ML20247K512

Requalification Exam Rept 50-456/OL-89-01 on 890417-21.Exam Results:All But One Senior Reactor Operator Passed Operating Exam & All Reactor Operators & Senior Reactor Operators Passed Written Exam

ML20247K512
Person / Time
Site:	Braidwood
Issue date:	05/23/1989
From:	Burdick T, Reidinger T, Shepard D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20247K500	List: ML20247K492 ML20247K512
References
50-456-OL-89-01, 50-456-OL-89-1, NUDOCS 8906010314
Download: ML20247K512 (75)

Text

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! U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No'. 50-456/0L-89-01.

Docket Nos. 50-456; 50-457 Licenses Fo. NPF-72; NPF-77 Licensee: Commonwealth Ecison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Braidwood Examination Administered At: Braidwood/ Production Training Simulator Examination Conducted: Requalification Examinations for six Reactor Operators and six Senior Reactor Operators RIII Examiner: /# f/.OM D. Shppard Date Chief Examiner: ((

T. Re'idinger (/ Date d #f Approved By: / jfff, /M Y.ll/f7 T. 3updick / ~ Date Examination Summary Examination Administered on April 17-21, 1989 (Report 50-456/0L-89-01):

Consisted of written and operating requalification examinations administered to six reactor operators and six senior reactor operators.

Results: All but one senior reactor operator passed the operating examination.

All reactor operators and senior rez.ctor operators passed the written examination. The requalification operating' examinations were administered to three crews; two crews passed the operating portion of the examination. The licensee's requalification program is declared satisf actory in accordance with the program performance criteria in NUREG-1021 " Operator Licensing Examiner Standards," ES601, " Administration of NRC Requalification Program Evaluations."

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PDR V ADOCK 05000456 PNU

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i Facility: Braidwood Examiners: Reidinger, Shepcrd, Victor, Stein Date(s) of Evaluation: Week of April 17, 1989 Areas Evaluated: X Writter. X Oral X Simulator Examination Results:

R0 SR0 Total Evaluation Pass / Fail Pass / Fail Pass / Fail (S, M, or U)

' Written Examination: 6/0 6/0 12/0 S Operating Examination:

Oral 6/0 5/1 11/1 S Simulator 6/0 6/0 12/0 S Evaluation of facility written examination grading: S Crew Examination Results:

Crew 1 Crew 2 Crew 3 Evaluation Pass / Fail Pass / Fail Pass / Fail (5, M, or U)

Operating Examination Pass Pass Fail S Overall Program Evaluation Satisfactory Submitted: Forwarded: pproved:

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T. Reidinge T. Surdick Examiner Section Chief Branch Chief 2

1 REPORT DETAILS j l

1. Examiners T. Reidinger, NRC*

D. Shepard, NRC M. Stein, Sonalyst F. Victor, Sonalyst

• Chief Examiner 2 .~ Examination Developme.nt The facilities effort and timeliness in developing the requalification examination material in accordance with the Operating Licensing Examiner Standards.(ES-601) was generally commendable in regards to both quality and quantity of material was appreciated by the NRC. The examination team, which consisted of NRC examiners, facility representatives and production training representatives, was able to develop all phases of the examination entirely from the facility's developed materials.

However, the NRC identified some material that did not meet the requirements of ES-601 and could not be used without revision. The identified material deficiencies were brought to the attention of the facility representatives on the examination team so that the appropriate revisions were made to the material for examination. The following are a few examples of material deficiencies:

• Some written examination Part B, " Limits and Controls," questions developed could be simply answered from memory. These types of questions are not appropriate for open-reference examinations.

• Some Part B developed questions were direct "look up" type questions.

This type of questions did not meet, as a minimum, the comprehension level o " understanding, and therefore are not appropriate for open-r; a ?nce examinations.

• All dynamic siniulator scenarios failed to meet the criteria of ES-601, Attachment 6, " Simulator Scenario Review Checklist," in that they did not contain at least one emergency plan implementing procedure.

• Some Part A and B questions exhibited the following deficiencies:

a. Ambiguous wording, i.e., answers to the written question did not match the intent of written question.

b. Open ended questions that did not have one specific answer.

c. Multiple choice questions hd non-discriminate distractors, i.e., all of the above, a and b above.

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i Ld. Double jeopardy questions.

e. Trick questions (ask for effects when there is none).

f. Superfluous-wording - unnecessary information.

Some written examination questions-and the Joo Performance Measure's (JPM's) questions had deficiencies in the inconsistent application of the correct knowledge and ability (K/A). In most cases, the correct system was used but the incorrect ability or knowledge was assigned. All of identified deficiencies.were discussed with the faci.lity representatives of the exam team in detail.

It is important to get objective questions on the written examination to ensure consistent grading. This would ensure that parallel grading is within allowed tolerance as contained in ES-601.

The facility should conduct a thorough review of the written examination question Bank and JPM Question Bank in order to identify _

all the additional questions which contain the identified deficiencies-and revise the questions'as necessary to fully meet the requirements of ES-601' prior' to future examinations. 1 Time validation of the written question encountered some immediate NRC concern in regards to the proper time validating as required per Exam Standard ES-601._ Subsequent re-validation times displayed no significant statistical deviations from the original tih:9 validation.

~The.following'contain several observations that were made by the NRC concerning the dynamic simulator scenarios that were developed for use during'requalification examinations.

• The facility identified all knowledge and abilities (K/A) as contained in NUREG 1122, " Knowledge and Abilities Catalog for Nuclear Power Plants," with an importance rating of greater than/ equal to 3.5.

• The facility and production training center representatives incorporated the required Team Dependent and Time Critical tasks in a well developed format and exceeded the minimum standards invoked by ES-601. NRC generally accepted all the recommendations made by the examination team for essigning the critical tasks with the exception that the NRC holds the Senior Reactor Operators responsible for the Generating Station Emergency Plan (GSEP) classification, i.e. , the SR0 will classify the scenario event per appropriate declaration in accordance with approved Station procedures. It is the NRC's position that this will be an SRO critical task for all requalification examinations.

• All the scenarios incorporated an overview of the malfunctions and the transients which comprised the scenario. This enhanced the simulator operator's comprehension of each scenario set. Also the 4

simulator setup Guide for initializing, lining up control boards, and cues for_ load swing instructions / sign offs was generally excellent.

• A formal change was required from the original scenario design.

This required developing the STA position and responsibilities in each of the six scenarios. Originally, the STA duties were incorporated into the SR0 evaluation guide but the STA is required to have its own evaluation guide for requalification examinations.

• All the scenarios were generally lacking in the concise developed descriptions of er .ted actions for evaluating operator performance for all emergency esponses. The facility modified all the scenarios to incorporate ' .Jor emergency procedure recovery action.s or immediate actions per added attachments (i.e., the facility modified the scenarios to outline major subsequent recovery actions, one example is EP-2 Faulted Steam Generator Isolation:

o Check main steam isolation o Check for at least one non-faulted S/G o Identify and isolate faulted S/G o Check for S/G tube rupture)

• All the scenarios generally exhibited a tendency for the expected actions of the Senior Reactor Operator to refer to the applicable Technical Specification strictly by number. The facility revised all references to Technical Specifications to incorporate a more meaningful description of the Tech / Spec LC0 and action statements.

• All the scenarios generally exhibited a common tendency in the expected actions for all the crew members to refer to the applicable abnormal procedure strictly by number, i.e., SR0 refer to 0A INST-2.

The facility revised a?1 references to abnormal procedures to incorporate an overview of all the major actW. steps outlined in the applicable procedure.

• The preparation week of running the simulator scenarios for examination review identified a scenario which required a crew to complete a normal evolution prior to encountering a specific malfunction in that normal evolution. The agreed upon corrective actions to allow the crew to address this malfunction and mitigate it was not accomplished due to an incorrect timing error in running the simulator.

• The validation team, during preparation week for t.he dynamic simulator scenarios, included a Production Training Principal Instructor who was cognizant of the subtleties and nuances of some of the scenarios and the changes required to correct them.

This individual was absent the days the scenarios were executed for the examination. It is rcccmmended that the simulator evaluator 5

included in the original preparation examination team be present during the examination week to ensure all commitments for corrections are implemented; as a minimum, changes to scenarios to allow proper crew interaction need to be clearly identified n the simulator operator's copy of the scenario. Examiner Standard ES-601 recommends that a certified instructor provide assistance in the development and review of the simulator scenarios during requalification examinations.

One or two scenarios did not exhibit a series of malfunctions which should be logically related or linked events as required per the Examiner Standard. The scenarios overall did attempt to.have related malfunctions in addition to a linked major plant transient.

• Time and procedure validation of the seven dynamic scenarios was accomplished in one working day. This required an accelerated critique of all the steps or procedures used during normal evaluations. This is undesirable as this time restriction based on simulator availability did affect the thoroughness of examination review.

• The scenarios need to incorporate operator options outside of the textbook anticipated actions to fully develop all the operatcr's expected actions.

The following contain several observations that were made by the NRC concerning the Job Performance Measures (JPM) that wer laveloped for use during requalification examinations.

• The preparation week of reviewing the JPM's for requalification examinations identified several JPM's which required corrective actions prior to using the JPM's for examination purposes. The corrective actions to modify the JPM's as agreed were not accomplished in a few cases.

• All JPM's exhibited excessive time validation for individual completion due to the JPM's questions at the end of the JPM being incorrectly incorporated into the validation time.

Examiner standards ES-601, Requalification Program,"

Attachment 12, excludes the questions related to the JPM from the time validation.

• The JPM's addressed all the requirements in that none of the examination JPM's had any identified deficiencies relating to the required references, task standards, task conditions, cues and critical elements.

• .ne JPM's met the performance standard in that all criteria was specified for the successful and required completion of steps and f

were identified either by closed (required completion) or open (optional completion) bullets.

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• Critical steps in the JPMs's in general were' accepted by the NRC.

Some critical steps were identified by the examination team as non-critical steps for the JPM's successful completion. Some critical steps had zero tolerance for operator error, i.e., a

' dropped rod had a JPM critical step which had the operator drive the rod in exactly ten steps, one operator drove the rods in 11 steps which per the critical step standard, the operator failed to meet and execute that required specific step. Failure to perform one critical step in a JPM constitutes a mandated failure for that JPM completion. However, the rod drop step tolerance was modified after discussions with the examination team members. Post JPM modification changed the tolerance to allow a few additional steps for the rod insertion. The majority of JPM's however did incorporate a better tolerance range for critical steps required by specific parameters, i.e., boration had a range of 300 gallons i 50 gallons.

• The JPM's required two question minimum was generally exceeded by the facility. The required question / answer references were not incorporated as required per Examiner Standard. Some JPM questions needed substantial revision prior to being used for the examination and it appears that some part A/B written questions were inadvertently written into some of the JPM's replacement question bank. This is undesirable as it might lead to a potential compromise of the written examination. However, the questions were not exactly identical in construction and design to the original question.

3. Examination Administration The facility was responsible for examination administration with the NRC observing the process. The following are a few specific program strengths and deficiencies that were identified by the'NRC during examination administration:

• Formal checklicts had been developed by the facility which were utilized to brief the operators prior to each phase of the examination. The formal briefing checklist enhanced the ability of the facility representative to provide consistent information to each group of operators to ensure they fully understood the examination process.

• Precautions, Limitations, and Setpoints (PLS) document was not initially available for operator examination purposes during the static simulator. It was subsequently provided to the operators when it was identified as missing. A PLS document is available for use in the station's control room.

• Transportation coordination and security accommodations of the crews for the written examination, Part A and Part B were excellent.

Fluid synchronization by the crews between the Production Training Center and Braidwood station er.hanced the timely completion of the written examinations.

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• During the dynamic simulator scenarios one Auxiliary Feedwater pump (A) did not start when the operator manipulated the switch and SI-8801A valve did not open when actuated by the operator to increase charging flow. In uoth instances, no major perturbation was noted on the crew or the scenario overall.

