ML20127E680

Exam Rept 50-333/OL-85-04 on 850211-15.Exam Results:One Reactor Operator & One Instructor Certification Candidate Failed Both Written & Oral Exams.One Reactor Operator & One Senior Operator Candidate Failed Written Exam Only

ML20127E680
Person / Time
Site:	FitzPatrick
Issue date:	04/24/1985
From:	Joshua Berry, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20127E649	List: ML20127E647 ML20127E680
References
50-333-OL-85-04, 50-333-OL-85-4, NUDOCS 8505200159
Download: ML20127E680 (100)

Text

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U. S. NUCLEAR REGULATORY COMISSION REGION I

.0PERATOR LICENSING ~ EXAMINATION REPORT EXAMINATION REPORT NO. 85-04 (OL)

J FACILITY DOCKET NO. 50-333 FACILIT.Y LICENSE NO.-DPR-59 LICENSEE: . Power Authority of the State of New York P. O. Box 41 Lycoming, New York 13093 FACILITY: James A. FitzPatrick Nuclear Power Plant EXAMINATION DATES: February 11 - 15, 1985

. REVIEWED BY: [f,#.M j .4 . h /) /7 m . #//h'/Y8I J. A. Berrf, ~LeadfRe(ctoWEngineer (Examiner) Daf.e REVIEWED Y:

. M. KePler, Chief, Project Section 1C Date

& f' [

APPROVED BY: '/

H. B. Kister7 Chief, Project Branch No.1 d[ 7 0(te

SUMMARY

Operator licensing examinations w'ere conducted at FitzPatrick during the period of February 12-14, 1985. Three Reactor Operator candidates, four

. Senior Operator candidates, and three Instructor Certification candidates were admin.istered written and oral examinations. One Reactor Operator candidate and

" ore Instructor Certification candidate failed both the written and oral examin-ations. One Reactor Operator candidate and one Senior Operator candidate failed the. written examination only.' During the oral examinations, all candi-dates . were noted to be very knowledgeacle in the use of Technical Specifica-tions'-and the Emergency Plan. Weaknesses were noted in the candidate's -

' abilities to use piping and instrument drawings as well as basic logic dia-

. grams. During gracing of the written examinations, the Rector Operator candi-dates were noted to be weak in the area of Plant Design / Instrument Controls.

No generic weaknesses were noted during the grading of the Senior Reactor Operator written examinations.

8505200159 850425 0REG 1 ADOCK 05000333 PDR LL

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. REPORT DETAILS TYPE OF EXAMS: Replacement. X EXAM RESULTS:

.l R0 l SR0 l Inst. Cert I-I Pass / Fail l Pa ss/ Fail l Dass/ Fail l' I I I I l- .

I I l  !

l Written Exam I 1/2 1 3/1 1 2/1 1 I I I I I I I I I I

.l0ral Exam i 2/1 l 3/1 1 2/1 l

~l , I I I I I I I I I

l l 1 I I l0verall i 1/2 1 3/1 l 2/1 1 1Results l l l l

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1. CHIEF EXAMINER AT SITE: D. Lange, U.S. NRC-Region I

2. OTHER EXAMINERS: W. Cliff, PNL-Battelle

3. FITZPATRICK ENTRANCE MEETING:

NRC Attendees D. Lange, U.S. NRC Region I Facility Attendees F. Catella, FitzPatrick Training Department

.D. Simpson, FitzPatrick Training Coordinator M. Curling, FitzPatrick Training Superindentant An entrance meeting was conducted immediately following the start of the RO/SRO written exam.

The tentative schedule for the oral exam assignments was discussed. The two hour exam review was scheduled on site from 3:00 PM - 5:00 PM. A tentative exit meeting was set up for Friday morning.

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1. Summary of strength's and deficiencies noted on oral exams:

A. strength was noted by'both' examiners in the area of Technical

? Specifications and use'of_the Emergency Plan.

An overall weakness was noted by both' examiners in the' candidates ability to use' piping.and instrument' drawings and bas _ic logic diagrams. .

- 2. . Summary of deficiencies-noted from grading of written exams:

Trie RO candidates had an overall weakness in tne area of Plant Desirgn and Instrument and Control.

The SRO candidates had no generic weaknesses noted. The candidates failing the exam showed consistently lower grades in the area of procedures and administrative controls.

- 3.

' Comments on availability of, and candidate familiari:ation with plant reference material in the control room:

Candidates did very- well using the control rocm plant reference procedures and technical specifications. Overall weaknesses were noted in the candidates ability to 'ise control room P&ID's and logic diagrams.

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Exit Interview Details Personnel Present at. Exit Interview:

NRC Personnel Dave Lange - Chief Examiner Larry Doerflein - Senior Resident Inspector NRC Contractor Personnel William Cliff,.PNL-Battelle Facility Personnel Donald Simpson - Training Coordinator Douglas Lindsey - Assistant Operator Superintendent W. Fernandes -.0perator Superintendent F. Catella - Nuclear Training Specialist

, M. Curling - Training Superintendent R. Converse - Superintendent of Power

. Summary of Comments made at exit interview:

'The Chief Examiner advised the facility of the preliminary results of the oral examinations.

The' Chief Examiner rioted the generic strengths and weaknesses observed during the oral exams, d..

The Chief Examiner commented on well written learning objectives that had been used during the ' candidates training program. The facility is in the process of developing lesson plans and exam bank questions based on these objectives. Emphasis should be placed on putting them in place.

Attachments:

1. Written. Examination (s)'and Answer Key (s) (SRO/RO)

2. Facility Comments on Written Examinations made during Exam Review

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k' ATTACHMENT 2. .

t i: - During ~ the SRO exam review, the following ccaments were raised by the utility.

Resolution of these com e.ts r are incorporated in the Master Answer Key. The

? following are the acceptea comments / additions / deletions f rom the -test answers.

l l

l l Question 5.5b Answer should be "Decause the high differential notch rod worth of edge rods can cause short periods OP65 pg.6." i 1

Ouestien 6.la Answer should, include: greater than 2.7 psig and 59.5 inches. l r Question 6.lc May'get the answer that if one diesel does not start, the second RHR- pump will not start. Celete answer " core spray pump at speed".

Question 6.2a Manual not necessary. If not included, 0.25 points for each correct answer. If manual is included, 0.2 'for each. correct

.. answer. .

t Question 6.2c Cand.idate may also say " reduced speed will reduce cooling later to barametric condenser and lube oil. (No credit-taken off for this answer.)

! 10uastion.6.3a 0.5 for answer in Exam plus for "all rods in per E0?-3 (0.5).

Question 6.3b RWCU system isolates. (acceptable additional answer) h/ Question 6.5c Steam flow indicaticn 9-5 panel. -(acceptable additional answer) '

Question 6.6 Steam flow feedflow mismatch and level decrease. - Plant efficiency decrease. (Other acceptable answers. )

Question 6.7 Discharge Valve Not Open - The reason for the runback is.to prevent excess axial thrust on pump.

.The reason for 44% limit is to get you in the range of one feedpump to prevent low level scram.

Question 6.8 High temp in pump rocm should be T ambient + 40cF.

Question 7.I' Add "only if instrument nitrogen cannot be restored."

1 Oupstion 7.2 Add " throttle servico water flow and if still can't reduce temperaturo.."

Questica E.1 Add "unr;no.4n condittens, contaminated icvai 500,000 DPM/cm 2 ,10R/hr Beta gamma, maintenanco in rad area."

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-ATTACHMENT 2 J.

LDuring :the RO exam review, . the following comments were raised by the utility.

All comments Twere resolved at the review with documentation provided by the facility training department.

Changes Necessary -NRC RESOLUTION Question No. Change REASON 1.09 Should be 10 not 40 Typo error-accepted.

2.03b . Answer Key should say Accepted with documentation.

"all'of it".

2.03d. Answer Key should include Accepted with 'documentation.

, (mini purge) A&B are reset alike.

2.05b Consider level control to Considered dur.ing grading.

torus in answer.

! 2.07b Answer will be "yes" if Modification accepted with candidate says from a documentation.

trip from high-level.

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3.05b-#2 ' Answer should include _ Accepted with documentation

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• to give the operator provided.

l enough time to respond to the steam flow to avoid a-

, scram.

I i 4.02b Consider expanding on. Documentation provided.

L answer to include RPV Considered during grading.

Control guidelines.

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U. S.' NUCLEAR REGULATORY COHH105 ION REACf0R OPERATOR LICENSE EXAMINATION

-FACIL11Y: FITZPATRICK REACTOR FYPE BWR-GE4 DATE ADMINISTERED: 05/02/12 EXAMINER: LANGE, D.

APPLICANT: ___

INSTRUCTIONS TO APPLICANT:

Hsc separate-paper for the answers. Write answers on one side only.

Staplo question sheet on top of the answer shoots. Points for each quest'ionLarc indicated in parentheses after the question. The passing Stade requiros at-1 cast-70%'in each category and a final .gr ade of at.

1 east'80%. Examination papers will be. Picked up six (6) hours after

~

the:cxamination starts.

%-OF

,CAfECORY  % OF APPLICANf'S ' CATEGORY

-VALUE- 10TAL SCORE VALUE CATEGORY

_ 1. L _ _ i ___________ ________

1. ' PRINCIPLES OF NHCLEAR POWER PLANI OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUlD FLOW - - - - - -

_ i __ _ 1 ___________ ________ 2. PLAN 1 DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS

.- I __ _ I ___________ ________ 3. . INSTRUMENTS AND CONTROLS

_ i __ _ _1 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL

'100.00- 100.00 TOTALS FINAL GRADE _________________%

~

All work.dono on this exanination is my own. I have neither siven nor received aid.

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

dPPLIC5OT I5~555U5TURE o ,. . . . . _ . . . . . _ . . . . . . _ . _ _ _ . . _ _ _ . .

