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Initial Operator Exam Repts 50-277/96-10OL & 50-278/96-10OL on 960913-20.Exam Results:Eleven of Twelve Applicants Passed Exam & Were Subsequently Issued Licenses & One SRO Applicant Did Not Pass Operating Portion of Exam
	ML20129H411
Person / Time
Site:	Peach Bottom
Issue date:	10/24/1996
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20129H398	List: IR 05000277/1996010; ML20129H394;
References
	50-277-96-10OL, 50-278-96-10OL, NUDOCS 9610310278
	Download: ML20129H411 (142)
	v • d • e

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

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Docket No ;50-278 License No DPR-44; DPR-56 Report No (OL)

Licensee: PECO Energy Company P. O. Box 195 Wayne, PA 19087-0195 Facility: Peach Bottom Atomic Power Station, Units 2 and 3 Dates: September 13- 20,1996 Chief Examiner: D. Florek, Sr. Operations Engineer, Region 1

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Examiners: C. Sisco, Operations Engineer, Region 1 B. Ferguson, NRC Consultant Examiner Approved by: Glenn W. Meyer, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety

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9610310278 961024 PDR ADOCK 05000277 V PDR

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EXECUTIVE SUMMARY Peach Bottom Atomic Power Station, Units 2 and 3 Inspection Report 50-277/96-10 (OL) and 50-278/96-10 (OL)

Operations The applicants were well prepared for the operating test. As a result,11 of 12 applicants passed the examination. Eleven applicants were subsequently issued licenses. One SRO applicant failed the operating portion of the examination. The examiners noted areas of consistently good performance in the simulator, including crew teamwork, crew communications, and SRO prioritization techniques, but also noted some minor areas of weak performance on the written examination and walkthrough (Section 05.1).

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', L Operations s

05 Operator Training and Qualifications ,

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05.1 Operator Initial Examinations

. SCODe i The examiners administered initial examinations to six RO and six SRO applicants in accordance with NUREG-1021, " Examiner Standards," Revision Observations and Findinas ,

I The results of the initial examinations are summarized below:

SRO RO Pass / Fail l

Written 6/0 6/0 12/0 l Operating 4/1' 6/0 10/1 Overall 5/1 6/0 11/1 ;

One SRO applicant was granted a waiver of the operating tes j

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The PECO Energy staff reviewed the written examination and assisted in the ;

validation of the operating examination during the week of September 3,1996. The l PECO Energy staff provided high quality comments on the examination that I significantly improved the examination. The PECO Energy staff who were involved '

l with the examination review signed security agreements to ensure that the initial examinations were not compromise At the exit meeting, the PECO Energy training management representative indicated that the answer key to common SRO-49 and RO-60 question had an incorrect answer. The chief examiner agreed and corrected the answer ke Based on the grading of the written examination, the following questions were missed by more than half of the applicants, indicating a weakness in understanding on the subjec SRO-8/RO 9 Knowledge of the restrictions on use of continuous rod withdrawa RO-11 Ability to predict the final reactor pressure during a heatup controlled at the administrative limi SRO-39/RO-47 Ability to predict the control room ventilation system response to a set of condition SRO-41/RO-52 Ability to predict the E12 bus response to an undervoltage conditio _ . - - ._ - ... ~ , . -

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SRO-98/R)-98 Knowledge of the use of check off list SRO-65 Ability to determine conditions that would require entry into procedure T-104, " Radiation Release."

. SRO-87 Knowledge of the allowable storage locations for rejected new ;

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, During the walkthrough portion of the operating test, several applicants performed poorly in each of the following areas:

! Restarting a recirculation pump at powe A 1 Determining the technical specification operability requirements for the RHR system during refuelin Determining the correct actions to respond to a high skimmer surge tank leve Implementing the SE-11 procedure consistent with the SE procedure structur Bypassing the APRMs when the APRM cabinets were opene Reporting degraded equipment observations to the control room when identified during the plant walkthroug During the dynamic simulator test, the following items were significant and consistent positive observation Teamwork within the crews was very goo Cornmunication within the crews was very good. Crew briefings were concise, timely, and appropriat SRO prioritization techniques were effective such that all crew members understood the SRO-directed important task During performance of the operating test, the examiners noted the following items for further PECO consideration. Each item resulted in some applicant difficulty during the examination:

Procedure ERP-130 was confusing as to whether the shift manager was personally required to notify the PEMA and MEMA of a site evacuation or whether the PECO communicator could carry out this notification under the direction of the shift manage l l

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i The recirculation flow unit comparators were not labeled on the inside of the cabinet. (The labeling was on the outside of the cabinet.) This caused one

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applicant to incorrectly diagnose an instrument malfunction.

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j Review of UFSAR Commitments

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j l A recent discovery of a licensee operating their facility in a manner contrary to the '

. updated final safety analysis report (UFSAR) description highlighted the need for a l

. special focused review that compares plant practices, procedures and/or parameters l to the UFSAR descriptions. While performing the examination activities discussed I in this report, the examiners reviewed portions of the UFSAR that related to the j i selected examination activities, questions or topic areas. The particular sections j reviewed were Section 7.5, Section 7.7 and Section 13.2. The specific areas reviewed were consistent with the UFSAR.

! Conclusions

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The applicants were well prepared for the examination, and as a result,11 of 12 were subsequently issued licenses. One SRO applicant failed the operating portion ;

of the examinatio j Management Meetings i

X1 Exit Meeting Summary l

At the conclusion of the examination, the examiners discussed their observations of the )

examination process with members of PECO Energy management. PECO Energy ,

management acknowledged the examiner observations. The PECO Energy personnel l present at the exit included the following: l R. Artus, instructor I P. Cromwell, Nuclear QA j L. MacEntee, Operations Training Coordinator i J. McElwain, Acting Plant Manager )

D. McClellan, Peach Bottom, Manager Operations Training l T. Mitchell, Vice-President Peach Bottom 1

R. Smith, Regulatory

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Attachments:

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1. SRO Examination and Answer Key 2. RO Examination and Answer Key 3. Simulator Fidelity Report

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l ATTACHMENT 1 l SRO EXAMINATION AND ANSWER KEY

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l U. S. NUCLEAR REGULATORY COMMISSION i SITE SPECIFIC EXAMINATION

SENIOR OPERATOR LICENSE j REGION 1

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APPLICANT'S NAME:

! FACILITY: Peach Bottom 2 & 3

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REACTOR TYPE: BWR-GE4 l

i l DATE ADMINISTERED: Seotember 13. 1996

INSTRUCTIONS TO ALLPICANT

use the answer sheets provided to document your answers. Staple this cover i sheet on top of the answer sheets. Points for each question are indicated in

parentheses after the question. The passing grade requires a final grade of

. at least 80.00%. Examination papers will be picked up four (4) hours after i the examination starts.

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TEST VALUE APPLICANT'S SCORE FINAL GRADE' !

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100.00

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All work done on this examination is my ow I have neither given nor received ai l

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i Applicant's Signature

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SENIOR REACTOR OPERATOR Page 2~

ANSWER SHEET .

Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blan MULTIPLE CH0 ICE 023 a b c d !

001 a b c d 024 a b c- d 002 a b c d 025 a b c d 1 003 a b c d 026 a b c d 004 a b c d 027 a- b c d !

005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d '033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d 017 a b c d 040 a b c d 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d

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ANSWER SHEET

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Multiple Choice. (Circle or X your choice)

If you change your answer, write your selection in the blan !

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046 a b c d 069 a b c d i

047 a b c d 070 a b c d 048 a b c d 071 a b c' d 049 a b c d 072 a b c d *

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050 a b c d 073 a b c d __

051 a b c d 074 a b c d ;

l 052 a b c d 075 a b c d __

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053 a b c d 076 a b c d

054 a b c d 077 a b c d

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055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d 081 a b c d ,

059 a b c d- 082 a b c d

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060 a b c d 083 a b c d 061 a b c d 084 a b c d 062- a b c d 085 a b c d ,

063 a b c d 086 a b c d 064 a b c d 087 a b c d )

065 a b c d 088 a b c d .

066 a b c d 089 a b c d 067 a b c d 090 a b c d 068 a b c d 091 a b c d

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SENIOR REACTOR OPERATOR Page 4 l A.NSWER SHEET- ~!

l Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blan a b c d ,

l t

093 a b c d i

094 a b c d

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095 a b c d l

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096 a b c d 097 a b c d 098 a b c d !

099 a b c d ,

100 a b c d i

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NRC RULES AND GUIDELINES FOR' LICENSE EXAMINATIONS During the administration of this examination the following rules apply: I Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . 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 examinatio l 1 Restroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts with anyone outside the examination ;

room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG The point value for each question is indicated in parentheses after the questio !

l If the intent of a question is unclear, ask questions of the examiner i onl . When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In 1 addition, turn in all scrap pape l 1 Ensure all information you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the '

examinatio . When you are done and have turned in your examin'ation, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may

be denied or revoke ;

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QUESTION: 001 (1.00)

WHICH ONE (1) of the following explains proper usage of the SRM shorting links?

The SRM shorting links... ;

a. are installed during refueling outages to bypass the SRM downscale rod bloc b. are installed during refueling outages to allow the SRM high-high scram to be functiona c. are removed during shutdown margin testing to bypass the !

non-coincidence logic of the SRM trip syste ,

d. are removed whenever it is necessary to place the neutron monitoring system in the non-coincidence mode of operatio .

QUESTION: 002 (1.00)

WHICH ONE (1) of the following will result in an IRM HI-HI half scram signal being generated?

