IR 05000293/1990004
| ML20012D462 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/05/1990 |
| From: | Conte R, Todd Fish NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20012D461 | List: |
| References | |
| 50-293-90-04OL, 50-293-90-4OL, NUDOCS 9003270366 | |
| Download: ML20012D462 (94) | |
Text
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REGION I
Examination Report No.: 90-04 (OL) < ' Facility Docket No.: 50-293 .; , , . I Facility License No.
DPR-35 i !' Licensee: Boston Edison Company '
! ' Facility: Pilgrim Nuclear Power Station w l Examination Dates: January 29 - February 2,1990 . l l Examiners: Todd Fish, Senior Operations Engineer Carl Sisco, Operations Engineer Nick Conicella,. Senior Operations Engineer
I' Chief Examiner: / Otl/ Mv
90 ! 'Tob6 Fis, Sen40 A per(tT &s Engineer 'date '
'- Y 3nllanl 3f5 f90 . Approved by: l [v - Richard onte 24flef BWR Section / dater
Operatio s-Branch, Division of Reactor Safety i EXAMINATION SUMMARY
Initial written examinations and operating tests were administered to ten Senior , , Reactor Operator (SRO) Candidates (6 Instant SR0s and 4 Upgrade SR0s). All i ' L candidates passed the written and operating examinations and their performance indicated good preparedness for the examination.
? However, 'a number of specific strengths and weakness were identified as a result t of examination grading.
These were documented for licensee self-review and , remediation as appropriate. Also, a number of discrepancies were noted with the licensed-reference material submitted to the NRC. These comments are also provided to the licensee for self-review and corrective actions, as appropriate.
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, ? 9003270366 900313 PDR ADOCK 05000293 - '
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t C . , t DETAILS 1.
INTRODUCTION AND OVERVIEW , The NRC examiners conducted this initial examination for ten Senior Reactor Operator (SRO) candidates (6 Instant SR0s and 4 Upgrade SR0s).
The examinations were administered in accordance with NUREG 1021, Rev. 5, dated January 1,1989, Examiner Standards (ES). The results are summar- ! ized below.
l l SR0 l l l Pass / Fail l l l l
l l l Written l 10/0 l l Operating l 10/0 l
I I l l l i l Overall l 10/0 l l l l , 2.
EXAMINATION RELATED FINDINGS / CONCLUSIONS The following is a summary of general strengths and deficiencies noted during the administration of the tests. This information is beirg provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
2,1 Operating Portion Strengths Communications between all crew members; good teamwork - -Response to alarms - Procedure adherence ' - Deficiencies i ,Several applicants were unaware of the location of the tools and - ! equipment required to perform E0P Procedure 5.3.23, " Alternate ' Control Rod Insertion."
l E0P Procedure 5.3.20, " Alternate Boron Injection," requires that the - l high Reactor Building Closed Cooling Water (RBCCW) temperature trip of the Reactor Water Cleanup (RWCV) pumps to be defeated by lifting leads in an electrical junction box. The location of this junction l
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! . box is not specified in Procedure 5.3.20.
This procedural step in-adequacy caused an applicant to spend over an hour trying to locate this junction box.
F - ! 2.2 Written Portion Strengths , ( All questions that are not listed below as a generic deficiency were answered correctly by at least 70% of the applicants, indicating satis-factory knowledge as a group in those subject areas.
. Deficiencies The questions listed below were answered incorrectly by at least 3 of the 10 applicants.
Question 003: Ability to distinguish between a level 1 and level 2 Radio- ' logical Occurrence Report (ROR) ' 007: Knowledge of independent verification guidelines of Proce-dure No. 1.3.34 (Conduct of Operations) e 010: Ability to utilize technical specifications to determine ~ if active or tracking LCO's ere to be entered if an IRM and an APKM oecome inoperable.
043: Knowledge of the normal current reading for a SLC squib valve continuity current meter.
050: Knowledge of which RPV level instruments have the majority of their reference legs located external to the drywell.
052: Knowledge of the maximum generator load allowed if one stator water heat exchanger is removed from service.
065: Knowledge of the fact that E0P caution statements do not contain action statements.
074: Knowledge of the bases for performing E0P-07 (Alternate RPV i Depressurization) if area temperatures exceed the maximum ' safe operating values in two or more areas of the secondary containment and a primary system is discharging into the secondary containment, , , 2.3 Other Findings The licensee sent to the NRC training lesson plans and learning objectives that dealt with the plant's Hydrogen Injection System.
The NRC prepared
several questions on the written examination dealing with this system.
During the pre-examination review conducted on January 25, 1990, the licensee's ;epresentatives stated that the applicants had not been trained l
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) i ' on this system and requested that these questions be removed from the written examination.
The licensee's representatives stated that training i had not been conducted on this system since the system was a recent plant , modification and was still in the pre-operational testing phase. The only l training that had been conducted on this topic was a general overview of ' the theory of hydrogen injection corrosion control. The NRC removed one , of the two questions that dealt with hydrogen injection system operation " and_ replaced it with a question from a different system.
The NRC did not remove the second question since the question dealt with the theory of hydrogen injection corrosion control and the applicants did receive training on this topic.
The reason for this confusion was that the licensee did not adequately review, in part, the materials sent to the NRC for preparation of the licensing examination. Although the hydrogen
injection system lesson plans were not taught as part of the initial operater license class, the licensee has recently incorporated these lesson plans into the operator license requalification training program; therefore, all currently licensed operators will soon be trained in the operation of this system so that they will be able to operate this system once pre-operation testing and system turnover to operations is completed.
Further, during the pre-examination review on January 25, 1990, the licensee representatives requested that a change be made to question number 007 of the written examination. The NRC incorporated the change and the question was given to the applicants as the licensee representa-tives requested. Af ter the written examination was administered, the
licensee presented to the NRC formal comments stating that stated question number 007 was not clear as written and should be changed.
The change the licensee proposed after the examination was exactly as the NRC had written , the question in the first place prior to the pre examination review.
(See Attachment 3).
Also, lesson Plan 0-RO-02-04-02, Condensate and Feedwater System, has-conflicting information on pages 21 and 35 regarding whether the steam flow signal or the feedflow signal is the most inaccurate input to the 3-element feedwater level control system during low power operation.
(See Attachment 3).
3.
Meetings l-A pre-examination review was conducted at the NRC Regional Office on i January 25, 1990, with members of the licensee's staff identified in Attachment 5.
Comments on the written exam were discussed.
An exit meeting was conducted on February 2, 1990 at Pilgrim Nuclear Power Station with members of the licensee's staff identified in Attachment 5.
[ The Chief Examiner discussed observations made by the examination team and l thanked the staff for their strong support during the exam week.
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' ' ' ' ' - , ph # ' / 15-(; -ru O / ATTACHMENTS, ' , , .. j i I_ , - ( i , '1.: Examination and Answer Key - 2, Facility Comments on Written Exam . .,. - >3.- Resolution of. Facility Comments b.
4.-,
- Simulator Fidelity Report 5.
Persons ^ Contacted ' ' ^ ' "
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V' AT7ACHMcNT 1 ,. ... U. S. NUCLEAR REGULATORY COMMISSION '~ ~~ . r SENIOR REACTOR OPERATOR LICENSE EXCMINATION ! REGION
? L' FACILITY: Pilgrim 1 ! REACTOR TYPE: BWR-GE3 , h . DATE ADMINSTERED: 90/01/29 ! > t ! ): CANDIDATE: I [ ll,IC4 C 706' r > wopy if INSTRUCTIONS TO CANDIDATE: , ,. ? ' Points for each Question are indicated in parentheses after'the question.
To
pass this examination, you must achieve an overall grade of at l ea s t 00*/..
Examination papers will be picked up four and one half (4 1/2) hours after ' the examination starts.
':
l l NUMBER l TOTAL l CANDIDATE'S l CANDIDATE'S l l l QUESTIONS l POINTS l POINTS l OVERALL l l l l l GRADE ( */. )
- _____________;__________;_______________;_______________.
l l l l l ' l 100 l 100.00 l .l l l l l l l l l l l l , e
All work done on this examination is my own.
I have neither given nor received aid.
T
Candidate's Signature r
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, " NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS ' . During the administration of this examination the following rules apply: 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
This must be done after you complete the examination.
3. Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the-examination ' 'N room to avoid even the appearance or possibility of cheating.
4.
Use black int or dark pencil only to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.
, ! 6.
Fill in the cate on the cover sheet of the examination (if necessary).
7.
You may write your answers on the examination Question page or en a separate sheet of paper.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
8.
If you write your answers on the examination Question page and you need more space to answer a specific question, use a separate sheet of the paper provided and insert it directly after the specific question.
DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.
9.
Print your name in the upper right-hand corner of the first page of answer, ' sheets whether you use the examination question pages or separate sheets of paper.
Initial each of the following answer pages.
10. Before you turn in your examination, consecutively number each answer sheet. including any additional pages inserted when writing your answers on the examination Question page.
11. If you are using separate sheets, number each answer and skip at least 3 lines between answers to allow space for grading.
12. Write "Last Page" on the last answer sheet.
. 13. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer.
Write it out.
I E 14f*.The,coint v31u] for Cach CuOatiCn 10 indicSt;d in COrCnth;OO3 Oft;r the Cu;OtiCn.
Th] Cmrunt of blcnk Cp3c? Cn Cn OxCminOtiOn Cu?OtiCn pig] 10 I "
(., NDT cn in"Jication of th) C;pth of CnOw?r PCQuirCd.
t 15., Show all calculations, methods, or assumptions used to obtain an answer.
i 10. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION ' AND DO NOT LEAVE ANY ANSWER BLANK.
NOTE: partial credit will NOT be < I given on multiple choice questions.
17. Proportional grading will be applied.
Any additional wrong information , that is provided may count against you.
For example, if a question is worth one point and asks for four responses, each of which is worth O.25 i points, and you give five responses, each of your responses will be worth ' O.20 points.
If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.00 instead of 1.00 even though you got the four correct answers.
18. If the intent of a Question is unclear, ask questions of the examiner only.
.
- 19. When turning in your examination. assemble the completed examination with examination Questions, examination aids and answer sheets.
In addition.
' turn in all scrap paper.
20. To pass the examination, you must achieve an overall grade of 80% or greater.
21. There is a time limit of (4 1/2) hours for completion of the examination.
(or some other time if less than the full examination is taken.)
22. When you are done and have turned in your examination, leave the examin- . ation area as defined by the examiner.
If you are found in this area ! while the examination is still in progress, your license may be denied or revoked.
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1.
, . t Y A"' QUESTION: 001-(1.00) For an SRO' license to remain ac'tive, one must. stand a minimum number of watches-per calendar quarter. SELECT the one choice that would satisfy-t requirement-for' maintaining an SRO. license active.
a '. Four'B-hour shifts as an RO and four 8-hour shifts as an SRO \\' b. Six 12-hour shifts as an RO and two 12-hour shifts as an SRO il c. Two 12-hour shifts as an RO and four 12-hour shifts as an SRO
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d. Two 8-hour shifts as an RO and seven B-Meur shifts as an SRO' N QUESTION: 002 f.).OO) .' Select the order concerning operation of components which if stated j ve r ba l'l y is in accordance with SI-OP.OO6 (Operations Communications).
' a. Start alpha RHR pump b. Place alpha-RHR pump in service c. Start RHR pump alpha d.
Take alpha RHR pump to-start ' . ,
4 QUESTION: 003 (1.00) . SELECT the item which would NOT be considered a Level 1 ROR as defined-i! Procedure No. 6.1-209 (Radiological Occurrence Reports).
a. A major RWP violation b.
Radiation exposure that would require a 24 hour report to the. NF! c. Radiation exposure that would require a 30 day report to the NRC d.
A radiological event with potential for a press release ! ! !
___ _ _ _. _ _ _....... _ -- 1-- '4 '. Pegar S , . , . QUESTION -OO4 (1.00) SELECT the statement regarding the requirements of a locked high radiation = area which is in accordance with Procedure No. 6.1-012 (Access Control to High Radiation Areas), a. A single individual can have access to work if he is a qualified RP technician b.
'P' keys may be issued to operations personnel with approval from the Chief Operating Engineer c.-Entry is allowed for two NPEOs providing a radiation monitoring device which continually indicates dose rate is provided d.
Independent verification that the door is closed and locked after work is complete is required QUESTION: 005 (1.00) 'The 'D' SLC pump must be tagged to repair leaks on the pump discharge block. The pump will require disassembly. SELECT the method of tagging that Procedure No. 1.4.5 (PNPS Tagging Procedure) directs to be used for this repair, a.
A yellow tag should be placed on the common control switch to allow . operation of the 'A' SLC pump b.
A red. tag should be-placed on the common control switch but the 'A' SLC pump could be operated if needed c. A red tag should be placed on the common control switch but neither the 'A' nor the 'B' SLC pumps can be operated d.
Since the control switch is a common switch, a tag is only required on the motor breaker not the control switch .
