Text

Exam Rept 50-220/85-03 on 850311-15.Exam results:35 Reactor Operators & 32 Senior Reactor Operators Passed Exam & 4 Reactor Operators & 1 Senior Reactor Operator Failed
	ML20205A698
Person / Time
Site:	Nine Mile Point
Issue date:	04/18/1985
From:	Joshua Berry, Keller R, Kister H; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20205A678	List: IR 05000220/1985003; ML20205A673;
References
	50-220-85-03, 50-220-85-3, NUDOCS 8504260035
	Download: ML20205A698 (103)
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U. S. NJCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.

85-03 (OL)

FACILITY DOCKET NO. 50-220 FACILITY LICENSE NO.

DPR-63 LICENSEE: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 FACILITY: Nine Mile Point, Unit 1 EXAMINATION DATES: March 11 - 14, 1985 I 7 ')- [ I PREPARED BY:

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Berry, Lead Reagtor Engineer (Examiner)

Date REVIEWED BY:

h

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. K'eller, Chie Prd ects-Section 1C Date APPROVED BY:

/g 7x'

H~. B. K1'steh Chief, Projects Branch No. 1

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SUMMARY:

As part of the NRC's programmatic evaluation of Requalification Training at Nine Mile Point, Unit 1, NRC prepared written examinations were administered, in parts, to all facility personnel taking the Niagara Mohawk prepared annual requalification examinations the week of March 11, 1985.

Additionally, oral requalification examinations were given to 11 licensed personnel, 7 SR0s and 4 R0s.

8504260035 850422 PDR ADOCK 05000220 G

PDR

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REPORT DETAILS TYPE OF EXAMS:

Requalification EXAM RESULTS:

l RO l

SRO l

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Pass / Fail l

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l Written Exam l_

35/4 l

32/1 l

l Partial Exams I l

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6/1 l

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32/1

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CHIEF EXAMINER AT SITE:

J. A. Berry, U.S. NRC - Region I

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OTHER EXAMINERS:

D. J. Lange, U.S. NRC Region I F. 4. Crescenzo, U.S. NRC - Region I T. L. Morgan,' EG&G Idaho, Inc.

D. E. Hill, EG&G Idaho, Inc.

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REPORT:

As part of the NRC's programmatic evaluation of-Requalification Training

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-at Nine Mile Point - Unit 1,

NRC prepared written examinations were-administered in parts, to all facility personnel taking the Niagara Mohawk prepared annual Requalification examinations the week of March 11, _1985.

Additionally, oral requalification _ examinations were given to 11 licensed

- personnel, 7 SR0s and 4 R0s, :-

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-3-The NRC written examination sections were administered as follows:

Honday, March 11 - R0 Section 2 to 17 people

- SR0 Sections 5 & 8 to 11 people Tuesday, March 12 - FG Section 3 to 13 people

- IRO Section 6 to 12 people Wednesday, March 13 - R0 Sections 1 & 4 to 9 people SRO Section 7 to 11 people Overall, examination results were good.

Five people failed NRC adminis-tered sections of the examinations, four R0's and one SRO, and one SR0 failed the oral examination.

The comparison of scores on NRC sections vs. the facility sections indicated that the overall average score on the NRC exam (if sections were together) and facility exam were _ within 4% of each other.

This is con-sidered an acceptable range.

Individual section comparisons indicated a wide disparity.

Section 8 of the NRC and Facility exams were within.5%

of each other in average score, but Sections 2, 3, and 6 were off by

- 6.71%, 9.91% and 6.56% respectively, with the NRC section scores being

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lower. Also, Sections 4 and 7 on the NRC exam had average scores 6.4% and 3.3% higher than the facility's sections.

The reasons for this disparity are not evident.

It appears that the higher scores on the NRC Sections 4 and 7 may be due to the facility's sections being overly long, but the other section differences cannot be so explained.

Probable causes may be the tension involved in taking an NRC exam, more operationally oriented (not memorization) type questions on the NRC exam, _ or the difference in question " style" between the two exams.

In _ addition to the conduct of examinations, the evaluation also consisted of a review of the NMP-1 Requalification Program Annual examinations-prepared by the facility, and discussions with licensed operators and training. staff members regarding the Requalification Program.

The Annual. Requalification examinations prepared by the facility were considered to meet NRC requirements, but were not considered to be of high quality.

Problems with the examinations included; double jeopardy questions, excessive length, many unnecessary theory calculations and questions having no relation to an operator's job, and simplistic short answer questions which failed to provide an adequate measure of depth of knowledge.

The facility's Requalification examination question bank is poor, and it is felt its use contributed to the problems with the exam-ination.

To Niagara Mohawk's credit, they have identified the problems with the existing exam question bank, and have begun a task to upgrade it.

Significant improvement is expected in next years exam.

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-4-Discussions with licensed operators indicate that there is dissatisfaction with the Requalification. Program.

Problems sited included; to much emphasis on theory that is not operationally oriented, too much self-study or reading, and unchallenging and uninteresting presentation of subject matter.

These matters have been previously brought to the attention of Niagara Mohawk management by other reviews of the program.

' Niagara Mohawk has committed to a course of corrective action. NRC Region I will monitor the progress of the action over the next year.

It is felt that the addition of a plant specific simulator training program to the Requal program will aid in improving the program.

Overall, the Nine Mile Point, Unit 1 Requalification program is satis-factory.

NMPC has already begun to correct many of the programmatic problems identified. No further NRC involvement in the program is planned this year, other than monitoring of the changes being made to improve its quality.

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4.

Personnel Present at Exit Interview:

NRC Personnel J. Linville, Chief, Reactor Projects Section 2C, DRP J. A. Berry, Lead Reactor Engineer (Examiner) DRP D. J. Lange, Reactor Engineer (Examiner), DRP F. J. Crescerzo, Reactor Engineer (Examiner), DRP A. J. Luptak, Resident Inspector, NMP-1 NRC Contractor Personnel D. E. Hill, EG&G Idaho, Inc.

T. L. Morgan, EG&G Idaho, Inc.

Facility Personnel T. W. Roman, Station Superintendent - NMP-1 K. F. Zollitsch, Training Superintendent, Niagara Mohawk J. C. Aldrich, Operations Supervisor, NMP-1 T. Wood, Training Supervisor, NMP-1 J. T. Pavel, Asst. Training Superintendent, Niagara Mohawk R. Seifried, Operations Training Instructor M. Dooley, Operations Training Instructor M. Jones, Operation Supervisor, NMP-2 i

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_ Summary of Comments made at exit interview:

The_ Chief Examiner noted that there was one person who was not a

clear _ pass on.the oral examinations.

A discussion was held regarding~ Niagara Mohawk's commitment to

implementation of upgrades in their Requalification Program based on previous audits.

Attachments': Written Examination (s) and Answer Key (s) (SR0/R0)

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U.S.

NUCLEAR REGULATORY COMMISSION

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REACTOR OPEFATOR RECUALIFICATION ErAMINATION FACILITY:

NINE MILE POINT

_________________________

REACTOR TYPEI BWR-GE2

_________________________

DATE ADMINISTERED: 85/03/11

_________________________

EXAnINER:

BERRY.

J.

_________________________

NAME:

_________________________

INSTRUCTIONS

____________

Uso. separate paper for the answers.

Write ansvers on one side only.

Steple question sheet on top of the answer sheets.

Points for each question are' indicated in parentheses after the question. The passino 3rade requires at least 70%

% OF CATEGORY

% OF CATEGORY UALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

__"_" ____

'"*00 100.00

PROCEDURES - NORMAL, ABdORhAL.

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___________

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ENERGENCY AND RADIOLOGICAL CONTROL 25.00 100.00 TOTALS

________ ______

___________

________

FINAL GRADE _________________%

All work done on this e:-t a m i na t i o n is my own. I have neither given not rece2ved aid.

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

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QUESTION

'4.01 (3.00)

In accordance with procedure NI-SOP-32, Failure of Reactor to

' Ser am, wha t si:: (6) immediate actions vovld you take to reduce power and insert all control rods in an ATWS situation?

(3.0)

GUESTION 4.02 (2.00)

Dascribe, in general, the four things you would co to reset a high pressure ecolant injection (HPCI) initiation, assuming that the initiation signal has cleared.

(2.0)

GUESTION 4.03 (2.00)

s.

Why is an operator instructed to ' reduce reactor power to 802 of the original power level with Reactor Rectreviation flow'

BEFORE removing a feedvater heater string?

(1 0)

b.

When two condensate booster pumps are required, the preferred linup is with 411 and 413 runninsi eben one booster pump is required, #11 or $13 should be in service.

Why is this pre-ferred?

(1.0)

GUESTION 4.04 (3.00)

c.

Assuming the CSO and the NAOE vere able to accomplish NOTHING in the wey of securing the staticn prior tc an evstortion. he, is th+ rescto* shutdoun ArJD how is the snutdoer, ccrified?

tov-answer should include where the CSO and NAOE proceed to and their subsequent actions.

(1,0)

-b.

After veriflestion of a turbine trip. the SSS is to proceed tc powerboard 11 E 12 What actions are to be performed at power-board 11 E 125 (1.0)

c.

How can RAW WATER be supplied to feed the reactor 5 (1.0)

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND.

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GUESTION 4 05 (3.00)

Concerning Procedure SOP-19 (Unexplained Reactivity Change);

a.

List sia (6) plant par ameter s/indic ations that should be checked if an unenplained reactivity char.se should occur at rateo power.

(1.5)

b.

Deperiding on the magnitude of the reactivity change, list three alar ms that may be t ra i t i a t e d. (PRIOR to a reactor scram?

(0.75)

c.

If this reactivity chan3e is a result of decreased temperature. due to a loss of a feedwater heater string, what is your immediate action and what two (2) adverse conditions are you trying to protect against? (.75)

ullESTION 4.06 (3.00)

Concerreirig Pr ocedur e N1-Op-14. Containnient Spray System; s.

What two (2) signals are required to automaticallv start the c ont ai nme ret spiay pumps?

(0.5)

b.

What action shov1d be taken following a confirmed high radiation alarm in the containment spray raw water system?

(0.5)

c.

The containment spray Raw Water Pumps avst be manually started b.v the control room oper ator ? TRUE or FALSE.

s0.25:

d.

This procedure directs you not.to manually override or shut this svsten down after an auto. Initiation unless two conditions are met. What ar e these two conditions and who is author 1:ec tc mate this decision 5 1.25 00ESTION 4.07 (2.00i During the 4:00 pm to 12:00 midnight shift. at ratad power. vou receive two alarms:

1 Off GAS line high pressure.

2 0ff GAS line hign temper atur e You notice that the condenser vaevum is decreasing.

c. Based on the above indications / conditions, WHAT HAS OCCURED?. and what sde:': ens: sutomstic sc*: ens rsn be er: r : * s t ~

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EEseC o r, tae soovs s i t u s t i ori list youf I fti n> 9 G 1 E * e oper &l oi 5-0 1 1 o n s.

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

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~~00ESTION 4.08-(2.50)

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ConcerninS N1-SOP-29, Pipe Break Inside Drywell; w

c.

U r.de r uhat conditions can the automatic controls of an Emergency Core-Cooling System oe placed in its manual mode? (be specific)

(1.0)

Think about the overall purpose of this procedure; ---; List at least three (3) operational functions, with respect to the Core and its Containment, you are expected to achieve to assure that the HEALTH and SAFETY of the public-is protected.

(1.5)

uuESTION 4.09 (2.50)

According to N1-SOP-3, Feedeater Halfunction (Decreasing FW Flow);

c.

What immediate actions would you take if feedwater flow rapidly decressed due to a loss of the Shaft Feedvater Pump?

(1.5s b.'Due to the above transient RX. Vessel level is decreasing at a very rapid rate. As the Shift Supervisor, at what Vessel level vovld you direct your. operators to depressurire the vessel'

'.s0.25)

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c.

Is it necessary to close the MSIVs during this transient?

( 0.25 )-

d.

List thr ee (3) conditions that could cause HPCI to evtomatically initiate as a result of this transient.

(0.5->

OUESTION 4 ~. 1 0 (2.00)

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Concerning Procedure 50P-15 Malfunction of the Control Rod Drsve E.' s t e n..

s.

During s-power ascension,(RX. Power appron. 30 %),

the selected control rod starts to drift. What Automatic responses. iei a l a r m s.'i n-dications. would be affected?

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(1.0)g:4 b.

What-criterion is used to define a control too as being inopersele?(0.Sf '

c.

Hcs could you verifv that a control rod has become uncovoled?

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AN5WERS -- N1NE MILE POINT-85/03/11-BERRY, J.

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3.00)~

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ANSWER 4.01

Place Mode Switch in shutdown (This inserts an additional s t r a n, signal)

. Trip recirculstion pumps

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Fully insert control rods using 'Emer, er cy Rod'In'

4.

Feset RPS tcip. danvally scram the

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s 5. Individually scram rods from

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panel r 6 Isolate and vent scram air heacer locally (0.33 each)

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REFERENCE

JCK-174 NJ-SOP-32, Rev.

4'_ Pg 8

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2.00)

ANSWER 4.02 (

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'J e r i f y a) Feedvater flov on til and #12'is '

1.0 mil'11on lbm/hr.

b) Reactor low level trip is clear (.5)

Suitch feedeater pump til and #12 M/A stations to'eanval (.5)

3.

Adjust the manual outputs until the deviation meters on the ill

12 M/A stations are nulled.

(.5)

F r,e s s t h e / F eeQ4t e r Return to Normal After HPCI' p u s h bu t t or, on the reactor control console.

(.5)

REFERENC'E

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W : -O F -4 6., p g. ic EDH-324

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ANEWER 4.03 (2.00)

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a.

This vill' prevent the other feedvater heater strings from being overloaded and will preclude possible over p'ower of the nuclear f uel._ 'Also power increase due to' increased inlet subcoo]Ing.

(1.t b.

This preferred lineup will preclude a system feedwater 'olitur-

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bance.dve to the loss of poverboard #101.

Also-wesores FMI (1.0)

c...levalaLy.

REFERENCE N:-Or-1:. eg. 10 EDF-I".-

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ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

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j ANSWE.~t 4.04 (3.00)

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c.

Toe.C50 proceeds to shutdown panel #12 and tr::s MG set 141 The NADE proceeds to shutdown pane 1 til and trips nG set 131.

Verif-ication i s :- t h e 'All Rods In' ehite light on tnear respective panels.

(1.0)

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b.

Verifies.that a condensate and feedwater booster pump are opera-ting and starts feedwater pump til, if HPCI has failed to initiate (1 0)

Also manual transfer of PB-11&12 if auto transfer fails c.

By installing an available spool piece between the feedwater sys-tem and the fire protection water svstem.

(1.0)

Also cross-connect;to containment spray raw water through inter-s

. tie. valves M cerdF g FEFERENCE N1-SOP-11, pg.'3-5

EDH-318 I

QNEWEI 4.05 (3.00)

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s, A.

1.

Control Rod position.

4.

Steam flow or temp.

2.

Recircolation flow.

5.

Feedwater flow or temp.

Reactor. pressure.-

6.

Turbine Generator load.

7.

Bspass, relief or safety valve flow.

(any si: at 0.25 for each correit-ans.)

_3.

5CC BloCr E15rm.

(0.25 e5Ch)

b.

1.

_ K ". alarm

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APRM alaf8..

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c. Reduce power to 80 7. of the power level prior tc the change using Recarc

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flow. -(0.25) This will prevent bundle overpower t.0.25) and overload 1.3 of the other feedwater heater strings. (0.25)

REFERENCE NhP. Ni-SOP-19 i p g, 2 E.3 )

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

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ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

ANSWER.

4.06 (3.00)

a.

A ecmbination of lo-lo reactor vessel water level and high dryecil pres-sure (3.5 psig.)

(0.50)

b..The rae water pump and the containment spray pump in the affected loof should be secured. (0.25)

The loop svetion and discharge valves should be closed. (0.25)

e.

TRUE (0.25)

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.d.

1. Sufficient evidence shows that the system is not performing its intended function. (0.50)

2.

Continued operation will prolons or produce an unsafe condition. (0.50)

Shutdown of the system will be at the direction oi' the Station Shift Supv.

(0.255 REFERENCE NMP. #1-OP-14,.pages I thru 5

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ANSWEF 4.07 (2.00)

s.

EXPLOSION in the Air Ejector discharge piping. (0.50)

.Avtomatic Actionsi 1. Valves BV-76-12/13 elose and off gas flow goes to :ero.

(0.50i 2. Reactor Scram at 23 Hg.

(0.25)

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1<R=cyce reactor load by decreasing rec 1rc. ficv.

2. C '. c s e main steam supp1v valve to air ejector 5 and m 1::i n g get.

3. Insert control rods per rod pattern until vaevue decreases to near scram point..

4.hanvally scram the reactor.

5.

Initiate emergency condensers, as necessary to remove the decay heat.

e.

Inform station per sonnel of conditions.

. Notifv Plant Superintendent.

( 7 correct ans, at 0.1785 es.)

REFERENCE NMP.

SOP-18

,' Explosion in the Air Ejection Disch. Piping'

I n *. cme avi:n t-. : A c ; ; ;a i O *- e E. : - - :?..:

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ANSWERS -- NINE MILE POINT-85/03/11-DERRY, J.

ANSWER 4.08 (2.50)

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1.

Misoperation-in automatic is confirmed by at least two i nde p e nd ere t process parameter indications.

(0.50)

2.

Core cooling is assured AND this procedure,(50P-2o), directs you.(0.50 b.

1.

haintain Core Cooling.

Limit the release of off sas radiation.

3. Place the Reactor Core and Containment in a SAFE STABLE condition.

4.

Keep the Torus bulk temp. vithin specified SAFETY limits.

(any three at 0.50 each)

NOTEi Other specific Operational objectives related to SAFETY accepted.

REFERENCE NMP. N1-50P-33. Pipe Break Inside Dryvell., Cautions, Limitations.and over-all purpose and objectives. pst 1-5

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ANSWER 4.09-(2,50)

a.

1.

Shift mode switch to refuel.

2.

Check all. rods are fully inserted.

3.

Observe power level decreasing.

4.

Check'for HPCI operation. Ensure that both moter driven feedvater pumps ar e r unreing.

5.

Cheek that the emergency condensers are in operation.

6.

Chec6 that the Core Spray pumps ar e r unr.ing and recir ev1 sting back to the t o -.i s.

(0 25 for each correct a r, s v e r,

b.

Low-Low-Low Level (-

10 inches)

(0.50)

c.

Yes, To conserve coolant inventory. (0.40)

d.

1.

Runout flow of 1.9 :: 10(6) or 3800 spa. (0.20)

Turbine Trip (0.20)

3. Low R:.. water level (0.20i REFERENCE i

i NMP. SOP-3, and Simulator scenerio Objectives t it:

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ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

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ANSWER 4.10 (2.00)

c.

1.

LPRM output in the vicinitv of the drif' ting rod.

2.

Positior. indication for that rod.

NOTE: The ROD DRIFT alarm vill not come in due to the rod that 25 drifting is the one selected.

The RWM alarm will not come in due to being > 25 % power.

(1.00)

b. A control rod which cannot be moved with controi rod drive pressure.(.5)

c.

Control Rod overtravel alarmi ( Ann. eindow F2-6,

Computer point 80-12)

when the contro; rod has been fully withdrawn.

(0.50)

d d '+ D E ^ "f--

tem.

REFERENCE

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Malfunction of CRD sys NMP., 50P-15,

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U.S.

NUCLEAR REGULATORY COMMISE!ON PEACTOR OFEFATOR REDUALIFICATION Ex'MINATION FACILITY:

N1NE MILE POINT

_________________________

REACTOR TYPE:

BWR-GE2

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

DATE ADhINISfERED: 85/03/11

___________.._____________

EXAnINER:

BERRY.

J.

_________________________

NAME:

_________________________

-INSTRUCTIONS

____________

Uso separate paper for the answers.

Write answers on one side only.

Staple. question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the questiore. The passing gr ade r equires at least 70%

least 80%.

Examination pape.

% OF CATEGORY

% OF CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

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'".00 100.00

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INSTRUMENis AND CON 1ROL5 25.00 100.00 TOTALS

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FINAL GRADE

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All work oone on tnis examination is my own. I have neither given not received aid.

