Text

NRC Operator Licensing Exam Rept 50-293/97-06 (Including Completed & Graded Tests) for Tests Administered on 970505-09
	ML20149K131
Person / Time
Site:	Pilgrim
Issue date:	07/18/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
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Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

. License No. DPR-35 -

Report No.97-006

Jocket No.- 50-293 .

Licensee: Boston Edison Company

800 Boylston Street

Boston, Massachusetts 02199

Facility:~ Pilgrim Nuclear Power Station

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Exemination Period: May. 5 - 9,1997

Examiners: D. Florek, Seninr Operations Engineer .

C. Sisco, 0,serations Engineer ,

S. Dernh., Examiner in Training l

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S. Willoughby, Contract Examiner

, Approved by: G. Meyer, Chief

Operations and Human Performance Branch ;

Division of Reactor Safety j

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PDR ADOCK 05000293

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EXAMINATION SUMMARY

Examination Report 50-293/97-006 (OL)

Initial examinations were administered to six senior reactor operator (SRO) instant

applicants during the period of May 5 -9,1997, at the Pilgrim Nuclear Power Station.

OPERATIONS

Five of six applicants passed the examination. One SRO instant applicant failed the written

and operating portion of the examination. The five applicants that passed were well

prepared for the examinations. The applicants consistently understood and implemented

the emergency operating procedures well. Some weak areas of understanding were

identified during the written exam and operating test.

Two of the applications were found to be deficient in that the applicants had not performed

the five significant control manipulations on the plant as required by 10 CFR 55.31(a)(5).

The applicants' qualification records did not support performance of five significant control-

manipulations. The root ceuse for this problem appeared to be that the BECO program

guidance inappropriately permitted multiple significant control manipulation credit for a

single, extended power change.

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Details

05.1 Operator initial Examinations

a. Scope

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The examiners administered initial examinations to six instant SRO applicants in

accordance with NUREG-1021, " Examiner Standards," Revision 7.

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b. Observations and Findinas

The results of the initial examinations are summarized below:

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SRO

PASS / Fall

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Written 5/1

Operating 5/1

Overall 5/1

The Boston E-jison Company (BECO) staff reviewed the written examination and

assisted in the validation of the operating examination during the week of

April 21,1996. The BECO staff provided comments on the examination that )

significantly improved the examination. The BECO staff, who were involved with

the examination review, signed security agreements to ensure that the initial

examinations were not compromised.

In a letter, dated May 16,1997 (see Attachment 2), BECO provided six comments

on the written examination. The NRC accepted two of the six comments. As a

result, one question was deleted from the examination and two correct answers

were accepted in one question. The NRC resolution of facility comments is

summarized in Attachment 3.'

The following summarizes the written examination questions that were missed by at

least three applicants, indicating a weakness in the understanding of the subject.

Ques 3 Knowledge of the normal indication for the core spray line

break detection monitor.

Ques 33 Knowledge of the method to move an MSIV by use of the

MSIV test push-button.

T Ques 36 Knowledge of the air ejector off gas radiation rnonitor signals

that willinitiate the 13-minute timer,

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Ques 38 ' Ability to use technical specifications related to inoperable

IRMs.-

Ques 43 Ability to determine procedure' entry to a given set of. ,

conditions.

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Ques 61 Knowledge of the number of drifting rods in a nine-rod array

that require placing the mode switch in shutdown. .

- Ques 76 Ability to determine the method and reason for depressurizing

the reactor to a given set of conditions. !

Ques 85 Knowledge of the method to track the duration of

surveillances.

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During the operating test, at least two applicants performed poorly in each of the ,

'following areas: !

Refueling operations

Recognizing a loss of control room annunciators

- The above test items represent areas of weak understanding or performance and are

provided to enable improvement of the training program.

During.the dynamic simulator test, the following item was significant and a

consistent positive observation. j

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Knowledge and understanding of the emergency operating procedures

(EOPs).

During the development and administration of the examination, the examiners noted

the following item for further BECO consideration of possible procedure

improvements.

Emergency Operating Procedure 5.3.21 page 34 of 58 indicated that the

installation of the jumper in panel C915 from jumper location DD-24 to DD-

25 defeated the high drywell pressure and low RPV levelisolation signals for

MO-47, Shutdown Cooling Outboard Isolation Valve. This jumper also

affected isolation signals for MO-29B LPCIinjection valve. The procedure

did not provide a note that this valve was also affected by installation of the

jumper.

c. Conclusions j

Five_of six of the applicants were well prepared for the examination, and as a result, j

five applicants passed the examination. One SRO instant applicant failed the l

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examination. ;

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05.2 Reactivity Manipulations

a. Scope

The inspector reviewed the BECO records to determine how the applicants complied

with 10 CFR 55.31(a)(5). This section of 10 CFR requires that applicants must

perform five significant control manipulations on the plant that affect reactivity or

power level.

b. Findinos

BECO Document 0-RO-04 "NRC Licensed Nuclear Plant RO/SRO initial

Qualification," dated August 1996, required a minimum of five significant reactivity

manipulations with amplification that effort should be made to diversify the

reactivity manipulations. Ten examples were identified for meeting the requirement.

Four of the examples related to 10% power changes with control reds or

recirculation flow. The inspector considered each of the ten examples as an

appropriate significant control manipulation.

Based on review of the individual applicant qualification records, four of the six

applicants had performed five significant cont'ol

r manipulations on the plant,

although one of these four applicants did not have diverse manipulations.

Based on review of the individual applicant qualification records, the NRC examiner

identified on May 5,1997, that two of the six applicants had not performed five

significant control manipulations on the plant. Although the BECO guidance

specified the minimum conditions for a manipulation, the minimum conditions had

inappropriately been used to credit more than one manipulation when a single,

extended power change occurred. For example, one applicant reduced power with

recirculation flow from 100% to 68% over 56 minutes. BECO considered this to be

three of the five significant control manipulations. The NRC staff disagreed with

BECO and considered this to be one significant control manipulation. Another

applicant reduced power from 100% to 50% initially with recirculation flow and

then later with control rods over 77 minutes. BECO considered this to be all five of

the required five significant control manipulations. The NRC staff disagreed with

BECO and considered this to be two significant control manipulations. BECO was

informed of the examiner's conclusion and informed that this would not impact the

administration of the remainder of the examination. The resolution of this issue was

pursued after the examination was administered.

The final applications submitted on April 18,1997, indicated that these two

applicants had performed their five required significant control manipulations. After

the NRC staff review of the supporting data for the application, the .NRC staff

concluded that one applicant had performed two of the five significant control

manipulations and the other applicant had performed three of the five significant

control manipulations. These two applicants did not meet the requirements of 10

CFR 55.31(a)(5).

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In discussions with BECO and the NRC on June 4,1997, the NRC reiterated the

NRC position and informed BECO that these two applicants passed the examination

but would not be issued licenses until the applicants and BECO submitted revised

Form NRC-398s after five significant control manipulations were performed on the

plant. BECO acknowledged the NRC staff finding and indicated that they had

initiated actions to have the applicants perform additional significant control

manipulations on the plant after the examination was administered and would

submit revised Form NRC-358s.

BECO submitted revised Form NRC-398s in a letter dated June 1,1997. BECO also

provided the details of how the applicants satisfied the 10 CFR requirement for

significant control manipulations. Based on the revised applications and supporting

data, the NRC subsequently issued licenses for these individuals.

c. Conclusion

The BECO guidance and examples of how to meet the requirements of 10 CFR

55.31(a)(5) were acceptable. However, the BECO practice of giving multiple

significant control manipulation credit for a single, eendad power change was not

acceptable. The examiner concluded that BECO had violated 10 CFR 55.31(a)(5),

which requires that applicants for operator licenses must have performed five

significant control manipulations on the plant that affects affect reactivity or power l

level. With the multiple manipulations removed, one SRO applicant had performed I

two significant control manipulations, and another SRO applicant had performed

three significant control manipulations. (VIO 97-06-01)

E8 Review of UFSAR Commitments

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

special focused review that compares plant practices, procedures, and/or

parameters to the UFSAR descriptions. While performing the examination activities

discussed in this report, the examiners reviewed portions of the UFSAR that related

to the selected examination activities, questions or topic areas. The particular

section reviewed was Table 5.2.4. The specific question reviewed was consistent

with the UFSAR.

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V. Manaaement Meetinas

X1 Exit Meeting Summary

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

examination process with members of BECO management. BECO acknowledged the !

examiners' observations. The BECO personnel present at the exit included the following:

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J. Alexander, Training Manager .

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M. Briggs, Principal Instructor

K. DiCroce, Sr. Regulatory Affairs Engineer

L. Olivier, Vice President Nuclear l

M. Santiago, Operations Training Manager 1

(T. Sullivan, Plant Manager .

.T, Trepanier, Operations Department Manager !

T. Venkataraman, QA Group Manager

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NRC Personnel-

S. Dennis, Operations Engineer l

D. Florek, Sr. Operations Engineer ]

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R. Laura, Senior Resident inspector

C. Sisco, Operations Engineer

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Attachments: !

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1. SRO Examination and Answer Key

2. Facility Comments on Written Examinations ;

3. NRC Resolution of Facility Comments i

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4. Simulation Facility Report I

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ATTACHMENT 1

SRO Examination and Answer Key

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U. S. NUCLEAR REGULATORY COMMISSION

SITE SPECIFIC EXAMINATION

SENIOR OPERATOR LICENSE

REGION 1

APPLICANT'S NAME:

FACILITY: Pilarim 1

REACTOR TYPE: BWR-GE3

DATE ADMINISTERED: May 5,1997

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%.

Examination papers will be picked up four (4) hours after the examination starts.

TEST VALUE APPLICANT'S SCORE FINAL GRADE %

100.00

All work done on this examination is my own. I have neither given nor received aid. l

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Applicant's Signature

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SENIOR REACTOR OPERATOR Prga 2

ANSWER SHEET

Multiple Choice (Circle or X your choice)

. lf you change your answer, write your selection in the blank.

MULTIPLE CHOICE O23 a .b cd __

001 a b c d 024 a b c d

002 a b'c d _

O25 a b c d

003 ' a b c d 026 a b c d

004 a bc d 027 a b c d

005 a b c d 028 a b cd

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

1 014 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 cd 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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SENIOR REACTOR OPERATOR - Pzgn'3

ANSWER SHEET-

Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank. ,

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046'.a b c d 069 a. b c d l

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047 a'b c d_ 070 a b c-d- )

048 ab c d 071 a b c d

049 a bc d 072 a b c d )

050 a b c -d 073 ab c d

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074 a b c d l

051 a b-c'd

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052' a b c d 075 a b c d

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053 a b c d 076 a b c d i

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054 a b c d 077 a b c d

055 a b c d 078 a b c d

056. a b c d 079 a b c d 1

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

, 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 cd

068 -a b c d 091 a b c d

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SENIOR REACTOR OPERATOR Prg3 4

ANSWER SHEET

Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

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092 a - b c. d

093 a b c-d _

094 a b c d _

095 a b c d _

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 _

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( * * * * * * * * * * END OF EX AMIN ATION * * * * * * * * * * )

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS

During the administration of this examination the following rules apply:

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1. Cheating on the examination means an automatic denial of your application

and could result in more severe penalties.

. 2. . After the examination has been completed, you must sign the statement on '

the cover sheet indicating that the work is your own and 'you have not

received of given assistance in completing the examination. This must be

done after you complete the examination.

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3. 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. cheating. ,

! 4. Use black ink or dark pencil ONLY to facilitate legible reproductions.

5. Print your name in the blank provided in the upper right-hand corner of the

examination cover sheet and each answer sheet.

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~ 6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER

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PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

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7. The point value for each' question is indicated in parentheses after the ;

question.

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t 8. If the intent of a question is unclear, ask questions of the examiner only.

9. When turning in your examination, assemble the completed examination with

examination questions, examination aids and answer sheets in addition,

, turn in all scrap paper.

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10. Ensure allinformation you wish to have evaluated as part of your answer is

on your answer sheet. Scrap paper will be disposed of immediately

following the examination.

i 11. To pass the examination, you must achieve a grade of 80% or greater.

12. There is a time limit of four (4) hours for completion of the examination.

13. 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 stillin progress, your license may be

i denied 'or revoked.

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SENIOR REACTOR OPERATOR Pcgs 7

. QUESTION: 001 (1.00)

The HPCI system has automatically initiated due to a low reactor water

level. Drywell pressure remains within normal limits. The HPCI turbine

slowly lowers reactor pressure. Reactor pressure continues to decrease

and reaches 80 psig.

Which ONE of the following is the expected automatic response?

a. Group IV isolation, but no HPCI turbine trip and no Group Vil

isolation

b. Group IV isolation and HPCI turbine trip, but no Group Vil

isolation

c. Group Vilisolation and HPCI turbine trip, but no Group IV

isolation

d. Group IV solation, Group Vil isolation, and HPCI turbine trip

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OUESTION: 002 (1.00)

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Which ONE of the following signals will NOT require resetting the Trip

and Throttle Valve to restart the RCIC turbine?

! a. RCIC Turbine Mechanical Overspeed

b. Reactor high water level

c. Manual trip pushbutton on 904 panel

d. High Steam Supply line differential pressure

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- SENIOR REACTOR OPERATOR _ P ga 8

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QUESTION: 003 (1.00)

With the plant operating at 100% power, the 'A' Core Spray Line Break

Detection Monitor is reading approximately -3.0 psid.

This reading is:

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a. Indicative of an 'A' Core Spray Line break inside the shroud.

b. indicative of an 'A' Core Spray Line break outside the shroud.

c. normal due to the differential pressure across the dryers and

separators being approximately -3.0 psid at 100% power.

d. normal due to changes in water density after the instrument was

calibrated to read zero under cold conditions.