• The facility provided a sufficient number of' examination proctors during administration of the written examinations. The two part examination was administered in two separate locations, with two facility representatives available to proctor both examination rcoms while at the same time providing escorts to any individual wanting to leave the examination rooms ensuring that they did not interact with any other individual participating in the examination.

• Good simulator execution and coordination was generally displayed by all the new simulator operators not originally involved during the preparation week. The simulator operators need to be " heads up" on any phone calls made from any member of the crew. In some cases phone calls were placed to auxiliary operators and/or to Health Physics, but no further acknowledgement was made back to the specific crew member. While it is recognized that this was not anticipated and not reflected on the simulator setup guide, the simulator operators need to "ad lib" phone calls from the crew members in a realistic technical fashion without impacting the execution of the scenarios.

The static simulator for the steam break (Part A of the wr:tten) had containment pressure of 13.5 psig vice 15 psig as required by the facility answer key. Also during initialization and simulator setup for SX failure, a reactor trip occurred and required additional setup time to finalize the scenario.

• The Facility and Production training center exhibited excellent coordination for JPM's completions at the plant and at the simulator, the scheduling of JPM's enhanced the possibility that only one operator would be stationed in a particular area or needing a specific procedure / equipment at any one time. This provided for a timely completion of this phase of the examination.

A. Examination Evaluations _

Coevaluation by the NRC examiners and the facility evaluators of 1

the operators performance on the examination was incorporated.

Coevaluations provided the NRC with the necessary information to assess the individual operators performance as well as the facilities requalification program performance.

The overall evaluations on the written examinations vith parallel grading of the written examination by the NRC and the facility l resulted in gens. rally consistent evaluations for all 12 operators.

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The overall evaluations on the operating examinations; which consisted of dynamic simulator scenario operations and job performance measure (JPM) plant walkdowns, were generally consistent between the NRC examiners and the facility evaluators for all 12 operators. In addition, the overall facility individual performance evaluation during the dynamic simulator examination was more conservative in that the facility failed one senior reactor operator in addition to a crew failure identified by both the NRC and facility.

The following are some observations made by the NRC following the operational examinations concerning individual / crew evaluations.

• Communications.between crew members during dynamic simulator scenario events needs to be evaluated for improvement. There were several instances of "open loop" communications between the crew members. Some individual crew members failed to ensure that all the crew members they were addressing heard and understood all the transmitted information which resulted in required actions being delayed.

• The Senior Reactor Operators generally conducted " side conversations" with the applicable crew member during normal and abnormal plant evolutions, not all crew members were involved in the crew analysis l of the plant conditions.

• Examiner's standards require that the crews shall be evaluated by both the NRC and facility evaluators with the NRC observing the facility led critique. The facility opted to give a general simulator crew evaluation to each of the crews without'giving specific examples of individual crew performance. This option exercised by the facility was to minimize impacting the operator's morale for the balance of the requalification examinations which consisted of the JPM's and the written examinations. The final critical crew evaluation without the crew present was accomplished at the completion of the crew's Job Performance Measures. The crew and individual critique without the crew present was generally critical l in nature but not to the degree the NRC expected for crew evaluations.

The required facility led critique of the crew was not witnessed by the NRC due to the conclusion of the requalification examinations on April 21, 1989. It is recommend that the facility led crew critique be held immediately following the dynamic simulator scenarios to immediate provide feedback to the crew members of their observed deficiencies. The critique is designed to determine if there are areas in which retraining is needed to upgrade licensed operat3r and senior operator knowledge.

• JPM No. N-32, " Perform a Radwaste Liquid Release," had a JPM i question which had a generic weakness in that 50% of all the l operators incorrectly answered the question.

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• ' JPM No. N-17, " Response to a General Warning Alarm," had a JPM question which had 75% of the operators incorrectly answering the question. Of the 75%, the SR0's had a 66% failure rate and the RG's had a 83% failure rate.

• An SR0 missed a transition from EP-1, Loss of Reactor or Secondary Coolant, Step 7 to ES-1.1, SI Termination, Step 1. With pressurizer level > 38%, criteria for transition was established.

• JPM No. N-59, " Transfer to Cold Leg Recirculation," was selected as one of the common JPM's for R0's and SR0's. This JPM had a overall failure rate of 33.3% which was the highest of all the common JPM's.

While it satisfactorily meets the conditions at least 50% of the examinees successfully completing this JPM per ES-601, some generic comments are in order. Operators generally didn't exhibit a high familiarity in executing the required procedural steps. i.e., one operator attempted to established both trains of RHR pump suctions from the containment recirculation sumps simultaneously vice using the procedure correctly in establishing RHR pump suctions on one train at a time. Other operators failed to " verify Auxiliary building and Control Room Ventilation Line-up," as required per procedure. Although this lineup step was not classified as critical by the examination team, it leads to an assumption that operators aren't accustome.1 to performing the applicable steps in this procedure.

B. Examiner's Concerns During administration of the operating examinations, the NRC identified several operational concerns which are described below.

• The SCRE role and responsibilities needs more definition in that the SCRE generally demonstrated a non-assertive position during all dynamic scenarios in either assisting .he Senior Reactor Operator or a junior crew members. The SCRE needs to assume more responsibilities i in assisting the Senior Reactor Operator in all facets of control room operations. The SCRE reactions varied from a stand off approach with very little supervisory assistance and needing considerable prompting by SR0 for assistance to a SCRE who aggressively I integrated himself into all malfunctions and transients in the

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control room providing assistance to all crew members.  !

• Completion of identified simulator JPM's was adversely impacted by the non-availability of the simulator during the examination week.

The facility adjusted the JPM's scheduled for simulator performance to be performed in the plant accordingly but recognizes that the simulator is a better tool for accomplishing JPM's specifically designed for use on the sialulator. '

• During the execution of the " Local Emergency Start of a Diesel l Generator," several operators omitted the critical steps of exercising either the Emergency Stop pushbutton or Emergency Stop Reset as required by the procedure.

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• GSEP classification by the SR0's during tne dynamic scenarios reflects a training deficiency. Training utilizes a method in SR0 classification of GSEP that is not identical to the method employed-at the plant. The NRC recommends an adoption of a consistent approach to GSEP classification for all the Senior Reactor Operators.

4. Evaluation of Facility Evaluators In addition to evaluating the operators p eformance, the NRC'also evaluated.the facility evaluators, using ES-601 as a guideline, in their ability to conduct consistent and objective examinations. Included in this evaluation is the ability of the facility evaluators to provide an unbiased evaluation of the facility operators.

The following are some examples of the observations made concerning the facility evaluators:

• During administration of the JPMs, a few of the facility evaluators at certain times used verbal cues and/or prompting which could lead the operator to a correct decision / action resulting in an adequate examination. When this occurred, the NRC examiner would privately counsel the facility evaluator and point out the minor deficiencies in his examination administration techniques.

• On different occasions several of the facility evaluators would substitute a completely different question than the previously approved JPM question. This decreases the base required per Examiner Standards to judge the requalification program as satisfactory or unsatisfactory on common questions.

• As the individual JPM task completion progressed throughout the day, the facility evaluators generally would not require as thorough an answer as he required during the individual's first several JPM completions. This could be attributable to working extended hours for examination purposes which may have had a cumulati"e tiring effect on the facility evaluators. This evaluation fatigue also reflected itself in how the facility evaluators positioned themselves during the execution of JPM questions. When an operator exhibited difficulty in locating a component or piece of equipment, the evaluators would move over and position themselves in the vicinity of the component which could assist the operator in locating the proper component. While the NRC noted the pre-disposition of the evaluators to position themselves after long standup time, accordingly, no operators were noted to take advantage of this momentary mental oversight.

• On several occasions during the JPM administration, the facility f evaluators failed to ask followup questions when operator's l knowledge was in question or were unable to re phase questions presented to the operators if the operator was not sure what the question was soliciting.

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• The facility evaluators generally exhibited excellent judgement in grading the JPMs during their execution in accordance with the Examiner Standard ES-601.

• The. facility evaluators exhibited excellent judgment in several cases by providing additional correct cues when solicited by the

. operators but were not originally designed into the JPM. This enhanced the performance of the JPM without distracting the operator being examined.

• The speed at-which the JPM's were performed was consistent, there was not a sense of being in a hurry to "get through" the JPMs.

& The facility evaluators ability to discern errors was generally good.

Even though the facility evaluators exhibited lapses in their evaluation techniques during phases of the requalification examination, the evaluators were regarded as good overall. The previously mentioned observations point out the need to up grade the formal training on how to conduct examinations to the personnel who will be utilized as evaluators during requalification examinations.

5. Requalification Examination Results Examination results are that the NRC passed crew's 1 and 2 and failed crew 3. Also the NRC failed one SR0 on JPM's.

Initially the facility passed all three crews on the operating examinations, fafled one SR0 in JPMs, failed one SR0 on the dynamic simulator examination, and failed one R0 on the written examination.

The NRC provided final crew and individual results to the facility upon receipt of the facility's initial operating test ~past/ failure results.

The facility then revised their initial simulator crew results by electing to fail crew 3 after further review of the failure criteria outlined in Examination Stindard 601. They transmitted these new results on simulator crew failure / pass to the NRC when the final exit was being conducted between the Facility and NRC. This occurred two working days after the facility initial results of simulator crew passes / individual failures were given to the NRC.

6. Program Evaluation Per NUREG 1021, ES-601, Braidwood requalification program satisfactory met the following criteria:

a. The y was a 90% pass / fail decisions agreement between the NRC and facility grading of the written and operating examination.

b. The program was judged satisfactory in accordance with criteria given for simulator evaluations.

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c. At least 75% of all the operators passed the examination.

d. The program' meets the requirements of 10 CFR 55.59(c)(2), (3) and (4), and is based on systems approach to training.

7. Exit Meeting An exit meeting was held on April 25, 1989,.between the facility and the NRC to summarize all of the observed requalification program and operator strengths, deficiencies and concerns.

Attendance List D. Shepard, Examiner, NRC T. Burdick, Section Chief OLS, NRC T. Reidinger, Chief Examiner, NRC K. Shembarger, Examiner, NRC R. Legner, Services Director, CECO L. D. Gerling, PTC Simulator Supervisory, CECO J. E. Browring, ILT Simulator Principal Instructor, CECO M. Olson, Requalification Principal Instructor, Ceco G. Vanderheyden, Operations Training Supervisor, CECO D. O'Brien, Technical Superintendent, CECO R. J. Ungeran, Administration Operational Engineer, CECO R. E. Querio, Station Manager, CECO K. L. Kofron, Production Superintendent, CECO T. Taylor, Resident Inspector, NRC l

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SIMULATION FACILITY REPORT Facility Licensee: Braidwood Nuclear Station Facility Licensee Docket No. 50-456-Operating Tests Administered At: Braidwood Nuclear Station ,

l During the conduct of the simulator oort'lon of th; operating tests, the following items were observed:

ITEM DESCRIPTION

-AFW pump:1A did not auto start when required

-SI-8801A valve. did not open when required t

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SCENARIO NUMBER  !

R1.D.EP-06-01 l SCENARIO TITLE MAIN STEAM LINE BREAK ON 1B STEAM GENERATOR - INSIDE CONTAINMENT  !

REVISION 1 i

APRIL 1989 .

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i Developed By l PTC/PWR Operations j Rev. 1 772M/1 4/89 i- A___ - _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ __ _ _ _ _ _ _

SETUP INSTRUCTIONS i

REFERENCE SEQUENCE SETUP ACTION (S) REQUIRED {

1. '

Initialize on 10-14, 50% power, BOL, equilibrium xenon.

2. Run the simulator to stabilize plant conditions.

3. Insert Malfunctions CRF-13A (D-4,0) and CRF-13B (F-2,0)

4. Insert MANUAL Rx trip and MANUAL SI

5. Insert Malfunction MSS-2B (1.5 E6, 120, 0) to initiate the fault in contain:nent.

6. Perform the actions as addressed in EP-0, EP-2 and EP-1.

Do not rush through the procedures. Isolate AF flow to to faulted IB SG only when directed to in EP-2.