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INSTRUMENTS AND CONTROLS PAGE

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QilESTION 3.01 (3.00)

o. What are two signals (including setpoints) that would cause the Feed System to shift to the HPCI Mode of operation and HOW would the results of a LOCA initiate each?

(1.0)

B.

When the Feed Pump is idle or not in the HPCI mode, the HPCI controller setpoint is blocked. This setpoint is not appliec to the controller until ____'____ is-produced at the ____?____.

(1.0)

C.

Why is the HPCI controller setpoint blocked initially?

(1.0)

QUESTION 3.02 (1.50)

For each of the following systems, list what TYPE of radiation detector is used and what AUTOMATIC ACTIONS occur when the tonitors trip. (NOTE: If no auto actions occur, indicate so.

Assume lineups are appropriate for auto actions to occur.)

A.

Air ejector offgas B.

RBCLC C.

Refueling Bridge (1.5)

OUESTION 3.03 (2.50)

Concerning the fuel :one level detectors:

s.

Undet what conditions does the valve displayed on the fuel j

one instrument come from the FUEL ZONE level transmitters (Include svstem initiation signal (s) in your answe*)

(2.0)

d.,What inclestion does the control room operator h a '. e that reference leg flashing 15 occuring in the fuel rone level instrument?-

(0.5)

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INSTRUMENfS AND CONTROLS PAGE

____________________________

QUESTION 3.04 (3.00)

Aosvae the REACTOR LEVEL' CONTROL SYSTEM is being operated in 3-element centrol using reactor level detector channel

'11'.

Reactor power is at AS%, STEADY STATE.

For each of the instrument or control signal failures listed below. 51A1E HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase, decrease, or remains constant) and BRIEFLY EXPLAIN WHY in terms of what is happening in the Level Control System immediately f ollowing the f ailur e.

(FOR EXAMPLE, your answers should include the following detail,

'Causes reactor level-to decrease due to a steam flow / feed flow error signal, steam floe < feed flow, resulting in a closure si 3nal to the feedwater control valve.')

= NOTE' A block. diastam of the Feedwater Control Svstem is on the following page for your use.

-a.

412 FEEDWATER FLOW transmitter FAILS HlGH (1.0)

b.

Channel

'11'

REACTOR LEVEL detector signal FAILS LOW.

(1.0)

c.

LOSS OF CONTROL SIGNAL to #13 FEEDWATER CON 1ROL VALUE.

(1.0)

.

DUESTION 3.05 (3.00)

For'esch of the followins,~ state whether a ROD BLOCK, HALF-SCRAM.

FULL SCRAh, or NO PROTECTIVE ACTION is generated for that condition.

NOTE *

IF two or more actions are generated, i.e.

rod block and a half-scram, state the-most severe.

1.e.

half-scram.

Assume NO oper-stor actions.

a.

APRn 11 Downscale, noce Switch in EUN (0.c.

-

b.

(4-LPRM inputs to APRM 15, Mode Switch in STARTUP (0.6)

c.

Both Flow Conv. Units Upscale (>107% flow), Moce Switch in RUN (0.6)

6.

APRM 12 and 16 Upscale. Mode Switch in STARTUP (0.6)

e.

Main Steam Line 111 ISOLATED, Mode Switch in RUN (0.6)

GUESTION 3.06 (2.00)

Describe fully how Reactor Building Closed Loop CoolingtRBCLC)

temperature is regulated as the heat load on the svstem

. : es5+5

.;..

-.

.

U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILIlY:

N1NE MILE POINT

_________________________

~

REACTOR TYPE:

BWR-GE2

_________________________

DAfE ADMINISTERED: 85/03/12

_________________________

EXAMINER:

LANGE, D.

NAME:

_ _

/

_

__

INSTRUCTION 9;

_____________

Uno separate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each qu2stion are indicated in parentheses after the question. The passing grade requires at least 70% in each cate3ory and a final gr ade' of at least 80%.

Examination papers vill be picked up s i :- (6)

hours after the e::a mi n a ti on stsrts.

% OF CATEGORY

% OF APPLICANT'S CATEGORY UALUE TOTAL SCORE VALUE CATEGORY

,

________ ______

___________

________ ___________________________________

__ 1

________.6.

PLANT SYSTEMS DESIGN, CON 1ROL,

__

___________

AND INSTRU.*.ENTATION 25.00 100 00 f0TALS

________ ______

___________

________

FINAL GRADE

_________________%

.: 1 1

..o r t cone on this e a a n. : r.s t 2 o n :s m.

co n.

I ha e ncitner 31ven nor receiveo aid.

_____________

_

______________

_,..

.

,.

_.,_ -

.

U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REDUALIFICATION EXAMINATION FACILI1Y:

NINE MILE POINT

_-------------_--_-

REACTOR TYPE:

BWR-GE2

.

__----_---_--_--_____

DATE ADMINISTERED: 85/03/12

__-------__--_--___-

EXAMINER:

LANGE, D.

NAME:

____---__________________

INSTRUCTIONS l

-____-_--__--

Uco ceparate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing 3 rod 3 requires at least 70% in each category and a final grade of at loost 80%.

Examination papers will be picked up six (6)

hours after othe enamination starts.

% OF ATEGORY

% OF APPLICANT'S CATEGORY

-

VALUE

. TOTAL SCORE VALUE CATEGORY

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

__---

-----_------------- ---------_--_-_

'5

_- I_00 100

.- l00 6.

PLANT SYSTEMS DESIGN, CONTROL,

_

___________

________

_-_

AND INSTRUMENTATION 25.00 100.00 TOTALS

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

_____-_----

_-__-___

FINAL GRADE _________________%

411.'ork done on this examination is my ovn. I have neither

,

'31 von not received aid.

'

-_-_ ----

_---

SIGNATURE

.

~,

_,

._

_., _,, -.

., -.,

.

CONTROL, AND'INSTRUMEN1ATION PAGE

6.___ PLANT SYSTEMS DESIGN,

__

_________________________________________________

QUESTION 6.01 (3.00)

A.

What conditions must be met to satisfy the logic for ADS Initiation ?

lo3 c arrangement')

(2.00)

NOTE:

( Include setpoints and trip i

B.

How would the system respond if MSERV #1 failed to open after proper logic actuation ?

( all other valves respond properly )

(0.50)

C.

What are tuo types of detectors used to provide positive indication of e leaking / lifted RELIEF VALVE ? ( e::c l u de lights and annunciator s )

40.50)

OUESTION 6.02 (2.50)

c. List six operational conditions that will cause an automatic closure of the Main Steam Isolation Valves. ( include setpoints and bypasses )

(1.50)

b.

List three (3) functions of the Main Steam Line Flav Restrictors. (1.00)

GUESTION 6.03 (3.00)

Concerning the High Pressure cooling injection system (HPCI) :

e. What prevents an idle feedwater pump from starting and pumping water thr ough a FULLY OPEN feedeater control valve following a HPCI inittst-ion signal ?

(0.75)

'

If s HPCI initiation occurs eith NO LOSS of 0FF-5ITE POWER. stste the c.

effect on the following pumps / valves or components.

1.

Condensste and feeduster pumps tnat are running.

60.25;

Idle feeowater pump.

(0 25)

3.

Feecester control system.

(0.25)

4.

Feeduster pumo controller 4 11 (0.25'

Feeoester Pump controller t 1 0.25 c.

In addition to a HPCI initiation being block.ed by protective pump lock-outs. list three (3) additional INiERLOCK5 thst elll also PREVEN 1 an automatic start.

(1.00)

,

I

\\

.

y 4.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMEN1A1 ION PAGE

______________________________________________________

i

' OllE ST ION 6.04 (1.50)

Concerning the Traversing In Core Probe Sys. (TIP) :

List two -(:2) specific areas of information that are obtained from signals 3enerated by the TIP system Be sure to include ho9 the signals are being

.

used and what information is being obtained.

(1.50)

GUESTION 6.05 (2.00)

Hou is the integrity of ECCS piping inside the reactor vessel versfled dur i r.g nor ma l o p e r a t i o r. ? In your answer include. SENSING POINTE. SPECIFIC SYSTEM (s) WHOSE PIPING IS BEING VERIFIED, WHY Il IS VERIFIED and the re-oponse of the instrumentation to a loss of integrity (2.00)

.

OllESTION 6.06 (2.25)

Concerr irig the G e r.e r a t o r Stator Cooling Water system ;

s.

What three (3) conditions will cause a Turbine Governor Runbsch ? (0.75)

( SETPOINTS ARE REQUIRED )

6.

Will an automatic Reactor Sctsm occur upon rece1Ft of a Governor runback trip signal? If yes, from what ? If not. how could a subsequent scr am be prevented ?

(0.75)

c.

What is the importance of r egulatirig f low within this system to maintair.

pressure beteeen 22-25 psi ?

(0.75)

GUESTION 6.0'

(3.00)

Concerning Reactor Vessel Level Instrumentation ;

s.

Using the attached figure 2-1. i R::.L ev el Inst.>.

Indicete what control

.

r c e n-l e '. e : I n s t r u m e n '. s are vEed to measure l e v s.' FirsmeterE

.I.E*J d 5.

Indicate what level parameter itsms 3,7 6.

,10 11, & 1 signity s1.20i b.

What plant condition (s) vill automatically initiate the fuel zone level detectors ?

(0.50)

c.

What indications would vou use to verifv reference leg flashing in the fuel zone level detectors ? ( be specific )

(0.70)

c.

What three (3) plant operation variables are used for compensation bv the Fuel Zone Level Indicators ?

(0.60)

-

.

6.

PLANT SYS1 EMS DESIGN, CON 1ROL, AND INSTRUhENTA1 ION PAGE

.. ___________________________________________________

GUESTION 6 08 (2.75)

Concerning the Standby Liquid Control Sys o. Once the SBLC sys, has initiated, what six (6) CON 1ROL ROOM indications could you use to verify that the system is operatin3 Properly AND in-jecting into the reactor vessel ?

( 1.50 )

b. After initiation of the SBLC. sys.,is it permissible to shut the system doen ? ( If not. WHY ? If so, under ehat conditions ? )

( 1.25 )

GUESTION 6.09 (2.25)

Concerning the CORE SPRAY system ;

o. What protective design feature, within the core spray system, allows for running the core spray pumps at shutoff head without overheating them ?

(Explain fully-be specific)

(0.75)

P ng is sensed in three different elaces.

i b.

Pressure in the Core Spray pi List these three sensing points, indicating what is being sensed and any automatic actions, alarms or indications that are provided from them. ( CCT*0!:"': 'RE ':GUIRED )

(1.50)

.

GUES' TION 6 10 (2.75)

According to procedure 4 N1-50P-5, ( Instrument Air Failure

)'

During your shift you enperience a complete loss of instrument air header pressure. For the fo. lowing air oPerateo valves, indicate the valve action

.

iopen or closed) and the station responses 1f any).

(2.75)

1.

Main Steam Isolation Valves.

2.

Feedwater Flow Control Valve.

Ma'e-up valve to concenser.

r

Fer:to-E.lo3 CLC-TCV

.

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,

U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILI1Y:

NINE MILE POINT

_________________________

REACTOR TYPE:

E:WR-GE2

_________________________

DATE ADMINISTERED: 85/03/12

_________________________

EXAMINER:

LANGE.

D.

NAME:

___

_ ___ ____ _

INSfRUCTIONS;

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

Use separate paper for the ansvers.

Write answers on one side only.

Stople question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing Stade requires at least 70% in each cate3ory and a final grade of at lenzt 80%.

Examination papers will be picked up six (6)

hours after th? examination starts.

% OF CATEGORY

% OF APPLICANT'S CATEGORY

_ _ 'J A L U ETOTAL SCORE VALUE CATEGORY

______ ______

___________

________ ___________________________________

__'______ ___ __

'5.00 100*00 7.

PROCEDURES - NORMAL, A E:N O R M A L,

___________

________

EMERGENCY ANO RADIOLOGICAL CONTROL 25.00 100.00 f0TALS

________ ______

___________

________

FINAL GRADE,__________,,___,%

1;; vori cone on this e nsn4: nat:on is ar,.

can. I have ne:the*

3:ven not rece:vec sid.

____________

_

______________

_ _ _

.

,

..-

.

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

-

RADIOLOGICAL CONTROL

-__--- -_---------_-

OllES1 ION 7.01 (2.50)

Concerning Procedure N1-OP-30, ( 4.16 Kv, 600 V, and 480 V House Service);

c. If low volta 3e is sensed on PB 11 or PB 12, its respective supply break-er-will open. Based on thise.at what point will the reserve supply breaker close ?( be specific as to value and reason why ).

(1 00)

b.. What is the normal,( oPen or closed ) Position and ehy, of feeder br eak -

ers to PB-101 (R-1014/1011 ? What is the interlock function of these breakers and how can this interlock be defeated ?

(1.00)

c.

What interlock e::ists between the feeder and tie breakers on PB 16 & FB

(0.50)

.

QUESTION 7.02 (3.00)

Concerning procedure SOP-10 ( Une::plained Reactivity Change ) ;

a.

List s i :- (6) plant parameters / indications that should be checked if an une::plained reactivity change should occur at rated power.

(1.50)

b.

Deperiding on the magnitude of the reactivity change list thr ee alarms that may be initiated. ( prior to a reactor scram )

(0. 5)

c.

If this reactivity change is a result of decreased temperature, due to

, what is your immediate action and a loss of a feedwater heater string what two (2) adverse conditions are you trying to protect against? t.75)

NUESTION 7.03 (2.50)

, Containment Spray System

Concerning procedure N1-OP-14 s.

What two (2) signals are reevired to automatically start the contain-ment spray pumps.

t C.!! )

6.

What action snould be taken following a confirmed high radiation alara on the containment spray taw water system ?

(0.50)

c.

The containment spray Raw Water Pumps avst be manually started by the c o r.t r o l room operator ? TRUE or FALSE (0.25)

.

d.

This procedure directs you not to manually override or shut this system dowri af ter an auto initiation unless twc conditions ar e met. What are these two conditions and who is authorized to make this decision ?

(1.25)

i

. _.

..

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

.

E7.

PROCEDURES - WORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~~~~~~~~~~~~~~~~~~~~~

~~~~ I65 L6G56 E~C STR6L R

____________________

GUESTION 7.04 (2.50)

a Ccncerning procedure N1-SOP-11 ( Control Room Evacuation, Fire )

k o. Where do Remote-Shutdown Panels 4 11 and 4 12 receive their power from ?

(0.50)

b.

Being forced to evacuate the control room, an attempt should sia d e to bring the P ant to a <afe shutdoen condit2on before leaving. List, in l

order of preference, eight (8) immedicte operator actions / verifications to be attempted prior to leaving.

(2.00)

DUESTION 7.05 (3 00)

Rocarding Shift Supervisor responsibilities and concerns involving the is-cuance and use of RADIATION WORK PERMITS,(RWP's).

involving work, that would a. List five (5) of_the sin (6) conditions,

require the issuance and use of an RWP (1.25)

.

b.

List three (3) specific qualifications, duties and/or criteria that apply to all personnel assigned as" LEADnAN on an RWP (0.75)

'

.

c.

If a maintenance activity must carcy over to the next shift, the RWP for that activity, must be approved by the appropriate Rad. Protection fech.,

s Leadman, and be re-initialed or sigreed by the Station Shift Supervisor.

( (RUE or FALSE i (0.25)

o.

What specific qualification criteria is required for individuals auth-or ized to use EXTENDED RWP.s ?

As a Shift Supervisor on the 12:00 mid

.

to 8:00 an shift, hoe could you verify that an individual meets this criteria ?

(0.75)

GUEETION 7.0c (1.50)

-During your 4: 00P to12:00 mid. shift there is an unexplained ( sloe ) de-crease in Primary Containment pressure. Other than a suspected loss of Pri-cary Containment, what additional events could have caused this pressure c2 crease, sExplain) ?

(0.25 for the everit, 0.25 for the r eason)

1

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

__. _ _ _

l

'

,

.

..

7.

. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~

~~~~~~~~~~~~~~~~~~~~~~~~

R b5bE 55C L bbOTR L

~~~~

____________________

QUESTION 7.07 (2.50)

During the 4:00 pm. to 12:00 mid. shift, at rated power, you receive two clarasi 1.0ff GAS line high pressure.

2.0ff GAS line high temperature You notice that the condenser vacuum is decreasing.

a. Based on the above indications / conditions. WHAT HAS OCCURED. and what add 2tional automatic actions can be expected ?

( 1.25 )

b.

Based on the above situation list your immediate oper ator actions.

( 1.25-)

QUESTION 7.08 (2.50)

Pipe Break Inside Dryvell i Ccncerning N1-SOP-29

,

a.

Under what conditions can ti.e automatic controls of an Emergency Core Cooling System be placed in its manual mode ? (be specific)

(1.00)

b.

Think about the overall purpose of this procedure

; List at least three (3) operational functions, eith respect to the Core and its Containment,that you are expected to achieve to assure tnat the HEALTH and SAFETY of the public is protected (1.50)

.

GUESTION

' 09 (3.00)

.

According to Procedure N1-SOP-3,Feedwater halfunction(Decreasing FW Flow;;

c.

What immediate actions would you take if feedwater flow rapidly decreas-ed due to a loss of the Shaft Feedvater Pump.

(1.50)

. Due to the above transient R *.. Vessel level is decreasing at a very rapio rate. As the Shift Supervisor, at what Vessel level voulo you direct your operators to depr essurl:e the vessel ?

(0.50)

c.

Is it necessary to close the MSIVes during this transient ? ( E::pl a in ) (. 4 )

d.

List-three (3) conditions that could cause HPCI to automatically init-1ste as a result of this transient.

(0.60)

.

_

, _.

-

..

.-

.

7.

! PROCEDURES.- NORMAL, AE:N O R M AL, EMERGENCY AND PAGE

~~~~ 5656E655C5[~C UTR6E~~~~~~~~~~~~~~~~~~~~~~~~

R

____________________

0UESTION. 7.10 (2.00)

Concerning Procedure SOP-15 Malfunction of the Control Rod Drive Sys.;

,

0.

Durin3 a power ascention, ( RX. power approx.30 %

),

the selecter control rod starts to drift. What Automatic responses, ief alarms /in-dications, would be affected ?

(1.00)

b.

What critetton-is used to define a control rod as being inoperable.iO.5)

c.

How.could you verify that a control rod has become uncoupled ?

(0.50)

__

.

U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION

>

FACILITY:

N1NE MILE POINT

_________________________

REACTOR TYPE:

E:W R - G E 2

_________________________

DATE ADMINISIERED: 85/03/12

_________________________

EXAMINER:

LANGE, D.

NAME:

,

__

INSfRUCTIONS*

____________.

Use separate paper for the answers.

Write answers on one side only.

Staple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.

Examination papers will be picked up six (6)

hours after the examination starts.

% OF CATEGORY

% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

III__ I !!!

___________

________ 8.

ADMINISTRA1IVE PROCEDURES.

_

CONDITIONS. AND LIhITAIIONS 25.00 100.00 f0fALS

________ ______

___________

________

FINAL GRADE _________________~

All work cone on this examination is av own. I have neither giver no- *eceived sic.

____________

______________

l

-

.

U.S.

NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY:

NINE MILE POINT

_________________________

REACTOR TYPE:

BWR-GE2

_________________________

DATE ADMINISTERED: 85/03/12

_________________________

EXAMINER:

LANGE, D.

_________________________

NAME:

_________________________

INSTRUCTIONS *

_____________

Uso separate paper for the answers.

Write answers on one side only.

Staple question sheet on top of the answer sheets.

Points for each qv9otion are indicated in parentheses after the question. The passiiig Stade requires at least 70% in each category and a final grade of at lacet 80%.

Examination papers will be picked up six (6)

hours after the e):anination starts.

% OF CATEGORY

% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

_['5.00 100.00

________ 8.

ADMINISTRA1IVE PROCEDURES,

______ ______

___________

CONDITIONS, AND LIMITAfIONS 25.00 100.00 T0fALS

________ ______

___________

________

FINAL GRADE _________________%

All ucrk cone on this e::a m i n a t i o n is my own. I have neither 31ven nor received aid.

____________

_

______________

.

,

,-

-,

--

r

-

,

c f

.

