of the following
- statements is correct regarding a main feedwater break as compared to a steamline break at power?
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c. A feedline break can be identified by the rapid depressurization of the affected S'/G prior to the Reactor Trip.
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! b. A cteamline break will only blow down one S/G, a feedline break will :
blowdown ALL S/Gs until FW isolation occur i i
1' c. The initial primary response to a feedline break is an INCREASE in T-ave; for a steam break, T-ave continuously DECREASES.
- d. The initial' secondary response to a feedline break will be an INCREASE l
- in main generator load; for a steam break, load will remain the sam ;
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i K-A # 000040EA2.01 4.2 \ Sample Plan EPE GRP 1000040
, Ref:rence System Ip chapt. 23, " Main Steam", p. 80 i
l COMMENTS revised distractor a and b per Max Bailey l
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t HIST N Used 3 ERO Q# 70 DIFF C Answer c.
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Given the following:
l - Unit 1 is operating at 100% power.
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- RCP Nq.1, SEAL LEAKOFF FLOW HIGH alarm is receive No. 2 sealleakoff high flow alarm has been PRINTED.
.; - RCP No.1 seal leakoff recorder indication is pegged HIGH.
j - Make-up to the RCS has increased 40 gpm to maintain pressurizer level.
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! Which ONE of the following has occurred?
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..c. The No.1 and No. 2 seals have failed and the RCS pressure drop is
- across the No. 3 sea b. The No. 2 seal has failed and is allowing water from the standpipe i to flow out the No.1 seal leakoff line.
i c. The No.1 seal has failed and the RCS pressure drop is across the j No. 2 sea d. The No. 2 and No. 3 seals have failed and the RCS pressure drop is f ccross the No.1 sea t i l
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l K-A # 000015EA1.22 4.0 \ Sample Plan EPE GRP 1000015 R:ference ILT Simulator Lesson Plan BOA RCP-1, " Reactor Coolant Pump Seal Failure", p. 9 COMMENTS deleted 6 gpm from stem per Max Bailey
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, HIST N Used
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SRO Q# 71 DIFF C Answer d.
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- Given the following
- - 1BCA-2.1. " Uncontrolled Deora==uriration of All Stamm Generators". has i been en,tered.
! - S/Gs 1 A;'1C, and 1D narrow range levels are 30%.
. - S/G 1B narrow range level is 10%.
, - RCS wide range cold leg temperature has decreased 130 deg F in the last I i hou )
- Containment pressure is 14 psi j
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- Which ONE of the following actions should be taken for the given conditions?
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c. Stop feed flow to all S/Gs until RCS cooldown rate is less than 100
- degrees F/ hour.
! b. Reduce feed flow to S/Gs A, C, and D to 25 gpm each until RCS
- cooldown rate is less than 100 degrees F/ hour.
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- c. Stop feed flow to S/Gs A, C, and D until cooldown rate is less than 100 degrees F/ hour.
- d. Reduce feed flow to each S/G to 25 gpm until cooldown rate i
! is less than 100 degrees F/ hou )
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K-A # 000040G.12 3.8 \ Sample Plan EPE GRP 1000040 l.
i Reference BCA-2.1, " Uncontrolled Depressurization of All Steam Generators", step 2.a. RNO
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COMMENTS no change
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HIST B Used no SRO Q# 72 DIFF M Answer b.
- Per the imrnediate actions of BCA-0.0. " Loss of AH AC Power" . the RCS is isolated by first verifying the PZR PORVs are closed. Which ONE of the following describes the subsequent verification sequences?
c. 2. Letdow'n line isol valves 3. Letdown orifice isol valves 4. Excess letdown isol valves b. 2. Letdown orifice isol valves I 3. Letdown line isol valves !
4. Excess letdown isol valves c. 2. Excess' letdown isol valves 3. Letdown line isol valves 4. Letdown orifice isol valves d. 2. Excess letdown isol valves 3. Letdown orifice isol valves 4. Letdown line isol valves
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K-A # 000055G.10 4.1 \ Sample Plan EPE GRP 1000055 Reference BCA-0.0," Loss of All AC Power", step 4 COMMENTS formatting to place PORV is stem revised per Max Bailey . _ . _ . . _ . _ _ . . - _ _ _ . _ _ . _ _ . . _ _ _ _ _ _ . . . . _ . _ -
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HIST B Ustd no ;
- SRO Q# 73 DIFF M Answer I Units 1 and 2 have just experienced a Loss of All AC power. Step 6 of BCA-0.0. " Loss of All AC Power" , requires that the Safeguards Loads be isolated from their respective buses prior to reestablishing power to the buse . Which ONE o' f the following describes the reason for defeating the automatic loading of the Emergency buses?
c. To prevent potential overload of the Emergency buses when they are re-energize b. To prevent damage to the Reactor Coolant Pump seals upon restoration of powe c. To prevent ESF pump starts without proper cooling water and auxiliary support equipment availabl d. To prevent RWST inventory depletion of water that may be needed for long term cooldow I i
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K-A # 000055EK3.02 4.3 \ Sample Plan EPE GRP 1000055 R:ference BCA-0.0," Loss of All AC Power", step 6 COMMENTS revised distractor c per Max Bailey
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HIST N Ussd l SRO Q# 74 DIFF Answer a.
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l Given the following conditions on Unit 1:
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- Si has actuated due to a large break LOCA.
, - RWST levelis 45% and DECREASING.
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-Train A'CN'MT recirc sump suction valve,1Sl8811A did NOT OPEN automatically.
