IR 05000245/1989026
| ML20006C272 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/25/1990 |
| From: | Conte R, Easlick T, Florek D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20006C271 | List: |
| References | |
| RTR-NUREG-1021 50-245-89-26OL, NUDOCS 9002070243 | |
| Download: ML20006C272 (66) | |
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U.SiNUCLEARREGULATORY' COMMISSION i
REGION I
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REQUALIFICATION PROGRAM EVALUATION..AND E0P/BWR POWER OSCILLATION INSPECTION
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- Combined. Report NO.
50-245/89-26 (OL)
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Facility Docket NO.
50-245
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Facility Licence NO.
DPR-21-q
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License:
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Northeast Nuclear Energy Company
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P.O. Box 270
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Hartford, Connecticut. 06141-0270 Facility:-
Millstone, Unit 1
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Examination Dates:
October 16-21, 1989 and October 23-26, 1989-l Examiners:
D. Florek, Senior Operations Engineer
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T. Eas11ck, Operations Engineer
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M. Morgan, PNL
- G. Buckley, PNL B. Orton,.PNL
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LRequalification Examination Chief Examiner:
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Theodore A. Eas11ck, Operations Engineer Date-
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. Replacement Examination Chief Examiner:
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l jufl]. d?e ikr/9a 06naldJ.FlorgK,Sr.OperationsEngineer
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. Approved By:
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Richard J. Cont y Chief, BWR Section Date
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Operations Branch, DRS
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EXECUTIVE.SUMHARY Written and operating examinations were administered to five Reactor Operators (R0s) and fifteen Senior Reactor Operators (SR0s). These operators were j
divided into five crews, which consisted of three operating and two staff
crews. The examinations were graded concurrently by the NRC and the facility
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training staff. As graded by the NRC, all five R0s and fourteen of the fifteen SR0s examined passed all portions of the examination. One SR0s did
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- not perform satisfactorily on the simulator evaluation as graded by the NRC and the facility. He passed the remaining two portions of the examination.
L All five crews that were evaluated performed satisfactorily as graded by the NRC,
'The licensee's licensed operator training program was determined to be satisfactory based on the criteria established in section ES-601 of NUREG-1021, Rev. 5.
Written and operating replacement examinations were administered to two Senior Reactor Operator applicants. In addition operating tests were administered to one Reactor Operator applicant and one Senior Reactor Operator applicant. All applicants passed these examinations.
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DETAILS
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1 Introduction During the examination period the NRC administered requalification examinations to 20 licensed operators (5 R0s and 15 SR0s).
Three operating crews and two staff crews were evaluated. The examiners used the process and criteria described in NUREG 1021, " Operator Licensing Examiner
Standard," Rev.
5., section ES-601, " Administration of NRC Requalification Program Evaluations." Additionally, the NRC administered replacement examinations to one reactor operator and three senior reactor operators.
The R0 received an operating examination only. One SRO received a simulator examination only. Two SR0s received a written and operating examination.
An entrance meeting was held with the licensee on August 18, 1989, in the Regional Office.
The purpose of the meeting was to brief the licensee on the requirements of the requalification program evaluation and to outline a prospective schedule for the examinations.
.The licensee personnel contacted during the examination are listed in Attachment 1.
The members of the combined NRC/ facility examination team, and.the facility evaluators are also identified in Attachment 1.
2. Examination Results-2.' 1 Requalification Individual Results The following is a summary of the individual examination results:
l NRC l
R0 l
SR0 l
TOTAL l
l GRADING l Pass / Fail l
Pass / Fail l
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15 / 0 l
20 / 0 l
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l l Simulator-l 5/0 l
14 / 1 l
19 / 1 l
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Walk l
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l l Through l
5/0 l
15 / 0 l
20 / 0 l
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2.2 Replacement Individual Results The following is a summary of individual examination results:
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SR0 l
TOTAL l
l GRADING l Pass / Fail l
Pass / Fail l
Pass / Fail l
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2/0 l
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3/0 l
4/0 l
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1/0 l
3/0 l
4/0 l
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2.3 Generic Strengths and Weaknesses-The following is a summary of generic strengths and weaknesses noted l
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examinations. This information is being provided to aid the licensee in upgrading the requalification training program.
No licensee response is required.
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STRENGTHS
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- Ability to recognize entry conditions and utilize Emergence
Operating Procedures, i
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- Ability to perform Emergence Plan Classifications.
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- Good uses of. summations by SC0/SS to keep the crew appraised of plant status during scenarios.
- Ability to locate and operate in plant equipment.
WEAKNESSES
- Failure to consistently use alarm response procedures.
- Knowledge of the Technical Specification requirements for the Feed Water Coolant Injection-System (FWCI).
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- Failure to consistently review precautions and limitat1En prior to performing JPMs.
- Ability to diagnose the failure of the Isolation Condenser
- Lack of formal communications between crew members Due to the limited number of applicants for the replacement l
examination no generic strengths or weaknesses were identified during the operating portion of the test or during the grading of the written examination.
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3. Requalification Program Evaluation Results The facility program for licensed operator requalification training was rated as SATISFACTORY in accordance with the criteria established is l
ES-601, paragraph C,3.b.(1), 0.3.b.(2), 0.1.c.(2)(c), 0.2.c.(2)(b), and D.3.c.(2)(b).
3.1 Examination Results i
.l The overall individual pass rate was 95% which meets the criteria of
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75% established in ES-601, paragraph C.3.b.(1)(b).
.j Five crews were evaluated and all were determined to be satisfactory
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which meets the criteria of no more than one third of the crews may be evaluated as unsatisfactory by the NRC, established in ES-601, l
paragraph D.I.c.(2)(c)(4).
All of the individuals passed the written and walk-through portions of the examination which meets the criteria of 75% established in ES-601, i
paragraphs 0.3.c.(2)(b), and 0.2.c.(2)(b)(2).
I The program is based on a systems approach to training as indicated by
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a review of the licensee submitted reference materials which meets the criteria of ES-601,- paragraph C.3.b.(1)(d).
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i 3.2. Analysis of Pass / Fail Agreement Both-the NRC and the facility found one SRO unsatisfactory on the overall exanination. The facility found one additional SRO unsatisfactory that the.NRC did not. This results in 100% agreement in the grading of-the written and operating examinations since the program will not be penalized for holding a higher standard of operator performance, as stated in ES-601, paragraph D 1.c.(2)(c)(3).
This meets the criteria of 90% agreement established in ES-601, paragraph C.3.b.(1)(a).
The' NRC found all the crews satisfactory on the simulator evaluations while the facility found one of the staff crews unsatisfactory. Again this results in 100% a paragraph D.1.c.(2)(c)greement and the requirements of ES-601, were met.
If the facility evaluators,iudge crew performance unsatisfactory and the NRC does not, remedial training is indicated but the program will not be penalized for holding a higher standard of operator performance.
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There was 100% agreement between the NRC and the facility on pass / fail decisions on the walk-through and the written examinations. This meets the program criteria for 90% agreement established in ES-601, paragraphs D.2.c.(2)(b)(1), and 0.3.c.(2)(b).
3.3 Common Job Performance Measures A review of the results of ten common Job Performance Measures (JPMs)
indicate the maximum percentage of examinees that missed a common JPM was 10%, Therefore, none of the JPMs were missed by more than 50% of the examinees.
Likewise none of-the common questions about the same
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JPM was missed by more than 50% of the examinees.
Therefore, paragraphs C.3 b.(2)(a) & (b) of ES-601 are not applicable.
All of the operators evaluated answered greater.than 80% of the common JPM questions correctly which meets the criteria of at least 75% of the examinees score over 80% on the common JPM questions. Therefore, paragraph C.3.b.(2)(e) of ES-601 is not applicable.
3.4 Licensed Operator Training
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The results of the requalification examinations indicate that the
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L facility does train and evaluate their operators in all positions permitted by their individual licenses. The operators were also trained for in plant JPM as indicated by the examination results and their familiarity with the walk-through process. Therefore, h
paragraphs C.3.b(2)(c) & (d) of ES-601 are not applicable.
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. The facility evaluators were found to be satisfactory in accordance with.the standards established in Attachment 5 to ES-601.
Therefore paragraph C.3.b (2)(f) of ES-601 is not applicable. At the beginning of the examination the facility evaluators were reluctant to supply the
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operatnrs with information that would normally be available to complete
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their task. Once the facility evaluators were' corrected this was no longer a problem. Overall they did an excellent job, i
I 4. Requalification Examination Preparation s
The reference material that was submitted by the licensee met and exceeded the requirements of Attachment 4 to ES-601. The material included a sample plan, a >700 question examination bank,15 simulator scenarios and 77 job.
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performance measures. All the material was well organized, which greatly L
aided in the review process.
The written examinations were reviewed and found to be well prepared and of good quality. 'They sampled a good cross section of information covered during the requalification year. A few questions needed to be reworded to ensure the question solicited the complete answer required. Only a small percentage of the questions were deleted and replaced with others from the bank, due-to their simplicity or lack of clarity.
-Four simulator scenarios were selected by the licensee, for the
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examination.
These scenarios were reviewed during the validation week.
The changes that were made to the scenarios included the addition of crew critical tasks and an increased number of expected operator actions. All the-scenarios were run as is, with the exception of one, that needed to be changed due to its similarity to a scenario taught during a requalification-class. Overall the simulator scenarios met all the requirements of ES-601.
An attention to detail problem was apparent during the validation week.-
The morning after the simulator scenarios were validated, an NRC examiner
.found a curtain open inside the simuletors instructor's booth.
The open curtain resulted in clear view of the simulator from a remote observation gallery.
In order to alleviate any question of scenario compromise, the
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examination team elected to develop a fifth scenario to be run at the Chief Examiner's discretion. The fifth scenario was run during the examination and there was no indication of compromise to any of the examination team
members.
The licensee selected ten job.perforn,ence measures (JPMs) and it was decided that they would all be common JPMs as opposed to the minimum five common JPMs, required by the standard ES-601.
For the most part the JPMs reflected the required steps in the procedures used to perform the task. Changes to the JPMs included the addition of a performance steps which require on operator to obtain the procedure and review the l
precautions prior to performing the task.
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The JPM questions were considered to be a weak area in the preparation of-the-examination.
Many of the questions needed to be rewritten due to clarity. A number of JPM questions were_ replaced due to the requirement of ES-601, paragraph D.2.b.(1)(g), which states the nature of the JPM questions shall be such that the answer cannot be determined simply by looking it up in a procedure.
Problems with the questions also appeared'
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first round of JPMs were complete.
Overall the validation week went very smoothly and the NRC examiners noted excellent cooperation by the facility examination team.
5. Requalification Examination Administration The examination was conducted without any major problems or delays.
For the most part the examination ran on schedule, with slightly longer days during the dynamic simulator portion. The only schedule change concerned the conduct of the JPMs.