. _.. _. _ i > r i pogg-6= ,u 1*'.{ -*- . :.. < ' GUESTION -006 (1'.00).
' ! The 'AYlRHR. pump has,been tagged out for motor replacement. SELECT.the person who-15 ultimately responsible for assuring the tagging-is proper and 3*
'. .ttvN conditions:are safe for work to commence as stated in Procedure No.
2"; 1.4;5 (PNPS. Tagging Procedure).
. a. The WE since he is always ultimately responsible for personnel
safety ] o.:The NOS that approved the tagout , -( c.- The operator that performed the independent verification d.'The individual for whom the tagout is applied - , QUESTION: 007 (1.00)
l Procedure No. 1.3.34 (Conduct of Operations) gives guidelines for performing system configuration verifications and independent verifications.. SELECT the statement which is NOT in accordance with the guidelines of Procedure - No.
1.3.34.
a.
Lifted-leads and jumpers ~ require independent verifications b. Two' individuals can be' permitted to work together when repositioning valves , l c. Valve positions for an.inacceesible valve is first verified by j checking. isolation of the power supply to that component d. If'a' component is inaccessible and a tagout-already exists for that - component, the independent verifier can use the existing tagout to determine position -
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QUESTIONt: 000. (1.00) Procedure No. 1.5.5 (Cutting, Welding, and Hotwork Fire Safety) has specia-requirements associated with TIG welding in the drywell. SELECT the reason why TIG welding-in the drywell is of special concern.
a.
TIG welding may cause spurious ESF actuations . b.'-J T he d ry we l l does not contain automatic fire suppression c.
The drywell does not contain automatic fire detection d..Appendiv: 'R' of 10 CFR 50 requires special requirements when TIG-welding.in containment structures OUESTION:-OO9 (1.00) A reactor shutdown is in progress due to unidentified leakage in the drywell. The drvwell is still inerted and-air purging is about.to commencc in accordance with Procedure No. 1.4.12 (Primary Containment Entry), SELEC the minimum oxygen concentration that must be established at all sample points prior to allowing the initial drywell entry.
a.
15%' b.'19% c. 21% d.
25% . .
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- QUESTION LO10' (1.00)
[ 'The reactor is at 100% power with all equipment operable when an I&C L supervisor reports that after reviewing surveillance data, the power-supplies'for APRMA' and IRM 'E' need replacement and these two instruments are technically inoperable. In_accordance with SI-OP.OOOB (Limiting Conditions of Operation Log), SELECT the action that-is required.
Y a. An active LCO is entered for APRM 'A' and an active LCO is entered " it for IRM 'E' [ Y " \\ b.
A tracking LCO.-is entered for APRM 'A' and a tracking LCO is-entered for IRM
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An active LCO is-entered for APRM 'A' and a tracking LCO is entered ' for IRM 'E' d.
A-tracking LCO is entered for APRM 'A' and an active LCO is entered f for. IRM 'E' K QUESTION: 011 (1.00) . SELECT the item which is NOT considered a temporary modification in accordance with Procedure No. 1.5.9 (Temporary Modifications).
~ a. Blank flange on a reactor-building ventilation damper b.
Floor drain plug installed in the reactor building c.
Gagged relief valve on an RHR heat exchanger d.
Jumper installed to' bypass MSIV isolation per EOPs . E _ _ _i-W i . -.... - -.. -. - - -
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,. , QUESTION JO12- (1.00) i ^ l.
_ Emergency. planning zones. consist 1of the Plume Exposure EPZ'and the i , ' Ingestion _ Pathway EPZ.' SELECT: the area dimenslons for these-EPZs.
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Plume' exposures 2 mile. radius t InOestionipathway 5 mile radius = $ b.. Plume exposure 5. mile radius ! - Ingestion. pathways:25 mile radius.
" . k c. Plume exposure: iO' mile radius i Ingestion pathway:' 50 mile radius ' ' s d.. Plume exposure:i5 mile radius ! Ingestion pathway: 75 mile radius ' , i
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- OUESTION:"O12
'(1.00).
t i SELECT the' emergency action level whose definition-is: "a-radiological ,. , release is likely-but no core degradation is indicated."
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' a.,Unusualievent '
- b. Alert c
. _ c.fSite' area emergency , ' d.
General; emergency, i
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QUESTION 014 (1.00) l SELECT the emergency' action level based on the following information .The reactor was-at 100%. power.when an }&C-technician inadvertently ' initiated HPC1 during-the performance of a logic surveillance test. The operator was quick to diagnose the problem and manually tripped HPCI. The- -pump discharge pressure peaked at 850 psig~before being. tripped.
- 6.
Unusual Event b.
Alert
4 c. Site area emergency l4- 'd.
None OUESTION: 015 (1.00) SELECT the emergency action level based on the following informations The reactor was at 100% power when a break developed in the torus which . caused torus water level to decrease.-Torus level was stabilized at 85 inches but could not be raised any higher. The reactor was manually scrammed and depressurized.
a. Unusual event-J b.
Alert c.
Site area emergency =d.
None-
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% 1 .. . CUESTION: 016: (1.00) , , . SELECT the, emergency action level based on the following information: The1 reactor lwas at~100% power when..while tak'ng-log readings,'the operator.
~ i ' drywell floor drain sump leakage increased from:1.5 gem to 6.0 noticed that gpm. The leakage has been,at 6.0 gpm for the past-four. hours-andcappears toc-ba:stwady.. - a.
Unusual event h b.
Alert c. Site area emeroency d.
None
. ObESTION LO17 - ( 1. O'O ) ) . . The process computer has various "on-demand" (OD) programs that the
operator can repuest;from the operator's console. SELECT the-OD program thatsis1NOT' demandable from the operator's console.. ,. 0D-3 core' thermal. power and APRM calibration - a '. b..OD-7 control rod position c.- DD-8,LPRM readings t d.;OD-19.LRPM fast scan- ! f 'l i l , ! a b cy-M
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p; a~ F' Pcg3 12 , 3 :.x. m . L ~QUESTIONI 010' (1.00) . SELECT-the order of-events that occurs when starting a recirculation pump p ' motor generator set.
a. Drive motor breaker shutst incomplete sequence timer starts: I exciter motor field power shifts to generator outputt exciter field UV protection is inserted b. Drive motor breaker shutst incomplete sequence timer starts; exciter field UV protectiors is inserted; exciter motor field power
shifts to generator output s c. Incomplete sequence timer startst drive motor breaker shutst.
. exciter. motor field power shifts to generator outputt exciter field-UV protection'is inserted ,. y,; d. Incomplete scQuence timer startst drive motor breaker.shutst exciter field UV protection is insertedt exciter motor field power shifts'to generator output-QUESTION: 019 (1.00) ' The plant is at 100% power when MG set lube oil pressure for the 'A' recirculation pump decreases to 22 osig and remains steady. SELECT the 'cutomatic action that will occur-to the 'A' recirculation MG set.
a.' Drive motor breaker trip only b.
Drive motor breaker and field breaker trips c.
Scoop tube lock only L .d.
No automatic-action since only the start permissive is not met and the recirculation pump is already running , .v I
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' ~ .. QUESTIONic020' (1 00') . C ncerning'the, reactor. vessel internals. SELECT the. component which . provides for-adequate core-flow to the high powered' fuel bundles.
a.. Lower tie' plate , - b.IFuel support.' piece - c.. Fuel? channel.
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Stu'b ^ tubes , QUESTION O21 (1.00)
m 1. ! Blade guidesgare usedLduring refueling operations when removing or lg ~. ' installing; af f ueli cell. -SELECT-the statement which describes the - installation of a blade guide.
~ -a..A blade guide:is installed to support the control rod'when it-Is , , fullyJ-insertedJand l ' two face-adjacent bundlest of a' cell are removed-L.
b. A blade guide ~is installed to support the control rod when it' is - . fully' withdrawn'and two face-adjacent bundles of a cell are removed - o' c. A bladesguide.is. installed to support the control rod when it"is d fully inserted and'two diagonal bundles of,;a. cell.are removed ' , ' - d. A; blade. guide is' installed ~to-support the control rod -. when : i t D is l' fully withdrawn and: two diagonal bundles of a cell are ' removed l-3, .l '
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,.] . ., .s ; QUESTIONa 022 (1.00) The refueling bridge has interlocks that provide for control rod blocks.
' SELECT.the set of conditions which will NOT generate a rod block.
a.
Mode. switch in REFUEL and bridge is.over the core-I b.
Mode. switch in STARTUP and bridge is over the core i ! c.
Mode switch in REFUEL, one control rod at position 02, and a second l control-roc is selected ' d.
Mode switch in STARTUP and the service platform Jib crane is loaded with 600 lbs.- j ,. " 'y OUESTION 023 (1.00) l J.
I SELECT the type of detectors that the main steam line radiation monitoring.
! ocystem uses.- l a.
Fission. chambers b.
lon chambers - l c.. Scintillation detectors l l i d.
Geiger-Mueller detectors l
! l-(L O'UESTION: 024 (1.00)
t 1' l SELECT the process radiation monitoring system which does-NOT provide for en. automatic plant response if its setpoint is exceeded. Assume an i L automatic plant response to be something other then alarms or annunciators.
j ! a.
Post treatment radiation monitor (AOG) ' b.
Radwaste liquid effluent radiation monitor c.
Refueling floor ventilation exhaust radiation monitor d.
Reactor building ventilation exhaust radiation monitor
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.-- QUESTION 025 _(1.00) "Concerning the standby gas treatment system (SGT). SELECT the order of component actuation signals.which occurs when a SGT initiation signal is received.
a.
Both fans starti fan-outlet dampers open; all other campers opent STANDBY fan' stops b.4 All dampers opent'only the AUTO fan starts: STANDBY fan starts if the. AUTO ~ fan trips c.
Both fans start; all dampers except fan discharge opent fan outlet dampers opent STANDBY fan stops d.
All campers except fan discharge opent both fans start; fan outlet dampers opent STANDBY fan stops
QUESTION: 026 (1.00) The plant is at 100% power when. control power is lost to the EPR of the main turbine MHC system. SELECT the expected plant response if no operator action is taken, a. No effect. Control transfers to the MPR and all plant parameters will remain the same as before the transient b.
Control-valves will close slightly-and bypass valves will open-slightly. Control transfers to the-MPR and all parameters will remain the same as before the transient c. Control valves will open slightly.-the reactor will depressurize and an MSIV isolation will occur d.zControl valves will close slightly. Control transfers to the MPR which will maintain reactor pressure slightly higher than before the transient !
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"! ' , > ' QUESTION /027' (1kOO) - ' %.
- .T'e(main'steamIlines contain four SRVs which are designed..to-. protect.the
" .h - rOactor--pressure vessel-and'associatedipiping from over-pressurization.
SELECTg the total steam flow capacity (% of rated) through1the-SRVs if.all gn ' SRVs were fully open with the plant.at 100% power.-
' a.s40% -> r b.
60%- .. . , lc.cBO%. t o <, . . d.'100%:
i I ' b ' QUESTION:t028-(1.00)' l . l e Thc1 reactor scrammed from 100%' power'due to a loss of offsite power. The ~ .; M
- following conditions exists.
' -allEDGsstarted-butA-5a'nd,A-6havebus' lockouts (d
- N 7 s-reactor pressure is cycling with relief valve actuations
' -reactor ~ power is 0% .' -reactor water, level is -52 inches and. decreasing at 1 inch / minute --RCIC is-injecting..at rated flow ~ . -HPCI1 initiated =thenstripped . i-drywell pressure'is 2.0 psig and slowly increasing . - -. . SELECTLthe' response of the automatic-depressurization system'(ADS) if ' no 1 operator action is taken.
- a. ADS will1 automatically initiate in 120 seconds-
' b'._ ADS will automatically initiate'in-5 minutes-t c. ADS will' automatically initiate.in 11 minutes . . ..d.
ADS will NOT automatically initiate-i
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. QUESTION: O29 (1.00) ,. SELECT the~ response of an ADS valve if the control switch on alternate Chutdown panel 156 or 157-is.left in the LOCAL position.
a.
The valve Will only. operate as a safety valve b.'The valve will only automatically operate as an ADS valve 't c.
The valve can only be manually operated from the control room as a , relief valve (; 'd.
There will be no effect on either the automatic ADS function or the manual valve operation from the control room < OUESTION: 030 (1.00) s A plant startup is in progress with the mode switch in the STARTUP ' position. SELECT the statement which does NOT correctly describe how the SRM system would respond.
a, A. rod block will occur if any SRM is not fully inserted and it reads 80 cps with all IRMs at range 2 b.
If the shorting links were removed, a 1/2 scram would occur if-only the 'A' SRM reads 6x10ES cos c. A rod block would occur if only the 'A' SRM experienced a low detector voltage with all IRMs at range 4 d.