I Which ONE of the following actions require that flow through 1A RH pump be temporarily stopped to l complete the switchover to Cold Leg Recirculation Mode?
c. RWST to A RH pump suction valve,1Sl8812A must be closed before 1Sl8811A can be opened manually, b.1 A RH pu'mp discharge to charging pumps,1CV8804A must be closed if 1Sl8811 A remains close c. RWST to 1 A containment spray pump suction valve,1CS001A must be closed before 1Sl8811 A can be manually opene d.1 A RH pump discharge crosstie valve,1RH8716 must be closed if 1Sl8811 A remains close K-A # 000011EA1.13 4.1 \ Sample Plan EPE GRP 1000011 R:ference 1BGP 100-1A2 l
COMMENTS replacement question for 74, need approval from Max Bailey l ok per Max Bailey 10/10/96
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HIST N Us:d
! SRO Q# 75 DIFF c Answer d.
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l A loss of instrument Bus 114 has occurred. No plant transients are in progress. Which ONE of the l following will require manual operator action to control?
l l a. Pressu,rizer Level, i
b. Pressurizer Pressur c. Main Feed Pump speed.
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l d. Volume Control Tank Leve j i
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I K-A # 000057EA1.06 3.5 \ 3.5 Sample Plan EPE GRP 1000057 Reference BOA ELECT-2, " Loss of Instrument Bus", Table D, " Loss of instrument Bus 114 Effects" item COMMENTS suggest adding load rejection from 90% and replacing MFP speed control with PORV,
- potential conflict with CERT exam see attached
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ok per Max Bailey 10/10/96 i
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HIST N Used SRO Q# 76 DIFF c Answer a.
Unit 1 control room was evacuated and BOA PRI-5. " Control Room inaccessibilitv". has been implemented due to a fire in the main control board. No other fires exist in the plant. As it was l evacuated; the reactor was verified tripped and an Sl was in progress. It is determined the Si can be I terminated apd,the crew has entered BOA PRl-5. Attachment F. "Sourious SI Termination". Which ;
ONE of the foll6 wing describes how the SI is terminated from this condition?
c. Use jumpers in the aux electric equipment room; then stop safeguards equipmen b. Use jumpers in the cabinet at the remote shutdown panel; then i stop safeguards equipmen c. Use the Si reset switches at the remote shutdown panel; then stop safeguards equipmen d. Stop safeguards equipment by local breaker operation I i
K-A # 000068EA1.12 4.4 \ Sample Plan EPE GRP 1000068 R:ference BOA PRl-5, " Control Room inaccessability", Att E, p. 48 and Att. F, p. 58 COMMENTS stem and distractor d revised per Max Bailey
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4 HIST N Used l SRO Q# 77 DIFF M Answer d.
i l Given the following:
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- - A LOCA has occurred
- - CETCs pre,1300 deg F
- ALL RCPs have been started per BFR-C.1. "Resoonse to inadeounte Core Coolina" l
! Which ONE of the following describes the criteria in BFR-C.1. "Resoonse to inadeounte Core Coolina" j that will direct the operators to stop ALL the RCPs after the procedure had directed starting them?
j o. CETCs indicate less than 1200 deg b. RCS pressure is less than 1370 psig.
i c. PORVs and Reactor Vessel head vents are all open.
l i- d. Two wide range Thot instruments indicate less than 350 deg F.
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K-A # 000074.EK3.04 3.9 \ Sample Plan EPE GRP 1000074 Reference BFR-C series lp 20 and 22; BFR-C.1 steps 16 - 19 COMMENTS replacement question, needs approval from Max Bailey ok per Max Bailey 10/10/96
. . HIST BM UsSd E-96-01-02 SA SRO Q# 78 DIFF M Answer ,
What adverse consequence could result from delaying feed and bleed cooling if the conditions to initiate feed and bleed are met in BFR-H.1. "Resoonse to Loss of Secondary Heat Sink"? (Select ONE of the following)
c. High te'mperature induced failure of S/G U-tube bend b. An overpressurization challenge to the Reactor vesse c. In:bility to provide sufficient injection for core cooling due t3 high RCS pressur d. Innbility to recover the S/Gs without damage from high thermal stresse l l
K-A # 000074EK3.11 4.0 \ Sample Plan EPE GRP 1000074 Reference ILT Simulator Lesson Plan BFR H Series, pp. 8,17 i
i COMMENTS made into multiple choice ok, no change I
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HIST s Used no t
- SRO Q# 79 DIFF M Answer !
i i I While performing 2BFR-C.1. "Resoonse to inadeaunta Core Coolina". which ONE of the following
! describes the basis for directing RCP restart after they have been previously tripped?
. i c. Allows,for, restoration of pressurizer pressure control using normal j spray valve l i b. Once subcooling is established; restarting the RCPs helps to collapse i i voids that may have formed in the reactor vessel hea l
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- c. Provides for cooling of the core when secondary depressurization has i
. not alleviated inadequate core cooling conditions.
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d. Provides mixing of the Si flow to protect the reactor vessel from cold )
4 water.
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1 K-A # 000074EK3.07 4.0 \ Sample Plan EPE GRP 1000074 ,
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R:;ference ILT Simulator Lesson Plan BFR C Series, p. 20 ,
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COMMENTS no change
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l HIST N Used SRO Q# 80 DIFF c Answer ,
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Given the following conditions: l
- Unit 1 in mode 5
- "A" train RH in shutdown cooling mode ;
- RCS temperature being maintained at 180 deg. F l
Which ONE of the following would indicate a tube leak in the "A" RH heat exchanger? l o. INCREASING CC surge tank leve b. INCREASING Pressurizer level.
l c. DECREASING RH retum flow to the RC '
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d. DECREASING RH system boron concentratio !
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l t l l 4 K-A # 000026EA2.01 2.9 \ Sample Plan EPE GRP 1000026 !
R:ference System Ip chapt.19, " Component Cooling Water", pp. 52, 54 COMMENTS ok, no change
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HIST M Used Bwrd NRC 96 tt SRO Q# 81 DIFF C Answer c.