It was decided to start an examination team in the plant and the simulator simultaneously.
This alleviated the problem of one team waiting for the first team to complete the simulator JPMs, before getting started.
During the dynamic simulator portion of the examination the NRC examiners were assigned to one individual, while the facility examiners were assigned to a certain position, ie: shift supervisor or balance of plant operator.
This did not result in any problems since the two examiners would confer
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after each scenario. The facility examiners decided to critique the crews after each set of scenarios.-rather than after each scenario.
This saved a considerable amount of time.
An instance of a lack of. attention to detail occurred when the facility lead examiner failed to mark and save the _ simulator traces. af ter the first day's dynamic simulator examination. This was requested by the Chief examiner-in order to use the traces to dispel any areas of disagreement in the grading of the scenarios. As it turned out there were no areas of disagreement and the traces were not needed. All the traces were marked and saved for the rest of the examination.
The dynamic simulator examination was video taped from a number of cameras These tapes proved to be an excellent tool in the grading of certain scenarios. On three different occasions the tapes were immediately reviewed following the scenario to help the examiners in the grading process.
The written portions of the examination were conducted without any problems, all operators completing the section A and B open reference examination
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during the first week of the two week examination. The time validation of the written examinations was adequate as evidenced by all operators completing the examination in the allotted time.
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The entire; examination process went very smoothly, considering the large i
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number of operators evaluated. Good cooperation between the examination
teams allowed any-problers that arose to be immediately resolved and l
. prevented them from impacting the examination.
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6. Replacement Examination Items Prior to administration of the written examination the~ facility operations and training representatives had an opportunity to review the written examination and provided valuable comments to assure-that the questions were appropriate for the facility. This review was conducted on October 19, 1989 with the facility reviewers under pre-and post-examination security agreements. The facility comments were incorporated which resulted in few comments of the applicants during the conduct of the examination. As a result of the pre-examination review the facility also had few post examination comments and provided these to the examiner at the exit meeting.
Prior to the administration of the simulator portion of the operating test t
the simulator scenarios were run on the simulator with the facility representatives, also under security agreements.
The administration of'the examinations went smoothly and no major problems were encountered. The examiner did identify to the facility during the preparation phase of the examination that some of the facility training material was missing and that the quality of the Process and Instrumentation Drawings (P&ID) was poor and in many cases unreadable. The facility provided acceptable' material when informed and agreed to assure that this situation would not occur in future NRC examinations.
7. Exit Meeting An exit meeting was held at the conclusion of the examinations on October 26, 1989. The personnel in attendance are listed in Attachment 1.
The NRC
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results of the simulator and walk-through portions of the examinations were presented. Requalification and Replacement Examination preparation and administration were discussed along with generic strengths and weaknesses of the program.
ATTACHMENT:
1. Persons Contacted 2. Requalification Examination Test Items 3. Simulator. Fidelity Report 4. NRC Resolution of Comments (Replacement)
5. Facility comments on written examination (Replacement)
6. SR0 Examination and Answer Key (Replacement)
'7. Northeast Utilities Letter to the NRC, MP-13810 dated December 5, 1989
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ATTACHMENT 1 PERSONS CONTACTED-Northeast Nuclear Energy Company S. Scace, Millstone. Station Superintendent (2)
J. Stetz, Millstone Unit-1-Superintendent (2)
B. Ruth, Manager, Operator Training (1), (2)
R. Leeneburg, Supervisor, MP1 Operator Training (2)
R. Palmieri, Millstone Unit-1 Operations Supervisor (2), (4)
G. Sturgeon, Millstone Unit-1 Instructor (1), (2), (3), (4)
C. Tabone, Millstone Unit-1 Instructor (2)
E. Berry, Millstone Unit-1 Shift Supervisor (2), (3)
G. Giles, Licensed Operator Requal. Coordinator (4)
S. Gilbert, Millstone Unit-1 Instructor (4)
D. Meckhoff, Millstone Unit-1 Instructor (4)
F. Tuttle, Millstone Unit-1 Instructor (4)
M. Jacobs, Millstone Unit-1 Instructor (4)
R. Payton, Millstone Unit-1 Instructor (4)
Nuclear Regulatory Commission B. Gallo, Branch Chief, Operation (1)
D. Florek, Senior Operations Engineer (1), (2), (3)
T. Easlick, Operations Engineer (1), (2), (3)
C. Sisco, Operations Engineer (1)
M. Morgan, Examiner, PNL (3)
G. Buckley, Examiner, PNL (3)
B. Orton, Examiner, PNL (3)
B. Raymond, Senior Resident Inspector (2)
M. Boyle, Project Manager Millstone (2)
Notes:
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(1) Attended Entrance Meeting, August 18, 1989 (2) Attended Exit Meeting (Examination), October 26, 1989 (3) Member - Combined Facility /NRC Examination Team (4) Facility Evaluator
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ATTACHMENT 2
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REQUALIFICATION EXAMINATION. TEST ITEMS I
. Written Examination - Part B :
- t Open Reference Examination No.. LOR-OR-1B (RO) and (SRO)
i QNUM TEST ITEM NOJ QVAL
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70 (R0 only)
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~135 2.0
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~- -247 2.0 7.
-249 3.0
259 2.0
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292 2.0
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(R0 only)
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- -314 (SRO only)
2.0 13-
-- 351 (SR0 only)
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Written' Examination - Part A
. Static Simulator' Examination No. 1 - Full' Load Reject resulting from the loss of>all'345 KV lines. (SSE11)
Examination Number: LOR-0R-1A (RO)-1 and (SRO)-1 0NUM-TEST ITEM NO.
QVAL
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132 2.0
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i Static Simulator Examination No. 2 - Spurious Group 1 Isolation, MSIV Closure -
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with an ATWS Condition. (SSE14)
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Examination Number: LOR-OR-1A (RO)-1.and (SRO)-1 QNUM TEST ITEM NO.
QVAL l
163 2.0
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168 2.0
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(SRO only)
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170 2.0
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351
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2.0
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172 (SR0 only)
2.0
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352 (R0 only)
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No.-Job 1 Performance Measures Task No.
Location
Return Isol. Cond, to Standby 024-705-01-01 Control Room 2 - Transfer.to MPR at Power 039-784-01-01 Control Room 3 - Transfer-RR Pumps from LOOP to Master 017-705-01-01 Control' Room-
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4 : Perform C.S. Sys. Operability Test 216-700-02-01 Control Room
5 - Maximize CRD Flow 018-717-05-01 Control' Room.
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6 -' Place ESW-& LPCI in Torus Cooling, 046-702-04-01'
Control Room
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L7 Transf. RPS' Bus"B"from RPMG"B"to IRP-1 076-713-04-04 In Plant'
8'--Vent Scram Air l Header During an ATWS 181-300-05-01
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9 - Shift S.W. Strainers (D/G)
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070-712-02-01 In. Plant
Manual Start From Local D/G Panel 070-702-04-01 In Plant-
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Dynamic Simulator Examination
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Scenarios LOR SES-06 - Fuel Element Failure with Isol. Cond. Tube Rupture
. LOR SES-10 - Loss of Feed with a Small Break LOCA LOR SES-12 - Complete. Loss-of Service Water LOR SES-14 - Loss of All Off-Site Power-LOR SES-11 - ATWS Condition with RWCU leak to Secondary Containment a
LOR SES-16 - Feed Water Line Rupture in the Drywell
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r ATTACHMENT _3-SIMULATION FACILITY REPORT
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Facility Licensee: Northeast Utilities Millstone 1 Nuclear Power Station
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Facility Docket-No:-50-245.
Operating' Test Administered on October 21, 1989 and Requalification Examinations'
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administered on the simulator October 16, 17, 18, 23 and 24, 1989.
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This form is to used-to report observations. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of ncn-compliance with 10 CFR 55.45(b). These observations -
do not affect NRC certification or approval of the simulation facility other than
~to provide information which may be used in future evaluations. No licensee action is' required in responte to these observations.
.During the conduct of the simulator portion of the operating tests, the following itemswereobserved.(ifnr,ne,sostate),
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NONE~
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HRC Resolutions of-Commments (Replacement)
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ATTACNMENT 5
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Facility comments on written examination (Replacement)
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I KffACHMEMT 6
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fi 590 Exeminat.',cn and Antwe" Kry (deplaawnt)
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i ATTACHMENT 7 i
NORTHEAST UTILITIES LETTER DATED DECEMBER 5, 1989
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l Millstone-1 - October 20, 1989 Resolutions to facility Corments Page 1 i
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Section 5 i
Question 2a Recommendation accepted and incorporated into answer key.
Question 9b
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Recomendation accepted and incorporated into answer key.
Recommend that facility procedure, ONP-525A, be changed to reflect that
' instrument rack 2206 is an alternate location to read reactor water level, j
Question 21 Reconnr.dation accepted and ince porated into the ar,$wer key.
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Recn nendation ace.epted and incorporated into the answer key.
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Question 26e
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Recoraendation accepted and incorporated into answer key, i
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Section 6
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Questior.12e Recommendation accepted and incorporated into the answer key.
l Question 13 Ij Recommendation accepted. Consideration for drawing will be made during grading.
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Question 14a Recommendation accepted and incorporated into the answer key.
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c ATTACWENT 5
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Facility comment 5 on written examin,tgon (p,p),c,,o t)
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t Ger+rni offices e Seloen Street, Dertin, Connecticut J
U.f.I$$'5.5IIO.~=~~
P.o Dox 270
...... ww. u= "
CCl C7.'.L*2',"1%
H ART F oRD. CONNECTICUT 0$141-0270 I
L L
J ( M "6'6000
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October 26, 1989 j
MP-13672 Mr. Robert M. Gallo Branch. Chief U.
S. Nuclear Regulatory Commission ion I Reg $ Allendale Road
i King of Prussia, YA 19406
.a.
Dear Mr Gallor
-
Attached is the ecmpilation of comocntr, on the written examin-
'
ation administered to Millstone Unit 1 1.ieonse candidates on October 20, 1989.
l These comments were the result of a review ot the examinations I
conducted by memlers of the Millstone Unit 1 training staff.
- References are'previded, where necest,ary, to substantiate the comments.
Please contact Mr. Raymond L. Lueneburg, Supervisor, Operator Training, Millstone Unit 1, with any questions concerning our
'
comments.
Sincerely,
.
Step >e/ #f
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cat.l n
E'.
Scace Station Superintendent Millstone SES/CJT/slk Attachment c
B.
W. Ruth, Manager, Operator Training USNRC Document Control Desk (w/o Attach.)
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On
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S Question 2a..
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This question required matching a system / component that would receive an isolation signal with a condition which may cause an isolation.
The question states the mode switch is in "RUN".
With the mode switch in RUN, the reactor would be at power and the shutdown cooling system would be isolated.
(It would have-
!
isolated at 350*r Rectre loop temperature.)
