ML18054B560

Exam Rept 50-255/OL-90-01 on 900312-16.Exam Results:One Reactor Operator Failed Written Exam & One Reactor Operator Failed Both Written Exam & Job Performance Measure Plant Walkdown Portion of Operating Exam

ML18054B560
Person / Time
Site:	Palisades
Issue date:	04/09/1990
From:	Burdick T, Lennartz J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18054B558	List: ML18054B557 ML18054B560
References
50-255-OL-90-01, 50-255-OL-90-1, NUDOCS 9004230263
Download: ML18054B560 (92)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGION II I Report No. 50-255/0L-90-0l Docket Nos. 50-255 Licenses No. DPR-20 Licensee: Consumers Power Company 1945 West Parnall Road Jackson, MI 49201 Facility Name:

Palisades Examination Administered At:

Midland Training Center/Palisades Examination Conducted:

Requalification examinations for four Reactor Operators and four Senior Reactor Operators Chief Examiner:

l~~<:...

~.nartz

. Approved By:

{J,fi~v(.,__

TClrilas M. Burdick, Chief Operating Licensing Section 2 Examination Summary Examination administered on March 12 throu h March 16, 1990

t/q/90 Date/

0 l1!r11' Date

• Report No.

0-255 L 1 Areas Inspected: Written and operating requalification examinations were administered to four Reactor Operators and four Senior Reactor Operators.

Crew performance as well as individual operator performance were evaluated on the dynamic simulator portion of the operating examination.

In addition, an evaluation of the licensees requalification program was conducted.

Results:

One Reactor Operator failed the written examination and one Reactor Operator failed both the written examination and Job Performance Measure (JPM) plant walkdown portion of the operating examination.

Four Senior Reactor Operators and two Reactor Operators passed the written and operating examinations.

In addition, both crews received satisfactory evaluations for their performance on the dynamic simulator examinations. These examinations combined with the requalification examinations administered in April 1989 (Report No. 50-255/0L-89-01) provide an adequate sample size to meet the requirements contained in NUREG-1021, "Operator Licensing Examiner Standards,"

ES-601, "Administration of NRC Requalification Program Evaluations".

Based on these results, the requalification program is evaluated as satisfactory.

9004230263 900410 PDR ADOCK 05000255 V

PNU

REPORT DETAILS

1.

Examiners

• J. Lennartz, NRC R. Eaton, NRC J. Hanek, INEL I. Kingsley, Sonalysts G. Weale, Sonalysts

• Chief Examiner

2.

Examination Development

a.

Written Examination The licensee's proposed written examination contained many items that did not meet the requirements of ES-601, and therefore had to be rewritten by the examination team or deleted from the examination.

Many of these deficient items were similar to deficiencies identified during the April 1989 written examinations (Part A and Part B) including:

0 0

Questions not limited to one topic (i.e.

11explain 11 or justify 11

)

resulting in open ended questions.

Multi-part questions required the operator to correctly answer first part of the question in order to correctly answer subsequent parts resulting in double jeopardy type questions.

The following are a few additional written examination bank deficiencies that were identified:

0 0

0 Questions did not contain enough information in the stem of the question to solicit the required answer.

Examination bank contained questions that were no longer valid due to system/equipment modifications.

A few questions were considered to be 11 direct lookup" type questions.

The fact that many of the identified deficiencies of the examination bank had been previously identified by the NRC, is an indication that the licensee did not conduct the needed review and subsequent revision to their written examination bank.

This review must be completed by the licensee prior to the next scheduled requalification examination in order to prevent delays and/or cancellation of the examination due to deficient materials.

2

b.

Dynamic Simulator Scenarios The following are observations made by the NRC concerning the dynamic simulator scenarios validated for use during the requalification examination:

0 0

One scenario, of the four that were selected, could not be used as written because it did not meet the requirements of,

11Simulator Scenario Review Checklist," in ES-601 in that it did not contain at least one time critical crew response.

The scenarios that were validated for use on the examinations did not 11 run 11 as written, (i.e., initial power levels were different, additional malfunctions occurred.that were not written into the scenarios, and some malfunctions did not occur as written). The licensee indicated the problems were due to doing a 11 load save 11 on the simulator software. The licensee should "dry run 11 the scenarios that are selected for the examination just prior to the examination prep week to ensure that the scenarios will run as written. This would reduce the time spent and the resources needed during the examination prep week while conducting the validation process.

c.

Job Performance Measures The following are examples of deficiencies relating to the requirements of Job Performance Measures (JPMs) as contained in Attachment 12, 11Job Performance Measure Quality Checklist", of ES-601.

. 0 0

The JPM questions did not contain the supporting K/A and value

• Some performance standards did not contain complete and correct system response cues such that the examiner could properly cue the operator when applicable.

d.

Sample Plan The following are examples of deficiencies concerning the sampling plan regarding the requirements as contained in ES-601 Attachment 4, "Facility Generated Reference Material Evaluation.

11 0

0 Safety related tasks were not identified Sampling Plan did not identify the percentage of items on Parts A and B of topics covered during the prior two years.

3

3.

Examination.Administration The licensee was responsible for examination administration while the NRC observed the process which allowed the NRC to evaluate the licensee's requalification program as well as the individual operators.

The following observations were made by the NRC concerning examination administration:

0 0

0 0

0 0

The licensee made good use of formal checklists to brief the operators prior to the start of each phase of the examination.

The licensee did a good job in maintaining examination security.

Pencils and scrap paper were not made available to the operators at the start of the Part B written examination, nor did the licensee proctor have enough of these supplies available to give to all of the operators upon request. The NRC proctor had to provide additional supplies.*

Dynamic simulator scenario events were well coordinated between the evaluators and the simulator operator. This was an improvement over the rigid timed schedule of events utilized on previous examinations.

During administration of dynamic JPMs, the licensee should ensure that the simulator setup matches the JPM task conditions, and the evaluators need to better coordinate with the simulator operators to

. provide indications that are as real as possible.

The licensee did a good job to ensure that the operators were not getting in each other's way or using the same procedure during performance of the in-plant JPMs.

4.

Evaluation of Facility Evaluators In addition to evaluating the operator's performance, the NRC evaluated the licensee evaluators' ability to conduct consistent and objective examinations as well as their ability to provide unbiased evaluations of the operators. This was an area which showed significant improvement over the previous examination. There were however, a few instances where the licensee evaluator failed to provide a needed cue to the operator and a couple of instances where the licensee evaluator provided cues that were not asked for by the operator.

In each of the cases, the NRC evaluator successfully counseled the licensee evaluator, pointing out the deficiencies, and asking the licensee evaluator to refrain from such practices.

4

The overall improvement in the ability of the licensee evaluators to conduct examinations indicates that the training program developed for the evaluators has been successful. The licensee should continue to provide training to their designated evaluators in order to ensure that fair, objective, and unbiased evaluations of the operators continue.

5.

Examination Evaluations Coevaluation by the NRC examiners and the licensee evaluators of the operators performance on the examination was performed.

Coevaluations provided the NRC with the necessary information to assess the individual operator's performance as well *as the licensee's requalification program performance.

a.

Operating Examination The overall evaluations on the operating examinations, which consisted of the dynamic simulator examination and the JPM plant walkdowns, for all eight operators, as well as the crew performance evaluation during the dynamic simulator examination for both crews were consistent between the NRC examiners and the licensee evaluators.

The following deficiencies in the operators performance were observed in a majority of the operators examined in each particular knowledge or ability:

0 0

Communications between crew members was poor.

The observed deficiencies included: 1) Shift supervisors failed to keep all crew members informed of plant status; 2) Control Operators (CO) performed manipulations on control boards that the other CO is "normally" responsible for and failed to inform each other of these manipulations; and 3) operators providing communi~ations failed to ensure that the person receiving the information heard and understood it resulting in delayed, "open loop",

and/or missed communications.

The senior operators ability to correctly implement contingency action steps within the EOPs was generally poor. There were many instances when a senior operator encountered an "instruction" step that could not be done, while verifying "immediate actions" step completion in EOP-1, "Standard Post Trip Actions," due to various plant/equipment malfunctions.

The correct action would have been to immediately implement the 11 contingency action" for that particular step that could not be satisfied. There were many instances when an 11 instruction 11 step could not be performed yet the senior operator continued on with the next sequential "instruction" step vice immediately performing the "contingency action 11 step as required. The 5

senior operators would mark the "instruction" steps that could not be met, and would come back to perform the 11 contingency action 11 for these marked steps after all immediate actions were verified completed.

b.

Written Examinations Parallel grading of the written examination by the NRC and the licensee resulted in consistent overall evaluations regarding pass/fail decisions for all eight operators. However, the licensee's graded examination scores were typically three to four percentage points higher with two licensee graded examinations about ten percentage points higher than the NRC parallel graded examinations.

These differences were due to post examination changes made to the answer keys by the licensee that were not accepted by the NRC.

The following are two examples of changes that the licensee made to the Part A written examination (two of two) answer key that the NRC did not concur with:

0 0

Question G16 required the operator to state which primary coolant pumps (PCPs) should be stopped based on plant conditions with a justification. The original answer key required, 110 PCP should be stopped due to high vibration with high seal bleedoff temperatures.

11 However, due to the simulator setup, all four PCPs had high control bleedoff temperatures indicated, which would require that all PCPs be secured.

The licensee accepted either the original answer, as per the key, or a revised answer of 11 all PCPs should be secured due to high control bleedoff temperatures 11 for full credit (1.0 pts.) The NRC allowed full credit for the revised answer only, and gave partial credit (0.25 pts) for the original answer key response.

Question G78 required the operator to choose the correct answer concerning the response of the pressurizer (PZR) pressure control system for the indicated system configuration. With the given system configuration, the correct answer was choice 11 b 11, (reactor wi 11 trip on high PZR pressure). The licensee also accepted choice 11 d" (reactor will not trip and PZR pressure will remain constant) due to simulator indications of steady PZR pressure.

However, with no further operator actions (as stated in the question stem) and the given system configuration, PZR pressure would eventually increase to the high PZR pressure trip setpoint, and therefore choice 11 b 11 is the only correct answer which was accepted by the NRC.

The licensee must ensure that they fully evaluate proposed answer key changes to ensure that full credit is awarded only for completely correct answers.

Objective and unbiased changes to the original answer key will ensure that parallel grading by the NRC examiners and licensee evaluators will result in consistent scores and overall results.

6

6.

Program Evaluation The NRC administered requalification examinations covered in this report, combined with the NRC administered requalification examinations in April 1989 (Report No. 50-255/0L-89-01) provide an adequate sample size to meet the requirements contained in NUREG-1021, "Operator Licensing Examiner Standards," ES-601, "Administration of NRG Requalification Program Evaluations," regarding requalification program evaluations.