8.-

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMI1ATIONS PAGC

__________________________________________________________

QUESTION 8.01 (1.00)

What action is required if the Core d a ::i mum Peaking Factor e::c e e d s the Dasign. Total Peaking Factor ? (Explain the reason for your answer)

(1.00)

DUESTION 8.02 (3.00)

Concerning Refueling Oper ations i s.

List sir (6) methods available to verify proper ruel bundle orientst:cn.

(1.50)

b. Consider the alarm-REFUEL INTERLOCK-located on the ROD E. LOCK h0NITOF PANEL. List two (2) conditionsr( including interlocks ),that*this alarm could be indicatins ?

'(1.00)

c. Under normal operationseptior to fuel handlingeprocedure N1-OP-34 has a prerequisite which states, The Fue1 Pool key lock switch on the '

C*

panel shall be placed to the Refuel position ehen handling fuel or it-adiated fuel casts. What is the pur,>cse of doing this ?

(0.50)

vhESTION 8 03 (2.00)

List six (6) pnysical interlocks and/or administrative conditions which avst be satisfied prior to starting a Reactor Recirculation Posis.

(2.00i QUEETION G.04 (2.50)

At Nine Mile Pointe Unit-1 the placement of electrical jumpers, changin-or removal of leads and the blocking of relays ar e p er f or med ONLY under Frecific soministrativley controlled circumstances i t.

List five ( 5 of thes? circumstanses.

. ;_

b. Tne places.ent anc restoration of jumperseblocks or lifting of lesos shall be a c c o m p l i s h e d b y --------------------; Complete the sentence with the appropriate personnel and the administrative requirements they hase to adhere to.

(1.00)

,.

,

.'

.

'____LADMINISTRATIVE' PROCEDURES, 8.

CONDITIONS, ArJD LIM 11 A110NS PAGE

______________________________________________________

f GUESTION 8 05 (2.50)

c. _The lowest point at which the Reactor Water Level can normally be mon-itored is approx.-----------

below minimum normal water level, or

above_the top of the active fuel. (fill in the appropriate levels)

(1.00)

b. What is the si 3nificance of the above Vessel Location Tap.

(0.50)

c.

The actual Low-Low-Low, water level trip pointe ( - 10' ),is 6 ft.-3 in.

below the minimum n o r nia l water level,(elev.302-9'). A General Electric service inf or a.a ti on letter resulted in raising this trip set point 2C

to conservatively account for what possible adver se condition's) and

?

<1.00)

resulting descripencies GUESTION 8.06 (1.00)

At Nine Mile Point, procedural controls will assure that the IRh scram is oointained up to 20 % Flow. How is this accomplished ?

(1.00)

.

GUESTION 8,07 (1.50)

Concerning the Limiting Condition For Operation ( Operability Requirements >

as described in-Technical Specifications. When a system. subsysteni, t r a a r..

I component or device is determined to be inoperable solely because its emer-Soncy power source is inoperable, or solely because its normal power source is inoperable, it may be considered operable for the purpose of satisfving the r equir ements of its applicable Limiting Condition for Operation only if the folloving two (2) conditions are satisfiec. LI5T THESE TWO C0dDIi10NE.

(1.50)

GUESTION 9.0E

'2.501 A s sun.e the Reacter has fallec to scram (

50F-3; b :. msnval er automatic

-

ceans. What CRITERIA would you use to determine when to initiate the STANDBY LIQUID CON 1ROL SYSTEh ? (Be specific).

(2.50)

.

_

.'

.

8.-

ADMINISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIONS PAGE

__________________________________________________________

GUESTION 8.09 (2.00)

In accordance eith the Technical Specifications, the reactor was scrammed due to Suppression Chamber water temperature being greater than 110 degrees F.

The reactor is now in HOT SHU1DOWN, suppression Pool Cooling is ON, and Suppression Chamber water temperature is 92 degrees F.

Using the attached section of Technical Specfications can you commence a startup ?? (Fully E:: Plain)

(2.00)

GUESTION 8.10 (2.00)

a NOTE: USE THE ATTACHED SECTION OF THE TECHNICAL SPECIFICATIONS TO z

ANSWER THE FOLLOWING QUESTION.

FULLY REFERENCE ALL SECTIONS YOU USE. *

During a shift turnover, with the plant operating at 75% power, you are informed that the BI-Weekly Closure Surveillance Test has exceeded the ma::imum allowable extension interval and will be performed on your shift.

Halfway through the test, ONE' Outboard MSIV FAILS te meet

the specified closing time. In accordance with the Tech Specs:

o. What situation e::ists due to the surveillance test being octside of the test frequency schedule.

(0.75)

b. What actiens must be taken due to the fact that the hSIV has failed it's closing time test ?

(1.25)

GUEETION 8.11 (2.50)

Concerning NhF-Technicial Specificaticn DEFINIT 10N5, ansser tne t-11oving either TRUE or FALSE.

a.

The TEST INTERVALS that are specified in Tecn. Spec.s are or.1v valid outing perlocs of power oper ation and do not apply in the event of en-t e re c e d Station Shutocen.

0.50'

. A r. OPERATING CYCLE is that portion of s t a t i o r, oper stlon c.e seer tne e r. d cf one operating cycle and the end of the ne: t operating c 3 cle.

(0.50)

. CORE ALTERATION is the addition. removal. relocation, or other manual movement, including control rod movement with the control roo drive hvd-reulic system. of fuel or controls in tne r eactor core.

(0.50i d.

A FIFE WATCH PATROL is a patrol that requires an area with Inoperable fire protection equipment to be inspected at least every four t4) hrs.

(0.50)

e.

A TRIP SYSTEh is an arrangement of sensor s and av::illar y equipment re-e. ed *: rr e-P.s r-d * snsm-t r -: I--

t ime's- *: sr.

- a : r-

.

c n a r.nfi f or tne purpose of satisfying a component response, t0.50

-

.

3.

ADMINIS1RATIVE PROCEDURES, CONDITIONS, AND LIMITA1 IONS PAGE

__________________________________________________________

OllE S T ION 8.12 (2.50)

During your shift, 12:00mid - 8:00AM, with the plant at 100 % steady state power, one of.your operators informs you that the closed position indicat-

.icn light for MSERV 6 6 is not indicating. The problem is determined not to b2 a burned out light bulb but a maintenance problem. You are informed that this proble anot be worked on until 8:00 am, when maintenance personnel at:

../. e.

..

, can the plant continue to l!s i ng tr. : attached sections of Tech. Specs oper ate' under this condition ? Fully reference all sections of T.S related to this situation, giving a brief description and bases for anv actions vou would take.

(0.50)

-

y

.

hg$.

f.

'

-

.-

5.

. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 5.01 (2.75)

A.

NO,

Thermodynamic efficiency is a comparison of energy in versus energy out. CO.53 The increase in generator output resvited from decreasing-the smount of steam diverted to the HP FW heater. [0.5]

This condition requires additional energy output from the reactor to r aise FW temp to the same saturation temp as before CO.53 Thus, thermodvnamic eff leiency of the plant has gone down. [0.5] note delta T across the heater would have caused more extraction steam to have been removed from the turbine.

E. Reactor Power,(CMWT), increase, (0.25), due to the core inlet temp-decreasing thus causing more heat to be added to reach the same core onit enthalpy (0.50)

.

REFERENCE HnP. Operations Tech. Module o, Chapter 46.

L ANSWER 5.02 (3.00)

e.

Adds negative reactivity E0.25] due to the increase in reevtr on leakage - Moderator temper atur e coefficient. [0.503 b.

Adds negative reactivity [0.25] due to the increase in neutron capture in the fuel - Doppler coefficient. [0.503 c.

Adds positive reactivity [0.25] due to the decrease in neutron Moder ator temperature coefficient. [0.50]

leakage

-

d.

Adds negative resetivity [0.253 due to the increase in neutron leakage - Void coefficient. CO.50 REFERENCE HnF.--

Fesctor Theorv ANSWER 5.03 (2 25)

s. VOID COEFFICIENT (0.50). adds negative reactivity (0.251 b.

FUEL TEMP. COEFFICIENT (0.50), cdds negative resctivit- (0.25)

c.

n0DERATOR TEhPERATURE COEFFICIENT (0.50), adds positive reactivity (0.25 I

REFERENCE NhP.

Reactor Theory

.

L

.

.,

.-

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUlDS, AND PAGE

_-__

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 5.04 (2.00)

At 100 % power.

(0.75)

At 4 % power, you are at operating pressure but at low feed 9ater flow rate. HPSH is los due to T inlet being high. As power increases, Pump inlet temperature is r educed due to mi::ing in the downeomer. T-inlet is lower so P-sat. at inlet is lower, therefor HPSH is higher.

(1.25)

REFERENCE NMP. Thermodynamics and fluid flow.

ANSWER 5.05 (2.00)

First, convert psis. to psia. by adding 14.7 psi. Then,refering to the 532 des.F eteam tables; 900 psia.

=

488 des.F 610 psia.

=

532 des.F - 488 des.F =44 des.F / half hour, or 88 des./hr (1.5?>)

NO. The cooldown limit of 100 deg.F/hr has not been e::c eeded.

(0.50s REFERENCE NdP. Module 4 9 Part 2 Properties of Matter, pg. 17 thru 23

.

ANSWEP 5.co (1.50)

The cold eater injection SUE:-COOL 5 the moisture separator drains. This suo-cooling prevents TWO-PHASE FLOW in the moisture separator dr ain tant pip-2ng to the feedvater heaters.

(1.50)

EEFERENCE snF rG -O F - 31, T a n c e n.

ConiFouno Reheat Turbine.

Re..

?.

pg4 75.

ANSWER 5.07 (2.25)

Floerste is proportional to speed; (speed)2-15 proportional tc head:

'

ispeed)3-is proportional to power; From the above relationships, since the discharge head oecreased by a fac-tor of four (4),

the folloeing new figures m. ply; 15 000 GPM (0.75)

CAPACITY

=

150'. RPd

'O

~5~

? PEED

=

= cecreases tc (

1. E > the original va;oe.

POWEE

,

...

.

G.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

g3

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

REFERENCE NMP-Fluid Llow. and BWR-Technology Basic Pump Law Relationships.

ANSWER 5.08 (2.50)

After about 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months the SDM will increase to a FPr o:

3 % as a resv]t of

.

Xenon peaking. After 20 to 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months the SDh will be back at 1 % as Xenon decays to its equilibrium value.From this point on the SDM decreases eith the Yenon concentr ation and the Reactor u111 go critical if no other action is taken.

(2.50)

-REFERENCE

<

NMP..Oper. Tech. Module 1-Ch. 16, pg.128 and Module 1,Ch,7,ps.55.

ANSWER 5 09 (2.25)

c.

13.4 Kv/ft-To limit elad plastic strain to 1 ~ during transient oper-ation.

(1.00)

b.

1.

Radial position in the core. (O.75)

2.

A::ial position in the core. (0.25)

(any 3 at 0.?U each)

.

9.

Position of the rod in the bundle. (0.25)

4.

hanfacturing tolerances. (0.25)

LJ W A "

c.

To protect the fuel claddins during a DRY-OUT.LOCA

- 2200 deg.F Limit.

Due to heat radiation problems in the fuel nodes.

(0.507 PEFERENCE NMP-OPER. TECH. Module 10, pg. 40 to 45.

AW5 WEE 5.10 (1.504

-

TRUE. (0.50) If a fuel bundle dries out during a LOCA

, the edge and corner rods could-disipate more heat easily than central rods. fhe edge and corner rods can radiate heat away from the fuel bundle while the central rods rad-late much of their heat to other central rods. (1.00) ( The primary heat transfer mechanism is thermal radiation.)

'

REFERENCE HnP-Oper. Tech. Module 10, ch.47 pg.t-41

_

.

__

-

.

,.

.

r 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

-

ANSWERS -- NINE MILE. POINT-85/03/12-LANGE, D.

ANSWER 5.11 (1.50)

YES. (0.50) To maintain the removal of non-condensable gasses produced from activation products and noble gasses produced the decomposition of water

,

in the fuel and leaking into the coolarit via. cladding cracks. (1.00-REFERENCE rJ M P. Oper. Tech. Module 5 pg. 41 to 45.

ANSWER 5.12 (1.50)

This accident was analy ed for three reactor operating modes. The HOT STAND E:Y condition results in the most severe condition. (0.50) This is because of the higher reactivity worths than at full power,and because of the larg-er concentration of fission products than at cold conditions.

(1.001-REFERENCE rJnF - Oper. Tech. Module #12, pg. 59.

l r.

I-f l

!

I l

!

I.

!

,

l'

L

,

t I

-...

,

_._.

,_

_. _

_, -.. -

. _ _

-__..

, __.__

.

e

..

,

..

.

6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUM

_______________________________________________ENIATION PAGE

_______

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

ANSWER 6.01 (3.00)

A. I H'o DW pressure (.25), 3.5 psig(.25), one out of two taken 244*

~(.2)

Low-Low-Low gat.er level (.25), -10 inches (.25), one out of

two taken tu,ss,et.2)

/

Time delay timed out(.2). 120 seconds (.2),

lo21e is Jr of 2 '. 2 > l' 2 C'"

B. hts secor ds after the 120 second timer started (.25),

if the

-

T4dtM O* t *

Primary valve il was not open, its backup valve #2 vov1d open(.25)

-C. Acoustic monitor (0.25)

Temperature elements (0 25)

REFERENCE Operation Technologyr Module 4, Part 8, ADS ANSWER 6.02 (2.50)

Lov-Low-Lov condenser vacuum (

7'

Hg ).

Bypassed when : 600 psig. and mode switch is in STARTUP or REFUEL.

2. Main steam line high radiation. ( 5: NFPB

).

3.

hsir, steam line high flow ( 105 psid or 120 % steam flow ).

.

Reactor lov pressure. ( 850 psig. with mode svitch in run ).

5. High stea temperature in the steam tunnel. ( 200 des.F ).

Loe-Lov vessel level.

( + 5'

).

( 0.25 for each correct ans. )

o.

1. Restricts discharge and protects vessel internals from lar ge D/P. (.32

Provides signal for M S I'.'

c l o s u r e.

(0.33'

3. Provides steam flow signal to FWCS. (0.33i

,

EErERENCE NnP. Oper. Tech. Module #

2.

Chapter o

.

6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

______________________________________________________

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

ANSWER 6 03 (3.00)

c. The flow control valve is prevented from opening following a HPCI initiation si 3nal until feedwater pump discharge pressure is sensed.

(0.75)

b.1.

Remain in-operation.

(0.25)

appron.(if 10 see

.25)

The idle FW Pump vill start and be up to speed in 3.

The FW control sys. will switch to single element control it had been in three element. (0.25)

Mf'}

'#9'

4.

Attempt to maintain RX. level atf65 (0.25)

'

5.

f 71)'

(0.25)

automatic start ik blocked by aux. oil pressure less thanljeed pugo The c.

1.

8 psig.) (0.33)

17 p. The r#eed pumps will trip if suction pressure drops belov 200 psig. or if aux. oil pressure drops below 3 psis.

(0.33)

3.

The feedwater booster pumps automatic start is blocked by suction pressure less than 35 psis.

( 0. 3 3 'e REFERENCE NnP. N1-OP-46, HPCI. and OPER. TECH. Module #4, part 19.

ANSWER 6.04 (1.50)

-

1.

Used to calibrate individual LPRM detec tc* s. (0.50) to map the core 2. Used by the prosess computer (0.50) to determine hCPR and local heet five conditions. (0.25)

REFERENCE NMP.- Oper. Tech, Module 3 part 5.

ANEWER 6.05

'.00)

A differential pressure sensor is used to confirm the, integrity of the CORE SPRAY piping within the reactor vessel ( between the inside of the vessel and the cor e shrovd).

sprl?* piping. a Delta p To continuously monitor tre integrity of the core switch measures the pressure difference between the 19o loops. which is offectively the inside of each Core Spray sparager pipe, just outsice of the Rx vessel.

If the core spr ay spar ager is intact, this pressure difference vill be zero.

If integrity is lost, this pressure differential will include the pressure drop across the steam seperator. Alarms at 5 psid

-

'he ::m'r:: - en

.~C REFERENCE 9nc. Ocer. Tech. Moovle #

4.

c's - t 10 Cc e 5 : r r-p?.4-1

.

6.-

PLANT SYSTEMS DESIGN, CONTROL, AND INSIRUMENTATION PAGE

______________________________________________________

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

ANSWER 6.06 (2.25)

O.

Nish temperature- > 83 des.C (0.25)

Low pressure - < 17 psis.

(

Low system flow < 442 spm.

(0.25)

Immediately reduce Rx. Recire. flow to minimum, in an b.

NO, (0.25),

attempt to prevent a scram. (0.50)

c.

System flav is regulated to msintain inlet pressure lov enough to pre-vent water from entering the stator windings in the event of a leak.

( if a leak develops hydrogen elll leak into the cooling vater).

(0.75>

REFERENCE NMP. N1-OP-44, Gen. Stator Cooling Water Sys., pg. 1-5

.

ANSWER 6.07 (3.00)

&

t. Lev-*

A'a"

s.

1.

Fuel Zone Ind.

7. High Level Alarm.

2.

Lo-Lo-Lo-Ro 3.

Hi/Lo, Lo-Lo Rosemount.

9. Low Level Scram.

correct ans.)

4.

Narrow Range GE-MAC.

10. Lov-Low isolation.

(1.2C for full 5.

Flange Level GE-hAC.

11. Instrument Zer o.

credit )

6.

Turbine Trip Signal.

Low-Low-Lov ADS.

b. This sys. is initiated by manual or automatic tripping of all five (5)

Rectre. MG set drive motor breakers.

(0.50)

c.

1. Cor e Level / Torus hon. Sys. Trouble annunciator will alarm. (0.35)

The digital level indicators on panel

F elll flash on and off.

'

(0.35)

d.

Drywell pre dete (0.20)

2.

Resctor P essure (0.20'

2.

Reactor builcing temper ature

'O.20)

REFERENCE NnP. Oper. Tech. Module 2, Ch. 2r pg.16,17,a flg.2-.'

..

.

6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

.

- ANSWER 6.08 (2.75)

'n.

1.

Continuity lights so out.

2. Alarm indication.

3. Milliamp meters on back of 09-03 indicate everent flow to firing ekt.

4.

Decrease in power.

5.

Selected pump has red light indicating pump is running.

6.

SBLC pressure reactor pressure.

<

7.

SELC tank level decreasing.

( any sin at 0 25 each )

b.

YES. (0.25) If the SBLC. tank level approaches reto sO.50) or the SBLC pump begins to loose discharge pressure. (0.50) This is indicsted by fivetuation of pump amperage and press. Need SS approval.

REFERENCE NMP Oper. Tech. Module 4-81 and N1-OP-12

.

ANSWER 6.09 (2.25)

_

s.

A relief valve in each core spray loop provides a Minimum Flow Recire-ulation path to the TORUS when the pumps are running at shutoff head.

(0.75)

b.

1.

Pressure on the suction side of the Core Spray pumps. (0.25)

ALARM on panel K at(2.5 psis) LOW SUCTION PRESSURE.

(0.25)

2. Pressure is sensed downstreag of the flow orifice.

(0.25)

at(225psis.) CORE SPRAY LOOF LOW PRESSURE. (0.25)

ALARM on panel H

Pr es sur e is sense on the disch.of the core spray topping pumps.(0,25 Provides ALARn at 445 psis CORE SPRAY LOOF HlGH PRESSURE and remote

'

indication of core spray system pressure.

(0.25)

REFERENCE NnP.Oper. Tech. Module # 4 pg. 50-61 and N1-OP-2 ANSWER 6.10 (2.75)

1. Main Steam Isolation Valves CLOSE (0.25) causing a REACTOR SCRAn on valve position. (0.50)

Feedvater FCV - locks-up as is (0.25) causing a LOSS OF HPCI FLOW CON-l TROL. (0.50)

,

3. MAHE -UP valve to Condenser - CLOSES (0.25) causing HOTHELL LEVEL to DECREASE. (0.50)

RX. Bldg.CLC-TCV, OPENS (0.25) Temp. decrease on closed loop vater.(0.25)

,

REFERENCE Instrument A3r Failure, pg.4 2,3 4 riF N1-500-5

.

l-I

, -,

---r

--

,,

-r----~r

,

- ~ -, - - -,. -,, - - - - - - - - - - - - - - - - -

,

.

-

.