QUESTION: 004 (1.00)

The following conditions exist:

- SBLC Tank Temperature 45 Degrees F

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SBLC Tank Volume 4000 gallons

- SBLC Tank Concentration 9.1% weight %

- B-10 Isotope Enrichment 53 %

What is(are) the MINIMUM required action (s) that you as the NWE should

immediately initiate?

a. Perform a SBLC flow test.

b. Determine whether the sodium pentaborate solution meets the

original design criteria.

c. Perform a SBLC flow test and determine whether the sodium

pentaborate solution meets the original design criteria.

d. Immediately commence a plant shutdown such that the plant can

reach cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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SENIOR REACTOR OPERATOR Paga 9

QUESTION: 005 (1.00)-

Given the following conditions:

- The plant is in cold shutdown.

- An RHR system test is in progress.

- The LPCI Override Control Switch (S178) has been taken to

MANUAL OVERRIDE.

- Drywell pressure is at O psig and steady.

- As part of the test, reactor water level is simulated at -60

inches.

- The next step in the procedure is to take the control switch

for the Torus Spray Valve MO-1001-37B to open.

Which ONE of the following explains why MO 1001-37B will NOT

open when the control switch is taken to open?

a. The 15 minute time delay is not timed out.

b. Drywell pressure is at atmospheric.

c. The RPV Level Override Keylock Switch (S188) is not in MANUAL j

OVERRIDE.

d. The 5 minute time delay has not timed out and the MO-1001-28B i

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is not closed.

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SENIOR REACTOR OPERATOR Pegs 10

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QUESTION: 006 (1.00)

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The following conditions exist:

- The plant is operating at 100% power.

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The "A" SBGT Fan is in AUTO. -

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The "B" SBGT Fan is in STBY. 3

- A valid SBGT Initiation signal occurs.

- The "A" SBGT Fan initially starts, runs for 10 seconds, then

trips for an unknown cause.

Which ONE of the following describes the expected automatic response of

the "B" SBGT Fan?

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The "B" SBGT Fan will:

a. start when the initiation signal is received, run for 65

. seconds and then stop, then restart.

b. start immediately after the "A" SBGT Fan trips and continue

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running uninterrupted

c. start after 65 seconds and continue to run uninterrupted.

d. start when the initiation signal is received and continue

running uninterrupted. j

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SENIOR REACTOR OPERATOR Prgs 11

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QUESTION: 007 (1.00). ,

The following conditions exist:

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The plant is at 100% power.

- it is determined that the 3A SRV will NOT open under ANY

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condition.

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Which ONE 'of the following states the MINIMUM action REQUIRED by

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Technical Specifications?

a. Place the plant in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Reduce reactor coolant pressure below 104 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. Provided HPCI is operable, enter 14 day LCO. When this LCO is

expired, place the plant in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. Provided HPCI is operable, enter 14 day LCO. When this LCO is

expired, reduce reactor coolant pressure below 104 psig within

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24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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QUESTION: 008 (1.00)

During a high drywell pressure condition, a valid ADS signal exists and

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. the ADS system has initiated. With the initiation signal still present

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both initiation Signal Timer Reset Pushbuttons are depressed.

Which ONE of the following describes the expected automatic response of

the ADS system?

j All ADS valves will:

a. remain open.

b. close and remain closed indefinitely.

c. close and remain closed for 105 seconds then reopen.

d. close and remain closed for 11 minutes then reopen.

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SENIOR REACTOR ' OPERATOR P:ga 12

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QUESTION: OO9 (1.00)

The following conditions exist:

- The "A" and "B" Reactor Feed Pumps are in service.

- - Both Reactor Recirculation Speed demands are at 60%. -

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- An instrument failure causes the Feedwater Regulating Valves to

reduce feedwater flow to 3 Mlbm/hr with reactor water level

reaching a minimum of + 17 inches.

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- The operator takes manual control of the Feedwater Regulating

Valves and is returning water level to normal with a current

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level of + 18 inches.

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- No operator action is taken on the Reactor Recirculation

System.

Which ONE of the following describes the expected response of the l

. Recirculation Flow Controllers? '

'The Recirculation Flow controllers will demand lowering speed to

a. 44% without a rate limitation signal.

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.b. 44% at a rate of 1.5% per second.

c. 26% without a rate limitation signal.

d. 26% at a rate of 1.5% per second.

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' QUESTION: 010 (1.00)

An Emergency Diesel Generator (EDG) has started due to a LOCA signal.

Which ONE of the following signals will cause an EDG trip?

a. Engine Overspeed

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b. Engine Low Lube Oil Pressure

c. Engine High Lube Oil Temperature

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d. Engine Crankcase High Vacuum

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SENIOR REACTOR OPERATOR: ' Piga 13 )

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QUESTION: 011 (1.00) l

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The following conditions exist:

- HPCI is injecting water from the CST to the RPV. l

- The HPCI Suction Valves From Suppression Chamber MO-2301-35 and-

MO-2301-36 Control Switches are in Auto.

- The CST Low Level Alarm comes in.

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Which ONE of the following describes the expected response of the HPCI

system?

The HPCI Suction From CST MO-23016 will receive a close signal-

)

a. as soon as both the MO-2301-35 and the MO-2301-36 valves reach

. full open.

- b. as soon as both the MO-2301-35 and MO-2301-36 valves come off

their closed seats,

c. as soon as either the MO-2301-35 or the MO-2301-36 valve

reaches full open. l

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d. at the same time the MO-2301-35 and MO 2301-36 valves receive ;

an open signal. )

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QUESTION: 012 (1.00)

Which ONE of the following states where the RCIC turbine receives stearn

and where the RCIC pump discharges?

a. Steam from "C" Main Steam Line and Discharge to "A" Feedwater

Line

b. Steam from "D" Main Steam Line and Discharge to "B" Feedwater

Line

c. Steam from "D" Main Steam Line and Discharge to "A" Feedwater

Line

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d. Steam from *C" Main Steam Line and Discharge to "B" Feedwater

Line

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l SENIOR REACTOR OPERATOR

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OUESTION: 013 (1.00)

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Which of the,following SRM rod block (s) is(are) bypassed by moving IRM

range switches from Range 2 tc Range 3? l

a. SRM Downscale Rod Block only -

b. SRM Inoperable Rod Block only ;

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c. SRM Downscale Rod Block and Detector Retract Not Permitted Rod

Block

d. SRM inoperable Rod Block and SRM Downscale Rod Block

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QUESTION: 014 (1.00) ,

With Reactor Power at'100%, an SRV spuriously lifts. Action to close !

the valve are successful. Immediately after valve closure, the j

downstream temperature is checked.

Which ONE of the following is an expected approximate downstream

temperature?

a. 212 degrees F

b. 295 degrees F..

c. 375 degrees F .l

d. 525 degrees F

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SENIOR REACTOR OPERATOR P gs 15

-QUESTION: 015 (1.00)

The following conditions exist:

- A half scram exists on RPS "A" due to APRM testing.

- A fire caused a loss of RPS Bus "B" and a full scram.

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- The half scram testing was stopped and APRMs were returned to

normal.~

- The SCRAM DISCHARGE INSTRUMENT VOLUME HI LEVEL SCRAM BYPASS

switch is then taken to bypass.

Which ONE of the following describes when the RPS "A" half scram may be

reset?

a. immediately.

b. after the air dump test switch is placed in isolate.

c. after the SDIV vent and drain valves come fully open.

d. after RPS "B" is energized.

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QU' - ION: 016 (1.00)

The mode switch is in RUN. Which ONE of the following scram signals is

automatically bypassed 2 seconds after taking the mode switch to

SHUTDOWN?

, i

a. Mode switch in shutdown

b. Main steam isolation valve closura

c. Turbine stop valve closure

d. Scram discharge volume high level

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

'

SENIOR REACTOR OPERATOR. Pags 16

QUESTION: 017 (1.00).

. The ATWS logic system has automatically initiated due to low reactor

water level.

~

Which ONE of the following actuations will be delayed by 9 seconds?

a. Rod insertion

b. Reactor Recirc Pump Field Breaker Trip

'

c. Reactor Recirc Pump Drive Motor Breaker Trip

d. Reactor Feed Pump Trip

{

QUESTION: 018.(1.00)

'

. During a reactor shutdown, the control rod selected on the Rod Select

Matrix is NOT in the rod group of the latched step. As reactor power

' decreases, at what point will this condition cause an insert and

withdraw block?

a. Steam Flow drops below 35%

b. All APRM readings drop below 20%

c. Steam Flow or Feed Flow drops below 20%

d. Steam Flow and Feed Flow drop below 35%

,

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SENIOR REACTOR OPERATOR P:gs 17

QUESTION: 019 (1.00)

The following conditions exist:

- The plant is operating at 100% power.

- APRM "C".is bypassed for maintenance.

- APRM "E" then fails giving a constant reading of 95% regardless

of input.

- A half scram already exists on RPS "B"

Which ONE of the following meets the action REQUIRED?

a. Initiate insertion of operable rods and complete insertion of

all operable rods within sixteen hours.

b. Reduce power level to IRM range and place mode switch in the

startup/ hot stand'y position within eight hours.

c. Reduce turbine load and close main steam isolation valves

within eight hours,

d. Reduce power to less than 45% of design.

QUESTION: 020 (1.00)

A TIP trace is being performed when a high drywell pressuro signal

occurs. Select the expected automatic action.

a. The shear valve fires with the detector stillin the core.

b. The ball valve closes with the detector stillin the core.

c. The detector withdraws into its shield and the ball valve

closes,

d. The detector withdraws into its shield and the shear valve

fires.

,

e- A ,m. asa _ -ire.d e 4 # ww .e4.1 - ,a 4 4 +4 -

4- beM iJ---'a-

,

SENIOR REACTOR OPERATOR Pi:gs 18

.

QUESTION: 021 (1.00)

.

'

While operating at 80% power, an instrument failure causes the throttle

pressure sensed by the EPR to fail high. No operator action is taken.

Which ONE of the following is the expected result? i

a. The reactor would scram on a high pressure scram signal. t

b. The MPR would take control and pressure would increase by

approximately 10 psi.

c. The reactor would scram on a low pressure scram signal.

'

d. The reactor would scram on a MSIV closure scram signal.

QUESTION: 022 (1.00)

At 500 psig during a reactor startup and heatup, the #1 Bypass Valve

(BPV) comes partially open. 1

i

'

Which ONE of the following errors is the cause?

Failure to maintain the:

,

l

a. EPR 40-80 psig below reactor pressure l

l

b. EPR 40-80 psig above reactor pressure ;

I

c. MPR 40-80 psig below reactor pressure

d. MPR 40-80 psig above reactor pressure

!

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. . . . -

SENIOR REACTOR OPERATOR P:ga 19

-

QUESTION: 023 (1.00)'

With the plant operating at 20% power both Reactor Recirculation Pumps

irip. The operator manually scram the reactor. Post scram, fuel zone

level indicators read:

a. falsely high since less flow exists through the jet pumps than

existed during calibration conditions.

, b. falsely low since less flow exists through the jet pumps than ;

existed during calibration conditions.

c. falsely low due to decreased density of the water in the vessel

against calibrated conditions,

d. falsely high due to decreased density of the water in the

vessel against calibrated conditions.

QUESTION: 024 (1.00)

1

With the "A" Loop of RHR in Lo oling, RPV level decreased to 12 l

inches. The Shutdown Cooling Outtsumo I:,ulation Valve MO-1001-47 l

stopped in mid-stroke. All other valves have responded as expected. i

!

Which ONE of the following is REQUIRED in order to open the "A" Loop !

LPCI Injection Valve #2 MO-1001-29A?

a. The MO-1001-47 valve must be closed,

b. The MO-1001-29A must be manually reset,

c. Reactor coolant pressure must be greater than 76 psig,

d. The Group ll isolation signal must clear and the Group 11 logic l

must be reset.

I

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SENIOR REACTOR OPERATOR PJgn 20

QUESTION: 025 (1.00)

A reactor scram has occurred. Electrical busses A-5 and A-6 have

transferred to the Start-up Transformer. Which ONE of the following

-describes the drywell cooler response?

a. The running drywell coolers will trip and start after a 45

second time delay. The drywell coolers in standby remain in

standby,

b. The running drywell coolers will trip. The drywell coolers in

standby will start after a 45 second time delay.

c. The running coolers will stay in service. The drywell coolers

in standby willimmediately start when A-5 and A-6 are

reenergized.

d. The running coolers will stay in service. The drywell coolers

in standby will start after a 45 second time delay.

!

QUESTION: 026 (1.00) l

Primary Coolant Temperature is 245 degrees F when Shutdown Cooling is

placed in service, immediately thereafter, a fire disables the Shutdown

Cooling Outboard Isolation Valve MO-1001-47 motor operator. The valve 1

is in the open position. I

Which ONE of the following meets the MINIMUM REQUIRED action?

a. Verify the ability to manually close the MO-1001-47 valve, then

reestablish shutdown cooling. )

)

b. Verify the ability to close the MO-1001-50 valve, then

reestablish shutdown cooling,

c. Close either the MO-1001-47 or MO-1001-50 valve and open the

respective breaker.

d. Station an operator to manually close the MO-1001-50 valve if

required and continue in shutdown cooling.