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1. PRINCIPLES 10F NUCLEAR POWER PLANT-OPERATION, PAGE 2

~~~~5sERs557sisiC5- sEEi iREssiER Es5 FLUiB Ft5s QUES 110N 1.01 1(2.25)

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a. Using.the, attached Power to Flow operating mapr name items l'thru 14'.

(explaniation'not'requirod). (1 40) b.-What.is the. significance or item i 10. (0.25)

c. ' Item 4 11-( .the entire line-) is slightly concaver from 9hore it inter-

-sects (Estarts)- at item t i to where it ends at item t 7. Explain the reason f^or this. (0.60) utlEST10N 1.02 (2.00)

'During a cooldokn of the reactor vessel from outside the control roome reactor. pressure decreased from 805 psis. to 595 psig. in one half hour.Has your reactor cooldown. limit been exceeded ? { show all work } (2.00) .

GUESTION 1.03 (2.00)

Concerning' control-rod worth during a reactor startup with 100% peak Xenon versus a startup with Xenon froo conditions, WHICH STATEMENI IS CORRECT?

.HISTIFY YOUR CHOICE.

a. PERIPHERAL control rod worth will bc LOWER during the 100% peak-Xenon startup than during the Xonon froo star tup.

.b. CEN1RAL control rod. worth will bc HIGHER during the 100% peak Xonon-startup than during the Xenon free startup.

c. PERIPHERAL control rod worth will bc HIGHER during the 100% peak Xenon startup than during the Xonon free startup.-

d. BOTH CENTRAL and PERIPHERAL control rod' worths WILL BE 1HE SAME regardless of core Xenon concentration.

.GUES110N 1.04 (2.00)

~ Indicate whether the following will INCREASE or DECREASE reactivity _

during operation AND briefly EXPLAIN why?

a. Moderator temperature increases while below staturation temperature. (0.5)

b. Fuel temperature increases. -(0.5)

c. Loss of a feodwater heator . (0.5)

d. A' sudden reduction in reactor primary system steam pressure. (0.5) k_ Z'E Z --'.-d

1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3

~~~ TAEss557sEsiCs? sEEi iREssFER Es5 EEUIo FtGs QUESIION 1.05 (2.50)

NOTE Answer the following question from a theoretical standpoint, not from a Fitzpatrick system desi3n standpoint.

With the plant operating at 90% power, extraction steam to the highest pressure feedwater heater is removed. An enginect. observing that turbine load increased by 15 MWe after the extraction steam r e ni o v a l ,

concludes that this action has improved the plant's thermodynamic efficiency (N01 heat rate). Do you agi ec with this conclusion? Caplain your answer fully. (2.0)

HUES 110N 1.06 (2.50)

The concept of Suberitical Multiplication is used to describe the behavior of the reactor during ref ueling oper ations or startup.

a. In a suberitical reactor, if the source Icvel doubles, what will happen to the neutron level ? (0.50)

b. What three variables affect the subcritical neutron 1cvel ? (0.75)

c. In a suberitical reactor, if a reactivity of 0.003 dk/k is added to the reactor,will it take longer to reach equilibrium if the initial k- eff-ective is 0.92 or if k- offective is 0.992 ? Explain the reason for your answer. (1.25)

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GUESTION' 1".'07 T3^.~00 F ^'"~

When the reactor is at full power and a feedwater controller mal-function results in a loss of foodwater flow, a reactor scram will occur (due to low reactor water Icvel) within a short period of time.

During the time period JUST PRIOR TO THE SCRAh, is reactor power expected to INCREASE, DECREASE or REMAIN CONSTANT? Give TWO REASONS for your answer. (3.0) ,

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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4

--- inEER55isERICE- REEi iEEsiFEE EE5 FEDi5 FE5E GUESTION 1.08 (3.00)

Assume the.rcactor is operating at 100% power and one recirculation

. pump trips. Indicate how each listed indicated paramotor would first chango (Increase or Decrease) and briefly explain why the change occurs.

3. Pcactor poWor (onc reason) (1.0)

.b.. r eactor water icvel 'two reasons) . ( 1. C '

c. feedwater flow (two reasons) (1.0)

GUESfION 1.09 (2.25)

For cach-of.the pairs of conditions listed below, state which condition

. would have a CREATER-DIFFERENIIAL. ROD WORTH and briefly EXPLAIN WHY .

.a. Reactor moderator temperature of 150 des.F or 500 dos.F o (0.75) bf. For an inserted' rod next to a fully withdrawn control rod or next to La folly inserted control rod. ( assume average core fiv:- is constant)

(0.75)

c. For s rod'at position 10 or position 40 of a core operating at 100 %

power ? (0.75)

-QUESTION 1.10 (2.00)

W111'the Recirculation pumps have.more NPSH at 4 % or 100 % power ?

( EXPLAIN YOUR ANCHER FULLY-) (2.00)

GUESTION 1.11 (1.50)

'For-each of the events listed below, state which reactivity coefficient will respond first and if.it adds positive or negative reactivity. (1.50)

a. Relief Valve opening-at.100 % power . (0.50)

"b . Rod drop at 100 % power. .

(0.50)

c. Isolation of a feedwater heater string at 75 % power. (0.50) e 4

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'2. PLANT DESIGN INCLUDING SAFETY AND EHERGENCY SYSTEMS PAGE 5 GUESTION 2.01 (2.50)

With the_ mode switch'in:the REFUEL position, what conditions (01HER than_those initiated by the Neutron Monitorin3.and Rocirc. Flow Control Sys)

Lwill initiate a; CONTROL ROD BLOCK ? Include setpoints as appropriate.(2.50)

GUESTION 2.02 (2.00)

'Concerning the Scram Discharge Volume ( SDV.)

a.Is it~ permissable to close the SDU vont_and drain valves during normel operation ? If not, why. If so, under what conditions / limitations. (0.75) eb.Following a Scram, from full-power, what will be the internal pressure of the SDV.'? (0.25) c.On your shift,during rated power conditions, you discovered that the SDV.HI-HI water 1cvol BYPASS switch has been in the BYPASS position since startup. Could.this have prevented a VALID high instrument volume scram from occuring-? (briefly explain). 26 (0.75).

d.With the MODE-switch-in SHUTDOWN.the HIGH -(#6 sal.) SDV. ROD BLOCK trip-is BYPASSED ? ( True or Falso ) (0.25)

GUESTION '2.03 (3.00)

a. What are.the. normal values for CRD HYDRAULIC SYSTEM FLOW, DRIVE WATER DIFF. PRESS. and CHARGING WATER HEADER PRESS., indicated in the Control Room. (1.00)

b. Approximately what porcentage of the flow in 'a' above is supplied to the cooling water heador? (0.50)

c. . Explain HOW/WHY requesting. single rod insortion-causes cooling header flow to vary (include by how much the flow varios).- (1.00)

'd. .The systoni flow in 'a' above is loss than the normal flow ou_tput of one pump. List two (2) taps off the CRDH system upstream of the' flow sensing clomont. (0.50)-

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I-2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE -6 QIlrET10N 2.04- (3.00)

Concernin3 the Recirculation System ;

a. InLaddition to.a Recire. Pump runback, what operational conditions will cause.a Rocire. Pump Trip ? (setpoints are required) (1.00)

b. Prior'to starting an Idle Recire. Pump what coolant limitations have

.to be adhered to ? (be specific). (0.75)

c. When starting up the Recire. Pump and opening the discharge valvor what specifi~c R,X.paramotor must be monitored and why ? (0.75)

.d. For. 1 Operation of the Scoop Tube Positioner with the Handcrank  :

1. How is the Electr ic Dreak rolcased on the positioner motor f or:

both ttur A-MG set and B-MG set. (0.50)

GUESTION- 2.05 (3.00)

Answer the following with regard to the RHR system and its various-modos of operation

-a. Match the following actions, events, or interlocks in Column A with the item in Column D that initiates that item. (0.75)

Column A Column B

1. . Shutdown cooling isolatos 50 psig.(ine)

7. LPCI auto initiation (in conjuction 75 psis Cine) with~hi DW pressure) 125 psis (inc.)

3. Input to the Auto Blowdown Sys. 420 psig (inc.) j 450 psig.(doc.)

b.- Explain the purpose of the.RHR.- CONDENSING MODE of operation . (1.00)

c. lotus Cooling may be initiated at any time, reguardicss or whether or not a LPCI initiation signal is present . ( TRUE or FALSE ) ?? (0.25)

d. With a LPCI' initiation signal-present list TWO (2) separate sets of conditions that would allow you to initiate CONTAINMENf SPRAY. (1.00) uuES1 ION- 2.06 (3.00)

a. What are three (3) signals that will cause a diesel sencrator to r automatically Emergency Start (exclude manual, setpoints ARE required)? (1.00) b.- When the Emers. D/G is in the MAINfENANCE mode of operation it can ONLY be manually startedrLOCALLY ? ( TRUE or False ) (0.50)-

c.  : List'six'( 6 ) Emers. D/G ongine favits that would provent automatic initiation and cause a shutdown of the Diesole indicating whether or not the trip would occur if a VALID LOCA signs 1 wer e present . (1.50) i

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2. PLANT-DESIGN' INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 g'

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'0HES110N-l2.07 (2.50)-

Concernin3 the Reactor Core Isolation Coolins.Sys. (RCIC) i

a. List all the conditions that will cause an automatic isolation of the steam line isolation' valves, 13-MOV-15 & 16 . (1.00)

b. If the.RCIC sys. lurbine had boon shutdown by an automatic trip signal, which i9adverntantly had come-in and cleared, will the Turbine re-start on_ a -VALID initiating signal with no operator action. ? E:rplain. (0.75)

c. What is the reason for thcl ' CAUTION','Do not oper ate the RCIC turbine at a speed below 2200 rpni. f or en e:: tended period of tinie * ? -( 0. 75 )

00ESTION 2.08 (2.75)

Concerning.the Standby Liquid Control Sys.;

a. Once-the SBLC.'sys. has initiated, what six (6) CONTROL ROOM indications could you use to verify that the system is operating properly AND in-jecting into the reactor vessel ? ( 1.50 )

b. After initiation.of the SBLC. sys.,is it permissibic to shut the system down ?-( if not, WHY., if'so, under what conditions ? ) ( 1.25 )

L uuESTION -2.09 (1.75)

Concerning the OFF-GAS system i

.a. List threc(3) conditions that will cause an automatic shutdown or the off sas recombinor.- (1.00)

b. What undesirabic condition could exist with the loss of the recombiner and subsequent f ailure of the load dilution f an to star t ? (0.50)

c. Who-(.by titic ) should be notified in the event the off sas sys..is operating with the recombiner. isolated ? (0.25)

OUESTION. 2.10 (1.50)

Concerning the 1raversing In Core Probo Sys. i

s. List two (2) valuable operational parameters that are developod'from signals sencrated by.the TIP. sys. (1.00)

( include in your ans. how the signals are beins used and what information is bains obtained )

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3. INSfRUMENTS AND CONTROLS PAGF 8

- QUESTIDH' 3.01 (2.50)

.No.1- Indicate at what RX. water level, ABOVE 1HE 10P OF THE ACTIVE FUEL ,

_each of tho._following actions is directly initiated . If more than one

' level appliose indicato all of the applicable Icvels. (1.50) as . Direct reactor scram'

b. Standby Gas 1reatment Systeni starts

.c. RCIC starts

d. Reactor Low Water level alsrm annuncistos

o. LRocirculation pumps' trip

f. RFPs trip No.2 What'is the high~ pressure trip sotpoint for _the ATWG Recirc. Pump' Trip and how does this compare to pressure sot point of the Relief Valvos lifting.?