The condition for part "a." is an RPV level of +4' inches.
One of the. required responses for-part "a."
is "3. - Shutdown Cooling Isolation".
.since Shutdown Cooling is already_ isolated, this choice may not be selected.
a Ren amend m
'
Due to the confusian which may occur-due to the initial condition stated 3n the question, do not require "3.
Shutdown Cooling-Isolation" for full'eredit.
Accept for full credit an answer which only lists
"$".
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Question 9b.
This question requires matching the individual with the location
'
end his function during control room evacuation in ONP 525A.
The procedure ONP 525A directs the BOP to check water level on instrument rack 2205..
Instrument rack 2206 also has. reactor water level indications, is on the same floor in the reactor building and is equally reliable under the plant conditions i
described.
Recommendt Do not deduct any credit for also listing
"L-7" instrument rack 2206-in addition to
"L-6" instrument rack 2205.
There are no adverse consequences to checking level on both instrument racks even though the procedure does not direct it.
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Question 21 This question required a_ list of the four alternate methods to
,
insert control rods described in top 590.18.
One of the methods
,
required is " venting over piston volume on CRD units".
This is i
accomplished by-venting the withdraw header (102 header) through
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the CRD-r102 valve (ref P&ID 25202-26024 SH.5).
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Recommend
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Accept also for this method a substitution of the words "102 l
[
header" for "over piston' volume".
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Question'24e.
This questiot. required matching a component with its failure mode on a loss of instrument air.. The meenanical_Vawum Pump Euction-i valve fails closed (if open) (ref P&iD 2!<202-26033 sh3).
DNP S12 a
Rapid and Total Loss of' Instrument Air listL the failure mode.as l
" stays closed", which was nrat a choice given in the question.
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. A.
Recommendt
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Accept'"3" - Fall Closed since the valve will fail closed if open, and " stays closed" was not a choice.
Note that " Fall as f
_
is" is not c.orrect since if the valve were open, it would not fail.as is.
For future questions of this type, include responses of " stays closed" and " stays open".
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L Question 26a., c.
E This question asks how components will respond to an auto
transfer of Vital AC from its normal to its alternate power
'
source.
Part "a." concerns the EPR.
The EPR fails high on a r
loss of power.- When transferring Vital AC, the procedure directs i
the operator to transfer pressure control to the MPR manually i
prior to the power transfer (ref OP 343, pp. 6).
The training
"
text warns that loss of power to the EPR would cause it to fail
'
with the MPR taking control; however, on a power transfer, the
!
EPR should remain in control (ref Text 1343, pp. 15).
This is
<
not certain, however, since it has not been proven at the plant.
,
.
Recommends Accept for full credit an answer which indicates knowledge of hou the EPR would respond if it does not remain in control
,
Part "e." concerns the Reactor Protection System (RPS).
RPS l
respond with a B channel trip causing a half scram.
'
i Recommend:
,
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Accept for full credit " Half-scram".
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Question 12e.
(Plant systems section)
l This question requires matching Radwaste system inputs with receiving tanks.
The input " Discharge flow from floor drain i
collector pump" is matched with the Waste Surge Tank.
The Waste Surge tank can be lined up to the floor drain collector pump's discharge; however, there is no procedure for doing this and, in
,
fact, is never done.
OP 312, pp. 4, warns in a precaution that
'
no water will be pumped to the waste surge tank without the Unit Superintendant approval.
This makes it a very unlikely choice.
A procedure exists for directing water from one floor drain
,
,
collector tank to another (ref OP 312D).
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Recommend:
f Accept for full credit a response which only lists
"3" Ploor
Drain Collector Tanks (A,B,C,D).
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i Question 13 This question required 4 indications that a fuel assembly is
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correctly oriented in the core.
This can be shown in a drawing
'
showing.4 fuel bundles.,
I Recommend:
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Accept a correct drawing of orientation for full credit.
i I:
e
. Question 14a.
(Plant systems section)
This question required appropriate classification for a given 1,
event.
,
"A spent fuel bundle has been dropped in the Spent fuel pool.
Area Radiaticn Moniter reads offscale."
The event given.does not say how high a -tadiation level there is or indicate the exte:it of
.
fuel damage.
This leaves it to the-judgement of the Shift Supervisor.
For an " Alert" the EAL tables list " ARM offscale" as a symptom, with the condition being "Hi Red levels onsite i
including damage to spent fuel."
.
t For a " Site Area Emergency" the EAL table lists " ARM on refueling
,
. floor read > 1000 mr/hr" as a symptom, with the condition being
,
" report of a fuel handling accident, ranjor damage to spent fuel.
of thp 5 refuel floor ARMS, offscale could be anywhere from 100
to 10 mr/hr depending on the ARM.
Recommend:
Accept for full credit either " Alert" or " Site Area Emergency"
since the decision will be based on the SS's opinion of the fuel
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damage.
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Page 4 of 4 m
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ATTA;PWdNT 6 SRO Examination and Anwer Key (Replacement)
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,w l'1 QQ
DRAFT COPY bek i
U. S. NUCLEAR REGULATORY COMMISSION
SENIOR REACTOR OPERATOR LICENSE EXAMINATION F
REGION 1 L
u FACILITY:
Millstone 1
's REACTOR TYPE:
BWR-GE3 h.:
ll DATE ADMINSTERED: 89/10/20
)
(
CANDIDATE:
' INSTRUCTIONS TO CANDIDATE:
- Points for each question are indicated in parentheses after the question.
i
' The' p&ssing grade requires at -least 80% correct overall.
Examination
,
- papers will be picked up four and one half (41/2) hours after the
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examination' starts.
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CATEGORY % 0F.
' NUMBER
.
F VALUE'
TOTAL CORRECT CATEGORY
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48.50 60.63 I
EMERGENCY AND ABNOP, MAL PLANT EVOLUTIONS
!
(43%)
_l A 31.50 39.38
'
PLANT SYSTEMS (40%) AND PLANT-WIDE GENERIC
~
RESPONSIBILITIES (17%)
}
80.00
~
OVERALL
% CORRECT OVERALL i-i
. All work done on this examination is my own, l'have neither given i
nor' received-aid.
l Candidate's Signature DRAFT COPY
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la Page 2 I!.
' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS (43%)
QUESTION: 01 (1.00)
From the list of Reactor Building Closed Cooling loads (RBCCW),
(1.0)
SELECT _the ONE (1) that 1syessential4cao(er os h Drywell Equipment Drain Sump Cooler a.
b.
Reactor Building Equipment Drain Tank Cooler Reactor Water Cleanup Non-Regenerative Heat Exchangers c.
~d.
Shutdown Cooling Heat Exchangers
,
ANSWER:-
0 '.
(1.00)
'
a.
(+1.0)
,
REFERENCE:
Operator Training, Systems Vol. 2, PC text p. 69, 1.
MPl:
Objective 46 and RBCCW text fig. 1.
-KA Numbers 22300lK601 (3.8), 295018K101 (3.6), and 295018K201
.
~ 2.
(3.4).
223001K601 295018K201 295018K101
..(KA's)
1
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- EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 3 (43%)
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QUESTION: 02 (3.00)
'
MATCH the plant conditions in Column I with the system / component in Column 11 that should receive an isolation signal. The
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systems / components in column !! can be used more than once or not i;
at all. ASSUME mode switch is in the "RUN" position, f
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COLUMN I COLUMN 11
'
(Condition)
(System / Component)
.a.
RPV Level 44 inches 1.
,
System Isolatior.
- b.=
RPV Pressure 1086 psig
,
n,
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2.
MSIV Isolation
?.
c..
Main Steam Lir.e
!
Pressure 800 psig 3.
!
q I
Isolation d.
Drywell Pressure 2.5 psig i
,
e.
Main Steam Tunnel System Isolation
,
. Temp. 205 deg F 5.
TIP System Withdrawal
f.
"C" Main Steam Line and Isolation Flow 115% of normal 6.
None F
ANSWER:
(3.00)
g
'4r Gep.acM
,,
b.:
.c.
'
,
d.
.
e..
f. -
(+0.5) each lettered respense l
(***** CATEGORY 5 CONTINVED ON NEXT PAGE *****)
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 4 (437.)
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REFERENCE:
k 1.
'MP1: Operator Training Systems Vol. II, PC text, pp. Ill-
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113,-Objectives 21 and 22, (RWCU) 3, (TIP) 4.
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2.
KA Numbers 295024K207 (3.9), and 295031K212 (4.5).
295024K207 295031K212
..(KA's)
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' QUESTION: 03 (1.00)
.
L Concerning ONP 512, " Rapid and Total loss of Instrument Air,"
h
~ SELECT the INCORRECT automatic action.
(Instrument Air header pressure is 50 psig.)
(1.0)
a.
Reactor Scram b.
Instrument to Service Air. Isolates c.
Drywell In:itrument Air Isolates d.-
Standby Instrument Air Compressor Starts t-ANSWER:
(1.00)
'
c.-
( + 1. 0 ) -
REFERENCE:
1.
MPl: Operator Training, ONP Text, pp. 104 107, Objective 90.
2.
KA Numbers 295019K214 (3.2).
,
295019K214
..(KA's)
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' EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 5 (43%)
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s QUESTION: 04 (1.00)
Following a Main Generator load reject, SELECT the CORRECT automatic
action that decreases reactor power to compensate for the reactor
. power increase that will occur when cold feedwater returns to the L
reactor.
(1.0)
a.
recirculation pump runback to 28%
b.
APRM high flux setdown c.
reactor scram
,.
d.
select rod insert
!
ANSWER:
(1.00)
.d.
[+1.0)
REFERENCE:
1.
MP1:. Operator Training, ONP Text, p. 26, Objective 21, 2.
KA Numbers 295005K301 (3.8).
L 295005K301
..(KA's)
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 6 (43%)
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QUEST 10N: 05 (1.00)
'
While operating at 100% power, a Group I isolation and a reactor
!.
scram occur. Data collected from the plant computer and plant t
operators indicates the following occurred:
!-
The Group I isolation was caused by a Technician error.
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The reactor scram was caused by high reactor. pressure.
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The Isolation Condenser initiated.
-
All operating feedpumps tripped and FMCI Inhibit alarm
-
was received.
Reactor water level decreases and FWCl auto started and
!
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injected to the reactor vessel.
F An operator reset Feedwater Flow Control and shifted to
-
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level control to maintain reactor vessel level at 37".
Continuing to follow Millstone procedures, the plant is
-
placed in a normal shutdown condition, j.
SELECT the CORRECT statement below.
(1,0)
a.
Power operation cannot resume because FWCI auto-initiated
!
and injected to the reactor vessel.
b.
Power operation cannot resume because a safety limit may have been violated, c;
Power operation cannot resume because the feedpumps should not have tripped.
d.
Power operation cannot resume until permission is received
,
from State authorities and the NRC because the Emergency i
Plan was implemented.