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a. Mode switch is in STARTUP: IRM A is indicating 100/125 of full :

scale on range ;

b. Mode switch is in RUN; IRM B is indicating 108/125 of full scale ,

on range 8: APRM B is failed downscal c. Mode switch is in RUN: IRM G is indicating 120/125 of full scale on range 10: APRM E is failed downscal d. Mode switch is in STARTUP: IRM H is indicating 108/125 of full scale on range i b

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SENIOR REACTOR OPERATOR l

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0UESTION: 003 (1.00)

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A control rod scram a

REGION 1 APPLICANT'S NAME:

FACILITY: Peach Bottom 2 & 3 REACTOR TYPE: BWR-GE4

DATE ADMINISTERED: Seotember 13. 1996 INSTRUCTIONS TO APPLICANT:

Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80.00%.' Examination papers will be picked up four (4) hours after the examination start TEST VALUE APPLICANT'S SCORE FINAL GRADE 100.00

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All work done on this examination is my ow I have neither given nor :

received ai !

Applicant's Signature

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REACTOR OPERATOR Page 2 I

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ANSWER SHEET '

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Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blan !

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MULTIPLE CH0 ICE 023 a b c d :

001 a- b c d 024 a b c d l

002 a b c d 025 a b c d !

003 a b c d 026 a b c d 004 a b c d 027 a b c d l 005 a b- c d 028 a b c d ]

006 a b c d 029 a b c d 007 a b c d 030 a b c d I

008 a b c d 031 a b c d 009 a b c d 032 a b c- d 010 a b c d 033 a b c d

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011 a b c d 034 a b c d

012 a b c d 035 a b c d l l

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013 a b c d 036- a b c d 014 a b c d 037 a b c d 015 c a b d 038 a b c d

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016 a b c d 039 a b c d I 017 a b c d 040 a b c d 3 018 a b c d 041 a b c d 019 a b c d 042 a b c d 1 020 a b c d 043 a b c d l 021 a b c d 044 a b c d 022 a b c d 045 a b c d i

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REACTOR OPERATOR . Page 3

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ANSWER SHEET ,

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Multiple' Choice (Circle or X your choice)

If you change your answer. write your selection in the blan .

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i 046 a b c d 069 a b c d

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047 a b c d 070 a b c d l 048 a b c d 071 a b c d 049 a'b c d 0/2 a b c d

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050 a b c d 073 a h c d l 051 a- b c d 074 a b c d 052 a b c d 075 a b c d !

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l 053 a b c d 076 a b c d 054 a b c d 077 a b c d l 055 a b c d 078 a b c d l ,

056 a b c d 079 a b c d ;

057 a b c d 080 a b c d I 058 a b c d 081 a b c d

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059 a b c d 082 a b c d ,

i 060 a b c d 083 a b c d l

061 a b c d 084 a b c d 062 a b c d 085 a b c d I

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! 063 a b c d 086 a b c d f

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! 064 a b c d 087 a b c d 065 a b c d 088 a b c d 066 a b c d 089 a b c d l 067 a b c d 090 a b c d 068 a b c d 091 a b c d

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ANSWER SHEET-Multiple Choice (Circle or X your choice)

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If you change ~your answer, write your selection in the blan l

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092 a b c d ;

i 093- a b c- d j

094 a b- c d

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095 a b c d I

!

096 a b c d i i

097 a b c d l

!

098 a b c d 099 'a b c d l

100 a b c d

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Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS M %16 During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . 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 examinatio . Restroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND 00 NOT WRITE ON THE BACK SIDE OF THE PAG . The point value for each question is indicated in parentheses after the l

'

questio . If the intent of a question is unclear, ask questions of the examiner onl . When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap pape !

1 Ensure all information you wish to have evaluated as part of your answer i is on your answer sheet. Scrap paper will be disposed of immediately ,

following the examinatio l

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1 To pass the examination, you must achieve a grade of 80.00% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may ,

be denied or revoke :

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l QUESTION: 001 (1.00)

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A reactor startup is in progress on Unit 2. All IRMs:are on range 2 and *

all SRMs are being withdrawn when the following SRM indications are receive .

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SRM A 110 cps

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SRM B 125 cps

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SRM C 120 cps :

- SRM D 95 cps !

WHICH ONE (1) of the following actions should occur under these conditions?

t a. An SRM "HI/IN0P" rod out block will be generate b. All SRM detector drive motors will deenergiz c. Only SRM D detector drive motor will deenergize.

j d. An SRM detector wrong position rod out block will be generated.

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l QUESTION: 002 (1.00)

WHICH ONE (1) of the following will result in an IRM HI-HI half scram signal being generated?

l a. Mode switch is in STARTUP: IRM A is indicating 100/125 of full l scale on range b. Mode switch is in RUN: IRM B is indicating 108/125 of full scale i

on range 8: APRM B is failed downscal c. Mode switch is in RUN: IRM G is indicating 120/125 of full scale on range 10: APRM E is failed downscale.

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d. Mode switch is in STARTUP: IRM H is indicating 108/125 of full scale on range 6.

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- 0UESTION: 003 (1.00)

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WHICH ONE (1) of the following will mitigate the consequence of a red ,

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drop accident?

l l a. Velocity Limiter !

b. Rod Block Monitor '

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c. Collet Piston d. Uncoupling Rod ,

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l OVESTION: 004 (1.00) 1 A control rod scram accumulator nitrogen pressure indicates O psi :

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WHICH ONE (1) of the following statements explains how the scram response of the control rod will be affected?

l a. The control rod will be hydraulically locked in place since the

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accumulator is valved ou ,

b. The control rod scram time will increase'as reactor pressure increases.

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c. The control rod will not fully scram if the reactor pressure is l

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less than 400 psi d. The control rod will fully scram regardless of reactor pressure.

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REACTOR OPERATOR Page 9

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QUESTION: 005 -(1.00)

A Unit 2 reactor startup is in progress. Reactor power is ap3roximately 1%. The following conditions exist on the RWCU system which 1as just been placed.in service:

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- RWCU Pump 2A 200 gpm flow

- RWCU Pump 28 Secured

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Non-Regen HX outlet temp 120 degrees F

- RWCU F/D 2A flow 200 gpm WHICH ONE (1) of the following actions is required to be taken?

a. Increase the flow from the 2A RWCU pump, b.~ Place the second RWCU pump in servic c.-Increase RBCCW flow to lower the non-regen HX outlet i temperatur l d. Reduce the flow through RWCU F/D 2 l I

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l QUESTION: 006 (1.00)

Unit 2 power is 38% and preparations are underway to startup the 2B reactor recirculation pump. The 2A reactor recirc pump is running at 40%

l rated speed. Plant conditions are as follows:

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Reactor vessel dome pressure 980 psig i

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Loop 'A' temperature 530 degrees F i

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Loop 'B' temperature 500 degrees F

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Bottom head coolant temperature 390 degrees F WHICH ONE (1) of the following describes the limitations, if any.

l imposed on starting the 'B' reactor recirc pump under the given l conditions?

a. The pump may be started immediatel b. The pump should not be started because the bottom head coolant *

l temperature is too low, c. The pump should not be started because the loop differential temperature is too hig d. The pump should not be started because the 2A reactor recirc pump is running too fast.

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QUESTION: 007 (1.00)

A power ascension was in progress on Unit 2 when the 2B reactor recirc pump tripped. Plant conditions are as follows:

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Reactor power 50%

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- Core flow 42%

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WHICH ONE (1) of the following actions should be taken?

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a. Restart the tripped recirc pump and increase core flow, b. Immediately insert a manual reactor scram, c. Reduce the speed of the running reactor recirc pump to lower reactor powe d. Insert control rods to lower reactor powe i l

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REACTOR OPERATOR Page 11

QUESTION: 008 (1.00)

A reactor startup is in progress. The Mode Switch is in STARTUP and

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Group 1 rods are being withdrawn. Control rod 50-31 was being withdrawn when it lost position indication. After stopping rod motion it was determined that the rod was at position 12 and a substitute position was entered into the Rod Worth Minimizer (RWM).

WHICH ONE (1) of the following explains how RWM control is affected by these conditions?

l a. The RWM will function normally and will initiate a rod block if l control rod 50-31 is mispositioned.

I b. The RWM will not enforce any rod blocks associated with the movement of control rod 50-3 c. The RWM will prevent control rod 50-31 from being moved in any

, direction.

d. The RWM will enforce a rod out block only on control rod 50-3 .

I QUESTION: 009 (1.00)

l A reactor startup is in progress on Unit 2. The Rod Worth Minimizer has failed and is procedurally bypassed. The reactor startup is continuin WHICH ONE (1) of the following describes the restrictions placed on the use of continuous notch override when withdrawing controls rods under these conditions?

Continuous notch override . ..

a. cannot be used under these condition b. can only be used if the rod target position is position 4 c. can be used to move any control rod as long as a second licensed l operator is present to verify rod movement.

l d. can only be used if a " black and white" rod pattern has been achieve f- + - 'vy

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QUESTION: 010 (1.00)

. Control rods are being withdrawn to bring the reactor critical. WHICH

ONE (1) of the following would require the URO to stop withdrawing

control rods?

i a. SRM period is 45 seconds.

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b. SRM period is 125 seconds.

j c. SRM count rates are 1000 cps.

i l d. SRtt count rate has doubled from the initial count rate.

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l QUESTION: 011 (1.00)

l Control rods are being withdrawn to perform a reactor heatup on Unit 2

! per 3rocedure GP-2. Normal Plant Startup? Reactor pressure is 85 psig.

If tie heatup continues for two hours at the procedural heatup rate

limit. WHICH ONE (1) o' f the following would be the maximum expected

! value of reactor pressure.