(1.00) i

- 00ESTION- 13.02- (3.00)

For EACH of the following conditions, state whether a scram, half-scram, rod _ block, or no, action is directly_sonorated.

~

For conditions that' pro-duco more,-than one action, state the more severo action (i.e. half-scram is more sovere'than a rod block). (3.00)

a. Loss-of one RPS MG' set Turbine trip at 25% power b.

c. . -Two main steam lines isolated, Hode switch in RUN

d. APRM B downscale, Modo switch in RUN

o. Scran discharge volume lovel is at 40 gallons,. Mode switch in S1ARTUP

'f. Recire. Flow Comparator Inop. ( mode switch in shutdown )

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3f. INSTRUMEN15 PAGE 9

________________.AND~CONTROLSt ____________

41IE Ci 10N. 3.03. ('3.00)

Ma'tch the' radiation.ldotectors in-COLUMN A with the appropriato

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, application (s) Land characteristic (s) .from COLUMN B.. All i temsfin

f:ULUMN.B apply. (3.0)

COLUMN A. " COLUMN O a.. lon chamber- 1. Used primarily as a devico.for'preciso

- % measuronients or whoro.high sensitivity

.b.= ceintillation i s ; r oquir ed.

c. Geiger-Muellor' 2 '. The si o of the-olectron avalancho is'

, proportional to the original. incoming radiation enorsy.

5m-:- e 3. Usos a photocathode to convert light into free electrons.

~4. Normally used to set dose ratos.

5. :lhe same si=c pulse is sencrated regardless of the type and sPocific-ionization characteristics of the radiation.

6. Normally roads out in counts'por minuto.  ;

7. Main Steam Lino Radiation Monitors- '

1

8. Main Stack flow Radiation Monitor .

-QUESTION- 3.04 -(3 00)

, :What effect.will a complete loss of Reactor' Protection Sys.( RPS ) power

have on the followins systems 1or. paramotors'? ( fully explain )-

n. . Recirculation' system. ( 0.75 )

b. EHC system. ( 0.75 )

c.~Roactor vessel lovel. ( 0.75 )

d. Reactor level instruments. ( control room and local inst. racks ).(0.75) '

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3. .'INSTRUMEN1S AND CONTROLS PAGE 10 QUES 110N 3.05 (2.50)

'Concernin3 the Rocirculation Systemi

a. Under normal operating conditions,'what signals will cause a MG Set Scoop ~ Tube Lockr WITHOUT tripping the MC Sc.t ? ( l.00 )

b.-List;two ( 21) conditions that will cause a Recire. Run-Back, the l percent-the pumps run back to, and the reason for the runback.(1.00-)

e.eVov'are in the process of starting up ( RX. Power is approx.00 % ).What p two ( -2 ). control room indications could you use to verify a Recirc.Run -

Back. ( 0.50 )

GUESTION 3.06 (2.25)

During the.4:00 pm. to 12!00 mid. shift , at rated power, you receive two alarms 1on panci 09-6 3

-1.0ff GAS line high pressure. ( 20 psis. )

2.0ff GAS line high temperature ( 300 dos.F )

You notice that the condenser isolation valves 00-A0V-113 A and 113 B arc' shut and condonscr vacuum is decreasing .

a.10ased,on the above indications / conditions, WHAT HAS OCCURED ?.and what additional automatic actions can be expecto3 ? ( 1.75 ) ,

b. Based on the above situation what would bc your first ae*f3Mid action ?

( 0.50 )

QUESfION 3.07 (3.00)

'Concerning Refuel Operations and Fuel Servicing Equipment i a.nThoro-are no interlocks to provent the refueling platform from travers-J _ing in the forward ( from the RX.vossel to the spent fuel pool ) dir-g ection. ( TRUE or FALSE ) (0.50) 6.-What safety precaution is taken prior to using the availiary hoist to handle contaminated equipment that must be kept below a specific water 1cvel '? (0.50)

c. ' What purposo does the Refuelins Bellows accomplish during refuel oper-ations, and RX.vossci hoatup and cooldown ? (0.50)

,a d. What two (2) conditions will-provent the refuel platform from tr aveling toward the core, when the modo switch is in refuel ? (0.50)

QUESff0N 3.08 '(2.00)

How is the integrity of ECCS piping inside the reactor vessel verified during normal operation (include sensing points, specific

.svstem(s) who's piping is verified, why its verified, and response of the instrumentation to a loss of integrity in your answor)? ( 2.00)

,t 1

i t^'*- - * * *

- FP'* - .. N ** _ Qalget e>= e m e * *w ** 9* - - - * * * * ' * * * * * * * --*""******-****h****d"'***** N- ,

I:

,3. > INS 1RUMEN15 AND CONTROLS PAGE 11

' OllE 5110N . 3.09 (2.00)

During your shift r-a Relief Valve fails open . Following the Reactor Scram

and ful l. P/C' Isolation the vessel rapidly depressuri cs to below 500 psis.

a. What control room Icvel instrumentation is accurate and what IcVel inst-rumentation would notHbe reliable during the above transient ? ( 2.00 )

( BRIEFLY EXPLAIN )

OUESTION ~3.10 (1.75)

Con'cerning the hain Steam System i

s. List three ( 3 ) functions of the Main Steam Line Flow Restrictors.(1 00) b.How'many solenoid operated pilot valves are associated with each MSIV and

.what is-the purpose of each. (0.75) 4 br

\

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=

4. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12

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

~~~~R5D56L66265E~CUUiR6L QUESTION -4.01 (2.00) c.- -According to Fit = patrick,s Radiation Protection procedure, for scl-ection'of survey instruments, list four of the fivo considerations that should be taken into account prior to.their use. .

(1.00)

b. Match the fo11owin3 HP. survey instruments in column 4 1 with the type -

of~ radiation / contamination detcetion being performed i n column # 2.

Column 4 1 Column 4 ?

1. . Portable Gh survey motor. a. Commonly used for measuring bota gamme

(~Eberlino- E-120 ) exposure rates. Has a range of 1 to-

2. Portable Ion Chamber. 25,000 mrad /hr.

-( Victorcon, 740-F ) 6. Very sensitive instrument used for

3. Portable ion chamber. '

sniffing

• type surveys to detect and

( Tolotector ) pinpoint the presence of beta a gamma.

4.: Portable G.M. detector. c. Not recommended for determining beta

( digi / master- 305,and contact exposures in high radiation auto-digi / master-305-B) . areas, but would be useful for checking an off gas' leak in a large area.

d. Commonly used for measuring bota/ gamma exposure rates in High-Rad areas. Has a range of 0.1 mr/hr to 1000 r/hr.

UESTION 4.02 (1.75)

Concerning F- E0P-33 , Small Break Accident.

-a. Undor.what' conditions'can the automatic controls of an ECCS sys. bo placed in manual ? (1.00) b.: List three of the four basic objectivos you are expected to achieve,in the event of a pipe break, with respect to the REACTOR CORE and its CONTAINHEHf

. (0.75) 1

'UESTION 4.03. (2.00)

.Given the set of indications listed below, which exist following a valid LOCA, state whether or not adequate core cooling can be assured. Justify your answer. (2.0)

-HPCI has ISOLATED due to low steam supply pressure.

-All reactor water IcVel instruments are off-scale LOW, with the exception of the. fuel cono instrument which is off-scale HIGH.

-Both core sprav pumps have started, subsequently tripped on over-load, and CANN0f be restarted.

-RHR pump 'A' is RUi'NING with an injection path to the RPV (minimum-flow valve closed in loop 'A'. All other RHR pumps have feiled to start, C

+ e= g = == em v 4 * .e .C w w. mmme seene ege

p

4. PROCEDURES - NORMAL, AE: NORMAL, EMERGENCY AND PAGE 13

~

, ~~~~R56 5L E55hL 5UUTR L'~~~~~~~~~~~~~~~~~~~~~~~

1 .____________________

OllES110N 4.04 (2.'25)

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

' decrease.= Explain Why . (0 25 for the event. 0.50 for the reason)

IE01 ION 4.05 (1.50)

c. In accordance with the Fitzpatrick Generating Station Emergency Plan Implomontation Procedures, what are the EMERGENCY EXPOSURE GUIDELINES for the following situations:( WHOLE BODY ONLY)

1. Life Saving and Roduction of Injury (0.5)

2. Operation of Equipment to Hitigate an Emergency (0.5)

3. Eithor of the above two conditions if adequate planning and protection permits. (0.5)

ESTION 4.06 (2.00)

According to F-AOP-1,( Reactor Scram ), list the immediate Operator actions you are to perform in the event of a reactor scram . (1.50) lES110N 4 07 (2.50)

While controlling reactor pressure following a reactor scram and isolation, (MSIV closure), reactor pressure approaches the safety / relief valvo' set.-

. point.following an earlier auto- blowdown; According to F-OP-1 ( Main Steam

,Sys.) you are directed to reduce pressurc~to approximately 900 psig.

o.How could you have possibly avoided this pressure increase ? (0.75)

.b.Now.'having to select a safety /rolief valvo for pressuto reduction,what is the next valve to be selected and WHY ? (0.75) c.'As soon as conditions permit , ie: Scram / Isolation reset. list two more conventional means of controlling reactor pressure.(bc specific) (1.00) 2 N: a.a '

s s , ns. w .M m a.~ e. v' e

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9 4.. PROCEDURES NORMAL, ABNORhAL, EMERGENCY AND PAGE 14

~~~~Rd556L5cICdC C5DTR5C-----------------~~-----

QUESTION 4.08 (2.50)

When operating the RHR System in the Shutdown Cooling Moder Proceduto F-OP-

'l states that anytiac the reactor vessel is in a no or low flow condition, to increaso vessel'lovel to 234.5 * .