.,

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIM 11 A1 IONS PAGE

__________________________________________________________

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

m ANSWER B.01 (1.00)

Action required-The scram trip setting must be adjusted (0.25)

.

Reason -

To ensure that the LHGR limit is not violated for any combination of M1 PF and REACTOR CORE THERMAL POWER.

(0.75)

REFERENCE NdP-1, Tech.5pec. baser f or 2 1.2'

I ANSWER 5.02 (3.00s o.

Fuel assembly serial 4 are readable from the associated control rod.

2. Lv3s on the fuel assembly bail handle point at the associated contr ol rod.

Channel spacer buttons are above the associated control rod.

4.-Channel fastener spi ir'3 C11Ps are above the associated cor.tr ol rod.

5. Gadolinive rods have longer end plugs which protrude thr ough the upper tie plate.

6.

Overall core symmetry.

(0.25 for each cor r ect ansver)

b.

1. The mode switch is in refuel with one control r od withdr awn. (0.25i

.

An attempt to move the refuel platform eith a fuel element over the core will result in de-ener 91:ing of the hoist motor. (0.25)

2. The mode svitch is in refuel eith the refuel platform loaded and over ~

the cor e 8(0.25) This condit2.on inserts a rod block to prevent c or.t t c l rod eithdrawal. (0.25)

c. This places the fuel pool h2sh radiation mon 2 tor, on the refueling br i dge. or. the emergency ver.tl etion c ir cuit talarms at 1000 sir /hr),(0.50)

EEFEEENCE Operation: T e c hr.o l c g y, nodule 2, Chapter 3.,

anc N1-OP-34, R e f v e l i r.g Pr c : d.

ANSWER B.03 (2.00)

1.

The 86 relays are reset.

2.

The Svetion valve is open.

5.

The discharge valve is shut.

-

4.

The discharge bypass valve is open.

5.

The scoop tube is reset.

6. Local hA station in manual and set at 20 %

.

. Perrte-Re : c :... n e-trort net be shutdown 2 hes. dur:ng power o c,e e s t i o r.

.

6.

D o r.o t C%CeGo 7'S KW power C orts v&p t i oI.. (any sir at 0.33 each)

-_.,

__ _

.

S.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

_________________________________________________________

ANSWERS -- NINE MILE POINT-85/03/12-LANGE.

D.

RFFERENCE NMF - N1-OP-1 Rev. 23, pgf 12.

ANSWER B.04 (2.50)

c.

1.

In accordsnee eith approved u.aintenance procedur es.

2.

I r.

test or surviellence pr ocedur es.

'

In sn approved aiodification.

4.

To fstilitste the coridvet of tests and checks.

5.

To pr eser ve the ssf ety, function, and/or integritv of the rtstion er systen.s 6.

For non-routine activities during refueling outage.

( any 5 at 0.30 each)

6.

Appropriste personnel----> A licensed operator OR a qualified instrument

'

technician or electrician.

(0.25)

Admiinistrative requirements i 1.

The sction talen must be verified by a second sveh OUALIFIED person.

2.

Only with pr ior appr oval of the Station Shift Super visor.

O.

Onl eith the knowledge of the Chief Shift Operator.

( three (3) r equir ed f or ful1 eredit 0 0.25 et i REFERENCE NMP. i APN.-7A. Placement of jumpers / blocks or lifting of leads.

ANSWEF S.05 (2.50)

s.

Seven

<'s ft. eleven (11)

In.

beloe m i n i m u s.

no r s.a l ester level. (0.50 Four rc' ft. eight (Si in. above the top of the active fuel.

(0.50i c.. This is the locstion of the Reactor Vessel Tsp for the Low-Lov-Lov water level instrumentation.

(0.50 %. 7N57

'C ~. a

,W TA, E M J 42.$. Set s*nN7.

. Th;s trip pc2nt esi r 2sec 20 intnes to conser.stive1y secount fo-per-sible dif f er ences in ACTUAL to INDICATED water level (0.50) due to pot-entiallv HIGH DRYWELL TEMPERAlURES (0.50).

REFERENCE NHF-Tecn. Spec.

2.1.1, Bases for Fuel Cladding-Safety Limit, pg. 15

- ANSWER 8.0c (1.00)

lhls is sceomplished by keeping the Reactor Mode Switch in the Startup pes-ition until 20 % flow is e::e e e d e d (0.50) and the APRM's are on scale (0.50s.

I

-

,

-.-

.

B.

ADMINIS1RATIVE PROCEDURES, CONDITIONS, AND LIMI1A1 IONS PAGE

__________________________________________________________

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

REFERENCE NMP. Tech. Spec. 2.1.2 Bases for Fuel Cladding. ps-16.

CNSWER 8.07 (1.50)

1.

Its Corresponding NORMAL or EMERGENCY power source is operable. (0.75)

2.

All of its r edondent s y s t e ni ( s ), subsystem (s). tr ain( s). component (si and devicris) ar e - oPer able.

(0.75)

REFERENCE NhP. Tech. Spec. Sec.

3.0, Opersb112ty Requ2rements. pg. 25.

ANSWER B.08 (2 50)

c. Increasing Reactor Power (0.50); or if two or more adjacent control rods (0.50) or thirty or more control rods cannot be inser ted below position 06 (0.50) and reactor ester level cannot be maintained (0.50) or if sup-pression pool temper atur e cannot be maintained (0.50).

REFERENCE NMF-N1-SOP-32, revt4, pg. to.

ANSWEF.

8.09 (2.00)

NO.(.75.' Operation shall not be resumed until the pool temperatur e is re-duced to below the power operation limit specified within the shaded ares cf figurc 3.3.2a when doencomarcr submergence is : 4 ft, or figure 3. 3. 2t-wher downc ommer soonergence i s :, 3 ft.

(1.25)

REFERENCE Technical Speczfication 3.3.2 specfication (E)

T e c h r. i c E l Specification 3.3.2 F29ures alb.

ANSWER 8.10 (2.00)

c. The LCO for the MSIVs are stated in Section 3.2.7.

Two main steam line isclation valves per main steam line shell be oper able with closing times gr eater than or equal to 3 sees and less than or equal to 10 sec.

Oper ating outside of this specification is a violation of T.S. and therefore is a reportable occurance.

(0.75)

b.

Action

'b'

of 3.2.7 states that with the one MSIV INOP due to exceeding the alloesble closing tisie, the affected steam line shall be isolated

.

If the problem is not corrected, initiate an orderly shutdown within ene heu-Enf * a.' e the eacte- : r.

ecid shutdoor c;th:n 10 h e p t.

.N

'

.

8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE

__________________________________________________________

ANSWERS -- NINE MILE POINT-85/03/12-LANGE, D.

RFFERENCE NMP-TS-sec.3.2.7 Reactor coolant system isol. valves.

c

' ANSWER-8.11 (2.50)

01 TRUE b.

FALSE c..

FALSE d.

FALSE

e.

FALSE (0.G< for each correct ans. )

REFEF:ENCE NMP-Tech. Spec. definitions.

ANSWER 8.12 (2 50)

With no indication of valve position and possible logie or maintance prob-les e::isting and unable to be corrected, the valve has to be considered in-operable according to the definition of operability. (0.50)

The LCO for pressure r elief systems-Solenoid Actuated Relief Valves

, as requires all s i:: valves to be operable when you stated in section 3.1.5

,

ar e at oper ating temper ature and pr essure. (0.25)

The pr imary bases is for depressurl stion to allov for full flow core spray operation in the ovent of a small line breat. (0.25)

The LCO for pressure relief systems-as stated in section 3 2.9 requires that only 5 of the e valves need to be oper able. (0.25)

The primary bases for this is to limit reactor over pr es sur e below the lov-ost safety velve set point in the event of r apid reactor Isolation. (0.25)

The two conflicting.LCO's are determined for different bases requir ement s.

The conservettve LCD for section 3.1.5, where all 6 valves shall be oper-oble, is the most limiting and should be adhered to. (0.50) The action tc be fellowed is the stne for both LCO's-----

E. e 110 ps13 or less and sa -

urat:en temperature, respectively, eithin ten (10:' hours. (0.50)

!

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E u_

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PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE

~

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RI55 LE55CIL C5sTR5L

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.____________________

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 7.01 (2.50)

t o. The respective supply breaker will not close until voltage has decayed to 20 % of the normal value. (0.25). This delay precludes r e-ener gi:ing motors on PB-11 or PB-12 when their voltage may be cons 2derably out of phase with the incoking voltage (0.75)

.

6.

R-1014 is normally closed and R-1011 is normally open. (0.25). This pro-vides a better balance of loading between T-1015 and T-101N (0.25,

.

R-1014 and R-1011 are interlocked such that onlv one can be closed at a time. (0.25) This interlock can be defeated by a control switch in the control room. (0.25)

e.

Only teo (2) of the three can be closed at any one time. (0.50)

REFERENCE NMP. N1-OP-30 Rev. #6,pages le 5,

.

ANSWER 7.02 (3.00)

a.

Control Rod position.

4.

Steam flov or temp.

2. Recirculation flow.

5. Feedwater flow or temp.

3.

Reactor pressure.

6.

Turbine Generator load.

7.

Bypass, relief or safety valve flow.

(any s i:: at 0.25 for each correct ans.)

.,

2.

APRM alarm., 3.

Rod Block s l a r s..

(0.25 each)

b.

1.

LPRM a l a r a, c.

Reduce power to 80 % of the power level prior to the change using Rectre flow. (0.25) This e111 prevent bundle overpower (0.25) and overloadinc of the other feedwater heater strings. (0.25)

R E F E F:E r.'2 E NhF. N;-SOF'-ic ( pg,2E3 )

,

,--

, ~ -

c

. - -

, -

,

-w n-- - - - - -.

.--:,

f:

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PROCEDURES - NORMAL, AE: NORM AL, EMERGENCY AND PAGE

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~~~~ ID5dL55iEXL E5sTR5L R

____________________

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 7.03 (2.50)

c. A combination of lo-lo reactor vessel water level and high drywell pres-sure (3.5 ps23.)

(0.50)

b.

The rav eater pusip arid the containaient spray p unip in the affected locp should be secured. (0.25)

The loop suction and discharge valves should be closed. (0.25)

c.

TRUE (0.25)

.

d.

1.

Suff2c2ent evidence shoes that the system is not performing its intended function. (0.50)

2.

Continued operation vill prolong or produce an unsate condition. (0.50)

Shutdown of the system will be at the direction of the Station Shift Supv.

(0.25':

REFERENCE NriP. N1-OF-14, pages 1 thru 5

.

ANSWER 7.04 (2.50i s.

Panels 4 11 & 12 are povered from. RPS continuous power motor generstors 162 and 172 (0.50)

.

b.

Scrat the Reactor 2. Tr ip the 345 KV breakers and trip machine 3.

Verify the R ::. S c r a s..

4.

Verifv the Turbine Tripped 5.

Initiate emergency cooling 6.

Opertte Manual Isoletion Switches - Vessel Isolation Channel 4 l' 1 anc 4 12 on the console l verify MSIV E R ).. Water CleanvF-Isclatacr..

7.

Sound Fire Alara, and 2dentify area if F.nown.

B. Verify HPCI initiation.

( eight correct anseers at 0.25 each )

REFERENCE NMP., N1-50P-11 Control Room Evaevation. and 0F Tech. Module # 4-87,8c

_ _

_ _ _ - _ _ _ _ _ _ _ _ _ _ _

.

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PROCEDURES - NORMAL, AE:NORhAL, ENERGENCY AND PAGE

~~~~ I656L5EiEAE E5UTR6L

~~~~~~~~~~~~~~~~~~~~~~~~

R

____________________

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 7.05 (3.00)

O.

1.

Contamination levels greater than 10,000 dpa/cm2

.

2.

Airbor n radioactivity requiring the use of respirator y equipment.

3. _ Neutr on r adiation e::posure.

High r adiat ion ar es e rit t l e s.

5.

Unkr oere t.nditions.in an area to be entered.

6.

h a i rst e ria n c e of equipment, coretr ols or i n s t r ume rit a t i ori i r. R a d i a t i or.

eress or High radiatior. areas.

( any five (5) at 0.25 for each correct a r. s. i b.The LEADMAN must be qualified in Radiation Protection. (0.25)

7.

The LEADMAN or a qualified alternate must be ON THE JOE: Al ALL TIMES

.

3. The LEADMAN may be responsible for ONLY ONE RWP at a time.

(0.25 for each correct ansver)

c.

FALSE - The SSS sigreature is only r equir ed upori a s sve of the RWP. (0.25)

d.

Only these people who meet the f ollowing criteria i 1. Gus11fied in Radiation Pr otectiori.g (O.20)

s knowled geg(Du al. Cour s eMonitor1res) by taking and pa 2.

Demonstrated their Self a comprehensive Self honitoring (0.30)

The names of those who are approved to use the EXTENDED RWP are contained in e LOG in the Control Room.

(0.25)

REFERENCE NMP. Procedure 4 RP-2 r Radiation Work. permit Procedure, pg.,2,3,6,~.

ANSWEF 7.06 (1.50)

Three events;

De reste i ri Supp. Pocl level: As level decreases the nitr ogen e-ut t e-copy a lar ger volume, r esulting in decreased Drywell pressur e.

2.

Increase in barometric pressure; The Dryeell pressure instruments are all refrenced to attesphere ther ef or e an increase in atmospheric press.

causes a decrease in D/W r ss. gc#

2.

(, decr ease ir. D/W or chilled wate'r tempi An incr ea se in c oolirig e sp-scity i re the D/W from the U/W coolers and chillers vill decrease D/W pressure.

(Other * e a s o ria b l e ans. accepted if sub s t ariti a t e d )

( 0.25 for each event; 0.25 for each reason )

REFERENCE NM'-

m

.

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

?

RA5i5L55iCAE E5NTR L

'~~~~

~~~~~~~~~~~~~~~~~~~~~~~~

j

ANSWERS -- N1NE MILE POINT-85/03/12-LANGE, D.

ANSWER 7.07 (2 50)

c. EXPLDSION in the Air Ejector discharge piping. (0.50)

-

Automatic Actionsi

.

1. Valves BV-76-12/13 closeand(off gas floe goes to cero.)

(0.50)

.

'

2. Reactor Scr am at 23 H3 10 25)

6.

1. Reduce reactor load by decreasins recire. flov.

2.Close as s i r.

stean. supply valve to Elr ejectors a nd n,1 x i r.3 j e t.

3. Insert control rods per rod pattern until vacuum decreases to near scram point.

4. Manually scram the reactor.

5.

Initiate emergency condensers, as necessary to remove the decay heat.

6. Inf or m statior per sonnel of conditions.

7.

Notify Plant Superintendent.

( 7 correct ans. at 0.1785 es.)

REFERENCE NMP.

SOP-18

,*Euplosion in the Air Ejeetron Disch. Pipine'

( Symptoms / Automatic Actions / Oper ator Actions )

ANSWER 7.08 (2.50)

a.

1.

Misoperation in automatie is confirmed by at least two i nd epende ret pr ocess parameter indleations.

(0 50)

O. Core' cooling is assured AND this procedure,(SOP-29), directs you.(0.50

'

b.

1.

haintain Core Cooling.

2. Lis.it the release of off sas radiation.

3. 8'l a c e the Reactor Core end Containment i r. a SAFE STABLE condition.

4.

keep the Torus bul6 temp. eithin specified SAFET) l i s. i t s.

(any three at 0.50 each)

NOTE; Other specific Operational objectives related to SAFETY accepted.

REFERENCE

'.

NMP. N1-50P-33, Pipe Break Inside Dryvell., Cautions, Limitations and over-all purpose and objectives. ps4 1-5

.

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PROCEDURES-NORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~ I65dE6656hL'E3 IR5E------------------------

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ANSWERS -- NINE MILE POIN1-85/03/12-LANGE, D.

C:NSWER 7.09 (3.00)

c.

1.

Shift mode switch to refuel.

2.

Check all rods are fully inserted.

3.

Observe power level deereasing.

4.

Check for HPCI operation. Erisur e that both moter dr iven f eedwater punip s are running.

5.

Chcek that the emergency condensers are An operation.

6.

Check that the Core Spray pumps ar e running and recirculating bete to the torus.

(0.25 for each correct ansver)

b. Low-Low-Low Level (-

10 inches)

(0.50)

c.

Yes, To conserve coolant inventory. (0.40)

d.

1.

Runout flow of 1.9 :- 10(6) or 3800 spm. (0.20)

2.

Turbine Trip (0.20)

3.

Low Rx. water level (0.20)

REFERENCE NMP. 50P-3, and Simulator scenerlo Objectives i 142

.

ANSWER 7.10 (2.C0)

a.

1.

LPRh output ir. the vicinity of the drifting rod.

2.

Position Indicatiori for that rod.

NOTE: The ROD DRIFT alars vill not come in due to the rod that is driftireg is the ores selected.

The RWM alars. vill not come in due to betreg 25 % power.

(1.00)

'b.

A C o n t i c '. Tod Wh1Ch CEnnot be boved with C o rs t r c '. tod di i v e p r e s s '..t e.

5, c.

Control Rod overtravel alarmi ( Ann. vindov F2-6,

Computer point 80-12)

when the control rod has been fully withdr awn.

(0.50)

REFERENCE NMP., 50P-15, halfunction of CRD syste r

r

.

.

...

.

-

.

U.S.

NUCLEAR REGULATORY COMMISSION REACTOF. 05'ERATOR REOUALIFICATION EXAMINATION FACILITY:

NINE MILE POINT

_________________________

REACTOR TYPE:

BWR-GE2

_________________________

DATE ADMINISTERED: 85/03/11

_________________________

EXAMINER:

BERRY, J.

_________________________

NAME:

_________________________

INSTRUCTIONS

____________

Uso separate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each quOstion are indicated in parentheses after the question. The passino grade requires at least 70%

,

% OF-CATEGORY

% OF CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

-*

_II_.00 100.00

PROCEDURES - NORMAL, ABNOFnAL.

____ ______

___________

________

ENERGENCY AND RADIOLOGICAL CONTROL

_'_______ ______

25.00.

100.00 TOTALS

___________

________

FINAL GRADE _________________%

All work done on this e::a m i na t i on is my own. I have neither 31 v o r, not received aid.

____________

______________

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

R Si5t55iEAE 55sTR5E----------------------~~

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CUESTION 4.01 (3 00)

In accordance with procedure NI-SOP-32, Failure of Reactor to Scram, what s i:: (6) immediate actions vov1d you take to redvee power and insert all control rods in an ATWS situation)

(3.0)

GLIESTION 4.02 (2.00)

Dascribe, in general, the four things you would do to reset a high pressure ecolant injection (HPCI) initiation, assuming that the initiation signal has cleared.

(2.0)

DUESTION 4.03 (2.00)

c.

Why is ari oper ator instructed to " reduce reactor power to 80%

of the original power level with Reactor Recirculation flow'

BEFORE removing a feedwater heater string?

(1.0)-

b.

When two condensate booster pumps are required, the preferred linup is with $11 and $13 running; eben one booster pump is required, #11 or 413 should be in service.

Why is this pre-ferredS (1.0)

00ESTION-4.04 (3.00)

s.

Assuming the CSO and the NADE vere able to accomplish NOTHING in the say of securin3 the station orlor to an esacuation. n c i. It the reBCtor shutdown AND hov is the shutooPr. veritteo!

tour answer should-include where the CSO and NADE proceed tc and their subsequent actions.

(1.0)

b.

After verification of a turbine trip, the SSS is to proceed tc powerboard 11 & 12 What actions are to be perforneo at pover-board 11 & 12o (1.0)

c.

How can RAW WATER be supplied to feed the reactor 5 (1.0)

.

. -, -

.

..,---_.,,e

--.

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h

.-

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.

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY ANO PAGE

R D bL GECdL'CUNTR6L

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~~~~~~~~~~~~~~~~~~~~~~~~

____________________

_

QUESTION 4.05 (5.00)

Concerning Procedure SOP-19 (Unexplained Reactivity Change);

-

a.

L i s t 51 >:.(6) plant parameters / indications that should be checked if an une::pl a ined r ea c t i vi ty change should occur at rateo power.

(1.5)

b.

Deperiding on the magnitude of the reactivity change, list thr ee slarms that may be inittsted. (PRIOR to a reactor scram)

(0.75)

c. If this reactivity change is a result of decreased temperature, due to a loss of a feedwater heater string, what is your immediate action and what two (2) adverse conditions are you trying to protect against? (.75)

,

utlESTION 4.06 (3.00)

Concernin3 Procedure N1-OP-14, Containment Spray System:

a.

What two (2) signals are required to automatically start the containment spray pumps?

(0.5)

b.

What. action should be~taken following a confirmed high radiation alarm in the containment spray raw water system?

(0.5)

e.

The containment sprav Rav Water Pumps must be manuall. started by the control room operator? TRUE or FALSE.

s0.25)

c. This proceovre directs you not to manually overrace or shut this svstem cown after an auto. initiation unless two conditions are ni e t. What are these two conditions and who is authori:ed tc a ss.e this cecision 1.25-QUESTION 4.07 (2.00)

During the 4:00 pm to 12*00 midnight shift, at rated power, vou r eceive two alarms ;

1.Off GAS line high pressure.

+

2 0ff GAS line high temper ature You notice that the condenser vaevum is decreasing.

o. Based on the above indications / conditions, WHAT HAS OCCURED?. and what

'

s e d i t. : e n s ! sutomstic sctions csr os eurecte5i

1 b.

bliec On tnE EbcVe situation list you? lameCIEtt opFT E tor EC1;ons. (1.01

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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

--- EK5iBE55iEEE 55 sis 5E------------------------

-GUESTION 4. 08 (2 50)

~

Ccncerning N1-SOP-29, Pipe Break Inside Drywelli e. Under uhat conditions can the automatic controls of a r. Eniergency Core Cooling System be placed in its manual mode? (be specific)

(1.0)

Think about the overall purpose of this procedurei

' List at least

thr ee (3) operational functions, with respect to the Cor e a rid its Containment, you are expected to acnieve to assure that the HEALTH arid SAFETY of the public is protecteo.

(1.5)

uuESTION 4.09 (2.50)

According to N1-SOP-3, Feedvater Halfunction (Decreasing FW Flava, s.

What immediate actions would you take if feedwater flow rapidly decreased due to a loss of the Shaft Feeduster Pumps

<1.5)

b. Due to the above transient RX. Vessel level is decreasing at a very rapid rate. As the Sh.ift Supervisor, at what Vessel level would you d. rect your operators to depressuri:e the vessel (0.;5)

c.

Is it necessary to close the MSIVs during this t r an s l er,t >

6: 0.25)

d.

List thr ee (3) conditions that could cause HPCI to automatically initiate as a result of this transient.

(0.5)

00EETION 4.10 (2.00)

Cor.cc-ning Procedure 50F-15 ns1 function ci tne C c-r 01 Arc Dri.e 5. sten s.

Dur ing a pouer ascension,tRx. power appr o2:. 30

%,,

t.,9 selectec control rod starts to drift. What Automatic responses. le; alsres/in-dications. 9ould be affected-(1.0)

b.

What criter2on is used to define a control roc 5s being inopersolet(C.5)

c. How could you verify that a control rod has become uncoupled; to.5)

x-

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PROCEDURES NORMAL,-ABNORMAL, EMERGENCY AND PAGE

~5'

RdDi5E 55iCEE C 5 s T R5E------------------------