A rod' block would occur if only,the 'A' SRM failed downscale.with all IRMs at range 2 . f
' ib f' ' p _P g3 18 '- r p( aq [ A., p OUESTION: 031 L ( l '. OO ) - [M' hThe-plant 11s at 95'/. power, steady. state, with all~ normal' plant parameters . K Lfor that power level when an operator 'slightly. closes the control.- rod: drive hydraulic system pressure control. valve. SELECT the statement which 'Cescribes'how this, action would' affect =the_controlLrod drive, hydraulic ' . mystem_ operation.-
a.. Increases control rod withdrawal speeds only ~ b.
Increases control rod insert speeds only
-~c.' Increases control rod insert and withdrawal speeds i ~ d.
Increases control' rod insert and withd'rawal-speeds and decreases control rod scram times , A OUESTION: 032 (i.00)
- Procedure-No. 2.1.4 (Approach to Criticality) directs the operator-to-give.
01 NOTCH OUT or. CONTINUOUS WITHDRAW signal to all control rods-once they are withdrawn to~ position 48. SELECT the primary reason,for performing these , 'octions, l ! a.
To ensure the control rod is coupled to its mechanism lb'.0To vent the control rod mechanism so its drive speed-will be= satisfactory c.
To obtain stall flow data i d..To ensure the' cooling water flowpath for the control, mechanism is clear- < E- 'f . ! r l .. , > .
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _... ___ _3; ':.y 2* Poga 19' , - ..- . ' O'UEST I ON : 033 _ ( 1. 00 ) The scram pilot valve air. header (SPVAH) system contains two backup scram [ -valves. SELECT the statement which~ describes the operation of the backup' - scram-valves.: 'a. Both RPS trip systems must trip for either valve to open b. The' valves are normally energized and-will open on a loss of power c. The valves are aligned such that both must open to vent the SPVAH.
d. Each valve contains two 120V AC solenoids-OUESTION: 034 (1.00) L' ' . SELECT the combination of LPRM inputs to an APRM (if any), that is required - for that-APRM to automatically generate a rod block, a.
At.least one level less than 2 only b.' Total inputs less than 11 only . c. Total inputs less than 11 OR at least one level less than 2 d. No combination. LRPM inputs are only an administrative' limit and do not generate automatic rod blocks .- - __ i
.. ..... .... ... _..... . - - - - -.- - - - - -
pi, p : ;)l, Pcg7 20.
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.c a
-
, 2- , b 7; - QUESTION: 035 (1.00) ",The full core display contains a red light' labeled DRIFT for each control rod. SELECT the condition which will cause'the DRIFT light to illuminate.
a. A' control rod closing cx11 y an odd reed switch with no motion
command present b. A control rod closing only an even reed switch with no motion command present c. A control rod closing either an odd or an even reed switch with no ' - motionEcommand present d. A control rod closing both an odd and even reed switch simultaneously with a motion command present-r OUESTION 036 (1.00) SELECT the: thermal limit which the rod block monitor (RBM) is designed to protect against exceeding.
a.
MFLPD b. MCPR c.
APLHGR d.
LHGR , QUESTION: 037 (1.00) ' , ' T h2 plant is.at 100% power when a complete loss of 24v DC occurs (both pnnels D-25 and D-26 deenergize). SELECT the statement which describes the cutomatic plant response.
a.
Rod block only b.
1/2 scram only c.
Scram xy d. No automatic response other than alarms
. ..
- '
' Pcg1 21- .i i.jj , .: , ~.
lll :' ' - DUESTIONs.030 (1.00) .A, reactor-startup._.is in progress with power,at 18%'. SELECT the' condition which will cause 'the:= rod worth minimizer (RWM)' to generate a rod block, a '. All rods in the currently latched group and all lower groups are withdrawn to their group withdrawal limits j 4, -t > b.
One withdraw error exists c.
Two' insert errors exist , ' d. A rod is at its alternate withdrawal limit because of a failed reed
switch ' ,, e' > QUESTION:-039.
(1.00).
ALGroup III PCIS isolation will' isolate-RHR (LPCI.) injection' valves J
- MO-29A(B) 11 low reactor water level or high drywell pressure exists when j
- RHR is in'the shutdown cooling (SDC) mode of operation. SELECT the
- '
EOxplanation of why a Group Ill PCIS: isolation does not occur, preventing LPCI from injecting,-if the plant was at 100% power and a major LOCA: occurred.
, -. l
- Valve ~ Names
- MO-7A(B C.D) - suppression chamber suction
' MO-43A(B,C D) - SDC suction - MO-47'- SDC' suction outboard isolation MO-50 - SDC suction inboard-isolation a.
SDC mode is activated: anytime MO-43A(B,C.D) is not. full closed and-
reactor; pressure is less than 100 psig-b..SDC mode is activated any time MD-47 and MO-50 are noti f ull closed
and reactor pressure is less than 100 psig .c.-SDC mode Lis activated anytime reactor pressure is less than ' l l 100 psig and the mode switch is in REFUEL or SHUTDOWN r SDC mode is activated anytime MO-7A(B,C,(D)j MW) _a d.
and reactor pressure is .less.than 100 psig . ' l.
! , ' !^ L l
, < -.1 .
\\' Pcg3 22- ... [!*,v ' ' , .l.
-QUEETION: 040: (1.00) . The) plant is at'100% power when a ground fault develops on 4160v AC bus A-6.
Bus-fault protection caused a complete deenergiration of the bus.' Immediately thereafter, the 'B' recirculation loop ruptures. (assume.a-
double-ended pipe shear). SELECT the statement which describes the automatic r3sponse.of,the LPCI system.
a..The 'A' and 'C' LPCI pumps will start and inject to the 'A' recirculation loop b.
The.'B' and<'D' LPCI pumps will start and inject to the 'B'
. recirculation. loop.
r r c. The 'A' and 'C' LPCI pumps will start'but.will NOT inject into
- sither of the recirculation loops d.
The 'D' and 'D' LPCI pumps will start but will NOT-inject'into , l; either of the recirculation ' loops l-L ., QUESTION: 041.
(1.00) The plant was at 100% power when the 'A' main steam.line ruptured inside ' . the drywell. All plant parameters were as would be expected for this event end.all automatic actions that were expected occurred. One minute-into the -; ovent, with reactor level rapidly recovering due to low pressure ECCS cystems" injecting, the operator was immediately directed to secure:LPCI ' Jinjection as necessary to maintain the normal water level band. The c-operator took-the control switch for RHR outboard injection valve MO-2SB to E 'CLOSEEbut'the valve did not reposition. SELECT the reason for the valve not
rspositioning.. > a.
The initiation signal must be RESET before this valve can be closed
b.
This valve ~is interlocked open for 2 minutes c. This valve is-interlocked.open for 5 minutes d. There are no interlocks associated with this valve.-The valve should have closed, therefore, a problem with the valve or its L control circuit exists , l l: l l l-I i l - - - - . - - ,
, c. ;;,,ss.
'-. P0g3 23 l ... , - ' -: -
l ~n
- .,
, . -QOESTIONar042 '(i.00).'- ~ ~ ' , (ThelHPCIcsystem uses'a ramp generator to. prevent. turbine overspeed on:an cutomaticLinitiation. SELECT.the component which.directly provides the-l - , J input to1 actuate the ramp; generator;on'an-automatic HPCI initiation.
1'! b a..MO-2301-3 turbine steam supply isolation. valve , ! ! b.
HO-2301'-i turbine'stop valve
-
- j c. - HO-2301-2. tur bine con trol-val ve.
- s d'. : Auxiliary - oil pump ' i i < , "' '0UESTIONecO43 '( 1.'00 ) : . An operator reports-that the 'A' standby [ liquid' control (SLC) Squib' valve . continuity meter indicates 3.5 ma'of' current. SELECT the action that the NOS should take.. > t ~ a.
Do nothing-since 3.5 ma is a normal indication , b.
Declare 1the squib valve-inoperable since 3.5 ma indicates current
is tooflow:and the squib valve may not fire . '; - - -t c.
Declare 1the squib' valve inoperable-since 3.5 marindicates-current = ' is tooihigh.and the'ignitor element'may have decomposed ,. ' d.' Declare the. squib valve inoperableLsince 3.5 ma indicates the squib valve has' fired' t l ' .- i h- ' s.
'. , . -- .
^ ~ ' . .3 POgi 24 ' y,gg *35 ; , +.; . ., s --QUESTION E044' ( 1. 00 ).
- TheLplantiwas at'100*/.: power when a smallusteam-leakLdeveloped whichEcaused
, Jdrywell, pressure to, increase to 3.01psig.~ The reactor scrammed ~andLduring- ~ ~ ' y Eth7 transient reactor water? level decreased to +5 inches before being. E rCOtoredito'the normal' band.:SELECTxthe PCIS-group isolations thetishould~ _hovax occurred.
+ _ a. _ Groups _'l,2 ' a.
b.
Groups 2,6: 'I
- c.:: Groups - 4, S , d.l Groups?3 ~, 4 ~ QUESTION: 0 4 5 ' '( l'. 0 0 ). !An immediate operator-action for Procedure No. 2.1.6-(Reactor Scram)Ris to p!'Oce theimode-switch'in the REFUEL or SHUTDOWN positions. SELECT the-primary reason for this immediate operator action.
a.-An additional ~scramisignal is generated j-b.'A Group I isolation signal is bypassed . -c, The bypass for MSIV' isolation on high reactor water level-is removed " d. The bypass for a' Group II isolation-signal.is removed
I
a
6 ? i _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _, t
!* ~- J > -e ! , "[ Pcg3 25, <.c . , m I.g.. A QUESTIONE046 ( l'. OO ) - 0: CHncarning.the-AC electrical distribution system, SELECT the choice which-h001the components listed with their correct bus power supplies.
~ a. A-13 reactor feedpump C b.-A-1: condensate pump B .A-2: condensate pump A A-2: reactor feedpump A A-3, sea water pump A' A-3: sea water pump'B ". A-43 recirc MG' set B A-4: recirc MG set A c..A-13-condensate pump A d.
A-1 reactor feedpump A- ' A-23 condensate pump C A-2: reactor feedpump B A-3: recirc MG set A A-3: 1recirc MG-set B A-4: sea waterEpump B' A-4 sea water pump A ? OUESTION: 047 -(1.00) Aftar a special test on the 'A' emergency diesel generator (EDG), the .l ffollowing switches were inadvertently ~left in the positions indicated: LLocal Auto / Man Voltage Regulator - MAN , I Droop /Isochronous Switch - DROOP . Norm / Test' Switch - TEST SELECT-the statement-which describes the automatic response of the 'A' EDG i ~ifcit=were to start on a valid LOCA signal.
] > a. :. The operational. mode would switch' to ISOCHRONOUS regardless of the
- switch: positions above'
, =b.
The operational mode would remain in DROOP because the Local l Auto / Man Voltage Regulator is not in the normal' standby position j ! ' E c. Thesoperational mode would remain in DROOP'because.the ! L . Droop /Isochronous switch is not in the normal standby position { ! E d.
The: operational mode would remain in DROOP because the Norm / Test ! ( . Switch is'not in the normal standby position ! l b - ! ! i.
l l l t; I ' . - - - - - - - - - - - -- --- .- ...
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Pcg3 2b; [,,; "', e
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3.
kQUESTI'ON: 048(1.00) . 'Th2fplant'is-at 100% power. Concerning oower sources for a 4160v'AC p TCmergency-bus -SELECTothe item which lists the preferred order of power ' h '.OCurces'from-the-normal supply--to the last available supply, a '. Startup' transformer; unit auxiliary transformer; emergency diesel generator; shutdown transformer lb. Unit auxiliary transformer; startup transformer; emergency-diesel generator; shutdown transformer c.
Startup transfermer; unit auxiliary transformer; shutdown transformers emergency diesel generator d. Unit auxiliary' transformer; startup transformer;. shutdown transformer; emergency diesel generator l-QUESTION: 049 (1.00) U TEloctrical protection assemblies (EPAs) are used to provide protection for ! Oyptems and components powered by RPS. SELECT the condition which will NOT f is 'cutomatically, trip an EPA.
a..Overvoltage-L b.
Undervoltage-
. .c.
Underfrequency g i L
- d. Overcurrent
.s- '
, . -
m- - - - .
~ - ,:
' Pcga'27 Y; ' sal
/17 ,
JOUESTIONibo50 ( 1.~ O0 ) ; {
, r All reactor vessel water level instruments, except for one,. have the , mCjority,ofEtheir reference legs located-external to the drywell to - minimize-the effects'of-_drywell-temperature changes on_ level indication.
s SELECT ~the. Instrument that does NOT have the majorityEof its reference-leg- , ,loccted. external to the drywell, ' p a'...Feedwater! level (O toc +60 inches) , b.
Narrow range (-50Eto +50 inches) c. Fuel 4 zone level.
d.-Shutdown level 't ' ,
. QUESTION JO51~ ' ( 1. 00 ) ' ..Tha plant 11 seat 100% power with all plant parameters normal for:-that power,. i w'h3n n t he ' ' C ' steam flow detector for the feedwater level control system; ' fails low:(its output-indicates O lbm/hr steam flow). SELECT the1 statement which describes the expected' automatic-plant response.