~

c. What lovel instrument should be used to verify 234.5' ? (0.50)

6. Why'is this level incrosso important ? (0.75)

c. List three problems that could occur if lovel was not increased. (1.00)

QUES 110N 4.09 (2.75)

Concerning F-AOP-36, ( Stock Open Rolief Valvo ) i

e. List five (5)-control room instrument indications, including back panels that you could use to verify a Relief Valvo is stuck opon.(1.25)

b. Having unsuccessvily attempted to shut the stuck open R.V., by cycling.

the'valvo control switch on panel 09-4, what further action can you take to got the valvo shut ? (0.50)

c. It is determined that the problem exists at the Remote Relief Valvo Panel. Concerning this, answer the following;

1. Whore is this' panel located ? (be. specific) (0.50)

2. 'With the panel energi=od, what further action can be taken ? (0.50)

QUESTION 4.10 (3.00)

Concerning Proceduto F-EOP-28,' PLANT SHUTDOWN FROM OUTSIDE 1HE CONTROL ROOM 't

c. What are the immediato operator actions if the main control room becomes uninhabitable ?(3 required) (1.5)

b. Where is RX. lovel and press. monitored outside the Control Room ? (0 5)

c. If the NCO is unable to shut down the plant prior to 1 caving the Control Room, list Four (4) ways to Scram the RX. (in order of profrence).(1.00)

.Q T,

_ .-.7- __..y___._7_.____ _ _.__--_____.-_7_

Y

\s J4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE' 15

~~~~ -

RA6iUL5EicIt c6NiRULI~~~~~~~~~~~~~~~~~~~~~~~

QUESfION 4.11 (2.75)

According to F- E0P-2 (RPV control);and F-EOP-4 (Primary Containment Con-trol);

a. List'_the' entry conditions for RPV Controle (include set Points) (1.50)'

b. List the entry conditions for Primary Containment Control.' (1.25) t.

finclude'setpoints) b

?

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6 TABLE II-3-1 P80PERTIES Of' SATURATED STEAM AND SA'IURATED WATER (TEMPERATURE)

Votome, ft'/ib Enthatpy 8tu/ib Entrog StwM, a f Water Evap 5 team Water Evap 5 team Water [ var Steam

[

vr v m vg he hq h, sg s, o s, 32 . 0.08859 0 01602 3305 3305 -0.02 1075.5 1075.5 0.0000 2.1873 2.1873 32 35 0.09991 0.01602 2948 2948 3.00 1073.8 1076 8 0.0061 2.1706 2.1767 35 40 0.12163 0.01602 2446 2446 8.03 1071.0 1079.0 0.0162 2.1432 2.1594 40 45 0.14744 0 01602 2037.7 2037.8 13.04 10681 108'.E 00262 2.1164 2.1426 45 50 0.17796 C.01602 1704 8 1704 8 18 05 1065 3 1083 4 0.0361 2.09:1 2.1262 50 60 0 2561 0.01603 1207.6 1207.6 28.06 1059.7 1087.7 0.0555 2.0391 2.0946 60 70 0.3629 0.01605 BES 3 868 4 38 05 1054.0 1092 1 0 0745 1.9900 2.0645 70 80 0 50'E C 016 7 6333 6?? 3 4E 04 104F4 1364 0C932 19426 2 03 % 63 90 0 69:i 001610 46E 1 465.1 56 02 1042 7 1100 6 0.1115 1.89M 2.0056 90 100 0.9492 0.01613 3504 3504 ES 00 1037.1 1105.1 0.1295 1.8530 1.9825 100 110 1.2750 0.01617 265 4 265 4 77.98 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 157.32 157.33 97.96 1019.8 1117.8 0.1817 1.7295 1.9112 130 140 2.8892 0 01629 122 95 123.00 107.95 1014.0 1122 0 0.1985 1.6910 1.8835 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 77.27 77.29 127.96 1002.2 1130.2 0.2313 1.6174 1.8487 160 170 5.993 0.01645 62.04 62.06 137.97 996.2 1134.2 0.2473 1.5822 1.8295 170 150 7.511 0.01651 50 21 50.22 148 00 990.2 1138.2 0 2631 1.5480 1.8111 180 190 9.340 0 01657 40 94 40 96 159 04 984 1 1142.1 0.2787 1 514! 1.7934 190 200 11.526 0 01664 33 62 33 64 168 09 977.9 1146.0 0 2940 1 4824 1.7764 200 210 14 123 0 01671 27.80 27.82 178.15 971.6 1149 7 0.3091 1.4509 1.7600 210

, 212 14.696 0.01672 26 78 26 80 18017 970 3 1150 5 0.3121 1.4447 1.7568 212 220 17.186 0.01678 23.13 23 15 188.23 965.2 1153 4 0.3241 1.4201 .L7442 220 230 20.779 0.01685 19.364 19.381 198.33 958.7 1157.1 0.3388 1.3902 .1.7290 230 240 24.968 0.01693 16.304 16.321 208.45 952.1 1160.6 0.3533 1.3609 1.7142 240 250 29.825 0.01701 13.802 13.819 218.59 945.4 1164 0 0.3677 1.3323 1.7000 250 260 35 427 0.01709 11.745 11.762 225.76 9366 1167 4 0.3519 . 1.3041 -1.6862 260 270 41.856 0 0171E 10.042 10.060 238.95 931.7 1170.6 0.3960 1.2769 '1.6729 270 280 49.200 0.01726 8.627 8 644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 280 290 57.550 0.01736 7.443 7.460 .259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 300 67.005 0.01745 6.448 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 310 77.67 0.01755 5.609 5.626 280.0 902.5 1182.5 0 4506 1.1726 1.6232 310 320 89.64 0.01766 4.896 4.914 290 4 894 8 1185.2 0.4640 1.1477 1.6116 320 340 117.99 0 01787 3.770 3.788 311.3 878.8 1190.1 0.4902 1.0990 1.5892 340 360 153.01 0 01811 2.939 2.957 332.3 862.1 1194.4 0.5161 1.0517 1.5678 360 380 195.73 0.01836 2.317 2.335 353 6 844.5 1198 0 0.5416 1.0057 1.5473 380 400 247.26 0.01864 1.8444 1.8630 375.1 825 9 1201.0 0.5667 0 9607 1.5274 400 420 308.78 0.01894 1.4808 1.4997 396.9 806.2 1203.1 0.5915 0.9165 1.5080 420 440 381.54 0.01926 1.1976 1.2169 419 0 785 4 1204 4 0.6161 0.8729 1.4890 440 440 466.9 0.01 % 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0.8299 1.4704 460 480 566.2 0.0200 " 0.7972 0.8172 464.5 739.6 1204.1 0.6648 0.7871 1.4518 480 500 6809 0.0204 0.6545 0.6749 487.9 714.3 1202.2 0.6890 0.7443 1.4333 500 520 812.5 0.0209 0 5386 0.5596 512.0 687.0 1199.0 0.7133 0.7013 1.4146 520 540 962.8 0 0215 0.4437 0 4651 5368 657.5 1194.3 0.7378 06577 1.3954 540 560 1133 4 0.0221 0.3651' O.3871 562.4 625.3 1187.7 0.7625 0.6132 1.3757 560 540 1326.2 0.0228 0.2994 0.3222 589.1 589 9 1179.0 0.7876 0.5673 1.3550 580 600 1543 2 0 0236 02438 02675 617.1 550 6 1167 7 0 8134 0 5196 1.3330 600 620 1786.9 0.0247 0.1962 0.2208 646 9 506.3 1153.2 08403 0 4689 1.3092 620 640 2059.9 0.0260 0.1543 0.1802 679.1 454 6 1133.7 0.8686 0.4134 1.2821 640

• 660 2365 7 0.0277 0.1166 0.1443 714.9 392.1 1107.0 0.8995 0.3502 1.2498 640 600 2708.6 0.0304 0.0808 0.1112 758.5 310.1 1068.5 0.9365 0.2720 1.2086 680 C 700 3094.3 0.0366 0.0386 0.0752 822.4 172.7 995.2 0.9901 0.1490 1.1390 700 705.5 3208.2 0.0508 0 0.0508 906 0 0 9060 1.0612 0, 1.0612 705.5 28

. _.~

MM

1. PRINCIPLES OF NUCLEAR POWF.R PLANI OPERAIION, PAGE. 16

~ ~

~~~~T EEREU6iU555C5I~UE5i~iR555fER 5UD fLUf6"FLUU ANSWEFS -- FITZPAIRICK -85/02/12-LANCE, D.

,/

/

ANSWER 1.01 (2.25)

a. 1. natural circulation line . 8. flow control range.

2. min. pump speed lino, 28 %. 9. Recire. pump NPSH limit lino.

3. min.' power line. 10. Jet pump NPSH limit lino.

4. 50 % load line. 11. APRM rod block lino.

5, 75 % load line. 12. APRM thermal scram.

6. 100 % load lino. 13. Thermal scram trip-Clamp

7. Pump constant speed line. 14. APRM- fixed scram.

(0.10 for each correct ans. )

b. Recire. pump NPSH limit , to protect against cavitation. (0.60)

c. This line is slightly concave becausci as core flow increases core inlet subcooling decreases, therefor core thermal power is slightly loss at full flow than it would be if the core inlot temperaturo did not chango.

(0.50)

REFERENCE

. l A.F . LP. book el Tab-C, Soc. E. Recire. and Recire. Flow Control Sys.

Performance Objectivo- Describe the operation of the Rocire. Flow Control System, Power to flow operating map FIG 47.

ANSWER 1.02 (2.00)

'First, convert psig. to psia. by adding 14.7 psi. 1honerofering to the steam tablesi 900 psia. = 532 dos.F 610 psia. = 488' dog.F 7 532 des.F - 488 dog.F =44 dog.F / half hour, or 80 dos./hr (1 50)

NO. The cooldown limit of 100 dos.F/hr has not boon exceeded. (0.50)

RFFERENCE J.A.F. Thermodynamics; MET.-222, pg.1-15. Enabling objectivos,222.7.1 thru 222.7.5 . Tech. Spoc. Thermal Limitations pg. 136 .

g[ANGWER 1.03 (2.00)

C is the correct answer CO.53. The highest Xenon concentration will be in the centor of the coro [0.53, the high flux region from the previous operating period CO.53. 1his will increase the flux icvols in the arca of the periphoral control rods CO.53, thus increasing their worth.

REFERENCE J.A.F. LP. 4 237.5 pg. 4 10 Significant effects of Xenon.- Effect on CRW .

a 4

9 E2 -* A h M h a mm ee rep __ s _ g _e>et@g M e _Cmpeg og **'**=4

C-0 5.10 During startup (power = 1 watt) a rod is pulled and a 60 second period is observed.

a) With no further pulling of rods, can the operator maint'iTn this Reactor period for Fminutes?

Explain your answer. (1.5) b) By pulling rods can the operator maintain the 60 second period for 30 minutes from the 1 watt power level?

Explain your answer. (1.5)

Answer:

p , p ,t/T = 1 watt e 1800/60 sec = 1.06 x 10 13 watt 7

= 10 mw No (0.5)

Moderator coefficient would turn it. (1.0) b) No (0,5)

Power at end of 30 minutes would even exceed max. MWT for Fitzpatrick. (1.0)

Ref. Equation sheet.

5.10 9 *- - -

4 5.11 Fitzpatrick is operating at full power when the Feedwater Controller malfunctions causing a complete loss of feedwater.