Based on the combined results of the requalification examinations, the licensees requalification program is evaluated as satisfactory.

7.

Examiner Concerns Licensee representatives that were to be dedicated to the examination team during the examination development phase, were also required to participate in vendor presentations and INPO audits. These collateral responsibilities significantly reduced the time available for examination development which was evidenced by the poor quality regarding "attention to detail" in the revis~d written examination that was submitted to the NRC for approval. A few examples of this poor quality include:

0 0

0 0

Two questions had no point value indicated.

Two questions had no reference indicated.

One question. had no answer *on the answer key.

Exam summary sheet did not match the actual exam in that it contained questions that had been deleted by the examination team during the examination prep week and failed to contain the information pertaining to the replacement questions.

In addition, the NRC identified deficiencies, concerning the written examination question bank, similar to those identified during the April 1989 requalification examinations, (see Report No. 50-255/0L-89-0l).

These deficiencies seem to indicate either a lack of staffing availability or a lack of commitment on the part of the licensee necessary to ensure that the requalification program is successful.

The written examination question bank will be affected anytime plant/system modifications and/or procedure revisions are made, and therefore must be considered as a living document.

The licensee must ensure that the necessary resources are available to make improvements to and provide maintenance of the examination bank in order to prevent delays and/or cancellations of future requalification examinations due to deficient materials.

8.

Exit Meeting An exit meeting was held on March 3, 1990 between the NRG and facility management to discuss the requalification program and operator deficiencies and strengths, as well as the NRG concerns.

7

I NRC representatives in attendance were:

J. Lennartz, Examiner I. Kingsley, Examiner G. Weale, Examiner R. Eaton, Senior Reactor Engineer, OLB E. Swanson, Senior Resident Inspector Licensee Representatives in attendance were:

R. Rice, Operations Manager J. Hanson, Operations Superintendant D. Rogers, Training Administrator T. Hagan, Director of Nuclear Training, General Offices R. Frigo, Operations Staff Support Supervisor D. Vandewalle, Plant Technical Director D. Armstrong, Senior Nuclear Instructor, Simulator P. Schmidt, SeQior Nuclear Instructor R. Tucker, Operations Support Coordinator The facility management acknowledged the examiners observations discussed in Section 2-7 of this report.

8

Facility: Palisades Examiners:

Lennartz, Eaton, Hanek, Kingsley, Weale Dates of Evaluation: April 17-21, 1989; March 12-16, 1990 Areas Evaluated:

X Written x

Oral x

Simulator Examination Results: Total for April 1989 and March 1990.

RO SRO Pass/Fail Pass/Fail Written Examination 6/2 8/0 Operating Examination Oral 7 /1 8/0 Simulator 8/0 8/0 Evaluation of facility written examination grading Crew Examination Results:

Crew 1 Pass/Fail Operating Examination 4/0 Crew 2 Pass/Fail 4/0 Overall Program Evaluation Satisfactory Crew 3.

Pass/Fail 4/0 Total Evaluation Pass/Fail (S or U) 14/2 s

15/l s

16/0 s

s Crew 4 Evaluation Pass/Fail (S or U) 4/0 s

The above program evaluation is based on the combined results of the April 1989 requalification examinations (Report No. 50-255/0L-89-0l) and the results of the requalification examination covered in this report.

Submitted:

Chief

/

SIMULATION FACILITY REPORT Facility Licensee:

Palisades Facility Licensee Docket No. 50-255 Operating Tests Administered At:

Midland Training Center During the conduct of the simulator portion of the operating tests, the following items were observed (if none, so state):

ITEM

1)

Malfunction ED36, "Loss of 125 VDC Panel llA"

2)

Miscellaneous Initial Condition (IC) setups DESCRIPTION Annunciator panel K-05 window 52, "D/G 1-1 START SIGNAL BLOCKED," failed to energize when malfunction ED 36A was inserted.

This annunciator should have energized for the given malfunction.

The synchroscope circuit on panel C-04 cannot be turned on if the backfeed ITC switch is in the "Transfer Cutout" position vice the normal position for some ICs.

The synchroscope circuit should always be able to be turned on during these conditions.

PERSONAL ANO CONFIDENTIAL ff\\c:l ::ifc,' C'cp-(j E'>l..c:.111

~ 0-b -L-Ro }.S(2o

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RO/SRO STATIC SIMULATOR EXAMINATION COVER SHEET NAME SOC SEC NUMBER --~~~~-

WORK LOCATION Palisades Plant DEPARTMENT -~O~p~e=r=a~t=i=o=n=s~~-

CLASS NAME EXAM ~2~ OF 2

(ANSWER KEY)

Static Exam #1B Scenario.#J APPROVAL

~ ~

3/,¥('21.

=--====--======================--=:=======

DATE ADMINISTERED --~---

ADMINISTERED BY GRADED BY GRADE-----

==========================================--===========

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

Signature

=====================================================

I was given the opportunity to review the correct responses to this examination.

Signature

SCENARIO #J MISC. EQUIP. FAILURES SETUP: IC-10 SETUP CHECKLIST:

MALF:

1. Advance recorders prior to start of scenario

2. Place recorders in "fast speed" for those so equipped

3. Ensure all recorder pens inking properly

~-

4. Turn off all chart drives at "freeze point"; leave "Recorder Power ON"

5. Complete attached Scenario Data Sheet

6. Calculate answers for appropriate questions on answer key.

l,EDOSC,,l 2,CCOJ,,1,,100 3,RC160,,1 4,RX06A Loss of preferred A/C bus Y-30 ccw system leak (lOOgpm)

P-500 high vibration Pzr. PT-OlOlA xmtr fail low I/O OVRD 1,LI-1110,,15 2,LIA-1400,,57 3,LIA-1416,,60 4,MG-VARS,,65 5,PI-1419,,50 TURNOVER: 1.

PCS boron =.686 ppmB

2.

Plant Age = 3 GWD, Equilibrium Xe

3.

First Out alarm= EK-0544 (Preferred AC bus' No. 3 Trouble

INSTRUCTOR SPECIFIC SETUP Scenario #J (Total Run Time= 15 mins.)

1.

2.

3.

4.

5.

Bring out of freeze Let run for 5 mins. to develop pre-scenario traces Insert Remote # 1 Freeze simulator just after PCP "Vibration Danger" alarm come comes in.

~-

Perform the following prior to freeze:

a.

b.

c.

d.

e.

f.

g.

h.

i.

j.

k.

1.

Bypass all RPS trip units on "C" channel eteeept VHP'.F f;:.. 1.,..,

Using ED2 Remote Function 20, adjust grid voltage (C-01 panel indication) to 360 KV

--A.djust:-~e-DG-M+/--H.-i-amp-Meter-to-+~us:i:nq-the-Ae-Adjuster--ff;.,/'/...

Place P-55B control select switch on C-12 to the MANUAL position.

Display a gualif ied CET reading on the PIP Provide a PIP printoutd""of incore thermocouples Ensure that PRC-0101;~is selected for this scenario.

Set West ESS Room Sump pumps to OFF Ensure TIC-0203 (L/D Temp. Controller) set at 110 °F Ma~ch the DEMAND signal of l/LRC-OlOlA to that of 1/LRC-OlOlB

~lz.../.,,,

Ensure that V-14A i~~s~fected for standby Ensure that the *t:[.'Cf't1n SMM is displayed at time of freeze. (Note: It will be blinking for this scenario)

SCENARIO #J (cont.)

QUESTIONS:

Gl6. EK-0913 alarm came in one hour ago.

Based on plant indications, which PCP' s, (if any),

are required to be stopped. Justify your answer. If no pumps should be stopped, answer NONE. ( 1. 0 pts. )

,:Jcc:c*1 1r

. ")/L**h"

(;. (.~

Ans : m: OD should be stopped. (o. § pt&.-> Justification:

P-5 OD l"'-*-n:... J vibration increase of > 2 mils in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> corroborated by C:'c-s1:J*

high seal bleedoff temperatures. (or P-SOD secured due to ARP-

"'!;j,~ts*J

_uida_ncr,e) ~*§cl pt~

..,,-.... 'o.:.-J "'!;.:'),..

<-*~...,....,

t-Ji~.....,11.y,.,,1*i;;.

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1.,_.... "

t:1;..,;.-*

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~

v,

• ti;... /,***b. (*..:./

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""*I**~.. 1 ("(';.,.,.,

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0

• J I).:,

.:;.., *.,... " -* \\i

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~* '

,~

Ref: ARP 5 (K-09) ~"'"',.."'~ -kr*'f:i *

~:µ...r<> (/TIME: 4. 5 M NS.

1

"./

G22. Determine the control rod insertion limit (in inches withdrawn)

(per Tech.

Specs.)

for the present plant conditions. (use the highest indicated TMM nuclear power) (1.5 pts.)

Ans. Find the highest indicated TMM nuclear power from the scenario data sheet (pt. #39); using Fig. 1.9 from Tech. Data Book, determine T.S. PDIL for the existing power level. (Acceptable answers should be within +/- 4 inches of above calculated answer~)

r-o..u @

,, G )

t;\\> *.

\\rt*

JOO r

~

f.~~

Ref. Tech. Data Book, Fi9. 1.9

.~ul'J**

TIME: 3.5 MINS.

G26. Assuming PCS temperature remains constant at it's present value, using the most conservative value from the following indicators, (CET printout,* PTR-0122) what is the minimum pressurizer pressure at which 25 degrees subcooling can be maintained? (2.0 pts.)

J ~

c.,~C~I i"'~r *1a...*, *.,

Ans: Using attached data sheet/'find thi~highest PCS temp. from J.r/

1. 21 and #22. Locate this PCS temp. on operator aid (OA-88)or cc:P c.tb.:*

(EOP P/T Limits Curves). Follow temp. line straight up to the

1. :s;-

25 degree subcooled curve and draw a line over to left axis

'j/L"/1.-

to find the corresponding PCS press.

Acceptable Answers should be within +/- so psia of instructor calculated press.

(

f),.* ':).,_, *.. ~

'(,,,,, * ' r :_,,.., ~:- 'JC',., ~.,._ J:c13.. 2,. o)

EOP 7.0, Attach. 3 TIME: 4.0 MINS.

Ref:

G57. Are the charging pumps properly aligned in accordance with plant operating procedures? Explain your answer. (l.O pts.)

Ans: No, (0.25 pts.) P-SSB control select switch is in the MANUAL position (whereas it should be in the AUTO position.) (0.75 pts.)