~~~~

____________________

ANSWERS.-- NINE MILE POINT.

-85/03/11-BERRY,~J.

l ANSWER f4.01 (3.00)

1.. Place. Mode Switch in shutdoen (This inserts an additional scram signal)

2.

Trir recarevlation pumps 3. Fully insert. control rods using ' Emergency Rod In'

4.

Reset RPS trip. Manually scram the reactor-5. Individually scram rods from

'N'

panel 6.

Isolate ard vent scram air header locally (0.33 each)

REFERENCE N7-SOP-32, Rev. 4e ps 8-JCU-1 4 ANSWER 4 02 (2.00)

1.

Verify a) Feedwater flow on 411 and #12 is 1.5 million Ibm /hr.

b) Reactor low level trip is clear (.5)

2.

Seitch feedeater pump 411 and $12 h/A stations to manual

.(.5)

,

3.

Adjust the manual outputs'until the deviation meter s or, the 911

M/A siations are nulled.

(.5)

4.

.F

'st the 'Feedwater Return to Normal After HPCI' pushbuttori on the reactn-control console.

(.5)

REFERENCE N1-OP-lo. pg. 10 EDH-32 ANSWER.

4.03 (2.00)

a.-

This vill prevent the other feedwater heater strings from being over loaded and will pr eclude possible over-power of the nuclear fuel. Also power increase due to increasec Inlet subcooltreg.

(1.0)

b.

This preferred lineup will preclude a system feedwater di stur -

bance due to the loss of poverbosed $101.

a h wesurer HPCI (1 0)

....levility.

REFERENCE

" -OF-1:.

e I.

1:

I: L-:;-

m

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_.

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'4.

PROCEDURES - NORMAL, ABNORMAL, EhERGENCY AND PAGE

~~~~ db5bLb65 BEL E5UTRUL'~~~~~~~~~~~~~~~~~~~~~~~

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R

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

ANSWER 4.04 (3.00)

c.

The CSO proceeds to shutdown panel #12 and trips MG set 141 The NA0E proceeds to shutdown panel 411 and trips MG set 131.

'J e r i f -

ication is the 'All Rods In* vhite light on their respective panels.

(1.0)

b.

Verifies that a condensate and feedvater booster pump are opera-ting and starts feedwater pump 411, if HPCI has failed to initiate (1.0)

Also manual transfer of PB-11&12 if auto transfer fails e.-

By-installing an available spool piece between the feedwater sys-tem and the fire protection vater system.

(1.0)

Also cross-connect to cor.tainment spr ay r a9 water through inter-tie valves M cerc Fg PEFERENCE N1-50P-11, ps. 3-5 EDH-318 ANSWER 4.05 (3.00)

a.

Control Rod position.

4.

Steam flov or temp.

2.

Rectrevlation flow.

5.

Feedwater flow or temp.

Reactor pressure.

6.

Turbine Generator loac.

7.

Bypass, relief or safety valve flow.

(any s 12: at 0.25 for each correct ans.)

3.

Fcd Elect sistm.

i0.:5 es:h.

2.

AFFm alarm.

c.

1.

LFon alstm

.

c. Redvee power to SO % of the power level prior to the change using Rectre flov. (0.25) This vill prevent bundle overpover (0.25) and overloading of the other feedwater heater strings. (0.25)

REFERENCE NhP. N1-SOP-19 ( P3,2E3 )

l u

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~ kb5bEU555 L C TRUL

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ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

m ANSWER 4.06 (3 00)

n.

A combination of lo-lo reactor vessel eater level and high drywell pres-sure (3.5 psig.)

(0.50)

6.

The rav vater pump and the containment spray pump in the affected loop shovid-be secured. (0.25)

,

The loop svetion and discharge valves should be closed. (0.25)

c. TRUE (0 25)

.

d.

1.

Sufficient evidence shows that the system is not performing its intended function. (0.50)

~2.

Continued operation will prolons or produce an unsafe condition. (0.50)

Shutdown of the system will be at the direction of the Station Shift Supv.

(0.25)

REFERENCE NMP<.N1-OP-14, pages 1 thru 5

.

ANSWER 4 07 (2.00)

a.

EXPLO5 ION in the Air Ejector discharge piping. (0.50)

Automatic Actions; 1. Valves BY-To-12/13 elose and off gas flow goes to :ero.

(0.50)

2. Reactor Scram at 23 Hg.

(0.25)

b.

1.:ecvee reacter loac by cecreasing recire, f1:..

2.Clote main steam s up p 1;..alve to air ejector s anc mining je.

'

3. Insert control rods per rod pattern until vaevum decreases to near scram point.

4. Manually scram the reactor.

'

G. Initiate emergency condensers, as necessary to remove the decay heat.

6.

Ir. form station personnel of conditions.

Notify-Plant Superintendent.

( 7 correct ans. at 0.1735 es.)

REFERENCE NMP.

SOF-18

,' Explosion in the Air E ection Disch. Piping'

f i.nst:.na A sten it.: A::1;r s Ope at:- 4:.;:r t

-

--_________ _ -_- _ _

-

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PROCEDURES -. NORMAL, ABNORMAL, EMERGENCY AND PAGE

~

~~~~~~~~~~~~~~~~~~~~~~~~

R365UL 55CIL E UTR6L

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____________________

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

ANSWER 4.08 (2.50)

o.

1.

Misoperation in automatic is confirmed by at least two Independent process parameter indications.

(0.50)

2.-Core coolins is assured AND this procedure,(SOP-29), directs you.(0.50 b.

1.

haintain Core Cooling.

2. Limit the release of off sas radiation.

3. Place the Reactor Core and Containment in a SAFE STABLE condition.

4.

Keep the Torus bulk temp. vithin specified SAFETY limits.

(any three at 0.50 each)

NOTEi Other specific Operational objectives related to SAFETY accepted.

PEFERENCE

'NMP. N1-SOP-33, Pipe Break Inside Dryvell., Cavtlons, Liettations and over-all purpose and objectives. p34 1-5

.

-ANSWER 4.09 (2.50)

a.

1. Shift mode switch to refuel.

2.

Check all rods are fully inserted.

3.

Observe power level decreasing.

4.

Check for HFCI operation. Ensure that both noter driven feedwater pumps are running.

5. Cheek that the energency condensers are in operattor,.

o.

Check that the Core Spray pumps ar e running and rectr eviating back t: the t:rvs.

i0 25 for each correct anseer/

b. Low-Low-Low Level (-

10 inches)

(0.50)

c. Yes, To conserve coolant inventory. (0.40)

6.

1. Runout flow of 1.C n 10(6) or 3800 3pm. (0.00)

Toroine Trip (0.20 3. Low R 2:. water level (0.20)

REFERENCE NNP. 50P-3, and Simulator scenerio Objectives 4 la:

-.