,._

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SENIOR REACTOR OPERATOR Paga 21

QUESTION: 027 (1.00)

l

When valving in a CRD hydraulic control accumulator, the 305-102

'(5Nithdraw Riser Isolation Valve) and the 305-112 (Scram Discharge Riser

isolation Valve) are required to be open prior to opening the 305-101 1

l

(Insert Riser isolation Valve). This prevents-

a. .a single rod scram when opening the 305-101 valvo. I

b. excessive scram time of that rod in the event of a reactor

scram. l

!

c. damage to the accumulator in the event of a reactor scram. I

d. damage to the drive mechanism in the event of a reactor scram. 1

1

)

Il

l

QUESTION: 028 (1.00) l

l

While operating at 100% power, a control rod is determined to be i

uncoupled. Attempts to couple the rod have been unsuccessful. l

l

Which ONE of the following states the MINIMUM REQUIRED actions?

a. Verify that the control rod can be moved with drive pressure

and maintain the control rod at the target position.

b. Fully insert the control rod and hydraulically disarm the CRD.

l

'

c. Fully insert the control rod and electrically disarm the l

directional control valves. l

l

d. Fully insert the control rod, electrically disarm the )

directional control valves and then declare the rod inoperable. i

!

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1

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, SENIOR REACTOR OPERATOR Pzga 22 i

QUESTION: 029-(1.00)

With the plant at pnwer, it is determined that the MO-1001-37B (B Loop

Torus Spray) and MO-1400-25A (A Loop Core Spray Inboard injection)

valves have failed their operability test. Both volves are currently

closed.

-

The maximum time allowed before the plant must be in COLD SHUTDOWN is:

a 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (1 day).

96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (4 days),

c. ~ 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months (7 days).-

d.192 hours0.00222 days 0.0533 hours 3.174603e-4 weeks 7.3056e-5 months (8 days).

QUESTION: 030 (1.00)

A tagout, which has been in effect on the "A" Reactor Recirculation Pump

for 7 days, has just been cleared. The "A" Reactor Recirculation Pump

is started and immediately manually tripped. On the second start

p attempt, the pump starts and runs for 10 minutes and then is manually

,

tripped.

When is the SOONEST that another start of the "A" Reactor Recirculation

'

Pump may be attempted?

a. Immediately

b.15 minutes after the second trip

c. 45 minutes after the second trip

d. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the second trip

i

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SENIOR REACTOR OPERATOR Pags 23

I

QUESTION: 031 (1.00)

l

The plant is operating at 100% power when the "B" Reactor Recirculation i

Pump trips. No operator action is taken.

Which ONE of the following describes the initial steady state to final

steady state change in the "A" Reactor Recirculation Loop Jet pump flow

and the reason for the change?

The "A" Reactor Recirculation Loop Jet pump flow will: l

l

a. increase due to lower core pressure drop. I

b. increase due to decreased core voiding. l

l

c. decrease due to higher core pressure drop.

d. decrease due to increased core voiding.

1

QUESTION: 03 (1.00)

While operating at 0% power, it is determined that th Main Steam Line

High Flow switches o the "B" Main Steam Line will N trip under a high

flow condition.

Which ONE of the following the MINIMUM RE IRED action? r

a. Direct l&C personnel to m ually trip t inop blejpi Eiles. f

b. Direct l&C personnel to manu ly i ert a ha f

isolation on the "B" Group 1 Cha 1. (j p s

c. Initiate an orderly shutdown db in Cold Shutdown Condition

within a MAXIMUM of 30 h urs afte he instrument failure.

d. Initiate an orderly shutd n and have th Main Steam Lines

isolated within a MAXI UM of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months a r the instrument

failure.

. . . - .-. - . . . - . - - - . ~ . . . .

SENIOR REACTOR OPERATOR. P:ge 24

QUESTION: 033-(1.00)

Depressing a.n outboard MSIV test pushbutton will:

a. energize the AC test valve and vent air from the underside of

the piston.

b. energize the AC test valve and admit air to the underside of

the piston.

,

c. deenergize the AC test valve and vent air from the underside of

the piston. -

' d. deenergize the AC test valve and admit air to the underside of r

the piston.

QUESTION: 034 (1.00)

Which ONE of the following conditions requires Rod Block Monitor

Operability?

a. MCPR is 1.35 and Reactor Power is 25%.

b. MCPR is 1.45 and Reactor Power is 75%.

c. MCPR is 1.55 and Reactor Power is 95%.

d. MCPR is 1.60 and Reactor Power is 100%.

t

w-

. SENIOR REACTOR OPERATOR P g) 25

QUESTION: 035 (1.00)

A loss of 120V Bus A (Y-3) occurs.

Which ONE of the following describes the effect on the RWCU system?

a. Half of the logic for closing the MO-2 and MO-5 valves is made

up.

b. MO-2 goes closed. As soon as MO-2 comes off the open seat, the

operating RWCU pump (s) will trip. MO-5 remains open.

c. MO 2 goes closed. As soon as MO 2 comes off the open seat, the ,

operating RWCU pump (s) will trip and MO-5 will go closed. j

d. MO-5 goes closed. As soon as MO-5 comes off the open seat, the

operating RWCU pump (s) will trip. MO-2 remains open.

QUESTION: 036 (1.00)

The OFF GAS ISOL CH PRM SEL switch is in position 2. Which ONE of the

following conditions of the Air Ejector Off Gas Radiation Monitors will 4

cause tiie 13 minute timer to initiate?

a. Hi radiation signal on both channels

b. Hi Hi radiation signal on one channel

c. Hi radiation signal on one channel and Downscale Trip on the !

other channel

]

d. Downscale trip on one channel and inop trip on the other

channel

1

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_. _ . . _

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l- SENIOR REACTOR OPERATOR P ga 26

!

QUESTION: 037 (1.00)

l

A reactor startup is in progress with reactor power in the intermediate

range. IRM "A" then starts to intermittently swing upscale and then

downscale.

Which ONE of the following conditions on IRM "A" will cause a Rod Block

but NOT cause a Half Scram?

The IRM reads:

a.1 (on the 0-40 scale) while on range 1. *

,

b. 3 (on the 0-40 scale) while on range 3. i

1

I

c. 36 (on the 0-40 scale) while on range 5.

d. 39 (on the 0-40 scale) while on range 7. !

QUESTION: 038 (1.00)

With the Mode Switch in Startup, at 1200 on 5/5/97, the Downscale Trips

for IRM Channels "A", "B", and "E" are made inoperable. ]

Which ONE of the following is the LATEST that one of these channels must

be placed in a tripped condition?

a. 1300 on 5/5/97 i

b.1200 on 5/6/97

c.1200 on 5/12/97

d. 1300 on 5/12/97

,

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SENIOR REACTOR OPERATOR PIga 27 '

OUESTION: 039 (1.00)

The plant was. operating at 100% power with the "B" CRD pump in service.

Subsequently, a valid LOCA signal generated a scram. The plant

responded as expected except, the startup transformer feeder breaker

to bus A 5 failed to close. A-5 has been automatically energized from the

shutdown transformer.

Which ONE of the following describes the status / availability of the CRD pumps?

a. "B" CRD pump is running.

"A" CRD pump can be started since no load shed signal was

generated.

b. "B" CRD pump is not running.

"A" and "B" CRD pumps cannot be started due to load shed

signal.

c. "B" CRD pump is not running.

"A" and "B" CRD pumps can be started since no load shed signal

was generated.

,

d. "B" CRD pump is running.

"A" CRD pump cannot be started due to load shed signal.

l

1

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- SENIOR REACTOR OPERATOR Pign 28

,

QUESTION: 040-(1.00)-

Which ONE of the following administrative precautions related to valves

are required when lining up RHR "A" loop for shutdown cooling?

a. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed.

MO-1001-7C "RHR PUMP C TORUS SUCTION" red tag closed. ,

MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag

closed.

MO-1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow

tag closed.

b. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag closed.

MO-1001-7C."RHR PUMP C TORUS SUCTION" red tag closed.

MO-1001-438 "RHR PUMP B SHUTDOWN COOLING SUCTION red tag

closed.

MO-1001-43D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag

closed.

c.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" red tag

closed.

1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" red tag

closed.

MO-1001-43B "RHR PUMP B SHUTDOWN COOLING SUCTION red tag

closed.

MO-100143D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag

closed.

d.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" yellow tag

closed.

1001-366A "RHR PUMP A SUCTION VALVE FROM THE TORUS" yellow

tag closed.

MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag

closed. ;

MO 1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow tag

closed.

_ _

.. _ _ ,

_- _

SENIOR REACTOR OPERATOR Pags 29

1 QUESTION: 041 (1.00)

_

Given the following conditions:

.

The plant is in cold shutdown

No recirculation pumps are in service

RHR pump "A"is in shutdown cooling

RWCU is in service

Reactor shutdown level instrument indicates 40 inches

Which ONE of the fo:iowing describes reactor coolant temperature

indication if the "A" RHR pump trips. Assume no operator action,

a. Recirc loop "A" temperature indicator is representative of

reactor coolant temperature.

b. Recirc loop "B" temperature indicator is representative of

reactor coolant tempesture.

c. RWCU bottom head drain temperature indicator is representative

of reactor coolant temperature.

d. No temperature indicator is representative of reactor coolant

temperature,

,

i

d

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

,

SENIOR REACTOR OPERATOR Prgs 30

.

- QUESTION: 042 (1.00)

.The following conditions exist:

!' - EOP-02 is being executed _. .

-

The Mode Switch is in Shutdown and ARI has been initia'ted.

'

- The MSIVs are closed.

- Reactor power is 2.5% and no boron has been injected.

- ~ Alternate _Depressurization is required by EOP-04.

- Four SRVs can be opened.

Which ONE of the following actions should be taken to control reactor '

I water level?

4

a. Secure all sources of injection. When pressure decreases below

200 psig, slowly inject with LPCI.

b. Secure all sources of injection. When pressure decreases below

. 400 psig, slowly inject with the Condensate Pumps.

c. Secure all sources of injection except CRD and RCIC. When

, pressure decreases below 200 psig, continue injection flow rate

with RCIC and CRD.

d. Secure all sources of injection except CRD and RCIC. When

pressure decreases below 270 psig, slowly inject with the

.

Condensate Pumps.

4

l

..

0

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

1

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SENIOR REACTOR OPERATOR Pcgs.31 i

,

l

.)

QUESTION: 043 (1.00)

The following conditions exist: l

- A manual' scram was inserted from 20% power. l

- No other scram signals exist. !

l

- Reactor power is on intermediate range 6 and decreasing.

- Three control rods are at position 06. All other rods are ,

fully inserted.

Which ONE of the following is the required action?

a.' ' enter PNPS 2.1.6. No EOP entry is required.

b. ' enter EOP-01, then exit EOP-01 and enter EOP-02 at R-1. ,

c. enter PNPS 2.1.6, " Reactor Scram", then exit PNPS 2.1.6 and

.

enter EOP-02 at R 1.

d. enter PNPS 2.1.6, " Reactor Scram", then enter EOP-02 and

execute concurrently with PNPS 2.1.6.

i

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< SENIOR REACTOR OPERATOR pig 3 32-

s

QUESTION: 044 (1.00)

The following conditions exist:

- EOP-02 is being executed.

-- Boron is being injected with the SBLC system. '

- Initial SBLC tank level was 4100 gallons. ;

Reactor Water Levelis being lowered to reduce reactor power. ;

- Current SBLC tank level is 3000 gallons.

- Torus water temperature is 112 degrees F.

- Reactor water level is'-100 inches.

Which ONE of the following actions is REQUIRED 7

a. Reise reactor water level to the +12 to +45 inch band and

perform Alternate Depressurization.

'

b. Raise reactor water level to the + 12 to +45 inch band. Do not

perform Alternate Depressurization. !

!

c. Maintain reactor water level at its current value and perform

. Alternate Depressurization.

d. Maintain reactor water level at its current value. Do not

perform Alternate Depressurization.

QUESTION: 045 (1.00) {

l

While operating at 100% reactor power, reactor pressure starts to

oscillate approximately 10 psi peak to peak and pressure control is

shifting alternately from the EPR to the MPR and back to the EPR.

Which ONE of the following actions are REQUIRED?

a. DJace the EPR control switch to off,

b.' Reduce reactor power to approximately 75%.

. c. Raise the MPR setpoint to prevent pressure control from

.

swapping between regulators.

,

d. Lower.the MPR setpoint to allow the MPR to take control of

pressure.

. . . -. . .- .

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

SENIOR REACTOR OPERATOR Pcga 33

,

t

QUESTION: 046 (1.00)

With the plant in Cold Shutdown, some solvent that is improperly stored

in a Control Room locker ignites. The Nuclear Watch Engineer makes the

decision to evacuate the Control Room and to call for off-site

assistance to put out the fire. Control is established at remote -

shutdown stations 20 minutes after the Control Room evacuation.

What is the MINIMUM event level classification?

a. Unusual Event

b. Alert

'

c. Site Area Emergency

d. General Emergency

,

QUESTION: 047 (1.00)

A LOCA has occurred. Which ONE of the following REQUIRES exiting the

RPV level control leg of EOP-017

a. Reactor water level is -165 inches and increasing with Reactor

Pressure at 200 psig.

b. Reactor water level is -125 inches and decreasing with Reactor

Pressure at 175 psig.

c. Reactor water level is -125 inches and increasing with Reactor

Pressure at 100 psig.

d. Reactor water level is -125 inches and decreasing vdth Reactor

Pressure at 75 psig.

,.

,

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- SENIOR REACTOR OPERATOR P gs 34

QUESTION: 048 (1.00)

~ Which ONE of the following conditions REQUIRES Alternate Reactor- ,

Pressure Vessel Depressurization assuming a primary system is ' ,

dischaiging into secondary containment?

.

a. RCIC torus piping area temperature is 300 degrees F and RCIC

'

,

turbine area temperature is 195 degrees F.

b. HPCI compartment water levelis 8 inches and HPCI turbine area ,

temperature is 195 degrees F. .:

c. RHR "B" and "D" pump area temperature is 300 degrees F and RHR

."A" and "C" pump area temperature is 195 degrees F.