' A Unit 1 startup is in progress. Unit 1 reached 35% power at 0500, August 13. At 0800, August 13, Chemistry reports the following sample results:
- Dose Equivalent 1-131..... 50 pci/ gram
- Gross Activity .................. 25 ci/ gram
- E-ba r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.75
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At 0800, August 15, Chemistry reports that sample results indicate Dose Equivalent 1-131 has decreased and stabilized at 0.83 pci/ gram. Using the attached LCO 3.4.8 "Soecific Activity" and current plant conditions, which ONE of the following applies?
l c. Plant may. remain at this power level indefinitel q b. Perform isotopic analysis for lodine once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> c. Be in HOT STANDBY with Tave < 500 F by 1400, August 1 l
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d. By 0900, August 15, initiate action to place the unit in HOT STANDBY ]
by 1500, August 1 '
K-A # 000076G.03 2.6 \ 3.5 Sample Plan EPE GRP 1000076 Reference Tech Spec 3.4.8, pp. 3/4 4-27 thru 29
- COMMENTS Supply the LCO and chart stem, answer and 2 distractors revised per Max Bailey
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HIST BM Used no SRO Q# 82 DIFF M Answer b.
.
! Which ONE of'the following is the basis for the time delay between a turbine trip and a generator trip?
c. RCP overspeed can occur during a major LOCA resulting in a high enough RCS flowrate to damage the reactor vessel intemal .
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b. The time delay reduces the impact on core heat removal by ensuring forced flow through the core for a short period of time immediately following the reactor trip.
j i c. The thermal stresses imposed on the RCP shafts are far greater if the l RCP's are stopped immediately (due to high RCS delta t) after turbine I
- tri , j d. The time delay allows circulating transformer currents induced by the
, trip transient to stabilize following the plant trip; reducing the probability of damage to the RCP motor l K-A # 000007G.07 3.6 \ Sample Plan EPE GRP 2 000007 R;farence System Ip chapt. 5, " Main Generator", p. 46; System Ip chapt.13, "Reator Coolant Pump", pp. 62, 63 COMMENTS converted to multiple choice look at improving distractors and answer ok per Max Bailey 10/10/96
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. HIST N Used SRO Q# 83 DIFF C Answer l
' Given the following plant conditions:
- - Reactor power is STABLE at 90%
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- Pressurizer levelis DECREASIN VCT lev'elii DECREASING.
- . - The following annunciators are actuated
) - PZR LEVEL CONT DEV LOW j - LTDWN TEMP HIGH
- REGEN HX LTDWN TEMP HIGH l
{ Which ONE of the following events would cause these indications? j e. Small-break LOCA on the loop B Hot Leg.
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b. Isolation of CVCS letdown.
l c. Charging header rupture, i
d. Letdown header ruptur !
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- K-A # 000022EA2.01 3.2\ 3.8 Sample Plan EPE GRP 2 000022
Reference BOA PRI-1, " Excessive Primary Plant Leakage", p.1; BAR 1-9-E2, "LtdwnTemp High",
BAR 1-9-A1," Regen Hx Ltdwn Temp High" i i
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COMMENTS stem revised per Max Bailey
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l HIST B Used no SRO Q# 84 DIFF M Answer A reactor trip and Si have occurred, and the control room operators are responding to a small-break
. LOCA. All RCP's are tripped. The operators have proceeded to the recovery stage in BEP ES-1.2.
j " Post-LOCA Cooldown and Deoressurization". A PZR PORV is used to depressurize the RCS until PZR level is greater than 21% (50% for adverse containment).
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In addition to ensuring that RCS conditions are under adequate operator control, the basis for this pressurizer level ensures ... (Select ONE of the following)
O. ... sufficient inventory such that PZR level does not drop low l when an RCP is starte b. ... that a reduction in subcooling does not occur when Si flow is reduce c. ... that starting an RCP will not cause letdown isolation.
! d. ... adequate PZR steam space to absorb pressure fluctuations during RCP start.
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K-A # 000009EK3.21 4.2 \ Sample Plan EPE GRP 2 000009 Reference ILT Simulator Lesson Plan BEP Series, pp.156,157 l
COMMENTS revised answer and distractor b per Max Bailey i !
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HIST su Used E-96-01-01 l SRO Q# as DIFF M Answer )
Given the following potential Unit 1 plant conditions:
1. Core exit thermocouples: decreasing 2. Core exit thermocouples: stable or increasing 3. RCS hbt leg temperature: stable or decreasing 4. RCS hot leg temperature: increasing l 5. RCS subcooling: less than 25 degrees F l 6. RCS subcooling: greater than 25 degrees F l 7. RCS cold leg temperature: at saturation for S/G pressure j 8. RCS hot leg temperature: at saturation for S/G pressure if ALL RCPs were tripped due to a small break LOCA, which ONE of the following combinations of the -
above indicatiohs would verify the existence of natural circulation?
c.1,4,5,7 b.2,4,6,8 l
l c.1,3,6,7 d.2,3,5,8 l
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K-A # 000009EA2.37 4.2 \ Sample Plan EPE GRP 2 000009 Reference BEP ES-1.2, " Post LOCA Cooldown and Depressurization", Att. B, p. 33 COMMENTS stem revised per Max Bailey ;
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HIST BM Used no SRO Q# 86 DIFF A Answer The following plant conditions exist fcr Unit 1:
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- Reactor Trip and Safety injection have occurred
- MSIVs have Just closed due to Containment pressure
- RCS pressure is 1700 psig and stable
- CETCs indicate 570*F
- ALL S/G Narrow Range levels are 40%
- PZR levelis 42%
- Figure 1BEP 1-1 is attached
' Based upon these conditions, the operators should ... (Select ONE of the following)
c. ... verify all RCPs are stoppe b. ... terminate Safety injectio c. ... maintain ECCS fiow until Si termination criteria are reache d. ... Initiate Containment Spray as a result of increasing containment pressur l
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K-A # 000009EA2.34 3.6 \ 4.2 Sample Plan EPE GRP 2 000009 l l
Reference BEP-1, " Loss of Reactor or Secondary Coolant", step 6, pp. 7,24 ,25 COMMENTS Figure 1BEP-1.1 is attached ok, no change
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HIST N Used l SRO Q# 87
DIFF M Answer d.
l The CAUTION before Step 1 of BEP ES-3.1. " Post SGTR Cooldown Using Backfill", wams that the i RCP in the ruptured loop should NOT be started first. Which ONE of the following explains the reason
- for th'a caution?