A Reactor Scram will occur on low level. During the time period just prior to the scram, is the Reactor Power expected to increase or decrease? Give two reasons for your answer. (2.5)

Answer:

Decrease (0.5)

1) Loss of feedwater causes less subcooling, more negative AK/K due to more voiding. (1.0)

2) Recirc. Runback on (20% feedflow decrease core flow. (1.0)

Ref. NET 237.3 pg. 8, F-0P-27 pg. 3.

End of Section 5 l

i 5.11

r: -

_ -g y

11. PRINCIPLES OF NUCLEAR POWER PLANT.0PERAff0N, . PACE 17

~

~~~~iEERh66iU555C5I"5E57~iE5U5 FEE ~5UD FLU 16~fUl0'  ?

ANSWERS -- FIf2PAfRICK -05/02 /12-L ANG : . D.

ANSWER. 1.04_ (2.00) 4 Adds. negative.rcact'ivity CO.253.duc'to the increase in neutron 1cakago ' Moderator temperature. coefficient. CO.253 '(0.5) b .' Adds negative roactivity C0.25] due to the increase in neutron capturo in the fuel - Doppler; coefficient. CO.253 (0.5)

.c. Adds positive reactivity Co.253 due to the decrease in neutron leci.ogo - hodcritor temperaturc coefficient.-CO.25] (0.Si d.LAdds negative reactivity CO.253 doo to the increase in neutron leakage'- Void coefficient. CO.25] (0 5)

REFERENCE-J.A.F. Reactor lhoory LP.4 237.4 Objective 4 237.3.1.5,and1.6 .pg.2 i and P3. 16 Effects of Coro paramotors.

ANSWER ,-1.05 (2 50)

NO.CO.53' Thermodynamic officioney is a comparison of. energy in versus energy cut.-[0.53 :The increase in Sonorator output resulted from.docreasin3

^

the' amount of steam diverted to the HP FW heator. E0.53 1his condition requires additional energy output from the reactor to raise FW temp to the same saturation temp as before CO.53 1hus, thermodynamic officiency of the. plant has gone down. [0.53 More delta T across the heator would have caused more extraction steam to have boon removed from the turbinc.(0.5).

REFERENCEL

.lAF. Heat Transfer 4 228.1.1.10 and 11 .

JAF. MET.'222.9 and 222.10 Thermodynamic Cycle Analysis.

JAF. HET. 1222.10.0.3 a thru de enableing objectives. Explain the Gross Thormal Efficiency, Not Thermal Eff.,and Cross and Not plant Elec.out-put . .

m .#

• a t

^~ -"-

__ ___t _ . - m -+tn_ *

• mb GOL __

{5 c . -

6

' 1. . : PRINCIPLES OF.NilCLEAR POWER PLANT OPERATION, PAGE 18

--- isiss557sisiE5- sEEi iEEs5 FEE Es5 FE5i5 FE5s ANSWERS - .FI12 PATRICK -85/02/12-LANCE. D.

ANSWER 1.06 (2.50)-

a._The suberitical neutron level is directly proportional to the neutron source strength. If the source strength doubles, the neutron level doubles. (0.50)

b. The three variables arci Source Strength, K-off.,and Time. (0.75)

c. The case ~ w hen K-eff.= 0.992 will take longer to reach ce;uilibrium. The reason for this is that o ric unu s t wait for the last ter m in the series expansion to become insignificant. That is, until K becomes insis--

nificant. This takes more tornos as K - 1. (1.25)

REFERENCE JAF. NET. 237.7 Subcritical Mult. and 237.7.1.4 Subcritical Power Level.

AF. Enablein3 Objectivos 237.7.1.1&2.

ANSWER 1 07' (3.00)

DECREASE.CO.53 REASONS!1( 2 of 3 required at 1.25 each)

1. Immediately the loss of foodwater flow causes a decrease in moderator subcooling which introduces negative dk/k into the core..

2. When feedwater flow drops below 20%, the recire. pumps will auto runback to 20%. The decrease in core flow causes an inercase in voiding which also adds negativo dk/k into core.

3. Decreasing level in the downcomer will reduce the available head for core circulation and will result in decreased core flowe.and thus reactor power will decresso.

REFERENCE

.lAF. NET- 237.4 Reactor lhcory, moderator coefficient .

JAF. Rceire. and Foodwater sys. 26% flow control runback < 20% FW. flow, F-OP-27 Recire. Flow Control.

9

r

,4-

1. PRINCIPLES OF NUCLEAR PnWER PLANT OPERATION, PAGE 19

~

~~~~~IUEEU56 UhESC5,~5Ehi~iREUEEER hYb'EEU56'iL50

' ANSWERS -- FI12PA1 RICK -85/02/12-LANCE, D.

ANSWER 1.08 (3.00)

O. Decrease (0.5) due to increased void content in the' core as flow

'decreasos'(0.5).

b.-Increase'(0.34) due to increased voiding in the core (0.33) and recite pump no longer taking a suction on the annulus (0.33).

c. Decrease (0.34A due to steam flow decrease (0.33) sr.d icvel increase (0 33).

REFERENCE JAF. SDLP. 02-H, 021 Recirc sys. , F-OP-27, Recire. Sys.

SDLP. 02B-F RX. Vessel Inst. and SDLP 02-A RX. Vossol Internals.

NSWER 1.09 (2.25)

o. At 500 des.F (0.25),As moderator temperature increases, neutron Icakage out of the fuel bundles is increased . The control rod is exposed to a higher neutron flux, thus rod worth increasos. (0.501.

b. lhe withdrawn rod. (0.25), Nootron flux is higher in this area, thus rod worth is greater. (0.50).

c. A t p o s i t i on JMP. ' ( 0. 25 ) , Voids at the top allow more fast 1cakage and less thormal/ keutron flux, thorfore greator rod worth at the bottom.

(0.50)

REFERENCE

.lAF. Reactor Theory. NET. 237.4 , Enabling objectivos, 237.4.5.3, 1 thru 5.

.Wi4SWER 1.10 (2.00)

Atl100 % power (0.75), At 4 % power you are at operating pressure but low feedwater flow rato. NPSH is low due to T- iniot being hi 3h. As power in-creases , pump inlet temperature is reduced due to mixing in the downcommer

-T- inlot is lower so P- sat. at inlet is lowore thorofor.NPSH is higher.

(1.25)

REFERENCE

.lAF. HE1, 214.9.12, pg. 12.

l L

h

.s

-+- --h-- * * *

• 5 a MMMMh_ _

e g. - - -

v. .

m , y

+ 1, ' PRINCIPLES OF NUCLEAR POWER PLAN 1 OPERATION, PAGE 12 0

--- isiss557sEsiE5- REEi iEEssFER Es5 FE515 FE5s ANSsERS'-- FI1ZPATRICl! -85/02/12-LANGE, D.

SWER .1 11 (1'.50) 2o. VOID. COEFFICIENT-(0.25), adds negative reactivity (0.25) b.. FUEL TEMP. COEFFICIENT (0.25), adds negative reactivity (0 25)

c. MODERATOR TEMPERATURE COEFFICIEN1 (0.25),. adds positive reactivity (0 25 REFERENCE 3 .lAF. NE1. 237.4 Reactivity Coefficients.

(

l i

+,

/

s i

A i

r e

4 i

TT ,

'6

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 ANSWERS -- FIf2 PATRICK -05/02/12-LANGE, D.

(jhNSWER 2.01 (2.50)

1. Scram discharge volumo high level scram bypassed

?. Refoc1 Platform over the core and fool grapple not full up

3. Grapple' fuel loaded (400 lbs)

4. Framo hoist fuel loaded (400 lbs)

5. Trolly hoist fuel loaded (400 lbs)

4. Service platform loaded (400 lbs)

7. SELECTION of a second ROD ( with one rod not full in )

( Seven correct answers at .357 cach ) = 2.50 for full credit.

REFERENCE

.lAF. LP.. book i 2 Tab L- Reactor Man. Control & RPlS systems. Refueling Interlocks / Mode Switch Position.

ANCWER 2.02 (2.00)

a. Yes. These valves may be closed for periodic testing and maintenance under strict Administrative. Controls. (0.25)

With either valvo closed, if the not drained alarm should come in, the valvos should be opened. If the valvos cannot be opened, commence a con-trolled shutdown. (0.50)

b. Reactor Pressuto . (0.25)

c. NO. (0.25) 1his switch is only active in the Refuel or Shutdown MODE switch positions. (0.50)

d. FALSE (0.25) This Rod Block trip is never bypassed .

REFERENCE JAF.LP.. Book i 1 lab E - CRD.HYD., Tech .Spces. sec.3.3 CRD. sys. Opor.

and LP. SDLP-071-RMC &RPIS sys. Objectivo # 4 pg.6 5.

g/ ANSWER 2.03 (3.00)

a. 59 spm, accept 55 - 60 spa (0.33)for flow; 260 psid.,arcept 250-270psid (0.33) for flow.; 1390-1400 (0.33) for charging water press,

b. Approx.' M % or 2 W spm. (0 50)- att #g ST- /

c. When a rod is inserted, one set of stabilizing valves, & valves.

-close (0.5) to direct 4 spm (0.5) to the CRD and away from the cooling water heador.

d.- minimum flow lino (0 25)

- recirculation pump seal purge (0 25)

REFERENCE VAF- Procd. F-OP-25 and LP.tGDLP-035 CRD. HYD. SYS.

3

' I

t. . - --

r.

7

~

'. 's

2. . PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS .PAGE 22 ANSWERS -- FITZPATRICK -05/02/12-LANCE, D.

ANSWER' 2.04 -

(3.00)V* /M

• c.. Reactor Low Level'( -38 ) and/or Reactor High-Pressure (1120 psig.)

(1.00)

b. Verify the coolant-temp. in the idle loop is within 50 dog. F of the coolant' temp. of the RX. vessel. (0.50)

Verify that the' coolant temp. between the upper and lower regions of vessel a,re.'within 145 des.F (0.50)

c. Closely monitor the APRH cher t recorder s. (0.25)

The increased flow wi,ll greatly affect RX. Power. (0.25)

d. On ' A

• stake the break. release Icvor to the release position. (0.25)

On

• B, tighten both red knobs-on top of the break. (0.25)

REFERENCE-

.IAF. dF-OP-27 Recire. sys. sPs. 12 -17. +

A SWER 2.05 (3.00)

a. 1. - psig

2. psig

'.~

3. psig (0.25 cach) ##I g b.. During Reactor Isolation (.25), th RHR sys. in the Condensing Mode- is oper ated in conjunction with the IC sys.(.25 ), in the caso of. lass of main feedwater flow (.25), to reduce or maintain RX. press.(.25).

c. TRUE . (0.25)(

d. Condition 4 1- Containment Spray contyo,1, switch in Manual (.166) and containment pressure of 2.00 psig.(.166) and containment spray override

• keylock switch in~0verride (.166). (0.50 total pts.)