Ref: SOP-2A TIME: 1.0 MINS.

SCENARIO #J (cont.)

QUESTIONS:

G63. Given present conditions, which one of the following is correct concerning annunciator window K-07-03 (Letdown Ht Ex Tube outlet Hi Temp.)? (l.O pts.)

a.

Annunciator K-07-03 is invalid

b.

TIC-0203 controller is improperly set

c.

Automatic functions of annunciator K-07-03 occurred as expected.

d.

Annunciator K-07-03 js alarming due to loss of ccw to Letdown Hx.

Ans: d. Annunciator K-07-03 is.due to loss of ccw to Letdown Hx.

Ref: ARP-4 (K-07)

TIME: 4.0 MINS.

G64. Maintenance request a tagout of both isophase bus cooler fans at this time. Is this tagout allowed under existing plant conditions? Justify your answer. (1.0 pts.)

Ans: No, (0.25 pts.)since generator output exceeds 14,400 amperes per phase. - (0.75 pts.)

Ref: SOP-8, Sect. 4.0 TIME: 4.0 MINS.

G78. With the plant operating under the present system configuration, which one of the following is the correct response of the Pzr. Press. Control Sys. (PPCS) assuming no further operator actions are taken? (1.0 pts.)

a.

Reactor will trip on TM/LP -

b.

Reactor will trip on High Pzr. Press.

c.

Reactor will not trip and Pzr. press. will increase to Code Safety Valve setpoint.

d.

Reactor will not trip and Pzr. press. will remain constant Ans. b.

Reactor will trip on High Pzr. Press.

Ref: SOP-1 TIME: 4. 0 MINS.

G79. Given a loss of Y-30, and present indications, what additional action (if any), is required to properly align the RPS? (0.5 pts.)

Ans. Trip "A" Channel TM/LP trip bistable Ref. ONP 24.J--\\

TIME: 4.0 MINS.

~

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c,*c.~. t

SCENARIO #J (cont.)

QUESTIONS:

Jl.

Is containment pts.)

isolation required? Justify your answer. (1.0

{.'. L Ans: No, (e. 25;f.ts.) there is

(~.75 p-~)

Ref: ONP 24.3, Attach. 1 G;*.-1en ')

no valid CHP"°or CHR condition present-1.<tpr.s)

TIME: 5.0 MINS.

J3.

Based on current indications, what is the CIS initiation logic for CHR? (l.O pts.)

a.

1/2

b.

1/3

c.

2/3

d.

2/4 Ans. b. CHR logic: 1/3 logic for left & right channel isolation Ref: ONP 24.3, Attach. 1 TIME: 2.0 MINS.

J7.

What is the maximum allowed MW output based on existing indicated generator parameters? (2.0 pts.)

Ans. Using SOP-8 Attachment #11, find 300 MVARS out on left axis; follow line across to the right to a point equal to 50 psig Hydrc:>gen p~ess. (interpolate b~tween 45-60 P~).glJf*~~hen draw straight line down to bottom axis and readcif49~ + io MW.

7'.)i:)

Ref: SOP 8, Attachment 11 TIME: 5.0 MINS.

Jl4. With the present plant conditions, IF* a total loss of offsite power were to occur, what operator actions (if any) would be required to restore power to Bus lC? (1.0 pts.)

d--°Jlw/"

Ans. Manually see~e (675-pts.) amt close D/G 1-1 output Breaker (brkr. #152-107)

(~t-s

.... ) 1.crrs*

.y-Ref

ONP 24.3 TIME: 4.0 MINS.

STATIC SIMULATOR QUESTION -

CROSS REFERENCE QUESTION #

OBJ #

TASK#

SCENARIO GUIDE Gl6 SIM11930 000 137 05 01 SGEOP 9.0 G22 SIMOG107 RELATED TO NUMEROUS TASKS G26 SIM00064 000 007 05 01 SGEOP 1.0 SIM00016 000 008 05 05 SGEOP 4.0,4.0C SIM34409 344 011 05 03 SGEOP 2.0,4.0C SIM34409 344 o~ 05 03 SGEOP 2.0,4.0C G57 SIM11923 011 001 01 01 SGEOP 1.0,6.0A G63 SIM00430 004 015 01 01 SGCOT-13 G64 SIM00003 045 002 01 01 SGCOT-11 G78 G79 Jl SIM11938 000 007 05 01 SGEOP 1.0 SIM00040 000 136 05 01 SGEOP 9.0D J3 SIM00040 000 136 04 01 SGEOP 9.0D J7 SIM05201 045 005 01 01 SGCOT-12 SIM00003 045 005 01 01 SGCOT-li Jl4

Exam tlB (SCENARIO fJ)

ITEM I. D.

K/A GROUP POINTS TIME (in min.)

G16.

EOP/ONP I 1.0 4.5 G22.

Sys. I 1.5 3.5 G26.

Generics 2.0 4.0 G57.

Sys. II 1.0 1.0 G63.

Sys. I 1.0 4.0 G64.

Sys. III 1.0 4.0 G78.

Sys. II L(J 4.0 G79.

EOP/ONP III 0.5 4.0 Jl..

EOP/ONP II 1.0 5.0 J3.

EOP/ONP I 1.0 2.0 J7.

Sys. III 2.0 5.0 Jl4.

EOP/ONP I 1.0

4.0 TOTALS

14.0 45.0 DISTRIBUTION:

Systems EOP/ONP Generics GROUP I:

2(18%)

3(27%)

1(9%)

GROUP II:

2(18%)

1(9%)

GROUP III:

2(18%)

1(9%)

PERSONAL AND CONFIDENTIAL fflo, ::i4u-Cvf(}

E-u..._,,...

I "'J "-

f.Z o) ~~o r i-. A RO/SRO STATIC SIMULATOR EXAMINATION COVER SHEET CLASS NAME CLASS NUMBER NAME SOC SEC NUMBER ~~~~~~~-

COMPANY CPCo WORK LOCATION Palisades Plant DEPARTMENT

~

Operations EIS/NUTREC EXAM ~l~ OF ~2~ (ANSWER KEY)

Static Exam #lA Scenario #C APPROVAL

~

~

J/-'~h.

=====================================================

DATE ADMINISTERED ADMINISTERED BY DATE GRADED GRADED BY GRADE

=====================================================

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

signature

=====================================================

I was given the opportunity to review the correct responses to this examination.

Signature

SCENARIO #C SETUP:

IC16 SETUP CHECKLIST:

MALF:

1. Advance recorders prior to start of scenario

2. Place recorders in "fast speed" for those so equipped

3. Ensure all recorder pens inking properly

4. Turn off all chart drives at "freeze point"; leave "Recorder Power ON"

5. Complete attached Scenario Data Sheet

6. Calculate answers for appropriate questions on answer key.

l,SGOlB,~l,,l00,:05 2,RM07A,,l,,100,:01 3,MS02B, I 2 B S/G Tube Leak @ 575 gpm RIA-0631 ramp to full scale Closure of B S/G MSIV I/O OVRD 1,LI-8946,,40 2,LIA-2020,,90 3, LIA-2 0 21,, 9 5 4,LIA-2022,,95 5,CV-0867-G,,OFF 6,CV-0867-R,,ON 7,M0-2140-g,,ON 8,M0-2140-R,,OFF TURNOVER: 1.

2.

3.

4.

5.

6.

7.

Plant Pre-trip operation conditions: 76% power, core Age of 6.4 GWD, PCS boron of 465 ppm.

First out alarm "Chg. pumps seal clg. lo press."

a SGTR occurred on B S/G Power reduction to approx. 70% at which time plant was manually tripped due to low pressurizer level (per ONP 23.2)

PCS manually depressurized to present indicated pressure for approx. 2 mins. prior to freeze EOP-1.0 Immediate and Contingency actions were completed at.time of trip with the exception of starting an AFW pump Approx. 4 mins. has elapsed since pla~t trip

INSTRUCTOR SPECIFIC SETUP Scenario #C (Run Time= 15 mins.)

1.

Ensure all Malfunctions are properly entered, then bring out of freeze. Advance all charts to show steady state traces.

2.

Insert Remote #1

3.

At 1 min 30 secs., start reducing Rx./turbine power by using full control rod insertion and matching Tave/Tref by use of the valve limiter until Pzr lev,,el drops to 35%, at which time manually trip the Rx.

4.

Trip 1 MFW pump immediately after Rx. trip.

5.

Trip the 2nd MFW pump at 535 degrees F Tave as indicated on selected Tave/Tref recorder.

6.

start a manual.depressurization immediately after Rx. trip to 1150 psia (+/- 50 psia) using max. pressurizer spray from the two spray cvs. Stabilize for 2 mins.

7o Trip A & D PCPs at 1300 psia.

8.

Allow AFAS to start AFW pump (s).

9.

Trip one or both Condensate pumps when (and if) their high temp. alarms come in.

10.

Perform the following just prior to freeze:

a.

Close CV-0767, CV-0771 (A S/G bottom blowdown valves)

b.

Ensure CV-0703 is partially open (B S/G FRV)

c.

Ensure CV-0768 & CV-0770 are open (B S/G bottom blowdown valves)

d.

Ensure all other S/G isolation valves are CLOSED for the "B" S/G.

e.

Insert Remote # 2 (to close CV-0501)

f.

Ensure V-68B and V-70B selected for standby

g.

Reenergize E bus

11.

Freeze when Pzr. pressure is at 1150 psia for 2 mins.

NOTE: If question Cl is used on exam MUST ensure that "Off Gas Monitor" is in alarm condition.

SCENARIO #C (cont.)

QUESTIONS:

Gl.

Based on current simulator indications, determine the required Shutdown Boron concentration. (l.O pts.)

Ans: Determine core age and number of PCP's running; identify if any control rods stuck out; use Tech. Data Book to find rqd.

boron and ADD 225ppmB for each stuck rod (if any). Answers should be within +/- 20ppmB of calculated ans..

Ac'-t'ffr-bl~-

~~nJ...,.:**5: liS:D*; '81.:>-rf*"(.r'J.ff'*"') or 0.:D:: c/:lOff*'Q::.tvpf") or '8'jopp*~:.z..,

Ref: Tech. Data Book, Fig. 1.2 TIME: 2.0 MINS.

b't.>c7d "").

G3.

C:f"I-O't*-Z..'!f c;,-..i<._._,1.. 11:.~

Based on present indications, which Emergency Classification~

should be declared? (SRO only) ( 1. O pts.)