Starting thisRCP could cause:
l c. A slug of cold water to pass through the core and cause a retum t3 criticality.
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b. An internal pressure surge, resulting in further failure of damaged tube c. A rapid cooldown of the reactor vessel, and present a challenge to the Integrity critical safety functio d. A clug of unborated water to pass through the core and cause a return to criticalit .
l K-A # 000038EK3.06 4.2 \ Sample Plan EPE GRP 2 000038
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Ref rence ILT Simulator Lesson Plan BEP-3 Series, BEP ES-3.1, " Post-SGTR Cooldown Using Backfill", p. 2 Caution COMMENTS ok, no change l
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HIST BM Us@d 92 NRC st SRO Q# 88 DIFF C Answer The following plant conditions exist:
- Unit 2 entered Mode 3 on 8/16/95 at 0500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> A cooldown to Mode 5 with the RCG at 130-140*F was completed on 8/18/95 at 1700 ho0r At 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on 8/21/95 a total loss of RH cooling occurre The following pertinent plant conditions existed prior to the loss of RH cooling:
- RCS Tavg ....... 135'F
- RCS level ... ..... mid-loop Which ONE of the following represents the MINIMUM time it will take the RCS to reach saturation?
(Fig. 2 BOA PRl-10-1. " Loss Of RH Cooling" may be used).
O 7 minutes b. 9 minutes c.11 minutes d.13 minutes K-A # 000025G.12 3.3 \ Sample Plan EPE GRP 2 000025 Reference BOA PRI-10, " Loss of RH Ccoling", figure BOA PRI-10-1, p. 9 COMMENTS QUESTION REQUIRES FIGURE 2 BOA PRI-10-1 ATTACHE selection formatting revised per Max Bailey
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HIST N Used :
SRO Q# 89 DIFF M Answer If PT-455 fails, BOA INST-2 Attachment B. "Pra==urizer Pressure Channel Failure" directs the operator to refer to ath ADDITIONAL Technical Specification; bey'ond the ones referred to if any of the other three channels has failed. The Technical Specification that applies to only one of the PZR pressure channels is .;. (Select ONE of the following)
,
e. ... LCO 3.4.4, Relief Valve b. ... LCO 3.4.9.3, Overpressure Protection System c. ... LCO 3.3.3.5, Remote Shutdown Instrumentation.
I d. ... LCO 3.3.3.6, Accident Monitoring Instrumentation.
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K-A # 000027G.03 3.1 \ Sample Plan EPE GRP 2 000027 Reference Tech Spec 3.3.3 3 ; ages 3/4 3-50 to 3-52; BOA IN3T-2, " Operation with a Failed instrument Channer', Attachment B step 7, page 13 l
l COMMENTS added LCO titles per Max Bailey, revised 1 distractor and psychometrically rearranged i changed answer to c from a l
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HIST B Ussd no
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SRO Q# 90 DIFF . M Answer d.
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l The following conditions exist:
- U-1 Component Cooling Water Heat Exchanger is aligned to U- .
- U-0, and U-2 Component Cooling Water Heat Exchangers are aligned to U- U-0 Cornponent Cooling Water Heat Exchanger Outlet Radiation Monitor is j in the INTERLOCK condition due to exceeding the ALARM setpoint.
l Which ONE of the following automatic actions occur in addition to receiving an audible ALARM on the l RM-11?
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. c. The U-1 surge tank vent valve remains OPEN, the U-2 surge tank vent v lve remains OPEN.
- b. The U-1 surge tank vent valve remains OPEN, the U-2 surge tank vent
.v Ne CLOSE c. The U-1 surge tank vent valve CLOSES, the U-2 surge tank vent valve
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remains OPEN.
l d. The U-1 surge tank vent valve CLOSES, the U-2 surge tank vent valve i
- CLOSES.
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l K-A # 000060EK2.02 2.7 \ Sample Plan EPE GRP 2 000060 R ference BAR RM11-1-0PR09J COMMENTS replacement for question 90
replaced "1" and "O" in U-1 and U-0, ok per Max Bailey 10/10/96 t
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i HIST N Used SRO Q# 91 DIFF c Answer b.
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i i Unit 1 C3 evaluating a Steam Generator tube leak with the following plant parameters:
- Letdown flow is at 75 gpm.
l - One (1) charging pump is runnin Pressurizer levelis STABLE.
l - Seal injection and leakoff flows are NORMAL.
i - Charging flow is 110 gp Preexisting RCS leakage was identified as 0.8 gpm to RCD LCV-112A is in the AUTO positio Which ONE of the following is the approximate amount of primary to secondary leakage?
a.10 gp b. 22 gp c. 34 gp d. 46 gp I
!
I K-A # 000037EA2.12 3.3 \ Sample Plan EPE GRP 2 000037 Reference BOA SEC-8, " Steam Generator Tube Leak", step 4, p. 3; System Ip chapt.15A,
"CVCS", p. 56 COMMENTS 110 - 87 - 0.8 = 22.2 gpm ;
ok, no change ;
[ l
_ - . _ _ .
HIST N Us:d SRO Q# 92 DIFF C Answer c.
A SGTR coincident with an RCS LOCA has occurred on Unit 1. The crew has implemented BCA- ,
'SGTR with Loss of Reactor Coolant. Subcooled Recoverv Desired". The following plant conditions '
cre reported while evaluating step 12, " Check if Subcooled Recovery is Appropriate":
- RWST Leviel - 50%
- Ruptured S/G Level - 51%
- Containment Pressure - 1.8 psig
- Containment Floor Water Level - 4.5 inches
- RVLIS is UNAVAILABLE
)
Using the attached Figure BCA-3.1-3, determine the correct procedure to continue the recovery, and l the reason for u, sing that procedure. (Select ONE of the following) l i
o. BCA-3.1 since RVLIS is needed if a Saturated Recovery is performe i b. BCA-3.1 since Ruptured S/G overfill is NOT imminen c. BCA-3.2 since this procedure conserves makeup wate ;
d. BCA-3.2 since RCS pressure, temperature, and level are easier to i contro !