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a. 615 psig j b. 660 psig i c. 710 psig

d. 855 psig i

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QUESTION: 012 (1.00)

Unit 2 is critical with a plant startup in progress after a refueling '

outag WHICH ONE (1) of the following describes when the Reactor -

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Operator can expect IRM response to become significantly more pronounced

during individual rod withdrawal? .

a. After achieving a " black and white" rod patter '

b. Whenever a peripheral control rod is withdraw c. When reactor power is between IRM range 3 and IRM j

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d. When the reactor starts to pressuriz '

OUEST10N: 013 (1.00)

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The shell warming mode of the main turbine has just been selected. WHICH ONE (1) of the following describes the valve alignment during this operation?

a. The turbine control valves are open fully: the intercept valves are closed; the intermediate stop valves are close b. The turbine control valves are partially open: the intercept valves are open fully: the intermediate stop valves are close c. One turbine control valve is open fully: the intercept valves are closed; the intermediate stop valves are fully ope d. The turbine control valves are closed: the intercept valves are partially open: the intermediate stop valves are partially ope ,

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OVESTION: 014 (1.00)

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, During shell warming, turbine first stage pressure increases to 140

psig. WHICH ONE (1) of the following is the expected response to this

! condition?

a. A Group I isolation will occur.

f b. A reactor scram will be initiated.

j c. The main turbine will overspeed.

d. Main condenser vacuum will start to decrease.

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QUESTION: 015 (1.00)

Reactor power is indicating approximately 9% on the APRMs with

] preparation being made to roll the main turbine. WHICH ONE (1) of the

following is used to verify that the APRMs are indicating properly?

a. 3D MONICORE P-1 printout.

J b. Turbine bypass valve position.

! c. Main steam line flow indication.

d. IRM indication.

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c l QUESTION: 016 (1.00)

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Preparations are being made to p' lace the Unit mode switch to "RUN".

Plant conditions are as follows:

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IRM C Failed upscale (NOT Bypassed)

- Highest reading APRM (D) 13%

i - Lowest reading APRM (C) 5%

l - PAM pressure 860 psig

- Main condenser vacuum 22 inches HG

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WHICH ONE (1) of the following would occur if the mode switch was placed in "RUN" under these conditions?

'

a. Half scra b. Full reactor scram onl '

c. Group I isolation onl d. Full reactor scram AND Group I isolatio ;

QUESTION: 017 (1.00)

WHICH ONE (1) of the following explains why the IRMs should be placed on a range which will maintain them onscale when the reactor is operating 1 at full power?

a. To extend the operational life of the IRM detectors.

b. To support the operability of the Rod Block Monitor syste c. To support the operability of the APRM downscale RPS functio d. To prevent inadvertent IRM HI-HI trip signals from being generate I I

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QUESTION: 018 (1.00)

Power ascension is in 3rogress. WHICH ONE (1) of the following describes a condition where the ow Power Set Point (LPSP) has been cleared?

a. One APRM channel in each RPS trip system is indicating greater than 10% powe b. At least one APRM is indicating 15% powe c. Feedwater flow is 17%: Steam flow is 21%.

d. Feedwater flow is 14%: Steam flow is 18%.

QUESTION: 019 (1.00)

Unit 2 is o>erating at 100% power. Reactor Building Ventilation has been placed on t1e SBGT system due to a high radiation condition. WHICH ONE'

(1) of the following is the possible consequences of continued operation in this lineup?

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a. A Group I isolation may occur from high MSL tunnel temperatur b. Group IV isolation may occur from high HPCI room temperature, c. Group II isolation may occur from high RWCU area temperatures d. Group V isolation may occur from high RCIC room temperature i

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QUESTION: 020 (1.00) '

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A reactor shutdown is in progress on Unit 3. Reactor recirc flow is ,

being lowered. WHICH ONE (1) of the following is the maximum rate of '

power reduction r GP-3, " Normal Plant Shutdown"? (Assume no abnormal -

conditions exist . '

Maximum rate of power reduction is ...

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a. 5 MWe/ minut b. 13 MWe/ minut c. 25 MWe/ minut d. 30 MWe/ minut :

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QUESTION: 021 (1.00)

WHICH ONE (1) of the following describes when it is allowable to ) urge the primary containment using the Reactor Building Ventilation Ex1aust System?

1. When one train of the Standby Gas Treatment System is inoperabl b. When reactor power is less than or equal to 15% rated thermal powe c. When the reactor is in a cold shutdown conditio d. When the primary containment is required to be operabl !

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QUESTION: 022 (1.00)

! WHICH ONE (1) of the following describes a condition when the Rod Block Monitor is required to be in operation?

] a. The reactor mode switch is placed in STARTUP.

l b. The Rod Worth Minimizer is inoperable.

j c. Anytime control rods are being moved.

j d. Reactor power is greater than 30%.

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OUESTION: 023 (1.00)

Given the following conditions on Unit 3:

- An ATkS has occurre Power is lost to 3R4-R- The SR0 has directed the URO to initiate SBL The URO positions the keylock switch to the " Pump B Run" positio WHICH ONE (1) of the following describes the status of the SBLC system under these conditions?

a. 3A pump is 0FF: 3B pump is ON: BOTH explosive valves fired: RWCU is isolate '

b. 3A pump is ON: 3B pump is 0FF; ONE explosive valve fired: RWCU is-isolated, c. 3A pump is ON: 3B pump is ON: 33fH explosive valves fired: RWCU is isolate '

d. 3A pump is 0FF: 3B pump is 0FF: NEITHER explosive valve fired:

RWCU is not isolate .

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j OVESTION: 024 (1.00)

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! The HPCI system was in a normal standby alignment when a valid automatic

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initiation signal was received. Shortly after HPCI initiation the PRO notices that the HPCI system has realigned to the alternate suction ,

'

t source. WHICH DNE (1) of the following conditions could have caused the '

HPCI system to realign?

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a. Low torus level.

A b. Low booster pump suction pressur c. Low CST leve I l

d. A group IV isolation signal.

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QUESTION: 025 (1.00)

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I l WHICH DNE (1) of the following describes a condition where the reactor

coolant system pressure safety limit has been exceeded?

a. Recirc loop pump suction pressure is 1260 psig.

. b. Recirc loop pump discharge pressure is 1330 psig.

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{ c. Reactor steam dome pressure is 1350 psig.

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d. Vessel bottom head pressure is 1360 psi ,

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) OUESTION: 026 (1.00)

?

WHICH ONE (1) of the following supplies power to the Backup Scram i Valves?

a. 24 VDC bus power 4 b. 125 VDC bus power c. RPS Bus A and B

d. 120 VAC Instrument Bus

d i

i OUESTION: 027 (1.00)

! WHICH ONE (1)'of the following will result in an APRM Inop trip signal

being generated?

a. A 10% or greater difference between two APRM Flow Units.

{

b. Bypassing two LPRM inputs from a single level.

j c. Placing the APRM Mode Switch to the " POWER FLOW" position.

i d. The LPRM count circuit senses only 16 LPRMs.

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QUESTION: 028 (1.00) ,

The E2 Diesel Generator tripped following a start on high drywell pressure. E2 Diesel Generator operating conditions just prior to the trip were as follows:

- Lube oil pressure: 21 psig

- Engine cooling water outlet temperature: 210 degrees F -

- Jacket coolant supply pressure 19 psig

- Generator load current: 550 amps

- Generator voltage: 4250 volts WHICH ONE (1) of the following signals caused the diesel generator to trip?

a. Low lobe oil pressur b. Low cooling water pressur c. Generator differential overcurren j d. Engine cooling water high temperatur l QUESTION: 029 (1.00)

While at 100% power a trip of a running condensate pump on Unit 3 has occurred. Condensate pump discharge header pressure dropped to 475 psi WHICH ONE (1) of the following actions should occur in response to these conditions?

a. The CRD pump suction valve will clos b. Recirc pumps will run back to 20%.

c. The condensate reject control valve will clos d. An 85% maximum speed signal will be sent to the running feed pump . _ _ .

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QUESTION: 030 (1.00)

Unit 2 is operating at approximately 35% reactor power when the following annunciator alarm is received:

- MOIST SEP HI LEVEL TRIP Assuming the alarm is valid, the main turbine will tri a. immediately and the reactor will scram, b. 'immediately but the reactor will remain on lin c. after a 30 second time delay and the reactor will scra d. after a 30 second time delay; t'h e reactor will remain on'lin QUESTION: 031 (1.00)

Unit 3 is operating at power when the following annunciator alarm is received: -

- A RECIRC PUMP SEAL STAGE 2 HI-LO FLOW Other indications on the 3A recirc pump are as follows:

- No. 1 seal pressure 990 psig and steady

- No. 2 seal pressure 785 psig and rising WHICH ONE (1) of the following is the cause of the alarm?

a. Failure of the first stage sea b. Failure of the second stage sea c. Plugging of the seal internal orific d. A loss of seal purge flo . . . - . - . .- . . - - . _ _ . - . . - - _ - - .- - . . .. . _ . . . . - . . - _ . _

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QUESTION: 032 (1.00)

While operating at 100% power a valid high steam line flow was sensed in the "A" main steam line only. WHICH ONE (1) of the following is the expected MSIV response?

a. All MSIVs will clos b. Only the MSIVs in main steam line 'A' will clos c. Only the inboard MSIV in main steam line 'A' will clos d. Only the inboard MSIVs in all four main steam lines will close.

t QUESTION: 033 (1.00)

A loss of reactor water level control has resulted in a reactor scram on

.

low water level. Reactor water level dropped to - 5 inches before l control was reestablished. WHICH ONE (1) of the following isolation i

valves should have closed?

a. Main steam line drain valve b. Recirc sample valve c. Drywell equipment drain valve d. RHR sample valve L

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QUESTION: 034 (1.00)

The URO has noticed a sudden rise in the A loop drive flow. WHICH ONE (1) of the following indications would be consistent with a jet pump failure?

a. A rise in core thermal powe b. A drop in core plate differential pressur c. A slight rise in recirc pump spee d. An increase in indicated dp on the.affected jet pum QUESTION: 035 (1.00)

Unit 3 is operating at approximately 80% power when the 'A' Narrow Range reactor water level instrument (which was auto selected for control)

fails downscale. Other reactor water level instruments indicate as

'

follows:

-

B' Narrow Range + 22 inches

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'C' Narrow Range + 24 inches

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'A' Wide Range + 15 inches

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B' Wide Range + 16 inches WHICH ONE (1) of the following describes how the Feedwater Control System will control reactor water level under these conditions?