Condition # 2- Containment spray control switch in manual (.125) AND

-containment pressore of 2.00 psig. (.125) AND vessel icvel '

0 * (.125)

. AND LPCI initiation signal SEALED-IN.(.125). (0.50 total pts.)

REFERENCE JAF. LP . Book i 2- lab-0 , SDLP t 10 RHR. Sys. and F-OP-13, RHR Sys.

p.

b L_ _ . _ _ .,,..-.m m-_...y_m. _ _ . _ . . , , _ , , , .____m. _ , , _ .. _____._.__%

u.

2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- FIT 2 PATRICK -05/02/12-LANGE, D.

F

(/ANOWER 2.06 (3.00) q a.1 - high drywell pressure > 2*e6 psis.

2 - low reactor water level (* 10 *) abw* Td ' E' 3 - 4 kv omorg. bus low volta 3c-(0.33 each) i b.TRUE (0.50)

c. 1. Ensino Overspood

2. Loss of generator field.

3. Generator protective relays

4. High circulating current ( roverse power )

5. Low lobe oil press. BLOCKED IF LOCA SIGNAL PRESENT

6. High Jacket water temperature. BLOCKED IF LOCA SIGNAL PRESENT l -

i (0 25 for each correct answor)

REFERENCE l

.l AF . Proed. 4 F-OP-22, D/G Emers. Power. and LP. 4 SDLP-71 A & De 90 .

t NSWER 2.07 (2.50)

a. 1.High area temp.

, 2.High steam lino flow. ( 4 required at 0.25 each )

! 3. Low steam supply pressure.

4.High turbino exhaust diaphragm pressure.

f b. NO.(0 25) Following a turbine trip from any reason, the trip throttle L valve must be manually roset at the turbino. ( 0.50 ) - eset/7' M /**W -

f- c. To minimize the possibility of a water hammer or flow roversal in the l turbino exhaust line. ( 0.75 )

REFERENCE

.lAF. F-OP-19, pg.4 4 & 8.

j hANSWER 2.08 (2.75) h a. 1. Continuity lights so out.

2. Alarm indication.

3. Milliamp motors on back of 09-03 indicato current flow to firing okt.

L

4. Decrease in power.

5. Selected pump has red light indicating pump is running.

6. SBLC pressure > reactor pressure.

7. SBLC tank level decreasing. ( any six at 0.25 each )

b. YES. (0.25) If the SBLC. tank level approaches =cro (0.50) or'the SBLC pump begins to loose dischargo pressure. (0.50)

REFERENCE

.lAF. F-OP-17 pg. 4 thru 7.

4 7* * ~ ~ ~ ' * * * ' ' - '

' -== ~ ~ N C __._

2. PLANT DESIGN INCLUDING SAFE 1Y AND EMERGENCY SYS1 EMS PAGE 24 ANSWERS -- FIT 2 PATRICK -85/02/12-LANGE, D.

I gj ANSWER 2.09 (1.75)

a. 1.~R-4A or 40 high di)otion flow (.33)

2. R-SA or 40 low dilution flow (.33)

3. R-'4A or 4B low outlet temp. (.33)

b. A detonable mixture of hydrogen and oxygen may accumulate under these conditions. (0.50)'

'c. Radiation Protection Department. (0.25)

REFERENCE JAF. F-OP-24.A Off-Gas Sys. pg. 1 - 23 ANSWER 2 10 (1.50)

a. 1. Used to calibrate individual LPRM detectors. (0.25) to map the core axial flux profile. (0.25)

2. Used by the-prosess computer-(0 25) to determine MCPR and local heat flux conditions. (0.25)

REFERENCE JAF. .SDLP.-07F 1IP sys. pg. 6.

- . 1 2

f.

• 4-

3. INSTRUMEN1S AND CONTROLS PAGE 25 ANSWERS'-- FI1ZPATRICK -

-85/02/12-LANCE, D.

ANSWER 3.01 (2.50)

a. Rx. Scram = 177'

b. SGIS starts'= 177'

c. RCIC Init. = 126.5' ( 0.25 for cach correct ans. )

d. Low Level alarm = 196.5'

o. Recire. Pump 1 rip = 126.5'

f. 'Rn. Feed Pump Trip =222.5' No. 2- lho ATWG Recire. Pump 1 rip is set at 1120 psig.(0.20 . The first two relief valves lift at 1090 psi 3., (0.25), the second two at 1105 psis. (0.25) and the last seven at 1140 psis (0.25).

REFERENCE

.l AF . SDLP. 02 B, C, D, E, F,- Reactor Vessel Instrumentation

' ANSWER 3.02 (3.00)

'a. half-scram

b. no action

!i c. half scram

d. rod block

e. scram

f. rod block

-(_ (0.50 for each correct ans.) (3.00)

REFERENCE

. .lAF . SDLP.-071 RMCS.

SDLP.-02H,021. Recire. sys.

GDLP.-07A-E Neutron Monitoring Sys.

ANSWER 3.03 (3.00)

a. 2, 4, 7, 4F*

b. 1, 3, 9

c. 5, 6 (0,3G cach) 41f REFERENCE JAF. SDLP.- 17,18'( Process and Arca Rad. Monitors )

SDLP. 38 8 01A ( Condenser Air Removal and Off Gas )

c

r 4

3. INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS'- FIl2 PATRICK -85/02/12-LANCE, D.

ANSWER 3.04 (3.00)

'a. A complete _ loss _of RPS will cause a scram . The Recire. pumps will run back to 26 % when less than 20 % feed flow is sensed, or vessel Icvol is less-than 196.5 without both feedwater pumps running. (0.75) b.: .The EHC pressure regulator will sense a decreasing pressure at the avs.

press.1 manifold and close the TCV,s . The tur bine will trip.on reverse

. power. The BPV,s will open if the avs. manifold press. rises abovo 920 psis. (0.75) c '. Lovel will initially shrink duo'to void collapse and then return to near normal with the smaller void reformation following the scram . (0.75)

d. RPS does not supply power to RX. lovel instrumentation.No effect.(0.75)'

REFERENCE'

.I A F . SDLP.171 a & b , & 93. . RX. Level & Inst.

SDLP- 94-C, EHC sys.

F-OP-27, Recire. Sys.

ANSWER 3.05 (2.50)

a. Loss of control signal.

MG. Set' control power transfer ( 3 correct ans G .30 each )

Loss of power to the scoop tube' positioner

b. 1. FW flow < 20 % runback to 26 % sp'ced'to protect against cavitation.

[oge) (0.50)

2. RX. . level < 196.5" without both'feedwater pumps running, runback to '

44 % to protect against cavitation. (0.50)

c. . Rod runback light on panol 09-4 (0.25)

Recire. Flow Limit Annunciator. (0.25)

REFERENCE JAF. Procd. 4 F-OP Recite,

. Sys. pg. 11-13 .

'\s4HSWER 3.06 ( 2 .,2 5 )

a. . EXPLOSION in'the AidIEjector discharge piping. (0.75)

_. Automatic Actions;

1. Air Ejection steam supply valve, 29-PCV-107 shuts. (0.25) 2.Possible. isolation of. Turbine Blds, on high airborn activity. (0.25)

'3.Hain Turbine and RX. feed pump trip. (22.5'hs) (0.25) 4.HSIV and Bypass valve closure on low vacuum ( 8'hg ) (0.25)

b. Manually Scram the Reactor. (0.50)

REFERENCE JAF. , Special Procd.F- AOP-4,'E:<plosion in the Air Ejection Off Gas Linc.'

( Symptoms / Automatic Actions / Operator Atticm )

=. _._._.___ _ _ _ . _ , . _ _ _ _ . _ _ _ . , . _ _ ..___ _

+

3. INS 1RUMEN1S AND CONTROLS PAGE 27 ANSWERS -- FI12 PATRICK -85/02/12-LANGE, D.

ANSWER 3.07 (3.00)

O. TRUE (0.50)

b. A stainicss stcol jamin3 button must be installed on the hoist cable.(.5

c. During refueling it provides a seal between the vessel flange and the

-drywell. (0.25)

During heatop and cooldown it accomodates the differential expansion that occurs. (0.25)

d. 1..Hore than one control rod withdrawn. (0.25) 2.1 Ref ueling platf or m position switch open. (0.25)

REFERENCE JAF. F-OP-66, Refueling Equipment. pg. 51 JAF. SDLP-19,-Fig. 19-2 Refuelin3 Bulkhead and associated Bellows.

JAF. SDLP-97/A, Fool Handeling. Equipment.

ANSWER' 3.08 (2.00)

A differential pressure sensor is used to confirm the integrity of the 1

-CORE SPRAY. piping within the reactor vossol ( betwoon the inside of-the vessel and the core shroud).

To continuously monitor the integrity of the core spray piping, a Delta P l cwitch measures the pressure difference between the two. loops, which is effectively'the inside of each Core Spray sparager pipe, just outsido of the Rx vessel. If the core spray sparagor is intact, this pressuto

-difference will be =cro. If integrity is lost, this pressure differential eill include the pressure drop across the steam seperator. Alarms at,5 psid in the control room (2.00)-

REFERENCE

.lAF. SDLP.-14 , Core Spray Sys.

JAF. Performance Objective 4 4014 .3-02, Lic. Oval Std.

JAF. Annunciator response pracd. Vol.-1, alarm 9.3.3-1

'\/(NSWER 3.09 (2.00)

c. GEMAC. and FUEL ZONE Icvel inst. are accurate. ( 0.50 each )

The WIDE RANGE YARWAY lovel inst. are not reliable. (0.50) reasoni Flashing in the refrence Ics. (0.50)

REFERENCE l

• JAF. F-EOP-1 pg. # 15 .

u

.I:

-l

~ 3. INSTRUMEN1S' A'ND CONTROLS PAGE 28 ANSWERS -- FITZPATRICK -85/02/12-LANGE, D.

ANSWER- 3.10' (1.75)

.a. 1.cRestricts discharge and protects vessel internals from large D/P.

(0.33) 2.'Provides. signal for MSIV closure.. -(0.30)

3. Provides~ steam' flow signal to FWCG. '(0.30) b.'Three solenoid operated pilot valves. (0.25) 1.OnoLAC and one DC for' normal operation. . (0.25) 2.'One? test pilnt solenoid.for slow closure testing.-(0.25F REFERENCE '

-JAF. F-OP-1. Main Steam, pg. 3-6.

)

1 c

a e

n 7

4

.4.-. PROCEDURES - NORMAL, ABNORMAL,. EMERGENCY AND PAGE- 29

'~ ~~~~~~~~~~~~~~~~~~~~~~~~

~~Rbb blUU5UbL'UUU5RUL ANSWERS -- FI1ZPATRICH -85/02/12-LANGE, D.