, i .a.gReactor water. level will decrease and stabilize at a lower than normal value but:the' reactor will not scram . b.
Reactor water level will decrease and' stabilize at a l'ower than 7' ' normal value but.the reactor will scram:due to lowfwater level, .Y . increase and stabilize at a-higher than' c.. Reactor: water level will . .. normaltvalue but the main turbine will not trip d.' Reactor water level will increase'and stabilize.at a higher than' normal value but tha reactor will scram due to a main turbine trip ., _ .j ' n - l mE l-
h[' ' ~ o " t , Pcg3 28;; [ :;jf'.( . ' e/Fr:
(; , s A I QUESTION: 052'~(1'00)- . , N ' . . . p ,Thei'A' stator water cooling heat' exchanger has been-isolated to repair [ tube leaks.LSELECT.the maximum generator capability with this heat , exchanger isolated. (Assume generator hydrogen pressure'is at its-normal- " 'value fore 1u11 power operation) , a a.~SO%~ reactor power b.f175 MWe (4410 stator amps)
, . ! c.,80%. load-(624,000 KVA). i , ' . d.
100%-load-(each heat exchanger-is 100% capacity) c , N E - b - , P ' OUESTION: 053 (1'.00) Procedure No.-2.1.1-(Startup from' Shutdown) directs the operator to ' transfer-FWLC~to "3-element control" at 30% power. SELECT the bases for not transferring.to "3-element control" before130% power.
a. An interlock-prevents placing."3-element, control" when 1eedflow is ~ . .less than 20% of rated , b. At'ilow power' levels'the FWLC level-input is inaccurate - c. At low power 11evels the FWLC steam flow input is inaccurate d. At low power levels the FWLC feedflow input-is inaccurate , h I
!- 1, I-l m,1 - s-- -. .
-_ , , ' PcgG 29' J 'YS4
, l s r
-GUESTION: 054 :(1.00) i SELECT.'the feedwater heater (s), which if removed from service, will-cause-im0in generator-output to increase-based only on the direct effect of the
- h3ater(s)'on the main turbine.-
l .a.
1st. point only b.'2nd,3rd,4th, or'5th point l ? c.
5th. point only e
d.
Any heater. All heaters will increase generator output by the same amount i .
QUESTIONS.055 (1.00) ! 'The'olant is at 100% power when a complete loss of essential instrument air occurs. SELECT the automatic _ valve response for the following condensate- ' cnd feedwater system air-operated valves. (Assume valve response isEbased only on its loss of air supply:and not on the' response of the plant)
a., Condensate pump min flow valves - FAIL CLOSED-Reactor feedpump min. flow valves - FAIL CLOSED
Feed regulating valves - FAIL CLOSED ' b.
Condensate pump min' flow valves - FAIL OPEN-Reactor feedpump: min flow valves - FAIL OPEN ! Feed regulating valves - FAIL AS IS .
c. Condensate. pump min flow valves - FAIL CLOSED ' Reactor'feedpump min flow valves - FAIL CLOSED . g Feed regulating' valves - FAIL AS-IS-d.
Condensate pump min flow valves - FAIL CLOSED Reactor feedpump min' flow valves.- FAIL OPEN Feed regulating valves - FAIL AS IS , E5
i ? A
a- -
- - - - - - - -. - - - - - . - - - - - - - - --- -
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Pcg3 30~ ' '. - ' ... ; . ! . .4 QUESTION: OS6 (1'.00) ' '
- The plant is at.100% power when an operator performing his rounds reports
h.
.that-core spray loopA' line break detection local indication at rack'2207 ~ic reading +8 psid. SELECT the statement which describes what this. reading . indicates. (Assume all other plant parameters are normal for this power lOvel) a.
This is a normal reading for operation at 100% power (normal DP across'the separator and dryer) b.
A core spray line break has occurred inside the shroud c.
A-core spray line breat has-occurred inside the vessel but outside the shroud d.-A core-spray line break has occurred either inside the drywell or inside the vessel but outside.the shroud - QUESTION: 057 (1.00) The plant was at 100% power when a loss of feedwater event occurred causing roactor. water level to decrease to -6S inches. HPCI failed to-initiate but RCIC'did automatically initiate and reactor water level is now increasing.
' SELECT the statement which describes the automatic RCIC system response if- -i no operator action is taken.
' a.-At +48 inches a RCIC isolation signal is generated. The isolation , ! will automatically clear at -49 inches and decreasing.
b.
At +48 inches a RCIC turbine trip signal ~1s generated. The turbine trip will automatically clear at +48 inches'and decreasing.
c. At +48 inches a RCIC. turbine trip signal is generated. The turbine .i trip will automatically clear at -49 inches and decreasing, d.
At.+48 inches the RCIC steam supply valve shuts. The valve will automatically reopen at -49 inches and decreasing.
.. --
Pcgg 3,; q y ' , T .- QUESTION: 058 (1.00) f 1The plant is-at 100'/.. power when an' operator, while explaining RCIC system- ^b-ii - operation to's trainee. Inadvertently-initiates RCIC. SELECT the statement 3(. which describes reactor power and reactivity response as a result-of RCIC 'N-injecting!to-the reactor at power.
4.
Power decreases slightly since RCIC injecting will cause a drop in pressure.
b.
Power remains'the same since only void content or control rod position'affects reactor power when operating at high power levels c. Power remains the same since'the negative reactivity from the ' pressure drop offsets the positive reactivity of the core-inlet coolant, temperature drop d.
Power increases slightly since the core inlet coolant temperature drops E QUESTION: 059 (1.00).
The offgas-system contains a 30 minute holdup pipe that allows for decay o*; shorter lived radioactive isptopes. SELECT the radioactive isotope which j would.NOT be effectively. removed by the 30 minute holdup pipe.
! a.
N-13 b..N-16 c. 0-18 d.
Kr-81 , b
- i
! L
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~ POg] 32
< .. ' ' ( , QUEST!ON: 060 (1.00) A problem with hydrogen injection chemistey is that steam line radiation + lovels increase. SELECT the statement whicn describes why steam line j rCdiation levels increase when hydrogen is injected to the reactor.
Increvted hydrogen produces more non-volatile nitrites atu1 nitrates a.
. Increased hydrogen produces more volatile nitrites and nitrates b.
c.
Increased hydrogen produces more non-volatile ammonia di increased hydrogen produces more volatile ammonia QUESTION: 061 (1.00) Cartain steps in the emergency operating procedures (EOPs) direct the Operator to take actions which may deviate from the requirements of tOchnical specifications. SELECT the statement which describes the actions cn operator shall take and the reason for allowing these actions,11 while i oxecuting an EOP, an EOP action statement violates a technical Opecification requirement.
a.
Technical specifications must not be intentionally violated since i it is a higher priority document as explained in Administrative i Procedure No. 1.3.4 (Procedures). Follow technical specification ' guidance.
- b.
Technical specifications must not be intentionallv violated since ' it-as part of the station's operating license. Follow technical specification guidance.
s c. The EDP flowcharts contain caution statements which state that EOPs take precedence over all other station procedures. Follow EOP , guidance.
a. Administrative Procedure No. 1.3.6 (Technical Specification Adherence and Clarification) allows for deviations from technical , specifications durang emergencies. Follow EOP guidance.
t I; - ,. -_
c - ,
'e-lo Is GUESTION: 062 (1.00) m , SELECT the statement which is NOT correct regarding EOP entry conc, .s " 'N,\\ . a. EOPs are designed such that anytime an EOP is entered, tnt i ' - N, emergency plan will also be entered.
If executing an EOP, a second entry condition for that eof! N b. that entire EOP must be re-entered from the beginning.
s ' - '
c. If an EDP is entered on one entry condition, that entry ci; . , clears, then that same entry condition is met again, the af' l1 must be re-entered from tne beginning.
I d. If an entry condition is met that is an entry condition 1t' different EOPs, both entire EOPs must be executed.
i OUESTION: 063 (1.00) SELECT the choice which is NOT considered a viable means of assur adequate core cooling in accordance with the EDPs.
i a.
Core submergence
b.
Spray cooling , c. Steam cooling and injection of makeup to the reactor vest ; d. Steam cooling
QUESTION: 064 (1.00) which contain the Minimum Alternata RPV Flooditj SELECT the EOP(s) Curve (MARFP).
, a. EOP-01 (RPV Control) and EDP-03 (Primary Containment Con b. EOP-02 (Failure to Scram) and EOP-06 (RPV Flooding)
c. EOP-01 (RPV' Control) and EOP-06 (RPV Flooding) d. EOP-06 (RPV Flooding) only
g ' , , .; h' Pcg] 34 . , < , %o l, - * U OUESTION: 065 (1.00) , " i SELECT the statement which does NOT accurately describe EOP caution Otatements.
I a. Located inside flowchart elements I b.
May contain instructional steps ! \\ \\ c. Referenced in the EOP by circled numbers C d. Identify potential hazards to personnel and eouipment j s . I DUESTION: 066 (1.00) i The RPV pressure leg of EOP-01 (RPV Control) directs the operator to i otabilize RPV oressure below 1005 osig with the main turbine bypass valves.
i SELECT the bases for choosing 1085 psig as the pressure limit.
, -a.
To avoid an ATWS-RPT , b. To ensure at least i bypass valve is partially open at all times c. To reset the scram signal d. To avoid opening the safety valves and discharging directly to the i
drywell ' ! OUESTION: 067 (1.00) The RPV water level leg of EOP-01 (RPV Control) directs the operator to
inhibit ADS if level cannot be maintained above ~49 inches. SELECT the Ocses-for inhibiting ADS in this situation.
a. ADS actuation could lead to a loss of adequate core cooling ! b. ADS actuation could cause reactor power surges c. ADS is never permitted to operate automatically in any of the EOPs > d. ADS actuation will violate the technical specification cooldown j,. rate and-intentional violation of technical specifications is not _ permitted . ) > $ ,
__ PO;g 35.
.c 'E , '
, ! s i CUESTION: 068 (1.00) t EOP-02 (Failure to Scram) recuires RPV water level to be intentionally ^jN ' lowered if certain conditions are met. If EOP-02 was entered and RPV water . i level was being-lowered. SELECT the plant condition that would allow the ' ' lowering of RPV water level to be stopped.
i I a. Reactor power 5% and drywell pressure 2.5 psig and steady ' , b.
Reactor water level -126 inches and reactor power 10% ! c. 1 SRV intermittently opening and drywell' pressure 2.0 psio anc steady d. Reactor. cower 5% and ~ reactor water-level at +9 intnes t i f .OUESTION: 069 (1.00)- SELECT the.drywellr temperature setpoint which is the entry condition for
EDP-03 (Primary Containment Control).
-i a.
148 degF b.
152 degF t c.
158 degF
d.-162 degF , ! OUESTION: 070 (1~.00) , SELECT the " Normal Band" for torus water level as stated in EDP-03 (Primary Containment-Control).
l a. Between 125 and 135 inches
b.:Between 127 and 130 inches ! c. Between 130 and 135 inches d. Between 130 and 137 inches I s ) S P . .- _ . . _ _
, .o .'.. . * i PC9] 33
e i
i i CUESTION: 072 (1.00)
EOA-03 (Primary Containment Control) directs the operator to secure drywell i sprays, if they were initiated, once drywell pressure is reduced to 2.5 .i psig. SELECT the bases for securing crywell sprays at that pressure.
a. The combination of drywell temperature and drywell pressure is no longer on the " safe" side of the drywell spray initiation limit ! curve.
] b. The scram can be reset so pressure reduction in the drywell is no { longer needed.
I c. Drywell sprays are secured so that suppression pool sprays can be i initiated since suppression pool sprays are preferred.
d. Drywell sprays are secured to prevent imploding the drywell.
, OUEST}ON: 07C (1.00)
EOP-03 (Primary Containment Control) directs the operator to enter EOP-07 (Alternate RPV Depressurization) 11 drywell temperature cannot be , maintained below 281 degF. SELECT the bases for this temperature limit.
' a. This is the design temperature of the containment and it may fail ' at this temperature and above.
b. Parameters cannot be maintained on the " safe" side of the drywell ' spray initiation limit curve at this temperature and above.
c. Parameters cannot be maintained on the " safe" side of the RPV , Saturation limit curve at this temperature and above.
d. The heat capacity tempersture limit of the suppression pool cannot be maintained if the drywell is at this temperature and above.
,
, P;ga 37 %. !
,
- =
QUESTION: 073 (1.00)
EOP-03 (Primary Containment Control) directs the operator to initiate i drywell sprays if containment hydrogen and oxygen content cannot be ,N maintained below certain limits. SELECT the bases for initiating drywell q sprays under this situation.
' t-a. Drywell sprays will cause condensation of the hydrogen and oxygen to reduce the pressure in the drywell.
b. Drywell sprays will extinguish any fires caused by hydrogen f igniting.
, c. Drywell sprays will filter any iodine in containment so there will' [ be no iodine release if the drywell is vented.
, d.
Drywell sprays will mix the drywell atmosphere to reduce localized buildup of cases.