"Ji'l--/'1-.;

Ans: At least ALERT (based on PCS Integrity). Site Area Emergency is also acceptable (conservative interpretation of PCS Integrity or Miscellaneous category) f Ref: EI-1; "PCS Integrity" or "Secondary Side" TIME: 6.0 MINS.

GlO. Under current conditions, what is the most practical means used to lower PCS pressure. (0.5 pts.)

Ans: Normal Pzr. spray (from the discharge of B & c PCPs.)

Ref: EOP 5.0, Sect. 4.0 TIME: 1.0 MINS.

Gl4. Using available indications, determine the minimum PCS cooldown rate to soc entry conditions. (Note: Use PTR-0112 for Tc value) (P-936 is available) (T-90 is not* available) (1.5 pts.)

Ans: 10 degrees F/hr +/- 1.0 degrees F/hr Re : EOP 4.0, Att. 9.0; EOP 5.0, Att. 7.0; EOP 7.0, Att. 4.0 TIME: 9.0 MINS.

Gl9. Given present conditions, identify ALL available boration flowpaths to the PCS. (1.0 pts.)

i.::---("'.or 1-e:" "I:

Ans: SIRW tank to suction of HPSI pumps ( O. 5 pts.) and Cone.

B/A~"'",:L;ti*. 1:*

to suction of Charging pumps via Gravity Feed valves (Gravityb-'11 -~c;;!:/$"

Feed Emergency Boration) (0.5 pts.)

. Ac"'-<"F C1j c.~J;f,.l<-"!'1! C;:.-_'"_!c.:,+

r-<~-'-?:I'>

R f Op t

11 C~!>w*.n"-

.;;f'.

1,**-,.**t.i...-

"l-Cf>.,,~ f<\\k:l;,1.

e : s

-2A, Sec. 7.5 D S; R~ ~

....__ -k.> ck. \\'J*"'f ~-cf>._.,,.,

rJ T-l..

=.

1J..J~oo'i (!:.:Z.~o"J) (0.2.}

T-'07--::

JOS,11' (~ 6000(5..._1) (c.1..)

~- pJ-y:> *'1rv

[N-.!tl.'">71... !t: tif'-\\r-t-'ZI *.-,

~

3/"l... -j~,

T-81

"'7~; u<J,, (t:. 1-..)-v.:ilfJ) ( o. ~)

'L

'Sen;,*, i.,J~

1~.-.t :- t? '"""Jr.I

,,.,,c".,, -NJ' = ~ ~or.

.* o ~-.I oc.*.::> ":JO\\ V ( 0. )

~~._.1 1-k~-t-r r""""'"* I

,:,,c!,,""1 I-

'l.:l-~1 '.L"' 1.J= rJ, "~J*'~~oco. l)

'...,brol".:,I t=".:J-.i.-t"*..... t.!",,_.,.\\.'.-.1..... _'?,'1./ :i... 7-*:*J.... ~r.'\\.1'~.J\\

J SCENARIO #C (cont.)

QUESTIONS:

G25. Assuming PCS temperature remains constant at it's present value, using the highest indicated Tc value, to what pressure (psia) can the PCS be lowered before PCP operation is inhibited. (1.0 pts.)

Ans: Using the highest indicated value of Tc from data sheet pt.

1,_;i..11 ;2..1..

-~ locate this temp. on the OA-88 (Pressure & Temp. Limits

~ -.J/Z*/'l" curve). Follow the specific vertical temp. line upwards until it intersects the "Minimum Pressure for PCP Operation" curve.

Then proceed from this pt. of intersection to the left vertical axis to determine the min.

pressure for PCP operation. (Acceptable answers are within+/- 50 psia.)

A11~...... - :

i 1 ~o r

-:,"l) p;;,;....

'!ti-3J1'-)'>;,

Ref: EOP 5.0, Attach. 3.0; EOP 6.0, Attach. 6.0 TIME: 2.5 MINS.

G43.

Ans:

With the aid of the Safety Function Status Checksheet from the appropriate Emergency Procedure, identify any acceptance criteria for any safety function that is not being met.

~2.0 pts.)

Using the attached scenario data sheet and EOP 5.0 Att. 1, the instructor should determine if all acceptance criteria are being met. Partial credit for this question is determined by equally weighing the number of appropriate answers in straight percentage. terms...

') G t

)

~1

(' t-i.* c.

11..-*. :*t-De._n,:,.., 1j *.

.U:.:'

l*'\\C*.;o.*-rtc 1<":>1A',.IX.. -

5.J'Ol-~1"t v. '- DG nn~11.,,_ *.s 11..'c;.*,1 e '- 1'-~

-s::..

w T

• EOP 5.0, Attach. 1 TIME: 7.5 MINS.

G54. Would the SI pump minimum flow stop valves (CV-3056 & cv-3027) close if the SIRWT level dropped to the RAS auto setpoint? Justify your answer. (l.O pts.)

Ans: No, (0.25 pts.) since HS-A on both valves is in the OPEN position. (or HS-A is.not in t~e CLOSE position) (0.75 pts) p\\,*c,"->-L=>) v.:11.;(.s,,_,fq re,*,,,-,,.

r?**,-,.,,_,..;.,<1 ~,.~..,_,,,,

~*-;,:) (?,~ 1/~l,;i:.; f7l.v->r,!,;:_

M-1,...x,J/

t!/v..;id (e>.z.;

Ref: EOP 9. 0, IC-2 TIME: 2. O MINS.

~.'"""" th-~ ~J,...

t:l-"r.

7.,

)

\\1'*11,,.,-, 16,.,, cC*,*, c.._,..,T

J.

C3.

If you allow the affected S/G to go solid to *~i.~!Jiilf~-i;°a'dwaste 31 t*Ji.:1 discharges to Misc. Waste, what actions must be taken to prevent challenging the S/G Code Safeties? (1.0 pts.)

Ans: Cooldown, /'.and depressurize the PCS to < 1000 psia.( O,& *. -p:a) ( S**b*"J. ~.);..:.

c~, L~)

~~u~,

<:;,t.}y. h..,;-)

Ref: EOP 5.0, Sect. 4.0 TIME: 5.0 MINS.

M-

"J11.<.j1 /

SCENARIO #C (cont.)

QUESTIONS:

C4.

Assuming the primary to secondary leak is < the capacity of the S/G blowdown system, what is the preferred method of cooling the affected S/G: (l.O pts.)

a. S/G feed and bleed to the clean waste receiver tanks

b. S/G feed and bleed to the misc. waste hold tanks

c. monitored steaming of the affected S/G

d. S/G feed and bleed to the filtered waste monitor tanks

~-

Ans: b Ref: EOP 5.0, Sect. 4.0 TIME: 5.0 MINS.

Cl4. Assuming that the present cooldown rate would remain constant for the next 5 minutes if nothing were changed, what effect would reducing PCS pressure by 100 psia have on the cooldown rate for the next 5 minutes, and why? Assume that this is the only action taken. (l.O pts.).

a.

The cool down rate would increase due to increased SI flow.

at lower PCS pressure.

b.*

The cooldown rate would decrease due to decreased loss of coolant via the ruptured steam generator tube.

c.

The cooldown rate would increase due to increased steam flow through the TBV at lower PCS pressure.

d.

The cooldown rate would decrease due to decreased specific enthalpy of water at lower pressure.

Ans.

a

REFERENCE:

EOP 5.0 Basis TIME: 4 MINS.

STATIC SIMULATOR QUESTION -

CROSS REFERENCE QUESTION #

OBJ #

TASK#

SCENARIO GUIDE Gl SIM 000428*

004 008 01 01 SGONP 23.1,23.2 G3 SIM 34403 344 019 05 03 SGEOP 1. 0 GlO SIM 01001 000 106 05 01 SGEOP 8.0,8.0A G14 SIM00163 001 102 01 03 SGEOP 1.0 SIM00214 002 202 00 01 SGEOP 4.0 SIM00216 002 202 01 03 SGEOP 4.0 G19 SIM00090 000 024 05 01 SGEOP 1. O,ONP 2.1 G25 SIM11930 000 007 05 01 SGEOP 1. 0 SIM11947 000 007 05 01 SGEOP 1.0 G43 SIM00064 000 038 05 01 SGEOP 5.0 G54 SIM00086 000 137 05 01 SGEOP 9.0B SIM34401 344 024 05 03 SGEOP 9.0B C3 SIM00059 000 038 05 01 SGEOP 5.0 SIM11961 C4 SIM00059 000 038 05 01 SGEOP 5.0 C14 SIM00059 000 038 05 01 SGEOP 5.0 SIM11961 000 038 05 01 SGEOP 5.0

Exam f lA (SCENARIO #C)

ITEM I. D.

K/A GROUP POINTS TIME (in min.)

Gl.

Sys.. I 1.0 2.0 GJ.

Generics

1. 0 6.0 GlO.

EOP/ONP II 0.5

1. 0 Gl4.

Sys. I 1.5 ~-

9.0 Gl9.

EOP/ONP I

1. 0 2.0 G25.

EOP/ONP II 1.0 2.5 G43.

EOP/ONP II 2.0 7.5 G54.

EOP/ONP I 1.0 2.0 CJ.

EOP/ONP II 1.0 5.0 C4.

EOP/ONP II

1. 0 5.0 Cl4.

EOP/ONP II 1.0

4.0 TOTALS

12.0 44.0 DISTRIBUTION:

Systems EOP/ONP Generics GROUP I:

2(18%)

2(18%)

1(9%)

GROUP II:

0 6 ( 54%)

GROUP III:

0 0

PERSONAL AND CONFIDENTIAL EXAMINATION COVER SHEET Name Social Security Number Company Work Location Department Course Class No.

EIS/NUTREC Exam No of

~p~r~v~l~ :_tr!<J': <;fll??l'l~D- -

Date Administered Administrated by Date Graded Graded by Grade rn Ct.)~er C-cfcr Ro e-1-..C'-tri

\\J+-. B All work done on this exam is my own, I have neither given nor received aid.

Signature I was given the opportunity to review the correct responses to this examination.

Signature

PAGE: 1 EXAM: NRC90RO

1)

PV:l.O If the plant is at 100% power and a failure occurs resulting in the depressurization of one of the high pressure air systems in the safeguards room; what procedural requirement ensures a failure of one High Pressure Air System does not disable both High Pressure Air Systems in the Safeguard Rooms?

PAGE: 2

2)

PV:l.O Given:

PZR pressure = 2000 psia PZR temperature = 636°F Quench Tank pressure = 20 psia PORV Isolation Valves are open PORVs are leaking by slightly for PCS leak check Determine the PORV tail pipe temperature.