The Feedwater Control System will ..

a. use a default value of + 23 inches to control reactor water leve ,

b. shift to the 'B' Narrow Range instrument to control reactor water leve c. shift to the 'C' Narrow Range instrument to control reactor water level . ,

d. use the average of the wide range instrument to control reactor water leve i l

, _ . . . .. _ . _ . 4 m am1... . . _ _ . . mm. . _a , mm _m. ..c.. _ _ _ m ._ . -. _m..um_____u._ _ _ .. .g_. _._m._.__& . __._ _ _ _ . _ _ . .

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QUESTION: 036 (1.00)

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WHICH ONE (1) of the following describes a condition:that will result in the greatest difference between all of the wide range reactor water level instruments and actual reactor water level?

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a. When core flow is less than 10%.

b. During single loop operations.

l c. When operating at 100% recirc flo d. When the reactor is in Mode !

OUESTION: 037 (1.00)

Unit 2 is operating at 80% power when the following annunciator alarm !

comes in:

- A RECIRC SPEED CONTROL SIGNAL FAIL Investigation reveals that the recirc pump controller output has failed to zero. WHICH ONE (1) of the following describes how the affected recirc pump should respond to this failure?

a. The associated scoop tube should lock up and the recirc pump speed should remain the sam b. The affected recirc pump should run back to 45% spee c. The affected recirc drive motor breaker should trip ope d. The affected recirc pump should run back to minimum spee .

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QUESTION: 038 (1.00)

The following conditions exist on the Control Rod Drive Hydraulic System:

- Drive water pressure 260 psid

- Drive water flow 0 gpm

- Charging water pressure 1500 psig

- Cooling water dp Upscale high > 60 psid

-

Cooling water flow 64 gpm

- CRD system flow 65 gpm WHICH ONE (1) of the following describes the aossible consequences of operation under these conditions during a CRDi system startup?

a. Overheating of the CRD mechanism b. CRD pump trip on low flo c. Inability to move control rod d. Control rod drifts occurrin QUESTION: 039 (1.00)

Drywell cooling has been maximized per T-223. "Drywell Cooler Fan Bypass" due to high drywell pressure. Shortly after starting the fans in fast speed a LOCA causes reactor water level to drop below -160 inche WHICH ONE (1) of the following describes how operation of the drywell !

cooling fans is affected by this condition?

a. The fans will automatically shift to slow spee b. The fans will remain running in fast spee c. The fans will stop and cannot be restarte d. The fans will stop and but can be restarted immediately by using the trip bypass switc .

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QUESTION: 040 (1.00)

Unit 2 is operating at 100% power. During the performance of ST-0-020- '

560-2. " Reactor Coolant Leakage Test". the reactor operator determines that the primary containment unidentified leakage pump out rate changed from 1.5 gpm to 4 gpm over the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Total leakage to the containment is 8 gpm. WHICH ONE (1) of the following describes the status of compliance with the Reactor Coolant System (RCS) Operational Leakage LCO?

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The RCS Operational Leakage LCO is ..

a. Het and plant operation may continu b. Not met due to excessive total unidentified leakage.

c. Not met due to an excessive increase in unidentified leakag d. Not met due to excessive. total leakage from the containmen QUESTION: 041 (1.00)

While operating at 50% power the following annunciator alarms:

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SAFETY RELIEF VALVE OPEN WHICH ONE (1) of the following indications confirms that an SRV is open?

a. Turbine control valves open slightl b. Steam flow - Feed flow mismatc c. Turbine bypass valves openin d. An increase in main turbine exhaust pressur l

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! OUESTION: 042 (1.00)

Due to a transient condition. T-227. " Defeating RWCU Isolation

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Interlocks" has been completed and RWCU is being used to control reactor i pressure. Approximately fifteen minutes later the following annunciator j alarms:

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l - CLEANUP RECIRC PUMP SUCTION LINE BREAK i WHICH ONE (1) of the following describes how the RWCU will respond?

a.'The RWCU will continue to operate in the recirc mode.

J b. Only the INBD inlet isolation valve (M0-2-12-015) will close.

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i c. Only the OUTBD inlet isolation valve (M0-2-12-018) will clos d. A full RWCU isolation will occu QUESTION: 043 (1.00) ,

WHICH ONE (1) of the following describes the consequences of a loss of 125 VDC panel 2002111 on the ADS System?

a. Channel 'A' of ADS is deenergized: ADS can initiate through l Channel 'B'. !

b. Both channels of ADS are deenergized: ADS will not operate under any conditio c. Automatic initiation of ADS is defeated; manual ADS initiation is still possibl ;

d. Channel 'A' of ADS will automatically shift to its alternate power supply: ADS will function as require __

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OVESTION: 044 (1.00)

At 100% power. RCIC is operating in the CST to ' CST mode with its flow controller in "AUT0". RCIC Turbine speed is 2200 RPM. WHICH.0NE (1) of f the following will result in raising RCIC Turbine speed?

<

a. Opening M0-2-13-27. "RCIC Minimum Flow Valve".

.

b. Throttling open M0-2-13-030. "RCIC Full Flow Test Valve".

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c. Placing the RCIC Flow Controller in " MANUAL" then throttling open M0-2-13-030. "RCIC Full Flow Test Valve".

d. Throttling open M0-2-13-021. "RCIC Feedline Injection".

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QUESTION: 045 (1.00)

The Standby Gas Treatment System (SBGT) has automatically started on Unit 3. System operating parameters are as follows:

'

- SBGT system total flow 5.000 scfm

- Secondary containment dp - 0.1 inches WG WHICH ONE (1) of the following explains how the SBGT System is operating?

a. The system is operating normall b. SBGT system total flow is too lo c. SBGT system total flow is too hig d. Secondary containment dp is not adequat !

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i QUESTION: 046 (1.00) i Vacuum is lowering on Unit 3. WHICH ONE (1) of the following is an indication of air in-leakage into the main condenser?

, a. A decrease in "Off-Gas Flow" indication.

i l b. A decrease in air ejector after condenser outlet pressure.

I c. An increase in Guard Bed differential pressure.

d. Automatic closure of the recombiner recycle valv j i

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QUESTION: 047 (1.00)

The following indications and annunciator alarms have been received on

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Unit 2:

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CONTROL ROOM RAD MONITOR DIV I INITIATED i CONTROL ROOM RAD' MONITOR DIV II INITIATED

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MCR Fresh Air Supply Flow Recorder (FR-0765) indicates 3150 scf MCR Radiation Monitors (RI-0760A/B) Red High Lights are NOT lit.

MCR Radiation Monitors (RI-0760C/D) Red High Lights are li ! WHICH ONE (1) of the following is the expected status of the Main

Control Room Ventilation System? (Assume 2 minutes have passed).

i j a. The Control Room Ventilation dampers and fans have shifted to

! the Purge Mod b. The Control Room Ventilation System ~has shutdown and isolated.

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c. The selected Control Room Emergericy Vent Fan (OAV-30 or OBV-30)

should be runnin '

i l d. The Control Room Ventilation dampers and fans are aligned for

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normal operation.

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! l l QUESTION: 048 (1.00)

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A LOCA has occurred on Unit 2. The El and E3 emergency diesel generators have failed to start. WHICH ONE (1) of the following describes the status of ECCS area cooling? i a. All ECCS room coolers are being cooled by the Service Water l Syste !

b. All ECCS room coolers are being cooled by the Emergency Service Water Syste c. All ECCS room coolers are being supplied by the Emergency Cooling Water Syste d. The 'A' RHR and 'A' Core Spray room coolers do not have cooling; i all other ECCS room coolers are being supplied by the Service Water Syste QUESTION: 049 (1.00)

Unit 2 is in Mode 4. WHICH ONE (1) of the following conditions would require the Secondary Containment to be operable?

a. The Reactor Building Ventilation System is removed from servic b. The SRM's are being removed from the core using the control room control c. A CRD mechanism is being removed from the vesse d. Both SBGT System trains are taken out of service for maintenanc l

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REACTOR OPERATOR Page 32 i QUESTION: 050 (1.00)

Unit 3 is operating at steady state 50% power when an EHC malfunction causes a turbine bypass valve to open. WHICH ONE (1) of the following describes how the plant will respond to this event after plant parameters stabilize? l l

a. Reactor power will increase: Feedwater heating will increas b. Reactor power will decrease: Feedwater heating will decreas c. Reactor power will decrease: Feedwater heating will increas d. Reactor power will increase: Feedwater heating will decreas OUESTION: 051 (1.00)

Unit 2 was operating at 100% power when a LOCA occurred. With the LOCA signal present. the SR0 has directed you to open valve M0-38(A) (Torus Spray Valve) to spray the torus. WHICH ONE (1) of the following i conditions must be met to allow this valve to be opened?

a. CTMT SPRAY VLV CONT switch taken to " MANUAL": Drywell pressure greater than 1 psig: RPV water level above 2/3 core heigh b. CTMT SPRAY VLV CONT switch left in "0FF": Drywell pressure greater than 2 psig: RPV water level above 2/3 core heigh c. CTMT SPRAY VLV CONT switch taken to " MANUAL": Drywell pressure greater than 1 psig: RPV water level below 2/3 core heigh d. CTMT SPRAY VLV CONT switch taken to " MANUAL": Drywell pressure greater than 2 psig: RPV water level above 2/3 core height: MO- ,

39(A) (Supply to Torus) fully close '

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l OUESTION: 052 (1.00)

Unit 2 has scrammed on high drywell pressure. Shortly after the scram !

the following annunciator alarmed

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E12 BUS UNDERVOLTAGE-l Voltage has decreased to 88% of normal and is steady. WHICH ONE (1) of the.following describes the response to this condition? !

a. The E-12 bus will transfer to its alternate source if this condition exists for greater than 10 second j b. The E-12 bus will transfer to its alternate source immediatel ,

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c. The E-12 bus will remain energized from its normal source since !

a LOCA signal is presen d. The E-12 bus will be completely deenergized if this condition j exists for greater than 10 second j i

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QUESTION: 053 (1 00)

Unit 2 is operating at 25% power when the 'A' MSL Radiation Monitor !

fails upscale. Shortly after the failure a loss of 'B' RPS occurs. WHICH !