+ \\ OUESTION: 074 (1.00) ' EOP-04 (Secondary Containment Control) directs the operator to enter EDP-07 ! (Alternate RPV Depressurization) when area temperatures exceed the maximum cafe operating values jn two or more areas and a primary system is . discharging into the secondary containment. SELECT the bases for entering EOP-07 for this situation.
a. The RPV is depressurized to prevent exceeding radiation limits for the secondary containment.
l b.
The RPV is depressurized to prevent failure of secondary containment integrity.
c. The RPV is depressurized to prevent failure of equipment which is
located in primary and secondary containments.
d. The RPV is depressurized before loss of SRV/ ADS operability occurs.
, ,
' hI y ' 'c .Pcg] 38
- -
.. , s ! l.
QUEST 3ON: 075 (1.00) SELECT the number of entry conoitions for EDP-04 (Secondary Containment Control).
a.
S' , ' .b.'6 , ! , C.
> d.
!, , .d ' "' . OUESTION: 076 (1.00) . EOP-06 (RPV Flooding) oces not list core spray as a preferred injection ' I cource to flood the RPV during an ATWS. SELECT the reason why core spray is
- not a preferred injection source for this situation.
a. Core sprav does not have the required flowrate to assure adequate core cooling during the floodup.
s b.
Core. spray cannot maintain the required 52 p51d with three SRVs j-opened that is needed to assure adequate core cooling.
_ p c. The SBL boron concentration requirement of technical specifications l . assumes that no systems are injecting inside the shroud and , diluting the mixture.
d.
Core spray could cause power excursions since it injects inside the ' - shroud.
. d .- r t
= ' _' I _ _,
p -
. Pcg] 37
.l,,.- ,s o ! , a- , CUESTION: 077 (1.00) ' The plant is at 100% power wnen during a surveillance test, the manual scram pushbutton for RPS 'A' is found to be inoperable. (Will not cause a ,', s 1/2 scram when depressed) SELECT the action required by technical specifications. Assume all other systems and components to be operable.
I a. No action is required since all automatic RPS scrams are operable.
, p b.
Trio RPS 'A' and operation at 100% power is permitted.
c.
Trip RPS 'A' and insert all operable control rods within 4 hours.
I ' d.
Place the mode switch to SHUTDOWN to scram the reactor.
QUESTION: 078 (1.00) The plant is at 100% power when the operator just receives the most recent P-1 computer printout anc notices that MAPRAT is greater than 1.0. SELECT the action required by technical specifications. Asbume all other systems cnd components to be operable.
, i a.
Initiate action within 15 minutes to restore MAPRAT to less than 1.0.
b.
Commence a reactor shutdown such that the reactor will be in cold , shutdown within 36 hours.
c. No action required since technical specifications does not have a MAPRAT limiting condition for operation.
d.
No action required since technical specifications requires MAPRAT to be greater than 1.0 at all times.
A
l [ . .
- , , ' ,e P g] CO '. o
_, I -. _ , EUESTION: 079 (1.00)
! The plant is at 100*/. power when during a HPCI valve stroke surveillance l
test. MO-2301-5 (HPCI steam to turbine outboard isolation valve), stroked ! closed in 30 seconds. SELECT the actions required by technical l ! specifications. Assume all other systems and components to be operable.
a. No action required since no data is provided to indacate that HPCI failed any surveillance requirements.
' . b. Initiate an orderly shutdown to be in Cold shutdown within 24 hours.
l c. Initiate an orderly shutdown to be less than 104 psig RPV pressure within 24 hours.
J Jews.w) C l os b,,,,MO-2301 -4. ( HPC I d.
steam to turbine inboard isolation valve) and declare HPC) inoperable.
QUESTION: 000 (1.00) SELECT the condition which is NOT considered an alteration of the reactor core as defined in technical specifications. Assume the mode switch is in REFUEL. the vessel head is removed, and the cavity is flooded. Also, consider DNLY the event stated below.
a. Removal of a CRD mechanism i b. Removal of an LPRM string c.
Installation of a fuel support piece d.
Installation of a blade guide
!
i 8.
Pcg] 41 ' ?e q - , , b' ^ . i, CUESTION: 081 (1.00) ! i SELECT.the set of conditions which meets the criteria for " hot standby . condition" as defined in technical specifications.
7,. "25J a. Coolant temperature greater than 212 degF Pressure below BOO psig 3', MSIVs closed Mode switch in ST ART 4JP or SHUTDOWN b. Coolant temperature greater than 212 degF Pressure below 600 psig , " MSIVs closed Mode switch in STARTUP c. Coolant temperature greater than 212 degF Pressure below 080 psig MSIVs closed Mode switch in STARTUP ' d. Coolant temperature greater than 212 degF - ' Pressure below 600 psig MSIVs closed Mode switch in SHUTDOWN , , OUESTION: 082 (1.00) i Technical specifications requires recirculation pump speeds to be within 10% of each other when the reactor is greater than 80% power. SELECT the
bases for this speed mismatch limit.
a. With a. mismatch of greater than 10%. excessive APRM noise fluctuations may occur b. With a mismatch of greater than 10%, excessive flux tilts may occur f across the core = c. With a mismatch of greater than 10%. recirculation pump vibration is too high for continued operation at greater than 80% power d. With a mismatch of greater than 10%, the LPCI loop select. logic may not function properly i _
. - - - .- Pc;o 42
' ., ... . F QUESTION 'OB3 (2.00) I The plant is at 100'4 power when a complete loss of vital AC (Y-2) occurs.
>: An immediate operator action of Abnormal Procedure No. 5.3.6 (Loss of vital AC Y-2) is to monitor reactor water level and pressure. SELECT the specific ' instruments that this procedure directs +,he operator to monitor, a. Pressure -HPCI instrument on panel 903 Level: Narrow range on. panel 905 ' b. Pressure: HPCI instrument on panel 903 Level Shutdown level on panel 904 L c. Pressures Barton gages on racks 2205 or 2206 Level: Narrow range on panel 905 d. Pressures Barton gaQes on racks 2205 or 2206 Level: Shutdown level on canel 904 OUESTION: 084 (1.00) You are the refueling SRO with core alterations in progress when the' refueling area radiation monitor alarms.-You carry out the immediate-operator actions of Procedure No. 5.4.3 (Refueling floor high radiation) and as a subsequent steo, you verify that the reactor building vent did not exceed 710 cps. SELECT the bases for verifying that the reactor building vent did not exceed 710 cos for this situation.
This is an EOP entry condition a.
b. This would. require an evacuation of the refuel floor and the reactor builcing c. This would indicate that standby gas should have automatically started d. This would indicate that the high efficiency filtration systems for the control room, cable spreading room, and the computer room must , be placed in service i ! i l ! I
I Pcg] 43 L ,1 % c-
, . r l DUESTION: 085 (1.00) l-l A station blackout.has occurred and the SBO diesel failed to start. You e were controlling reactor level and pressure.with HPCI and RCIC when both , , these systems tripped and became unavailable. Procedure No. 5.3.31 (Station '
- N blackout) directs you to Procedure No. 5.3.26 (RPV injection during
' cmergencies) for alternate injection methods. SELECT the alternate injection system listed below that.could be used for'this situation.
[ ! a.
Fire water cross-tied to RHR , L b.
SSW cross-tied to RHR ' c.
Demineralfred water transfer cross-tied to SBLC f d.
Condensate transfer cross-tiea to ECCS fill lines , GUESTION: OB6 (1.00) i , A failure to scram event has occurred. All actions have been taken up to the point of venting the over-piston area of the control rod drives in , cccordance with Procedure No. 5.3.23 (Alternate rod insertion). SELECT the- ' CRD piping in which the over-piston vent valves are installed.
a.
Exhaust b.
Withdraw , c.
Insert d.
Drive . F V
f,i[ " PCg] 44 , * '
,- .- . f 54-
- .
i D OUESTION: 087 '(i.00) ~ l J The plant was at lOOX. power when a fire occurred in the cable spreading ! ' room (CSR) and the watch engineer directed an evacuation of the main j control room (MCR) in accordance with Procedure No. 2.4.143 (Shutdown froa.
, p outside the control room). SELECT the action listed below which is NOT en ! 'immediate operator action of Procedure No. 2.4.143. Assume at the time of i Cvacuation the fire had not affected the operation of any plant equipment.
h a. Reduce recirculation pump speeds to minimum b !. b. Place modo switch to the SHUTDOWN position ! l'
" ~ Shut the MSIVs c.
, f- ' , d.
Trap the feedpumps . l QUESTION: OBB (1.00) , 1The plant was at 100% power when train
- B'
4th point heater isolated due to cn internal tube rupture. An immediate operator action of Procedure No.
2.4.150 (Loss of feedwater heating) is to reduce reactor power. SELECT'the i , ' method of reducing reactor power that is in accordance with the immediate operator actions of Procedure No. 2.4.150.
4. Runback recirculation flow to reduce core flow to 40 Mlbm/hr t regardless of reactor power level i b. Insert control rods to below.the 80% rod line' c. Runback recirculation flow to reduce reactor power to 75% d. Utilize both recirculation flow and insertion of control rod to , reduce reactor power to 25% i ) , 'h l
1
' 1'
F1 - l
yp.,
Pcg] 45 F ' i
DUESTION: 089 (1.00) .The 'olant was at 100% power when, while performing Procedure No. 2.1.15 ' OPER-09 (Daily surveillance log), a problem was noted with one of the non-calibrated Jet ocmps. SELECT the condition which is NOT a symptom of a j jet pump' failure.
.y-c: a. Decrease in core olate dP indication N '
> b.
Decrease in indicated reactor power
c. Decrease in indicated steam flow , , d.
Decrease in recirc loop flow for the loop with the inop jet pump , f f f L ! QUESTION: 090 (1.00) The plant was at 100% power wnen the 'A' recirculation pump tripped.
' Procedure No. 2.4.17 (Recirculation pump trip) was entered. SELECT the ' statement which describes the recuirement given in Procedure No. 2.4.17 regarding when to manually scram the reactor. Assume the reactor did not i ' automatically scram and no operator action has been taken.
a. Immediately if APRM oscillations of 5% peak-to-peak occur b. Immediately.since the reactor is operating above the 80% load line .i c.
Immediately if periodic LPRM downscale alarms occur . d.
Immediately since core flow is less than 24.5 Mlbm/hr
i s t 'I.~ I E , -- -. - . -. - - - -. - - - - -.. -. -
n k'. [~ , ., P0g] 43 ? m .s i l bu _ i ' I l QUESTION: 091 (1.00) , P The plant was at 1007, power when a small crack developed on the side of the ' [ main condenser causing vacuum to decrease. SELECT the statement that is an t-appropriate immediate operator action in accordance with Procedure No.
2.4.36 (Decreasing condenser vacuum).
' a. Reduce reactor power by reducing core flow to not less than 37.5 . _ Mlbm/h- , b.
Reduce reactor power by reducing core flow to not less than 31.5 Mlbm/hr i c.
Recute reactor power by reducing core flow to not less than 20.5 - !' Mlbm/hr d.
Reduce reactor power by reducing core flow to not less than 24.5 , Mlom/hr t QUESTION: 092 ( 1. OC' i i The reactor is in cold shutdown, coolant temperature is at 125 degF, and the 'A' RHR pump is in_ shutdown cooling when a bus fault develops on 250 vDC bus D-9.
Procedure No. 5.3.30 (Less of 250 vDC power bus D-10) is entered and this procedure directs the operator to secure shutdown cooling.
SELECT the reason for securing shutdown cooling for this situation.
a.
D-9 supplies control power for_the RHR pump breakers , b. D-9 supplies operating power to MD-1001-47 (shutdown cooling
suction isolation-outboard) c. D-9 supplies operating power to MO-1001-50 (shutdown cooling suction isolation-inboard) d. D-9 supplies logic cower for PCIS Group III (RHR shutdown cooling isolations) ,
yt Pcg] 07 . . .; f-QUESTION: 093 (1.00)
- A reactor startup is in progress. The main generator is at 1600 RPM and is ready to be synchronized to the grid when the TURBINE HIGH VIBRATION L
Cnnunciator' alarms. In accordance with Procedure No. 2.4.46 (Turbine bearing malfunction). SELECT the bearing vibrat2on setpoint at'which the Operator woulo immediately trip the turbane if the vibration continued to N-. increase.
.g a.
O mils L b.
10 mils c.
10 mils ! o.
14 mils DUESTION 044 (1.00) The plant was at 100*/. power when reactor' pressure began decreasing. The operator-Quickly diagnosed the problem to be a failure of the EPR and turned the EPR "off", as. directed by Procedure No. 2.4.37 (Turbine control , cystem malfunctions). This action was successful in stabilizing the . transient and the MPR is now in control. SELECT the statement which Cescribes final reactor power level as compared to the pre-transient power level.