EXAM: NRC90RO

PAGE: 3 EXAM: NRC90RO

3)

PV:l.O What indications are available in the Control Room to monitor the High Pressure Air System?

A. Air pressure gauge located on C-13.

B. Hi-Lo Pressure Alarm located on C-13.

c. High Pressure Air Compressor control~switches on C-13.

D. Pressure Indicator Alarm(PIA) located C-11.

PAGE: 4

4)

PV: 1. 0 Given:

PCS is in cold shutdown Shutdown Cooling is in operation PCS level is at the 620 1 elevation EXAM: NRC90RO What administrative requirements exist ~o identify the actions necessary to isolate the Containment?

PAGE: 5 EXAM: NRC90RO

5)

PV:l.O The PCS is in cold shutdown and drained to the centerline of the hot leg. Shutdown cooling flow is 2900gpm.

A. What instrument is required to be read hourly for PCS level?

B. What instrument is required to be read shiftly for PCS level?

/

PAGE: 6 EXAM: NRC90RO

6)

PV:l.O The PCS Pressure is 280 psia, as indicated on PI-0104A/B. The Shift Supervisor directs you to open M0-3015 and M0-3016. Why won't the valves open at this pressure?

~-

PAGE: 7 EXAM: NRC90RO

7)

PV:l.O Given:

Reactor initially at 100% power with all rods out A loss of condenser vacuum requires reactor power and turbine load to be reduced to 50% power After power is stable at 50%, Group 4 rods are at 20 inches and the Group 4 PDIL alarm is alarming.

A. What is the time requirement for restoring rods above PDIL after an emergency power reduction?

B. If reactor power is then increased back toward 100% power and the ASI deviates from the target ASI, at what value of deviation must the power escalation be halted?

PAGE: 8--

8)

PV: 1. 0 Given:

Reactor is at 100% power PCS Xenon is 75uCi/kg EXAM: NRC90RO An S/G tube leak is suspected Off-gas flow is reported to be 4 CFM.

RIA-0631 (Condenser Off-Gas Monitor) i~-indicating 4 x 103 CPM.

Determine the estimated primary to secondary leakage rate.

PAGE: 9

9)

PV: 1. 0 Given:

Loop Tc's are at 400 degrees F PZR pressure is at 500 psia Actual PZR level is at 45%

PCS heatup is in progress with LTOP system is NOT armed EXAM: NRC90RO 3 PCPs operating Which ONE of the following statements would apply to the above conditions?

A. LTOP should be armed with both channels selected to the "SOC" setpoint.

B. Dedicated licensed operator shall be stationed in the Control Room to terminate an inadvertent HPSI pump start and stop CHG pumps as necessary to limit PCS pressure.

c. Both HPSI pumps shall be operable.

D. At least two Charging Pumps shall be rendered inoperable.

PAGE: 10 EXAM: NRC90RO

10)

PV:l.O Given:

Reactor at 100% power LIA-0702A (S/G level instrument) fails high What effect, if any would there be on the position of the following valves?

A. CV-0701 (FRV for 'A' S/G).

B. CV-0703 (FRV for 'B' S/G).

PAGE: 11

11)

PV: 1 Given:.

Reactor at 100% power P-55A is tagged out P-55B and C are operable PZR level is at 48% and decreasing PCS leakage is d~termined to be 120 A. Why is a reactor trip required?

EXAM: NRC90RO gpm

~

B. If EOP 1.0 was completed with PZR pressure at 1500 psia with all appropriate contingency actions completed, what action would be required after EOP 4.0 was entered AND PZR pressure decreased to 1300 psia?

PAGE: 12 EXAM: NRC90RO

12)

PV: 1. 0 The plant was operating-at 100% power when the following ALARMS are received:

1.

LOOP TAVE TREF DEVIATION.

2.

LOOP TO LOOP TAVE DEVIATION.

J.

S/G HIGH LEVEL.

The operator observes the following parameters:

1.

"A" S/G Level at 65% and STEADY and "B" S/G Level at 86% and TRENDING UPWARD.

2.

"B" S/G Feed Reg Valve at 95% OPEN, "A11 S/G Feed Reg Valve at 65%

OPEN.

J.

Main Feed Pumps speed at "A" 4200 rpm and "B" 4300 rpm.

4.

TA VE at 5 5 5 ° F.

Why is a manual reactor trip required?

PAGE: 13 EXAM: NRC90RO

13)

PV:l.O What is the maximum amount of time allowed to complete an independent verification of a switching and tagging order on P-BB?

~-

PAGE: 14 EXAM: NRC90RO

14)

... 1i~1<;~

PV~~/--

)I

.o Given

Reactor is at 80% power.

1. 1 and #2 Battery Chargers are in service.

1. 3 Battery Charger is inoperable and is to be tagged out.

The operator first opens Breaker 52-285~and then he mistakenly opens Breaker 72-15.

As soon as he does this, he sees a FIRE BALL blow out of Breaker 72-18 and Breaker 72-18 trips.

The plant trips.

Which of the following methods best describes how steam pressure will be controlled.

Assume no operator action.

A.

Each steam generator's pressure will be controlled by the steam safety valves on its steam header.

B.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0779 and CV0781.

c.

Steam generator pressure will be controlled by the Turbine Bypass Valves.

D.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0780.and CV0782

PAGE: 15 EXAM: NRC90RO

15)

PV: 1. 0 The Reactor has tripped automatically; EOP 1.0 has been completed and the following information is known.

Using the Diagnostic Flow Chart, determine which EOP should be entered.

YOl is energized

- All left-hand criteria are not satisf}ed No fire is occurring

- Control Room is habitable Bus "C" is energized Bus "D" is not energized DC Buses normal

- Pzr pressure at 1760 psia Both Steam Generator pressures > 800 psia

- Containment pressure is 0.5 psig and stable No steam line radiation

- Pzr pressure and level stable AFW normal No PCP are operating

- Cooling Towers normal

- all rods inserted and power decreasing

PAGE: 16 EXAM: NRC90RO

16)

PV: 1 List the buses that are fast transferred from station power to startup power on a turbine generator trip with standby power available. Assume no additional failures.

~-

PAGE: 17 EXAM: NRC90RO

17)

PV: 1. 0 While performing the "C" shift portion of SH0-1, the control room operator notices that the Control Room Air Temperature wasn't recorded during "B" shift.

Assuming SH0-1 was completed by 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> on A shift, answer the following:

A. State the tech spec/standing order 54-requirement, if any, which was violated?

B. What is the maximum allowable control room air temperature?

PAGE: 18

18)

PV: 1. 0 Given:

Reactor is at 2% power.

P-55A is in service pumping 44gpm P-55B is off and in automatic P-55C is tagged out for seal replacement EXAM: NRC90RO Explain why the above conditions are not allowed by Tech Specs.

PAGE: 19 EXAM: NRC90RO

19)

PV:l.O Annunciator EK 13-72 "Containment Iso. and Safety Inj. Left Side Containment CKT Undervoltage 11 has activated.

Further investigation reveals that breaker #3 on Y-10 is tripped and cannot be reset.

Determine the impact this condition has on the AUTOMATIC CONTAINMENT ISOLATION FUNCTION.

EXAM NRC90SRO GENERAL INFORMATION

-- EXAM INFORMATION --

me:~~,,,,,.

(~,e-t s~o.ck-1.>..,,.i t'" ~t' O-f +. G

~-~-~----~-------------------------------------------~------------------------

EXAM NO.: NRC90SRO DATE GENERATED: 03/08/90 i 7,.s-o TOTAL POINTS: 8. 5&6:t-

"'J/z.:.. tJ_,

RESPONSE TIME (min) w98. 4:;.-, o MC QUESTIONS: :L POINTS:

~F j. :).;;

or-

fl*(}*

t*</5~

TF QUESTIONS:

0 POINTS:

o.oo ES QUESTIONS:

16 POINTS:

16.00

ANSWER KEY PAGE: 1 EXAM: NRC90SRO

1)

PV:l.O Q#:3897 RT:6.0 LP:RHAAOAl.20, RHAAOAl.02, RHAAOK5.05, RHAAG28.ll,

2)

Given:

Reactor is at 100%

One PZR code safety valve begins to leak slightly Pzr pressure drops and stabilizes at 1900 psia Quench Tank pressure is 20 psia.

~

That is 587 degree F.

Determine the following:

A. What is the PORV 'tail pipe' temperature?

B. What is the new subcooled margin in terms of temperature?

ANSWER :

(0.5 pt. each)

A. 227 degree F plus or minus 5 degree F B. 41 degree F plus or minus 1 degree F Ref: Steam Tables PV:0.5 Q#:3881 RT:3.0 LP:ASBDOK4.0 The normal and backup air supply valve) is? (CHOOSE ONE)

Normal Backup A.

IA B.

IA C.

HP Air D.

HP Air CT:OB (HPSI subcooling

ANSWER KEY PAGE: 2 EXAM: NRC90SRO ANSWER :

Ref: P&ID M-225, sheets 1 and 2

3)

PV:0.5 Q#:3880 RT:3.0 LP:ASBDOK4.0l CT:OB

~-

The normal and backup air supply for CV-0742 (Feedwater Block Valve) is? (CHOOSE ONE)

Normal Backup A.

HP Air IA B.

IA HP Air

c.

HP Air None D.

HP Air N2 CORRECT RESPONSE :C ANSWER :

Ref: P&ID M-225, Sheets 1 and 2

4)

PV:l.O Q#:l204 RT:5.0 LP:ADAOG14.12, ISAAG13.0l, REQUAL CT:OB What is the maximum amount of time allowed to complete an independent verification of a switching and tagging order on P-8B?

ANSWER KEY PAGE: 3 ANSWER :

Eight hours (1.0)

Reference:

Admin 4.10 sect. 6.1

5)

PV:l.O Q#:l037 RT:5.0 LP:ASCCOG7.09, REQUAL CT:OB Given:

PCS is in cold shutdown Shutdown Cooling is in operation PCS level is at the 620 1 elevation EXAM: NRC90SRO What administrative requirements exist to identify the actions necessary to isolate the Containment?

ANSWER :

"Inoperable Containment Penetrations" sheet and "Temporary Service Lines Through Containment Openings" sheet (also accept GOP-14, Attachments 1 and 2) (0.5 for each)

Ref: GOP 14.6.2.f

ANSWER KEY PAGE: 4

6)

PV:l.O Q#:1076 RT:6.0 LP:ASCCOG7.09 CT:OB Given:

PCS is in cold shutdown Shutdown Cooling is in service EXAM: NRC90SRO PCS level has been drained to the centerline of the hot leg.