ANSWER:
(1.00)
,
b.
(+1,0]
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 7 (43%)
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REFERENCE:
1.
MPl: Technical Specifications 2.1.1.C.
2.
KA Numbers 295005K201 (3.7), 295005A206 (3.6), and 295005G003
,
F (3.9).
!
2950050003 295005A206 295005K201
..(KA's)
!
c
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QUESTION:.06 (1.00)
The Emergency Operating Procedures refer to the " Heat Capacity Temperature Limit Curve." SELECT ONE (1) of the following statements i
~,
that CORRECTLY identifies the three (3) parameters used to determine
!
the Heat Capacity Temperature Limit.
(1.0)
a.
Torus Water Temperature, RPV Pressure and Torus Water Level b.
Torus Water Temperature, Drywell Pressure and Torus Water Level
,
t c.
Drywell Temperature, RPV Pressure and Torus Water Level
'
'
d.
Toras Water Temperature, RPV Pressure and Drywell Pressure P
ANSWER:
(1.00)
,
a,
.(+1.0]
!
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REFERENCE:
1.
HPl:
E0P 570, "RPV Control," figure 8
'
2.
KA Numbers 295026A203 (4.0),
i 295026A203
..(KA's)
,
QUESTION: 07 (3.00)
. A manual scram must be inserted whenever a limiting safety system
!!
setting has been exceeded and no scram has occurred. STATE SIX
{i
.(6) additional plant conditions that require a manual scram, (3.0)
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-EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 8 (43%)'
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k ANSWER:
(3.00)
'
.
1.
Both recirculation pumps trip.
2.
Feedwater conductivity is at, or abo'e, 0.5 micro mhos (or
!
15 micro mhos at condensate pump discharge).
L'
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3.
Reactor water cleanup conductivity is at or above 2.0 micro mhos.
[-
4.
High energy pipe rupture, unisolable from the reactor.
5.
Loss of TBCCW.
6.-
Loss of TBSCCW.
- .
l-1.
Loss of RBCCW.
I 8. _
Loss of Service Water.
"
r 9.
Loss of both CRD pumps,
,
any six (6) [+0.5) each
[
REFERENCE:
[
1.
MP1: Operator Training, Procedure, ONP Text, pp. 7 and 123, L:
_ Objective 6.
L 2.
KA Numbers 29500lG0ll (4.2), 295002G0ll (4,0), 295018G011
'
(4.1).
f 29500lG001
..(KA's)
g
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page '9 (43%)
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' QUESTION:. 08.
(2.00)
!
!
L MA1CH.the characteristics of fission product release in Column !
ii
[-itithtypeofreleasemechanismsinColumn!!.
(2.0)
[
.
a I,
COLUMN 1 COLUMN !!
j
.
a.
Release rates are not 1.
Tramp uranium in core l
,
~
proportional.to power changes material j
lL b.
Release rates are exponential 2.
Pin hole defects in
,i i
with power changes fuel cladding
!n
!
c.
Release rates are proportional 3.
Cracks or splits in
!
..
to power changes fuel cladding a
..
,
d.
Release ~1evels are low l
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ANSWER:.
(2.00)
!
a..
2 (3)
t
b. -.
3-
!
c.
I
!
n d.
!
[40.5)each REFERENCE:
1.
MPl: Operator Training, Systems Volume 7, MCO Text, pp. 33
,
through 35, Objectives 8 and 9.
2.-
KA Humbers 295017K201(3.3),
i<
'295017K201
..(KA's)
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L IRGENCY AND ABNORMAL PLANT. EVOLUTIONS Page 10 (43%)
i L QUESTION: 09 (3.00 hyded)
l ONP 525A, "Oc.:y:d Fire in Control Room or Cable Vault," states
'
[
that if a Control Room Evacuation is required, specific-individuals are required to go to different plant locations and perform-c l
activities required by ONP 525A.
MATCH the activities in Column I l'
' tith the locations and individuals in Column 11 and Column 111.
The locations and individuals may be used more than once or not
,
at all.
(3.0)
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-COLUMN I COLUMN 11 COLUMN 111 (Activity)
(Location)
(Individual)
a.
Valve out scram L 1.
MCC 101 AB-1 1-1.
L air header (42'6" level)
I
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L 2.
SCO i
b.
Check reactor (14'6" level)
water level by Narrow Range Yarway L-3.
CR0 flow Control 1 3.
CRP 905 Station Control Operator c.
Control reactor L 4.
MCC F 3 pressure between 1-4.
BOP Control 900 & 1000 psig L-5.
1 10 3 Area Operator d.
Open breakers for L 6.
Instrument Rack 1-5.
Plant 1 1C-3 & 1 1C-10 2205 Equipment to allow local Operator valve control L 7.
Instrument Rack 2206 e.
Open circuit breaker for L 8.
Main Steam Tunnel 1-MS 6.
.
f.
Place CRIS/IC-1
& CRIS/IC 4
switches to
>
Emergency
,
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 11 (43%)
,
ANSWER:
(3.00)
a.
- L 3,1 4 b.
L 6, 1-4 s
c, L 5,12
'
d..
L 1, 1 3
c.
L-2, 1-4 f.
L4,I2
-[+0.5)eachletter
+
,
-REFERENCE:
M
'
l.
WPl: ONP 525A, p. 2-3.
2.
KA Numbers 295016K202 (4.1) and 295016G006 (4.1).
,
295016G006 295016K202
.,(KA's)
QUESTION: 10 (1.50)
The ATWS System will automatically (in)itiate when reactor vessel level is less than or equal to
inches for at least (2h-seconds on both channels of onq division or reactor
>-pressure is greater than or equal to ( 3.T psig on both channels of one division.
(1.5)
,
' ANSWER:
(1.50)
1.
-48
.
2.
3.
1150
[+0.5]each (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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. EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 12 (43%)
I:
REFERENCE:
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, 1.
MP1: Operator Training, Systems Volume 7, ATWS Text, p. 19.
2.
KA Numbers 295037K203 (4.2).
.
f i
295037K203
..(KA's)
,
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f QUESTION: 11 (3.00)
I Given figures 4 and 5 from the Emergency Operating Procedures (EOP)
'
and the following conditions, STATE why :dequate core cooling is i
NOT. o':nvided.
(3.0)
'
e l
Drywell Temperature 300 deg F
!
-RPV' Pressure 400 psig
'
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,
t'
RPV' Indicated Level-140 inctes
i Core Spray Pumps are running
!
I
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ANSWER:
(3.00)
D\\ f;
!
. Reactor--level is below Top of Active fuel
'
31.0-(Core Spray system is not injectin)g
-,41.0c L
Reactor level is abuve half core L +1. 0 >
-
(Reactor Pressure is above 350 psi permissive f il. r-for Core Spray admission valves to oPen)
'
!
!
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REFERENCE-i 1.
MPl: Operator Training, Procedures, E0P Text, p. 26, Objective 14.
,.
2, MPl:
E0P 570, RPV Control, p. 30, table 7.
F l-3.
KA Numbers 295031K302 (4.7), 295031K303 (4.4), and 295031K304
'
.(4.3).
>
l-295031K304 295031K303 295031K302
..(KA's)
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 13 (43%)
QUESTION: 12 (2.00)
Column 11 lists the consequences of system trips and isolations.
Column I lists different combinations of system trips and isolations.
MATCH the combinations in Column I with the consequences in Column II.
The answers in Column 11 may be used more than once or not at all.
ASSUME reactor power 4e MM and APRM channels are not bypassed.
(2.0)
as 40%
COLUMN I COLUMN 11 a.
No Trip A and 0 isolate
.
2.
Half Scram b.
Main Steam Lines C and D isolate 3.
Full Scram c.
APRM Channels 1 and 6 trip
-
d.
APRM Channels l
2 and 3 trip
ANSWER:
(2.00)
'
a.
b.
c.
!
d.
[+0.5]each l
REFERENCE:
1. -
MPl: Operator Training, Systems Volume 7, RPS Text, p. 23, i
Objective 17.
2.
KA Numbers 295006K201 (4.4) and 212000A211 (4.1).
l 212000A211 295006K201
..(KA's)
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EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 14 o
(43%)-
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a L
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L QUESTION: 13 (1.00)
'In consideration of uncontrolled power oscillations, ONP 503D
t
" Response to a Request-For Generation Rqduction" states:
'
CWhenever core flow is less than (1.1 pounds per hour,
'
reactor power must be below the ( 2.5
% rod line of the
power / flow operating map."
(1,0)
,
ANSWER:
(1.00)
1.
31 ud11 ion (+0.5]
~
2.
50% [40.5]
!
, o.
.
l REFERENCE:
?
!
1.
.MPl: ONP 503D, Rev. O. Change 1.
,
2.
KA Numbers 295001A201 (3.8).
'
295001A201
..(KA's)
i QUESTION: 14 (2,00)
[~.
Reactor is' operating at 100% power and one recirculation pump trips off line. Do the following plant parameters initially j
INCREASE, DECREASE, or REMAIN THE SAME7 (2.0)
a.
Reactor Power-b.
Reactor Level
,
c.
Reactor Pressure l-r d.
-Total Core Flow (***** CATEGORY 5 CONTINUE 0 ON NEXT PAGE *****)
=-
,
n
.e b.
l EMERGENCY AND ABNORMAL PLAN 1 EVOLUTIONS Page 15
(43%)-
,
t f.' ANSWER:14'
(2.00)
!
L-a.
decrease
-
[
b..
increase
!
[
c.
decrease
d.
decrease 1[<0.5)each i
REFERENCF:
'
1, MPl: _ Operator Training, Systems Volume 1. RR Text, Figure 2.
2.
KA Neobers 29500lK301 (3.6), 29500lK302 (3.8), and 295>0lK306
,
(3.0).
.
295001K306 29500lK302 29500lK301
..(KA's)
.
QUESTION: 15 (1.00)
i SELECT the ONE (1) statement that explains the reason that pressures
[
in the drywell and torus must be above 5 psig before Primary Containment Sprays can be initiated.
(1.0)
a.
To prevent chugging in the downcomer discharge to the
,
suppression pool, b.
To _ prevent thermal shock to critical instrumentation in drywell.
c.
To ensure adequate Net Positive Suction Head for the LPCI pumps, d.
To ensure that sufficient differential pressure will be i
developed between the torus and the drywell.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
I
e
-
- .
.*
i EMERGENCY AND ACNORMAL PLANT EVOLUTIONS'
Page 16
(43%)
F ANSWER:
(1.00)
>
[+1.0)
c.
REFERENCE:
I'
l.
MP1: Operator Training, System Volume 5, LPCI Text, p. 30,
!
_
Objective 30.
s 2.
.KA Numbers 295024K203 (3.9), 295024K211 (4.2), and 295024K215 (3.9).
I l~
295024K215 295024K211 295024K203
..(KA's)
N i
,
.