ONE (1) of the following describes the expected plant response? !

t a. All MSIVs will close: the reactor will scra i b. All MSIVs will remain open: the reactor will scra c. All MSIVs will remain open; a half scram will occu d. Only one MSIV will close; a half scram will occu {

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QUESTION: 054 (1.00)

Unit 3 was operating at 100% when a loss of off-site power occurre Shortly after the loss of power a break in the fire main reduces pressure to 120 psig. WHICH ONE (1) of the following fire pumps will be operating in this condition? (Assume no operator actions taken.)

a. Only the diesel driven pum b. High pressure lube water pump AND diesel driven pum c. High pressure lube water pump AND motor driven pum d. Motor driven pump and diesel driven pum I

QUESTION: 055 (1.00)

l A Traversing In-Core Probe (TIP) trace is in progress when an instrument !

malfunction causes a spurious Group II D Isolation. WHICH ONE (1) of the following is the expected automatic response of the TIP system?

a. The TIP Probe will withdraw from the core and the ball valve will.clos b. The TIP Probe will withdraw from the core and the shear valve will fir c. The TIP Probe will remain in place and the shear valve will fi r d. The TIP Probe will withdraw from the core and both the ball valve will close and the shear valve will fir _ _

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QUESTION: 056 (1.00)

Core reload is in progress at CCTAS step 1150 w'ith six (6) new fuel bundles remaining to be loaded into the core into the "B" quadran After loading two of the six new fuel bundles, the CCTAS step 1152 reading and the earlier CCTAS step 1150 SRM readings were as follows:

SRM A SRM B SRM C SRM D CCTAS 1150 75 100 75 75 CCTAS 1152 90 200 90 90 WHICH ONE (1) of the following states the expected results of loading the remaining bundles? (Assume all 6 bundles have equal reactivity worth.)

a. One more bundle will cause a local criticalit b. Two more bundles will cause a local criticalit c. SRM "B" will indicate 400 cps when the core is fully loade d. SRM "B" will indicate 600 cps when the core is fully loade l

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QUESTION: 057 (1.00) ,

I WHICH one of the following correctly describe interlocks which will prevent an unexpected RPV level decrease while in shutdown cooling? l a. M0-10-17. "S/D Cooling Outboard" and M0-10-18. "S/D Cooling i Inboard" close on low RPV level of 1-inch and high drywell i pressure of +2 psi ;

M0-10-15A(B)(C)(D). "A(B)(C)(D) S/D Cooling" close on low RPV j level of 1 inch and high drywell pressure of +2 psi l b. MO-10-17. "S/D Cooling Outboard" and MO-10-18. "S/D Cooling J Inboard" close on low RPV level of 1 inch and high drywell I pressure of +2 psi '

MO-10-15A(B)(C)(D). "A(B)(C)(D) S/D Cooling" have no auto close signal c. M0-10-17. "S/D Cooling Outboard" and MO-10-18. "S/D Cooling Inboard" close on~10w RPV level of 1 inch but not on high ,

drywell pressure of +2 psi I MO-10-15A(B)(C)(D). "A(B)(C)(D) S/D Cooling" close on low RPV l level of 1 inch but not on high drywell pressure of +2 psi j d. M0-10-17. "S/D Cooling Outboard" and M0-10-18. "S/D Cooling Inboard" close on low RPV level of 1 inch but not on high

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drywell pressure of +2 psi M0-10-15A(B)(C)(D). "A(B)(C)(D) S/D Cooling" have no auto close signal .

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QUESTION: 058 (1.00) ,

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Given the following core information at Peach Bottom: Unit #2:

Quadrant A SRM A reads 8 cps: 50% of the fuel bundles are

, remove . l

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Quadrant B A special movable detector connected to the normal SRM B circuit reads 4 cps: 60% of the fuel bundles are removed.

! Quadrant C SRM C reads 10 cps: 50% of the fuel bundles are

removed.

i Quadrant D SRM D reads 2 cps; all fuel has been removed from the quadrant except 2 bundles adjacent the SRM.

I Assuming the Signal to Noise Ratio for all detectors is 2.1. WHICH ONE

! (1) of the. following is correct regarding the operability of the SRM instruments?

a. NO SRMs are operabl ,

l b. ALL SRMs are operabl c. SRMs A. B and C are operable. SRM D is inoperabl d. SRMs A. C and D are operable. SRM B is inoperable.

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! QUESTION: 059 (1.00)

i WHICH ONE (1) of the following systems is used to provide cooling to the l Fuel Pool Cooling Heat Exchangers in the event of a loss of Service i Water?

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a. RHR System l b. HPSW System

, c. ESW System d. RBCCW System i

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l REACTOR OPERATOR Page 38 QUESTION: 060 (1.00)

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A complete loss of 125 VDC to the EHC cabinet has occurred. WHICH ONE (1) of the following describes how plant operation is affected?

a. The main turbine will tri b. The main turbine will remain on line: the turbine can only be tripped at the front standar c. The main turbine will remain on line: the EHC cabinet will

automatically transfer to its alternate power sourc d. The main turbine will remain on line: the turbine can only be tria)ed at the front standard or by using the manual trip pus uutto :

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QUESTION: 061 (1.00)

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Unit 2 is at 90% power with the Drywell Chilled Water System (DWCW)

,

aligned as follows:

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'A' and 'B' Drywell Chilled Water Pumps ' running: 'C' in standb 'A' and 'B' Drywell Chillers running: 'C' in standb The No. 2 Auxiliary Bus becomes deenergized. WHICH ONE (1) of the following explains how the DWCW System will be affected? (Assume no operator actions are taken).

a. RBCCW will automatically align to the DWCW System and supply all RBCCW and DWCW System load j b. RBCCW will automatically align to the DWCW System: non-essential i RBCCW loads will isolat c. Both the 'A' and 'B' Drywell Chilled Water Pumps will trip: the

'C' Drywell Chi.lled Water Pump will auto star d. The 'A' Drywell Chilled Water Pump will remain running: the 'B'

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Drywell Chilled Water Pump will trip: the 'C' Drywell Chilled Water Pump will auto star ,

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QUESTION: 062 (1.00)

L A Unit 3 startup is in progress with reactor power at 1% and all MSIVs

are open. WHICH ONE (1) of the following actions would occur if a l

complete loss of both divisions of the 24 VDC power system? '

l a. Both recirc pumps would tri b. All EDGs will auto star c. The HSIVs would clos d. The reactor would scra QUESTION: 063 (1.00)

WHICH ONE (1) of the following explains how the CRD System will respond to a loss of instrument air?

>

a. The scram discharge volume vent and drain valves will fail ope b. The CRD drive water pressure control valve will fail close c. The CRD flow cor: trol valve will fail close d. The CRD suction valve from the CST will fail ope I

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OUESTION: 064 (1.00)

, Unit 2 conditions are as follows:

- Reactor is scrammed, all rods are in.

- RPV water level is - 45 inches and lowerin RPV pressure is 800 psig and lowering.

. - Drywell pressure is 4 psig and rising.

- Torus spray is in service using the 'A' RHR pump per T-203.

j " Initiation of Torus Sprays Using RHR"

!

l WHICH Of4E (1) of the following describes how the torus spray lineup will l

be affected if these conditions continue?

l The torus spray valves will ...

a. automatically close when reactor pressure drops below 450 psi b. automatically close when reactor water level decreases to - 160

inche c. automatically close when reactor water level decreases to below l 2/3 core height.

d. remain open until closed by the operato '

i l l

l; QUESTION: 065 (1.00) l 1 1 j The following annunciator has alarmed on Unit 2:

! - INVERTER TROUBLE I

WHICH ONE (1) of the following indications would confirm that panel

20Y050 has deenergized?

j a. Recirc pumps run back to 45% spee '

] b. 'A' and 'C' Narrow Range level instrument fail upscal c. Reactor Feed Pumps run back to 85% spee d. Control rod position indication is los . - _ _ - - - - . - . . - . . _ . . - - _ . . . - - - ._ . -. -

O REACTOR OPERATOR

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L QUESTION: 066 (1.00)

A loss of all shutdown cooling has occurred in Mode 4. A010.12- " Alternate Shutdown Cooling" has been entered. WHICH ONE (1) of the ,

following systems is used to remove the heat being generated in the reactor?

a. Fuel Pool Cooling heat exchanger b. RBCCW c. Emergency Service Water d. High Pressure Service Water

QUESTION: 067 (1.00) i A loss of a reactor feed water pump on Unit 2 has occurred. Reactor water level is +20 inches and lowering slowly. The URO is reducing recirc flow per GP-9-2. " Fast Reactor Power Reduction" when the following annunciator alarms: '

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- APRM HIGH '

WHICH ONE (1) of the following actions should be taken?

a. Stop reducing recirc flow and manually insert control rods per GP-9-2 Appendix b. Stop reducing recirc flow and manually scram the reacto c. Continue reducing recirc flow until minimum speed is reache d. Trip the recirc pumps then manually scram the reacto l

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I QUESTION: 068 (1.00)

A loss of shutdown cooling has occurred. WHICH ONE (1) of the following is the MINIMUM reactor water level that will promote natural l circulation?

a. + 30 inches as read on the Narrow Range instrument ,

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b. + 50 inches as read on LI'-86 (Refuel Range).