7. Run the simulator = 10 minutes. The following conditions should exist prior to freezing the simulator:

a. Containment pressure must be > 20 psig with the Hi-3 Annunciator lit.

b. RCS pressure should be increasing.

c. Faulted S/G pressure stable or increasing.

d. CS has actuated and the RCP's have been stopped.

e. RCS Ta vg < 5640F.

8. Adjust variable ACNMAIR as necessary to insure containment pressure is > 20 psig before freezing the simulator.

9. Freeze the simulator.

10. Have EP-0 open to the diagnostic kickout step for EP-2 and have EP-1 open at step 8.

11. Turn OFF all chart drives.

12. /I-/f%RecordPzrLevelforKEY I f O @ psig Record RCS pressure for KEY Y O psig Record 1B S/G pressure for KEY Rev. 1 772M/5 4/89

. l SCENARIO

SUMMARY

Unit'l had been operating at 50% power with all systems functioning properly in automatic. No LC0AR's or abnormal out ot services existed at the time of the turnover.

The following plant conditions existed at the time the shift turnover occurred:

l Mode 1 Rx Pwr 53  %

Turbine Pwr 580 MWe Boron Concentration 800 ppm Rod Position Bank D at 208 steps Chg Pmp PD .l_A_l 'B RCP's l_A_l l_E_l l_C_l l_D_l Sx Paps B CCW Pops l_A_l B . O

.l_A_l 5/4 Cond. Pmp l_A_l l_B_l l_C_l D Feed Pmps A l_JLl l_C_l Htr Drn Pumps l_A_l l _B,_ l C Cir Wtr. Paps l_A_l l_D_l l_C_l Electrical 4 kv 141 lAatl DG 241 6.9 kv 156 luatl sat sat 142 l1A1l DG 242 157 l5EEl 158 uat l tat l 143 sat DG lua1l 159 uat laall 144 sat DG luatl 10 minutes ago containment parameters indicated a faulted steam generator within j

containment. The operators initiated a manual reactor trip sud safety injection. The operators implemented EP-0, progressed to EP-2 and then progressed to EP-1. The crew is

[ ' currently at step 8 of EP-1. Only the actions dictated in the Emergency Procedures have I .been performed.

Rev. 1 772M/4 4/89

QUESTION # R1.C-FW-04B-01 K/A # 059000 K4.16 f

K/A RATING R0-3.1 SRO-3.2 TIME PTS 2.0 min .5 pts.

QUESTION:

Determine the initiating signals that could have tripped the Main Feedwater Pumps.

(2 answers required)

ANSWER:

Safety Injection Actuation. (.25) -

P-4/Rx trip with 5640/ Low Tayg. (.25)

REFERENCES:

BwEP-0 Step 5, Rev. 2A; 20E-1-4030FW33 Rev. L

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

NO YES / NO YES /

SME: PWR License Supervisor:

Rev. 1 772M/6 4/89

QUESTION # R1.C-CV-06D-01 K/A # 006000 K1.08 K/A RATING R0-3.6 SRO-3.9 TIME PTS 3.0 min .5 pts.

QUESTION:

Under present-plant conditions, explain why you are not able to keep open the " seal water return isolation" valves, CV8100 and CV8112.

ANSWER: (.5 pts)

Phase A (OA) signal. -

1.

REFERENCES:

20E-1-4030CV12

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

NO YES / NO YES /

SME: PWR License Supervisor:

l Rev. 1 772M/8 4/89

QUESTION # R1.C-CV-06D-02 l K/A # 013000 K1.11 K/A RATING R0-3.3 SRO-3.8 TIME PTS 3.0 min .75 pts.

. QUESTION:

Prior to the Faulted Steam Generator, the NSO reported that charging was going to Loop 1A. Explain why a manual SI caused CV-8146 to be in it's present position.

/NSWER: (.75 pts)

The SI caused a phase A isolation signal (.25) causing CV8146 to open on a loss of instrument air (.5) l l

l l

REFERENCES:

6/20E-1-4030CV18, Rev. H l 1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 772M/10 4/89

QUESTION # R1.D-EP-06A-01 K/A # 026000 K4.01 K/A RATING R0-4.2 SRO-4.3 TIME PTS 2.0 min .5 pts.

QUESTION:

Determine the source (s) of water in the containment recirculation sumps.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • (*********

ANSWER: (.5 pts)

The refueling water storage tank / containment spray (.25) and the faulted steam generator (IB). (.25)

REFERENCES:

Main Control Board indications

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

a YES / NO YES / NO SME: PWR License Supervisor:

l Rev. 1 772M/12 4/89 i

QUESTION # R1.D-EP-06A-02 K/A # 006000 K5.06 K/A RATING R0-3.5 SRO-3.9 TIME PTS 2.0 min j.5 pts.

9 QUESTION:

Explain whether or not the Safety Injection (SI) Pumps have injected any water into the RCS.

ANSWER: '(.5 pts)

The SI pumps have not. injected (.25) because RCS pressure is above the shutoff head (1590 psig) of the SI pumps (.25)

REFERENCES:

Main Control Board RCS Pressure Recorders 1PR-403

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • J I

i

1. Does the question have a direct relationship to RO/SRO job performance? {'

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

• l YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 ,

772M/14 4/89 )

, QUESTION # R1.D-EP-06A-03 K/A # 039000 K1.01 K/A RATING R0-3.1 SRO-3.2 TIME PTS 2.0 min 1.0 pts.

QUESTION:

Determiv which one of the following best describes the physical location of the fault on the 1B steam generator:

a. On the Main Feedline downstream of FW-009B. (Main Feedwater Isolation Valve)

b. On the IB S/G manway.

c. On the Main Steamline downstream of the steam flow transmitter taps.

d. On the Main Steamline upstream of the eteam flow transmitter taps.

ANSWER: (1.0 pts) .

c. On the Main Steamline downstream of the steam flow transmitter taps.

REFERNCES : Main Control Board Indications P&ID M-35 Sheet 1, Rev. AK

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • =*********************

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

/ NO YES / NO YES SME: PWR License Supervisor:

Rev. 1 772M/16 4/89

i .

QUESTION #~ R1.D-CS-02A-01 K/A # 013000 K1.05 K/A RATING R0-4.1 SRO-4.4 TIME PTS 3.0 min .75 pts

• • • • • • • • • • • • • • • • • w***********************************************************************

QUESTION:

When conditions permit, what must be done to stop the Containment Spray pumps and place them in standby?

ANSWER: (.75 pts)

CS actuation reset (.5) and control switch to trip /after-trip. (.25) l l

i

REFERENCES:

20E-1-40300S01, CS02 1

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the ansvt 2 key match the requirements of the question?

5. Is there only ine right answer?

YES / NO YES / NO SME: WR License Supervisor:

Rev. 1 l 772M/18 4/89

QUESTION # R1.C-QG-06D-01 K/A # 010000 K4.02 K/A RATING R0-3.0 SRO-3.4 TIME PTS 2.0 min .5 pts QUESTION:

Explain why the pressurizer backup heaters (Groups A, B, and D) are not energized even though pressurizer pressure is less than 2210 psig?

ANSWER: f.5 pts) (either response acceptable)

Pressurizer level is < 17%.

Or Pressurizer level is lY-l fI % (Read simulator value. Recorded on page 772h/. of Key) l

REFERENCES:

6/20E-1-4030RY04, RYO6, RY10

1. Does the question have a direct relationship to RO/SRO job performance? l l 2. Is the question stated precisely and unraSiguously? I

3. Is the question written at the knowledge level appropriate for requal? ,

4. Does the answer key match the requirements of the question?  !

5. Is there only one right answer?

i YES / NO YES / NO SME: PWR License Supervisor:

l l

{

Rev. 1 i 772M/20 4/89 )

l

QUESTION # R.1.C-CC-04F-01 K/A # 000026 EK3.02 K/A RATING R0-3.6 SRO-3.9 TIME PTS 4.0 min 1.0 pts.

QUESTION:

. Determine the two (2) possible reasone why the "CC from RC pumps THERM BARR ISOL VLV" ,

(CC-685) automatically closed.

ANSWER: (1.0 pts)

1. Phase B isolation signal /20 psig containment pressure (Hi-3) (.5)

2. Hi return flow from the thermal barriers (192 apm) (.5) .

U '

Hi return flow due to start of additional CCW pump l

l l

)

i

REFERENCES:

20E-1-4030CC04

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • <.t*********************

1. Does the question have a direct relationship to R0/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer? j YES / NO YES / NO SME: PWR License Supervisor:

i Rev. 1 l 4/89 '

772M/24

QUESTION # R.1.D-EP-19D-01 K/A # 000040 EK3.04 K/A RATING R0-4.5 SRO-4.7 TIME PTS 4.0 min 1.0 pt.

QUESTION:

Explain why SI termination criteria was not met when step 6 of EP-1 was performed.

'(Include applicable cetpoints) i

$**************************************************************************************** j ANSWER: (1.0 pt)

Pressurizer level (.5) < 38%, (.5) .

REFERENCES:

EP-1 Step 6 Rev. 2A

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 l 772M/34 4/89

O

• l .

QUESTION # R.1.D-EP-19D-02 l

K/A # 000040 EK3.04 K/A RATING R0-4.5 SRO-4.7 TIME PTS 2.0 min 1.0 pts.

QUESTION:

Explain whether or not the SI signal has been reset.

ANSWER: (1.0 pts)

SI has not been reset (.5) .

Auto SI blocked permissive light is not lit (.25)

SI actuated permissive light is lit (.25)

REFERENCES:

EP-1 Step 9 Rev. 2A

1. Does the questzon have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 772M/36 4/89

- 1 QUESTION # R01.D-EP-06B-01 K/A # 000009 EK3.21 l K/A RATING R0-4.2 SRO-4.5 TIME PTS 2.0 min .5 pts.

QUESTION:

Explain whether or not containment spray may be stopped per present plant conditions.

(Include all conditions which are applicable in your discussion).

)

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ce*********************

ANSWER: (.5 pts)

G ...y nu6 ~oe securea (.zb) until uud. press is ( ia psig (press is currencly i 15 v.ig) (.431 Tvv1C t 1. </ 7 an W2l-$ 9 4

L .k, m c a a m ,~4 p e cs-

• rJg 8 3 , r p s ,~g y n m .d Ac mea *<.. . J ue d W ro b , A s a ns w er .1 d*M 4

.. e s m h< see w ) (.2 d Lee w c4 fuss +Ane ' 6kr / U'd'", 'h*

l a J . q nay L <>c h n n . 4- s fe og (ess inne

\:n:k ts n A .y 1:e . L ie C. t r) g W CM.

RE!ERENCES: EP-1 Step 8 Rev. 2A

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • e*********************************o****************

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

l 3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

I

(

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 772M/38 4/89

i QUESTION # Rel.D-EP-06B-02 4 K/A # 000009 EK3.21 K/A RATING R0-4.2 SRO-4.5 TIME PTS 3.0 min .75 pts.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 3 QUESTION: l i

Determine which safeguards equipment would have to be manually started after Safety l Injection has ha-r reset if offsite power is lost. )

I ANSWER: (.75 pts)

RH pumps (.375)

SI pumps (.375) .

REFERENCES:

EP-1 Step 9 Caution Rev. ZA 20E-1-4030EF01 Rev. 5

1. Does the question have a direct relationship to RO/SRO job performance?

l 2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate fer requal?

4. Does the answer key match the requirements of the question? I

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

l Rev. 1 772M/40 4/89

e

• ~

QUESTION # R.1.D-EP-060-03 K/A # 000009 EK3.23 K/A RATING R0-4.2 SRO-4.3 TIME PTS 2.0 min .5 pts.

QUESTION:

Explain what plant conditions required the NSO to manually stop all Reactor Coolant Pumps.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ce********

ANSWER: (.5 pts)

CC water was lost Phase B isolation (20 psig enst pressure - Hi 3)

REFERENCES:

EP-1 Operator Action Sunniary Rev. 2A

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 772M/48 4/89

QUESTION # R.1.D-EP-06C-04 K/A # 000040 GO.07 K/A RATING R0-3.7 SRO-4.2 TIME PTS 5.0 min 1.0 pts.