-
(1.00)
- QUEST 10N
- 116-
'
r
SELECT the ONE-(1) CCRRECT anner.
If another entry condition
.;
.
occurs-for an Einergency Cperating Procedure (EOP) that has y
been completed, then the completed E0P:
(1.0)
y
,
a.
is Snmediately performed again, b.
is performed again, after completing the E0Ps that have been started, c.
is not required to be performed again, d.
can be performed again, but is entirely at the discretion of the Shift Supervisor.
'
ANSWER:
(1.00)
a.
[+1.0]
!
.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
,
g
.
- ;.
.
.
g o
.
EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 17 (43%)
[
REFERENCE:
~1.
MPl: Operator Training, Procedures, E0P Text, p. 5-6,
.
Objective'4.
2.
KA Numbers 295009G0ll (4.5).
'
,
295009G0ll
..(KA's)
QUESTION: 17 (2.00)-
.Upon loss of feedwater heating, do the following plant parameters initially INCREASE, DECREASE, or REMAIN THE SAME?
(2.0)
a.
Reactor Power
,
-b.
foedwater Temperature c.
Generator Output
r d.
f t:i:ter Me:4ee--L&v4.1.
N
'
Recincula.6lm b e de.h b lHx L c//en /h.( /,(psH j f
,,
.
ANSWER:
(2.00)
I a.
increase b.
_ decrease
!
c.
decrease
?
d.
' increase (
(+0.5] each
,
REFERENCE:
1..
MPl: Operator Training, procedures, ONP Text, p. 160,
,
Objectives 10 and 129,
'
2.
KA Numbers 295014K206 (3,5), 295014A203 (4.3), 295014A106 (3.4).
295014A106 295014A203 295014K206
..(KA's)
.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
.
?
p.-
-
<
z
.
.
e a o
i EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 18 (43%)
QUESTION: 18 (2.00)
l I..
For each condition listed in Column 1 MATCH the action in Column 11 that is required by the Technical Specification.
Each action
,
cay be used more than once or not at all.
(Technical Specifications
!
are provided.)
(2.0)
COLUMN I COLUMN II (Condition)
(Action)
>
>
a.
Suppression Pool Temp 1.
the reactor shall be in Cold
is 110 deg F Shutdown or Refuel condition
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
'
b, during Relief Valve testing at full peuer 2.
Reactor Coolant Temperature Suppression Pool Tomp sha'l ba Setow 330 des F
>
,
reaches 100 @ g f within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
,
'
c.
Trrut level instrumtntation 3.
Scran the reactor
,
has been out of service
.
'
for 6 1/2 hours 4.
depressurize the RPV to less than 200 psig at normal de the reactor W 3 crammed cooldown rates due to a G %up I isolation and the %ppression Pool-S.
Wer Suporession Pool T'ar lemp is 120 deg F to below the normal operating temperature within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
'
,
ANSWER:
(2.00)
a.
b.
c.
I c.
t
[+0.5] each (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
.
.,
yW
'V..
...
.i 5 ]-
OIMERGENCY"ANDABNORMALPLANT, EVOLUTIONS w@st 'h(t3%)
'
'
Page 19'
>
=,
W
-
Mk. 3 y
.
se. 7,-
b :REff,RINCE:
.
l _ :.1, MPl: To::hnical Specification 3.7 A.1
.
'
2.
KA Numbers 295013G008 (4.4).
i f,
295013G008
..(KA's)
,
b
.
.t
.I y, QUESTION: '19 (1.00)
'
SELECT the statement that best describes the Boron Injection c
+1nitiation Temperature.
(1.0)
.
a.
Torus Water Temperature is 110 deg F L
b.
Torus Water Temperature is 110 deg F and increases as
!
power decreases
.
c.
Torus Water Temperature is 155 deg F and decreases as power decreases
,
,
d.
Torus Water Temperature is'155 deg F i
l ANSWER:-- 19 (1.00)
- b. -
[+1 0)
.
L REF' ERENCE:
l 1.
MP1:.EOP 575, " Failure to Scram," Figure 11, i
2.
KA Numbers 295037K302 (4.5), 295037A204 (4.1) and
.295037A201.(4.3).
295037A201 295037A204 295037K302
..(KA's)
.
("*** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
'
,
j y
.
- c.;
--
,
,
y
, ;.
e <.
n
,
't
<
EMERGENCY 1AND. ABNORMAL PLANT. EVOLUTIONS Page 20 t
.(43%)?
I
!
QUESTION: 20 (3.00)
-
'For each condition listed in Column 1 MATCH ALL the Emergency
m Operating Procedures (EOPs) that should be entered.
The E0Ps can
,'
be.used more than once'or.not at all.
(3.0)
' COLUMN I COLUMN II
-
- (Conditions)
(EOPs)
!
as. Primary Containment Hydrogen 1.
RPV Control, E0P 570 Concentration is 1.5%
t 2.
Failure to Scram, E0P 575
' b..
' Pressure is 0 inches of water 3.
Primary Containment Reactor Building Differential
~ Control, E0P 580 c.
Torus Water Level is 14.1-feet 4.
,
d.
. Drywell Pressure is 2.2 psig Control, E0P 585 e.
Reactor Water Level is 5.
None
.+5 inches f.
Drywell Temperature is '
-
'160 deg F-
' ANSWER:
'20-
-(3.00)
.
.
a
.5 b.-
.
.c.
.d.
1, 3-e.-
f.
3'
>
[+0.5] each lettered response
'
.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
>
,
w=
'
'
,
.
,
..g.
.l6
,n.
y i< ' EMERGENCY AND' ABNORMAL PLANT EVOLUTIONS" Page 21
L(43%):
'
,
.
L I -REFERENCE:
Ir
MPl: LEOP 570, "RPV Control," p. 1.
-
-
2.
MPl:- E0P!580, '" Primary Containment Control," p. _1.
' 3 '.
MPl:. E0P 585, " Secondary Containment Control," p.1.
'
. _4...KA Numbers 295024 Goll-(4.5), 295028G0ll (4.4), 295031G011
(4.6),:and 295035 Goll ~(4.2).
,
?
295035G011-295031G011.
295028G011 295024G0ll.
..(KA's)
p QUESTION: 21 (2.00).
STATE the F0VR'(4) alternate methods to insert control rods described
'
in E0P 590.18 " Alternate Methods of Control Rod Insertion".
(2.0)
LANSWER:
(2.00)
- 1.-
scramming' individual control rods (+0.5]
- 2.'
using maximum cooling water pressure [+0.5]
3.
using maximum drive. water pressure (+0.5]'
4.
. venting over piston volume on CRD units (+0.5] ( ve.J.'g to t 43 e.-
r
- REFERENCE
' l '.
MPl:
E0P 590.18 " Alternate Methods of Control Rod Insertion".
-2.-
KA 295037K307 (4.3)
295037K307
..(KA's)
.
QUESTION: 22 (' 2. 00)
' STATE.FOUR (4) alternate systems for injecting so(ium pentaborate
.
into the RPV. (Spec.*f.c. F/w PA &, roh, 9 lnec[.)
(2.0)
,
I
-.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
s
. = - -
g
,
.
.
,; c.
...
,
-
,
EMERGENCY'AND ABNORMAL PLANT EVOLUTIONS Page 22 i
'(43%)-
,
'
'
.
s
,
y
-
,
- ANSWER
- ' -22 (2.00).
,
1.
Feedwater [+0.5]
'
'
2.
CRD [+0.5]
3. -
Reactor Water Cleanup- [+0.5].
g h ' 4.
Hydro pump.(into SLC injection drain line)
[+0.5]
REFERENCE:
-
1.-
MPl: E0P 590.2 " Injecting Sodium Pentaborate using the CRD System"
'2.
MPl:
E0P 590.3 " Injecting Sodium Pentaborate using the Cleanup Filter" 73.
MPl:
E0P 590.4 " Injecting Sodium Pentaborate using the SLC
- '
.
Injection Line Drain" 4.-
MPl:
E0P S90.5 " Injecting Sodium Pentaborate using the Feedwater System"
- 5.'
KA 295037K213 (4.1)'
,
.295037K213
..(KA's)
,
t:
\\
i e
l l
.
.
<
l-
-(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
L r
W
~
,
-
..
p g..
4 '
.
f EMERGENCY: AND ABNORMAL PLANT EVOLUTIONS Page 23-(43%)!
,
,
r
!
i LQUESTIONi 23 (2.00)
MATCl the automatic action in Column I with the condenser vacuum
!
<
setpoint in Column 11. The setpoints in. Column 11 can be used i
{;
tore than once or not at all.
(2.0)
'
COLUMN I COLUMN 11
.
Il La.
Turbine Trip 1.
7" Hg
,
b.:
Bypass Valve Trip 2.
15" Hg
c.
Low Vacuum Alarm 3.
20" Hg
.
'
- di- ~ Reactor Scram-4.
21.5" Hg
!
v
.
5. -22" Hg I^
6.
22.5" Hg c
7.
23" Hg
8.
25" Hg
- -ANSWER
'23 (2.00)
a.-
6.
>
b.D
.1 -
- i-o c.
d.
i
,[+0.5] each
!
- .;
i!
-if Ii
,-
l (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
!
.
--
_. ~
_
f.
s.$
'
'
f.
'
I EMERGENCY AND ABNORMAL 1 PLANT EVOLUTIONS Page 24<
(43%)-
!
y L..
"
' REFERENCE:
p 1.
'MPl: Operator Training, Procedures, ONP Text, pp. 75 through 77, Objective 62.
'
'2.-
KA Numbers 295002K301 (3.8), 295002K302 (3.4), and 295002K304 (3.6).-
,
p 295002K301 295002K304 295002K302'
..(KA's)
j b ' QUESTION: 24 (3.00)
,The plant has experienced a total loss of Instrument. Air (Instrument
--Air. Header' Pressure is O psig). Column I lists the various
,
components that may be affected. MATCH the components from o"
. Column I with the appropriate failure mode listed in Column II.
-
- Each failure _ mode may be used more than once or'not at all. ASSUME
!
reactor is:at 100% power.
(3.0)
t
,
'
COLUMN I COLUMN II-(Component)
-(Failure)
.
c.
Outboard MSIVs 1.--Fail As Is
b.
Inboard MSIVs 2.
Fail Open
'
,
tc.
Isolation Condenser Vent 3.
Fail Closed
.d.-
RBCCW Surge Tank Fill Valve 4.- Not-Affected (i.e.,
g has independent e.
Mechanical Vacuum Pump-source or. backup g
Suction Valve, system).
"'
N.JJ:~3 f.
..tr::1 Valves
,
-(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
,
$q
%%
ppn v
-;..
+
.
3, : 4;.. *
-
,
-
EMERGENCY AND ABNORMAL PLANT-EVOLUTIONS Page 252
' ' ( 43%) ;-
.