, d. Main Steam Tunnel area temperature is 300 degrees F and RHR "A"

and "C" pump area temperature is 220 degrees F.

,

QUESTION: 049 (1.00)

Following a Nitrogen Line leak in the drywell, AO-4356 (Nitrogen / Air

Isolation Valve to the Drywell) was closed. By calculation, how many

times over the next eight hours can each SRV be actuated?

NOTE: COUNT EACH OPEN AND CLOSE CYCLE AS ONE ACTUATION

a. 5 ,

b. 10

c. 20

4

d. 40

.

J

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,a ,. - y a-

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SENIOR REACTOR OPERATOR P ga 35

QUESTION: 050 (1.00)

in the event that torus water level cannot be maintained above 95

inches, HPCl is secured in order to prevent:

a. exceeding the Primary Containment Pressure Limit. '

b. exceeding the Pressure Suppression Pressure.

c. exceeding the Heat Capacity Temperature Limit,

d. isolating HPCI on high exhaust pressure.

' QUESTION: 051 (1.00)

The following conditions exist:

- Reactor pressure is 10 psig.

- Drywell pressure is 4 psig.

- ' Torus bottom pressure is 15.2 psig.

-

Torus water level is 303 inches.

Select the correct action and its reason.

.

Under these conditions:

a. Alternate RPV Depressurization is required to prevent SRV Tail

Pipe failure,

b. Suppression Chamber Spray initiation is required using enly

those RHR pumps not required to provide adequate co cooling.

c. Suppression Chamber Spray initiation is not allowed sinc. the

Torus Spray Sparger is covered.

d. Suppression Chamber Spray initiation is not ai. owed since the

Torus-Drywell Vacuum breakers are covered.

.

,

' SENIOR REAC f0R OPERATOR P;gs 36

QUESTION: 052 (1.00)

A trip of the "A" Reactor Recirculation Pump has occurred. The plant is

operating in Region ll of the Power / Flow Map after the immediate actions

2

of 2.4.17 have been completed.

Which ONE of the following is REQUIRED 7

Exit Region 11 by:

a. manually scramming the reactor.

b. restarting the "A" Reactor Recirculation Pump.

c. increasing the speed of the "B" Reactor Recirculation Pump.

d. inserting control rods in reverse order of the pull sheet.

l

QUESTION: 053 (1.00) !

l

The following conditions exist: l

l

-

Torus water level is 105 inches.

- Torus water temperature is 180 degrees F.

- Reactor pressure is 700 psig.

Which ONE of the following states whether Alternate RPV Depressurization

is required, not required, or prohibited and the reason.

Under these conditions, Alternate RPV Depressurization is

a. not required since primary containment limits are not exceeded,

b. required to ensure the energy released during an RPV blowdown I

can be accepted.

.

. c. required since the downcomers are now exhausting to the torus

free air space.

d prohibited since the SRV Tail Pipes are now exhausting to the

torus free air space.

!

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SENIOR REACTOR OPERATOR Paga 37

OUESTION: 054 (1.00)

To initiate a reactor scram when the control room has been evacuated, it

is undesirable to deenergize the RPS busses as the means of scramming

because:

,

a. ' nuclear instrumentation needed to monitor reactor power will

become denergized.

b. pressure control using turbine bypass valves will be lost after

the scram.

c. RPV level control will unnecessarily transfer from feedwater to

HPCI.

I

d. groups I, ll, Ill, and VI isolations will be defeated. j

QUESTION: 055 (1.00)

The following conditions exist:

- The plant is operating at 75% power.

- At 0800 one Safety Relief Valve opened.

- At 0802 EOP-03 has been entered due to torus water temperature

reaching 80 degrees F.

At what point should a Reactor Recirculation pump speed reduction and

manual reactor scram be performed?

a. Immediately when it is determined that the SRV cannot be

reclosed.

b. At 0810.

c. When torus temperature reaches 120 degrees F.

d. When the " unsafe"_ region of the Heat Capacity Temperature Limit

curve is entered.

-

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SENIOR REACTOR OPERATOR P:go 38 '

i

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,

QUESTION: 056 (1.00)

Drywell spray was initiated in accordance with EOP-03. As drywell 1

temperature and pressure are decreasing, the unacceptable region on the

Drywell Spray initiation Limit curve is entered at a Drywell temperature

of 250 degrees ~ F.

Which ONE of the following is the REQUIRED action?

a. Secure drywell spray when drywell pressure drops below 2.2

psig,

b. Secure drywell spray when torus bottom pressure drops below 2.2

psig.

c. Adjust drywell spray as necessary to maintain operation within

the Drywell Spray Initiation limit curve. ;

l

d. Immediately secure drywell spray. l

l

QUESTION: 057 (1.00)

A loss of feedwater heating has occurred. Which ONE of the following is

the REQUIRED immediate operator action?

Run back Reactor Recirculation flow until:

a. reactor power has been reduced 25% below its pretransient level

without regard to current total core flow,

b. total core flow has been reduced to 36 Mlb/hr without regard to

the current reactor power level,

c. reactor power has been reduced to at least 25% below its

pretransient level AND total core flow has been reduced to at

least 36 Mlb/hr.

d. reactor power has been reduced 25% below its pretransient level

OR total core flow has been reduced to 36 Mlb/hr.

-- -

_ , . , , . .. . .

!

I

- SENIOR REACTOR OPERATOR Paga 39 -

,

QUESTION: 058 (1.00)'

A startup is in progress with reactor pressure at 900 psig when the "A"

CRD pumps trips and the "B" CRD pump cannot be started. .Two accumulator

alarms, both in the same nine rod array, illuminate.

D

Which ONE of the following is the required. action?

,

s. Manually scram the reactor.

b. Determine the cause of the alarms. If both alarms are due to

low gas pressure then manually scram the reactor,

c. Fully insert one of the rods with an accumulator alarm and 1

-

disarm its directional control valves.

, d. Enter LCO to be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .'

QUESTION: 059 (1.00)-

While operating at 100% power, a recirculation pump seal failure causes

EOP-03 entry on high drywell pressure and high drywell temperature.

Following initiation of suppression chamber spray, drywell pressure

- stabilizes at 4 psig, torus bottom pressure stabilizes at 8 psig, and

drywell temperature stabilizes at 175 degrees F.

Which ONE of the following actions is REQUIRED?

a. Declare an Unusual Event and initiate drywell spray in

'

i

accordance with the Primary Containment Pressure leg of EOP-03.

b. Declare an Alert and in:tlate drywell spray in accordance with

the Drywell Temperature leg of EOP-03.

. c. - Declare an Unusual Event. Do not initiate drywell spray.

d. Declare an Alert. Do not initiate drywell spray.

. .-

J

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

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

SENIOR REACTOR OPERATOR Prgs 40

1

)

.

- QUESTION: 060 (1.00) I

1

With the plant operating on the 65% load line, condenser vacuum starts

to decrease. Reactor Recirculation Flow is reduced in accordance with -i

plant procedures. !

After the Reactor Recirculation Flow reductions the plant will be

operating,

a. in the scram region.

b. In the exit region.-

I

c. In the caution zone.

d. above the MELLA line.

<

QUESTION: 061 (1.00)

l

The plant is operating at 100% power when control rods start to drift.

i The MAXIMUM number of control rods in a nine rod array that are allowed

to drift WITHOUT REQUIRING tho mode switch to be placed in shutdown is:

a. one rod without regard to whether the rods are drifting in or

out.

b. two rods if rods are drat;ng in and one rod if rods are

- drif ting out. ,

l

c. two rods without regard to whether the rods arr liiting in or I

out.

[ d.~ three rods if rods are drifting in and two rods if rods are

drifting out,

t

0

l

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ta

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.,. . . , , - . -

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

SENIOR REACTOR OPERATOR Pcgs 41

OUESTION: 062 (1.00)-

The plant is operating at power when a total loss of TBCCW occurs,

immediate actions are complete in accordance with plant procedures.

Which ONE of the following describes RPV pressure and level control?

a. ' RCIC is being used in the' level control mode and HPCI is being

.used in the pressure control mode.

b. HPCIis being used in the level control mode and RCIC is being

used in the pressure control mode,

c. HPCI is being used in the level control mode and SRVs are being

used to control pressure. RCIC remains shutdown.

d. RCIC is being used in the level control mode and SRVs are being

used to control pressure. HPCI remains shutdown.

QUESTION: 063 (1.00)

The plant is operating at 100% power when a loss of Bus A5 occurs.

Which ONE of the following action (s) is(are) required? ,

a. Reduce reactor power to maintain steam tunnel temperature below

'

160 degrees F.

b. If steam tunnel temperature exceeds 160 degrees F scram the

reactor and close the MSIVs.

c. If steam tunnel temperature exceeds 160 degrees F scram the

reactor. Maintain the MSIVs open.

d. If steam tunnel temperature exceeds 160 degrees F commence a

normal plant shutdown.

I

SENIOR REACTOR OPERATOR Pcga 42

I

,

QUESTION: 064 (1.00).

With the plant at 100% power on the 100% load line, reactor water level j

starts to decrease due to unknown causes. . Level is currently + 25 inches

and is trending down at 1/2 inch per minute. I

l

Which ONE of the following is the required action assuming water level

continues to fall? :

a. Insert rods using the RPR rods until below 70% load line, then l

reduce core flow to 36-40 Mlb/hr. .)

I

b. Insert rods using the RPR rods until below 70% load line, then

'

reduce recirculation pump speed to minimum.

c. Reduce recirculation pump speed to minimum, then insert rods as

necessary to exit the caution zone,

d. Reduce core flow to 36-40 Mlb/nr, then insert rods using the

RPR rods until below 70% load line, then reduce recirculation

pump speed to minimum.

.

!

QUESTION: 065 (1.00)

The plant is operating at 100% power with the "B" Reactor Recirculation

Pump scoop tube locked when a reactor scram occurs.

Which ONE of the following actions are REQUIRED?

a. Direct a licensed operator to manually position the "B" Reactor

Recirculation MG set scoop tube to minimum speed.

b. Direct any member of the operating crew to manually position

the "B" Reactor Recirculation MG set scoop tube to minimum

speed.

c. Unlock the scoop tube, if possible, then run the "B" Reactor

Recirculation pump to minimum speed.

d. Trip the "B" Reactor Recirculation Motor Generator Set.

,

, - , , ,

SENIOR REACTOR OPERATOR Pigs 43

QUESTION: 066 (1.00)

Given the following conditions:

- A fuel leak occurs and as a result the reactor is manually

scrammed.

- Due to the fuelleak, the CRD HCU east and west areas radiation

levels reach 1200 mR/hr and 1250 mR/hr respectively.

- The west Scram Discharge Volume vent and drain valves have

failed open.

Under these conditions, Alternate RPV depressurization is:

a. not required since the CRD HCU east and west areas are

considered the same area,

b. required in order to protect secondary containment from

failing.

c. required to allow the scram to be reset and the primary system

leak isolated,

d. not required since there is no primary system discharging into

secondary containment,

i

1

i

.

i

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

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' SENIOR REACTOR OPERATOR Prga 44

4

OUESTION: 067 (1.00)

.

The following conditions exist:

-

A seismic event has caused the torus suction lines to both Core

Spray loops to crack downstream of the Core Spray Suction (MO-

1401-3) valves.

-

These cracks result in the water level in the SE and NW

Quadrants to reach 8 inches and 10 inches above the floor

respectively.

- Efforts to lower the water level are only able to maintain

level.

- There is no primary system discharge into secondary

containment.

!

Which ONE of the following is required by EOP-047 ;

1

,

a. Isolate the Core Spray suction from the torus.

1

b. Maintain the Core Spray suction aligned to the torus,

c. Perform Alternate RPV Depressurization.

d. Transfer the Core Spray suction for both loops to the CST.

OUESTION: 068 (1.00) i

Which ONE of the following conditions violates secondary containment

integrity?

a. Both drywell personnel access doors are open.

b. Reactor water cleanup MO-1201-2 (RWCU Suction) valvo is failed '

open.

~

c. Reactor building ventilation is secured due to dampers failing

closed.

d. One refuel floor exhaust isolation damper is failed open with

the other refuel floor exhaust isolation damper open and fully

operable.

-

-. _

SENIOR REACTOR OPERATOR Pega 45

OUESTIONi 069 (1.00)

The following conditions exist:

- A reactor startup is in progress

- The Reactor Mode Switch is in "Startup/ Hot Standby"

- Reactor pressure is 850 psig

The main turbine is tripped

-- A valid Group Iisolation has occurred

- All systems operated as designed

Which ONE of the following conditions caused the Group Iisolation?

a. Low main steam line pressure

b. Two main steam lines isolating

c. High main steam tunnel temperature

d. High reactor water level

1

l

QUESTION: 070 (1.00) i

l

A steam leak in the drywell has occurred and the control room crew has

entered EOP-01 and EOP-03. TI-9019 and TRU-9044 on panel C903 are l

'

broken. In accordance with the data contained in the attached 2.1.27,

which ONE of the following is the instrument run temperature for the "A"

channel instruments?

a. 208 degrees F

b. 210 degrees F

c. 216 degrees F

d, 220 degrees F

!

t - , .- - , -,

. . . . . . .. . _ . .

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

SENIOR REACTOR OPERATOR Pag 3 46 ,

!

i

OUESTION: 071 (1.00) ,

!

,

initiating suppression chamber spray prior to torus bottom pressure l

'

reaching 16 psig prevents fatigue failure of

l

a. SRV Tail Pipes. -

b. Torus Drywell vacuum breakers.