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c. + 60 inches as read in LI-85 (Wide Range). '

d. + 50 inches as read on the Narrow Range instruments OUESTION: 069 (1.00)

Unit 3 is operating at approximately 90% power when the PRO recognizes that main condenser vacuum is decreasing slowly. WHICH ONE (1) of the following conditions could be the cause for the lowering vacuum?

a. SJAE supply' pressure is 95 psi b. Seal steam header pressure is 1.0 psi c. Auxiliary Steam System pressure is 200 psi d. Circ water pump bay level is 110 fee .

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QUESTION: 070 (1.00)

! During a GP-3 shutdown following a transfer of house: loads, a loss of )

level control on Unit 2 has resulted in reactor water level reaching +

60 inches and is risim very slowly. WHICH ONE (1) of the following -

automatic actions should have occurred? - l

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a. Both recirc pumps trip.

i b. All condensate pumps tri !

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c. A HPCI turbine tri r d. RCIC steam line isolation valves (M0-15.16) clos !

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I QUESTION: 071 (1.00)

Unit 2 is operating at 60% power with a HPCI surveillance in progres l SPOTMOS temperature indicates 95 degrees The following annunciator alarm is received: l

1

- SAFETY RELIEF VALVE OPEN WHICH ONE (1) of the following describes when the crew is required to initiate a manual reactor scram?

a. When Torus temperature reaches 105 degrees b. When Torus temperature exceeds 110 degrees '

c. Immediately after verifying that an SRV is ope d. After exhausting all attempts to close the SR . _ -

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OUESTION: 072 (1.00)

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An EHC failure has resulted in reactor pressure increasing to 1090 psig but the UR0 reactor initiated manually did NOTthe scra In res>onse to the failure to scram the Alternate lod Insertion (ARI) System and all control rods inserte t WHICH ONE (1) of the following describes how the reactor recirc pumps

} will respond to a manual initiation of ARI? ,

i a. The recirc pumps will tri j b. The recirc pumps will run back to minimum speed.

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c. The recirculation pumps will remain running at their present

,

speed with the scoop tubes locked u d. The recirculation pumps will remain running at their present speed with the scoop tubes operationa QUESTION: 073 (1.00)

T-101. "RPV Control" has been entered on high reactor pressure. The !

MSIVs are closed and SRVs are cycling. The SR0 has directed the PRO to l

control reactor pressure between 950 and 1050 psig using SRVs. WHICH ONE :

(1) of the following is the reason for stabilizing reactor pressure l below 1050 psig? '

a. To allow the scram logic to be reset, b. To minimize the inventory lost from the reacto c. To maintain operability of reactor water level instruments, d. To maintain cooldown rate below 100 degrees F per hou .

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QUESTION: 074 (1.00)

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i l An automatic reactor scram has occurred. All control rods were full out prior to the scram. Two of the control rods did NOT fully insert. One control rod.is at position 04 and the other control rod is at msition 08. All scram solenoids have deenergized and all scram valves lave

opene WHICH ONE (1) of the following methods of control rod insertion would be the most likely to be effective for inserting these rods?

a. Use the individual scram test switche !

b. Vent the scram air heade c. Raise CRD drive water pressure and manually drive in the control ,

rod ;

i d. Initiate the ARI Syste '

l QUESTION: 075 (1. 00'; l WHICH ONE (1) of the following conditions would require entry into T-104. " Radioactivity Release"?

a. When any off-site radiological release is occurrin b. Whenever a main steam line break outside the containment has been confirme c. When the off-site gaseous radiological release rate exceeds the Unusual Event Emergency Action Leve d. Whenever a primary system is discharging into the secondary containment.

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QUESTION: 076 (1.00)

T-104. " Radioactivity Release", has been entered. The Turbine Building Ventilation System has been restarted? WHICH ONE (1) of the following is reason for restarting the ventilation system?

a. To maintain negative pressure in the Turbine Buildin b. To prevent an unmonitored radiation release to the environmen c.,To provide a filtered release path to the environmen d. To assure maximum safe temperature limits are not reache QUESTION: 077 (1.00)

WHICH ONE (1) of the following plant conditions requires entry into T-103. " Secondary Containment Control"?

a. Excessive input causes the Reactor Building Floor Drain Sump HI-HI level alarm to come i b. Excessive input causes the Turbine Building Floor Drain Sump HI-HI level alarm to come i c. Reactor Building differential pressure is 0 inches of wate d. Reactor Building Refuel Floor differential pressure is .25 inches of wate .

REACTOR OPERATOR Page 47

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QUESTION: 078 (1.00)

A loss of drywell cooling occurs. Drywell Pressure increases to 3.2 psig and Drywell Bulk Average Temperature increases to 145 degrees WHICH ONE (1) of the following identifies the Trip Procedures that should be entered, a. T-101, "RPV Control" onl b. T-102. " Primary Containment Control" onl c. T-101, "RPV Control". AND T-102 Primary Containment Contro d. T-102. " Primary Containment Control". AND T-103. " Secondary Containment Control".

QUESTION: 079 (1.00)

WHICH ONE (1) of the following is the lowest RPV water level at which the core can be adequately cooled by Steam Cooling with N0 injection?

a. -160 inches b. -200 inches c. -210 inches l

d. -240 inches l

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QUESTION: 080 (1.00)

A failure to scram has occurred and reactor power is approximately 35%.

WHICH ONE (1) of the following would be the consequence of tripping the reactor recirc pumps above minimum speed?

a. A turbine trip on high reactor water level, b. A loss of turbine bypass valve c. Reactor pressure exceeding the SRV lifting setpoint d. A Group II isolation may occu QUESTION: 081 (1.00)

Unit 2 power is 70% when a control rod at position 24 starts to drift out. WHICH ONE (1) of the following actions should be taken?

i a. Momentarily turn off RMCS powe b. Give the control rod a momentary insert signa c. Drive the control rod in to its original positio d. Fully insert the control rod using EMERG I .

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REACTOR OPERATOR Page 49

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QUESTION: 082 -(1.00)

An ATWS has occurred and SRVs are being used to control 3ressure. WHICH ONE (1) of the following describes conditions where the deat Capacity Temperature Limit is being approached?

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a. Reactor pressure decreasing: Torus temperature steady: Drywell I temperature increasin b. Torus pressure decreasing: Torus temperature decreasing: Torus !

level increasin c. Reactor pressure increasing: Torus temperature increasing: Torus ;

level stead d. Reactor, pressure steady: Drywell temperature steady: Torus level increasin l

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OUESTION: 083 (1.00) )

An ATWS has occurred on Unit 3. The URO is implementing T-220- " Driving Control Rods During Failure To Scram". WHICH ONE (1) of the following explains why the Charging Water Hdr Blk Valve (HV-3-56) is closed?

To increase the ...

a. rate at which the CRD Over-piston area vent b. rate at which the scram discharge volume drain c. rate at which the CRD accumulators repressuriz d. CRD flow to the drive water header for driving in control rods.

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REACTOR OPERATOR Page 50

. i OUESTION: 084 (1.00)

A transient on Unit 2 has resulted in a rupture of the torus causing torus level to decrease. Plant conditions are as follows:

- Reactor power 0% (all rods inserted)

- Reactor pressure 800 psig and decreasing

- Torus level 10.5 feet, dropping slowly

- Torus pressure 3.5 psig

- Torus temperature 90 degrees F WHICH ONE (1) of the following statements describes the effect that this I torus water level will have on plant operations? l a. RHR and Core Spray pumps cannot be run due to cavitation concerns on low suction pressur b. HPCI can only be operated if it is aligned to the CS c. The SRVs may be damaged during operation due to water backing up into the discharge pipin d. The reactor should undergo an emergency blowdown since the l drywell downcomers may become uncovere l OUESTION: 085 (1.00)

WHICH ONE (1) of the following describes the consequences of operating with torus level above the SRV Tailpipe Limit Curve?

a. The SRV tailpipe could fail during actuatio b. The SRV would not open if require c. The SRV tailpipe discharge is uncovere d. The SRV lifting setpoint will be higher than~ norma .

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REACTOR OPERATOR Page 51

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OVESTION: 086 (1.00)

Evacuation of the main control room is required due to toxic gas. WHICH ONE (1) of the following describes operator actions that should be taken (if possible) prior to evacuating the control room?

a. Trip all recirc pump l

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b. Initiate RCI c. Open all turbine bypass valve !

d. Place torus cooling in servic !

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QUESTION: 087 (1.00)

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Torus sprays have been placed in service due to high drywell temperature. WHICH ONE (1) of the following explains why torus sprays have to be secured if torus pressure drops below 2 psig?

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a. To prevent exceeding the Torus-to-Drywell vacuum breaker !

capacit j b. To prevent operation of the Reactor Building-to-Torus vacuum breaker I c. To prevent exceeding the drywell downcomer differential pressure limit d. To prevent runout of the RHR pum . _ - - . = . . - _ _ . _ _ - - . - . - . _ - . _ - - - - . - - ._- .

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QUESTION: 088 (1.00)

WHICH ONE (1) of the following describes how the lock and chain should

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be left when a manually locked valve is unlocked?

! a. The chain is wrapped around the pipe next to the valve with the lock being in the locked condition.