Reactor power will remain constant since reactor pressure will be a.
the same as it was before the transient.
b.' Reactor power will increase slightly since reactor pressure will be ', .slightly higher than it was before the transient.
c. Reactor power will decrease slightly since reactor pressure will be slightly lower than it was before the transient.
d. Reactor power will increase slightly based only on the fact that there is less feedwater heating than before the transient.
g l 1.
~i l l ! i ! .1
m- , Pcg] 48 , ,. .o e - l,,
' , ' #UEST l' N's ' 095 (1.00) O
iThe plant'was at'100% power when a pipe rupture occurs in the 'A' loop of l .9CCW. SELECT the component (s) which will NOT be directly affected by a i 1000 Of the 'A' loop of RBCCW.
' c. VAC-205 A-F drywell air cooling coils i b.
A& B RCIC pump area coolers ' c. A & C.RHR pump mechanical seal coolers ,; . d. A RHR heat exchanger . : l !
,UESTION: 096 (1.00) -The plant is at 100% power with the 'B' TBCCW pump tagged for motor .e'calocement when the 'A' TBCCW pump trips on overcurrent. If the operator- ~10110 to carry out the immediate operator actions of Procedure No. 2.4.41 lLO o of TBCCW) the reactor will eventually. automatically scram. SELECT the
cvCnt 'which will initially cause the reactor to scram if no operator action
10~tcken.
' o.' Turbine generator trip due to loss of stator cooling , b. Turbine generator trip due to loss of isophase bus cooling , c. Condensate pump trip due to loss of cooling to motor coolers . causing low reactor water level ' d. Reactor feedpump trio due to loss of cooling to lube oil and seal water coolers causing low reactor water level ., h t-f . [ _ _ . -.
P:QJ GO
- 'a
c , e I '
. ' i
- UESTION
- 097 (1.00)
l'The plant is at 100%~ power when a break develops on the instrument air heCC2r. Instrument air header pressure continues to decrease. SELECT the l-CDCu;nce of automatic actions that would occur, as air header pressure [czntinues to decrease, if no operator action is taken.
!
- c.-AO-4365 shuts to isolate non-essential 1 instrument airl AD-4350
shuts to isolate the service air header Feedwater control valves ! lock-upt Scram pilot valve air header trip valves open ' i ' b. AO-4350 shuts to isolate the service air headert AO-4365 shuts to ) isolate non-essential instrument airt Feedwater control valves lock-up Scram pilot valve air header trip valves open c. A0-4350 shuts to isolate the service air headert AO-4365 shuts to f isolate non-essential instrument air Scram pilot valve air header trip valves open Feecwater control valves lock-up i , d. A0-4365 shuts to isolate non-essential instrument airl AO-4350 I shuts to isolate the service air headert Scram pilot valve air . header trip valves open Feedwater control valves lock-up i JESTION: 098.(1.00)
Th2' plcnt is at 100% power when the RECIRC PUMP A SEAL STAGING HI FLOW-annunciator alarms. The operator performs the actions per ARP 904C-H3 and t crcc; dure No. 2.4.22 (Failure of recirculation pump seal). and gathers the . 011owing informations i CRD seal flow: 4 GPM Rocare pumo seal cavity #1 pressure: 1010 psig R; circ pump seal: cavity #2 pressure: 1005 psig - l Drywell: pressure: 0.7 psig Drywell floor drain leakage 1.25 GPM No other-Abnormal alarms I ss 3 ELECT the type of failure that has occurred.
L c. Failure of #1 seal , b ' b.. Failure of #2 seal i c.-Plugging of #1 internal restricting orifice d.
Plugging of #2 internal restricting orifice
. i
Pcq3 50' '. ' + . > l
t
1UESTION: 099 (1.00) ' i'. r~0ctor Startup 'is in progress with power at 25% when the operator
- nztic;s=the in-service CRD flow control valve oscillating causing CRD l
SyCtCa flow to oscillate. Proceoure No. 2.4.11 (Control rod drive m0lfunctions) is entered. In accordance with Procedure No._2.4.11 SELECT ! th] c;ndition that would require the operator to immediately scram the t rCCctor for this situation, i O. 2 or more accumulator trouble alarms r b. 2 or more control rod drifts occur in a 9-rod array [ c. 2 or more CRD high temperature alarms d. CRD system flow is lost and cannot be immediately restored f i UESTION: '100 (1.00) l r Drac;fure No. 2.2.19 (Residual heat removal) contains a precaution for the i minicum RPV water level to maintain during a cooldown in case forced
circulation (retire pumps or RHR pumps in shutdown cooling) is lost. SELECT [ the cinimum RPV water level which will ensure a natural circulation 'flCwpoth'in thefreactor.
! , c. +35~1nches
b.
+41 inches ! c. +46 inches
Ld. +52 inches y i' . P
! , D $ v P (********** END OF EXAMINATION **********) ,
- - . - -
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- LPs, Procedure,No. 6.1-012 (Access Control'to High Radiation Areas) pages 7-9,16 JOBJ None Kit 294001 K1.03 (3.8)
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t ., i_ 'NCWER:- 035 ~ ( 1 ; OO ) : , , . c.
.. EFERENCE:
': P:; O-RO-02 06-03 Control Rod Drive page 31 OBJ: None -<A:c201002 K4.03!(3.6) 201002-A2.02 (3.3) 201002A202- ~~201002K403 ..(KA's) i
- il b
,
,
h,.. -
li; :ir.
' , gsg,,.. ~. x ' pcq3 31 "s , <.. w - 4 -
"- , , , 0, i 70 5 ' l: ' '!-il, i i NSWERs.
036' ( 1. 00 ) - -bJ ,
i h >, 'lEFERENCE: , i ,LP 0-RO-02-07-01' Neutron Monitoring-page 24
iLOBJ :: ELO-48 -
- KAir21SOO2 K3.02.(3.6)
21DOO2K302 ..(KA's)- > I - .. - ... . . WSWER : 1037.(1.00) i d , a
2FERENCE: iip.silO-RO-02-01-01' DC Electrical Distribution pages 17,18
- 1BJ ELO-1,2 '
U$AE263OOO-K3.03 (3.8) 215003 K6.02'(3.8) 215004 K6.02 (3.3) ' 21SOO4K602 263OOOK303 21SOO3K601 ..(KA's) , 'l NbWERaj 038 (1.00) t x b ' ( . . .IFERENCE: . , A-r -- , . 'PtyO-RO-02-06-03. Control Rod Drive pages 59-63 t, . .3BJa" ELO-29 . > YA s ": 201006 k4.02'(3.5)- l 201006K402 ..(KA's) 't , y . o JSWER ' 039c '(1.OO L ? b;l > ' i , , r > t t ._lJ,. ' - k ' s \\ i t i <
')~ 1;
V ' (b ....'....u
- PCg3( 62;
. -
'" [.' ;,f
' e,. INEFl ENCE:- p , LP:~0-RO-02-09-011LPCI and RHR page'12 ' . p3-RO-02-09-01~ Primary Conteinment page 35-OBJaiELO-2,13.14 ' - ELO-27,28.29- . ..KA:E205000'K4.03 (3.8) -205000 K4.02 (3.8) 205000K403-205000K402! ..(KA's) < \\, hNSWET!:- '040' (1.00) [ '6 . t n {jEFERENCE: ELP: 0-RO-02-09-01 LPCI and RHR pages 44,45 OBJ LELO-8,10,12 ! KA: '203000 A2.04 (3.6) 203000 A2'.16 (4.5) 203000'K4.11-(3.5) L 203OOOA204-203OOOA216 '203OOOK411 ..(KA's) "tNSWER: 041 (1.00) c I
- 'EFERENCE:
' ~ LP:.O-RO-02-09-01 LPCI and RHR pages 40,41 ?OBJ: 1ELO-8,10,13- 'kAsJ203OOO K4.10-(4.1).
- 203OOOK410 ..(KA's) lNSWER: 042 11.00) b i' LEFERENCE: PLP LO-RO-02-09-03 HPCI pages 14,15 i
- OBJ a,' ELO-2,5
' KAsc-~206000~K4.09 (3.9) -206000 K4.11 (3.5) 206000 GOO 7 (4.2) ! 9' 206000K409, 206000K411 206000 GOO 7 ..(KA's) . I lt .. i; ; ! \\+ by'L . j
d " [.m (. J.
Pcg3~63
. n.
.. - '
- p
- lkNSWER: 043 .(1.00) d "EFERENCE-
- iP. -lO-RO-02-06-06 Standby Liould Control oage 14
'OBJ -ELO-4 <KA '211000 A4.03-(4.1) -211000A403 ..(KA'fs) nNSWER: 1044- (1.00) b' EFERENCE: LP's.0-RO-02-08-01. Primary' Containment pages 33,35,42 OBJ.ELO-27- 'KA '223OO2 A1.02 (3.7) 223002 A2.09 (3.7) 223OO2 G011 (4.1) 223OO2A209 223OO2A102 223OO2G011 ..(KA's)
- NSWER 045 (1.00)
. b i <EFERENCE. . [LP lO-RO-02-08-01 Primary Containment-page 31 TOBJ: ELO-27.29 ' t<As/223002 K4.04 (3.6)
- .]
223OO2K404.
..(KA's) ,ii ~l LNSWER-046 (1.00) 'i ' c
4 t{ )
k
- {
< . l,. A ' Pcgo 64 ., ,
- \\
3EFERENCE: LLP LO'-RO-02-01-02 AC-Electrical Distribution .TP-5-OBJa~ELO-13.
- KAs 262001 K3;Ol'(3.7).
1262OO1K301 ..(KA's) ANSWER:- 047- ' ( 1. 00 ) l - e, ~
1EFERENCE: ,. LP : 0-RO-02-09-06 Diesel Generator System pages-23,24 OBJ ELO 20,21 KAs-264000 A2.10 ( 4. 2 )'- .264000A210' ..(KA's)' , wNSWER:- 048 (1.00) i b 'EFERENCE: LP: 0-RO-02-09-06'Diese1IGenerator page 10 OBJ: ELO-12 ' .KA:f.264000 K3.02-(4.0) 264000k302 ..(KA's)- uNSWER: 049- -( 1. 00 ) =. d !EFERENCE: ' JLP: 0-RO-02-07-02 RPS and ATWAS~ page 17 LOBJ: ELO-4' KAa1212000 A2.02 (3.9) 212000 GOO 7 (4.2)
- 212OOOGOO7 212OOOA202
..(KA's) n.
.
g a - ' , - < q,_..g N;g;. ~ ' i, . Pcg2 65' c , . . 5:s: + r ,, ! -LNSWER ' s.050 (1~.00)- ' . "c- ,
il i.
- LEFERENCE:
t-
=LPs.O-RO-02-06-01 Non-nuclear. Instrumentation -page 32 EOBJ: ELO-2.10 EKA31216000 K5.06.(3.6): 216000 K5.07-(3.8) D 21'6000K506' 216000K507 ..(KA's) k , > , LNSWER: '051 (1.00)
- o.
~ i .EFERENCE. I;iLP's- 0-RO-02-04-02 2. Condensate and Feedwater page 50 10BJ : ELO-81 .,KA '259002 K6.03 (3.1) i
259002K603 ..(KA's) - I ~ ii . s WSWER. 2052 (1.00) -j
. .. .; c.
' l: , .IFERENCE:
- q
.,P lO-RO-02-01-03-Main Generator page 36 I .O-RO-02-01-04 Generator Gas Control page 10 -]BJEELO-4: ELO-6.
' t<A3J2450001K6.05-(2.9) 245000 G010 ( 2. 9 ).
245000G010 245000K605 ..(KA's)
! q 1 S M E R_:' <053-(1.00) ' ,- t l, C Q . i , l i '! ! wj ,
q.= + --- ,-
, a ., . , t. y.. .
- ,7
>, = PCg3 16~ =,, o, ..: , N' ,d, lEfERENCE
(LPiiO-RO-02-04-02 Condensate'and'Feedwater page 32
!;OBJ :l ELO-59,70. 7 5. 79 - .cKA s.c 259002 ' K1'.02 - ( 3; 3 ) 259002-K4.10 (3.4) -259002K410-259002K102' ..(KA's) ,
bNSWER:. 054 (1.00) , , .. O- . r IEFERENCE:
- LPDO-RO-02-04-02.
, . Condensate and,Feedwater page 49 (OBJ: ELO-29 ' Ek A V259001 A2.02 (3.'3) 259001 A2.04 (3.4) 245000 A2.06 (3.1) > .259001A202 '245000A206 259001A204 ..sKA's)
rt t [NSWER . 055 I -(i.00)
- d e
'EFERENCE ' , 7.