.

c. downcomers. l

-

d. the Reactor Building-Torus vacuum breakers.

!

l

QUESTION: 072 (1.00)

a

l The following conditions exist:

- A core off-load is in progress.

- The Refuel Bellows Seal Rupture alarm is received followed 2

minutes later by the Fuel Pool Low Level alarm.

- Currently an irradiated bundle has been removed from the core

but is still above the reactor vessel.

Which ONE of the following is the REQUIRED action?

a. Immediately evacuate the refuel floor and leave the bundle

hoisted above the reactor vessel,

b. Return the bundle to the in-core position that it came from, j

c. Place the bundle in the nearest open in-core position.

d. Place the bundle in the nearest open fuel pool position.

i

- . -_ .-. _

SENIOR REACTOR OPERATOR pig 3 47

OUESTION: 073 (1.00)

The following conditions exist:

-

The reactor is shutdown.

-

At 1600 all RPV water level indication was lost due to

electrical problems and EOP-16 was entered.

- At 1630 conditions to flood the RPV were established with A, B,

and D SRVs open and RPV pressure 52 psig above torus pressure.

- At 1640 electrical power was restored and water level can be

determined.

Which ONE of the following actions are REQUIRED?

a. Immediately exit EOP-16 and enter EOP-01 at L-1 and P-7.

b. Continue vessel flooding until 1819 then immediately exit EOP-

16 and enter EOP-01 at R-1,

c. Concurrently execute EOP-16 and EOP-01 at L-1 and P-7.

d. Continue vessel flooding until 1819 then stop all injection.

Verify that RPV level decreases before the MCUTL is reached, i

then exit EOP-16 and enter EOP-01 at R-1.

I

,

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SENIOR REACTOR OPERATOR- P gs 48

QUESTION: 074 (1.00)

"

The following conditions exist:

A failure to scram has occurred.

'

.-

- No boron has been injected. '

-

Reactor power is 30%.

- The Main Turbine is tripped.

-

The Main Condenser is available. ,

l -

.orus water level is normal.

Due to difficulty in establishing suppression pool cooling, the

'

-

. Heat Capacity Temperature Limit (HCTL) was exceeded.

-

Which ONE of the following states the proper method of controlling

reactor pressure? ,

j

a. Reactor pressure should be reduced using the main turbine 1

bypass valves to stay below the HCTL curve.

b. Reactor pressure should be reduced using the SRVs to stay below l

'

the HCTL curve.

,

c. Alternately depressurize using the main turbine bypass valves. i

!

d. Altemately depressurize using the SRVs.

,

OUESTION: 076 (1.00)

,

Which ONE of the following actions allow the operator to disregard NPSH

limits?

a. After a successful reactor scram, Core Spray is being used to l

maintain level between -125 to +45 inches.

b. After a successful reactor scram, LPCI is being used to

maintain level between + 12 to +45 inches.

c. Durin0 an ATWS, LPCI is being used to maintain level between

-155 to -140 inches after level was lowered until reactor power

dropped below 3%.

d. During an ATWS with Alternate RPV Depressurization required and

all SRVs INOPERABLE, LPCI is being used for injection.

,

)

, ,

. . . _ _ _ _ _

SENIOR REACTOR OPERATOR pig 3 49

QUESTION: 076 (1.00)

The following conditions exist:

- A steam leak occurs just upstream of the Main Turbine Stop

Valves with ooth MSIVs in the "A" main steam line failing to

close.

- A reactor scram is successful in inserting all rods fully.

- Both Main Stack Process Radiation Monitors have been reading

2.5E+4 for the last 25 minutes.

- Off-site release rate projections are 2 R/ hour Whole Body at

the site boundary.

Select the correct action and its reason.

Under these conditions the preferred method of depressurizing the RPV is

using:

a. SRVs because of the scrubbing potential of the torus water,

b. SRVs because the heat removal capability is greater than the

Main Turbine Bypass Valves,

c. Main Turbine Bypass Valves because the hotwell is the preferred

heat sink.

d. Main Turbine Bypass Valves because the heat removal capability

is greater than the SRVs.

s

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i

SENIOR REACTOR OPERATOR Pcgs 50

,

4

QUESTION: 077;(1.00)

- With a Reactor Building Vent Radiation Hi-Hi Alarm present, EOP-04 ,

directs the operator to verify secondary containment isolation of 1

Reactor Building Heating and Ventilation and the initiation of Standby

'

. Gas Treatment System. j

.'

\

-

'This verification will ensure that:

!

a, the Reactor Building atmosphere is contained at a positive

pressure until it can be treated and released.

I

b. a trected and controlled ground release of the activity is j

provided. j

i c. a treated and controlled elevated release of the activity is

provided. l

d. both the primary and secondary containments are maintained at a

slightly negative pressure,

i

'

l

. 1

i OUESTION: 078 (1.00)

The following conditions exist: i

1

- A reactor startup is in progress with RPV pressure at 500 psig.

" - It is determined the "A" Channel of Group i PCIS has one

reactor high water level switch (16A-K105A) that will NOT trip.

The MINIMUM time allowed to place 16A-K105A in the tripped condition is:

a. one hour,

b. two hours.

c. six hours,

d. twelve hours.

0-

- SENIOR REACTOR OPERATOR Pcge 51

- QUESTION: 079 (1.00) ,

~ The following conditions exist:

- ' A successful automatic reactor scram occurred on high reactor ;

pressure.

- The main condenser is ava'ilable but not currently in service.

- The operator is attempting to stabilize pressure between 900-

1060 psig using SRVs.

l

Re-establishing the main condenser as a heat sint::

a.- is not allowed. 1

i

b. is preferred but is allowed only if no valid MSIV isolations

exist.

c. is required immediately after valid MSIV isolation signals are

overridden.

d. is only allowed if the SRVs become unavailable. ;

l

QUESTION: 080 (1.00) l

With the plant operating at 100% power, the control room becomes

uninhabitable because of toxic gas. Evacuation is ordered and only the

immediate Actions of Pf4PS Procedure 2.4.143 were carried out.

At this point reactor water level is being maintained by:

a. Reactor Feed Pumps and CRD.

b. RCIC and CRD.

c. HPCI and CRD.

d. CRD only.

>

M

SENIOR REACTOR OPERATOR Prgs 52

OUESTION: 081 (1.00)- *

With the plant at 100% power, an MPR and EPR f ailure caused the turbine

stop valves to close and the turbine bypass valves to remain closed.

Reactor pressure peaked at 1330 psig at which time the reactor scrammed

on high flux. -

Select the statement below that correctly describes the transient.

a. No safety limit violation occurred. The Stop Valve closure

scram was the only RPS trip failure. J

b. No safety limit violation occurred. The Stop Valve closure

scram was not the only RPS trip failure,

c. A safety limit violation occurred. The Stop Valve closure

scram was the only RPS trip failure.

d. A safety limit violation occurred. The Stop Valve closure

scram was not the only RPS trip failure.

l

1

QUESTION: 082 (1.00)

Following a reactor scram, the Mode Switch should be taken to Shutdown

as soon as possible in order to:

a. disable the low steam pressure isolation.

b. enable the high reactor water level isolation.

c. insert another scram signal for 2 seconds,

d. allow MSIV closure without generating a scram signal.

_

SENIOR REACTOR OPERATOR P:ga 53 '

OUESTION: 083 (1.00)

The following conditions exist:

- The reactor was shutdown at 0230.

- Due to loss of level indications, EOP-16 was entered at 1030.

- At 1100 flooding conditions were established with 3 SRVs open.

- Flooding was stopped as soon as Flooding Cornpletion Time was

reached,

t

1

Assuming RPV levelinstruments do not respond, which ONE of the

16116 wing is the LATEST time at which injection must be reinitiated?

- a. 1214

b. 1217

c. 1254

-

d. 1257

OUESTION: 084 (1.00)

A worker in the Emergency Response Organization had 100 mrem TEDE for

the current year and 2.5 Rem TEDE lifetime prior to the declaration of

an emergency. Which ONE of the following is the MAXIMUM TEDE this

worker can receive over the course of the emergency without special

authorization?

a. 2.4 Rem

b. 2.5 Rem I

c. 4.9 Rem

d. 5.0 Rem i

!

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e

-SENIOR REACTOR OPERATOR Pago 54 ;

i

l

1

l

.

QUESTION: 085 (1.00) j

1

~

A surveillance on the Reactor Water Cleanup High Flow Isolation is due. l

.

Which ONE of the following describes how the duration of the

surveillance is tracked and when the inoperability clock begins and

ends?

a. The surveillance is tracked in the NOS Logbook.

The clock starts when the system is removed from service and

ends when the system is returned to normal lineup.

b. The surveillance is tracked in the NOS Logbook.

The clock starts when the system is removed from service and

ends when the NWE signs off the surveillance. l

c. The surveillance is tracked using an LCO Maintenance Planning

i

Checklist. The clock starts when the system is removed from

I

i service and ends when the system is returned to normal lineup.

d. The surveillance is tracked using an LCO-Maintenance Planning

Checklist. The clock starts when the syt. tem is removed from :

-, service and ends when the surveillance is signed off by the I

l

' work group.

T

5

s

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

SENIOR REACTOR OPERATOR P:gs 55

,

QUESTION: 086 (1.00)

With the plant at 5% power, a closed motor operated valve located in the

drywell must be tagged in the closed position.

Which ONE of the following is the proper method for determining that the

valve is in the closed position?

a. The position should be first verified by the indirect method

before power is isolated. The isolation of the power supply

may then be performed. Independent verification of the power

supply is not required.

b. The position should be first verified and independently

verified by the indirect method before power is isolated. The

isolation of the power supply may then be performed and

independently verified.

c. The first verifier should enter the drywell for verification of

valve position. The independent verifier may perform an

indirect verification of remote valve position.

d. The first verifier and the independent verifier should make

separate drywell entries for verification of valve position.

QUESTION: 087 (1.00)

Which ONE of the following may enter the Controls Area without receiving

permission from the NWE/NOS or Control Room Operator?

a. Operations Department Manager

'

b. NRC Resident inspector

ej. Station Director

d. The Outside Nuclear Plant Reactor Operator (NPRO)

1

.-. . . . .-

SENIOR REACTOR OPERATOR P:gs 55

QUESTION: 086 (1.00) .

With the plant at 5% power, a closed motor operated valve located in the

drywell must be tagged in the closed position.

Which ONE of the following is the proper method for determining that the

valve is in the closed position?

a. The position should be first verified by the indirect method

before power is isolated. The isolation of the power supply.

may then be performed independent verification of the power

supply is not required,

b. The position should be first verified and independently

verified by the indirect method before power is isolated. The

isolation of the power supply may then be performed and

independently verified.

c. The first verifier should enter the drywell for verification of

valve position. The independent verifier may perform an

indirect verification of remote valve position.

d. The first verifier and the independent verifier should make

separate drywell entries for verification of valve position. l

l

QUESTION: 087 (1.00)

Which ONE of the following may enter the Controls Area without receiving

permission from the NWE/NOS or Control Room Operator?

a. Operations Department Manager

b. NRC Resident inspector

c. Station Director

d. The Outside Nuclear Plant Reactor Operator (NPRO)

i

l

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

SENIOR REACTOR OPERATOR Pigs 56

- QUESTION: 088 (1.00)

Absent a basis to assign a longer duration, Which ONE of the following

is the normal duration of a temporary modification?

a. Installation until the end of the shift

b.- 6 weeks following installation

c. 6 months following installation

d. installation until the end of the refueling outage

QUESTION: 089 (1.00)

The MINIMUM amount of parallel watchstanding REQUIRED in order to

reactivate an NRC reactor operator license is:

a. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

i

b. 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months . )

i

c. seven 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months shifts.

l

d. five 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shifts.

l

QUESTION: 090 (1.00)

1

Which ONE of the following is the MINIMUM REQUIRED protective equipment l

for handling Sodium Hypochlorite outdoors? 1

Safety Goggles and:

a. Rubber Gloves

b. Rubber Gloves and Apron, Rubber Safety Boots, Forced Air

Respirator i

c. Rubber Gloves and Apron, Rubber Safety Boots

d. Rubber Gloves and Apron, Rubber Safety Boots, Respirator

i

SENIOR REACTOR OPERATOR Prgs 57

QUESTION: 091 (1.00)

Which ONE of the following conditions would allow a fail open air

operated valve to be DANGER tagged in the closed position?

'

a. The valve is gagged in the closed position with a device to

ensure it does not change state.

'

b. The DANGER tag is only for equipment protection and no

maintenance will be performed under this tagout..

c. The air supply to the valve is also DANGER tagged in the open

position.

d. A " Human Red Tag" is assigned to monitor the status of air to

the valve.

-

QUESTION: 092 (1.00)

Which ONE of the following conditions PROHIBIT the use of a " Human Red

Tag"?

a. The only qualified tagger available to be a " Human Red Tag" is

- a member of the work group.

b. Two isolation points are required to provide isolation.

'

c. The work is expected to take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to complete.

d. The work is expected to take 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to complete with only 1/2

hour left in the current shift.

l

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+

- _ . _ . _

1

.

1

SENIOR REACTOR OPERATOR Prgs 58. '

l;

.

l

QUESTION: 093 (1.00) 1

1

An offsite fire department is responding to the site during a fire in a. /

- vital area. )

' Which ONE of the following describes the security reouirements in order !