QUESTION:

a. Calculcte the primary to secondary AP across the IB steam generator and

b. Compare this valve to allowable limits (Indicate all parameters and values uted to obtain your answer)

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • e**********

ANSWER: (.5 pts)

A 33C6 tpsi h>3 5%) (.2 pts) i

• Primary pressure i

• Secondary pressure 40p lg(*(i 5%) (.2 pts)

AP 2r76 ticopsid (i 5%) (.2 pts) .

Allowable limit is 1600 paid (.2 pts)

Exceeding allowable limit (.2 pts)

(* Obtain values recorded per page 772M/5 of KEY)

Allow for math errors carried forward

REFERENCES:

GP 100-1, Limitation and Action 2.g

• • • • • • • • • • • • • • • • • • • • • • • • • • • a*************************************************************

1. Does the question have a direct relationship to RO/SRO job performanc67

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does che answer key match the requirements of the question?

5. Is there only one right ar,wer?

NO YES / NO i YES /

SME: PWR License Supervisor:  ;

i kev. 1 772M/50 4/89

QUESTION # R1.D-EP-19A-02 I

K/A # 010000 Kl.02 K/A RATING R0-3.9 SRO-4.1 3 I

TIME PTS 2.0 min 1.0 pts. l l

cc****************e*******************************************e**************************

QUESTION:

If no further operator actions are taken over the next 20 minutes, RCS pressure will:

a. Be controlled by operation of Pzr heaters and sprays.

b. Be controlled by the Pzr PORV.

c. Be controlled by the Pgr safety valve.

d. Be controlled by the shut off head of the operating high-head injection pumps.

oce*******************************************************************************t******

ANSWER: (1.0 pts)

b. Increase to the pzr PORV setpoint.

REFERENCES:

P&ID M-60-5 M-64-5 e*****=*************************************************************************aa*******

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

/ NO YES / NO YES SME: PWR License Supervisor:

Rev. 1 772M/54 4/89

h

,4-d * .: g , -4

~

r y . ..s.,

,: , , D.

Mr) h.L 't.;ala CI'$15$ k SCENAkIO NUMBER R1.C.SM-03-01

. SCENARIO TITLE FAILED OPEN STEAM GENERATOR PORV 1A SX PUMP TRIP REVISION 1 APRIL 1989 .

L L.

I Developed By PTC/PWR Operations Rev. 1 788M/1 4/89 I

L

SETUP INSTRUCTIONS SETUP ACTION (S) REQUIRED REFERENCE l'EQUENCE.

1. Initialize IC #17 100% power, BOL, Equil Xe

2. RbN simulator to stabilize and setup traces. (2 minutes)

3. PLACE rod control in MANUAL, turbine with MW IN.

4. PLACE ICV-121 it. MANUAL.

5. INSERT MALF AUX-4A (0,0) to trip the 1A SX pump.

6. ACTIVATE ANNUN AUX 1 to cause the SX PUMP SUCT PRESS LOW annunciator to be lit.

7. RUN simulator additional 6 minutes.

8. INSERT MALF MSS-180 (1,0) to fail open 10 S/G PORV

9. PLACE MS0180 control switch to CLOSE position and the controller in MANUAL. .

10. RUN simulator approximately 2 more minutes

11. VERIFY the.following conditions exits:

- ICV 129 has bypassed the demins on high letdown temp.

- Program and actual PZR level traces have trend decrease f

- OTAT trip setpoiat trace has trend increase ,

- RCP seal outlet temp < 235'F

- RCP lower brg. temp < 225'F

- RCP status display is on MCB screen - dwg. #65

- FW flow traces have trend increase

- Tavg < Tref by more than 1.5'F.

- 10 S/G PORV TROUOLE alarm litt (1-15-C10)

(override ANNON MSS-11=1) l L

12. ACKNOWLEDGE all annunciators

\

13. FREE 7E simuintor.

14. STOP all chart driveo.

15. NO books to be opened on desk.

16. RECORD: RCP Seal Outlet Temp. /7f-/7$7 0F RCP Motor Bearing Temp. /6/-/43 0F Lower Radial Bearing Temp. /Cl-/f1L OF Rev. 1 788M/5 4/89

SETUP INSTRUCTIONS SEQUENCE SETUP ACTI6N(S) REQUIRED REFERENCE

1. Initialize IC #17 100% power, BOL, Equil Xe

2. RUN simulator to stabilize and setup traces. (2 minutes)

3. PLACE rod control in MANUAL, turbine with MW IN.

4 PLACE 1CV-121 in MANUAL.

5. INSERT MALF AUX-4A (0,0) to trip the 1A SX pump.

6 ACTIVATE ANNUN AUX-5 = 1 to cause the SX PUMP SUCT PRESS LOW annunciator to be lit.

7. RUN simulator additional 6 minutes.

8. INSERT MALF MSS-180 (1,0) to f ail open 10 S/G FORV

9. PLACE M50180 control switch to CLOSE position and the controller in MANUAL.

10. RUN simulator approximately 2 more minute i

11. VERIFY the following conditions exits:

- ICV 129 has bypassed the demins on high letdown temp.

- Program and actual PZR level traces have trend decrease l

- OTAT trip setpoint trace has trend increase I

- RCP seal outlet temp < 235'F

- RCP lower brg. temp < 225'F

- RCP status display is on MCB screen - dwg. #65

- FM flow traces have trend increase

- Tavg < Tref by more than 1.5'F. '

- 10 S/G FORV TROUBLE alarm lit: (1-15-C10) -

(override ANNON MSS-11=1)

" p.

7 V

12. ACKNOWLEDGE all annunciators FMl 4 z k%o

13. FREEZE simulator. h4 **

14.

[5. 6 & A-1 STOP all chart drives.  ;

15. NO books to be opened on desk.

i

16. BECORD: RCP Seal Outlet Temp. /7 7'/ OF RCP Motor Bearing Temp. /0 3-/ / d 0F Lower Radial Bearing Temp. /SW W OF Rev. 1 788M/5 '4/89

• 1

~'

SCENARIO

SUMMARY

Unit one has-been operating at 100% power for the last 3 months. Rod control is being

' controlled.in manual due to excessive rod ' hunting' in automatic. All other systems have been functioning normally in automatic. No LCOAR's or survelliances are in progress. At the time of shift turnover the following conditions existed, Mode 1 'Rx Pwr 99  %

Turbine Pwr 1174 MWe Boron Concentration 683 ppm Rod Position Bank D at 220 steps Chg Pmp PD l_A_l B RCP's l_A_l l_1_l l __C._ l l_JLl Sx Paps l_A_l B CCW Pops l_A_l B 0 Feed Pmps A l_J_ l l_C_l S/U Cond. Pop l_A_l l1l l __C._ l D Htr Drn Pumps l_A_l l_1_l C Cir Wtr. Pmps l_A_l l_JLl l_C_l Electrical 6.9 kv 156 luatl pat 4 kv 141 l5l DG 241 157 luatl sat 142 laatl DG 242 158 uat l mat l 143 sat DG lnatl 159 uat laatl 144 sat DG luatl l' Approximately 8 minutes ago the 1A SX pump tripped. The NSO attempted to start the IB f SX pump but was unsuccessful. Operators have been dispatched to check the SX pumps and

l. breakers. Within the last two minutes the shift has noted a decrease in Tavg along, with an increase in reactor power level. . Investigation has revealed the 1C steam generator PORV. 1MS-0180, open. No further actions have been performed at this time.

Rev. 1 788M/4 4/89 L _ _ _ . __ _ _ __ _ ___ _ _-______ _- _- - ___- - - - - ___ _ _ _

3 QUESTION # R1.C-QG-060-04 K/A # 035010 K1.09 K/A RATING R0-3.8 SRO-4.0 TIME PTS 5.0 min 1.0 pt.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • v*******************************************

QUESTION:

Explain why both actual and program pressurizer levels are trending as they currently are.

ANSWER: (1.0 pt)

(Increased steam flow causes Ta*L to decrease.)

Program Pressurizer level decreased (.25) due to lavg decrease (.25)

Actual Pressurizer level decreases (.25) (shrink) with the RCS cooldown (Tavg decrease)

(.25) l 1

REFERENCES:

MCB indications j

1. Does the question have a direct relationship to R0/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4 Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/6 4/89

4 QUESTION # Rl.D-EP-06A-04 K/A # 002000 K5.11 K/A RATING R0-4.0 SRO-4.2 TIME PTS 5.0 min 1.0 pt.

QUESTION:

Total steam flow in the secondary system has changed MQEE than was required to restore MW loading due to the steam generator PORV. Explain why. (Include any primary effects  ;

on the secondary)

)

l

• • • • • • • • • • • • • • • • a************************************************************************

ANSWER: (1.0 pt) -

Decrease in Tavg (.5) has caused steam temperature / pressure to decrease (.5) (rcequiring more steam flow for the same MW loading).

OR (also acceptable)

Q = UAAT with hconstant(.25)

UA constant (.25)

Tavg decrease (.25)

Tstm decrease (.25)

REFERENCES:

MCB Indications I

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor: .

Rev. 1 788M/8 4/89

4 QUESTION # R1.C-MS-06A-03 K/A # 035010 K4.05 K/A RATING R0-3.1 SRO-3.4 TIME PTS 5.0 min .75_pt.

QUESTION:

Explain why when you placed the 1C S/G PORV close switch to the CLOSE Position, annunciator 1-15-C10 "S/G IC PORV TROUBLE" illuminated. (Include in your response the SER alarm printout you would expect)

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • e*

ANSWER: (.75 pt)

Action dumps hydraulic / accumulator pressure (.5) SER point 2414/ STEAM GENERATOR 10 PORV ACCUMULATOR PRESSURE LOW (.25)

I l

REFERENCES:

AR 1-15-10 6/20E-1-4030-MS41

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

NO YES / N0 YES /

l SME: PWR License Supervisor:

Rev. 1 788M/10 4/89

i 1

QUESTION # R1.C-SX-02A-02 K// 076000 Kl.16 K/ *. RATING R0-3.6 SRO-3.8 TIME PTS 3.0 min 1.0 pt.

c****************************************************************************************

QUESTION:

Evaluate the cresent main control board valve lineuo and explain why:

c. The IB SX pump will AUTO start if a safety injection were to occur at this time.

b. The NSO was unsuccessful in his attempt to manually start the IB SX pump, e*********************************************************************.*******************

ANSWER: (1.0 pt)

c. . AUTO start only requires valve ISX-001B (pump 1B suction valve) to be OPEN (which it currently is)/All auto start interlocks are satisfied (.5)

b. Valve 1SX027B (RCFC outlet valve) is required to be OPEN (to manually start the IB SX pump, presently closed). (.5) *

REFERENCES:

6/20E-1-4030-SX01 l c****************************************************************************************

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

L 3. Is the question written at the knowledge level appropriate for requal?

l

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/12 4/89

~

E71, .

1 QUESTION # R1.D-0A-69D-02 K/A # 076000 K1.01.

K/A RATING R0-3.4 SRO-3.3 l TIME PTS 4.0 min 1.C pt.