Shutdown cooling flow has been reduced to 1000 gpm Two charging pumps are tagged out for m~intenance Shutdown margin is 3.9%

What requirement(s) necessary to prevent a boron dilution accident is(are) not being met?

ANSWER :

Close and caution tag MV-CVC2162 (also accept close and caution tag PMW Supply Isolation) (1.0)

Ref: SOP 3, Section 4.0.i

7)

PV:l.O Q#:1040 RT:6.0 LP:TBAOOA2.0l, REQUAL CT:OB PCS is in cold shutdown Shutdown Cooling is in service with P-67A as the operating soc pump PCS is at 120 degrees F Plant was shutdown 10 days ago PCS is being drained with level currently at 623' elevation The following indications/reports are received:

A.O. reports loud noises coming from P-67A.

soc Hx inlet pressure indicator, PI0360, is fluctuating.

soc flow indicator, FIC 0306, is fluctuating.

A. What problem exists with P-67A?

B. SRO ONLY Assuming that P-67A was tripped and ONP 17 was activated, classify the event using the Site Emergency Implementing Procedures.

/

ANSWER KEY PAGE: 5 ANSWER :

A. pump is cavitating(or vortexing) (.25)

B. Alert(0.75)

EXAM: NRC90SRO

Reference:

SOP3 SECTION 7.3.3 rev 8, ONP 17 section 4.4.1 rev 11, E.I.

procedure E.I.-1 attachment 1 rev 15

8)

PV:l.O Q#:l348 RT:lO.O LP:RTBOOG5.05, REQUAL CT:OB Given:

Reactor initially at 100% power with all rods out A loss of condenser vacuum requires reactor power and turbine load to be reduced to 50% power After power is stable at 50%, Group 4 rods are at 20 inches and the Group 4 PDIL alarm is alarming.

A. What is the time requirement for restoring rods above PDIL after an emergency power reduction?

B. If reactor power is then increased back toward 100% power and the ASI deviates from the target ASI, at what value of deviation must the power escalation be halted?

ANSWER :

A. 2 HOURS. ( 0. 5 )

B. + OR -

0.05 FROM THE TARGET ASI.

(0.5)

REFERENCES:

ARP 5, Tech Spec 3.10.5.a I

EM-04-17 Rev 15

ANSWER KEY PAGE: 6 EXAM: NRC90SRO

9)

PV:l.O Q#:106 RT:6.0 LP:TBAFGl0.01, REQUAL CT:4, 7, OB Given:

Reactor is at 100% power PCS Xenon is 75uCi/Kg An S/G Tube leak is suspected Off-gas flow rate is reported to be 3 CFM RIA-0631(Condenser Off-Gas Monitor) is ~ndicating 7 x 104 CPM.

Determine the estimated Primary to secondary leakage rate.

ANSWER :

3 x 10-2 GPM

+ or -

0.2 x 10-2 (1.0)

(OR) 0.03 GPM

+ or -

0.002 reference ONP 23.2 attachment 2 Rev 3

10)

PV:l.O Q#:473 RT:6.0 LP:ASHAOGS.02, REQUAL CT:2, 6, 8, OB Given:

Reactor is currently at 70% power P-67A(LPSI Pump)* was deemed inoperable and taken out of service yesterday for a bearing replacement.

It has been 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since P-67A was declared inoperable.

What action, if any, is required?

ANSWER :

Rx placed in hot S/D in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (1.0)

Ref:

T.S. 3.3.2b, Admin. 5.01 section 5

ANSWER KEY PAGE: 7 EXAM: NRC90SRO

11)

PV:l.O Q#:320 RT:5.0 LP:ASLCOK6.02 CT:2, 3, 6, OB Given:

Reactor at 100% power LIA-0702A (S/G level instrument) fails high What effect, if any would there be on the position of the following valves?

A. CV-0701 (FRV for 'A' S/G).

B. CV-0703 (FRV for 'B' S/G).

ANSWER :

A.

Closes ( O. 5)

B.

None (also accept stay as is) (0.5)

Ref: M-207, Sh. 1

12)

PV:l Q#:404 RT:5.0 LP_:TBABGl0.03, TBABOA2.06 CT:4, 7., OB Given:

Reactor at 100% power P-55A is tagged out P-55B and C are operable PZR level is at 48% and.decreasing PCS leakage is determined to be 120 gpm A. Why is a reactor trip required?

B. If EOP 1.0 was completed with PZR pressure appropriate contingency actions completed, required after EOP 4.0 was entered AND PZR 1300 psia?

at 1500 psia with all what action would be pressure decreased to

'ANSWER KEY PAGE: 8 EXAM: NRC90SRO ANSWER :

A. Leak rate is greater than charging pump capacity (0.5)

2. Trip all operating PCPs (0.5)

Reference:

ONP 23.1 rev 15 EOP 1. O rev 1 EOP 4 ~ 0 rev 1

13)

PV:l.O Q#:431 RT:.6.0 LP:TBABOKJ.01, REQUAL CT:4, 7, OB Given:

Reactor tripped 15 minutes ago EOP 1.0 has been completed and EOP 2.0 has been entered T cold is 525 °F 2 PCPs are in operation T-2 level is 50%

T-81 level is 30%

T-939 level is at 41% and is available for use T-90 is unavailable T cold required is 300°F Determine the time interval for heat removal (in hours).

ANSWER :

Crcd,t c,IJov* i!6 Cl:> b., /Jw.-r~ !

"t&J.*ii:::>

).-.,~scio(: rove) o.1 13 (+ or -

1) hours (1.0)

REFERENCE:

EOP 2.0 ATTACHMENT 3 REV 1 y c O.:.',;) (!. /a.?1 o. L.

J

ANSWER KEY PAGE: 9 EXAM: NRC90SRO

14)

PV:~#:3954 RT:5.0 LP:ASABOA2.0l

'. Q Given:

Reactor is at 80% power.

CT:OB

1. 1 and #2 Battery Chargers are in service.

1. 3 Battery Charger is inoperable and is to be tagged out.

The operator first opens Breaker 52-285~and then he mistakenly opens Breaker 72-15.

As soon as he does this, he sees a FIRE BALL blow out of Breaker 72-18 and Breaker 72-18 trips.

The plant trips.

Which of the following methods best describes how steam pressure will be controlled.

Assume no operator action.

A.

Each steam generator's pressure will be controlled by the steam safety valves on its steam header.

B.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0779 and CV0781.

C.

Steam generator pressure will be controlled by the Turbine Bypass Valves.

D.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0780 and CV0782 CORRECT RESPONSE :A

ANSWER KEY PAGE: 10 ANSWER

Reference:

ONP 2.3 ONP 24.l EXAM: NRC90SRO

15)

PV:l.O Q#:702 RT:S.O LP:TBABOA2.02, REQUAL CT:4, 7, CB, OB The Reactor has tripped automatically; EOP 1.0 has been completed and the following information is known.

Using the Diagnostic Flow Chart, determine which EOP should be entered.

YOl is energized

- All left-hand criteria are not satisfied No fire is occurring

- Control Room is habitable Bus "C" is energized Bus "D" is not energized DC Buses normal

- Pzr pressure at 1760 psia Both Steam Generator pressures > 800 psia Containment pressure is 0.5 psig and stable No steam line radiation

- Pzr pressure and level stable AFW normal No PCP are operating

- Cooling Towers normal

- all rods inserted and power decreasing ANSWER :

EOP 8.0 (also accept Loss of Forced Circulation EOP) (1.0) reference EOP 1.0 attachment 1 rev 1

ANSWER KEY PAGE: 11

16)

PV:l Q#:3182 RT:3.0 LP:ASAAOK4.0l CT:OB EXAM: NRC90SRO List the buses that are fast transferred from station power to startup power on a turbine generator trip with standby power a~ailable. Assume no additional failures.

ANSWER :

lA, lB, lF, lG

(. 25 each)

~

lso "~v* a..S" &>,c;J.d?:,,,,.. t~ curr(.cr

Reference:

LP IE-8903 E-17 Sh 9 Rev 5

17)

PV:l.O Q#:1033 RT:5.0 LP:ASHFOG5.0l, REQUAL CT:OB SRO ONLY Given:

Plant is at 35% power with no inoperable equipment Air handling unit, V-95, is in service.

Condensing Unit vc-11 trips on High Lube Oil Temperature and will not reset(until oil temperature decreases).

If you declare VC-11 inoperable, how long can VC-11 be inoperable, before any action is required?

ANSWER :

7 days ( 1. O)

Reference:

Standing Order #54 section 3.14.1

ANSWER KEY PAGE: 12

18)

PV:l.O Q#:1166 RT:5.0 LP:ASFAOG5.0l, REQUAL CT:OB EXAM: NRC90SRO With the plant at 100%.power, the failed fuel monitor is taken out of service for maintenance.

It is estimated that the monitor will be isolated for 6 days.

What compensatory actions, if any, are required during this six day period?

ANSWER :

Daily sample of the PCS is required(l.O).

Reference:

Tech Spec table 4.2.l

19)

PV:l.O Q#:261 RT:3.0 LP:ASHBOG5.03 CT:S, OB Given:

Reactor is at 100% power CV-1358 (N2 to Containment) was stroke tested: closure time was 30 seconds Assuming the closure time on CV-1358 cannot be improved, what action, if any, is required?

ANSWER :

Declare CV-1358 inoperable (0.2)

Fail closed CV-1358 (also accept deactivate to isolation position) (0.4) within four hours (0.4).

(OR)

Restore CV-1358 to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (1.0).

Reference:

Tech Spec 3.6.l

MC TF ES EXAM NRC90RO GENERAL INFORMATION

-- EXAM INFORMATION --

EXAM NO.: NRC90RO DATE GENERATED: 03/08/90

. iq,oo TOTAL POINTS:

• 9. SO~~ 'J/z,ofl'i)

RESPONSE TIME (min) : 93.0

'J/1-A/'~

(Y\\ctsV r C-~f\\}

Ro f )ns.vt'.,. IUJ

?+. G QUESTIONS:

2 POINTS:

~1.00 QUESTIONS:

0 POINTS:

0.00 QUESTIONS:

17 POINTS:

17.00

ANSWER KEY PAGE: 1

1)

PV:l.O Q#:460 RT:4.0 LP:ASBDOG7.09, REQUAL CT:OB EXAM: NRC90RO If the plant is at 100% power and a failure occurs resulting in the depressurization of one of the high pressure air systems in the safeguards room; what procedural requirement ensures a failure of one High Pressure Air System does not disable both High Pressure Air Systems in the Safeguard Rooms?