. - -

-

- -

-

_,

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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

~~~~ 565bl6555hL~bbNiRUL

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ANSWERS -- NINE MILE POINT-85/03/11-BERRYr J.

ANSWER 4.10 (2.00)

a.

1.

LPRn output in the vicinitv of the drifting rod.

2.

Position Indication for that rod.

NOTE: The ROD DRIFT alarm elll not come in due to the rod that is drifting is the one selected.

The RWh alarm will not come in due to being > 25 % power.

(1.00)

b.

A control rod which cannot be moved with control rod drive pressure.(.5)

c.

Control Rod overtravel alarm; ( Ann. window F2-6,

Computer point B0-12)

when the control rod has been fully withdrawn.

(0.50)

d M 4 M * N p ^^ tem.

REFERENCE 50P-15, Malfunction of CRD sys NMP.,

.

-.

-.'

.

i U.S.

NUCLEAR REGULATORY COMMISE!ON FEACTOR OF EF ATOR REOUALIFICATION EX AMINATION FACILITY:

N1NE' MILE POINT

_________________________

REACTOR TYPE:

PWR-GE2

_________________________

DAfE ADhINISfERED: 85/03/11

_________________________

EXAnINER:

E:ERRY.

J.

_________________________

NAME:

_________________________

INSTRUCTIONS

____________

Uso separate paper for the answers.

Write answers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing gr ade r equir es at least 70%

least 50%.

Enamination pape.

-

% OF CATEGORY

% OF CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

_'I

3.

INSTRUMENTS AND CON 1ROL5

__

___________

________

_

23.00 100.00 TOTALS

________ ______

___________

________

IINAL G R A D E _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ *E All work cone on this examination is my own. I have neither given nor received aid.

,

____________

______________

,

A

7-.

.

?

.

(

-3.-

INSTRUMENTS AND CONTROLS PAGE-2

.____________________________

QUESTION 3.01:

(3.00)

c. What are two signals (including setroints) that would cause the Feed System to shift to the HPCI Mode of operation and HOW would the resvits of a LOCA initiate each?

(1.0)-

-B.

When the Feed Pump is idle or not in the HPCI Mode,-the HPCI

. coret r o l l e r setroint is blocked. This setpoint is not applied to the' controller until

?

____ ____ is produced at the ____?____.

(1.0)

C.

Why is the HPCI controller setpoint blocked initially?

(1.0)

GUESTION 3.02 (1.50)

For each of the fo11ceing systems, list what TYPE of radiation d2tector is used and what AUT0hATIC ACTIONS occur when the conitors trip. (NOTE: If no auto actions occur, indicate so.

Assunie lineups ar e appropriate for auto actions to occur.)

A.

Air ejector offgas B. RBCLC C. Refueling Bridge (1.5)

.

OUESTION 3.03 (2.50)

Concerning the fuel :one level detectors:

a.

Undet what conditions does the value displayed on the fuel

one instrument cose from the FUEL ZONE level transmitter (Include system initiation signalts) in vovr answer)

(2.0i c.

Wnt: Indi:1 tion does the control room operator have that reference les flashing is occuring in the fuel zone level instrument?

(0.5)

.

.,

..

.

3.

INSTRUMENIS AND CONTROLS PAGE

____________________________

QUESTION 3.04:

(3.00)

Aasume the REACTOR LEVEL CONTROL SYSTEM is being operated in 3-element ccntrol using reactor level detector channel

'11'.

Reactor power is at 95%, STEADY STATE.

For each of the instrument or control signal failures listed below. 51A1E HOW REACTOR LEVEL WILL INITIALLY RESPOND (increase. decrease, or remains constant) and BRIEFLY EXPLAIN WHY in terms of what is happening in the Level-Control System immediately-following the failure.

(FOR EXAMPLE, your answers should include the following detail',

'Causes reactor level to decrease due to a steam flow / feed flow error signal, steam flow < feed flow, resulting in a closure

. signal to the feedwater control valve.')

NOTE: A block diagram of the Feedwater Control System is on the following page for your use.

a.

12 FEEDWATER FLOW transmitter FAILS HlGH (1.0)

b.

Channel

'11'

REACTOR LEVEL detector signal FAILS LOW.

(1.0)

c.

LOSS OF CONTROL SIGNAL to 413 FEEDWATER CON 1ROL VALUE.

(1.0)

.

GUESTION 3.05 (3.00)

For each of the following, state whether a ROD BLOCR, HALF-SCRAn, FULL SCRAh. or NO PROTECTIVE ACTION is generated for that condition.

NOTE:

IF two or more actions are generated, i.e.

rod block and a half-scram. state the most severe, 1.e.

half-scram.

Assume NO oper-stor actions.

s.

AFFr. 11 Downscale, noce Switch in RUN (0.es b.

<4 LPRM inputs to APRM 15, Mode Switch in STARTUP (0.6)

c.

Both Flov Conv. Units Upscale (>107% flow), Mode Svitch in RUN (0 6)

d.

APRn 12 and 16 Upscale, Mode Switch in STARTUF (0 61 c.

Ma n Steam Line 111 ISOLATED. Mode Switch in RUN (0.6)

GUESTION 3.06 (2.00)

Describe fully hov Reactor Ov11 ding Closed Loop Cooling (RBCLC)

tooper atur e is regulated as the hect load on the svstem

est+i.

..;

.

N

c_

-

- ~ '

.

-3.

INSIRUMENTS AND CONTROLS ~

PAGE

____________________________

QUESTION 3 07 (2.00)

O. Will a normal transfer of an RPS Bus from its normal to its

'

emergency r-ower supply cause any protective action? WHY or WHY NOT?

(1.0)

B. Will a shift of a REACTOR TRIP Bus from its normal to its emergency power supply.cause any protective action? WHY or WHY NOTS (1.0)

OtlESTION 3.08 (2.50)

A.

The Main Steam Line flow restrictors are used to develop main steam flow si3nels. What are two (2) CONTROL FUNCTIONS provided by this signal?

(1.0)

B. Pressure switches on the main condenser will actuate the PCIS (.5)

to shut the-MSIV's if vaevum decreases to ____?____.

C.

WHEN and HOW is the low vaevum trip of the hSIV's bypassed?

(1.0)

GilESTION 3.09 (3 00)

Concerning Reactor Vessel Level Instrumentationi a.

Using the attached figure 2-1 ( R:' Level Inst.), list ehat level ~ instruments are used to measure the indicated parameters 1 2,

3, 4.

&5 ii.0)

b.

Indicate what level parameter items e through 12 signifv.

(1.25)

c.

What three ( !' )

plant vsrisoles are useo for compensst:en c '. the Fuel Zone Level Inoicators (0.75)

GilESTION 3.10 (2.50)

Concerning the Standby Liquid Contr ol System; s.

Once the SBLCS has initiated, ehat s 2 :: (6) CONTROL ROOn indications could you use to verify that the system is oper ating properly AND injecting into the reactor vessel?

(1.5)

b.

After initiation of the SBLCS is it permissible to shut the system

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U.S.

NUCLEAR REGULATORY COMMISSION

SENIOR REAC10R OPERATOR REDUALIFICATION EXAMINATION FACILI1Y:

N1NE MILE POIN1

_________________________

REACTOR TYPE:

BWR-GE2

_________________________

DAfE ADh1NISTERED: 85/03/12

_________________________

EXAMINER:

LANGE, D.

NAME:

__

_ _ __

INSTRUCTIONS:

_____________

Use separate paper for the answers.

Write ansvers on ene s i d e o r. l y.

Staple question sheet on top of the answer sheets.

Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.

E::a m i na t i on papers vill be picked up s i :- (6)

hours after the enaniination starts.

% OF CATEGORY OF APPLICANT'S CATEGORY

.

VALUE TOTAL SCORE VALUE CAlEGORY

________ ______

___________

________ ___________________________________

-

100.00

__"_____ ______

___________

!

________ 5.

THEORY OF NUCLEAR POWER PLANI OPERATION. FLUIDS, AND 1HERMODYNAnlC5 25 00 100.00 TOTALE

________ ______

___________

________

F!dAL GRADE _________________%

A;; rem con + c r. tais e r s n.i n s t i o n is n. y cen. I have neither 31ven not received aid.

_________-__

__

______________

.

.,

.

U.S. NUCLEAR REGULATORY COMMISSION

-

SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILI1Y:

NINE MILE POINT

-

_________________________

REACTOR TYPE:

BWR-GE2

_________________________

DAIE ADMINISTERED: 85/03/12

_________________________

EXAMINER:

LANGE, D.

_________________________

NAME:

_________________________

INSTRUCTIONS!

_____________

Uno separate paper for the answers.

Write answers on one side only.

Stople question sheet on top of the answer sheets.

Points for each quoction are indicated in parentheses after the question. The passing 3 redo requires at least 70% in each category and a final 3rade of at loost 80%.

Examination papers will be picked up six (6)

hours after the examination starts.

% OF

,

.ATEGORY

% OF APPLICANT'S CAfEGORY UALUE TOTAL SCORE VALUE CA1EGORY

._______ ______

___________

________ ___________________________________

. _ ' _

5*00

___I_00

________ 5.

THEORY OF NUCLEAR POWER PLAN 1 100

____

_

___________

OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 100.00 TOTALS

._______ ______

___________

________

FINAL GRADE

_________________%

11 work cone on this examination i s my own. I have neither Ivon nor received aid.

____________

__

______________

-

e v

-

.-

.,

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUlDS, AND PAGE

____

______________________________________

______________

OUESTION 5.01 (2.75)

NOTE: Answer the following questions from a theoretical standpoint, not frcm a Nine Mile Point system design standpoint.

c. With the plant operating at 90 % power

, e::tr action steam to the fifth stage feedwater heaters is removed.

An engineer. observing that turbine load increased by 15 hWe after the e ::t r a c t i o n steam removal.

concludes that this action has improved the plant's thermodvnamic efficiency (NOT heat rate).

Do you agree eith this conclusient Earlain your ansuer fully.

(2.00s b.

For the above condition, will reactor power (CdW1)

increase, decrease, (Briefly Explain)

(0.75)

or remain the same

.

GUESTION 5.02 (3.00)

Indicate ehether the following eill INCREASE or DECREASE reactivitv during operation AND briefly EXPLAIN why, s.

Moderator temperature increases ehile belou sattuation temperature.

t.75)

6.

Fuel temperature increases.

(,75)

c.

Loss of a feedwater heater.

( 75)

d.

A sudden reduction in reactor primary system pressure.

(.75, OUEETION 5.03 (2.25)

For each of the events listed below. state ehich reactivite coetfic:ent will respond first and if it ados positive or negative resetivitv.

s.. Relief Va3ve o:ening at 100 *: Power iO.

5>

.

!

. Roc or:- at 100 % power.

~5,

..

l c.

Isolation of a feedvater heater string at 75 % power, t.O.75s I

OUESTION 5.04 (2.00i l

Will the Recirculation pumps have more NpSH at 4 % or 100 % power Explain your enswer fully.

s2.OOi i"

C

l l'

L~

.

.,

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

_-_--_----

GUESTION 5.05 (2.00)

During a cooldoen of the reactor vessel from outside the control room,

!

roactor pressure decreased from 885 psis to 595 psis in one half hour.Has ycur reactor cooldoen limit been exceeded ? { show all work }

(2.00)

GUESTION 5 06 (1.50)

The-discharge of the Feedester Booster Pumps provide cold uater inject 2cn into the' moisture separator drain system. From a theoretical standpoint.

why'is this done ?

(1.50)

QUESTION 5.07 (2.25)

Relationships exist between Centrifugal Pump parameters that are refered to as. BASIC PUhP LAWS Use these pump laws to ans9er the follouing.

.

A' centrifugal pump is operating at 30,000 GPM, 3,000 RPM, eith a discharge head _of 200 psig. If the discharge head decreases to 50 psig. shst are the new CAPACI1Y, SPEED, and POWER REGUIREMEN15 of the pump.

(2.25)

DUESTION 5.08 (2.50)

The Reactor has been at 100 % power for 90 davs ehen a Reactor 5 cram occurs The temper ature is maintained at 540 deg.F,and the Shutdown hargin 10 min, sfter the scram is 1% delta K/K Describe what happens to the Shutdown

.

nar gin dur ing tne next three (3) days if the temperature is maintained at 540 degF. 0 (2.50)-

GUESTION 5.OC (2.25)

At NhP-Unit 1,

Fuel thermal safety limits are established for the purpose of protecting fuel clad integrity during accident or steady state power op-erition.

a.

Fr the 8x9 fuel, what is the LHGR limiteano why ess it specificallv.es-tablished ?

( 1. 0 0 '>

b.

The actual LHGR of a fuel rod is dependent upon what ( 3 ) variables ?

(0.75)

. Whv was-the (M)APLHGP the mal limit estab1:shet ?

iO.50'

a

.

.,

.

5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE

.. ______________________________;-_____________________

,

THERMODYNAMICS

______________

QUESTION 5.10 (1.50)

The Central fuel rods are more likely to exceed the 2200 deg.F limit

durinS a LOCA even thov3h the edge and corner fuel rods have higher LOCAL PEAVING FACTORS

.--->

TRUE or FALSE ? ( Justify your answer )

(1.50)

"

i GUESTION 5.11 (1.50)

If tne main Condenser and associated systems cere absolutely A:: TlGH1 would there be any need for the Steam Jet Air Ejectors during full power operation ? (Explain your answer).

(1.50)

QUESTION 5.12 (1.50)

A Rod Drop Accident-is expected to be more severe at'

1.

Power Operation

Hot Standby 3.

Cold Ccnditions i Choose the correct anseer and justify your decision )

(1.50)

.

..

.

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.

TABI.E II-3-1 PEPERTIES OF SATURATED STEAM AND SA'I'URATED WATER (TEMPERATURE)

i volume. ft'/te Entta'py. St.,ac Entrocs. St tre a r M s t.

water Evan Steam water Evap Steam wa:er Evac Steam r"

F psia v

h h

er ry

g r

og

g 3r sig s,

0.08559 C 01602 3305 3305-0.02 1075.5 1075.5 0.0000 2.1873 2.1873

35 0.09991 0.01602 2945 294E 3.00 1073.8 1076 8 0 0061 2.1706 2.1767

40 0 12163 0.01602 2445 2445 8 C3 1071.0 1079 0 0 0162 2 1432 2.159:

45 0.14744 0016 2 2037.7 2037.8 13 04 10681 10E;.2 0C252 2.1164 2.1 26

5 0.177H C.01E 2 1704 S 17D E la C5

0i53 10E34 OI~i; 2D3 ;

2.125, 50

02i6; D"16 3 12 7.6 12*? 6 25 06 IOi ~r 7 10E'.7 00 55 2 0391 2.5:5 l 60

l 02f79 C 0;Ki Efi 3 855 *

35 05 05 3

0: 2 :

C C'45 1.99:

2 Cf45

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65 00 37 1 1;*1. j012?5 1.E5*:

.91: 5. 100

.

' 113 1.271; C.C16 ;7 2ii 4 255 77.95 1C3; 4 113 3 0.1472 1.5;05 1.9577 110 123 1.65:7 C.C1520 20325 203 26 87 97 1025 6 1113 6 {01645 1.7653 1.9I3~ ; 120 133 22233 C C162i 157 22 15 33 97.95 1019.5 1117.5 C 1817 1.72H 1.9112 130

'

140 2.5552 0 01629 122.95 123.30 107.95 1014.0 11220 01955 1.69 :D 1.85 H 140 150 3.718 0 01634 97.05 9707 117.95 1006 2 11261 02150 1.6536 1.8656 ' 150 ISO 4.741 0.C1643 7727 77.29 127.96 10C2 2 113";.2 C2213 1.6174 1.S*E7 160 l

f'

170 5.993 C;;95 62 0:

62 36 137.97 9H 2 1 : 3..: ' c e. 3 1.5E22 18295 170 15; 7.511 00155; 50 21 5* 22 142 0 99 2

3E2 ' C 263

1.54E'

1.E;;l j 1E 190 t

9.3 G 3 C0:55?

40 9-4' H 15554 954 1

42: t 0275'

15;;E 1.793:

19:

l210

! 11 525

6 33 62 33 M 16E 0

977 ~

i ;

29 "

141;;

1 77 4 ' 23C 200

j 14.;23 i CI;t7; 27 E:

2'.E2

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43 7 C209;

45: 5 126:0' 210

!

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'5: 17 970 3

52 5 CJ:21 144 7 1.756E 212 220

1.7442 l! 22 17.1E6 0016'E 23 13 23 15

.5s.23 9652

!!53 4 0.324; 1.42 4 230 20279 C.01685 19.364 19 35; 195 33 958.7 1157.1 0.335E 13902 1.7290 230 240 24.962 0 01693 16.304 16 321 206 45 952.1 1160 6 03533 13609 1.7142 ' 240 250 29.E25 C.0170; 13.802 13 5;9 218 59 9454 1164 3 l 0.3677 13323 1.7000 [ 250

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1~ " 5; 235 H 93. 7

. ; 7: i C.3 H ~

l'ss 16 29 270

,

.

C 0172s 8.627 86:4 249 17 9246 11735 C 093 1.25;;

1.65H [ 280 i

28; j

4~ 23:

lC0173E 290 t 5735 7 443 7 46'

2594 9174 1:76 5 : C 4235 1.2235.

1.6473 ' 290 t

I t 0:745 6 445 6 46s 269 ~

91C0 1;'97 i C 4372 1.1979 1.535; 3 30:

300 1 67 30'

'

i 31:

77.67 C01755 5 639 5 E26 250 0 902 5 1;E~ 5 C4536 1.172f 1.6232 310

.

'

320 89.64 0 01766 4.896 4.914 29C 4 894.8 11852 0.4640 1.1477 1.6116 ! 320 340 117.99 0.01757 3170 3.788 311.3 878 8 1190 1 0 4902 1.0990 1.5892 I 340 I

360 153.01 0CIS11 2.939 2.H7 332.3 862.1 119' 4 C.5161 1.0517 1.567E 360

380 195 73 0 01536 2317 2.335 353 6 844 3 1195* ' O.5 ;6 1.0057 1.5473 ! 3a0 I

I' 400 l 2472E

! 0.0156 1.54 ' -

11530

e l 3751 825 9 12";

C 56E7 0 96:7 1.5274 ! 400

420 33E 75 3 0.C189:

1 ASOE 1.49 7 396 9 806.2 12 3.

C3915 0.9165 1.5350 I 42C 440 32;.54 0.0;926 1.1976 1.2169 419 0 785 123 4 C 6161 0.5729 1 489 i 440

'

460 466.9 0.C196 0.9746 0 09:2 441.5 763.2 120: 5 06C5 C 8299 1.4704 I 460 480 566.2 0.0203 *- C.7972 0.5172 464.5 739 6 1204 1 0.6648 0.787; 1.4518 430 500 f 68: 9 i 0.C20 065 5 0.6749 457.9 714.3 1202 2 : C 659:

C 74:3 1 4333. 500

'

8 520

. 5;25 CC239 0535i C 55M 512 0 657 I

99-C';33 07

13 14: sf 52C 54*

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C: ;i C 4:37

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535 E 657 5

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73 5

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{1133 4

! CC221 03651 03E'1 ' 5624 625.3 1:E' 7 C26 5 0.6132 1.3757 560

!'560,1326 2 l C.0225 0299 03222 ! 529 ;

529 9 1179 C7E'E C5573 1.255: ! 58; 580 60.-

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%

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f a ma vo s/t Cycle efficien:y o (Net-crx out)/(Energy in)

w = mg s = y t + 1/2 at o

E = mc KE = 1/2 mv a = (Vf - V )/t A=AN A = A e'AI g

PE = mgh V = V + at w = e/t 1 = an2/t f

o 1/2 = 0.692/t,/2 s

NPSH = Pin - Psat t./2eff = [(t1/2)(t j'

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p=pe

C SUR = 26.06/T SCR = 5/(i - K,,f

'

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-

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CR.(1 - K,,.' = CR.i'. -

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ce-9.

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' mw = 2.L'

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3 ;,,'. -

.

C en s > = ' pr ::;

Tr. = Z.5: :r a.. e c-,. c ;. %.1.

... c..., e, :/e..

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.

3.

INSTRUMENTS AND CONTROLS PAGE

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

' ANSWER 3.01 (3.00)

A.

1.

Lov level of 53'(.25) due to loss of inventory through the break (.25)

2.

Turbine trip (.25) due to the scram on DW pressure, level or valve position (0.25)

3.

Ic00000 lbm/hr on either motor driven feed pump 10.25) due to feed system tr ying to r e3ain level lost from the break (0.25)

B. discharge pressure (.5),

feed pump (.5)

(1.0)

C. To_ ensure the feed pump starts against a closed discharge valve.

(1.0)

nampart A.

Reactor can scram on High DW pressure, low level,MSL valve position

' REFERENCE OP-46 HPCI JCK-185 ANSWER 3.02 (1.50)

A.

ion chamber (.1) Offgas isolation valve " stack blocking valve'

and drain valve shuts (.3)

B.

scintillation (.1) None(.3)

C, G-h(.1) Nor mal R>, Blds ventilation trips (.3) and Emergency ventilation starts (.3)

REFERENCE Operation Technology, hodule 5 Part 15 Process Red Monitor JCK-185 ANSWER 3.03 (2.50)

c.

System is initiated by all five reactor recireviation motor gener-ator set drive motor breakers tr ipping.

At this point. the dis-played level comes from the triple-lov (Rosemount) transmitter.

Whenever one of the core spray isolation valves opens OR level falls beloe triple-lov (-10'),

the level displayed on the fuel rone meter comes from the fuel :one level transmitters.

(2.0)

b..

' Core Level / Torus Temperature Monitor Svstem Trouble' energi:es.

(0 5)

is-bl* :n

s:vi ec, Al se t, e c: ::: s 1 will flash.

,

es::e- :- r2 ;,t;:s;:-

LE=EVENCE Cp rs ;cn3 Ieennclogy, Moo.Ilt Chap.2. 4.4f.50,51 EDH 0:

-

-_ j

.

.,

.

-3.

INSTRUMENTS AND CONTROLS PAGE

____________________________

-ANSWERS --LNINE' MILE POINT-85/03/11-BERRY.

J.

ANSWER 3.04-(3.00)

o.

Causes reactor level to DECREASE E0.253 due to the Level Control System having a STEAM FLOW / FEED FLOW ERROR, SfEAM FLOW t FEED FLOW [0.53 resulting in a CLOSURE SIGNAL TO THE FEEDWATER CONTROL VALUES CO.253.

(1.0)

'

6.

Causes reactor level to INCREASE [0.25] due to the Level Control System havin3 a LEVEL ERROR, LEVEL SET > INDICATED LEVEL E0.53 resulting in an OPEN SIGNAL TO THE FEEDWATER CONTROL VALVES EO.253.

(1.0)

c.

-Reactor level should REMAIN CONSTANT CC.253 because the 413 FEEDWATER CONTROL VALVE WILL LOCK-UP C.753.

(1.0)

.

Note: Loss of electrical will not lock up valve.

REFERENCE Operations Technology Mod V pt.

6, N1-OP-16 pg. 16 EDH-312 ANSWER 3.05 (3.00)

rod block b.

half-scram c.

rod block d.

full scram e.

half-scram (3 0)

(0.6 each)

REFERENCE Orerat: ens Technelery. mod.III. pg. 20-22,34 EDH-314 ANSWER 3.06 (2.00)

Three valves are automatically positioned to regulate RBCLC temperature: RBCLC heat exchanger inlet valve. RBCLC neat e ::-

changer bypass valve, and RBCLC heat exchanger service water outlet valve (1.0). Initially, the controller modulates the RBCLC inlet and bypass valves (.5).

As the heat load increases further. the controller will cause the service water outlet valve to open more fv11vf.5).

(2.0)

REFERENCE

-

JC Ff -I S 2 Orerstiens Technology. Module 5, Chapter

,

.

.,

..

.

r 3..

INSTRUnENTS AND CONTROLS PAGE

____________________________

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

ANSWER 3.07 (2.00)

A.

No(.25).This power system has the capability to be synchro-ni:ed to its to its emergency supply (.75) prior to trans-ferring therefore there is no irteruption of power.

(1 0)

0. b '(.25) h Sah sc W dilk resul

'

because the RPS trip bus supplied by that MG set loses powerf.75)

(1.0)

REFERENCE NI-OP-48 Motor. Generator Sets pg 2-5 JCK-184 ANSWER 3.08 (2.50)

.

A.

Feed veter control system input (.5)

MSIV closure (.5)

(1.0)

i.5)

'

.B.

7'

Hg C.

Reactor Mode switch in Startup or Refuel (.5) and reactor

. pressure is less than 600 psig(.5)

(1.0)

REFERENCE Operation Technology, Module 2, Part 9, Main Steae Operation Technology, Module 4,

Part 2, Coolant Isolation JCH-187 ANSWEP 3.00 (3.00)

7.

Level Alarm.

2. Lo-Lo-Lo-M. Hign c.

1.

Fuel Zone Ind.

Los Level S : t a n..

F o s e n.o unt E

3.

H1 'L o. Lo-Lc R o s e nie v nt.

b.

4.

Nar-c. Range GE-MAC.

10. Lov-Lov isolstion.

5. Flange Level GE-MAC.'

11. Instrument Zero.

Turbine Trip Signal.

12.' Low-Low-Lov ADS.

c. Drywe11 h e1.. a craepps4are)

lg 6 lg( dm Resetor Pressure Reactor Building Temperature REFERENCE NMP. Oper. Tech. Module 2, Ch.

2, pg.16,17,8 flg.2-1

-

J

.

.-

.

3.

INSTRUMENTS AND CONTROLS PAGE

____________________________

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

ANSWER 3.10 (2.50)

o.

1.

Continuity lights so out.

2.

Alarm indication.

3. Milliamp meters on back of 09-03 indicate current flov to firing cFt.

4.

Decrease in_ power.

5.

Selected pump has red light indicating Pump is running.

6.

SBLC pressure ' reactor pressure.

7.

5BLC tank level decreasing.

( any s i:: at 0.C5 each )

b.

YES. (0.25) If the SBLC. tank level approaches :ero (0.50) or the SBLC pump begins to loose discharge pressure. (0.50) This is indicated by fluctuation of. pump amperage and press. Need SS approval.

REFERENCE Oper. Tech. Module 4-81 and N1-OP-12 NMP

.

a

m-_~

.

.-

.,

.

..

U.S.

NUCLEAR REGULATORY COMMI5EION FEACTOF' OF EPATOR REDUALIFICATION EX AMINATION FACILITY:

'N1NE MILE POINT

_________________________

REACTOR TYPE!

BWR-GE2

_________________________

DAfE ADMINISfERED* 85/03/11

-_________________________

EXAMINER:

BERRY.

J.

_________________________

NAME:

_________________________

INSTRUCfIONS

____________

-

liso separate paper for the answers.

Write answers on one side only.

Steple question sheet on top of the answer sheets.

Points for each question are indicated in pa entheses after the question. The passing 3rede requires at least 70%

% OF CATEGORY

% OF CAfEGORY

,

UALUE TOTAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

_1_00 100,00 2.

PLANT DESIGN INCLUDING SAFETY

____ ______

___________

________

AND EhERGENCi SiSTEn5

.

25.00

'100.00 T0fAL5

________ ______

___________

________

FINAL GRADE _________________%

All1eork done on tnis examination is av own. I have neither 91 ion.nor reteived aid.

____________

__

_______________

J

i.

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.,

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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

OllESTION 2.01 (3.00)

Answer the following with regard to the Emergency Cooling System:

a.

What are two sources of makeup to the Emer3ency Condenser Maker.ip Tank?

(0,5)

b.

What are the initiation signals for the system? Include setroints.

(1.0)

c.

How may the automatic initiation feature be overridden?

.5)

d.

After systen initiation and pressure is < 1000 psis, How is the cooldown rate controlled?

(1.0)

O!IESlION 2.02 (2.00)

WHAT partlevlar harard is involved with extended 12ght load operat2on of a Diesel Generator and HOW is this hazard minimited?

(2.0)

OllESTION 2.03 (3,.00)

Fill in the indicated blanks A-I'

AIR OPERATED i

NORMAL' POSITION I

VALVE ACiION I

SlA1 ION EFFECi VALVES I

DURING OPERATION I

ON LOSS OF AIP f

ON AIR LOSS I

I I

MSIV'=

OPEN CLOSED REACTOF SCRAM OH.

VALVE POSITION EhERGENCY CLOSED A.

B.

t'E N TIL A TION SPENT FUEL THROTTLING C.

RI5E IN LEVEL IF POOL LCV COND. XFER PunFS IN OPERATION TUROINE BLDG THROTTLING D.

E.

CLC - TCV EnERG. CONDEN.

CLOSED F.

G.

COND. RETURN nu TK TIE OPEN H.

I.

9ALVE J

~*

.

2.

-PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

----------------------------------

QUESTION 2.04 (2.00)

If the Main Condenser and associated systems were absolutely AIR TIGHT would there be any need for the Steam Jet Air Ejector s during f ull power operation? (E:: plain your answer)

(2.0)

QUESTION 2.05 (3.00)

A.

What conditions must be met to satisfy the logic for ADS inittstion?

Include setpoir.ts and trip logic arran3ement (2.0)

B.

How wov)d the system respond if MSERU $1 failed to open after proper logic actuation? (All other valves respond properly)

(0.5)

C.

What are-two types of detectors used to provide positive indication of a leaking / lifted RELIEF UALVE? (exclude lights and annunciators)

(0.5)

GUESTION 2.06 (3.00)

Concerning the High Pressure Cooling Injection system (HPCI);

a.

What prevents an idle feedwater pump from star ting and pumping water through a FULLY OPEN feedvater control valve following HPCI initist-loni (0.75)

b.

If a HPCI Initiation occurs with NO LOSS of 0FF-SlTE POWER, state the affect on the following valves, pumps or components.

1.

Condensate and ferdester pumps that are running.

(0.25)

2. Idle feedwater pump.

(0.25; 7; Feeduste-Cont- ] Svstem iO.:5 C.25i

Teedwater p u ng. controller 4 11 (

5.

Feedvater pump controller i 12

.(0.25)

c.

In addition to a HPCI initiation being blocked by protective pump lock-outs, list three (3) additional INTERLOCUS that e111 also PREVENT an automatic star t.

(1.0)

RUESTION 2.07 (2.00)

How is the integrity of ECCS piping inside the reactor vessel verified du-ing norna1 ooeration. In vove a n s i.ie r inc iv.,e : SEN? d? c0INTE. ErECITIC i ~i Er i. WnCEE : :: : ;- I E

.'E R : ~ :E;. Wh: Ii II

'E; IE: eno t..e esp:nse of tne instrumentation to a loss of i ntegr i ty.

(2.0)

--

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.

2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

_______________________________________________________

OtlESTION 2 08 (2.00)

Cencerning the Generator. Stator Coolins Water System; c. What three (3) conditions will cause a Turbine Governor Runback?

(Setpoints are required)

(0.75)

b. Will an automatic Reactor Scram occur on a Governor runbact trip signal?

If yes, from what? If not, ho9 could a subsequent scram be prevented?

(0.5)

c. What is the importance of regulating flow within this system to maintain pressure between 22-28 psi?

(0.75)

QUESTION 2.09 (2.00)

Concerning'the CORE SPRAY system; a. What protective des 13n feature, within the core spray system, allows for running the core spray posips at shutoff head without overheating them.

( E::Pl ein f ully-be specific)

(0.5)

6.

Pressure in the Core Spray piping is sensed in three different places.

List these three sensing points, Indicating what is being sensed and sny automatic actions, alarms, or indications that are provided from R E 5 n r.E U 7M them. (C '* C ~;

,ra (1.5)

GUESTION 2.10 (3.00)

Concerning Refueling Operations; s.

List si:

is; methods available tc e e,r i f t proper fuel bundle orientation.

(1 5)

b. Consider the REFUEL INTERLOCK alarm located on the ROD BLOCK MONIlOR PANEL. List two (2) conditions, including inter loet s.

that this alarm could be indicating?

(1.0)

c. Under normal operations, prior to fuel handling Procedure

.

N1-OP-34 has a prerequisite which states, 'The Fuel Poo!

Ley lock ' swi tch on the

'G'

panel shall be placed to the Refuel position when handling fuel or irradiated fuel casks.'

What is tne purpose of doing this?

t*.!'

l

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

ANSWERS -- NINE MILE POINT-85/03/11-BERRY, J.

. ANSWER 2.01-(3.00)

A.

CST via Condensate transfer system, Fire Water System (0.5)

B.

1080 psig(.33) or

.

loe loe. reactor-level of 5 inchesi.33) for a period of 10 seconos(.33)

(1.0)

C.

Shut the steam supply valve (s)

5>

D.

Alternate opening and closing one condensate return valve (1.0)

REFERENCE Operations Technology, Module 4, part 7, Emergency Cooling System JCH-162 ANSWER 2.02 (2.00)

The secumulation of all in the engine e::haust system called ' SOUPING *i.5, could resvit in a fire (.5).

The engine is operated at some minimum is clean (1.0).huntil load for a period of time 9P inspection shoes that the (2.0)

e::h s u st stack

\\gg KEFERENCE NI-OF-45 Rev 6, pg 11 J CI:- 16 9 ANSWEF 2.03 (3.00)

A.

Open B.

None C.

Open L

O p e r.

E.

Decrease of tempersture on closet loop water F.

Open G.

Emergency Condenser in operation

-

H.

Closed I.

Isolates Make-up Tanks (from each other'.

4 0.35 each )

REFERENCE N1-50p-5, Rev.

3, pg.

2.

3.48 EDH-307 j

,

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

____________________.__________________________________

ANSWERS -- NINE MIt.E POINT-85/03/11-BERRY, J.

ANSWER 2.04 (2.00)

YES..(0.75) To maintain the removal of non-condensable gasses produced from the decomposition.of water,

activation products and noble gasses produced in the fuel and leaking into the coolant via. cladding cracks. *1.25)

REFERENCE NMP. Oper. Tech. Module 5 pg. 41 to 45.

ANSWER 2.05 (3.00)

A.

m High DW pressure (.25), 3.5 psig(.25), one out of two tak en twice(.2)

Lov-Lov-Lov water levelf.25), -10 inches (.25), one out of two taken twice(.2)

Time delay timed outf.2), 120 seconds (.2),

logic is j of 2<

2)

B.

MR5 seconds after the 120 second timer started (.25). If the primary valve #1 was not open, its backup valve #2 vovld open(.25)

C.

Acoustic monitor (0.25)

Temperature elements (0.25)

REFEFENCE Operation Technology, hodule 4,

Part 8 ADS ANSWER 2.06 (3.00)

The flor contrcl

.alve is P r e v e r.t e d f r o n, opening felleving 5 HFC2

initiation s 13 r. s l until feedwater pus.p oischstge pressure is sensed.

(0.75)

b.1.

Remain in operation.

(0.25)

.

2.

The idle FW Pump vill start and be up to speed in appro:

10 sec.(.25)

.

!

3.

The FW control sys. will switch to single element control if it hst been in three element. (0.25)

'

~ -

~~

4.

Attempt to maintain RX. level at 65

._. 0. '.JD.__

(

71 (0.25)

'

c.

1.

The feed pump auton.atic start is blocked by av::. oil pr essur e less than 9 psig.

(0.33)

~ me "ee; ; u n. p s vil1 : : ;.: :-

5... :,1 0. - ; +sso e se:4e ce.ce ;:

si3 cr if av::. cil pressur e drops celow 3 psig.

(0 33i

The #eeovster boost +r comes sotomst:: sts*+

1:

r. l o c e e c e, so:t::r,

'i

* e i s '. 7 E lesi th&r.

7.33 j

"5;3

.

'

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE

_______________________________________________________

NINE MILE POINT-85/03/11-E:ERRY, J.

ANSWERS

--

REFERENCE NMP. N1-OP-46, HPCI. and OPER. TECH. Module 44, part 99.

ANSWER 2.07 (2.00)

A differential pressure sensor is used to confirm the integrity of the CORE SPRAY piping within the reactor vessel ( between the inside of the vessel and the core shrovd).

To continuously monitor the integrity of the core spray pipins, a Delta P ewitch measures tne pressure difference between the two loops. which is offectively the inside of each Core Spray sparager pipe, just outside of the Rx vessel.

If the core spray sparager is intact, this pressure difference vill be ero.

If integrity is lost, this pressure differential will include the pressure drop across the steam seperator. Alarms at 5 psid in the control room (2.00)

REFERENCE NMP. Oper. Tech. Module t 4, Part 10. Core Spray. pg.4-61.

ANSWER 2.08 (2.00)

e. High temperature-83 des.C (0.25)

Lov pressure 17 psig.

(

-

Low system flow 442 spm.

(0.25)

'

6.

NO, (0.25),

Immediately reduce R::. Recirc. flov to minimum, in an attempt to prevent a scram. (0.50)

c.

Svstem flov is regulated to maintain inlet pressure lov enough to pre-vent water from entering the stator windings in the event of a leak, i 1 a leak develops hydrogen vill lea 6 inte the cooling wateri.

(0. 5)

REFERENCE NMP. N1-OP-44, Gen. Stator Cooling Water Sys., pg. 1-5

.

ANSWER 2.09 (2.00)

a.

A relief valve in each core spray loop Provides a Minimum Flow Rectre-ulation path to the TORUS when the pumps are running at shutoff head.

(0.75)

b.

1.

Pressure on the suction side of the Core Spray pumps. (0.25%

ALARM on panel K a: : : *:rs. LOW SUCTION PREESURE.

0.25-

-

ass.'e is semset c o ~ n t ; ; c-a n, of ins ' ' c.. ort 1:e.

,.25

.

ALARn on panel K ev

_.a ps13 CORE SPRAY LOOP LOW PRESSURE. (0.25)

2.

essure is s-e n s e d on the disch.:' tne :or s sc ss ictoim3 ounms.t0

!

r.iaes A AF-c

-,.

ez.7.CO6E 5FRA:.OGF HiGh FREEEUEE snc ren.cte Indication of core spray system pressure.

(0.25)

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2.

PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYS1 EMS PAGE

_______________________________________________________

At>5WERS -- NINE MILE POINT-8 5 /03 /11-E:ERF Y,

J.

REFERENCE NMP.Oper. Tech. Module # 4 pg. 59-61 and N1-OP-2.

ANSWER 2.10 (3.00)

c.

1.

Fuel assembly serial 4 are readable from the associated control roc.

2.

Lugs on the fuel assembly bail handle point at the associated contr ol rod.

3.

Channel spacer buttons are above the associated control rod.

4.

Channel fastener spring clips are above the associated control rod.

5. Gadolinium rods have longer end plugs which protrude through the upper tie plate.

6.

Overall core symmetry.

(0.25 for each correct ansver)

b.

1. The mode switch is in refuel 91th one control r od withdr awr..

(0.25)

An attempt to move the refuel platform eith a fuel element over the core will result in de-energi Ing of the hoist motor. (0.25)

. The mode switch is in refuel with the refuel platform loadec and over the core ;(0.25) This condition inserts a rod block to pr ever.t control rod withdrawal. (0.25)

c.

This-places the fuel pool high radiation monitor, on the refueling bridge,on the emergency ventiation cirevit (alarms at 1000 nr/hr).(0.50)

REFERENCE Operations Technology, Module 2, Chapter 3.,

and N1-OF-34, Refueling Proco.

J

m x

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.

II.

S.

NUCL EAR REGUL ATORY COMn155 ION REACTOP OPERATOR REGUALIFICATION EXAMINATIOrd FACILITY:

NINE MILE POINT

_________________________

REACTOR fYPE:

BWR-GE2

_________________________

DATE ADhINISIERED: 85/03/11

_____________________..___

EXAhINER:

B E F.R i,

J.

_________________________

NAME

_________________________

INSTRUCTIONS

-

____________

Uso separate paper for the answers.

Write anseers on one side only.

Stcple question sheet on top of the answer sheets.

Points for each c,vastion are indicated in parentheses after the question. The passing 3rade requires at least 70%

% OF CATEGORY

% OF CATEGOI.Y VALUE 10TAL SCORE VALUE CATEGORY

________ ______

___________

________ ___________________________________

,

_E[if[_,[I'f 1.

PRINCIPLEE OF NUCLEAF POWER

___________

________

PLANT OPERATION. THERh0DYNAnICE.

HEAT TRAN5FER AND FLUlD FLOW

5.00 100.00 T0fAL5

________ ______

___________

________

FINAL GRADE _________________%

All work done on this examination is my own. I have neither 31 von nor received aid.

____________

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PRINCIPLES OF' NUCLEAR POWER PLANT OPERATIONr PAGE

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GUESTION 1.01 (2.00)

MATCH the Failure Mechanism from colven (1) AND the Limiting Condition from column (2) WITH the associated Power Distribution Limits (a-ci beloe. (A ' letter-number-number' sequence is sufficient.)

a.

Linest Heat Generation Rate (LHGE)

6.

Average Planet Linear. Heat Generation Rate (APLHGR)

c.

Minimum Critical Power Ratio (MCPR)

1 - FAILURE MECHANISM 2 - LIMITING CONDITION 1.

FUEL CLAD CRACKING DUE TO LACK 1.

1% PLASTIC STRAIN OF COOLING CAUSED BY DNS 2.

FUEL CLAD CRACKING DUE TO HIGH 2.

PREVENT 1RANSITION STRESS FROh PELLET EXPANSION BOILING 3.

GROSS CLAD FAILURE DUE TO DECAY 3.

LIMIT CLAD TEMP HEAT E STORED HEAT FOLLOWING TO 2200 F A LOCA (2.0)

GUESTION 1.02 (1.50)

Concerning control rod worth during a reactor startup eith 100%

poal ::enon ver sus a startup with ::enon f ree cor da tions, WHICH STATEMENT IS CORRECT? JUSTIFY YOUR CHOICE.

ii.5)

I '. PERIPHERAL control rod wor th will be LOWER during the 100%

peak ::e non startup than during the ::enon f ree startup.

CENTRAL centrol rod wortn will be HIGHER du ing tne 100% pest

< e no n startUP than during the xenon free startup.

3.

PERIPHERAL control rod worth will be HIGHER during the 100%

peak ::enon star tup than dur ing the :cenon f ree startup.

4. BOTH CENTRAL and PERIPHERAL control rod worth WILL BE THE SAME regardless of core xenon concentration.

QUESTION 1 03 (2.00)

5tseting st 35; on IRm Rense 2 resetor power ine esses for m:nutes

-

n a '; ser.onc yetloc. -nst sill oe tne I h r. Inoi:a;;cri a f *. e t n. i n v i e s What fraction of reactor poeer is this?

(2.0)

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.

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~~~~ 55E56DY5 b5C5I~555T~iRd55fER~556~FLb5D kEUU

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T

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l OUESTION 1.04 (.50)

h l

If the turbine exhaust pressure increases due to a change in circula-

'

ting mater temperature, which of the follouing is true?

(0.5)

a.

E::h a u s t quality increases and exhaust enthalpy decreases b.

E::havst quality decreases and exhaust enthalpy decreases c.

Exhaust quality decreases and exhaust enthalpy increases d.

Enhaust quality increases and exhaust enthalpy increases QUESTION 1.05 (A +tn ?

faI Determine the condenser hotwell subcooling (condensate depression)

if the condenser vacuum is 27.9' Hg. and the condensate tempera-

---

ture is 90 degrees F.

(1,0)

b.

What is one disadvantage of condensate depression)

(0.5)

c.

Hoe does increased condensate depression affect condensate pump net positive suction head?

(0.5)

d.

Give too examples of'hov you, as an operator, can increase condensate depression.

(1.0)

140ESTION 1.06 (2.50)

The reactor has been oFerating at 100% Power for one month eben a scram cceves In which seversl contiol rods FAIL TO FULLY INSEFT.

E ou3r o c. s DC Intert to b?Ing the reactCr subCriticsl at inc time of scram.

!?

reactor moderator temperature is maintained CONSTANT. and control roos are NOT moved, about HOW LONG vill the operator have to valt before he c a r. be reasonably sure that the reactor will remain suberitical?

EXPLAIN.

(2.5>

GUESTION 1.07 (2.00)

Although steam is known to be a poorer heat conductor than water, NUCLEATE BOILING is a BETTER heat tr ansf er mechanism than SINGLE

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PRINCIPLES OF NUCLEAR POWER PLANT JPERATION, PAGE