'
5NSWER:-'24-(3.00)~
-a.-
- b.
4.
,
c.
- 3 d.-
e..
2r 3
- w t
f.
'4-
[+0.5] each
'
.
REFERENCE:
[
1.
MPl: 0NP 512,'" Rapid and Total Loss of Instrument Air,"
'
.
.
. Figure 7.1.
2.
KA Numbers 295019K202 (3.0), 295019K203 (3.3), 295019K205 (3.4), and 295019K213 (3.2).
.
295019K213 295019K205 295019K203 295019K202
..(KA's)
TQUESTION 25'
(1.00)
tWhile' performing refueling a spent fuel-bundle is dropped in the
- . reactor cavity-area. STATE the areas of the plant that require u
'
>
evacuation per the:immediate actions of ONP 519, " Dropped fuel
'
Bundle".
(1.0)
s i
(ANSWER:
(1.00)
11.
. Refuel' Floor [+0.5]
12.
Drywell
[+0.5]
t'
ac:
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
.
7,-
.f s z.; e
,
p
r
,
,
L EMERGENCY AND ABNORMAL PLANT EVOLUTIONS Page 26
'
b 1 (43%) ~.
'
'
' '
q i
s,
-REFERENCE:
11. -.MPl: ONP!519, " Dropped Fuel Bundle," p.1.
'
L 2.' -
KA Numbers 295023K301 (4.3).
j-
'295023K301
..(KA's)
- j; p.
' QUESTION: 26-(3.00)
-
Millstone' Unit'1 is operating at 100% power when the vital AC bu's j
automatically transfers' from its normal to alternate. power source.
DESCRIBE how each of the following components and systems are affected by-the transfer.
,
.
a.)
Electrohydralic Pressure Regulator (EPR)-
(0.5)
,
b.L Reactor Building Ventilation-(0.5)
c i-c.-
~ Feedwater Regulating Valves (0.5)
i
~d.
Recirculation System-(0.5)
e.:
Reactor Protection System (0,5)
' f.
Rod Block Monitor (0.5)
'
ANSWER:
(3.00)
-a.
EPR remains in control.(with.only a minor perturbation an.y 1,54 PWe(
L, Aa.5,W
,bb Peacg4.,bc,.,%
q int system pressure) [+0.5],,.EP E
b.
~ isolates [+0.5]
- "C c..
FRVs lockup (fail as is)
[+0.5]
,i id; Scoop tubes lockup [+0.5]
e.
RPS channel B trips [+0.5](g,lf5%
i
- f.
.no-affect [+0.5]
l l
L 1-
~(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
<
.
i
, +.,,
--
.
,,
.
,
G < ar ; { ^. -. ;
-l I
-
'
,
'
' EMERGENCY AND' ABNORMAL PLANT.EVOLUTIONSL Page'27 ((43%);
r-t
.
'l REFERENCE:
-i J,
MPl:- Operator Training, Systems Volume 6, TX-1343, Vital
-
-
and Instrument AC Systems, pp. 14, 15, and 16, Objective 8.
- 2.
KA Numbers 262002A301 (3.1) and 295003K301 (3.5).
2'95003K301~
..(KA's)
,
l
r
-!
-- ;
i I
'
,33 k
\\
I
,
i
,
L(***** END OF CATEG09Y 5 *****)
'
,
<
m,
...
,
,
y,y
>,y
,
!
PLANTLSYSTEMS (40%) AND PLANT-WIDE. GENERIC-Page'28
RESPONSIBILITIES-(17%)-
,
-
.
(c-L
,
'
~
QUES. TION:'01. -(1.00)
Concerning the Standby Liquid. Control System, SELECT.the CORRECT
-analysis of the boron bearing solution that. meets Technical
' Specifications. TechnicalJSpecifications are provided.
(1.0)
,
a.
-Boron enrichment-
.56%
-Concentration (wt%). 13%
t-Tank Volume
'1900 gallons
- Temperature 60 deg F
b..
-Boron. enrichment-54%.
-Concentration (wt%)~ 10%
>
-Tank: Volume 1950 gallons-Temperature 75 deg F
c.
-Boron enrichmentL 51%
-Concentration (wt%)'.15%
'
-Tank Volume.
-1800 gallons-Temperature 95 deg F-n d.
-Boron enrichment 52%
-Concentration (wt%)
12%
_
^-Tank Volume-1950' gallons-Temperature 75 deg F
,
,
,
?
'
-ANSWER:
Ol'
(l.00)-
{
- d.'
[+1. 0)-'
' REFERENCE:
1..
MPl: Technical-Specification 3.4.C.
?2.
~KA Numbers 211000G006 (4.2)
,
s 211000G006'
..(KA's)
i
.(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
'
.
q v --
.<
.
.
g.
,
y, x <
.
PLANT SYSTEMS (40%) AND ' PLANT-WIDE GENERIC Page 29 k
RESPONSIBILITIES (17%)
~
[
i
,
,
i YQUEST!0N:t02-.(3.00)
.-MATCH the permissive or limitation' in Column I with the Mode Switch
>
Position-in Column:II'. The Mode Switch positions in Column Il may be used more'than once or not at all.
(3.0)..
COLUMN I
.
COLUMN 11
-
'
W.' w((*4.amd On ' IeseI uut.
-
Cry el, s;r,t,il:tia- =ad norg 1.
Refuel
,
a.
_4, m i., +. 4 s.. u s. u. m m. m... L.
- u..,_, n.
.
_
..
.,
1 b;,-.
.No control rod movement possible.
..
3.
Startup/ Hot Standby c.
-APRM 15% Scram in service.
4.
Run-d..
Discharge volume high water level bypass.
. TRyt $l,'tk Flu T".'es Bypsnet>.
-el-RPS e,,e,gi :d wits !PF byanst^_.
cf.-
One rod. free movement interlock in service, j:
!
.
.
!
'
-ANSWER:
(3.00)
'
1':
-a.
'
b.
2.
!
-!
I Lc.
1, 2, ' 3
'
-.
d.
1;12 e.
- f.~
-1~
-(+0.5].each lettered response
'
!
.
.
l (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
l I'
-
- - - - -
-
- - - -
p
.
to g '.
.
.,
.
,
lP
- ..f 3 ;
V
'
PLANTiSYSTEMS (40%)-AND PLANT-WIDE GENERIC Page 30
,
RESPONSIBILITIES-(17%);-
'
- REFERENCE:--
-
' " 1.
IMP 1: Operator-Training, RPS Text, pp. 57 through 59, Objective 28.
L.2.'
KA Numbers 212000K412_(4.1)-and 212000A216 (4.1).
212000A216 212000K412
..(KA's)
t QUESTION: 03 (1.00)
h For'each condition below SELECT the ONE (1) that will NOT have an;INOP trip. ; ASSUME all channels are unbypassed.
(1.0)
j,
.
a.
APRM channel-S has 4 LPRM inputs in "A" level; 3. LPRM inputs in *B": level; 3 LPRM inputs in "C" level and 1
LPRM-input in "D" level.
z Jb.
APRM_ channel 3-(with' function switch in COUNT) meter h
indicates 50L c '.' APRM channel 4 APRM mode' switch in POWER.
d.
APRM channel I circuit board removed.
'
,
'i
ANSWER:'
(1.00).
" a.'
['+1 '. 0 ]
~
.r
../ REFERENCE:
+
J
-1.
MP1: ~ Operator Training, Systems Volume 6, TX-1404A, Average Power Range Monitor System, pp. 9; 14, 15, 21 and 25, l0bjectives=4a and 14.
72.
KA Numbers 215005K104 (3.6) and 215005A203 (3.8).
I
,
e 215005A203 215005K104
._. (KA's )
.
L l
u
,
.
f
[
_(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- { y L
.
$
r Jb
<
.,.
,
-
-
_.
,
n s..
..
v
.
.
J,
. PLANT SYSTEMS-(40%) AND-PLANT-WIDE GENERIC Page 31-
- RESPONSIBILITIES (17%)'
"
- -
,
}
l
!
E QUESTIONi 04 (3.00)
L MATCH the major load in Column I with the-appropriate: 4160 VAC bus (
'
.in Column-II.' Each bus in Column II may be used more than once or
,
_
.
not'at all..
(3.0)-
.
F<4
. COLUMN I-COLUMN 11 l
O
- a.
LPCI Pump.lA 1.
14A J-b.
ESW Pump 101 2.
14B
'c; Core Spray Pump:lA'
3.
14C
- d.
' Emergency Condensat'e' Transfer Pump 4.
-140
'
_e.
RBCCW Pump 1A 5.
14E
f'
Service Water Pump ID 6.
14F
'
g.
Reactor Feed Pump IC 7.
14G h.
TBCCW: Pump 1A-1..
Recirculation Motor Generator 1B j.
iControl. Rod Drive Pump 1A
>
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
1,
,
,
_
,,
_ _,.
.
.
-
--_ _ _ _
____,__-___-__7_____.__
,
ji-:. 4 :.M * -
.
- .,.
,.e; PLANT SYSTEMS (40%) AND PLANT-WIDE GENERIC
.Page 32 r-1:,-
l RESPONSIBILITIES _(17%)
'-
'(ANSWER:
_04 (3.00)
-
a.-.
6-b.
6-
- i C.
6.
d.
,
e.
'5-
-
-
- f,
(
g.
t.
h ~.
.3
,
i,,
.!
-
j.
.(+0.3]'each
!
,
i REFERENCE:
L
.. l '.
MP1: Operator Training, System Volume 5, 4160 VAC Text, pp.
9 throughc13, Objective 8.
.
e 12.: -KA Numbers 26200lGG04 (3.5) and 262001G007 (3.5).
-
.'
r..
26200lG007 262001G004
..(KA's)
-;
!
!
.i (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
. _ = _ - _ - _ _ _ _ _ -.
_
-. -.
.-
.m
...
,,
.
' -
,
t y e, +,
,. i
,.
M
- PLANT ~ SYSTEMS-(40%) AND PLANT-WIDE' GENERIC Page 33
,l t-RESPONSIBILITIES (17%)--
!
.
!
U QUESTION: 05-(3.00)
-For'each' condition-in Column I SELECT the CORRECT action required
by:the Technical Specifications.from Column II.
The Technical
,LSpecifications.are provided. The actions in Column II'may be-
!
used more than once or not;at all. ASSUME the reactor is operating
.
t at.100% power.
(3.0)
COLUMN'l-
. COLUMN IL
$
a.
Isol ation' Condenser nd 1.
Orderly shutdown
.
EmergencyCondensatePIlmk initiated and the
-.both found inoperable, reactor coolant temp shall be less than -
b.-
'
LESW Pump 1A and.LPCI Pump 1C 330 deg.F within 24 hrs.
!
r both found inoperable, i
2.
Orderly shutdown f
c.