~-

ANSWER :

Not allowed to tie T-9A and T-9B together with the reactor critical.

reference SOP 20 section 4.0a rev 4

2)

PV:l.O Q#:3901 RT:3.0 LP:RHAAG28.ll CT:OB Given:

PZR pressure = 2000 psia PZR temperature = 636°F Quench Tank pressure = 20 psia PORV Isolation Valves are open for PCS leak check PORVs are leaking by slightly Determine the PORV tail pipe temperature.

')ji*/'t" ANSWER :

(!:Ji -236-degrees F + or -

5 F (1. O) 1..).. 1. 9~

Ref: Steam Tables

ANSWER KEY PAGE: 2 EXAM: NRC90RO

3)

PV:l.O Q#:3885 RT:2.0 LP:ASBDOA4.0l CT:OB What indications are available in the Control Room to monitor the High Pressure Air System?

A. Air pressure gauge located on C-13.

B. Hi-Lo Pressure Alarm located on C-13.

c. High Pressure Air Compressor controlAswitches on C-13.

D. Pressure Indicator Alarm(PIA) located C-11.

CORRECT RESPONSE :B ANSWER :

Ref: P&ID M-225. Sht. 1, ARP-7

4)

PV:l.O Q#:1037 RT:S.O LP:ASCCOG7.09, REQUAL CT:OB Given:

PCS is in cold shutdown Shutdown Cooling is in operation PCS level is at the 620 1 elevation What administrative requirements exist to identify the actions necessary to isolate the Containment?

ANSWER KEY PAGE: 3 ANSWER :

EXAM: NRC90RO "Inoperable Containment Penetrations" sheet and "Temporary Service Lines Through Containment Openings" sheet (also accept GOP-14, Attachments 1 and 2) (0.5 for each)

Ref: GOP 140602.f

5)

PV:l.O Q#:l075 RT:5.0 LP:ASCCOG7.09, REQUAL CT:OB The PCS is in cold shutdown and drained to the centerline of the hot leg. Shutdown cooling flow is 2900gpm.

A. What instrument is required to be read hourly for PCS level?

B. What instrument is required to be read shiftly for pCS level?

ANSWER :

A. LIA-0105 (also accept PCS Level Indicator in Control Room) (Oo5).

Bo LG-0105 (also accept PCS level sightglass in Containment) (0.5)

Ref: GOP 14.6o3

6)

PV:l.O Q#:l039 RT:4.0 LP:ASCCOK4.0l, REQUAL CT:OB The PCS Pressure is 280 psia, as indicated on PI-0104A/B. The Shift Supervisor directs you to open M0-3015 and M0-3016. Why won't the valves open at this pressure?

ANSWER :

PCS pressure has to be < 275psia in order to open the MOVs ( 1. 0)

(OR)

Opening interlock is not satisfied (1.0).

reference:

SOP 3 Section 7.3.2

ANSWER KEY PAGE: 4

7)

PV:l.O Q#:1348 RT:lO.O LP:RTBOOG5.05, REQUAL CT:OB Given:

Reactor initially at 100% power with all rods out EXAM: NRC90RO A loss of condenser vacuum requires reactor power and turbine load to be reduced to 50% power After power is stable at 50%, Group 4 rods are at 20 inches and the Group 4 PDIL alarm is alarming.

~-

A. What is the time requirement for restoring rods above PDIL after an emergency power reduction?

B. If reactor power is then increased back toward 100% power and the ASI deviates from the target ASI, at what value of deviation must the power escalation be halted?

ANSWER :

A. 2 HOURS. ( 0. 5)

B. + OR -

0.05 FROM THE TARGET ASI.

(0.5)

REFERENCES:

ARP 5, Tech Spec 3.10.5.a, EM-04-17 Rev 15

8)

PV:l.O Q#:1291 RT:6.0 LP:TBAFGl0.01, REQUAL CT:OB Given:

Reactor is at 100% power PCS Xenon is 75uCi/kg An S/G tube leak is suspected Off-gas flow is reported to be 4 CFM.

RIA-0631 (Condenser Off-Gas Monitor) is indicating 4 x 103 CPM.

Determine the estimated primary to secondary leakage rate.

ANSWER KEY PAGE: 5 EXAM: NRC90RO ANSWER :

2.7 x 10-3 +or -

0.3 x 10 - 3 (OR) 0.0027 gpm + or -

.0003

Reference:

ONP 23.2 attachment 2 rev 3

~-

9)

PV:l.O Q#:469 RT:6.0 LP:REQUAL, ASHAOG7.09 CT:2, 6, 8, OB Given:

Loop Tc's are at 400 degrees F PZR pressure is at 500 psia Actual PZR level is at 45%

PCS heatup is in progress with 3 PCPs operating LTOP system is NOT armed Which ONE of the following statements would apply to the above conditions?

A. LTOP should be armed with both channels selected to the "SDC" setpoint.

B. Dedicated licensed operator shall be stationed in the Control Room to terminate an inadvertent HPSI pump start and stop CHG pumps as necessary to limit PCS pressure.

c. Both HPSI pumps shall be operable.

D. At least two Charging Pumps shall be rendered inoperable.

ANSWER :

B (1.0)

Reference:

SOP3 section 4 and Standing Order 54 section 3.3.2

ANSWER KEY PAGE: 6 EXAM: NRC90RO

10)

PV:l.O Q#:320 RT:S.O LP:ASLCOK6.02 CT:2, 3, 6, OB

11)

Given:

Reactor at 100% power LIA-0702A (S/G level instrument) fails high What effect, if any would there be on the position of the following valves?

A. CV-0701 (FRV for 'A' S/G).

B. CV-0703 (FRV for 'B' S/G).

ANSWER :

A.

Closes (0.5)

B.

None (also accept stay as is) (0.5)

Ref: M-207, Sh. 1 B

• f\\ll-&Y\\... =tc..

~"-M c.v....,,;)7411 c.ioSi.:> 1 /hFP d.i~*r*

p~e~..r,-<......,;Jl ;Acrl*'*~

._......,;c.>,.,....;'.'

_~s.... ilt-i" C:..\\1-0/.;>"J

(.!...Jt:;"::.t*'l" Ge

{Y'l<t""'~*'""

1-t"~tti

.~ ~ ~*,~* (.a.s)

$1.,.,

PV:l Q#:404 RT:S.O LP:TBABGl0.03, TBABOA2.06 CT:4, 7, OB Given:

Reactor at 100% power P-55A is tagged out.

P-55B and c are operable PZR level is at 48% and decreasing PCS leakage is determined to be 120 gpm A. Why is a reactor trip required?

B. If EOP 1.0 was completed with PZR pressure at 1500 psia with all appropriate contingency actions completed, what action would be required after EOP 4.0 was entered AND PZR pressure decreased to 1300 psia?

ANSWER KEY

. PAGE:

7 ANSWER :

EXAM: NRC90RO A. Leak rate is greater than charging pump capacity (0.5)

2. Trip all operating PCPs (0.5)

Reference:

ONP 23.l rev 15 EOP LO rev 1 EOP 4.0 rev 1

~-

12)

PV:l.O Q#:407 RT:6.0 LP:TBABGl0.03, REQUAL CT:4, 7, OB The plant was operating at 100% power when the following ALARMS are received:

1.

LOOP TAVE TREF DEVIATION.

2.

LOOP TO LOOP TAVE DEVIATION.

3.

S/G HIGH LEVEL.

The operator observes the following parameters:

1.

"A" S/G Level at 65% and STEADY and "B" S/G Level at 86% and TRENDING UPWARD.

2.

"B" S/G Feed Reg Valve at 95% OPEN, "A 11S/G Feed Reg Valve at 65%

OPEN.

3.

Main Feed Pumps speed at "A" 4200 rpm and "B" 4300 rpm.

4

• TA VE at 5 5 5 ° F.

Why is a manual reactor trip required?

ANSWER :

S/G Level is above the High Level Overide (85%) and level is still increasing(l.O)

Ref:

ONP 10 STEP 3.2 REV 3

ANSWER KEY PAGE: 8 EXAM: NRC90RO

13)

PV:l.O Q#:1204 RT:S.O LP:ADAOG14.12, ISAAGlJ.01, REQUAL CT:OB What is the maximum amount of time allowed to complete an independent verification of a switching and tagging order on P-8B?

ANSWER :

Eight hours (1.0)

I*

k

h. i ti' f-

~

..... *./ +

11 c:'/\\d Oil.Ow' c;'r p.v

o.

c:.re. * * -

Reference:

Admin 4.10 sect. 6.1

.. b,-s;.;,~,_!' ~*S-)

rj--

,J;of1

>

ANSWER KEY PAGE: 9 EXAM: NRC90RO

14) 1.0 PV:~Q#:3954 RT:S.O LP:ASABOA2.0l CT:OB

~ jJ Given:

\\Vr'l Reactor is at 80% power.

1. 1 and #2 Battery Chargers are in service.

1. 3 Battery Charger is inoperable and is to be tagged out.

The operator first opens Breaker 52-285~~nd then he mistakenly opens Breaker 72-15.

As soon as he does this, he sees a FIRE BALL blow out of Breaker 72-18 and Breaker 72-18 trips.

The plant trips.

Which of the following methods best describes how steam pressure will be controlled.

Assume no operator action.

A.

Each steam generator's pressure will be controlled by the steam safety valves on its steam header.

B.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0779 and CV0781.

C.

Steam generator pressure will be controlled by the Turbine Bypass Valves.

D.

Steam generator pressure will be controlled by. Atmospheric Dump Valves CV0780 and CV0782 CORRECT RESPONSE :A

ANSWER KEY PAGE: 10 EXAM: NRC90RO ANSWER

Reference:

ONP 2.3 ONP 24.l

15)

PV:l.O Q#:702 RT:5.o LP:TBABOA2.02, REQUAL CT:4, 7, CB, OB The Reactor has tripped automatically; EOP 1.0 has been completed and the following information is known.

Using the Diagnostic Flow Chart, determine which EOP should be entered.

YOl is energized

- All left-hand criteria are not satisfied No fire is occurring

- Control Room is habitable Bus "C" is energized Bus "D" is not energized DC Buses normal Pzr pressure at 1760 psia Both Steam Generator pressures > 800 psia Containment pressure is 0.5 psig and stable No steam line radiation

- Pzr pressure and level stable AFW normal No PCP are operating Cooling Towers normal

- all rods inserted and power decreasing ANSWER :

EOP 8.0 (also accept Loss of Forced Circulation EOP) (1.0) reference EOP l.O attachment 1 rev 1

ANSWER KEY PAGE: 11

16)

PV:l Q#:3182 RT:J.O LP:ASAAOK4.0l CT:OB EXAM: NRC90RO List the buses that are fast transferred from station power to startup power on a turbine generator trip with standby power available. Assume no additional failures.