QUESTION: 1 Explain tha relationship between the loss of SX (pump trip) and the CVC systems ,

inability to maintain norrsi reactor coolant pump seal outlet temperatures.

l ANSWER: (1.0 pt)

Loss of SX cooling to CCW (.25)

Less heat removed from letdown heat exchanger (.25) causing Seal injection (charging) temp (and thus seal retura temps) to increase. (.5) )

l l

REFERENCES:

AR 1-2-05 P&ID's M"56-1B, M-42-2B

1. Does the question have a direct relationship to R0/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is'there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/18 4/89

I QUESTION # R1.D-0A-10E-02 K/A # 003000 K1.03 K/A RATING RO-3.3 SRO-3.6 TIME PTS 5.0 min 1.0 pts.

ce***************************************************************************************

QUESTION:

You have reported to the SRO that N0 normal reactor coolant pump operating parameters-have yet been exceeded that would require tripping the reactor coolant pumps as M[

INDIRECT RESULT of the loss of SX. Support your statement with parameters and trip cotpoints monitored. (2 required)

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 7 ANSWER: (1.0 pt - 2 of 3 responses required)) , p f' O JJ g74

1. RCP seal outlet temperatures (.25) are < 235'T /not>2350F/* OF jl6i id

2. RCP motor bearing temperatures (.25) are < 195'r /not>1950F/* Y(.25) (.25) h2 W0ll0

3. kf? lower radial bearing temperature (.25) is < 225'F /not 9 /* A 0F (.25) o Actual value recorded on key page 778M/5 also accepta .

$l 15)-lO 0 2-152 l!G TVNb spb 6E ME up w uA

$% TIS

REFERENCES:

OA RCP-2 OP-RC1 - bb oe***************************************************************************************

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO 7ES / NO SME: PWR License Supervisor:

Rev. 1 788M/20 4/89 l

s L--. __m.____-________.___m __

QUESTION # R100-CC-04A-01 K/A # 004010 K4.03 K/A RATING R0-3.1 SRO-3.6 TIME PTS 2.0 min 1.0 pts.

QUESTIDN -

Concerning the current status of valves ICC-130A & B (letdown temperature control valves):

a. CC-130A&B have failed shut causing 00-130 controller to demand 100%.

b. CC-130A&B are 100% open and responding correctly to letdown temperature.

c.00-130 controller is demanding an increase in letdown temperature so CC-130A&B are closing.

d.- CC-130A & B have failed open due to low instrument air receiver pressure.

ANSWER: (1.0 pts) .

b. CC-130ALB are 100% open and responding correctly, to letdown temperature.

REFERENCES:

P&ID M-66-2

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the' question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

, 5. Is there only one right answer?

l L

t YES / NO YES / NO ,

SME: PWR License Supervisor:

l l

l l

Rev. 1 788M/26 4/89

QUESTION # R1.0-RP-01E-02 K/A # 012000 K6.11 K/A RATING R0-2.9 SRO-2.9 TIME PTS 3.0 min .5 pts.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2*****

QUESTION:

The OTAT setpoint recorder shows a change in the OTAT setpoint. What is the dominate factor causing this change?

ANSWER: (.5 pts)

Tavg decreased I

REFERENCES:

Tech Spec Section 2.0 P.L.S.

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question? l S. Is there only one right answer?

YES / NO YES / NO SME: FWR License Supervisor:

Rev. 1 788M/28 4/89

c 1

t .

[ QUESTION # R1.0-MS-06A-02 l-h K/A # 035010 K6.02 K/A RATING R0-3.1 SRO-3.5 t

TIME PTS. 5.0 min .5 pts.

. QUESTION:

I Why would you expect the S/G PORV to close when taking the control switch to CLOSE if it did not respond by operation of the M/A controller?

l 1

ANSWER: (.5 pts)

The control switch dumps hydraulic / accumulator pressure (to close the valve).

4

REFERENCES:

6/20E-1-4030-MS41

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/32 4/89

O QUESTION # R1.0-CV-06C-03 K/A # 004020 K4.04 K/A RATING R0-2.6 SRO-3.0 TIME PTS 3.0 min 1.0 pts.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ,m**********

QUESTION:

T*- CVCS demin divert valve (CV-129) will return to the DEMIN positic a when:

a. Letdown temperature decreases below 133*F AND it's control switch is in the AUTO position.

b. Letdown temperature decreases below 1330F &HD electrical power is removed from the valve actuator.

c. Letdown temperature decreases below 1330F AND instrument air is removed from the valve actuator.

d. Letdown temperature decreases below the reset temperature (below 133'F) AND it's control switch is placed in the DEMIN position.

ANSWER: (1.0 pts) .

o. Letdown temperature decreases below the reset temperature (below 133*F) 6HD it's control switch is placed in the DEMIN position.

l

REFERENCES:

6/20E-1-4030-CV28

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

l YES / NO YES / NO l

SME: FWR License Supervisor:

Rev. 1 788M/34 4/89 l

4 QUESTION # R1.C-SM-03G-01 K/A # 041020 G0.11 K/A RATING R0-2.9 SRO-3.7 TIME PTS 3.0 min .5 pts.

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • w************

QUESTION:

The 1C S/G PORV will not/cannot be closed. Identify the specific valve (s) (by valve number) that must be closed to meet the requirements of technical specifications.

ANSWER: (.5 pts.)

1MS0190 l

l l

REFERENCES:

P&ID M-35-2

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/36 4/89

Y QUESTION # R1.D-0A-69A-01 i K/A # 076000 K3.07 K/A RATING R0-3.7 SRO-3.9 TIME PTS 5.0 min 1.0 pt oo******************************************eas************************************a*****

QUESTION:

Evaluate the effect(s), if any, that the loss of SX (pump trip) will have on the cperation of both the 1A and 1B auxiliary feedwater pumps should both the A W pumps be required to start.

oce***************en*********************************************ene***eemmeneene******** -

ANS".lER : (1.0 pt)

The motor driven A W pump (1A) will overheat / damage the pump is operation is continued

(.5)

) The Diesel Driven A W Pump (1B) continue to operate normally/no effect on pump operation

(.5)

REFERENCES:

P&ID M-42 o****************************************eea******s******eene****************************

1. Does the question have a direct relationship to RO/SRO job performance?

2. Is the question stated precisely and unambiguously?

3. Is the question written at the knowledge level appropriate for requal?

4. Does the answer key match the requirements of the question?

5. Is there only one right answer?

YES / NO YES / NO SME: PWR License Supervisor:

Rev. 1 788M/40 4/89

y. .

CEV F: SIES GMimTID EP0lff Omith1012 IRfCLEAR PSER STATISI 11EW K/Af 1 nrmtIPTI(Il III.C.CV 01-A 00402015G2 3.8/4.1 G! YEN A COPY OF THE APPROPRIATE PROCEDUE(S) DISCUSS HOW TO LINEUP THE CVCS FOR VARIOUS OPERATIONS. (SOLIDPLMTOPERATIONS)

!!!.C.MS-06-A 035010K602 3.1/3.5 DISCUSS THE OPEMTION OF THE STEM GENERATOR PORVs INCLdDING BOTH MANUAL, LOCAL Am AUTOMATIC OPERATION.

!!!.C.KS-06-8 000040(! c6 3.6/3.7 DISCUSS THE CAPACITY Cf THE STDM GENERATOR SAffTY VALVE.

!!!.C.RC-01-C 002000G011 3.3/4.0 GIVS A SET OF PLM CONDITIONS OR PAMMETERS INVOLVING THE REACTOR C00LMT SYSTEM AND THE APPROPRIATE SECTION OF TECH. SPECS., DETEMINE TECH.

SPEC. COMPLIANCE AE REQUllED ACTIONS.

III.C.lel-03-D 005000K411 3.5/3.9 GIVEN THE APPROPRIATE PMCEDUE, DISCUSS HOW TO PLACE 1HE IDI SYSTEM IN THE SHUTDOWN COOLING MODE OF OPElp.YION.

Ill.C.RY-06-8 011000K510 3.7/3.4 GIVEN THE APPROPRIATE SECTIONS FROM PROCEDUIES THAT APPLY TO DMWING A BUBBLE IN THE PRESSURIZER: DISCUSS TE BESES POR EACH PRECMION, LIMITATION A E ACTION.

III.C.SI-05-A-2 006000k506 3.5/3.9 DISCUSS THE BASES FOR EACH NOTE AND CAUTION. (FLOWEFFECTONRCS PRESSURE)

!!!.D.AF-01-A 0610004206 2.7/3.0 GIVEN A COPY OF THE AUXILIARY FEEDWATER CHECK VALVE LIAKAGE PIOCEDURE, DISCUSS THE BASIS OF EACH STEP, NOTE OR CAUTION IN TE PROCEDURE.

(BACKLEAKAGEOFMAINFEEDWATER)

!!!.D.EP-02-C 000000G012 3.9/4.1 GIVEN A SET OF PLM COWITIONS OR PAMMETERS INDICATING A VAPOR SPACE LDSS OF COOLANT ACCIDENT, AND A SET OF PLANT PIOCEDURES, IDENTIFY THE CORRECTPROCEDURE(S)TOBEUTILIZEDA10DISCUSSREQUIREDOPEMTORACTIONS.

(STUCK OPEN PZR SAFETY)

!!!.D.0A-09-D 000025G0ll 3.6/3.9 GIVEN A SET OF PLM COWITIONS OR PAMMETERS IEICATING A LOSS OF RH (DOLING AND A SET OF PLANT PROCEDUES, IDENTIFY THE CORRECT PROCEDURE (S) TO BE UTILIZED AND DISCUSS REQUllED OPOIATOR ACTIONS.

III.E.M-04-A 194000K103 2.0/3.4 DISCUSS THE CECO RADIATION DOSE LIMITS.

Ill.E.M-08-A 0100000011 3.2/3.9 GIVEN A SET OF PLM LOWITIONS OR PAMMETERS DETEMINE IF ENTRY INTO A TECHNICAL SPECIFICATION ACTION STATDENT IS REQUIRED. (LER 88-020)

Ill .E.M-08-B 000mamna 3.1/3.6 GIVEN THE APPROPRIA.*2 SECTION 0F TECHNICAL SPECIFICATIONS AND A SET OF PLANT CONDITIONS OR PARAMETERS IEICATING A POSSIBLE LCO V!0LATION, DETEMINE TECH. SPEC.COMPLIANCEANDREQUIREDACTIONS.

Ill.E.M-14-L 194000K101 3.6/3.7 DISCUSS Ti.T ACCEPTABLE ETHODS OF PERF0 MING AN 12EPEEENT '

VERIFICAT10h AND THE PROPER APPLICATION OF EACH.

Ill.E.M-18-0 194000A110 2.9/3.9 DISCUSS PRDPER USAGE OF PLANT PROCEDURES. (LIR 88-001) 005054G011 3.4/3.3 DISCUSS PROPER USAGE OF PLANT PROCEDURES (BWFR H.1)

TRAINING DEPARTENT ONLY l

Page 1

. l

' CIEW F: 580 EXMIITIM IEPORT Bin!!NOtB INCLEAR PWER STATIM TIIM K/M ,,,,jg.,,, nsvnIPTim III.E.M-26-8 194001K105 3.1/3.4 GIVEM GINTAllMENT PAMMETERS, DETEMINE IF CONTAlifENT ENTRY IS PEMITTED.

PIB-ZP-TK-007 194001A116 3.1/4.4 CLASSIFY / RECLASSIFY 906DICY EVENTS IEQUIRIIE BEIEENCY PLM IIFLEMENTATIONS. (GSEPCLAS$1FICATIONS)

Sl###RY CREW F: 10 EXAMINATION TOTALQUESTIONS=18 TOTAL VALIDATED Tilf = 91 MI,NLITES 1 .

l TRAINING DEPART!ENT ONLY Page 2 1

lEXAMINEFIN PROFIIE ;

EXAM TII12: SRD EXAMINATION EXAM NUMBER: 20-N-tuR-03-0043-89 i IOCATION : 20-N-fGR-0 EXAM HIS'KRY INS:R. ID: 20-IN-DRA REQUESIED ACIUAL DATE T EKAM: 04/20/ 'IUrAL ICINT VAWES: 18.00 18.00 TYPE T EZAM: M 'IUIAL NGEER T QUESTICNS: 18 18 l ANSWER FORM ID: 'IUIAL DIFFICULTY IEVEL: 54.00 54.00

'IUIAL TIME 'IO CCMPIETE: 91 NLMBER T TINES GENERA'IED: 1 NWEER T TINES USED: 0 NGEER & PARTICIPANTS: 0 ('IUrAL)

N(MBER T PARTICIPANTS: 0 (LAST EXAM)

F1-Help F2-Field Info F1061groff ESC-Return

~

m ,. - m-n r m -- .  ;-

Ngr ..~c, _,

,a*

a e

• IEEPCRP R2.4 SIO EIANDOEIW '

PME: 1 EKAKDerIm RET DME: 04/12/99 EKAN NLDEER: 20-N-fGR-03-0043-89 IIX2 rim 20-+H30ut-03 EKAN CXXNT 1 POINT VALIE  : 18.00 TOTAL QlESTIWS : 18

'1 URAL DIFF IEVELt 54.00 PT. VAWE EEHG: 20-020216 IIICRr06B (1.00) 1. BM RY-5, Drawing a Bubble in the Pressurizer, requires that the pressurizer be heated to a tanparature of 445'F before the reiction in charging flow la commenced, hat is the basis of this tamparature ,

requirement? 1 i

ANDER: This ensures that the PER is at saturations '(1.0 pt).