ANSWERp 0 ,

(2.00)

c. 1.'Sonsitivity of the instrument to the types of radiation.and/or con-tamination present.

f2. Ranges of the instrument with respect to expected 1cvels of rad -

iation.

3. .. Limitations of'the instrument with respect'to humidity, temperature, etc.

4. Current calibration of the instrument.

5.-Proper response to check sourecs. (any four.at .25 cach)

b. #1=b, 42=a, #3=d, 44=c (0.25 for cach correct ans.)

= REFERENCE-

.!AF.-Radiation Protection Procedure 2.4.3

.NSWER 4.02 (1.75)

a. 1. .Misoperation-in automatic is confirmed by at 1 cast two independent

-process parameter indications. (0.50)

2. Core cooling is assured._ (0.50)

b. 1. Maintain Core Cooling.

2. Limit the release of off gas radiation.

3... Place the-Reactor-in a safe stable condition.

4., Keep the Torus bulk-temp. below 120 deg.-F. (any 3 at .25 ca.)

' REFERENCE-

.l AF . F- E0P-33,-Small Break Accident.

ANSWER 4.03- (2.00)

Adequate core cooling is not assured. (1.00)

With both Core Spray loops out of service and only one RHR-pump available e .for injection, the definition of adaquate core cooling cannot be' met.'

REFERENCE Mitigation of Core Damage , JAF. RX. Core Cooling Bases. Sec. 3.5, pg. 126 of Tech. Spoes.

a

- w

~4

'4..' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE' 30-

~~~ EE5iBE5GiE3E C5 iE5E-------------------~~---

ANSWERS -- F112 PATRICK -95/02/12-LANGE. D.

ANSWER .4.04 (2.25)

Three events; 1.. Decrease in Supp. pool 1cveli As Icvel decreases the nitrogen must oc-copy a: larger volume, .resulting in decreased Drywell pressure,

2. Increase .in barometricpressurci The Drywell pressure ins tr onients are

. all '.r ef r enced to atmosphere therefore an increase in atmospnetic press.

rimses a :dect case in D/W-press.

3.-A decrease in D/W or chilled. water to nspi An increase in cooling cap-

.acity in.the D/W from.the D/W coolers and chillers wi11' decrease D/W pressure. (Other rossonable ans. accepted if substantiated).

( 0.25 for cach event; 0.50 for cach reason )

2 REFERENCE JAF. Procedure,F-AOP-9 Loss of Primary Containment Integrity.

ANSWER 4.05 (1.50) a.-1) .75 R

2) 25 R '

' 3 ). 12 R REFERENCE JAF. Radiation Protection Procedures 2.7.4 , Emergency Exposure Guidelines.

ANSWER 4.06 (2.00)

.Immediato O'perator Actions; 1.1 Place the Modo Switch in Refool. (0.33) 2.= Insert IRM and SRM detectors into the Core. (0.33)

3. Verify the Rn. is shutdown by observing Power decrease. (0.03) 4.-Confirm all control-rods are inserted to or beyond position 00 by i a.-bypass the SDV. high 1cvel trip.

~

(0.165)

'b; Attempt to reset the Scram signal. (0.165)

5. Once the RX. is shutdown Trip.the Main Turbine. (0.33)

6. Roset N-2 supply to the drywell if it had isolated from a, group -2 isolation. (0.33)

REFERENCE JAF.-F- APO-1, .Roactor Scram.

i t

a-** 'W e . *  %.

4 .1 PROCEDURES -nNORMAL, ABNORMAL, EMERGENCY AND' PAGE 31'

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

R bEUL5EiEst C5sTR5t

. ANSWERS.---FITZPATRICK -85/02/12-LANGE, D.

HSWER 4.07 (2.'50) a'. With HPCI or RCIC, if availabic . (0.75) b.'Tho'SRV that dischargos to the torus as.far away as possible from the first SRV that actuated. (0.25) To minimine local heating of the torus water.-(0.50)

L c.'ti- Open MSIV,s and control pressure with the turbine bv pass valves.

OR,' (0.50) 42- Operate the RCIC. sys. in the Steam Condensing mode.(0.50)

REFERENCE

.lAFNPP.F-OP-1 sec 5 pg 15 & 16 ANSWER- 4.08 (2.50)

a. Refueling.GENAC level indication. (0.50)

b. .To assure adaquatcl cool' ant mining thru natural circulation. (0.75)~

c.. Vessel stratification ..

Loss of valid temperature indication. (2.5)

Extreme case of. vessel boiling and pressuri:stion.

(any two (2) at .75 cach)

' REFERENCE.

F-OP.-13.RHR System pg 13 ANSWER. -4.09 (2.75)

a. 1. Acoustic monitor (alarm).

2. Tailpipe. temp. (alarm).

3. . Generator (cicetrical output). -( any five (5) G O.25 cach )

4. . Torus temp.

5 RPV.. pressure and water oscillations.

6. SORV' solenoid indicating light (showing energized state).

b. Direct an operator to panel 09-45 ( relay room )-(0.25) and have him pull the control power fuses for the SORV. (0.50)

c. 1. Reactor building - 300 ft. clev. North wall. (0.50)

2. Try-cycling the C/S'again, and if no action,open PS. ckt.brk inside

'the Remote Relief Valve Panel. (0.50)

REFERENCE

.lAF.'Procd. F-AOP-36, Stuck Open Relief Valve.

i_. _ _ .

_ . . . - - . . ..~...s __.mm- _ _ . _ _

7:

t

4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32

~~~ ~ R 5D 26L66565L "6 6 UiR UE ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~

ANSWERS -- FIT 2 PATRICK -85/02/12-LANGE, D.

NSWER 4.10- (3.00)'

a. 1-Scram the reactor (0.5) 2-Trip ~the. main; turbine (0.5) 7-Leave the Mode switch in Run (this insures MSIV closure)'(O.50)

b. RX. bids. 300 ft.-elev.. NE. inst.. racks. (0.50)

c. Manually. trip the turbine at the front. standard. (0.25)-

~

De-energico RF3 from-Dist. panels in the reley r oom. (0.25's

llpen RPS MG set-_ supply breakers. (0.25)

Isolato and vont inst. air to scram valves. (0.25)

REFERENCE JAF. F- E0P-28 Plant' Shutdown From Outside The Control Room.

- ANSWER 4.11 (2.75)

a. 1. -RPV water 1cvel below 177'

2. RPV press.above 1045 psig.

3. D/W press.above 2.7'psig.

-4. MSIV_isol.' required. .

5. RX. scram required, with RX. power > 2.5 % or cannot be determined.

6. When directed by another EOP.

(0.25 for each correct ans.)

b. 1. Supp. pool avs. water- temp.> 75 dos.F

-2. D/W avs.tcap. > 135 deg. F

3. D/W press. > 2.7 psis.

4. Supp. Pool water 1cvel > 0.0 in.

5. Supp. Pool water level.< 1.5 in.

(0.25 for each correct'ans.)

= REFERENCE

.l AF . E0P-2 and E0P-4.

'N

- .- - s -- 4 - , , , . . .-

ru' ' ^

3.,

1 .

' TEST CROSS REFERENCE PAGE 1 DUESTIDN' VALUE REFERENCE

-01.01 2425- DJL0000090 01.02 2.00 DJL0000091.

01.03' 2.00 DJL0000092-01.041 ~2.00 DJL0000093

?01.05 2.50' DJL0000094 01.06 2.50 .DJL0000095 01.07 3.00 DJL0000074:

01.00 13.00 DJL0000116 01.09- 2.25 _ DJL0000127 01.10 c 2 .' 0 0 DJL0000130

.01.'11 1.50 ' DJL0000131

- 25.00 02.01. 2.50 DJL0000097

-02~.~02 '2.00 DJL0000098 02.03 3.00 DJL0000099-H02.04 3400 DJL0000100-02.05 3.00 - DJL0000101

-02.06 3.00 'DJL0000102

-02.07' 2.50- DJL0000121 02.08 2.75 DJL0000122 02.09 '1.75 DJL0000133.

02.10 1 ~. 5 0 DJL0000134 25.00 3- 03.01~ 2.50' DJL0000103 03.02. 3.00 DJL0000104 03.03 -

3.00 - DJL0000105 03.04 3.00 DJL0000106 03.05 2.50 ' DJL0000124 03.06 2.25- DJLOOOO17.5.

.03.07_. 3.00 DJL0000126

-03.08' 2.00 DJL0000127 03.09 2.00 DJL0000128 03.10 1.75 DJL0000136 25.00 04.01 2.00 DJL0000108 04.02 '1.75 DJL0000110 04.03 12.00 - DJL0000111 04.04. 2.25 DJL0000112 04.05 1.50- SDJL0000113

- DJL0000117 l -04.06 2.00 H04 . 07' -2.50

'DJL0000118 04 0S' 2.50' DJL0000117' 04.09 -2.75 ' DJL0000123

.04.10 3.00 - DJLO.000132 t-

].

, , , **O

m. 4 %*=* _ - m h ,h sn.h -m._ -_(__

f

.y s

h' TEST CROSS REFERENCE PAGE 2 OUESTION VALUE REFERENCE 04.11 2.75 D J L 00 001::".

25.00 100.00 J

l i

. l Question and Answers to James A. Fitzpatrick SR0 Exam - 2/12/85 5.0 Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics (25.0) 5.1 Assume Fitzpatrick is at 100% power and flow. Condenser vacuum decreases thus causing the hotwell temperature to increase to the saturation temperature of the new vacuum. This in turn causes the temperature of the feed water to increase.

a) Would Reactor power increase or decrease? Explai n. (1.0) b) Would Core Flow increase or decrease? Explain. (1.0)

Answer:

a). Decrease (0.25) Increased temperature of feedwater would increase negative reactivity insertion of moderator coefficient (0.75).

b) Increase (0.25) Core Flow would increase due to reduced two phase pressure drop across core. (0.75).

Ref. NET 237.4 pg.13 H228.8 pg. 31 H228.8 fig. 3-4.

5.1 x

l

.4-5.2 'Will the recirculation pumps have more NPSH at 4% or 100% power? Explain your answer fully. (2.0)

Answer:

More NPSH at 100% power. (0.5)

At 4% you are at operating pressure but low feedwater flow the temperature at the recirculation pump inlet is nearer to saturation than at 100% power where there is a higher percentage of feedwater in the downcomer causing the temperature at the inlet of the recirculation pumps to be lower. (1.5)

Ref. MET 214.9 pg. 16.