! b. The chain is looped through the valve handwheel so as not to j obstruct valve movement with the lock in the locked condition.

c. The chain is wrapped around the valve yoke with the lock in the i unlocked condition.

} d. The chain is looped through the valve yoke with the lock in the

unlocked conditio l

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QUESTION: 089 (1.00)

WHICH ONE (1) of the following conditions will permit a procedure change to be processed via a Temporary Change (TC)?

,

The procedure change ... :

a. will eliminate 0V hold points, b. corrects a procedure error and time constraints prevent processing a procedure revisio c. will change the scope of the procedure and time constraints prevent delay of the performance of the procedur d. eliminates the acceptance criteria of a surveillance test but the affected system is still operable per Technical Specification .

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REACTOR OPERATOR Page 53

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QUESTION: 090 (1.00)

The Unit 2 R0 is collecting data for a required surveillance. One of the " black box" steps cannot be completed because the instrument associated with the data collection is inoperable. All other data taken was normal. WHICH ONE (1) of the following describes the action (s) that should be taken?

a. Record the as found data; declare the system operable; initiate maintenance on the inoperable instrumen b. Request engineering to make a operability determination on the syste c. Notify shift management and sign off the surveillance as unsatisfactor d. Record the instrument data as Out-of-Service; declare the system operable: initiate maintenance on the inoperable instrument.

QUESTION: 091 (1.00)

WHICH ONE (1) of the following is the lowest level of authority that may authorize radiation exposures in excess of 10CFR20 limits during an event requiring the activation of the TSC?

a. Emergency Response Manager b. Plant Manager c. Dose Assessment Advisor d. Emergency Director i

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Page 54'

l QUESTION: 092 (1.00) l WHICH ONE (1) of the following conditions must be met to allow the URO to be temporarily relieved of his duties while both units are operating at full power? l

a. There must be a minimum of two licensed operators (R0s) in the i Main Control Roo l b. The relieving RO must be briefed by the Shift Manage l c. The temporary relief period is expected to be less than one hou i d. If the relieving R0 was not at the Shift Turnover Meeting, the temporary relief must be documented in the Narrative Lo QUESTION: 093 (1.00)

An individual has just come on shift and has a current exposure of 1500 mrem TEDE. WHICH ONE (1) of the following is the maximum additional TEDE dose he/she is allowed to receive without obtaining a dose extension?

a. 1500 mrem TEDE b. 2500 mrem TEDE c. 3000 mrem TEDE d. 3500 mrem TEDE

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QUESTION: 094 (1.00) i In the Main Control Room, during full power operations, an annunciator i continues to alarm due to corrective maintenance associated with a corresponding system. The maintenance is scheduled to be completed at the end of the next shift. WHICH ONE (1) of the following actions could be taken? '

3. Remove the annunciator can and reinsert the can in an inverted ,

position as authorized by the Engineering Duty Manage '

b. Place the annunciator mode switch to manual as authorized by ,

Shift Management and initiate an ES '

c. For Process Computer associated alarms obtain the System

' Engineers permission to delete the alarm during the maintenance perio d. With the Shift Supervisor permission place a red triangle on the alarm window and make the appropriate EDL entr QUESTION: 095 (1.00)

WHICH ONE (1) of the following describes the indication that will satisfy Independent Verification (IV) that an excess flow check valve is in the normal position?

a. Red light is o .

b. White light is o c. Green' light is o d No lights are o .

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REACTOR OPERATOR Page 56

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QUESTION: 096 (1.00)

During a plant startup, a licensed control room o)erator assisting with the startup in the Main Control Room has worked tie following hours ,

excluding turnover time: l

Friday 1600 to 0400 l Saturday 1200 to 2400 1 Sunday 1000 to 1600 Monday 0700 to 1500 Tuesday 0900 to 2300 Wednesday 0800 to 1700 WHICH ONE (1) of the following statements below identifies the l violations of the overtime guidelines?

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a. The operator worked more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in 4 l l

b. The operator worked more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in 2 j c. The operator worked more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in 2 l

d. The operator worked more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in 4 l

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QUESTION: 097 (1.00) l l

Technical Specifications sets limits on chlorides in the reactor coolant system. WHICH ONE (1) of the following describes the basis for this ,

limit?

a. Chlorides increase the formation of insoluble metallic corrosion product b. Chlorides increase galvanic corrosion at dissimilar metal junction c. Chlorides catalyze the oxidation of carbon stee d. Chlorides contribute to stress cracking of the stainless stee .

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REACTOR OPERATOR Page 57

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QUESTION: 098 (1.00)

WHICH ONE (1) of the following describes the function of Check-Off Lists (COLs)?

a. COLs must be performed to ensure system readiness prior to operatio b. COLs are performed to align a system so that a specific task can be performe c. COLs are used to verify the adequacy of a system Clearanc d. COLs are used to verify a system is operable per Technical Speci fication i

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l OUESTION: 099 (1.00) l WHICH ONE (1) of the following describes the main control room instrumentation that is backed by DC power?

a. Red tags with white lettering, b. Yellow tags with black lettering, c. Black tags with yellow letterin d. White tags with red letterin I l

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e REACTOR OPERATOR ,Page 58 l

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! QUESTION: 100 (1.00) ;

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WHICH ONE (1) of the following correctly describes the storage of '

transient combustible materials? Transient combustible materials:

d I i a. cannot be stored in the control roo ;

i b. cannot be stored in the protected are I c. can be stored in the plant areas in closed metal container '

! d. can be stored in stairwell enclosures in closed metal e containers.

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(********** END OF EXAMINATION **********)

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SRM Instrumentation -

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3.3. .3 INSTRUMENTATION 3.3. Source Range Monitor (SRM) Instrumentation LCO 3.3. The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABL APPLICABILITY: According to Table 3.3.1.2- ACTIONS

'

CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more required Restore required SRMs 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months i SRMs inoperable in to OPERABLE statu MODE 2 with

intermediate range i

monitors (IRMs) on

Range 2 or below.

e B. Three required SRMs Suspend control rod immediately inoperable in MODE 2 withdrawal .

with IRMs on Range 2 or belo Required Action and Be in MODE hours associated Completion Time of Condition A or B not me (continued)

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PBAPS UNIT 2 3.3-10 Amendment No. 210

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$ SRM Instrumentation i

3.3. ,

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ACTIONS (continued)

I CONDITION REQUIRED ACTION COMPLETION TIME i . .

s One or more required Fully insert all I hour 1 SRMs inoperable in insertable control f MODE 3 or rod *

8 5 I Place reactor mode I hour

"

switch in the shutdown positio . One or more required Suspend CORE Immediately SRMs . inoperable in ALTERATIONS except MODE for control rod insertio tiQ Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblie _

.

SURVEILLANCE REQUIREMENTS

__.... ...___ ..____..------------NOTE--------------------------------------

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified condition ...____........____...........__ ___.......___............ ______........ __ SURVEILLANCE FREQUENCY SR 3.3.1. Perform CHANNEL CHEC hours (continued)

PBAPS UNIT 2 3.3-11 Amendment No. 210

. . . . . . _ __- . -- - _ _ . .

.

SRM Instrumentatien

- 3.3. *

I

!

SURVEILLANCE REQUIREMENTS (continued)

-

SURVEILLANCE FREQUENCY

.

,

SR 3.3.1. NOTES------------------ j

! Only required to be met during CORE :

ALTERATIONS.

! One SRM may be used to satisfy more

.... .. $.$.. $..$..$ ?............

,

! Verify an OPERABLE SRM detector is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

,

. located in:

i The fueled region;

' The core quadrant where CORE ALTERATIONS are being performed, when the associated SRM is inclus d in the fueled region; and 1 A core quadrant adjacent to where !

CORE ALTERATIONS are being performed, i when the associated SRM is included in the fueled region.

i SR 3.3.1. Perform CHANNEL CHEC hours

i

, (continued)

.

>

)'

PBAPS UNIT 2 3.3-12 Amendment No. 210

.

__. .- _ . _ . . . _ - _ . . _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ - _ _ _ - _ . _ _ _ _ _ . _ . - . _ _

SRM Instrumentation 3.3. .

SURVEILLANCE REQUIREMENTS (continued) :

SURVEILLANCE FREQUENCY l

1 SR 3.3.1. NOTES------------------ ; Not required to be met with less than !

or equal to four fuel assemblies j adjacent to the SRN and no other fuel 1 assemblies in the associated core i quadran l Not required to be met during spiral l unloadin )

l Verify count rate is: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during l CORE l : 3.0 cps; or ALTERATIONS i l Within the limits of AND Figure 3.3.1.2- hours I

i SR 3.3.1. Perfom CHANNEL FUNCTIONAL TEST and 7 days determination of signal to noise rati SR 3.3.1. NOTE------------- -----

Not required to be perfomed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRNs on Range 2 or belo .........................................

Perfom CHANNEL FUNCTIONAL TEST and 31 days detemination of signal to noise rati SR 3.3.1. NOTES------------------ Neutron detectors are exclude . Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after IRNs on Range 2 or belo ........ __........................__....

Perfom CHANNEL CALIBRATIO days

.

PBAPS UNIT 2 3.3-13 Amendment No. 210

-, ,,. - . - --,- .. , ,-

. . . . . - - . - .

.

SRM Instrumentatien

. 3.3. .