Core. Spray Pump 1B and LPCI initiated and the i
- Pump 1B both found inoperable, reactor shall be in cold-shutdown or refuel d.-
The Condensate Storage Tank conditon within 24-hrs.
.has currently 245,000 gallons
!
of usable water.
3.
Reactor Operations i.,
.
permitted only during
'
e.
The-Gas Turbine-Generator is the.suceeding 4 days.
>
found inoperable.
!
4.
Reactor. Operations is
-
,
i f. ~
Diesel Generator'and Core permitted only during
!
Spray Pump.lA both found the suceeding 7 days.
-inoperable.
5.
Reactor Operations is permitted only during the-suceeding 30 days.
6.
Power operation : hall i
be restricted to a traximum of 40% of full
power within 24 hrs.
7.
No reactor. shutdown required, t
.
-(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.
.
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-
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,
..
'l
,
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'
-PLANT; SYSTEMS (40%)-AND. PLANT-WIDE-GENERIC.
Page 34 Iu RESPONSIBILITIES (17%)
j g
- y
- y
- ,
'?
i ANSWERi
.05:
(3.00);
f cr a. < 1l
_
-
c b., ' 4
.
>
. c.
2-l
,
~d.
- X f'
.
e.~
$
' f.
,
d
[+0.5] each
.
-REFERENCE:
1.
?MPl:. Technical Specifications 3.5.
22.
KA Numbers 203000G005 (4.4), 207000G005 (4.2), 209001G005 i
(4.2), and 264000G005 (4.2).
<
L203000G005 264000G005 20900lG005 207000G005
..(KA's)
.. QUESTION:'06-(1.00)
SELECT the CORRECT statement that describes the primary purpose of the charcolLadsorber beds in the Augmented Off-Gas System.
(1.0)
i a. - Traps particulate matter.
.b.
Dilutes the hydrogen gas ccncentration, j
.c.
Adsorbs-entrained moisture.
-d.
- Delays fission products gases.
,
. ANSWER:-
06 - (1.00)
,
rd.
[+1.0]
(
.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
,
n w
'
.:
-
.
, y-
%...
,
..
o
,
. PLANT SYSTEMS ~(40%) AND PLANT-WIDE GENERIC Page 35 g
LRESPONSIBILITIES (17%)-
!'
,
[
REFERENCE:
'
le
.MPl:- Operator' Training, Systems Volume 4, A0GS Text, p.- 14,
-Objecti.ve 4.
,
l 2,
KA: Numbers 271000K406 (2'.9) and 271000G007 (3.4).
.i
'
p 271000G007
'271000K406
..(KA's)
g
,
,
- .
{;.-QUESTION:107 J
(1.00)
,
L
. henever both recirculation pump ( are in operation, pump speeds I
[
W
';-
~shall.be' maintained within
- ( 1.)
% of each other when power level:is greater than 80% and within (2.3
% of each
- l other when power level is less than 80%.
( 1. 0)--
,
'
,
!}
? ANSWER:-
(1.00)
. l.,
10%:
2, 15%
'[+0.5] each
,
-REFERENCE:
'
.- l.
-MPl: Technical Specifications 3.6.H, p. 6-14. -
02.
.KA Numbers 202001G005 (4.2).
i
- 20200lG005'
..(KA's)
i a
. -
i
l l
.(***** CATEGORY 6 CONT!NUED ON NEXT PAGE *****)
,,
3 -
j, 4;
l --<
PLANT SYSTEMS (40%)'AND PLANT. WIDE GENERIC:
Page 36-
,
,1
.
i;
~ RESPONSIBILITIES;(17%).
'i
.: -
i p i00ESTION: 08 (2.00)
-
LMATCH the functions in Column I with the appropriate component.
'
,
c
.of the Turbine Control System in Column II.
'
p j
> COLUMN'I COLUMN II.
.
!/
a.-
Provides backup overspeed
.l.
Pressure Control Unit-protection.
.
,
'
2.. Speed / Load Control Unit b.'
Performs the principal control intelligence for.
3.
Control Valve Relay'-
normal turbine control..
.
- c..
Converts.the mechanical.
linkage-signals from the 5.
Acceleration Relay speed relay.into hydraulic
. signals.
.6.
Intercept Valve Transmitters d.
Initiates rapid-closure of the turbine: control valves and-intercept
' valves;
' ANSWER:
.08 (2.00),
i
.
- a.
b-
-3
.
- c. -
e-
?d.
'5'
i
[+0'5] each
'
.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
- - _ _,
.
. -.
.
.. -
.
n 7,
,
.
,
,
,
I,
- &
+*'
Ie L PLANT SYSTEMS (40%) AND: PLANT-WIDE GENEitlC.
Page_37_
,
RESPONSIBILITIES (17%)'
(L
!
_ REFERENCE:,
1.
MPl: Operator Training, Systems Volume 3, Turbine Generator Text, p. 16-23 Objective 28.
T.
KA Numbers l241000K404 (2.8), 241000G007.(3.5) and 241000K403 (3.1).
- 241000K403 241000G007 241000K404
..(KA's)
(QUESTION:09 (1.00).
The mechanical vacuum pump is' not to operate above 5% thermal power.. SELECT the CORRECT reason.
(1.0)
a.
The pump is not designed to handle the additional load, b.
An explosion could occur inside the pump.
c.
The.-stack radiation release rate limit will be exceeded.
d.
The-pump is no longer required.
g
-!
1 ANSWER:
09-. (l'. 00)
U b.,
[+1 ~. 0]
.
,
REFERENCE:
1.
- MP1
- _ Operator Training, Systems Volume 4, CAR Text, p. 34, Objective 18, 2.-
KA Numbers.256000K409 (2.8).
,
!
,
256000K409
..(KA's)
,
L QUESTION: 10 (2.00)
'
The S utdown Cooling Sy(etem is designed to cool (3.)
the reactor from
'[1
._deg F to'
2.T deg F within hoyrq in a
,
l
- controlled manner and maintain reactor temperature at ( 4.)
(2.0)
.
n l
l l
,
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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- .,
- -
.
,
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L
.
.
p[ '
PLANT SYSTEMS (40%) AND' PLANT-WIDE GENERIC Page 38
RESPONSIBILITIES (17%).
'
si
>
c:
-
'
.
s
[ ANSWER:: 10 (2.00)
l
{
1 '..
280-
,
2.
125 o
t
[
- 3'
,
I-4.
125 e :.
.[+0.5] sach;' all temperatures shall be +/- 5 degrees.
E
!
REFERENCE:
>
1.
MPl: Operator Training, Systems Volume 2,.SDC Text, p. 1 f
.
0bjective 1, 2..
KA Numbers 205000G004 (3.7).
- l
-
205000G004
..(KA's)
[ QUESTION:-11 (1.00)
'
-The parameters for automatic initiation of the Automatic Pressure
~ Relief System are present and the l'20 second timer.is running.
.
l
' SELECT.the action.that..ill NOT reset the 120 second~ timer.
(1,0)
'
h
- a Drywell pressure drops below 2 psig.
-b.
Interruption of power tc the APR logic channels.
,
c. _. Reactor level is restored above -48 inches.
APR Tiber d.
Operation of the Ti 0_0 thy Reset pushbutton,
,
ANSWER:
(1.00)
,
a.
[+1,0]
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.
- - - -
_
,
_
,
j 3;,; :.' i
PLANT SYSTEMS-(40%) AND PLANT WIDE GENERIC-Page 39 S
RESPONSIBILITIES (17%)
REFERENCE:
' 11 MPl: Operator Training, System Volume 5, APR Text, p. 15,
- Objective 11.
.
~2.:
KA; Numbers 218000K403 (4.0) and 218000K501 (3.8).
,
"
218000K501 218000K403
..(KA's)
' QUESTION: 12 (3.00)
Concerning the Radwaste System, MATCH the plant inputs in
.. Column-1 with.the receiving tanks in Column II.
Each tank
.'
can:be used more than once or not all.
(3.0).
u COLUMN I COLUMN <Il (Inputs)
(Tanks)
a.
Reactor Building equipment drain-1.
Waste Collector Tank
'
b tank discharge.
2.
Waste Surge Tank-b.
.Drywell floor drain sump.
3.
Floor Drain Collector c..
Turbine Buildin floor drain Tanks (A,B,C, & D)
sumpBpumM81 harpe.
( Am Gavpresse Are4>-
4.
Decontamination d.
Drain water-from the Torus Solution Tanks
- e; Discharge flow from floor
.
drain' collector pump.
Lf.
Radwaste' floor drain sump.
'
i l
i h
"
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
{
i
!
'
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m.,
,
,
,
,
,
.
3 + e,.
'
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' PLANT.SY' STEMS.(40%) AND PLANT WIDE GENERIC:
Page.40 sj) RESPONSIBILITIES (17%)
,
E
)?
!
.
.
-l LANSWER:-. 12
'(3.00):
,
,
ij ta..
.b.
3-c.
4-
.i
.
'd.-
1:
t f3 e.
'
. f.
-3'
[+0.5) each i
- REFERENCE:
l '.
- MP1
- Operator Training,; Systems Volume 2, RW Text, pp. 65, 69, 88, and 99, Objective 10.
' 2. -
KA Numbers 268000G007 (3.1),
b 268000G007
..(KA's)
l'
. QUEST 10N: 13 (2.00)
-
.STATELFOUR (4)'~ indicators that a fuel assembly is correctly iorientated in the core.
(2.0)
n
.,
t
.
>.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
p.,
-
'
.
c
<,-
.v
<
(
.
-
-
.
-
D
.(PLANT = SYSTEMS'(40%) AND' PLANT-WIDE GENERIC Page 41'
.i (-
RESPONSIBILITIES.(17%)-
'
+
l ANSWER:
13-(2.00)
y l '.
E All: handle identifications are readable from cell' center
outward. -[+0.5)-
F-
- 2.. ~All orientation lugs are pointing towards the center of the
cell. _[+0.5)
h 3.
All channel fasteners are adjacent at= the center of the cell.
U
'
[+0.5]-
L4.
Spacer buttons are touching at the control rod gap.
[+0.5]
i REFERENCE:-
t-
-
-l.
Operator Taining,'Sytems Volume 1, Reactor Fuel, p. 14,
Objective 1..
~
2.' -
KA Numbers 234000K505(3.7).
ti
'
234000K505
..(KA's)
.
-;
L l
J (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
I
p.
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!
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,
[;
PLANT. SYSTEMS (40%) AND PLANT W10E GENERIC-Page 42
.
RFSPONSIBILITIES (177.)
'
t-
..
L-
!
F
~
(2.00)!
LQUESTION: 14
'
- MATCH the~ events in Column I.to the appropriate classification
in-Column II in accordance with EPIP 4701-1, " Millstone Emergency
,
- Action > Levels," (attached).
Each classification may be used more than once or not at all~.
(2.0)
- !