-LPsLO-RO-02-04-02 CondensateJand Feedwater pages 20,31,38
2-3 B J 1ELO-13,46,72 ' .KAs 259001'K6.01-(3.0) 259001 A2.05 (3.0) ' 259001K601 259001A205 ..(KA's) >
':NSWER: .056 (1.00) , t 'C , .\\EFERENCE. \\
- .P ;O-RO-02-09-02 Core Spray pages 12,13 13BJ 'ELO-6,7,12a
'
- U@t 209001 K4.04 (3.2)
209001 A2.05 (3.6) 209001A205.
209001K404' ..(KA's) ! , Y l . m- - - . -
,i ,
4 , y Pcga GB j ' R JiL '. fi ' ... {llEFERENCEs- 'I (LP: 0-RO-02-04-04 Hydrogen-Injection Chemistry- 'page 17 - i* OBJP ELO-15' , f zy KA 294001 Al'.14-(3.4). _ 10CFR55.43 (b)('4)' I J 294001A114 ..(KA*s)'. , ' ' ANSWER:- 061, (1.00) ,
- d
! 'iREFERENCEs-9 -_
. .. . j D LPI- 'O-RO-03-04-02 EOP Development'and Use pageJ12 [1 l Procedure No. 1.3.6 (Tech Spec Adherence and Clarification) o U 'OBJ:JELO-3.1.8 ... ( K A 294001 A1.02 (4.2)
['s 294001A102 ..(KA's)- bit - , w! ANSWER: 062.(1.00); I e , t: - ' n.
a.
' g'
- i dREFERENCE:
' LP,0-RO-03-04-2 EDP Developmen t and Use page 21 LOBJ: ELO-5: KA:l2940011A1.02 (4.2) <. c.. 294001A102~ ..(KA's) l y lEQi' , t ,jANSWER: 063 (1.00) b
> b WREFERENCE: lf,v i _ fij's: 0-RO-03-04-02 EOP Development and Use page 40 '
- N; OBJ ~s' ELO-19 t KA
- "295031'K1.01 (4.7)
y 295031K101- ..(KA's) ' , ? s f__ l ' , ' e Iky :'f[%. :, .. /S i 'D i t-g . l . . , t
,'1 7 , ,:4 f*,. Pag 3L69 , ! -n i L'ANSWERi ~ ~ 064-(f.OO).
, b IREFERENCE: lLPt.0-RO-03-04-02 EOP Development and Use page 54 - -0-RO-03-04-04.EOP-02 (Failure to scram)- .,-; OBJ t ; None.
. .. E L O - 1 2 ' 'KAs-295037 K2.09-(4.2)- ~295037 GOO 7 (3.9).
, 295037k209 295037 GOO 7 ..(KA's)
- ANSWER '
065 (1.00) b- " REFERENCE: ,[LPa[0-RO-03-04-02 EOP. Development and Use-page 13 OBJ ELO-3a' ' KA t' 294001 A1.02 (4.2) - 294001A102 ..(KA's) ANSWER: 066- (1.00) c
REFERENCE: ' ' LPt;O-RO-03-04-03 EOP-01-(RPV control) page 33-OBJa.None,
- KAsu295025 GOO 7 (3.7) 295025 K2.01 (4.1) 295025 K2.OO (3.7)
'295025 K2.04 (4.1)L295025 K2.05 (4.2) - 295025 GOO 7' 295025K201 295025K208 295025K204 295025K205 ...(KA's)
ANSWER: 067 (1.00) - a f - T 't
-
ag -.;; <l ;a.[ -1
- Pago!_70 :
, so 1 lL lg REFERENCE / ,
- LP' 10-RO-03-04-03.EOP-011RPV-Control _ page 21
'OBJ1ELO-le-1KAs 295031 GOO 7-(4.0) 295031 GOO 71 ..(KA's) ... JANSWER:- 068- (1.00)- -b . ' REFERENCES.
- LP 0-RO-03-04-04 EOP-02 Failure to-Scram page ~ 17
, OBJ:-ELO-8- - KA 1295037 GOO 7 (3'.9) 295037 K3.03 (4.5) /295037 GOO 7-295037K303 ..(KA's) 'JANSWER: 069 '(1.00)- b.
REFERENCES- - LP f0-RO-03-04-05'EOP-03 Primary Containment Control- - page 2 nOBJ VELO-2z - KA t: 295012 G011':(4.4) (295012G011- .-. ( K A * s ) ANSWER: 070-(1.00)- b > o REFERENCES-t- ELPt O-RO-03-04-05 EOP-03 Primary Containment Control page 9 OBJ -ELO-2 LKAs 295029'G011 (4.5) 295030 G011 (4.5) < W 295029G011 295030G011 ..(KA's) n1 ! t i
V e , ' > ,.4 1, . ',. > < ,Pcgof71' .. [,-.,. , - 'lk
r ! ' [' ANSWbR:- -071 (1.00) o - ' ' i I
- REFERENCE
- LP:VO-RO-03-04-05 EOP-03 Primary-Containment Control page 20
' OBJ: ELO-7; 104: 295024 : K3. 01 - ( 4. 0 )- ' -295024K301 ..(KA's) > t ' ~-ANSWER: 072. (1.00) a , REFERENCES, .t ' (J's : 0-RO-03-04-05 EOP-03 Primary Containment Control ~page 12 ' OBJr.-ELO-15 - ~
'KA '295028'K3.01-(3.9) - - ' 295028K301-' ..(KA's)' l,- 4-l ' ANSWERS' 073 ?(1.00).
d ', i _ REFERENCES > LP.: 0-RO-03-04-05 EOP-03 Primary Containment Control page 28 OBJt-ELO-5-KA /223OO1 G010-(3.6) 223OO1G010 ..(KA's).
-. > EANSWER - 074
- ( 1. 00 )
, y . . i [ -. - - - - - - - -- -
,;, - - . - - - ' Pco-) -72 > , -.,;.... y 1: */. ", . ?;> , iAEFERENCE: 3 p.. '. .. . -- . . . page 13 ,
- LPs..O-RO-03-04-06 EOP-04 Secondary' Containment Control
. i OBJ s.. ELO-5 - t ' KA 295032 K3.01;(3.8)-
295032K301 ...(KA*s)' - s ANSWERS,. 075 (1.00)' C -, c,-
- REFERENCE
h . P LPt.-0-RO-03-04-06 EOP-04 Secondary Containment Control.. page 42: .~-OBJ: ELO-1 - KA :295032 G011 (4.2)- > '29 5032G011 - ..(KA's) i 2 ANSWER: -076-(1.00) ! ^d s , ' REFERENCE:-
-LP: 1.0-RO-03-04-08 EOP-06. RPV: F 1 coding pages 12,38- . .OBJ:.ELO-8 ' ' , KAt: 295008' GOO 7 (3.3) d 295008 GOO 7 '.'.(KA:s) -i ' ANSWER: 077 (1.00) 'b . , a .. REFERENCES- . ' -1 l 'LP: UO-RO-06-01-O3' Limiting Conditions for Operations page 9
- Tech Spec 3.1 Table 3.1.1
, . OBJ. - ~ ELO-1,3 KA: 295015.GOOB ( 4. 2 ). , 295015 GOO 8' ..(KA's) ,o l l i ! ..
" y kI f., - -* ' - .Pcga.73: r .x li ' s ) ". [[ii k-ANSWER:: IO78-- ' ( 1~. 00 ) ' -i , .. - ; <
- > P . REFERENCE:. .
LPri-O-RO-06-01-03- -Limiting Conditions for Operations page 51 'OBJ ELO-3 KAs 295014'GOOO-(4'.3) h295014GOOO- ..(KA's1 d , t , ' i.
c, , ' C%.. . _.
.
iANSWER - .079 (1.00)' . a d-M !
- ' REFERENCE
' LP.: -0-RO-06-01-03 Limiting. Conditions for Operations pages-27,44 Tech Specs 3.5.c', 3.7.d' . -OBJ.'ELO-3' .. . J KA: 295020~GOOOD(3.9) ^295020GOOB' ...(KA's)-
l[ <- ANSWER:-. 000. (1.00) , t I-a- ?! t r
YREFERENCE:; . . ' i [_ .LP t0-RO-06-01-01 Tech Spec Definitions page 14 OBJ: ELO-3m ~KA::'295023 GOOB (3.9)
, -295023GOOO ..(KA's)
5 3.h-q ANSWER: 081-( 1. 00 ) ' -b f
- 1 r
? {' - 7
~~ w.
) ' E ". L e,.' . Pg,g7. 74 Lf .., p a u 'i., > , t ' r
' k ?}EFERENCEs_ '] ' LPs'O-RO-06-01-01~-Tech Spec Definitions page 12 l OBJslELO-3g ^f - KA 1295006 GOO 8'(4.3)-
295006GOOB ..(KA's)~ ] ' ANSWER: -082: (1.00)- ' d- ]
, < REFERENCE: ' l 'LP 10-RO-06-01-03 Limiting Conditions for Operations page 37 j OBJ: ELO-4-
KAs-295001-K2.05 (3.3) 295001. GOO 4 (3.7) ' ' 295001K205 295001 GOO 4 ..(KA's) ' ANSWER: 083 (1.00)- ' a -
. . REFERENCE:
, LP:JProcedure No.'5.3.6-(Loss of-vital AC Y-2) page 2 ! OBJ1 Procedure No. 1.3.34 (Conduct of Operations) section 6.12 , " ._ K A 1295003 G010'(4.1) - .. 295003G010 ..(KA's) . [ ANSWER: 084. (1.00) f 'a: ~ - -REFERENCE: ..
..LP: Procedure No.ES.4.3 (Refueling floor high radiation) page 7 O-RO-03-04-06 EOP-04 (Secondary containment control) ~OBJ: None ' ELO-1 =KA::295023 GOli-(4.2) 295023 GOO 7 (3.6) 295023G011-295023 GOO 7 ..(KA's) . ? T t-l '
. ... m
,. _. - n-o .? Pogo;7DJ .. . .t <= , .
- $i1
' ANSWER, 085 (1.00)- . . 't a ,(, ! n
- EFERENCE:
' LP s ' Procedure No.: 5.3.31 (Station Blackout) page 2 i! - . Procedure No. 5.3.26 (RPV injection during emergencies) page 4 OBJ 'None- , DKAs.295003 A1.03 (4.4) 295003 K2.03 (3.9) . -295003A103 295003K203 ..(KA's) I s.,_.. . . JANSWER: 086 (1.00) b.
,. _ '~ REFERENCE:
' LP: Procedure No. 5.3.23 (Alternate rod. insertion) page 10-
- OBJ: LNone-d 60 4 : 295037 K2.05 (4.1)
295037:K3.07 (4.3)
295037K205 295037K307 -..(KA's) s_. !: ANSWER: '087 (1.00) llL > j.
c= ,4 REFERENCE: - <; JLPt. Procedure _No. 2.4.143 (Shutdown from outside the control room) page 2 1.
lOBJ 2 Procedure No. 1.3.34 (Conduct of Operations) section 6.12 JT [ KA:?295016 G010 (3.6) 295016G010 ..(KA's) ' ' .l ' % ANSWER:- '088 (1.00) ' _ C.
L I (i.* t' ( ,
- !!
s' ,
_.. _ Mg 'PCg3-76L ' ' < , , s., t '. s. ; j _, REFERENCE: , , 'LP ProcedureLNo."2.4.150-(Loss of feedwater heating) _ page 2 - q 'OBJ'{ Procedure:No.~1.3.34-(Conduct of Operations) section 6,
"- KA 1295014 G010 (3.9) 295014G010' ..(KA's) ^ t ANSWER: 089 (1.00)' d REFERENCE: , LPs' Procedure No; _2.4.23'(Jet pump flow-failure) page 2 'OBJ -None-KAs->295001 A2.05 (3.4) 295001 K3.06 (3.0)- 295001A205 295001K306 ..(KA's)
- o
' ANSWER :- 090 _(1.00) s , c , o,i ~ REFERENCES-lLP -- ProcedureL No.12. 4.17. ( Recircul ation ~ pump.: trip ) D'!: OBJ:-Procedure No. 1.3.34 (Conduct ~of Operations)_section 6.12 KA s --295001 K1.04 (3.3) 295001: K3.04 (3.6) .295001K104.
295001K304 -..(KA's).
-4 , . ANSWER: 091 (1.00).
b' REFERENCE: .'LPs Procedure No. 2.4.36 (Decreasing condenser vacuum) page 4
- OBJ:. Procedure No. 1.3.34 (Conduct of Operations) section 6.12
- KA:~295002.G010 (3.7)
295002G010 ..(KA's) I
g.- ,y v
(7:= ' F _ . P;g].77 ,.s . , j= -j, ;'4 4 , ! 4, . I ' P.
v . <- ?N1'WER: 092. ~(1.00)- V a ! 3 30 + br !~ + , F t 'l .: REFERENCE: ' ' "
. . LP -Procedure.No. 5.3.30: (Loss of.250 vDC power bus'D-10) { ' " OBJ:1-None .
. . [ (KA 295004.K2.03~(3.3)l [ L '295004K203 ..(KA's)
i-^ ' { .p 'AN2WER: 093 -(1.00) C
- REFERENCE:
-f LP -Procedure'No. 2.4'46 (Turbine bearing malfunction) page 2 . - OBJ: Procedure No 1.3.34-(Conduct of Operations) section 6.12 KA:-295005 G010 (3.6) . 295005G010' ....'(KA's): J ANSWER t. 094 ( 1. 00 )- b'
REFERENCE:" - LP:LProcedure No. 2.4.37 (Turbine control system malfunctions) y ;OBJt'--None " ~KAsl295005 K1'.01'(4.l') 295005K101 ..(KA's) ' , l ' '- . ANSWER:- 095 (1.00)
, ? -l' , l f F 3,: m
- ,.