>
K-A # 000038G.12 3.8 \ Sample Plan EPE GRP 2 000038 R;ference BCA-3.1, "SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired", Figure
__BCA 3.1-3, " Containment Floor Water Level vs RWST Level", BCA-3.2 Bases COMMENTS Supply Figure CA-3.1-3 revised CNMT floor water level per Max Bailey 10/10/96
. - . -
!
HIST BM Used no SRO Q# 93 DIFF c Answer c.
The following plant conditions exist on Unit.1:
- PZR pressure control is in automatic with the control selector switch in the 457/458 position
- PZR pressure transmitter PT-457 has failed full scale HIGH.-
Which ONE of the following describes the plant response assuming NO operator action is taken?
c. RY-455A opens and cycles around 2185 psi b. RY-456 opens and cycles around 2185 psi c. RY-455A opens and closes when RCS pressure decreases below 2185 psig cnd a Rx Trip / SI will subsequently occur on Low PZR pressur d. RY-456 opens and closes when RCS pressure decreases below 2185 psig and e Rx Trip / SI will subsequently occur on Low PZR pressur l
- K-A # 000027EK2.03 2.6 \ Sample Plan EPE GRP 2 000027 Reference System Ip chapt.14, " Pressurizer", pp. 70 - 72 COMMENTS distractor a and b revised per Max Bailey
-. -. . _ . . __ - - . . - . _ _ . , - - .. - ._. .
,
l HIST N Us:d j
SRO Q# 94 DIFF A Answer ,
The operators have just entered BFR-H.1. " Response to Loss of Secondarv Heat Sink". with the following parameters:
l Reactor Trip Breakers ................. Open Reactor power ........................ .. 0%
i Wuje Range S/G Levels .............. A B C D l 60 % 58% 50 % 45 %
Containment Pressure ............... 7.2 psig l CETC (average of ten highest) .. 655 deg F Aux Feed flow ............... ............. O gpm Startup FW pump .................. .... Available Safety injection ........................... Actuated l RCPs .......................................... None running
- Which ONE of the following is the next required action for these conditions?
o. Initiate Bleed and Fee b. Restore Aux Feed flow to all Steam Generator c. Establish normal Feedwater flow using the Startup FW pum d. Prepare to feed one Steam Generator from the Essential Service Water syste K-A # 000054G.12 3.2 \ Sample Plan EPE GRP 2 000054 Reference 1BST-3, " Heat Sink"; 1BFR H.1, " Response to Loss of Secondary Heat Sink", Caution prior to step 2, p. 3 COMMENTS answer corrected / revised answer per Max Bailey revised stem per Max Bailey 10/10/96
-- ._ . .. -. . .- -.
HIST N Used SRO Q# 95 DIFF M Answer d.
i Which ONE of the following represents the maximum activity allowed in a gas decay tank, and the
, b: sis for this maximum?
.
c. 5,000 curies, to ensure offsite dose will be limited to a small fraction of the 10CFR100 limits in the event of an inadvertent release of the tank contents.
,
b. 5,000 curies, to limit site boundary dose to 0.5 rem in the event of an inadvertent release of the tank contents.
'
c. 50,000 curies, to ensure offsite dose will be limited to a small fraction of,the 10CFR100 limits in the event of an inadvertent release
>
of the tank content d. 50,000 curies, to limit site boundary dose to 0.5 rem in the event of cn inadvertent release of the tank content K-A # 000060G.04 2.4 \ Sample Plan EPE GRP 2 000060 R;ference System Ip chapt. 46, " Gaseous Rad Waste", p. 4; Tech Spec Bases 3/4.11.2.6, p. B 3/4 11-2 COMMENTS ok, no change
i HIST N Used l
SRO Q# 96 DIFF c Answer !
l l
Which ONE of the following events would result in a containment AREA radiation monitor alarm?
o. RCP #1 seal failur i b. PZR Shfety Valve Seat Leakag c. Steam Generator Tube Ruptur l d. RCS leak at the incore Seal Tabl !
.
K-A # 000061EA1.01 3.6 \ Sample Plan EPE GRP 2 000061 R;ference BEP-0, " Reactor Trip or Safety injection", step 30, p. 21; BAR RM-11-4-1 AR14J COMMENTS distractor a and c revised per Max Bailey
_ . _ _ . _ _.._.__.____ _ __._ .__._~.__._ _ ~.. ._ -
HIST N Us::d
SRO Q# 97 DIFF A Answer a.
i ,
f Given the following Unit 2 plant conditions:
i l -Train B RH is in shutdown cooling mode following a refuelin RCS temperature is 300*F.
- - RCS pr4ssure is 340 psig.
1_ -2B CV pump is in operation.
l- - PZR bubble has been established j - A loss of instrument air just occurred.
, Which ONE of the following describes what will initially occur and why if NO operator actions are
- taken?
I
i e
e. An RCS cooldown due to INCREASED RH flow through the RH H I i .
!
b. PZR level decrease due to INCREASED RH letdown flow.
- c. 2B CV pump will lose suction due to INCREASED charging flo I 1 l j d. An RCS heatup due to DECREASED RH flow through the RH Hx
,
!
!
- i
'
!
I i i
l
,
,
i
- *
.
- K-A # 000065EA2.08 2.9 \ 3.3 Sample Plan EPE GRP 2 000065
- Reference System Ip chapt.18, " Residual Heat Removal", pp. 24,26
't i COMMENTS formatting CAPS and distractor d revised per Max Bailey
.