Table 3.3.1.2 1 (pope 1 of 1) j

,

source Range Monitor Instrumentation l

l l

APPLICABLE MODES OR OTNER REQUIRED SURVEILLANCE FUNCTION SPECIFIED comITIONs CNANNELs REQUIREMENTS I Source Range monitor 2(*) 3 sa 3.3.1.2.1 l

SR 3.3.1. SR 3.3.1. sa 3.3.1. ,4 2 sR 3.3.1. sR 3.3.1. sa 3.3.1. , SR 3.3.1. (b)(c) sa 3.3.1. sa 3.3.1. SR 3.3.1. sR 3.3.1. sR 3.3.1. (a) With Inns on Range 2 or belo (b) only one SRM channet is reesired to be OPERABLE dJring spiret off tood or retoed when the fueled reglen includes only that SRM detecto (c) Specist movable detectors may be used 6n ptoce of Sans if connected to nonmal SRM circuit l

.

l l

PBAPS UNIT 2 3.3-14 Amendment No. 210 ;

)

i i

l

.

SRM Instrumentation 3.3.1.2 , I 3 ' I I j l l

'

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I .6 ! '

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I !

1 i

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I I I 2.0 \ I I I

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I \ I \ ! I I .9 - - - - -

- - -wce - - -

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2 6 10 14 18 22 26 30 Segnal to Noise Ratio Figure 3.3.1.2-1 (page 1 of 1)

Minimum SRM Count Rate Versus Signal to Noise Ratio PBAPS UNIT 2 3.3-15 Amendment No. 210

1 e

RCS P/T Limits

, 3. .4 REACTOR COOLANT SYSTEM (RCS) l I

j 3.4.9 RCS Pressure and Temperature (P/T) Limits j i l l

LC0 3. RCS pressure, RCS temperature, RCS heatup and cooldown l rates, and the recirculation pump starting temperature i requirements shall be maintained within limits.

!

l APPLICABILITY: At all times.

i ACTIONS

,

'

CONDITION REQUIRED ACTION COMPLETION TIME l

,

! NOTE--------- Restore parameter (s) 30 minutes

] Required Action to within limit shall be completed if i i this Condition is AHQ !

entered.

...................--- Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of the continued operation.

J LC0 not met in MODE 1, 2, or 3.

i

, Required Action and Be in MODE hours i associated Completion

~ Time of Condition A AND not met.

i Be in MODE hours

(continued)

!

.

i i

l PBAPS UNIT 2 3.4-21 Amendment No. 210

, . .

-. - . . -. - - . . . _ _ - ,- _ - . _ _ _ .

.

RCS P/T Limits

. 3. .

ACTIONS (contiw:ed)

CONDITION REQUIRED ACTION COMPLETION TIME NOTE--------- Initiate action to Immediately ;

Required Action restore parameter (s)

'

shall be completed if to within limit '

l this Condition is entere &HD

..._____.............. Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 ;

LCO not met in other operatio or than MODES 1, 2, and ;

i

<

!

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4. NOTE--------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testin ____.. ______.............__......__ ..._ Verify: 30 minutes RCS pressure and RCS temperature are within the applicable limits specified l in Figures 3.4.9-1 and 3.4.9-2; and RCS heatup and cooldown rates are s 100*F in any I hour perio (continued) )

,

.

PBArS UNIT 2 3.4<22 Amendment No. 210

Is RCS P/T Limits

.

3.4.9 a

SURVEILLANCE REQUIREMENTS (continued)

.

SURVEILLANCE FREQUENCY

SR 3.4. Verify RCS pressure and RCS temperature are Once within i within the criticality limits specified in 15 minutes i figure 3.4.9- prior to i

'

control rod withdrawal for j the purpose of achieving

-

criticality

a SR 3.4. NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump star ..__......___........ .....................

Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperature prior to each is s 145' startup of a recirculation pump SR 3.4. NOTE--------------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump star .._____..........................______....

Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is s 50* startup of a recirculation pump

. ,(continued)

PBAPS UNIT 2 3.4-23 Amendment No. 210 l l

. - _ - . . -. .

-. .. . . .

RCS P/T Limits 3. ,

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

l SR 3.4. NOTE--------------------

Only required to be performed when tensioning the reactor vessel head bolting stud ___________________________________________

Verify reactor vessel flange and head 30 minutes flange temperatures are > 70*F.

,

SR 3.4. NOTE--------------------

i Not required to be performed until

30 minutes after RCS temperature s 80*F in

,

MODE 4.

, ___________________________________________

.

Verify reactor vessel flange and head 30 minutes flange temperatures are > 70* SR 3.4. NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 100*F in NODE __________________________________________

Verify reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months flange temperatures are > 70* s_

,

l

PBAPS UNIT 2 3.4-24 Amendment No. 210

.

_ _ _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _

_ _ _ . - - . - - - - -- .- . . __ .

P

, RC5 P/T Linits

- .

-

3.4.9

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VALID 10 32 EFPY

'

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. .

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. . . . - -

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O O 100 200 300 400 500 600

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MNIMUM REACTOR VESSEL METAL TEWERATURE ('F)

Figure 3.4.9-1 (page 1 of 1)

i Temperature / Pressure Limits for Inservice Hydrostatic and Inservice Leakage Tests

.

PBAPS UNIT 2 3.4-25 Amendment NO. 210

. - . . . . _ . . . _ _ . . . _ . _ . . _ _ . _ - . _ . _ _ _ _ _ . . . - . _ _ . . . . . _ _ _ _ _ _ . . _ . _ . . _ . . . _ _ . _ _ .

RCS P/T Limits 3. * :

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FOR 32 EFPY OF OPERATION 80LTUP 70*F ! 32 EFPY BELTUNE CURVE

es LI5s uMmNG THAN

/ DISCONTINUITY CURVE B

a 4 0 100 200 300 400 500 600

~

MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 3.4.9-2 (page 1 of 1)

Temperature / Pressure Limits for Non-Nuclear Heatup and Cooldown Following a Shutdown PBAPS UNIT 2 3.4-26 Amendment No. 210

_ - _ __

. _ . . _ _ . . _ _ _ _ _ . _ . . . - _ . _ . _ _ _ _ _ . . . . . . . . _ . . _ . _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . _.

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RCS P/T Limits

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0 100 200 300 400 500 600

'

MINIMUM HEACTOR VESSEL METAL TEMPERATURE (*F)

Figure 3.4.9-3 (page 1 of 1)

Temperature / Pressure Limits for Criticality PBAPS UNIT 2 3.4-27 Amendment No. 210

. _ . . . - . _ . - . _- . _ _ . -- . _ . - ._ _ _ . _

? 4

1TES) TORU.t LEVEL

% - = = = = =e= =L ===

GETCEE214.5 FT AND 14.9 Fi ,

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BEiwEEN 8 4.5 F i A ND I 4.S F l THfW INITIATE A NORuAL SHUTOOWN U

............................... .........

L, 14-3 I3 L iLow TORUS

/*=============== LEVEL HICH OR t T/t-4 LD*

RESTORE AND WAINTAIN 70RUS LEVEL SETWEEN 14.5 Fi AND 14.3 FT USE SmLT ST3fEMS g st001Ric TO ASSURE ACC

. CONO IRANS (50 14A.I.A-2(31)

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HPSw ( T-231)

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I,c wuItf ESECUTING THE F0tt0w!NC STEPS. TORUS

LE VEL DROPS SELow 12.5 Fi -A ND T-101 HAS N01~'~ *

I ALREADY BEEN ENTERE *

.

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2. E N T E R i-101 AND E *ECUTE I N CONCURRE N T L Y l eliN 1-102 CURVE 1/L-1 3. PERF0Ru A NORWAL RPv OEFRESSURI2ATIDN [ *I I 1 1-101 L hf A f CAPaCI TY LE tf L Limi f * U" l

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REACTOR OPERATOR Page 1

ANSWER 'K E Y

MULTIPLE CH0 ICE 023 a t

001 d 024 c

'

002 c 025 c 003 a 026 b 004 c 027 c 005 d 028 c 006 b 029 d 007 d 030 a

,

008 b 031 a

.

009 b 032 a

010 a 033 c 011 a 034 b 012 a 035 b 013 a 036 c .

!

014 b 037 a l

, i 015 b 038 d 3 016 b 039 b 017 c 040 c 018 c 041 b

-

019 a 042 d 020 b 043 a l

021 c 044 a 022 d 045 d

. - . . _ . . . . . . . . - - -.= - - .- - ..~. - _ .._ -.-_-.- _. ._ ... -. - - -

i i

REACTOR OPERATOR Page 2

. .

ANSWER KEY l l

l l

'

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l f

046 c 069 a I

.

047 d 070 c 048 a 071 c

049 c 072 d i 050 d -

073 a 051 a 074 c l 052 b 075 c 053 a .

076 b l 054 a 077 a 055 a 078 c l

056 b 079 c l

057 b 080 a

!

058 b 081 d 059 d 082 c 060 / d 061 b 084 d 062 d 085 a 063 c 086 d 064 d 087 b i 065 d 088 a 066- d 089 b 067 a 090 c 068 b 091 d

- - . - - . . . _ ._

a

~

REACTOR OPERATOR , Page 3 O

ANSWER KEY 092 c 093 a 094

%h !

095 a

^

096 d !

l 097 d i l

098 b 099 b i 100 c

.

!

-

(********** END OF EXAMINATION **********)

. . .

- . .. .- -. .. . - _ . - - - . -

- . . ..-..~. . _--.-. ~~

!

,

a

!

t o ,

ATTACHMENT 3

SIMULATION FACILITY REPORT l

I Facility Licensee: Peach Bottom Units 2 & 3 I I

Facility Docket Nos: 50-277 & 278

) Operating Tests Administered from: September 16-20,1996 a

f i This form is used only to report simulator observations. These observations do not

, constitute audit or inspection findings and are not, without further verification and review,

, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC

] certification or approval of the simulation facility other than to provide information that

. may be used in future evaluations. No licensee action is required in response to these i observations.

! None i

l b

?

.

i

.

i i

l l

I I

l l

v t e Inspection Report - Peach Bottom - 1996010
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1996 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) Security Operator Exams Emergency Planning Other

IR 05000277/1996010

Text

REGION I

Attachments:

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