- to allow access to the protected area / vital area? l

i

a.' The fire truck and firemen must be searched prior to entering ;

~ the protected area. No additional search is required prior to

entering the vital area provided security escorts the' team.

b. No search is required of the fire truck or firemen prior to

entering the protected area provided security escorts the team, ,

however both the truck and firemen must be searched prior to >

entering the vital area. :

l

c. No search is required of the f;c - truck or firemen prior to l

entering the protected area or vital area provided security

escorts the team. l

l

d. No search is required of the fire truck or firemen prior to

entering the protected area or vital area provided security and

operations department escort the team. )

QUESTION: 094 (1.00)

You are working in a Hot Particle Control Zone (HPCZ) in a double set of

protective clothing. Which ONE of the following is the proper method of

removing the protective clothing when exiting the area?

a. Remove both sets of protective clothing at the step off pad at

the exit of the HPCZ.

b. Remove both sets of protective clothing at the step off pad at

. the exit of the buffer zone.

c. Remove the outer set of protective clothing at the step off pad

at the exit of the HPCZ and the' inner set of protective

clothing at the step off pad at the exit of the buffer zone,

d. Remove the outer set of protective clothing at the step off pad

at the exit of the buffer zone and the inner set of protective

clothing at the step off pad at the exit of the HPCZ zone.

4

... .- ,

_

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U

SENIOR REACTOR OPERATOR Pzg3 59 l

l

QUESTION: 095 (1.00) l

An ALERT has been declared. Which ONE of the following describes the l

REQUIRED emergency notification? I

l

a. The NRC must be notified within 15 minutes after the

declaration of the ALERT. State and local agencies must be 1

'

notified immediately thereafter, not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. State and local agencies must be notified within 15 minutes

after the declaration of the ALERT. The NRC must be notifind

immediately thereafter, not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

c. The NRC must be notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the declaration of

the ALERT. State and local agencies must be notified

immediately thereafter, not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes,

d. State and local agencies must be notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after

the declaration of the ALERT. The NRC must be notified

immediately thereafter, not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes

QUESTION: 096 (1.00)

Which ONE of the following describes the required manning of the Fire

Brigade?

The Fire Brigade shall consist of five members:

a. including the Brigade Leader. Two of these persons may also be

part of the crew required for safe shutdown of the plant.

b. including the Brigade Leader. These persons may not be part of

,

the crew required for safe shutdown of the plant.

c. excluding the Brigade Leader. Two of these persons may also be

part of the crew required for safe shutdown of the plant.

d. excluding the Brigade Leader. These persons may not be part of

the crew required for safe shutdown of the plant.

- - . . . . - ..

1

SENIOR REACTOR OPERATOR P:ga 60

l

OUESTION: 097 (1.00)

l

During an emergency, a reasonable action that departs from Technical l

Specifications must be taken immediately.

1

in accordance with PNPS procedures, which ONE of the following MUST ;

approve taking this action? 1

a. An on shift licensed Reactor Operator and on shift licensed

Senior Reactor Operator i

i

b. A licensed Senior Reactor Operator only

c. A licansed Senior Reactor Operator and the Operations

Department Manager

d. A licensed Senior Reactor Operator and the Operations

Department Manager and the Plant Manager j

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OUESTION: 098 (1.00)

You have worked the foDowing schedule:

- Thursday 1st scheduled day off l

-

Friday 2nd 7 am to 7 pm I

- Saturday 3rd 7 am to 7 pm )

-

Sunday 4th 7 am to 3 pm

-

Monday 5th 7 am tn 3 pm

- Tuesday 6th 7 am to 9 pm

-

Wednesday 7th 7 am to 3 pm

-

Thursday 8th 7 am to ?

Which ONE of the following represents the LATEST you can be required to

work on Thursday the 8th, without special approval being granted?

(Assume turnover time is NOT included)

a.3pm

b. 5pm

c. 7 pm

d. 9 pm

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SENIOR REACTOR OPERATOR Pzga 61

QUESTION: 099 (1.00)

The only individual available for a call-in for TSC staffing informed

the Nuclear Watch Engineer (NWE) on the phone that he has consumed

alcohol within the previous 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Which ONE of the following describes the individuals ability to work in

the TSC? :

a. not permitted to work in the TSC.

b. permitted to work in the TSC provided the individual informs

the NWE that alcohol has not impaired his ability to work in

the TSC. A blood alcohol concentration test is at the NWE

discretion, based on the NWE phone discussions with the ;

individual.

c. permitted to work in the TSC only if a blood alcohol

concentration test is performed upon arrival on site and

the concentration is less than 0.04.

1

d. permitted to work in the TSC only if a breathalyzer l

test is performed upon arrival on site and the blood to

alcohol ratio is greater than or equal to 4.0.

!

l

QUESTION: 100 (1.00)

A procedure is currently being performed which requires the installation l

of a jumper. It is discovered that the procedure directs the jumper I

placement in a position that would cause an unexpected ESF actuation. A

change to the jumoer position is required.

Which ONE of the following is the required method to revise this

procedure to change the jumper position? l

a. Editorial correction

b. SRO Change

c. Minor Revision

d. Major Revision ]

( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * *)

-- -

.. . - - -

,

' BOSTON EDIS0N RTYPE H6.02

.

PILGRIM NUCLEAR POWER ST. TION

Procedure No. 2.1.27

DRYWELL TEMPERATURE INDICATION

i

REQUIRED REVIEWS REVIEWERS AND APPROVERS

hE. A$ oves /NNW 6b7M

"" Writ *" '**

Thi"k '*d""*'

STAR

Act

8)

'wcynical Reviewer

'gg)q,V

Date

Review *# 9 # ^ * u' 9/#6/

Validator '

Date' W

SAFETY REVIEW E0"!9ED/

/i f

Procedure A' ner

'

e/d

/Dgte

NOT REQUIRED

N/+

QAD Man'ager

ORC REVIEW REQUIRED / Date

MT REQ'J:"C0

AA L lo /k 194

-

ORCC{plirman '

Date

0m Ibb smlk fobt9/94

sporpible anager / 1Dhte

Effective Date: /0/d8 9%

020095 2.I.27 Rev. 3

- -. .. .

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.

REVISION LOG

REVISION 3 Date originated 5/94

Paaes Affected Descriotion

4 Add PDC 92-58 to References.

~5,7,10,12,14,15 Revise Kaye nomenclature and delete channel points from old

Kaye recorder in accordance with PDC 92-58.

Editorial 2C Date Originated 3/93

Paaes Affected Descriotion-

7,9,11,13 Delete references to Station Honeywell Computer System as it

is obsolete.

Editorial 2B Date Originated

1

Paoes Affected pescription i

1

4,5,7 Editorial corrections to reflect new E0P numbers and entry l

conditions and to add new Editorial Correction rev bar l

identifications. I

Editorial 2A Date Originated

Paaes Affected Qgscriotion

4,5,7 Incorporated editorial corrections to Main Control Room

Panel Labels per PDC 87-78C.

REVISION 2 Date Originated

Paaes Affected Descriotion

All Reformat to comply with PNPS 1.3.4-1.2.

.

2.1.27 Rev. 3

Page 2 of 15

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I

, .IABLE OF CONTENTS

.

EA9A

1.0 PURPOSE AND SC0PE................................................. 4

2.0 REFERENCES........................................................ 4

3.01 . DEFINITIONS....................................................... 4

4.0 DISCUSSION........................................................ 4

5.0 PRECAUTIONS AND LIMITATIONS....................................... 6

6.0 PREREQUISITES..................................................... -6'

7.0 PROCEDURE......................................................... 7

8.0 ATTAC HME N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

' ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL

TEMPERATURE ESTIMATE..................................... 9

ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL

TEMPERATURE DETERMINATION............................... 10

ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN

T EM PE RATUR E E ST I MAT E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS . INSTRUMENT RUN

TEMPERATURE DETERMINATION............................... 12

ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 13

ATTACHMENT 6 .TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 14

.

2.1.27 Rev. 3

Page 3 of 15

_ _ -_

. . . . _ _.

1.0 - PURPOSE AND SCOPE -

'

This Procedure provides instructions for determining Drywell bulk temperature when

, ,

the Emergency Operating Procedures _(EOPs) require measurement of this parameter.

-

2.0 REFERENCES

,

2.1 DEVELOPMENTAL

[1] PNPS Technical Specifications Table 3.2.H

4

[2] PNPS Technical Specifications Tables 3.2.H and 4.2.H

-

[3] PDC 87-78C, Improvements to Labels, Nameplates on Main Control Room Panels

i

[4] PDC 92-58, Kaye Recorder Replacement !

2.2 IMPLEMENTING

[l] PNPS 2.2.49, " Primary Containment Cooling System"

[2] PNPS 8.7.1.4.2, ' Primary Containment Integrated Leak Rate Test"

3.0 DEFINITIONS

None

4.0 DISCUSSION

[1] The following sections of the Emergency Operating Procedures require

measurement of Drywell temperature:

, (a) E0P-1, RPV Control: RPV Water Level Instrument Run temperatures associated

with the RPV Saturation Temperature Figure of Caution 1.

(b) E0P-2, Failure to Scram: RPV Water Level Instrument Run temperatures

associated with the RPV Saturation Temperature Figure of Caution 1.

(c). E0P-3, Primary Containment Control:

.

(1)' Entry condition _(150"F)

,

(2) Drywell temperature path

(3) RPV Water Level Instrument Run temperatures associated with the RPV

,

Saturation Temperature Figure of Caution 1.

(4)- Figure 5: (SPDS 031) DSIL (Orywell Spray Initiation Limit)

.

2.1.27 Rev. 3

Page 4 of 15

. - . . .-.

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4.0 913GUSSION (Continued)

(d) E0P-4, Secondary Containment Control: RPV Water Level Instrument Run ,

temperatures associated with the RPV Saturation Temperature Figure of

_

.

Caution 1.

(e) E0P-16, RPV Flooding

E0P-26, RPV Flooding, Failure To Scram:

(1) Temperatures near the RPV Water Level Instrument Reference Leg

vertical runs.

[2] Drywell temperature is normally monitored in the Control Room by using

TRU-9044, DRYWELL TEMP / PRESS Recorder, and TI-9019, DW TEMP Indicator, on

Panel C903. TRU-9044 receives its input from a single temperature element

located at a relatively low elevation in the Drywell. TI-9019 receives its

input from a single temperature element located just below the neck of the

Drywell. Both of these temperature elements measure ambient Drywell air space

temperature.

[3] The TE-5050A through P temperature elements are used to evaluate Drywell

tem >erature with respect to Technical Specifications limits (refer to

Technical Specifications Table 3.2.H). The Drywell locations of these

elements are listed in Attachment 5. These elements are used to monitor ,

Drywell temperature for Technical Specifications requirements because of their I

reliability, location, and their redundancy (dual-element RTDs). In addition, I

these temperature elements are the primary elements used for the Primary

Containment Integrated Leak Rate Test.

[4] Local Drywell air temperature indication is supplied by the TE-8125 series I

temperature elements. The TE-8125 series temperature indication consists of

20 RTDs located throughout the Drywell which provide input to the Kaye Temp. 1

Computer (refer to Attachment 6).

[5] When TRU-9044 and TI-9019 are not available, selected Drywell temperature

elements are used to estimate an average temperature near the RPV water level

instrument runs and an average bulk Drywell temperature. Temperatures near

the RPV water level instrument runs are monitored by those thermal elements

which are located in the upper elevations of the Drywell since mest of the

instrument runs are found in this region of the Drywell. Bulk average Drywell '

temperature is a weighted average temperature based on the volume of the

Drywell. By averaging more readings from the lower region of the Drywell

(which contains most of the Drywell air space) than from the upper region of

the Drywell, a representative average Drywell temperature is obtained. More

sophisticated methods to calculate a_ weighted average Drywell temperature are

available, as part of the ILRT Procedure, PNPS 8.7.1.4.2. The method outlined

,

in this Procedure, however, attempts to balance the complexity and time

.

consuming aspects of the sophisticated approach against the requirement to

rapidly obtain a value for Drywell temperature suitable for use in the E0Ps.

i-

2.1.27 Rev. 3

'

Page 5_ of 15

1

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-.

5.0 PRECAUTIONS AfC LIMITATIONS

[1] The Drywell temperature shall be maintained within the following limits when

the reactor coolant temperature is above 212 F.

(a) Above elevation 40': $ 194*F ;

(b) Equal to or below elevation 40': s 150*F

Upon determination that the Drywell temperature at any elevation has exceeded

the above limits, the Drywell temperature at each elevation shall be logged

every 30 minutes. The Drywell temperature shall be reduced to within the

above limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise corrective action shall be as

specified in Technical Specifications Sections 3.2.H.2 and 3.2.H.3.

i (Tech Spec 3.2 H.1)

[2] If the Drywell temperature has exceeded either limit of Technical

Specifications Section 3.2.H.1 for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , an engineering

,

evaluation shall immediately be initiated to assess potential damage and

render a determination of ability of safety related equipment to perform its

intended function.

If either limit of Technical Specifications Section 3.2.H.1 has been exceeded

for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , ferther action to justify continued operation shall

be determined by an engineering evaluation which must be completed within one

week. (Tech Spec 3.2.H.2)

[3] If the requirements of Technical Specifications Section 3.2.H.2 have not been I

met, an orderly shutdown shall be initiated and the reactor shall be in a cold

shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (Tech Spec 3.2.H.4)

[4] If the Drywell temperature at any elevation exceeds 215*F and the temperature

cannot be reduced to below 215 F within 30 minutes, a reactor shutdown shall

be initiated and the reactor shall be in cold shutdown condition within

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (Tech Spec 3.2.H.4)

i

[5] When reactor coolant temperature is above 212*F, the Drywell air temperature '

limits will be determined by reading the instruments listed in Techaical

Specifications Table 3.2.H. These instruments shall be logged once per shift,

and each reading compared to the limits of Technical Specifications Section 3.2.H.1. (Tech Spec 4.2.H.1)

J

6.0 PREREQUISIT_E.ji

.