~~~~iUERE665U 55C5I~5EST iR U5i5R~ Ub'ELU56~ELUU

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QUESTION 1.08 (1.50)

ANSWER THE FOLLOWING THREE MULTIPLE CHOICE QUESTIONS REGARDING CON 1ROL ROD EFFECTS ON CORE POWER.

s.

Withdrawal of a ' deep control rod' generally has an sppreciable effect on total core po9er output because (0.54 1.

The power increase is spread throughout the core bv the relatively high.vold content in the area of withdrawal.

2.

The power increase is large due to the low void conter t present in the area of withdrawal.

3.

The power increase is spread throughout the core by the relatively high moderator temperature in the area of withdrawal.

4.

The power increase is large due to the minimal ' rod shadowins'

present in the area of withdrawal.

b.

Withdraval of a "shalloe control rod' generally changes the power shape, while affecting total core power very little. because (0.5)

1.

The poeer increase is small due to the shadowing effect of nearby control rods.

2.

The power increase is small due to the high void content present in the top of the core.

3.

The relatively large local power increase is off-set by an increase in void content.

4.

The relatively large local power increase II off-set bv a r.

increase in fvel temperstore.

c.

The ' reverse power effect' or ' reverse reactivity effect' occasionally observed when a shallow control rod is wi thdr awr, one or two notches is ove tc*

'O.5 1.

A relatively lar3e local power increase belns off-set b.

a void related power decrease.

2.

A relatively large local power increase being off-set by a moder ator temperature realted power decrease.

3.

A relatively small local power decrease ove to the 'shadewing'

effect of nearby control rods.

4.

A relatively small local power decrease due to increased local doppler effects.

.

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PRINCIPLES OF NUCLEAR POWER PL A'NT OPERATION, PAGE

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00E51 ION 1 09 (2 00)

___________________________________________________I

I CONDENSER i

i 1 89 PSIA I

I 26.07' HG VAC

!

I I

WATER LEVEll*~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~l----WAfERLEVEl.

I HOTWELL i

I i

120 DEG F I

I I

I 8 FT I

I I

I

__________

l I

L leondensatel______ i l__________________________I Pump l______

i i

Using the above figure, calculate:

s.

Pressure at the condensate pump inlet (1.0)

b.

Degrees F subcooling of the condensate in the hot 9 ell (1.0)

SHOW ALL ASSUhPTIONS AND WORK - STEAM TABLES ARE AiTACHED GUESTION 1.10 (1.50)

The moderator temperature coefficient is not considerec to be a significant reactivity coefficient because its affect is 12mited primarily to the reactor startup range. Why is it not considered significant in the POWER RANGE during NORhAL operation" (1.5)

a

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PRINCIPLES OF NUCLEAR POWER PLAN 1 OPERATION, PAGE

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QUESTION 1 11 (3 00)

A fuel pin, ever a period of time, has a uniform coating of corrosion products about 0.001 inches thick buildup on its sur,f a c e.

Assuming that power generation within the fuel pin REhAlNS CONSTANT during the time of the buildup, would you expect the follocing temperatures to increase, decrease, or remain the same during the buildup?

EXPLAIN EACH ANSWER.

a. Fuel temperature.

(1.0)

b. Cladding temperature.

(1 0)

c.

Coolant temperature surrounding the lower por tiori of the fuel pin (prior to the onset of boiling).

(1 0)

pub * *' L.'u QUESTION 1.12 (2.00)

a. If the thermocouple in the discharge of s' Safety Relief Valve measures only the temperature of the dischar geo steam, what temperature would you expect to see on the recorder if an SRV was leaking while the reactor was at 100% power? EXPLAIN.

(Steam Tables and Mollier Diagram are attached.)

(1.0)

b.

Would the steam flow through an SRV be seen on the Control Room stess flov instruments if the SRU FAILED CPEN during oper e t t or.0 EXPLAIN.

(1.0i 140ESTION 1.13 (1.50)

Give ONE undesiracle result of each of the f o 11 c 91 ng. * E.e nio r e specific than ' pump failure');

s.

Operatir.3 a centrifugal pump for extended periods of time eith the discharge valve shut.

(0.5/

b.

Startin3 e centrifugal pump with the discharge velve full open.

(0.5)

c.

Operating a motor driven pump under ' PUMP RUNOUT' conditions.

(0.5)

_ - -.

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TAB 1.E II-3-1

~

MtOPERTIES OF SATURATED STEAM AND SATURATED WATER (TEMPERATURE)

Volume, ft /lb Enthalpy, Stu/lb Entropy, Stu/It, s F

Temp Pre s s.

F psia

-

Steam Water Evap Steam water Evap Steam water Evap b,

h, se sg, s,

h

't

'rs

's e

r

0.08559 0 01602 3305 3305-0.02 1075 5 1075 5 0.0000 2.1873 2.1873

35 0.09991 0.01602 2948 2948 3.00 1073.8 1076.8 0.0061 2.1706 2.1767

40 0.12163 0.01602 2446 2446 8.03 1071.0 1079 0 0.0162 2.1432 2.1594 m

0.14744 0.01602 2037.7 2037.8 13.04 1068.1 1081.2 0.0262 2.1164 2.1426

50 0.17796 0.01602 1704.8 1704 8 18 05 1065.3 1083.4 0.0361 2.0901 2.1262

60 0.2561 0.01603 1207.6 1207.6 28.06 1059.7 1087.7 0.0555 2.0391 2.0946

70 0.3629 0.01605 86E 3 868 4 38 05 1054 0 1092.1 0.C745 1.9903 2 0645

83 C.5063 0.01607 633.3 633.3 4E C4 1048 4 1095 4 0.0932 1.9426 2.035 L to

0.6951 0.01610 455 1 465 1 55 02 1042 7 1103E 0.1115 1.8970 2.0056

100 0.9492 0 016;3 350 4 350 4 65 00 10371 11051 0 1295 1.8533 1.9525 100 110 1.2750 0.01617 265 4 265 4 77.9E 1031.4 1109.3 0.1472 1.8105 1.9577 110 120 1.6927 0.01620 203.25 203.26 87.97 1025 6 1113.6 0.1646 1.7693 1.9339 120 130 2.2230 0.01625 15732 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0.01629 122.98 123.00 107.95 1014.0 1122 0 0.1985 1.6910 1.8895 140 150 3.718 0.01634 97.05 97.07 117.95 1008.2 1126 1 0.2150 1.6536 1.8686 150 160 4.741 0.01640 7727 7729 127.96 1002 2 1130.2 0.2313 1.6174 1.8487 150 170 5.993 0.01645 62 04 62.06 137.97 996.2 1134 2 0.2473 1.5E22 1.8295 170 183 7.511 0 01651 5021 50.22 148 00 990.2 1133.2 02631 1.5480 1.8111 130 193 9.343 0 01657 40.94 40.96 ISB 04 9S4.1 11421 0.2787 1.5145 1.7934 19c 1.7764 f 200 203 11.526 0.01664 33 62 33 64 165 09 977.9 1145.0 0.2940 14E24 210 14.123 0C1671 27 SO 27.62 178 15 971.6 1149 7 0 3091 1453 1.763';.' 21c

-

I 212 14.696 0.01672 26 75 26 83 160.17 9703 1150 5 03121 1.4 c 7 1.7565 : 212

,

220 17.186 0 01678 23 13 23.15 183.23 965.2 1153 4 0.3241 1.4201 1.7442 I 220 233 20.779 C.01655 19.364 19381 195 33 958 7 1157.1 0.3388 13902 1.7290 230 240 24.968 0.01693 16.304 16321 208 45 952.1 1162.6 03533 13609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164.0 03677 13323 1.7000 250

,

260 35 427 0.0170c 11345 11.762 225.76 935 6 11674 0.3E19 1.3043 1 6562 26C 270 41.856 C01715 10.042 10.06";

23595 931.7 1172.6, 03960 12069 1.6729. 270

!