. - , --. - ._ .
-. _
HIST BM Used 94 NRC
, SRO Q# 98 DIFF C Answer d.
< You cre the SRO in CNMT for Fuel Handling operations with the following conditions:
.
- MODE 6 I - Fuel handling in progress in CNMT
- A fuel assehibly is dropped during removal from the core l - Bubbling is observed from the dropped assembly
' Which ONE of the following describes your FIRST required action for this event?
.
) c. Direct operations to Establish Containment Closur b. Direct the Control Room to Start CNMT Charcoal Filter units.
'
c. Direct the Fuel Handlers to place any fuel assembly in the transfer d vice into the change fixtur d. Direct ALL unnecessary personnel to evacuate CNMT.
,
a r
i
.
i K-A # 000036EK3.03 3.7 \ Sample Plan EPE GRP 3 000036
,
! R ference BOA REFUEL-1," Fuel Handling Emergency", pp.1,2 -Entry Conditions and Step 1 l
,
!
COMMENTS Modified stem conditions and 2 dists, same answer stem and selections revised per Max Bailey
!
i
.i
HIST N Ussd SRO Q# 99 DIFF A Answer a.
If pressurizer level transmitter LT-459 is selected as the controlling channel when the reference leg for LT-459 develops a slow leak; which ONE of the following correctly describes anticipated instrument or plant response?
L1-459 LI-460 LR-112 (Pzr level) (Pzr level) (VCT level) INCREASING DECREASING INCREASING j DECREASING INCREASING INCREASING INCREASING DECREASING DECREASING DECREASING INCREASING DECREASING l
l i
l K-A # 000028EK1.01 2.8\ Sample Plan EPE GRP 3 000028 R:farence System Ip chapt.14, " Pressurizer", pp. 28,30,32; Ip Appendix B review Questions, 38.c, pp.106,114 COMMENTS ok, revised format per Max Bailey
. . _ _ _ _ _ _ _ . _ . _ _ _ . _ _ _ _ . _ . _ ..... _._ ___ _ _._ _ _
$
HIST BR Used E-96-01-01 I SRO Q# 100 DIFF C Answer d.
i
}
! A natural circulation cooldown is in progress per BEP ES-0.2. " Natural Circulation Cooldown" The
' pl:nt is being depressurized using auxiliary spray.- As pressure decreases through 1300 psig, a rapid increase in pressurizer level is observed. Charging and letdown are in manual and are equa Which ONE 6f the following describes the expected operator actions?
I c. Isolate the ECCS accumulator b. Increase the cooldown rate to 50*F/hr.
. c. Reduce Charging flow, and place additionalletdown orifice (s) in i servic '
) d. Close the auxiliary spray valve, and energize PZR B/U heaters to i repressurize the RC I
,
i
,
K-A # 000056EK3.02 4.4 \ Sample Plan EPE GRP 3 000056 R ference BEP ES-0.2, " Natural Circulation Cooldown", step 14 RNO, p.10 COMMENTS stem, answer and distractor d revised per Max Bailey, need review of distractor c ok per Max Bailey 10/10/96
.. - . ,
_ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - . . _ ._ _ _._ _ . .
i i
!
i REV. 1A LOSS OF EMERGENCY COOLANT RECIRCULATION 1BCA- WDG-1B UNIT 1 ,
i ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l STEP l
/
l 15 :
. TE G4INATE J:
I
'
- Establish minimum ECCS flow
' to remove decay heat by
- RCS subcooling - ACCEPTABLE )
PER ATTACHMENT A (Pace 34). AND performing the following: i
! FIGURE 1BCA 1.1-4 (Pace 33)
! (+50*F) a. Deteraire minimum ECCS
'
flow required from FIGURE i * Check Reactor Vessel inventory: 1BCA 1.1-1 (Page 30).
!
' o With no RCP running: k>.. Manually operate ECCS j
equipment to reduce ECCS flow to minimu ,
i * RVLIS plenum region - 113, i OR GREATER
! c. GO TO Step 19 (Page 15).
o With any RCP running:
Average of ten highest
l core exit TCs - DECREASING
!
!
i
!
!. '
l
.
FOR REFERENCE l
,
\
l
!
i
!
!
!
!
i
!
i
\
'
APPROVED JUN 201996 i
' B.O.S.R.
i I Page 12 of 35
.
-
- ._ . _ - _ _ _ - _ _ . -_ _ - . . - - ___ . - . - . _ - - . _ _ - _ . _ ..
! REV. lA LOSS OF EMERGENCY COOLANT RECIRCULATION IBCA-1 1 i~ WOG-1B UNIT 1 FLOWRATE(GPM)
600(,
i
%
'
l 550 , i .
'l 500 1 hr 2 hr 4 hr 8 hr 24 hr
- -
'.
450
,
i , ,
,
,
n o
n n
.
.
,
. i li 'l i i 400 \ %
u u ,
'
o
h 'I ni l
-,' t,
'l
-, , .e ,
'
350 ,
,
n , i il
, i
- ( ' ll ll B
'
r
' l !
" l 300 ! , l ! E
'
'i II i , !i i i!
' i
" '
250
' '"
Q^ i
'
-
,
,, ! m ,, x .
" " '
200
'
a u i
,
-
..
is l
-
i, i i
'
l u a
-
u r
' '
150 .. m m g
' ,,_
'"
l ll ll 4 RIhm
_
j '"4m j l} [ E E "'lkm_
'
100- t u u ,
.
l, .. ..
..
~
ll ll ll l l 50 i ii u i l i
i n u a r
- ,
0-
. . . . . . . .
'
100
"
.'. . '.....
1000
"'
. . . . . . . .
10000 TIME (MINUTES)
APPROVED l FOR RBERENCE auna =
FIGURE IBCA 1.1-1 B.O.S.R.
'
REQUIRED ECCS FLOW VS TIME FROM TRIP l
Page 30 of 35
. _ _ _ _ _ _ _ _ . . . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _.