None

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2.1.27 Rev. 3

Page 6 of 15

l

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1

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-7.0 50CEDURE

. , .

[1] DETERMINE bulk Drywell temperature using one of the following methods (listed

in order of preference):

(a) SELECT the higher of the valves indicated on TI-9019, DW TEMP Indicator,

and TRU-9044, DRYWELL TEMP /F 'SS Recorder (Panel C903).

CAUTION ,

i

j The instruments listed below are not environmentally qualified for use in a harsh '

environment. Under accident conditions, they should only be used if either j

TI-9019 or TRU-9044 is not available for use.

1

(b) Highest probable Drywell temperature from EPIC points DRY 002 or DRY 004.

(c) For a more representative bulk teaperature, AVERAGE the TE-5050 series RTDs

using the computer points in accordance with Attachment 1.

(d) For a more representative bulk temperature, AVERAGE the TE-8125 series RTDs

using the Kaye Temp. Computer in accordance with Attachment .2.

1

(e) All of the TE-5050A through P series RTDs can be read locally at Panel C85, '

Reactor Building El. 23' East, for Attachment I data.

[2] DETERMINE RPV water level instrument run temperature using one of the !

'

following methods (listed in order of preference):

(a) SELECT the higher of the values indicated on TI-9019 and TRV-9044 !

(Panel C903). l

(b) AVERAGE the TE-5050 series RTDs in ecordance with Attachment 3.

(c) AVERAL .he TE-8125 series RTDs using the Kaye Temp. Computer in accordance

with " '.achment 4.

1

(d) The TE-5050A through P RTDs can be read locally at Panel C85, Reactor l

Building El. 23' East, for Attachment 3 data.

[3] Additional information on Drywell temperature elements and location is

contained in PNPS 2.2.49, " Primary Containment Cooling System". 1

1

.

,

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2.1.27 Rev. 3

Page 7 of 15

- . . - _ - .

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8.0 ATTACMENTS

ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE ESTIMATE

ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE. DETERMINATION

ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE ESTIMATE

ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE

DETERMINATION

ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION

ATTACHMENT 6 - TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION

4

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2.1.27 Rev. 3

Page 8 of 15 j

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ATTACHMENT 1

Sheet 1 of 1

TE-5050 TEMPERATURE ELEMENTS

BULK DRYWELL TEMPERATURE ESTIMATE

[1] SELECT one temperature element in each group of temperature elements Alg!

RECORD its temperature.

[2] COMPUTE the average temperature as follows:

(a) Average - (A + B + C + D + E + F)/6

T'.ME

TE-5050 EPIC

GROUP COMPUTER POINT

ELEMENT # TEMPERATURE ( F)

A

DRY 002

A -----OR---- -------------- ------- ------- ------- ------- ------- -------

B DRY 004

i

.

E

DRY 010

B -----OR---- ---- - --- ------- ------- ------- ------- -- ---- -------

G DRY 014

C -----OR---- -------------- ------- ------- ------- ------- ------- -------

H DRY 116

L l

DRY 122 '

D -----0R---- -------------- ------- ------- ------- ------- ------- -------

M DRY 124

K

DRY 120

E -----0R---- -------------- ------- ------- ------- ------- ------- -------

J DRY 118

N

DRY 126

F -----OR---- -------------- ------- ------- ------- ------- ------- -------

p DRY 130

-

AVERAGE

Performed By Date Reviewed By Date

2.1.27 Rev. 3

Page 9 of 15

ATTACHMENT 2

Sheet 1 of 1

TE-8125 TEMPERATURE ELEMENTS

BULK DRYWELL TEMPERATURE DETERMINATION

[1] SELECT one temperature element in each group of temperature elements 88Q

,

RECORD the temperature indicated on the Kaye Temp. Computer.

[2] COMPUTE the average temperature a: follows:

(a) Average - (A + B + C + 0 + E + F)/6

TlME .

TE-8125

GROUP

ELEMENT # TEMPERATURE (*F)

3

A ----0R----- -------- --------- --------- --------- --------- -------- l

l

4 l

l

9 i

8 .... 0R.... ........ ......... ......... ......... ......... ........ 1

10

..

11

C -----OR---- -------- --------- --------- --------- --------- --------

12

13

D -----OR---- -------- --------- --------- --------- --------- --------

14

15

E -----OR---- ----- -- --------- --------- --------- --------- --------

16

17

^

F -----OR---- -------- ------- *- --------- --------- --------- --------

18

AVE: GE

Performed By' Date Reviewed By Date

2.1.27 Rev. 3

Page 10 of 15

ATTACHMENT 3

Sheet 1 of 1

TE-5050 TEMPERATURE ELEMENTS

INSTRUMENT RUN TEMPERATURE ESTIMA1E

[1]. DETERMINE.the rack (s) of concern 8tlQ RECORD the indicated temperature for each

element in that group.

[2] COMPUTE the average temperature.

Instrument Runs for Rack 2205

A Channel Instruments

TLME

!TE-5050 EPIC

ELEMENT COMPUTER POINT

TEMPERATURE ( F)

.

A

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

DRY 002

. .....

Zw . . . ....... ....... ....... ....... .......

j

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

AVERAGE

Instrument Runs for Rack 2206

B Channel Instruments

TLME

TE-5050 EPIC

ELEMENT COMPUTER POINT

TEMPERATURE (*F)

..... ..... ...... . ..... . $b. ....... ....... ....... ....... .......

D DRY 008 2.Ir

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

4.....

E DRY 010 LIO .

AVERAGE

Performed By Date Reviewed By Date

2.1.27 Rev. 3

Page 11 of 15

, -

ATTACHMENT 4

Sheet 1 of 1

1

TE-8125 TEMPERATURE ELEMENTS

INSTRUMENT RUN TEMPERATURE DETERMINATION

[1] DETERMINE the rack (s) of concern A151 RECORD the indicated temperature for each l

element in that group. !

[2] COMPUTE the average temperature.

Instrument Run for Rack ?205

A Channel Instrument

T;:ME

TE-8125

ELEMENT # TEMPERATURE (*F)

4

...........

[b

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

-

10 -,g

AVERAGE

-

Instrument Runs for Rack 2206

B Channel Instrument

T::ME

TE-8125

ELEMENT # TEMPERATURE (*F)

4

...........

2.[h

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

9

Q{[] .

AVERAGE

Performed By Date Reviewed By Date

2.1.27 Rev. 3

Page 12 of 15

-

. .

ATTACHMENT 5

Sheet 1 of 1

.

LTE-5050 TEMPERATURE ELEMENTS LOCATION INFORMATION

,

.

Temperature EPIC Elevation Azimuth Area

Element Point ID (feet) (degrees) Monitored

TE-5050A DRY 002 86 0 2' out from vessel below

an exh. register

TE-50508- DRY 004 89 180 l' out from vessel.

- TE-5050C DRY 006 86 50 4' out from vessel above

supply register

- TE-50E00 DRY 008 90 330 2' below head exh. hole.

TE-5050E DRY 010 60 270 2' out from bio-shield.

'

4 out from bio-shield.

~

-TE~-5050F DRY 012 60 90

TE-5050G DRY 014 40 270 10' out from bio-shield under

Main Steam Line.

TE-5050H DRY 116 40 90 10' out from bio-shield under

Main Steam Line.

TE-5050J DRY 118 35 0 l' from CRD area inside wall.

TE-5050K DRY 120 35 180 l' from CRD area inside wall. l

l

TE-5050L DRY 122 22 205 13' out from CRD area outside !

wall.

TE-5050M DRY 124 22 45 13' out from CRD area outside ,

wall. l

TE-5050N DRY 126 15 270 8' out trom CRD rea nutside

wall.

TE-50500 DRY 128 15 0 On CRD area outside wall.

N

TE-5050P DRY 130 12 125 10' out from CRD area outside

wall.

~

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2.1.27 Rev. 3 i

Page 13 of 15 )

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._ _ _ __ _. . _ . _ _ _ , .

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ATTACHMENT 6

Sheet 1 of 2

4

TE-8125 TEMPERATURE ELEMENTS LOCATION INFORMATION

'

Temperature Elevation Azimuth Area

Element (feet) (dearees) Monitored

TE-8125-1 90 285 Head Exhaust

TE-8125-2 90 210 Head Exhaust

,

TE-8125-3 85 180 3' out from vessel

,

TE-8125-4 85 0 4' out from vessel

above ductwork

TE-8125-5 82 300 In exhaust duct

TE-8125-6 82 45 In exhaust duct

TE-8125-7 80 270 In annulus

TE-8125-8 80 90 In annulus

TE-8125-9 54 270 6' out from !

bio-shield l

'

l

TE-8125-10 54 90 6' out from '

.

bio-shield

TE-8125-ll 40 270 10' out from

, bio-shield under Main

,

Steam Line

TE-8125-12 40 90 10' out from

bio-shield under Main

Steam Line ,

1

l

TE-8125-13 25 315 On CRD area outside

wall

!

TE-8125-14 '25 135 On CR0 area outside

wall

TE-8125-15 19 225- 14' out from CRD

area outside wall

TE-8125-16 19 45 14' out from CRD

m es outside wall.

2.1.27 Rev. 3

Page 14 of 15

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

.

ATTACHMENT 6

Sheet 2 of 2

TE-8125 TEMPERATURE ELEMENTS

LOCATION INFORMATION

Temperature- Elevation Azimuth Area

Element (feet) idearees) Monitored

TE 8125-17~ 14 265 On CR0 area outside wall

TE-8125-18. 14 110 On CRD area outside wall

TE-8125-19 29 180 l' from CRD area

4

inside wall

TE-8125-105 29 0 l' from CR0 area

inside wall

.

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2.1.27 Rev. 3

Page 15 of 15

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SENIOR REACTOR OPERATOR Pegs 1

ANSWER KEY

MULTIPLE CHOICE O23 c

001 b 024 b

002 b 025 d

003 d 026 c

004 c 027 d

005 b 028 c

l

006 a 029 6 at A

007 b 030 c !

1

008 c. 031 a }

s 4

000 a M -

010 a 033 a

011 a 034 b

012 a 035 b

013 c 036 d I

I

014 b 337 c

015 d 038 d

l

016 a 039 d

017 b 040 b

018 c 041 d

019 b 042 c

020 c 043 d

021 d 044 b- ,

- 022 d 045 d

,

il

SENIOR REACTOR OPERATOR Pags 2

ANSWER KEY

.046 c 069 c

'

047 d 070)('C'

048 ' d 071 c-

t

'

049 c 072 c

050 a 073' a

051 c 074 d

052 d 075 d

A-

053 b 076 b

054 b 077 c

1

055 a 078 a l

l

056 a 079 b l

057 d 080 a

g 058 a 081 d

059 d 082 a I

-

060 c 083 d

'

061 a 084 d

062 a 085 a

063 c 086 b

064 d- 087 a

065 d 088 c

066 b 089 b

.

067 a 090 e

'068 d 091 a

>

. ].?

. .

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

<._.

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' SENIOR REACTOR OPERATOR

P:gs 3 >

ANSWER KEY

,

002 : b

- 093 'c'

,

!

094 c

~

. 095 b

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-_ (* " * * * * * * * END OF FXAMINATION * " "' ' * )

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ATTACHMENT 2

Facility Comments on Written Examination

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N

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e 10CFR50.55

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Boeton Edinon

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s.

Pilgrim Nudear Power Station

Rocky Hdi Road

- Plymouth, Massachusetts 02360

L J. Olivier

, Vice President Nuclear Operations

. and Station Director

May 16,1997

BECo Ltr. 2.97-054

4

U.S. Nuclear Regulatory Commission

Region I '

. 475 Allendale Ruad

King of Prussia, PA 19406

Docket No. 50-293 ,

License No. DPR-35

Pilarim's 1997 NRC Written Examination Comments

a

1 The written examination administered on May 5,1997, was considered to be an in-depth examination,

which fairly tested the six (6) SRO candidate's knowledge in the appropriate areas. After thorough

, . analysis of the content of the examination, it is clear that the use of misleading information, use of the

double negative context, and the asking of subjects not important to public health and safety were

avoided.

However, specific requests on several written exam questions are submitted for your consideration in

Enclosure 1. Enclosure 2 contains the reference documentation associated with each of the requests.

'

- Your consideration of these requests is greatly appreciated.

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L. J. Olivier

- PMKINRCEXCO.

Enclosure '

- cc:. See ne'xt page

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cc- Mr. Don Florek

Region 1.

475 Allendale Road .

King of Prussia, PA .19406

Mr. Alan Wang, Project Manager

Project Directorate 1-3

Division of Reactor Projects - 1/11

Mail Stop: 14B2

U. S. Nuclear Regulatory Commission

1 White Flint North

11555 Rockville Pike

Rockville, MD 20852.

'

' U.S. Nuclear Regulatory Commission

Attention: Document Control Desk

Washington, DC 20555

' Senior Resident inspector .,

Pilgrim Nuclear Power Station

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ENCLOSURE 1

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ENCLOSURE 1

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1. Question # 32

While operating at 100% power, it is determined that the Main Steam Lin'e High Flow

switches on the "B" Main Steam Line will t10T trip under a high flow condition.

Which ONE of the following is the MINIMUM REQUIRED action?

a. Direct I&C personnel to manually trip the inoperable switches. I

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b. Direct l&C personnel to manually insert a half Group i isolation on the "B" Group I -

Channel.

c. Initiate an orderly shutdown and be in Cold Shutdown Condition within a

i MAXIMUM of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months after the instrument failure.

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d. Initiate an orderly shutdown and have Main Steam Lines isolated within a

MAXIMUM of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the instrument failure.

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ANSWER: d.