-or-This ensures a PZR bubble will fans (1.0 pt).

-or-445'F is equal to the saturation temperature for 400 pela/385 psig

(* 15 psi) (1.0 pt).

References:

T/S 3.4.9.2 EB-NLM: 20-020256 IIICCV01A

) (1.00)

2. BwCP CV-1, Startup of CVCS, cautions that the letdown orifice isolation valves shall remain open when the plant is solid. Explain what added protection is provided by keeping these valves open ln this condition? ,

ANSHER: The Istdown Line Relief valve (600 psig) is available (1.0 pt).

References:

BWOP CV-2 M-64 S l

..w-

IEFGtr R2.4 50 EEAMDUEFIGl PAfE: 2 EIAMDGEFIGI 3Glf DME: 04/12 AJ

' PT. VAUE I EIHUt 20-020067 IIIDAF01A (1.00) 3. A check of TI-AF126 (local Amilia F*arktar pump discharge tmperature) 30 airartes after s down the 1A >=iliary Pa=rhter pump revealed that the taugurature has increased to 150'F.

A. Wiat could cause the temperature to increase?

B. Wiat are two possible adverse effects if the >=iliary Fearkter ptmp was started?

MUER: A. Check valve back leakage, (.5 pts). ,

B. Stammt bindirugdamage of Amiliary Feedwater pumps (.25 pts), and hterhauuuur/demage of AF piping (.25 pts).

)

i l

Peferences: BwCR SEE-7 EEHKM: 20-020270 IIIO606A (1.00) 4. Describe two sottods of isolating steam flow through the SG PGMs if they cannot be operated from the Control Ebein. (1 pt.)

l ANSER: (Any two .5 pt. each)

A. 16019kRCD) .

B. Iocally close the Incal Manual PORV of operaticm Isolation valve (Pump the Hydraulic / Hand Pump to s} mat the PCRV.

C. Icosen the metacrew on the Hand Pump (steam pressure will close the ve).

D. Tev'al centrol at the RSP to shut the PORV.

References:

BwOh Pri-5 P & ids

..uw

m g- g g y PT. VAU E EEHEM: 20-020086

.IIIO 906B

'(1.00) 5. thit 1 is operating at 97% power, when a safety valve on the 1A steam ganarator fails open. How nuch must turbine load (in pu. void.) be ramped back in order to mutintain reactor power at 97%?

A. 2%

B. 4%

C. 6%

D. 8%

MGER: C. 6%

(1.0 pt)

References:

Technical Specification Bases 3.7.1 j EB-M E: 20-020134 i IIICSIO5A1 (1.00) 6. In BwCP-SI-5 Raising SI h-'lator Inval with the SI punps, there

} is a caution that says wtan using the "A" SI pap, the discharge to the cold leg isolation valve (SI-8821A) can be left open if the RCS pressure is > 1700 psig. Mut are the adverse conseguances of being ln Made 5 and solid while performing this p.vc dare without closing this valve?

MGER: Water would be p H into the RCS f.5 pts) and

. potentially overpressurize the RCS/h RH reljefs or PZR PCRV

(.5 pts)

References:

BwCP SI-5

..vu l

PT. VAIAE EB-IUi: 20-020286 IIIO5W D (1.00) 7. The ICS is at 345 psig and 340'F. Shortly after placing the "A" Train of NR in shutdown cooling mode, the N90 notices ICS wide range pressure and PER level decreasing. The Equipent Attendant informs him that the B RH suction relief valve is lifting.

Wtich one of the fo daar ribes the cease of this event?

A. The ISI8809B, M 'mN to RCS Cold Iag Injection valve, was inadvertently closed.

B. The Ni Diach. Header X-Tie valve,1RX8716A, was not closed prior to starting the "A" Ni pump.

C. The Imtdtnet Pressure Cor-trol valve, PCV-CV131, was left in Mf!O for the "A" RH pap start.

D. The FC Icop A to Mi Pr A Suction valve, INW701A, inadvertently closed dias to an instament failure.

AM9E:R: B. The NI Disch.14aadar X-Tie valve,1RH8716A, was not closed prior to starting the "A" RH pump. (1 pt.)

}

References:

BwCF RH-6 P&ID M-61 arxi M-62 EB-NLM 20-020275 IIIDOA09D (1.00) 8. Unit 1 is in FA 5. The Wtit ISO manually trimad the 1B Ni punp l because of fluctuating motor amps armi pump flow near zero. Wlde range ICS pressure Indicates 325 psig on 1PI405 and CE7-SCAIE HIGi on 1PI403. Detenmine dat operator actions, other than startiry the 1B Mi pump, are required to restcace the "B" train 151 loop to service?

AN5tER: Open 1RH8702R/perfoon step 4 of PRI-10 (Ioss of Ni Cooling) (1 pt) 1

References:

BwOA PRI-10 Technical Specifications 4

..~

• ~ ~ -

EDempirfd EET tWEE: 04/12/99 I

PT. VAIDE EEHELM: 20-020068 IIIElWO4A (1.00) ^w has received 2.0 Run for the , ircluded in this total

9. An is Namen for the current week. 'Ibday had received an additional 80 semmt under a Type I 19@. Derar=he all administrative exposure control liinits the operator M. (Assume no achainistrati's )

actions were tahan.) (. Assouc z o o m g. ;4 iF ra i ND*  ;^ 66 used. e4 L' QTk AM9ER: He M the dall expecure lind.t of 50 meem (without 3 suparvisor's ) (.5 pts), and the weekly limit of 3 300 naam (

(without en Sgarvisor apptwal). (.5 pts) l

References:

BwRP 1000-Al B dP 1100 EB-NtM: 20-020320 PIBZP25015 (1.00) 10. Unit 1 is in Mode-1, the 1A diesel y. w .b,r has been 006 for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and is w i.ed back in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 'the Urt't 1 SIB 811B ffrain j Containment treaker tripped and ascizculation the thanuals Susp will notIsolation reset. Fbrty- Wdve)five g minutee after the breaker tripped Electrical Maintenance reports that the limitorque notar must be replaced. Estimated time for notor replacement is three days. Evaluate for the miniman GSEP classification and appropriate EAL, if any.

ANSWER: Unusual Event (.5 pts) condition 3a/6a (.5 pts) (a ICO requires a shutdom) .

References:

BwZP 200-1A1

--nore.. -

jg g; - g i Pr. v m E j EB4Est 20-020379 IIIIEF01B (1.00) 11. A hi$ PER pressure reactor trip has occurred on thit 1. Shortly after the trip a low PER pressure SI' actuates. Wtile ru.' Q step 26 of STEP 0 i REACIOR TRIP / SAFETY IIC1lCTICBI). The following conditions exist.

Ibth PER PORVg- cr m n indication One PER safety ved - OPEN indication.

RCS pressure- 1700 psig and decreasing RCS tasperature- 557'F and slowly decreasing PRP pressure- 16 peig and increasing

^

PRP t .-- - uu.& 198'F and increasing Based on the above information the SIO states that he is going to transition to BuGP-1 at this point. Explain Way you disagree with the SRO's statssent..

ANDER: He should transition at step 40/ PRE conditions (1.0 pt) (dure safety valve status is evaluated).

References:

BtEP-0 REACICH 'IRIP CR SAFEIT INJilCTIGI EB-nim: 20-020125 IIIEAM18D (1.00) 12. Ckt March 5,1988, the thit 2 ISO instructed an e^uic to unlock and renove the chain from 2 SIB 921B (2B Safety In on Pump Mamal for Mechanical Maintenance. Iatar that day, a -

Discharge B-nan alsoValve)d locke closed the valve by mistahm. The error was discovered on March 13, 1988 during rounds. Mutt operator actions could have p. ._ dad both adstahms frost occurring? (Two answers are required.)

AIGER: (Any 2 .5 pts. each)

1. Proper use of the Camponergt Abnormal Position Ing.

2. Proper une of repeat backs.

3. Independent verification of valve.

Refererres: BdP 340-2 B dP 330-3 b[!lR 8

-sme--

~ - ~ ~

M M i 04 N M

,- 4 PT. VAUE EEHet 20-420435 IIIEMO6B (1.00) 13. A contairemnt entry is planned for routine troubleshooting of an 1 RCP StMipe Fill valve, inside the missile barrier. The thit is at 75% power with a load increase of 0.5 MKAnin in progress up to 95% power. All other plant and contaiment oorditions are Irw=1 Wry wouldn't you authorise this entry?

( N reasons required)

MGER: 1. At power levels > 404/At this power level (.25 pts), entry into the missile barrier should be restricted to mentguncy entries only (.25 pts)

2. The reactor naast be operating at a stable power level /75% power

(.5 pts) (Stable power is defined as i 2%). '

-tr-Ioad increase is in progress (.5 pts)

References:

BuhP 1450-1

.. u.e

. -- ~~ -

p g; gg -

4 PF. VAIAE EIHUh 20-020127 IIIIAH180

, (1.00) 14. With BMP-1 in effect, the SCRE nonitoring the Critical Safety Functions rufas.the following indications:

Reactor Power is 0%

SLR is -0.2 IFM on both IR and SR channals Containment Pressure is 20 psig Core Exit O----:egles all between 350'F and 450'F Care Exit E- --egles murage of 10 highest is 410*F RCS Pressure is 480 psig Narrow Range S/G Invals A = 27% B = 30% -c = 22% D = 20%

Feed Flow to S/Gs A = 0 gua B = 0 gut C = 0 gua D = 0 gut Aux Food Flow to S/Gs A=0ga B = 0 gr C = 0 gun D = 0 gan RCS Cbid Img T_.g htres A = 390 F B = 400 F C = 385 F D = 380 F Mutt Burr sust be implemented?

ANSER: BWFR S H.1 (Response to Ioss of Secondary Heat Sink) (1.0 PT).

References:

B dP 340-1 Bd5T 1, 2, 3, and 4 BWFR H.1 l

)

..at o l

l M r

==morra. = o-:wum rr. vauE EDi e s 20-020327 IIIERIES (1.00) 15. thit 2 is in axle 3 at NOTMP preparing to enter Mode 2, idien the Pressurizer Par Pressure pan andDISCHARE I 1 i- 'IIBGENmEE HIGH Anran,iciator a1.o acto.t.. actuates.

= Pa : indicat.

closed. Both PGW +=ilpip= temperatures are 250*F and s10wly increasing. A11 safety valve tailpip= tasperatures are 190'F and steady. Wiat action (s), if arty, are required to contiran operations in mde 37 -

NSDER: Within one hour (.5 pts) restore the PGMs to ===hilityAshut both block valves (.5 pts).

-cr-Enter ICOAR 4.4-la (1.0 pts)

Referemoes: 3.4.4 EEHm: 20-020419 IIIEAM)SA (1.00) 16. Unit 1 is at 100% steady state power. Mtile m Ivu.d.ng a shift surveillance the Nio records the following readings:

Channel 455 PZR Pressure = 2190 ps g Channel 456 PZR Pressure = 2195 ps Channel 457 PER Preemxre = 2195 ps g Channel 458 PER Pressare = 2190 ps Mint, if any, actions should be initiated pertaining to these readings?