ANSWER :

lA, lB, lF I (ti bu Cl '-c..c-,~'.t R

Q;)

c.;d.d.i.Pwr..I eference:

lG

(. 25 each)

C: J 0, o,,~ E b.1:. '*! --lkd t:'P*r1:'i-~9~j',,.~~-

,~r(

E-17 Sh 9 Rev 5

17)

PV:l.O Q#:1046 RT:5.0 LP:ASHFOG5.0l, REQUAL CT:OB While performing the "C" shift portion of SH0-1, the control room operator notices that the Control Room Air Temperature wasn't recorded during "B" shift.

Assuming SH0-1 was completed by 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> on A shift, answer the following:

A. State the tech spec/standing order 54 requirement, if any, which was violated?

B. What is the maximum allowable control room air temperature?

ANSWER A. not verifying control room air temperature every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (also accept table 4.2.3 section la) (0.6)

B. 90°F(0.4) reference: standing order #54 table 4.2.3 section la rev 18

ANSWER KEY

' PAGE: 12

18)

PV:l.O Q#:1163 RT:S.O LP:ASFAOGS.01, REQUAL CT:OB Given:

Reactor is at 2% power.

P-55A is in service pumping 44gpm P-55B is off and in automatic P-55C is tagged out for seal replacement EXAM: NRC90RO Explain why the above conditions are not allowed by Tech Specs.

ANSWER :

Do not have a charging pump off each bus(l.O)

(Abo a..cc-t"ft "':'" r~ *th;~ CG.. l::l;.;,~ ~ 0

!..J~~J.e If*,,.,,..~ Co..11d

'Ref: Standing Order #54 and P&ID'E-1 sheet 1 d..'1;)1<*\\i,,,_ t>>tt.i B fl b'0 """ F-µ,j)

~

• "J/J.o/9o

19)

PV:l.O Q#:254 RT:4.0 LP:ASHBOG9.0l CT:3, 6, OB Annunciator EK 13-72 "Containment Isa. and Safety Inj. Left Side Containment CKT Undervoltage" has activated.

Further investigation reveals that breaker #3 on Y-10 is tripped and cannot be reset.

Determine the impact this condition has on the AUTOMATIC CONTAINMENT ISOLATION FUNCTION.

ANSWER :

Automatic CIS will not occur on the left channel.

(1.0)

(If CIS needed, manual isolation via individual control switches will be required.)

Ref:

ARP 8 rev 47 ONP 24.1 rev 13

• ,t **

Name PERSONAL AND CONFIDENTIAL EXAMINATION COVER SHEET Social Security Number ~~~~~~~~~~~-

Company Work Location Department Course Class No.

Date Administered Administrated by Date Graded Grade n 1" :>-le/ l ~ f*a-S'2-o L~~.'>'1 Pt-. B All work done on this exam is my own, I have neither given nor received aid.

. Signature I was given the opportunity to review the correct responses to this examination.

Signature

PAGE: 1

1)

PV:l.O Given:

Reactor is at 100%

One PZR code safety valve begins to leak slightly Pzr pressure drops and stabilizes at 1900 psia Quench Tank pressure is 20 psia.

~

Thot is 587 degree F.

Determine the following:

A. What is the PORV 'tail pipe' temperature?

EXAM: NRC90SRO B. What is the new subcooled margin in terms of temperature?

I

PAGE: 2

2)

PV:0.5 The normal and backup air valve) is? (CHOOSE ONE)

A.

B.

c.

D.

Normal Backup None IA and N2

,l t*ft.-k J.

'3/ z. I) /9 {) ~-

j)-

EXAM: NRC90SRO (HPSI subcooling

't.

PAGE: 3 EXAM: NRC90SRO

3)

PV:0.5 The normal and backup air supply for CV-0742 (Feedwater Block Valve) is? (CHOOSE ONE)

Normal Backup A.

HP Air IA B.

IA HP Air

c.

HP Air None D.

HP Air N2

PAGE: 4 EXAM: NRC90SRO

4)

PV: 1. 0 What is the maximum amount of time allowed to complete an independent verification of a switching and tagging order on P-BB?

-r*

PAGE: 5

5)

PV:l.O Given:

PCS is in cold shutdown Shutdown Cooling is in operation PCS level is at the 620 1 elevation EXAM: NRC90SRO What administrative requirements exist ~o identify the actions necessary to isolate the Containment?

PAGE: 6 EXAM: NRC90SRO

6)

PV:l.O Given:

PCS is in cold shutdown Shutdown Cooling is in service PCS level has been drained to the centerline of the hot leg.

Shutdown cooling flow has been reduced to 1000 gpm Two charging pumps are tagged out for m~intenance Shutdown margin is 3.9%

What requirement(s) necessary to prevent a boron dilution accident is(are) hot being met?

..tr.

PAGE: 7 EXAM: NRC90SRO

7)

PV:l.O PCS is in cold shutdown Shutdown Cooling is in service with P-67A as the operating SOC pump PCS is at 120 degrees F Plant was shutdown 10 days ago

• PCS is being drained with level currently at 623' elevation The following indications/reports are received:

A.O. reports loud noises coming from P-67A.

soc Hx inlet pressure indicator, PI0360, is fluctuating.

soc flow indicator, FIC 0306, is fluctuating.

A. What problem exists with P-67A?

B. SRO ONLY Assuming that P-67A was tripped and ONP 17 was activated, classify the event using the Site Emergency Implementing Procedures.

I

~

I I

I i

~

PAGE: 8 EXAM: NRC90SRO

8)

PV:l.O Given:

Reactor initially at 100% power with all rods out A loss of condenser vacuum requires reactor power and turbine load to be reduced to 50% power After power is stable at 50%, Group 4 rods are at 20 inches and the Group 4 PDIL alarm is alarming.

~*

A. What is the time requirement for restoring rods above PDIL after an emergency power reduction?

B. If reactor power is then increased back toward 100% power and the ASI deviates from the target ASI, at what value of deviation must the power escalation be halted?

PAGE: 9

9)

PV:l.O Given:

Reactor is at 100% power PCS Xenon is 75uCi/Kg An S/G Tube leak is suspected Off-gas flow rate is reported to be RIA-063l(Condenser Off-Gas Monitor)

EXAM: NRC90SRO 3 CFM is indicating 7 x 104 CPM.

~

Determine the estimated Primary to secondary leakage rate.

PAGE: 10 EXAM: NRC90SRO

10)

PV:l.O Given:

Reactor is currently at 70% power P-67A(LPSI Pump) was deemed inoperable and taken out of service yesterday for a bearing replacement.

It has been 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ~ince P-67A was de~lared inoperable.

What action, if any, is required?

'*.:i.*

PAGE: 11 EXAM: NRC90SRO

11)

PV: 1. 0 Given:

Reactor at 100% power LIA-0702A (S/G level instrument) fails high What effect, if any would there be on the position of the following valves?

A. CV-0701 (FRV for 'A' S/G).

B. CV-0703 (FRV for 'B' S/G).

. PAGE: 12

12)

PV:l Given:

Reactor at 100% power P-SSA is tagged out P-55B and C are operable PZR level is at 48% and decreasing PCS l.eakage is determined to be 12 o gpm -!!.

A. Why is a reactor trip required?

EXAM: NRC90SRO B. If EOP 1.0 was completed with PZR pressure at 1500 psia with all appropriate contingency actions completed, what action would be required after EOP 4.0 was entered AND PZR pressure decreased to 1300 psia?

PAGE: 13

13)

PV: 1. 0 Given:

Reactor tripped 15 minutes EOP 1.0 has been completed T cold is 525 ° F 2 PCPs are in operation T-2 level is 50%

T-81 level is 30%

T-939 level is at 41% and T-90 is unavailable T cold required is 300°F ago and EOP 2.0 has been entered is available for use Determine the time interval for heat removal (in hours)"

EXAM: NRC90SRO

PAGE: 14 EXAM: NRC90SRO

14)

PV~~.:.\\1 ;

/,0 Given:

Reactor is at 80% power.

1. 1 and #2 Battery Chargers are in service.

1. 3 Battery Charger is inoperable and is to be tagged out.

The operator first opens Breaker 52-285~_and then he mistakenly opens Breaker 72-15.

As soon as he does this, he sees a FIRE BALL blow out of Breaker 72-18 and Breaker 72-18 trips.

The plant trips.

Which of the following methods best describes how steam pressure will be controlled.

Assume no operator action.

A.

Each steam generator's pressure will be controlled by the steam safety valves on its steam header.

B.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0779 and CV0781.

c.

Steam generator pressure will be controlled by the Turbine Bypass Valves.

D.

Steam generator pressure will be controlled by Atmospheric Dump Valves CV0780 and CV0782

PAGE: 15 EXAM: NRC90SRO

15)

PV: 1. 0 The Reactor has tripped automatically; EOP 1.0 has been completed and the following information is known.

Using the Diagnostic Flow Chart, determine which EOP should be entered.

YOl is energized

- All left-hand criteria are not satisf~ed No fire is occurring

- Control Room is habitable Bus "C" is energized Bus "D" is not energized DC Buses normal

- Pzr pressure at 1760 psia

- Both Steam Generator pressures > BOO psia

- Containment pressure is 0.5 psig and stable No steam line radiation

- Pzr pressure and level stable AFW normal

- HQ PCP are operating

- Cooling Towers normal

- all rods inserted and power decreasing

PAGE: 16 EXAM: NRC90SRO

16)

PV: 1 List the buses that are fast transferred from station power to startup power on a turbine generator trip with standby power available. Assume no additional failures.

PAGE: 17

17)

PV:l.O SRO ONLY Given:

Plant is at 35% power with no inoperable equipment Air handling unit, V-95, is in service.

EXAM: NRC90SRO Condensing Unit vc-11 trips on High Lub~ Oil Temperature and will not reset(until oil temperature decreases).

If you declare VC-11 inoperable, how long can VC-11 be inoperable, before any action is required?

PAGE: 18 EXAM: NRC90SRO

18)

PV: 1. 0 With the plant at 100% power, the failed fuel monitor is taken out of service for maintenance.

It is estimated that the monitor will be isolated for 6 days.

What compensatory actions, if any, are required during this six day period?

PAGE: 19 EXAM: NRC90SRO

19)

PV:l.O Given:

Reactor is at 100% power CV-1358 (N2 to Containment) was stroke tested: closure time was 30 seconds Assuming the closure time on CV-1358 caitnot be improved, what action, if any, is required?