280 49 203 0 01726 E627 8.644 I 245.17 924 6 1173 6 04092 1.2501 1.6599 ' 28:

I 293 57.550 0 01736 7.443 7.460 255 4 917.4 1176.5 04236 12235 1.6473 l 290 300 67 005 0 01745 6 445 6 456 269 7 910 0 1179.7 ! C4372 1.1979 1.6351 30C l

310 77.67 0.01755 5 609 5.66 2500 902 5 11E2 5 fC4506 1.1726 1.6232 21 l

320 ES 64 0.01766 4 896 4.914 2904 E94 5 1185 2 i 04640 1.1477 1.6116 ' 32 340 117.99 0.01757 3370 3.755 311.3 878 8 1190.1 04902 1.0993 1.5892 340 l

360 153.01 0.01E!!

2.939 2.957 332.3 8621 1194 4 0.5161 1.0517 1.567E 360

380 195.73 0.01636 2317 2335 353.6 844.5 119E.0 0.5416 1.0057 1.5473 3ac 400 247.26 0 01864 1.B444 1.8630 375.1 825.9 1201 0 0.5667 0.9607 1.5274, 40C

.

420 305.78 C.0189 -

14805 1.4997 396.9 806 2 12031 0.5915 0.9165 1.50S3 420

!

440 381.54 0.01926 1.1976 1.2169 4190 785 4 12044 C.6161 0.8729 1.4890 44C

!

460 466.9 0.0196 0.9746 0.9942 441.5 763.2 1204.5 0.6405 0.8299 1.4704 460 0.7972 0.8172 464.5 739 6 1204.1 0 6648 0.7871 1.4518 4a0 480 566.2 0.0200

"

500 680.9 0.C204 0.6545 0.6749 4S7.9 714 3 1202.2 0 6890 03443 1.4333 500 520 812.5 0.0204 0.5336 0.5596 512.0 6870 11990 03133 0.7013 1.4146 520 543 962.8 0C215 0.4437 0 4651 536E 657 5 119 3 0.7378 06577 1.3954 54 5 40 1133 4

:::'.

03fi; C.3 5 ".

Ef: 4 62E ~

..E

036:t 06;3:

.3757 l 561 580 1325.2 0.0225 0.2994 03222 559.1 589 9 1179.0 l 03376 l580 C 5673 135501 600 15432 0C236 0.243E 03675 6171 55: 6 1167 7 0E134 05196 13330, sx

G2:

1 1756 9

4-0196:

0.2235 645 9 506 3 1:53 :

54:3

4659 1.3?92 62

640 2059 9 0.C260 0.1543 0.1802 679I 454 6 1133 7 0.8656 04134 1.2821 64C 660 2365 3 0.0277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 03502 12498 660 680 2706.6 0.0304 0.0808 0.1112 758.5 310.1 10685 0.9365 0.2720 1.2086 680 s

700 30943 0.0366 0.0386 0.0752 822 4 172.7 905.2 0.9901 0.1490 1.1390 700 705.5 320S2 0.0505

0.0508 906 0

906 0 1.0612 0,

1.0612 705.5 a

J

EQUATION SHEET

.

..

...

.

f = ma y=

s/t Cycle efficiency = (Network out)/(Energy in)

w = mg-s = V t + 1/2 at

E = mc

KE = 1/2 mv a = (Vf - V,)/t A = AN A=Aeg PE = mgh V

=V

+ at w = e/t x = En2/t

= 0.693/t f

g t

eff = [(t1/2)(t })

NPSH = Pin - Psat 1/2 b

[(t1/2)+(t))

b

,

m a oAV AE = 931 am

-*

I=Ieo Q = k pat Q = UAah I=Ie#

g I = I, 10-x/TVL Pwr = W ah

,

f TVL = 1.3/u P = P,10*"#I*}

'

HVL = -0.693/u P = P e /T t

O SUR = 26.06/T SCR = S/(1 - K,ff)

CR = S/(1 - K,ff,)

,

x CR (1 - K,ff)) = CR (1 - keff2)

SUR = 26c/t= + (s.-o)T j

-T = (t*/o) + [(s - o)/lo)

M = 1/(1 - Keff) = CR)/CR T = L/(o - 8)

M = (I - Keff,)/(I - Kef g ;

T = (8 - o)/(lo)

SDM = (1 - K.,,)/K

_

e..

e r,.

-

IK t' = 10"* secents, o = (Keff-1)/Keff * #Keff df

= 0.1 seconds-I

= [(t*/(T K,f f)] + [s,ff (1 + T))

/

_

I)d) = 1 d P = (I6V)/(3 x 1010)

I)d) 2 =2 2 1d22

1 = oN R/hr = (0.5 CE)/c (meters)

R/hr = 6 CE/d (feet)

NPSH = Static head - h) - Psat Water Parameters Miscellaneous Conversions i gal. = 8.345 lbm.

I curie = 3.7 x 10IC ps Igaj.=3.78 liters I kg = 2.21 lam, Stu/hr l

1 ft = 7.48 gal.

I no = 2.54 x 105 l

Density = 62.4 lbm/ft 1 mw = 3.41 x 10 Stu/hr

I Density = 1 gm/cm lin = 2.54 cm Heat of vaporization = 970 Btu /'bm

'F - 9/5*: + 32 Heat of fusion = la: Stu/lbm

= 5.'9 ' 'F-32 )

' Atm = 14.' :s: = 29.5 in. H;.

F:-

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. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION, PAGE

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--- isiEs557sisiEs-siii iEisiEEE Es5 EEUi5 FE5s

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ANSWERS --' NINE MILE' POINT-85/03/11-BERRY, J.

l l

ANSWER-1.01

.

(2.00)

Failure Mechanism Li mi ted Cor.'di ti on

.

,

A.-LHGP j-

1

3 B.

APLHGR

4 C.

MCPR

2 9 0.33each[

(6 answers-reo,.

REFERENCE-O erations Technology Manual, Module 10 JCK-150 P

,

ANSWER 1.02 (1.50)

i C is the correct choice (0.5)

The' highest Xenon concentration will be in the center of the core CO.333 due to it's being the high flox region from the previous. oper:3 ting per iod

.CO 333.

This will increase the flux levels in the area of the peripheral control rods, thus increasing their worths.CO.333 i

REFERENCE Operattonsl Technology Manual. Module le Chap. 14

,

t

>

~

k g

$

<

e

-

,

-

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

-~~~isEER66 REsiEs-sEEi TEEsiFEE Es5 FEUi6 FE5E

____________________________________________

ANSWERS -- NINE MII.E POINT-85/03/11-BERRY, J.

ANSWER 1.03 (2.00)

P=(Po)(e raised to t/T)

t=7 minutes = 420 seconds T=70 seconds P=(35)(e raised to 420/70)

P=(35)(403.4)'

P=14119% on ranse 2 P=1411.9 on ranse 4 P=141.19 on ranse 6 P=14.119% on range 8 or P=1.4119% on range 10 (1.5)

100 on ranse 10=10% of rated So, 1.4119% on range 10 = 1.4119% of 10%

Fraction power =.14119% of rated or.0014119 of rated (0.5)

REFERENCE OFerations Technology Manual, Module 1,

Chap 10 8 Module 3, Chap. 3 NhF Guestion Bank

I JCV-156 ANSWER 1 04 (.50i d - E :. h s"s t qualitv increases and e:: h s u s t enthalPy incresser 10.51 REFERENCE Operations Technology, Mod. IX, Chap.2 EDH-207 i

.J

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

~~~~ 5ER566YU555C57 EEdi~YRdU5EkR'5U6"FL 56 FLUU T

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' ANSWERS -- NINE MII.E POINT-85/03/11-BERRY, J.

ANSWER 1.05 C3. M T' 2C

--

2' Hs absolute (0.25)

A.

29.9' - 27.9'

=

.98 psia (0.25)

-

2He absolute

=

100 F (0.25)

Tsat for.98 psia

=

10 F condensate depression (0.25)

(1 0)

100 F - 90 F

=

B. Plant efficiency is reduced (0.5)

C. NPSH increases (0.5)

D. Reduce turbine load, Increase cire water flow, raise condenser pressure, Decrease cire water temp., Adjust tempering gate Position, Increase hotwell level (2 required)

(1 0)

REFERENCE Steam Tables, Fluid Mechanics and Thermodynamics Study Guide DNG102

. ANSWER 1.06 (2.50)

60 - 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months (1.0)

It will tak.e approx. 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months for the Xenon to peak and then decay after the scram.- If the positive reactivity inserted by the decay of Xenon is less than the shutdcen reactivity due to rods, then the reactor vill remain suberitical. (1.5)

REFERENCE SNPS Reactor Physics Module - Lesson 11, pages 7-154 to 7-161 Student Objectives 41 &2 ANSWER 1.07 (2.00)

The bubbles caused by nucleate boiling serve to agitate the stagnent fluid film ne::t to the surface, thus improving thermal conductivity.[1.03 Also, each bubble, as it leaves the surface. carries off more energy than is possible by natur al convection.01.03 REFERENCE Heat Transfer and Thermodynamics ANELEF

.CE 1 5; a.

- 1 [0.5J t

- ? PO.":

.. - 1 CO.E3 a

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

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ANSWERS -- N1NE MILE POINT

- 85 / 03 /11 - E'E R R Y,

J.

REFERENCE Roactor Physics Module ANEWER 1.09 (2.001 s.

P = Pate + Pvater column (0.5)

1 89 psia + 8 ft / 2.3 ft / psia (0.4)

=

5.4 psia (0.1)

=

b.

Interpolating from steam tables 1.8 - 2.0 Psia (0.4) yields Psat for 1.89 psia to-be 124 degrees F Subcooling = Psat - Pactual (0.4)

Subcooling = 124 - 120 4 de3rees subcooling (0.1)

=

REFERENCE Fluid hechanics Heat Transfer and Thermodynamies-ANSWER 1.10 (1.50)

Once the reactor reaches the power producing range, pressure is controlled, and the moderator temperature changes very little over the power range.

therefore the MTC effects are minimi ed.

(1.5)

REFERENCE Resetor Physics ANEJER 1.11

'. 3. 0 0 )

a.

Fuel temperature would increase CO.253 to get the needed delta T to tr ansf er the heat to the coolant. The corrosion layer will require some delta T across it to transfer heat [0.253 b.

Cladcing temperature would also increase [0.253 because the pin temperature increased and the cladding is nov transferring heat to the corrosion film instead of the coolant.[0.253 c.

Coolant temFerature remains the same [0 253 since it is a function of pressure, which is maintained constant by the EME system. CO.253 RErERECE d

Heat itsni fer J

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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE

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ANSWERS -- N1NE MILE POIN1-

-85/03/11-BERRY, J.

. ANSWER 1 12 (2.00)

s.

Since the steam would undergo adiabatic e::pansion, from the nollier diagr am the maximum temperature would be appron. 325 degrees F.

(1.0)

b.44e-(0.5) SRVs come off 6-#r e the flow restrictors which measure T6 steam _ flow. (0.5)

Mr

-

REFERENCE Rosetor. Theory hein Steam System Thermodynamics / Steam Tables-ANSWER 1.13 (1.50)

.A.

The pump will eventually add a sufficient amount of heat to the fluid to cause cavitation. Also will accept overheating of the pump.

(1.0)

,

B. Could cause e::cessively long starting currents or water hammer if the doenstream piping eas not filled.

(1.0)

C.

Cevses et:cessive motor amps to be drawn and the high current eculd cause damage to the motor vindings.

1.0)

(

REFERENCE Fluid Flov Basle Principles

.

.'-

.

l-Docket No. 50-220 Niagara Mohawk Power Corporation ATTN:

Mr. B. G. Hooten Executive Director Nuclear Operations c/o Miss Catherine R. Seibert 300 Erie Boulevard West Syracuse, New York 13202 Gentlemen:

SUBJECT:

EXAMINATION REPORT NO. 50-220/85-03 (OL)

This transmits the Examination Report of Requalification Examinations conducted by USNRC Region I at Nine Mile Point, Unit I the week of March 11, 1985.

At the exit interview held on March-14, 1985, the preliminary results of these examinations were discussed.

Although your overall Requalification Program rating is satisfactory, we note that some of the problems identified in this report have been brought to your attention by other reviews, and that you have committed to corrective action.

NRC Region I will continue to monitor the progress of these actions as they relate to the improvement of your Requalification Program.

No reply to this letter is required.

Your cooperation in this matter is appreciated.

Sincerely, Samuel J. Collins, Chief Projects Branch No. 2 Division of Reactor Projects

Enclosure:

Examination Report No. 50-220/85-03 (OL) w/ attachments

REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 85-03 (OL)

FACILITY DOCKET NO. 50-220 FACILITY LICENSE NO. DPR-63 LICENSEE: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 FACILITY: Nine Mile Point, Unit 1 EXAMINATION DATES: March 11 - 14, 1985 PREPARED BY:

J. A. Berry, Lead Reactor Engineer (Examiner)

Date REVIEWED BY:

R. M. Keller, Chief, Projects Section 1C Date AP ROVED EY:

H. B. Kister, Chief, Projects Branch No. 1 Date i

SUMMARY:

As part of the NRC programmatic evaluation of Requalification Training at Nine Mile Point, Unit 1, NRC prepared written examinations were i

administerec, in parts, to all facility personnel taking the Niagara Mohawk prepared annual requalification examinations the week of March 11, 1985.

Aaditionally, oral recualification examinations were given to 11 licensed personnel, 7 SR0s and 4 R0s.

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.9:

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REPORT DETAILS

'. TYPE OF EXAM 5:

Requalification EXAM RESULTS:

R0 l

SR0 l

l Pass / Fail l

Pass / Fail l

I I

I IWritten Exam I 35/4

32/1 l

l Partial Exams l I

I I

i 10ral Exam I

4/0

6/1

I I

l l

l I

I I

l

.

I i

l 10verall l

35/4 1.

32/I I

I l

l I

I I

I i

1.

CHIEF EXAMINER AT SITE:

J. A. Berry, U.S. NRC - Region I

.2.

OTHER EXAMINERS:

D. J. Lange, U.S. NRC - Region I F. J. Crescenzo, U.S. NRC - Region I T. L. Morgan, EG&G Idahc, Inc.

D. E. Hill, EG&G Idaho, Inc.

3.

. REPORT:

As part of the NRC's programmatic evaluation of Requalification Training at Nine Mile Point - Unit 1,

NRC prepared written examinations were acministered in carts, to all facility personnel taking the Niagara Mohawk cre;a e: arr.s' ;e:as':sti:t e x a?' n at' :r : ine ses :# Mz-:- * '.. 19EE.

.

Accitionally, oral recualification examinations were given tc 11 licensed personnel. 7 SR0s anc 4 R0s.

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__ _ _ _ _ _ _ _ _ _ _ _ _

_ _.

_ _ _ _ _ _ _. _ _

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

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-3-The NRC written examination sections were administered as follows:

Monday, March 11 - R0 Section 2 to 17 people

- SRO Sections 5 & 8 to 11 people Tuesday, March 12 - R0 Section 3 to 13 people

- SRO Section 6 to 12 people Wednesday, March 13 - R0 Sections 1 & 4 to 9 people SRO Section 7 to 11 people Overall, examination results were good.

Five people failea NRC admints-tered sections of the examinations, four RO's and one SRO, and one SRO failed the oral examination.

The comparison of scores on NRC sections vs. the facility sections indicated that the overall average score on the NRC exam (if sections were together) and facility exam were within 4% of each other.

This is con-sidered an acceptable range.

Individual section comparisons indicated a wide disparity.

Section 8 of the NRC and Facility exams were within.5%

of each other in average score, but Sections 2, 3, and 6 were off by 6.71%, 9.91* and 6.56% respectively, with the NRC section scores being lower.

Also, Sections 4 and 7 on the NRC exam had average scores 6.4% and 3.3% higher than the facili*y's sections.

.

The reasons for this disparity are not evident.

It appears that the higher scores on the NRC Sections 4 and 7 may be due to the facility's sections being overly long, but the other section differences cannc. be so explained.

Probable causes may be the tension involved in taking an NRC exam, more operationally oriented (not memorization) type questions on the NRC exam, or the dif ference in question " style" between the two exams.

In addition to the conduct of examinations, the evaluation also consisted of a review of the NMP-1 Recualification prograe Annual exarirations prepared by tne facility, anc discussions with licensec operators and training staff members regarding the Requalification Program.

The Annual Requalification examinations prepared by the facility were considered to meet NRC requirements, but were not considered to be of high quality.

Problems with the examinations incluced; double jeopardy questions, excessive length, many unnecessary theory calculations and questions having no relation to an operator's job, and simplistic short answer questions which failed to provide an adequate measure of depth of knowledge.

The facility's Requalification examination question bank is poor, and it is felt its use cont *ibuted to the problerrs with tne exam-i t a t i t r..

Tc M agara %tawi's :re:it. they have i:er ' 'e: tne :r:o', ems with the existing exam question bant, and have began a task to upgrade it.

Significant imcrovement is expected in next years eram.

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-4-Discussions with licensed operators indicate that there is dissatisfaction with the Requalification Program.

Problems sited included; to much emphasis on theory that is not operationally oriented, too much self-study or reading, and unchallenging and uninteresting presentation of subject matter.

These matters have been previously brought to the attention of Niagara Mohawk management by other reviews of the program.

Niagara Mohawk has com.nitted to a course of corrective action. NRC Region I will monitor the progress of the action over the next year.

It is felt that the addition of a plant specific simulator training program to the Requal program will aia in improving the program.

Overall, the Nine Mile Point, Unit 1 Requalification program is satis-factory.

NMPC has already begun to correct many of the programmatic problems identified. No further NRC involvement in the program is planned this year, other than monitoring of the changes being made to improve its quality.

4.

Personnel Present at Exit Interview:

NRC Personnel J. Linville, Chief, Reactor Projects Section 2C, DRP J. A. Berry, Lead Reactor Engineer (Examiner) ORP D. J. Lange, Reactor Engineer (Examiner), DRP F. J. Crescenzo, Reactor Engineer (Examiner), DRP A. J. Luptak, Resident 1 spector, NMP-1 NRC Contractor Personnel D. E. Hill, EG&G Idaho, Inc.

T. L. M:rgan, EG&G Idaho, Inc.

Facility Personnel T. W. Roman, Station Superintendent - NMP-1 K. F. Zollitsch, Training Superintendent, Niagara Mohawk J. C. Aldrich, Operations Supervisor, NMP-1 T. Wood, Training Supervisor, NMP-1 J. T. Pavel, Asst. Training Superintendent, Niagara Mohawk R. Seifried, Operations Training Instructor M. Dooley, Operations Training Instructor M. Jeres. Operatier S us viscr, NM;-2

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wo

,s.

o-5-5.

Summary of Comments made at exit interview:

The Chief Examiner noted that there was one person who was not a

clear pass on the oral examinations.

A discussion was held regarding Niagara Mohawk's commitment to implementation of upgrades in their Requalification Program based on previous audits.

Attachments: Written Examination (s) and Answer Key (s) (SRO/RO)

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IR 05000220/1985003

Text

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Enclosure:

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