MEi6ER: Enter (1@) lOOhR for DB parameters / Enter ICORR 2.5-la (1.0 pt).

-or-Restore Pzr to within its limit within two hours (1.0 pt)

(or redme 1 power to <5% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

References:

BwoS 0.1-1,2,3 IER 88-020 l

I anwJLv i

~

INKRP R2.'4 50 EUNDUEIN PAfE 10 EUNDUEIGE EEY DME: 04/12Ml9 !

o' PT. VAILE EIMUt 20-020148 IIIEAM14L (1.00) 17. Detandne the pcoferzed method of inkperr' M verification charing the performance of 1BwCS 7.4.a-1, (kilt 1 Esot .M Service Mater System Monthly surveillance for:

A. ISK46A, IA SK Rocza and IAabe Oil Cizs HK Outlet valve, was found closed.

D. 1MN-SK016A, RCFC Fan Coolers IA & 1C SK inist was fourri closed, with la E 1C the only RCFCs running.

ANSMER: A. Second w a^A (.1 pts) locally checking the valve throttled

(.4 pts)

B.

^

Second %.w (.1 pts) remotely 0- ving the open indication light /wu'uc1 power (.4 pts).

-or- .

Second % . w (.1

^

) locally check valve open via stam positiorV11mit swi (.4 pts). ,

1 i

References:

BwAP 300-1 EIHRE: 20-020171 IIICRC01C (1.00) 18. You have been directed to stablime reactor power at 20% following a rapid load decrease from 100%. 'Ihe resultant Xe build up has caused the P-12 BYPASS / PERMISSIVE to alarm. Delta-I has been driven out of the target band. You are diluting at the nnximasa rate in alternate dilute, rods have been withdrawn to 228 steps and the stone &ses are closed. Explain the basis of your current actions, include applicable  ;

time restrictions, if any apply.

AlGER: Restore Tavg to > 550'F (.5 pts) within 15 minutes (.5 pts).

-or-Enter ICORR 1.1.4-la (1.0 pt)

References:

T/S 3.1.1.4 DVR 20-1-88-216

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)

7-

s. .

.h

• CKW F: M EMMINATim EPORT BRL!!niON) WCLEAR POWER STATION TMW K/M IF DESCRIPTim Ill.B.CV-01-D 0040204401 3.8/3.3 GIVEN A COPY OF THE CYCS STARTUP PROCEDURE DISCUSS HOW TO CARRY OUT CRITICAL OR DIFFICULT STEPS M THE EXPECTED PLMT RESPONSE WILE PERFDMING THESE STEPS. (BORAT10N)

III.C.CV-01-A 004020k602 3.8/4.1 GIVEN A COPY OF THE APPROPRIATE PROCEDURE (S) DISCUSS HOW TO LINEUP THE CVCS FOR VARIOUS OPERATIONS. (SOLID PLMT OPERATIONS]

Ill.C.MS-06-A 035010K602 3.1/3.5 DISCUSS THE OPERATION OF THE STEM GENERATOR PORVs INCLUDING BOTH MANUAL, LOCAL A2 AUTOMATIC OPERATION.

Ill.C.MS-06-B 000040EA108 3.6/3.7 DISCUSS THE CAPACITY OF THE STEM GENERATDR SAFETY VALVE.

Ill.C.RC-01-C 0020000011 3.3/4.0 GIVEN A SET OF PLMT COW!TIONS OR PAMMETERS INVOLVING THE EACTOR COOLANT SYSTEM AND THE APPROPRIATE SECTION OF TECH. SPECS., DETEMIME TECH.

SPEC. COMPLIANCE AE REQUIRED ACTIONS.

li f.C.M-03-0 00$000K411 3.5/3.9 GIVEN THE APPROPRIATE PROCEDURE, DISCUSS HOW TO PLACE THE RH SYSTEM IN THE SHUTDOWN COOLING MODE OF OPERATION.

Ill.C.RY-06-B 011000K510 3.7/3.4 GIVEN THE APPROPRIATE SECTIONS FROM POCEDUES THAT APPLY TO DRAWING A BUBBLE IN THE PRESSURIZER: DISCUSS THE BASES FOR EACH PRECAUTION, LIMITATION AND ACTION.

Ill.C.S!-05-A-2 006000A202 3.9/4.3 DISCUSS T11E BASES FOR EACH NOTE AND CAUTION. (LDSS OF FLOWPATH)

I!!.C.5M-01-A 035000G013 3.4/3.5 GIVEN A COPY OF THE APPROPRIATE PROCEDURE, DISCUSS HOW TO PERFOM SG SYSTEN LINE-UPS. (H/U OF MAIN STEM LINES)

!!!.D.AF-01-A 061000A206 2.7/3.0 GIVEN A COPY OF THE AUXILIARY FEEDWATER CHECK VALVE LEAKAGE PROCEDURE, DISCUSS THE BASIS OF EACH STEP, NOTE OR CAUTION IN THE PROCEDURE.

(BACKLEAKAGEOFMAINFEEDWATER) lli.D.EP-02-C 000008G012 3.9/4.1 GIVEN A SET OF PLMT CONDITIONS OR PAMMETERS INDICATING A VAPOR SPACE LDSS OF C00LMT ACCIDENT, AND A SET OF PLMT PmCEDURES, IDENTIFY THE CORRECT P OCEDURE(5) TO BE UTILIZED A 2 DISCUSS REQUIRED OPERATOR ACTIONS.

(STUCK OPEN PZR SAFETY) 111.D.0A-09-0 000025G011 3.6/3.9 GIVEN A SET OF PLMT COEITIONS OR PARAMETERS INDICATING A LOSS OF RH COOLING A2 A SET OF PLMT PROCEDURES, IDENTIFY THE CORRECT PROCEDURE (S) TO BE LTTILIZED AND DISCUSS REQUIRED OPERATOR ACTIONS.

Ill .E.M-01-0 194001A103 2.8/3.4 DISCUSS THE REQUIREMENTS FOR M ON-COMING NSO TO READ THE CD/ UNIT LOGS AS STATED IN THE SHIFT TURNOVER PROCEDURE.

I!!.E.M-08-A 010000G011 3.2/3.9 GIVEN A SET OF PLANT CONDITIONS OR PARMETERS DETERMINE IF ENTRY INTO A TECHNICAL SPECIFICATION ACTION STATEENT IS REQUIRED. (LER 88-020) l lli.E.M-08-B 000008G008 3.1/3.6 GIVEN THE APPROPRIATE SECTION OF TECHNICAL SPECIFICATIONS AND A SET OF PLANT CONDITIONS OR PARMETERS INDICATING A POSSIBLE LCD VIOLATION, DETERMINE I TECH. SPEC. COMPLIANCE AND REQUIRED ACTIONS.

l TRAINING DEPARTENT ONLY Pqe 1 l

l 1

i CEV F: RD EXMIETION EPORT SmtDWom) INCLEAR POWER STATION TMW K/M DP DESCRIPTIM III.E.M-14-C 001000G001 3.7/3.8 DISCUSS THE REQUIREMENTS WHICH MUST BE MET FOR AN INDIVIDUAL TO MANIPULATE CONTICLS WHICH DIRECTLY OR INDIRECTLY AFFECT THE REACTIVITY OR POWER LEVEL OF THE REACTOR.

!!!.E.M-14-L 194000K101 3.6/3.7 DISCUSS THE ACCEPTABLE ETHODS OF PERf0 MING M INDEPDOENT VERIFICATION AND THE PROPER APPLICATION OF EACH.

!!I.E.M-18-D 194000A110 2.9/3.9 DISCUSS P10PER USAGE OF PLMT PROCEDURES. (LER 88-001)

!!!.E.M-21-A 194001A106 3.4/3.4 DISCUSS THE ADMINISTmTIVE CONTROLS ESTABLISHED FOR mp0 LING LONG TEM AMUNCIATORS OVE TO THE FOLLOWING: M ALAM DUE TO PLANT CONDITIONS.

Ill.E.M-26-B 194001K105 3.1/3.4 GIVEN CONTA! MENT PAIMETERS, DETEMINE IF (DNTAIMENT ENTRY 15 PEMITTED.

SUPf%RY CREW F: R0 EXAMINATION TOTAL QUESTIONS = 20 TOTAL VALIDATED TIE - 95 MINUTES I

l l

l l

TRAINING DEPARTENT ONLY Page 2 l

I

DHLL CDRP EUMWEL IRAIDNOCD l EUMDUTICM PIOFIIEl .s .

EDM TIILE: 10 EDMDUTICN EU M NLMBEA: 20-N-MIR-03-0042-89 IOCATICN : 20-N-NCER-0 EDM HISitRY D61R. ID: 20-IN-DRA REQUESTED ACIUAL DATE & EUM: 04/20/ TUTAL POINT VAIUES: 20.00 20.00 TYPE T EUM: M TUDL NLDEER T QUESTICNS: 20 20 l

ANSER FWM ID: TUIAL DIFFICETY IEVEL: 60.00 60.00 IUDL TDE TO CIMPIRIE: 95 i NLMBER T TDES GEERA1TDs 1 1

HLMER T TDES USED: 0 N(DEER T PARTICIPAN15: 0 (TUDL)

NLDEER & PARTICIPANTS: 0 (IAST EUM)

F1-Help F2-Field Info F10-Signoff ESC-Return

~

1. '[.

Z.~ %s'e r T 7 5*

i

. IEFGU R2.4 K) EDMIlptrIOf PAGE: 1 EDHDetrICM KET int 1E 04/12/99 EKAM MMER: 20-N-N2R-03-0042-89 IOCitrIO4 20-N-KIR-03 l EDM CXXMT : 1 POINT VAM E  : 20.00

'IUrAL QlESTICNS : 20

'IUPAL DIFF IJM!L: 60.00 PT. VAHE EB-NtM: 20-020216 IIICRr06B (1.00) 1. BwoP RY-5, Drawing a Bubble in the Pressurimr, requires that the pressurizer be hasted to a tenparature of 445'F before the rocketion in charging flow is cannenced. Met is the basis of this tasperature requirement?

ANSM!R: 'this ensures that the PZR is at saturation. (1.0 pt).

e

'Dtis ensures a PZR bubble will fom (1.0 pt).

-or-4450F is equal to the saturation temperature for 400 psia /385 psig (f 15 psi) (1.0 pt).

References:

T/S 3.4.9.2 EB-MM: 20-020256 IIIOCV01A (1.00) 2. BwCP CV-1, Startup of CVCS, cautions that the letdown orifice isolation valves shall remain open when the plant is solid. Explain what =**M protection is provided by loseping these valves open in this condition?

ANSM!R: 'Ihe Istdown Line F311ef Valve (b'M peig) is available (1.0 pt).

References:

BwT CV-2 M-64-5

...a-e-m_ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . . _ _

. REPCRT R2.4 PO EDMD9ff1W PACE: 2 EDMDutrIW KEY DPE : 04/12/99 PT. VAUE EB-N(M 20-027511 IIIBCY01D (1.00) 3. If BCS boaton vu witration is 700 ppm, what is the required setting on the potentiometer for the Boric Acid flow controller (_FK-110), to keep RCS boron concentration constant? (1 pt.)

ANDER: 3.0 i 0.05 turns. (1 pt.)

References:

BwCP CV-1 DwCP CV-7 BwGP 100-Al EB-N(M: 20-027460 IIICSIO5A2 (1.00) 4. Mtile perfoming BwCP SI-5, Raising SI h=$1ator level with the SI punps, explain how using the " B" SI to fill the SI accumulators with the plant at 3600F and 1350 peig m .Au both SI trains IIq mrable?

AIGER: Shutting SI8835 (.5 pts) isolates both SI trains (.5 pts).