!
i
)
!
!
, REV. lA LOSS OF EMERGENCY COOLANT RECIRCULATION 1BCA-1.1
WDG-18 UNIT 1
.
!
I e
- .*
( +50) j; , '
.
, , ,
26M
- .r I r J r I
- i s r F A J l' I I i
)
I .r .V r
,
i i i i 2200 '
ACCEPTABLE [ r f l
,
r r a
- i f I I B r J v v n 4' I E i r a
, , ,
!
1800 r n i
! I a v i ' '
E F j 1600 F
,
.. J l'
A W I ' r 3
! .,
1400 a r j W E .V
- E V A
' ' #
- 1200- J F J f .V i J i e ;r i NOT t:
>
2000 j
'
/ j' ACCEPTABLE [-
l e a r
,
a r r .r
'# #
!
.
8H > r F
>
' A'2V A
.
s' r >r Containment
!
l 600- Conditions: , ', '
,
'
Adverse- - "
i
'
400 *?) ,
lP f 4 -Saturation
200 j i
!
! 0 450 500 550 600 650 700 200 250 300 350 400
!
RCS Temperature (*F)
APPROVED
,
.
FIGURE 1BCA 1.1-4 JUN 20 996
! .
e RCS SUBC00 LING MARGIN +50*F
.
B.O.S.R.
1
'
Page 33 of 35
. _ . _ . . , _ - __
-. . . - . - - .. --- .
. . . . - . . - . - - - . _ . - - - . _ . . . _ . . _ . . - . . - . . . . -
.
f REV. 1A LOSS OF EMERGENCY COOLANT RECIRCULATION 1BCA- WOG-18 UNIT 1
i
!
]
!
" ATTACIDENT A :
- Determination of Post-Accident RCS Subcoolina
1) Check Cnnt for kDVERSE conditions:
i j o Cnnt pressure greater f.han 5 PSI !
- .og-
!
!
- o Cnnt radiation greater than 10 5 R/HR:
!
,
- l
..-
2) Check RCS wide range pressure (1PI-403A/405).
3) Determine temperature for RCS pressure using NORMAL or ADVERSE Cnnt curves per:
o FIGURE referenced in procedure step-OR-o FIGURE 1BCA 1.1-2 for Attachment B 4) Check average of ten highest core exit TCs (1TI-IT001/IT002) .
5) If average of ten highest core exit TCs is less than temperature for RCS pressure, then acceptable subcooling exist FOR REFEREEE
--
JUN 201996
.
B.O. Page 34 of 35
. _ . _ __. ____.__ _ _ ___ _ _ _ . - . _ _ . - - _ ._ __.. . . _ _ . _ _ _ _ _ _ _ _ . _-__
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!
3/4.4.8 SPECIFIC ACTIVITY i
- l j LIMITING CONDITION FOR OPERATION
l
,
i
'
3.4.8 The specific activity of the reactor coolant shall be limited to: i
i Less than or equal to 1 mircoCurie per gram DOSE EQUIVALENT I-131**, I and I Less than or equal to 100/E microcurles per gram of gross radioactirity.
, I
-
APPLICABILITY: MODEL 1, 2, 3, 4, and 5.
- ACTION:
MODES 1, 2 and 3*:
, With the specific activity of the reactor coolant greater than
! I microcurie per gram DOSE EQUIVALENT I-131** for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
during one continuous time interval or exceeding the limit line shown
- on Figure 3.4-1, be in.at least. HOT STANOBY with T , less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and i
' With the specific activity of the reactor coolant greater than 100/l microcuries per gram, be in at least HOT STANDBY with T , less than 500*F within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> . . - . - - - . . . .
_
= o ,. ; ,- ,
fa1%) r; .>c.-,'. L , ,*m ,
r. " <
-
- - '
e i
- With T., greater than or equal to 500* **For Unit 1, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gra l BYRON - UNITS 1 & 2 3/4 4-27 AMENDMENT N0. 77
..
. - . - - . .. _ _ . _ . - _ . _ _ .. - - - .
.
.
i REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION (Continued)
MODES 1, 2, 3, 4, AND 5:
With the specific activity of the reactor coolant greater than _ i 1 microcurie per gram DOSE EQUIVALENT I-131* or greater than 100/E I
'
. microcuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limit .
SURVEILLANCE REOUIREMENTS
'
4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4- j . L l
l
!
- For Unit 1, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gra BYRON - UNITS 1 & 2 3/4 4-28 AMENDMENT NO. 77
. . _ _ . _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ .
,
{ 'L i i :
s
'L i 38 '
s i
!
n i
! 'L
,
,
t
'
,
s
'
d L
'
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t ,. 3 %
'
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300 ', ,
-
ortunou k l'
'
L i
i 4
! L i 1 i l t
\
l 'L
'
t
' ,
150 t
I i l
{ s ;
l L i
!
i i
k k 1 ,
1 1 L
'4 l
't
%
'L f \s i
3,100 Accenma u === opewiou,on r uura\ .s i
i i
i t s
' n s s T
'
'
= , or 7 tur --
s s
! IM i k
n
~ ~ ~
acctetet A
- ~ _
~
OPDtAtcu 't
,___
.uwt i umi- - -
i -
'Jo a e 30 m a 80 to 100
,
. -, -,
i l N ,$h ' ~
MEk b 'h
~
j
"
d}IIIN'Ii)heEdibe..Ih)BP., ..
i t
.
FIGURE 3.4-1
!
1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY j LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTO COOLANTSPECIFICACTIVITY>1; SCI /GRAMDOSEEQUIVALENTI-131*
i
1 j
- *For Unit 1, Reactor Coolant Specific Activity >0.35 pC1/Gran DOSE EQUIVALENT I-131
! BYRON - UNITS I & 2 i 3/4 4-29 AMENDMENT NO. 77
!