- COLUMN.I COLUMN II
>
.(Event)=
(Classification)
.
a.
A spent fuel bundle.has been 1.
General Emergency dropped in the Spent Fuel Pool..
"..
- Area Radiation Monitor reads.
2.
Site Area Emergency
.
.offscale.
'
3.
Alert-b.1 The Recirculation System-Sample'line.has broken in the 4.
Unusual Event
'
Reactor Bu,ildingg,'Cr= p !
..3..__
ca
.
>-.
.Ed W sr 6e.lUIIIE2.
c;
. Security has notified th'e
' Control Room of a: bomb threat.
"di JSafetyJ Releif Valve MS-3-SS l
has inadvertentely opened
,
and cannot be reclosed.
,
,
- ANSWER
= 14 _
(2.00)
L
-
-2oJ3--
a.
- b.. J-2.
[L
.~c.
l^
d.
[+0.5) each
,
l l
l.
j.
>
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
-
-
-
-
-
-
g,.< ;
-
.
.
.
.
N
[:f )... : 9 ^ : 9
- PLANTL SYSTEMS: (40%) AND PLANT-WIDE GENERIC Page 43
<
RESPONSIBILITIES-(17%)-
<,
.
. REFERENCE:
1.'
MPl:' EPIP 4701, " Millstone Unit.0ne Emergency Action Levels,"
,
Form 4701-l',
p'. 1-4.
- 2.
KA Numbers 294001All6(4.7).
,~
.
e
,
L
'294001All6.
..(KA's)
l
,
,
..
!' ; QUESTION: 15-(2.00)
g
' Concerning station ta'gging, MATCH tag description Column I with the
'
, tag color in Column-11.
(2.0)
-
COLUMN I COLUMN 11
'
(Description)
(Color)
i l,
' a.
Tag placed on. equipment as a 1. ' Red Tag.
- L-caution;against improper operation where other tags do 2.
Blue Tag i-not apply.
3.
Green Striped Tag
' b'.
Prevents.re-energizing if.
(-
equipment trips.
'
4.
Yellow Tag.
c.1 Indicates that equipment is to 5.
White Tag beJenergized.by indj"'tr' Akdkoelgmog [,,,y _Q f-j%M/ulto whom the tag;was issued.
.d.-
Prevents operation of equipment that would endanger personnel or equipment.
"
ANSWER:
15-(2.00)
a.
.b) 13
.c.
e',
- d.
i
[+0.5]-each p
l:
p V.
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
~
.
. -
.
m/
.
.
-
y; Ti.
e: m* -
' /
PLANT. SYSTEMS'(40%) AND PLANT-WIDE GENERIC Page 44
.
RESPONSIBILITIES (17%):
L
_
iREFERENCE:
'
- ).<
MPl: ACP-QA-2.06A, p. 3-4,." Station Tagging."-
il
-: 2.
KA Numbers-29400lK102 (4.5).
,
I 29400lK102
..(KA's)
.
!
,
' QUESTION: 16 (2.50)
. MATCH the description in Column I with the dose in Column II.-
'i Dose may be used more than once or not at all.
(2.5)
l COLUMN.I'.
COLUMN 11
'
(Description)
(Mrem)
>
a..
> Federal' limit per calendar 1.
300 quarter.for skin of whole body. quar.terly exposure is 2.
500 i
known.
3.
1000 Jb.
Federal ~ limit per calendar quarter exposure to the 4.
1250
- extremities - quarterly exposure'is known.
5.
1500 c.
Federal limit to whole body 6.
3000
~
Jper calendar quarter if
' lifetime exposure history is.
7.
5000
.known via Form NRC-4..
.
.
8.
7500-Ld.
Administrative limit to
. radiation work ~per calendar 9.
12000
-
quarter,per site (e.g.
Connecticut Yankee and 10.
18750
Millstone). 5 =p: = r m 4*
Le hoon ramm e + n A
'
ME K a 4 w uieT i;idnau l; slo.3 JAdministrative limit to e.
i female, radiation worker who deciares herself pregnant.
Limit applies to the duration-of.the pregnancy.
L 1. <l'
.
l.
!
,
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
t
p,,;
.
-
' x 3. 4. ~ c e
.
,
,
,
-
t
'
- Y PLANT., SYSTEMS'(40%) AND PLANT-WIDE GENERIC
,Page 45 RESPONSIBILITIES-(17%)'
,
w
?
.;
'
! ANSWER:
,
j
.
..
,
.
.. L(2.50)
J
- .'a.
4
.b.
.10.
t
.'
c.;
'
d'
l e.-
' [+0.5] each
,
REFERENCE:
,
- .
l..
-MPl: SHP4 4902, " External. Radiation Exposure Control and
"
..
Dosimetry. Issue," Rev. 11, pp. 6-9 and 13.
'
-
~2.-
KA Numbers.29400lK103 (3.8).
29400lK103-
..(KA's)
~
-QUESTION' 17j (1.00)-
Concerning the Millstone Station Emergency Exposure Guidelines,
- WHAT.are the exposure limits in the.following cases?
,
'a.
Lifesaving actions such as search and removal of an injured-person,-
rem.
'
(0,5)
b.
Necessary entry-into a hazardous area to protect facilities and ' eliminate future escape-of effluents, rem.
(0,5)
ANSWER:
17'
(1.00)
a.
75 [+0.5]
b.
25 [+0.5]
,
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
ie
m7.
-
.
o g(;
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s
,
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r
'(.;c PLANTISYSTEMS (40%) AND PLANT-WIDE GENER'C Page 46:
!
t.
L'. RESPONSIBILITIES (17%)
!
e
.
l1 ' REFERENCE:
- m (1 1.
MPl:
SHP 4202, Rev. 11,'p. 15-16.
.
.
[t L2..
KA Numbers 2$4NIK103 (3.8).
4-k-
..
29400lK103
..(KA's)
-
[
t
.
.;
,
-;
,-
l t
l (***** END OF CATEGORY 6 *****)
i-(**********
END OF EXAMINATION **********)
j
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TEST CROSS REFERENCE Page
.1-(OUESTION.. VALUE. REFERENCE
'
!
- j
.
.
01.
1.00
. 90001 H
3.00'
90002, d
1.00 90003
'
?
1.00 90004'
'
705 1.00.
90005.
06L 1.00 90006 L
3.00 90007 L
' 08 2.00 90008 i
h" 09.
3.00 90009 l
'
p
'l.50-90010 I
.11 3.00 90011 l
o 12'
2.00
- 90012
13'
'1.00 90013-
2.00'
90014
-
i
'[
LIS 1.00 90015 l-16 1.00 90016.
!
L L17 2.00 90017
L
.18-2.00 90018 l
.
- (
19.
- ).00 90019 l
oLt
.20 3.00 90020'
r h"
2.00 90021 i
'
2.00 90022 l
'23 '
2.00
'90023
i
3.00 90024 l
^
1.00-90025 i
p 26-3.00L 90026
[
......
48.50
!
01
1.00'
-90029 i
'02 3.00 90030 i
- 03 1.00'
- 90031
!
3.00 90032 E
05'
3.00 90033-
u
.06-1.00 90034
!
F
- 07
1.00 90035 08~
2.00 90038 L
' 09 -
1.00-90039 i
L 10, 12.00-90040 l
b
~3.00 90042
!
1.00.
90041 i
b
'12
,
- 13 2.00 90043 I-14 2.00 90027 i
A 15.
2.00 90028 i
16.
2.50 90036 l
. I7-1.00.
90037 c
.
......
'
~
j-31.50
......
,
+
.......
.
80.00-i
!
b
.,
._
_
. -
. - -
. _..,. _,.... -__, _ _
_
-, _ -,
, _. _ r,
f
-
. _ _
'
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!
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.-
N=
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.q i
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<
~110
'
-
\\
'
120 fl
'
130
]
%
- 350+F
'
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"
A
,,
E-100
'
~ **
'
- 200*F-170
-
N A
- 150*F
'
.
,
~ 100*F
,
3go j-190-e00
200 400 000 000 1000 1200
.
RPV Pressure (psig)
!
Figure 4 - Correction for TAF
,
.
..
.
..
_
_.,, _.
_..
. - _ - _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _
__ _ __
'
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-
-170 i
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-180
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~190
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-200-210
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-220 IW 150 200 250 300 350
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Drywell Bulk Temperature (F)
Figure 5 - Correction for % Core Height
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ATTACHMENT 7 b
General offices o seiden sheet, Berlin, Connecticut
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U.I.U2$U,$iUrIN P.o. Box 270
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.a uws we* mu* co*=a H ART F oRo. CONNECTICUT 051410270 L
L J 0 % O",", C % ""
( W W 5'5 *
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December 5. 1989
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i Mr. Robert M.
Gallo Operations Branch
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Division of Reactor Safety l
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O.
S.
Nuclear Regulatory Commission l
Region 1
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475 Allendale Road
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King of Prussia,'PA 19406 j
Reference:
Millstone Nuclear Power Station Millstone Unit one
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Docket 50-245
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operator Requalification Exam m
Dear Mr. Gallo
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Enclosed are the results of the Millstone Unit one evaluation of I
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the operator Requalification Examinations given the weeks of f
October 16 and October'23, 1989.
A summary of the examination
results is provided along with the individual results for each i
operator examined.
These examination results indicate that the
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Millstone Unit One Requalification Program is satisfactory
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overall and that operator retraining is needed in specific areas.
We are taking the necessary corrective actions based on the
. examination results.
The individuals who failed the examination have been removed from licensing duty,,and an upgrade program is l
being planned.
Details of the proposed upgrade program are
enclosed for your review.
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i in our view, the new examination process worked well.
The NRC l
Examination Team interfaced effectively with the Millstone Unit t
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One personnel to administer the examination.
The professionalism of the NRC Examination Team was noted by both the Facility Examination Team, as well as the operators taking the exam.
This
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helped in significantly reducing the already high level of stress the operators were under during the exam process.
If there are any questions regarding our evaluation of the examination or our corrective actions, please contact Mr. Ray
Lueneburg, Supervisor, operator Training, at 203-447-5610.
l Very truly yours,
i M
Steven /t. 'S ace
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L.
Station Superintendent Millstone Nuclear Power Station
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'SES:GLS sik
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Enclosures
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10/26/89 MILLS 10nt tMIT ONE ES 601 EK M RESULTS 1.- overall Examination Results*
RO SRO Total Pass / rail Pass / Tail Pass / Tail Written 4/0 16 / 0 20 / 0 Simulator 4/0 14 / 2 18 / 2 JPM 4/0 16 / 0 20 / 0 overall 4/0 14 / 2 18 / 2 2.
Dynamic Simulator Crew Results*
OPS B OPS C OPS E Staff B Staff C IGC S
S S
S S
racility S
S S
S U
- tele These results indicate facility grading. Refer to Section 3 for NRC grading of exam.
i