References:

BwCP SI-5 P&ID M-61, Sheet 3 i

... a i

L_ _ _ _ _ _ __ ___ --

EUMDGEICM KEY [WrE 04/12/99 PT. VAUIE EEHG: 20-020286 IIIClum3D (1.00) 5. 'Ihe RCS is at 345 psig ani 340'F. Shortly after placing the "A" Train of NHR in shutckwt cooling node, the IGO notices RCS wide range pressure and PZR level decreasing. 'the mytir==rit Attendant

, informs him that the B RH punp suction relief valve is lifting.

Mich one of the following best describes the cause of this event?

A. 'Ihe ISI8809B, RH 'IRN to RCS Cold Iag Injection valve, was inadw=.6Lly closed.

B. The RH Disch. liaariar X-Tie valve,1RH8716A, was not closed prior to starting the "A" RH punp.

C. 'Ihe Istdown Pressure Control valve, PCV-CV131, was left in AUID for the "A" RH punp start.

D. 'Ihe RC Icop A to RH Pusp A Suction valve, IRH8701A, inadvertently closed due to an instrument failure.

ANSER: B. 'Ihe RH Disch. Header X-Tie valve,1RH8716A, was not closed prior i to starting the "A" RH punp. (1 pt.)

References:

IWOP RH-6 P&ID M-61 and M-62 EB-NLM: 20-020067 IIIDAF01A (1.00) 6. A check of TI-AF126 (local Auxil INudwater punp discharge t%et.ure) 20 minuteo after shutt cbun the 1A Anrilia71%edwater punp revealed that the temperature has incraamari to 1500F.

A. Wat could cause the t%etum to iru,..;;?

B. Wat are two possible adverse effects if the Auxiliary Feedwater pump was started?

AEWE:R: A. Check valve back leakage (.5 pts).

B. Steam binding / damage of Anrillag Feedwater ptmps (.25 pts), and Waterhanener/ damage of AF piping (.25 pts).

References:

BwCA SEC-7 ouw-

' REPWT R2.4 30 EEAMDOEIN PME: 4 EKAMDUTICM KEY DMtE: 04/12MF)

PT. VAUE EB-MM 20-020086 IIIO606B (1.00) 7. Unit 1 is operating at 97% power, when a safety valve on the 1A steam y:= =.=ior fails open. Ibw nuch must turbine load (in pw.u a.) be ranped back in order to naintain reactor power at 97%?

A. 2%

B. 4%

C. 6%

D. 8%

ANSWER: C. 6%

(1.0 pt) l

References:

Technical Specification Bases 3.7.1 EB-NCE: 20-020270 IIICM506A (1.00) 8. Describe two methods of isolating steam flow through the SG PORVs if they carrot be operated from the Control Room. (1 pt.)

AtEWCR: (Any two .5 pt. each)

A. DE019A/B/C/D).

B. Incally close the Incal Manual PORV of operation Isolation Valve (Punp/Nand Punp to shut the Hydranifr-the PORV.

C. Ioosen the setscrew on the Hand Punp (steau pressure will close.the ve).

D. Tne al sud.i.vl at the RSP to slutt the PCRV.

References:

BwCE Pri-5 P & ids

... a u--

u

. IEFGtr R2.4 !O EEAltDstrIGE PME: 5 EKANDetrIGE REY DittE 04/12/99 PT. VAUE EI H e t: 20-020379 IIIGW015 (1.00). 9. A high PER pressure reactor trip has occuzzed on thit 1. Shcrtly after the trip a low PER pressure SI actuates. Mile WIuudng step

- 26 of BEP 0 qHEACitR ' TRIP / SAFETY DUNCTION). 'the following conditions ex;st.

Both PER PG Ws- nmnyn indication One PER safety valve- CFEN indication.

K:S pressure- 1700 poig and decreasing MS ta g arature- 557'F and slowly decreasing Pltr pressure- 16 poig and increasing Pftr * .-_- -i.u v 198'F and increasing Based on the abme information the SFO states that he is going to transition to ad:P-1 at this point. Explain why you disagree with the SE's statement.

ANSER: He should transition at stap 40/PRT conditions (1.0 pt) (there safety valve status is evaluated).

Refermicos: BWEP-0 RimCitR 'IRIP GL SAFETY DUBCTIGE EB-MM: 20-020102 IIIEAN14C (1.00) 10. A RO trainee is withdrawing control rods during a reactor startup while the thit N90 is nonitoring various parameters on the plant computar. mat artninistrative violation, if any, has occurred?

ANSER: A trainee must be under direct ion (1 pt). (W an a trainee is WIvudng any control , the IU/SRO must observe the neoa==ary indications as i he perfnemari the manipulation himself.)

References:

B dP 300-1

.. u.e-0

. REPCRP R2.4 IO EKAMDUtTICM PACE: 6 EKAMDerICM Erf IRIE: 04/12/99 PT. VAU E EB-NLM: 20-020125 IIIUAM19D (1.00) 11. Ch March 5,1988, 'Je Unit 2160 instructed cc tor to unlock ard renove the chain fra 2SI8921B (2B Safety In on Pump Manual for Mechanical Maintenance.

Discharge B-nan also Valve)d locke closed the valve by mistake.Later that day,

'the error was a discovered an March 13, 1988 chiring rounds. Wat operator actions could have prevented both mistakes from occurring? ('no answers are required.)

ANDER: (Any 2 .5 pts. each)

1. Proper use of the Cuwa=d. Atromal Position Iog.

2. Proper use of repeat backs.

3. Ir@d-d. verification of valve.

References:

B dP 340-2 B dP 330-3 SPecial Operating Orders IER 88-001 EB-MLM 20-020128 IIIEAM21A (1.00) 12. thit 2 is in Made 1, operating at 100% power, when the FWIV HYD/P!EU PRESS ION annunciator alams. A *B" nen investigates erd reports that the hydraulic and pneumatic sures are sat.isfactory. After initiating a work request, t acti m should the Unit 2 PSO take to ensure the oncoming NSO is aware of the status of the FSIV HYD/P!EU PRESS IDW annunciator?

MEMER:

Place a work request sticker on the annunciator wirrkw (.5 pt.) and include the annunciator status in the shift turnover /t:ontrol board walktkwyturnover sheet (.5 pt.).

References:

B dP 380-2 B dR 0-39-A2

.. m.-

REPORP R2.4 10 EDHDUEIN PME: 7 EDHDUEIN REY ' DNEE: 04/12/89  ;

4

=

PT. VAME EEHOtt 20-020435 IIIEAM26B (1.00) 13. A Cbntairment entry is planned for routine troubleshooting of an HCP Stan@ipe Fill valve, inside the missile barrier. The thit is ya up at 75% power.

to 95% power with a load All other imta plant of contalment and 0.5 PMAninoorti in ftions are nonnal.

Wry wouldn't you authorize this entry?

('No reasons required)

ANSNER: 1. At power levels > 40%/At this power level (.25 pts), entry into the missile 1=rriar should be restricted to maargency entries only (.25 pts)

2. The reactor must be operating at a stable power level /75% power

(.5 pts) (Stable pomar is defined as i 2%).

-or-Ioad increase is in progress (.5 pts)

References:

B dP 1450-1 EB-NCE: 20-020244 IIIEAM01D (1.00) 14. Determine the requirements for the oncoming 100 to read the logs associated with his duties:

Today's Date: 01/06/89, Shift 1 Last Date on Shift: 12/26/88, Shift 2 ANSWER: Four days / 01A2A9 (1.0 pts).

l

References:

B dP 335-1 l autJL u l

RE30 T R2.4 10 EUNDetrICN PME: 8 EUNDUtrICM IGT DR5E 04/12/89 PT. VAIIE 1

EEHUt 20-C20148 IIIERN14L (1.00) 15. Detacmine the preferred method of independent verification durirg the perfammence of IBwCS 7.4.a-1, thit 1 Essential Service Mater Systen Mxsthly Surveillance for A. 1SK46A,1A SK Room and IAJbe Oil Clrs HK Oatlet valve, was found closed.

B. IMN-SK016A, RCPC Fan Coolers 1A E 1C SK inlet was found closed, with 1A & 1C the only RCPCs running.

ANEE R: A. Second operator (.1 pts) locally checking the valve throttled

(.4 pts)

B. Second w ter e (.1 pts) remotely observing the open indication light / w h 01 power (.4 pta).

-or-Second w e ^wi- ( .1 ) locally check valve open via stam positiorV11mit swi (.4 pts).

References:

B dP 300-1 ,

EB-NLM: 20-020419 IIIEMOBA (1.00) 16. thit 1 is at 100% steady state power. Wtile mfvuning a shift surveillance the N90 records the following readings:

Channel 455 PZR Pressure = 2190 ps Channel 456 PZR Pressure = 2195 ps g Channel 457 PZR Pressure = 2195 ps Channel 458 PZR Pressure = 2190 peig mat, if any, actions should be initiated partaining to these readings?

AIGHE:R: Enter (10puwk) IIIRR for [tB parameters / Enter ICORR 2.5-la (1.0 pt).

-or-Restcre Pzr e to within its limit within two hours (1.0 pt)

(or redtace power to <5% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). 7g- gg

References:

BwoS 0.1-1,2,3 IER 88-020

..vu,-

- 'HERRP R2.4 10 EKAMDERrICM PXE: 9 EKANDENTICN RET DME: 04/12/1l9 PT. VAWE EE H u ta 20-020327 IIIslee8B (1.00) 17. thtt 2 is in Itxie 3 at M7FAECP preparing to enter mde 2, dann the Pressurizer PCRV DISCHARGE 'IEMPERMURE HIGi Annunciator actuates.

PRT Pressure and Invel alaans also actuate. Both PORVk indicate closed. Both PCRV tailpipe temperatures are 250'F and slowly increasing. All safety valve tailpipe tenparatures are 180*F and steady mat action (s), if any, are required to crritinus operations ANSWER:

Within one hour block valves (.5(.5 pts)Th 3, '/49 the PCRVs to r==rnh111ty/ shut both pts).

w Enter I4XER 4.4-la (1.0 pts)

References:

3.4.4 EB-NtN

20-020275 IIIDOh09D (1.00) 18. Unit 1 is in mde 5. 'Ihe Unit NSO manually tripped the IB M pump because of fluctuating noter anps and punp flow near zero. Wide range RCS pressure indicates 325 peig on 1PI405 and 0FF6 HIGI on 1PI403. Detennine what operator actions, other than starting the IB M punp, are required to restore the 'B" train NH loop to service?

ANSWER: Open 1RH8702B/ perform step 4 of PRI-10 (Ioss of RH Oooling) (1 pt)

References BwCR PRI-10 Technical Specifications

.' , REPCRP R2.4 10 EIAMDRFIOt PAGE: 10 EEAMDUfFIGI NtY DME: 04/12/99 Pr. VARE EB-10(: 20-020221 IIIC901A (1.00) 19. 'Ihe MIV bypass valves am about to be %M to begin a Main Steam System heatup. Wat is the MINDEM time allowed to equalise pressum across the M IVs when main steem hem 4 p m ssure is initially 100 psig and steam generator pmssure is constant at 1000 psig? (Assume the steam lines are wanned up at 100'F/hr).

AE NCR: 2_.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (2.2 to 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).

(Includes 30 minutes wait period after EIV bypassa are crackad, plus approximately two hours to heat flun 100 psig to 950 psig [can open E IVs with 50 psig & ]) (1 pt.)

References:

BwGP 100-1 BwaP E -9 DVR 06-1-87-066 EB-MM: 20-020171 IIIOC01C (1.00) 20. You have been directed to stablize reactor power at 20% following a rapid load decrease from 100%. 'Ihe resultant Xe build up has caused the P-12 BYPASS / PERMISSIVE to alarm. Delta-I has been driven out of the target band. You are diluting at the neximum rate in alternate dilute, rods have been withdrawn to 228 steps and the steam dunps are closed. Explain the basis of your current actions, include applicable time restrictions, if any apply.

ANSWER: Restore Tavg to > 550'F (.5 pts) within 15 minutes (.5 pts).

-tx.-

Enter IOOAR 1.1.4-la (1.0 pt)

References:

T/S 3.1.1.4 DVR 20-1-88-216

-End of Print-