- -. . , _. - - . . .
- im m;
, ..:
, ', bi,?. .,i; P:q):78 p; *, < p(.' c , " ,,fi ... 1
[NCE :'.. IV J'r;.c ; dure ' No. - 2. 4. 47 t (Loss of RBCCW) pageL6
- Ning:
195010 K2.01 (3.4): [5218K201 ..(KA's) e rai
- 096
.(l'.00) .. cdc .. u, i r iNCE s.
! r:;cadu re No'. 2. 4. 41 (Loss of TBCCW) page 12 \\ N::.n 2 " 95018'K1'.01 (3.6) . 3301BK101 ..(KA's)- l l--
,t: 097 (1.00).. . .y _ .- , VCE4 ' -RO-02-02-04-Instrument and High-Pressure Air
- ocGdure No. 5.3.8-(Loss of instrument air)
page 2 . f!LO-7 , lony '?5019 A2. 02 - ( 3. 7 ) i 5019A202 ..(KA's) .,; ' s , !: >098 (1.00) n: , + j u .c Of'. i
! ! . .\\; - O, i
., ;, ' - m7 -- , t r y,. , ,,, "
- Y PCg3;79:
y :.p.; ;, # '-.; .j.' l'7EhECENCE: bLP 'O-RO-02-06 O2-Recirr.ulation' System TP-9.
d k . Procedure'No.
2...' 2 2 (Failure of recirculation pump seal)
40BJa-ELO-13- '-None: ['K :i295001 GOO 9-(3.4)- ' L J295001 GOO 9 ..(KA's) >. M -, la - , ANSWER: 099- (1.00) ,, n.
b' p ' PEFERENCEs-LP: Procedure No. 2.4.11'(Control rod drive system malfunctions)-page.2 OBJavProcedure No.!1".3.34 (Conduct of-Operations) section 6.12 KAt'295022 G010 ( 3. 5 )' - 295022G010 ..(KA's)
A ANSWER: 100 -(1.00) '.c REFERENCE: 'LP 1 Procedure ~No.-2.2.19-(Residual heat removal) page 126 -
- OBJ
- .None.
.
- KA's 295021~ K1.04 - ( 3. 7-)
295021 K3.01-(3.4) 1295021K104 .295021K301.
..(kA's) I
J ' -l l - ! , ' (********** END OF EXAMINATION **********) .i b '
. - _.
. " -j - -. .. * pr< ATT A CHM EM T 2- , ' L gg
. Pilgnm Nuclear Pomt station - Rocky Hill Road -) , Plymouth, Massachusetts 02360 , , Ralph G. Bird february 1, 1990 Senior Vice President - Nuclear BECo Ltr 90-020 J - Mr. Hilliam Russell, Regional Administrator J U.S. Nuclear Regulatory Commission j 475 Allendale Road ... j,. King of Prussia, PA 19406 > w Docket No. 50-293-License No. DPR-35 i
Subject: WRITTEN COMMENTS ON NRC SENIOR REACTOR OPERATOR LICENSE " EXAMINATION-
Dear Mr. Russell:
In accordance with NUREG-1021, Section ES-201,_ Attachment 2, Enclosure 4, the l l Boston Edison Company's (BECo) Operator Training Staff has prepared the ' L enclosed written comments for your review and consideration.
The NRC written examination was administered to ten BECo. Senior Reactor Operator license candidates on January-29, 1990.
<
' L If you have any further questions please do not hesitate to contact Mr.
h Harrison R. Balfour,-at (508) 747-8602.
' - p.
- D
_ R.
ird ' . JLK/dmc l Attachment ' ' cc: Mr. Robert Gallo w/o attachment Operations Branch Chief U.S. Nuclear Regulatory Commission ._ 475 Allendale Road , King of Prussia, PA 19406 L , .
a-
, - -:
.. . 8. -.6? . - 'is : Page 2-l cc: Mr. Richard Conte - w/o attachment - BHR Section Chief
U.S. Nuclear Regulatory Commission 475 Allendale Road
King of Prussia, PA 19406 . o _ Mr.LTodd Fish - w/o attachment License Examiner ' U.S. Nuclear Regulatory Commission 475 Allendale Road - - King of Prussia, PA 19406 .
Sr. NRC Resident Inspector - Pilgrim Station - w/o attachment
L ( r > . . - O d I -
.z o
- !
>
- b g{gg.50320/4-f
)
iOVESTION 3.0: . 1.00 Pbints) ( t . . Select the item which would NOT be considered a Level 1 ROR as; defined in l ' Procedure No. 6.1-209'(Radiological. Occurrence Reports).
! , a. A major RHP violation < -
h b. Radiation exposure that'would require a 24 hour report to the NRC ' a } *l fc. Radiation exposure that would require _a 30 day report to the NRC ' p .d. A radiological event with potential for a press release j ANSWER < i ' rE a.
O . + CDttiEf!15:
. . 3 !He recommend that this question be deleted from the exam.
The NHE is '. not required to know this information from memory.
- >
,. , n' Procedure 6.1-209 requires the;NHE to be able to process.R0Rs only if i L
Senior Radiological personnel cannot be contacted within 30 minutes of '"4 the occurrence, these personnel are available 24 hours a day.. The.NHE' ' is given a completed ROR and verified the responses of the questionnaire , lwhich has already'been completed by the originator.
The NNE is not
,T required to know what specifically constitutes a Level I ROR, but how to p ' ~ verify a Level I ROR exists. l Additionally, this:is not a task
-Identified within the job task analysis for either the Reactor =0perator ' ? .or Senior Reactor Operator position at Pilgrim Station, l; f } REFERENCES: .t y s
- PNPS Procedure 6.1-209 Sections 6.0.[1], 7.2.[2],- 7.3.[1] and 7.3.[2]
- Standing Order 89-02 of PNPS Health Physics-Department
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- INPO.Accreditted Job-Task-Analysis for the positions of R.O. and S.R.O.
at Pilgrim Station
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'00EST10N 7.0: ( 1.00 Points) Procedure No.-1.3.34 (Conduct of Operations) gives-guidelines for performing system configuration verifications and independent verifications.
SELECT the statement which is NOT in accordance with the guidelines of Procedure No. 1.3.34.
a. Lifted leads and jumpers-require independent. verifications b. Two individuals can be permitted to work together when repositioning valves c. Valve positions for an inaccessible valve is first verified by checking isolation of the power supply to that component .d. If a component is inaccessible and a tagout already exists for that component, the independent verifier can use the existing tagout to.
] determine position ANSHER: c.
,_ ~l COMMENTS: 7.:He recommend that answer (a) also be accepted as correct.
Procedure 1.3.34 specifically excludes lifted leads and jumpers from Independent , ' i Verification.
REFERENCES: j.
PNPS Procedure 1.3,34 Sections 6.5.[3] and 6.5.[63.(a).
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' q , L SELECT the emergency action level whose-definition is: "a radiological-release-is likely but no core degradation.is indicated".
, j a. Unusual event j b. Alert-c.LS1te' area' emergency- ! ~ 'T 'd. Genera 1l emergency.
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p .. COMENTS: , We' recommend that both answers B and C be. accepted for full credit.' n '., ' - ~'? PNPS EP-IP-120 and 130 define ALERT and SITE AREA EMERGENCY consistent with E the definitions in NUREG 0654. These references. imply that a releaseLis a-
E consideration in both of these EALs.
The difference is'that the release considered in an. ALERT is. expected to be a small fraction of-the EPA - Protective-Action Guideline Exposure Level.
The definition stated in p.. f question 13: refers to a release but does not specify its magnitude relative E . ? -to.these. Exposure levels.
In addition, NUREG 0654 11sts examples of an.
- Alert. ' Example 15 describes an offsite release.
As such the-definition stated.in the question. accurately describes =EITHER an Alert OR a-Site Area ' Emergency.
, ' References attached: NUREG-0654 pages 1-8, 1-9,.1-12 & 1-13 PNPS EP-IP-120 page 3 , p.; .PNPS EP-IP-130 page 4 , t .i
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. 950320/54 '*.e Y~ / ObESTION 53.0: f(1.00 Points) ,i . . Procedure No. 2.1.1 (Startup from Shutdown) directs-the operator to-i .c: . transfer FHLC to'"3-element control" at 30% power. _ SELECT the bases for i ' not transferring to "3-element' control" before 30% power.. < -a. An' interlock prevents placing "3-element control" when feedflow is less.
than 20%'of rated ! b, At low power levels the FHLC level input is inaccurate l c. At low power levels the FHLC steam flow input is inaccurate d. At low power levels.the FNLC feedflow input is. inaccurate.
ANSWER: , ,. , . ' c.
' ' COMMENTS: . ' " . . .. 53.
Recommend that c;. or d. befaccepted as. correct answers for the following- . reasons: i ! 1.
PNPS' Procedure 2.2.82, Reactor Vessel Water Level Control System, section 4.2'[5]: " Control system should be changed to'three: element, control when the steam flow and feedwater. flow'are significantly on ' scale, which isLusually 30%-power".
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Lesson plan 0-RO-02-04-02, Condensate and Feedwater System, page 35:- A "During-single element; control, reactor water level is the only input signal as-the steam flow and feed flow detectors are inaccurate during ~ .this time".
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Feedwater Level Control System Reference Text, page 16: "During reactor startup, since the steam and feed flow detectors have ' poor accuracy at low flow, the feedwater control is in single element automatic control".
t REFERENCES: 1. PNPS Procedure 2.2.82, Reactor Vessel Water Level Control Syste, Section '4.2[5] -2. Lesson Plan 0-R0-02-04-02, Condensate and Feedwater System, page 35 3. Feedwater Level Control System Reference Text, page 17 - - -. -. - t te-w -
. ? ' 3Q ATTACHMENT 3 Question 003: COMMENT NOT ACCEPTED Procedure 6.1 - 209 (Radiological Occurrence Reports) clearly _ states on pages 7 and 9_ that the Watch Engineer is one of the people designated'with the-responsibility of verifyingian ROR as a level 1.
Given this potential for having to make this verification as stated in the procedure, an SRO is expected to have a conceptual knowledge of what type of events constitute level 1 RORs. The four choices given to answer question 003 clearly discriminate between an SRO that has no concept of what a level 1 ROR 1s and one that does. The question does not probe for minute differences between level 1 and level 2 RORs which would~ require the procedure for reference.
Question 007: COMMENT NOT ACCEPTED Originally answer (a) was written as; " Lifted Leads and Jumpers are exempt from independent verification." At the pre-examina-tion review conducted on January 25, 1990 with the members of your facility as stated in this report, the NRC evaluators were told by the facility representatives that lifted leads and jumpers ~ were indeed required to nave independent verifications and that the NRC should delete the word " exempt" and state that they are required to have independent verifications.
The question and answers as given to the applicants are exactly _what the licensee requested and no further changes wi_11 be made.
. Question 013:- COMMENT ACCEPTED The answer key was changed to accept B or C.
^ Question 053: COMMENT ACCEPTED - The answer key was changed to eccept C or D, however, pages 21 and 35 of lesson plan 0-R0-02-04-02, Condensate and Feedwater system have conflicting information and should be corrected.
Page 21 states only the steam flow signal is inaccurate whereas page 35 states that both the steam flow and feed flow signals are inaccurate.
g z- _ p [b II* t. 3,'4 p-, t E , L < ATTACHMENT 4 SIMULATOR FIDELITY REPORT Licensee: Pilgrim Docket.Nof 50-293 Operating Test Administered on:: January 30, and February 1,1990 i During conduct.of the simulator portion of the operating tests, the simulator's .i . performance, throughout all scenarios, was satisfactory.
No deficiencies were i p noted-finis assessment does not constitute audit or inspection findings, nor: 'o does it affect NRC certification of the simulator.
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,.'b ; e NY: , g i ATTACHMENT 5 PERSONS CONTACTED "' . Individuals-Contacted Notes * h A, Shiever 0perator. Training Supervisor (1, 2, 3) _ ! T. Sullivan, Chief Operating Engineer (1, 2, 3) ^ ! [ LW. Green, Senior Nuclear Training Specialist (1,2,3) H. Balfour, Operations Training Manager (2, 3) ' K. Highf111, Pilgrim Station ' Director-(2) ( J. Alexander, Nuclear Training Dept. Manager _(2,3) j- .L.- Olivier, Operations Manager (2,3)' r.: -P. Hamilton, Compliance Division Manager (3) r R. Anderson, Plant Manager (3) J. Kelly, Compliance Engineer (3) ':
- Notes:
I .(1) Attended Pre-Exam Review on January 25, 1990 (2) Attended Entrance Meeting on January 29, 1990 (3) Attended Exit Meeting on February 2, 1990
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