<
>. -
. ._, . _ . - _ _
_ . . - . ..-_-_ _ ._. . _ _ _ _ . _ _ . _ _ _ . . _ - . _ . _ . . . _ . _ . .
. _ _ . _ _ _ . _ . _ . _ _ _ _ . . . _ _ _
, .
i TABLi. 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE .
AND ANALYSIS PROGRAM -
'
TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN W51CH SAMPLE i AND ANALYSIS FREQUENCY AND AlBLULL1_RE0 HIRED ,
1 Gross Radioactivity At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I, 2, 3, 4 Determination ** Isotopic Analysis for DOSE EQUIVA- Once per 14 days I LENT I-131 Concentration , Radiochemical for E Deter.nination*** Once per 6 months * 1
' Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, If, 2f, 3#, 4#, 5#
Including I-131, I-133, and I-135 whenever the specific activity exceeds I !
pC1/ gram DOSE EQUIVALENT I-131**** l or 100/E pC1/ gram i of gross radioactivity, and
,
b) One sample between 2 1,2,3 ;
and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following '
a THERMAL POWER change :
exceeding 15 percent ,
of the RATED THERMAL
'
POWER within a 1-hour perio FOR RWERBRE
,
'
BYRON - UNITS 1 & 2 3/4 4-30 AMENDMENT NO. 77 l t
i
._ _ . . _ . . _ . . _ . . . _ ._ _ _ . . _ - ._. .. .. . _ _ _ . _ . .
4 -
!
l TABLE 4.4-4 (Continued)
- TABLE NOTATIONS
.
- Until the specific activity of the Reactor Coolant System is restored
, within its limits.'
'
- Sample to be taken after a minimum of 2 EFPD and 20 days of POWER
- s e0PERATION h:.ve elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or
'
longe !
4 **A gross radioactivity analysis shall consist of the quantitative l
- measurement of the total specific activity of the reactor coolant except l
'
for radionuclides with half-lives less than 10 minutes and all
! radiciodines. The total specific activity shall be the sum of the l l degassed beta-gamma activity and_ the total of all identified gaseous a activities in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and i i , extrapolated back to when the sample was taken. Determination of the' !
contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95 percent confidence level. The latest ,
l available data may be used for pure beta-emitting radionuclide i
'
- A radiochemical analysis for E shall consist of the quantitative i
measurement of the specific activity for each radionuclide, except for
- radionuclides with half-lives less than 10 minutes and all radiciodines,
- which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the determination of E
! for the reactor coolant sample. Determination of the contributors to l E shall be based upon these energy peaks identifiable with a 95 percent j confidence level.
! ****For Unit 1, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gram.
!
,
)
i
'
i l
i
.
)
.
i a BYRON - UNITS 1 & 2 3/4 4-31 AMENOMENT NO. 77
i
. . _ . . _ _ _ . _ _ _ _ . _ _ _ _ . . . _ _ . . _ _ _ . _ . . . _ . . . _ _ . _ - _ . _ _ _ _ _ . _ . . _ . _ _ _ _ .
,
i l REV. 1 LOSS OF REACTOR OR SECONDARY COOLANT -
} WDG-18
-
UNIT 1 18EP-1
!
L
!
.
i (Nomunt)
!
l 2600 ,
, y ,
j .r i r i A F I
! ' J 2400 r J J
'
} A' I I
I .F . F
'
l 2200 ' '
! ACCEPTABLE [ r' f
, f J' J t 2000 s I I j , r , r
-
t r J 1 Y I E j 1800
-
' ' v r J I l J E F
'
i E J J
1600 0 8 8 4 X F F
} .F J J'
- i r I i 1400 ' '
- J r s r r .r l J' F J 1200 F ' J l e .r ,
- .r e r
- e e i NOT
-
'
/ rj
'
l 1000 j ACCEPTABLE
'
i e > r AV W J
!' No r> '
,
> r 1 ',r 2r
' #
l 600 Containment s'? ,
l Conditions: e E e
. s r 1
- Adverse-- -s f
{ 400 Norma g f
- r,e ur
- *'
e ---Saturation
! 200 , s'
!
)
i 0 f
i 200 250 300 350 400 450 500 550 600 650 70d i RCS Temperature (*F)
! APPRC//.3 i FIGURE IBEP 1-1 i RCS SUBC00 LING MARGIN APR 12 E35 f B.O.S.R.
e i
j Page 24 of 25
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...
l REV. 57A LOSS OF RH COOLING UNIT 2 2 BOA
'
PRI-10
,.
-
. . , . . ._
l l
FOR REFERER
., .
0 I TRCS = 100 F .-
TRCS = 140*F 2 . . _ _ _ _ . _ .
.
.-
1 /
5
! A %} 1 f
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i .0 6 .0 10 .0 14 .0 18 .0 Time Af ter Shutdown thes)
l l APPROVED i
l APR 021992
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FIGURE 2 BOA PRI 10-1 B.O.S.R.
i
. Time to Reach Saturation Following a l Loss of RHR at Mid-Loop I
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l Page 9 of 61
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(0290E/0018E/031492) -
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_ . _ _ - _ . _ . . _. _ _ _.. _.. _ . _.._.. _ _ _ _ _ ._ - _ .__. __.. _ __ _ . _ _ _ _ . _ _ _ _ . -
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REV. 1 SGTR WITH LOSS OF REACTOR COOLANT - -
WOG-18 SU8C00 LED RECOVERY DESIRED 18CA- ; UNIT 1
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' Indicated RWST Level (%) [LT930-933]
100, s
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. k s l . s s i i Expected s 60 ' Region '
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BCA- 't Subcooled s Saturated Recovery
\s Recovery Appropriate \'
s Appropriate '
s
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0 5 10 15 20 25 30 Indicated Containment Water Level (inches) [LI-PC006/007]
APpeoy, FIGURE 1BCA 3.1-3 APR 12 D-CONTAINMENT FLOOR WATER LEVEL VS RWST LEVEL B.O. Page 39 of 41
.