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DISCUSSION: l

The stem of this question states,"... it is determined that the Main Steam Line High Flow

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switches on the "B" Main Steam Line will NOT trip under a high flow condition..."

There are two trip systems associated with Group I PCIS, designated "A" and "B". Trip System

"A" has inputs from MSL High Flow switches comprising two instrument channels, and Trip

System "B" has eight inputs from MSL High Flow switches comprising two instrument channels,

each steam line is equipped with four switches each, one for each instrument channel

(Enclosure 2, Attachment 1, page 1).

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. The stem of the question states that all flow switches on the "B" Main Steam Une are

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inoperable. Since this is the case, there are less than two operable instrument channels for

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both PCIS logic trip systems. (See Enclosure 2, Attachment 1, Page 2)

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Since there are less than the minimum operable instrument channels for both trip systems,.

1 Attachment 1, page 3 states, "If the minimum number of operable instrument channels cannot

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be met for both trip systems, place at least one trip system (with the most inoperable channels)

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in the tripped condition within one hour or initiate thc appropriate action required by Table

3.2.A listed below for the affected trip function."

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Table 3.2.A requires action "B", which states, " Initiate an orderly load reduction and have Main

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. Steam Lines isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ".

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Since there is no grace period (of one hour) for the two trip system inoperability (vice the one

trip system inoperability), there is no obvious correct response.

Page 1 of 10

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REQUEST (Question # 321:

Since the correct response is not offered as a choice in the responses, we request that this ,

question be deleted from the examination.

REFERENCE:

PNPS Technical Specifications, Table 3.2.A and associated notes (Enclosure 2,

Attachment 1).

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! 2. Question # 76 l

The following conditions exist:

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- A steam leak occurs just upstream of the Main Turbine Stop Valves with both )'

j MSIV's in the "A" main steam line failing to close.

- A reactor scram is successful in inserting all rods fully.

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- . Both Main Stack Process Radiation Monitors have been reading 2.5+E4 for the last

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25 minutes.

- Off-site release rate projections are 2 R/ hour Whole Body at the site boundary.

, Select the correct action and its raason.

Under these conditions, the preferred method of depressurizing the RPV is using:

a. SRVs because of the scrubbing potential of the torus water. 'l

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b. SRVs because the heat removal capability is greater than the Main Turbine Bypass i

Valves. l

c. Main Turbine Bypass Valves because the hotwell is the preferred heat sink. .

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d. Main Turbine Bypass Valves because the heat removal capability is greater than

the SRVs.

j ANSWER: b.

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DISCUSSION:

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Because both answer "a" and "b" select the SRVs as the correct mechanism of depressurizing, !

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the question then becomes discriminatory as to the basis for doing so. Appendix B of the

Emergency Procedure Guidelines states that Contingency #2, Emergency RPV

Depressurization may be required to:

Minimize radioactivity release from the RPV to the primary containment and secondary

containment, or to areas extemal to the primary containment and secondary '

containment.

Additionally, Appendix G states that the purpose of the Radioactivity Release guideline is to

limit radioactivity release into areas cutside the primary and secondary containments.

Since distracter "a" implies that SRV's are used because they discharge to the primary

containment, "a" can be construed as the correct answer. That is, given the situation provided,

the fact that the SRVs tiischarge to the containment via the torus is more significant than the

fact that the SRV's heat removal capability is greatedhan the bypass valves.

Since Appendix B also provides generic guidance that SRV's are used because of their heat

. removal capability, "b" is also correct. .

Page 3 of 10

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REQUEST: (Question # 76)

Because both answers "a" and "b" are correct per the EPGs, we request that answers "a" and

"a" both be accepted as correct, and the question be retained in the examination.

REFERENCEi

1. Emergency Procedure Guidelines Appendix B, Section 11, Contingency #2 (OEl l

Document 8390-4B, [ Enclosure 2, Attachment 2]). l

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4, 1

l 3. - Qitestion #'28 )

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, While operating at 100% power, a control rod is determined to be uncoupled. Attempts

to couple the rod have been unsuccessful.

. Which ONE of the following states the MINIMUM REQUIRED actions.

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a. Verify the control rod can be moved with drive pressure and maintain the control rod ;

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at the target position.

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b. 'Fu!!y insert the control rod and hydraulically disarm the CRD.

c. Fully insert the control rod and electrically disarm the directional control valves.- j

d. Fully hsert the control rod, electrically disarm the directional control valves and then

declare the rod inoperable.

ANSWER: c. J

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DISCUSSION:

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The only differentiation between distracter "d" and the correct answer "c" is whether the

rod is declared inoperable. ]

If a control rod was uncoupled, it would be declared inoperable by Technical

Specifications when the inoperability was discovered. The control rod would then be

inserted and electrically disarmed to ensure control rod movement was precluded.

Taking this action does not eliminate the fact that the control rod was inoperable but does

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allow relief from the requirements of the associated Technical Specification actions for an l

uncoupled control rod. The control rod that was uncoupled would still be administratively I

controlled as an inoperable control rod, even though the action statement of Technical

Specification 3.3.F does not have to be applied. At PNPS, if an action has to be taken

on the part of Technical Specifications, the equipment inoperability is traced through the

application of an " Active LCO"in the LCO log.

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from a Tech Spec consideration only, the rod is not inoperable. However from an

administrative and practical standpoint, the rod is indeed inoperable, and the Active LCO

is maintained to control the status of the rod. Therefore, if the cand:date approached the

' question from this perspective, distracter "d" can also be considered as an acceptable

,

answer,

, While Procedure 2.2.87, 5.2.1[3] does state that a rod fully inserted and electrically

L disarmed is not inoperable, it references Tech Spec 3.3.A.2 that concems rods that

cannot be moved with drive pressure. This statement does not apply to the conditions

' identified in the question.

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REQUEST: (Question # 28)

We request that distracter "d" also be accepted as correct.

REFERENCE:

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1. PNPS 1.3.34.2 (See Enclosure 2, Attachment # 3)

. - 3.0[1]" Active LCO" Definition

. ' 4.0 " Discussion"

2. Operations Department Manager (Tom Trepanier, (508) 830-8364)

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4. Question # 50

in the event that torus water level cannot be maintained above 95 inches, HPCI is

secured in order to prevent:

a. exceeding the Primary Containment Pressure Limit.

b. exceeding the Pressure Suppression Pressure,

c. exceeding the Heat Capacity Temperature Limit.

d. isolating HPCI on high exhaust pressure.

ANSWER: a.

DISCUSSION:

The operators at PNPS are provided a " supplemental" approved handout for the study of

the EOP procedures (Enclosure 2, Attachment 4). In this handout, the basis for the 95

inches torus level securing of HPCIis not stressed as the PCPL. The fact the exhaust

will become uncovered is stressed, and HPCI will then directly pressurize the

containment. The wording for the PCPL statement is "may exceed the PCPL", and not

"the basis for the uncovery is the PCPL". Wnen this question is considered, the fact that

the primary containment would pressurize is a valid line of thought. From this direction,

scrutinization of the choices through the use of the supplied EOPs would lead a

candidaa to choose the most limiting curve between the PCPL and the PSP. This would

of course be the PSP curve. Based on this line of reasoning, response "b"is considered

also to be a valid response.

REQUEST: (Question # 50)

We request that distracter"b" also be considered as correct.

REFERENCE:

EOP-03 Supplemental Training Materials / Flow Charts (Enclosure 2, Attachment 4)

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Page 7 of 10

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5. Question # 29

With the plant at power, it is determined that the MO-1001-37 (B loop Torus Spray) and

MO-1400-25A (A Loop Core Spray Inboard Injection) valves have failed their operability

test. Both valves are currently closed.

The maximum time allowed before the plant must be in COLD SHUTDOWN is:

a. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ( 1 day)

b. 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months (4 days)

c. 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months (7 days)

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d. 192 hours0.00222 days 0.0533 hours 3.174603e-4 weeks 7.3056e-5 months (8 days)

ANSWER: b.

DISCUSSION:

PNPS 2.2.125, " Containment isolation System" lists the valves that are considered to be

primary containment isolation valves. An identical listing is contained within the FSAR.

Included in this listing are both the MO-1400-25A and the MO-1001-378 (see

Enclosure 2, Attachment 5). As containment isolation valves, the administrative

requirements require at least one valve in the line to be deactivated in the isolated

position, unless the valve receives any signals other than the isolation signal. Whether

the valve receives any other signals (other than the isolation signal) determines whether

the valve has to be deactivated electrically or otherwise administratively controlled. If the

requirements of this procedure are not met (and the questinn does not provide this

information), an orderly shutdown shall be initiated and the reactor shall be in Cold

Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,

This question was designed to test the applicants' ability to determine:

1) the impact of an Inoperable 378 valve on "LPCl* operability;

2) the impact of an inoperable 378 valve on the Containment Cooling Loop's Operability

and,

3) the overall effect of 1 and 2 when coupled with an inoperable Core Spray system.

At least one applicant, (during a followup interview), interpreted this question as a test of !

his ability to recognize that:

1) Both valves are PCIS valves

2) That at a minimum, the 25A would need to be deactivated since it receives an Auto

' Open signal and,

3) Determine the corrective actions for failed PCIS valves.

Since the questions asks for the maximum time allowed before the phnt must be in

COLD SHUTDOWN, if a candidate were to assume that the question is testing his

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knowledgs of PCIS, thin it is rs: sort:bla that ths candidits would choso "c" es tha

correct response, given that no other actions are taken.

REQUEST: (Question # 29)

,

Due to the two different ways that this question can be Interpreted, we request that both "a"

and "b" be accepted as correct.

REFERENCE:

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PNPS Procedure 2.2.125 (Enclosure 2, Attachment 5)

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6. Question # 27

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'When' valving in a CRD hydraulic control accumulator, the 305-102 (Withdraw Riser . 1

Isolatio'n Valve) and the 305-112 (Scram Discharge Riser Isolation Valve) are required to

be open prior to opening the 305-101 (insert Riser Isolation Valve). This prevents:

af a single rod scram when opening the 305-101 valve.

b. excessive scram time of that rod in the event of a reactor scram,

c. damage to the accumu5 tor in the event of a reactor scram.

d. damage to the drive mechanism in the event of a reactor scram.

ANSWER: d. .

DISCUSSION:

While it is stated in PNPS 2.2.87 that valve misoperation dunng the isolation or

restoration of a HCU can cause " severe damage to the mechanism", the isolation of the

102 (by itself) can also delay control rod insertion following a scram signal. As seen in

Enclosure 2, Attachment 6, with the 102 valve shut, the exhaust path from the

- mechanism is isolated. Since the question does not state the position of the associated

rod for the HCU being restored, the candidate could reasonably assume that the rod is in

a position other than fully inserted. If the exhaust path is isolated, any scram signal will

not permit the mechanism to scram at " normal" rates, if the control rod inserts at all.

REQUEST: (Question # 27)

'

Due to the fact that response "b" contain the phrase " excessive scram time of the rod in

the event of a reactor scram", we request that response "b" be also accepted as a correct

answer.

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REFERENCE:

Figure 4 from PNPS Training Material (Enclosure 2, Attachment 6)

Page 10 of 10

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ATTACHMENT 3

NRC Resolution of Facility Comments

Ques 28 Disagree with BECO comment. The question stem requested " MINIMUM l

REQUIRED actions" and the applicants had Technical specifications. The

question clearly related to interpretation of technical specification-required

actions. As specified in Technical Specification 3.3.A.d, control rod drives

that are fully inserted and electrically disarmeo shall not be considered

inoperable. Therefore answer d is incorrect. There was no change to the

answer key.

Ques 29 Agree with BECO comment. There was insufficient information provided in

the question to rule out consideration of containment isolation system

technical specifications. The answer key was revised to accept a or b as

correct answers.

Ques 32 Agree with BECO comment that there is no correct answer to the question

as written. There was no comment provided to this question during the

preexam review. The question was deleted from the examination. l

Ques 50 Disagree with BECO. The question asks for the reason HPClis secured at a

decreasing torus level of 95 inches. Enabling objective 10 required the

applicant to " state the significance of torus levelless than 95 inches as

regards the HPCI system." The significance, stated in O-RO-03-04-05, Rev

,

1, IG 3 is to prevent exceeding the primary containment pressure limit ?

d

(PCPL). The BECO response to the question is attempting to reword the

-question to determine the first EOP-03 curve limit reached if HPCI exhaust is

not secured at a decreasing torus level of 95 inches. This was not the

,

question asked. There is only one correct answer to the question asked.

While the pressure suppression pressure (PSP) will be exceeded, it has a l

- relatively small consequence. The PCPL will be exceeded, which has a large j

consequence, primary containment failure, and is the stated reason in the l

reference material for securing HPCI at a torus level of 95 '"ches. There was

no change to the answer key.

Ques 7 Disagree with BECO comment. As described in 0-RO-03 04-07, Rev 1, IG 9,

the purpose of performing alternate depressurization under the conditions of

' the question is to reduce the driving head and flow of any primary leak by

rapidly reducing the pressure. In 0-RO-03-04-09, Rev 1, IG 18 the SRVs are

used because the heat removal capability (40% power) is greater than the

main turbine bypass valves and the RPV will be depressurized sooner. Tha

basis for venting containment, when required, using the torus vents

considers the scrubbing potential of the torus water to support the torus

method as the preferred method. Venting of the primary containment was

not required based on the conditions given in the question. Therefore, the

only correct answer to this question was answer b. There was no change in

the answer key.

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ATTACHMENT 4

SIMULATION FACILITY REPORT

Facility License: DPR 35

Facility Docket No: 50 293

Operating Test Administration: May 6-9,1997

1

This form is to be used only to report observations. These observations do not constitute i

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

of a 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

observations.

IIfM DESCRIPTION

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None

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ML20149K131

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