5.2

... i 5.3 State whether the following increase, decrease or have no effect on Control Rod worth, a) Rod Density decrease. (2.5) b) Gadolinium burning out, c) Void fraction decrease.

d) Moderator temperature decrease.

e) Fuel temperature increase.

Answer:

a) Increase b) Increase c) Increase d) Decrease e) No effect (0.5 for each correct answer)

Ref: NET 237.4 pg. 19, 18.

A e

4 5.3

~ _

I l

5.4 While Fitzpatrick is operating at 90% power, extraction steam l to the highest pressure feedwater heater is removed. An i engineer observed that the turbine load increased by 20 MW l electric and concluded that this action has improved (increased) the plant's thermodynamic efficiency (not heat rate).

Is this conclusion correct? Explain your answer fully.

(Include what caused electrical output to increase). (2.5)

Answer:

No (0.5) thermo efficiency is a comparison of Energy In to Energy Out (0.5). The _ increase in output results from no steam being diverted to the high pressure feedwater heater (0.5). Because the feedwater is now cooler, more energy from the reactor is required to bring the water up to saturation temperature (0.5) thus thermo efficiency is down (0.5).

Ref. H-228.8 pgs. 22-23.

t 5.4 1

. .. . _ _ . _ J

i j

l 5.5 Fitzpatrick has been operating at 100% power for 3 weeks, when the Reactor scrams. You begin startup within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

a) Which control rods have the highest worth? Explain. (1.0) b) ~ Briefly explain why you must approach criticality very carefully under these conditions. (0.5)

Answer:

a) Edge Rods (0.5) Xenon poisoning is highest where the highest flux was, in the. center of the core - thus edge rods have the highest worth.(0.5) (1.0) b) Because the high differential notch rod worth of edge rods can cause short periods. (0.5)

Ref. NET 237.5 pgs. 2 and 4, OP65 pg. 6.

5.5

5.6 What are the occur with Ugmes ofduring fuel the twocore major neutron life? reactions For each which reaction state whether the reaction produces a positive or negative reactivity effect and which (if any) of the reactions occur as a result of theTppler effect? (2.0)

Answer:

Fast Fission (0.25); Resonance Capture.(0.25) (0.5)

Fast Fission + Positive reactivity (because it contributes (0.5)

.more neutrons to overall neutron population). Resonance Capture + negative reactivity (because it absorbs neutron - (0.5) thus reducing overall neutron population).

Resonance Capture.

(0.5)

Ref. NET 227.1 pg. 4 NET 227.2 pg. 4.

5.6 l

1 i

__ _3_.. _ _ ,__.-. . - - . _ -..

5.7 ~ During a backshift, Standby Liquid Control (SLC) was inadvertently initiated and almost immediately stopped while the plant was at 90% power. You determine that the total SLC pump run time was [15 seconds]. While making your reports to plant management, you are asked to determine.if SLC ran long enough to inject any boron into the Reactor vessel. From other sources available in the control room, you have the following information:

1) Length of pipe from SQUIB VALVE to REACTOR VESSEL - 100 FEET

2) SLC piping - 2 INCH INSIDE DIAMETER PIPING

3) SLC pump capacity - 40 GPM a) Did boron reach the vessel? (1.0) b) How long did it, or would it take for boron to reach the vessel? (ShoTa11 work, consider instantaneous pumn capacity). (1.5)

Answer:

a) No (1.0) 2 b) Volume - nr L = (3.14) 1 2(100) = 2.18 ft3 TE 3

(2.18 ft ) x 7.48 = 16.3 gallons. (1.0) ft (16.3 gallons)/(40gpm) = .408 min = 24.5 seconds. (0.5)

Ref. NRC Exam Bank 0005421.

MET - 214.3 pg. 5.

j 5.7 l

l l 1

- - ~ - _ ~ _

+

5.8 Fitzpatrick has three (3) thermal safety limits. List these limits and briefly describe the purpose of each. (3.0)

Answer:

1) MCPR (1.07) - to insure 99.'9% of the fuel rods do not experience transition boiling during a transient.

2) LHGR - to 13.4kw/ft EOC - to insure 1% plastic strain on cladding is not exceeded.

3) MAPLHGR - to insure that clad temperature remains below 2200*F during loca.

(0.25 for each limit, 0.75 for proper description).

Ref. Heat Transfer H-228.9 pg. 39.

i 5.8

, , . . . . . < ~**

9

5.9 Concerning the figure given below (for thermodynamic cycle analysis):

a) Identify the points corresponding to the high pressure turbine inlet, and low pressure exhaust to the hotwell. (0.5) b) Explain why there is no temperature increase between points 4 and 1. (0.5) c) What does the horizontal portion of the line between point 2 and point 3 correspond to. (Be specific). (0.5)

P O. 3

] P u

1 J 2, Entropy ,

Figure 1

. Answer:

a) 3, 4 (0.5) b) Vapor is condensing to liquid in a saturated environment which is at the saturation temperature in the condenser. (0.5) c) The vapor generation portion of the core where liquid and vapor are saturated. (0.5)

Ref. Thenno lesson Plan MET 222.10 pg. 5.

'I 5.9

5.10 During startup (power = 1 watt) a rod is pulled and a 60 second period is observed.

a) With no further pulling of rods, can the operator maintain this Reactor period for Wminutes?

Exclain your answer. (1.5) b) By pulling rods can the operator maintain the 60 second period for 30 minutes from the 1 watt power level?

Explain your answer. (1.5)

Answer:

1800/60

)

P = Po et /T = 1 watt e sec = 1.06 x 10 13 watt

= 10 mw No (0.5)

Moderator coefficient would turn it. (1.0) b) No (0.5)

Power at end of 30 minutes would even exceed max. MWT for Fitzpatrick. (1.0)

Ref. Equation sheet.

5.10

e 5.11 Fitzpatrick is operating at full power when the Feedwater Controller malfunctions causing a complete loss of feedwater.

A Reactor Scram will occur on low level. During the time period just prior to the scram, is the Reactor Power expectedfto increase or decrease? Give two reasons for your answer. (2.5)

Answer:

Decrease (0.5)

-1) Loss of feedwater causes less subcooling, more negative AK/K due to more voiding. (1.0)

2) Recirc. Runback on <20% feedflow decrease core flow. (1.0)

Ref. NET 237.3 pg. 8, F-0P-27 pg. 3.

End of Section 5 l-3 j 5.11

A

6.0 Plant System Design, Control and* Instrumentation (25.0) 6.1 Concerning the diesel generator system.

a) With the engine control switch in standby what conditions or actions will cause Emergency Diesel Generator No. 2 to Start? (Include set points). (1.5) b) If the field flash circuitry of the diesel generator failed to operate, will the diesel generator still start and produce power? Explain. (1,0) c) Explain the ECCS loading sequence on the Diesel generator from the time of initiation to full loading. (Give time, in seconds. Include times that diesel and loads start and are at speed). (1.0)

Answer:

a) High Drywell Pressure (0.25); greater than 2.7 psig (0.25) low Reactor level (0.25); 59.5 inches (0.25) 4KU emergency bus low voltage. (0.25)

Control room switch to start. (0.25) b) Yes (0.5)

The generators are still capable of energizing due to residual magnetism. (0.5) c) 0 second Diesel initiation.

10 seconds Diesel up to speed.

11 seconds First RHR pump starts.

16 seconds Start second RHR pump.

(First RHR pump up to speed).

21 seconds Start core spray pump.

(Second RHR pump at speed).

(0.2 for each correct answer, values acceptable 210%).

(may also get that if one diesel does not start, the second RHR pump will not start.)

Ref. F-0P-22 pg. 6, 7, and 39.

6.1

4 6.2 With regard to the Reactor Core Isolation Cooling System (RCIC).

a) What five (5) signals cause automatic isolation of the RCIC Steam line isolation valves 13-MOV-15 and 13-MOV-16? (1.0) b) If the RCIC turbine received and inadvertent trip signal, which immediately cleared, would the turbine restart on a valid auto initiation with no operator action? (Briefly explain). (1.0) c) There is a caution which states "Do not operate the RCIC turbine at a speed below 2,200 RPM for an extended period of time". What is the reason for this caution? (1.0)

Answer:

a) Manual (only if auto initiated)

High Area temperature.

High Steam line flow.

Steam Supply Lower Pressure.

Turbine Exhaust Diaphragm High Pressure.

(0.2 for each correct answer)

(Manual not necessary. If not included 0.25 for each correct answer) b) No,(0.25)

Following a Turb'ine trip from any cause, the trip throttle valve must be manually reset at the turbine. (0.75) c) To minimize ~ the possibility of water Hanner (0.5) or flow reversal (0.5) in the turbine exhaust (may also say reduced speed will reduce cooling water to barametric condenser and lube oil - no credit taken off for this answer)

Ref. F-0P-19 pg. 4 and pg. 8.

1 i

6.2 I l

1 4

6.3 As regards the Standby Liquid Control System (SBLC) Operating Procedure (F-0P-17).

a) Af ter initiation of the SBLC system, is it permissable to shut the system down? (If not why, if so under what conditions?) (1.5) b) After a valid initiation, list six (6) control room indications that you could use to verify that the SBLC system is operating properly and injecting into the reactor vessel. (2.0)

Answer:

Yes,(0.5) a) Control tank level approaches zero, pump begins to loose discharge pressure, or all rods in per E0P-3.

(0.5 for each of two correct answers) b) Continuity lights go out Alarm indication Milliamp meters mounted on back of 09-03 panel indicate Low current flow to firing circuits Power should decrease Selected pump has red light indicating running SBLC pressure >RV pressure SBLC tank level decreasing ,

RWCU system isotates.

(0.33 per correct answer, any 6).

Ref. F-0P-17 pg. 4, 5, 6 and 7.

6.3

6-6.4 With regard to HPCI:

a) .Where are the HPCI rupture diaphrams located and.

what are their purpose? (1.0) b) Describe the valving sequence of the HPCI suction valves from

-the CST and torus during an auto transfer from the CST to the torus. Briefly state the reason for this sequence. (1.0)

Answer:

I a) 0n, the exhaus*. line in the HPCI room (0.5) to protect the exhaust line from over pressure. (0.5) ,

b) The CST suction valve stays open till the torus suction valves are open. (0.5) This is to insure a continuous suction to HPCI during the change over. (0.5)

Ref. F-0P-15 pg. 6, SDLP 23 pg.13.

f.

i 6.4 d --

4 1M $ er Odu! 9 us g aw -

a-m % m -- ----- r-ww -- *-r-r

x :

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6.5 ~ Concerning the Main Steam System:

a) FolloEing a failure of the air supply to the accumulators how

• many times can an operator actuate (manually cycle open and close) a relief valve from the air supplied by the valve's accumulator? (0.5) f b) What are the two (2) purposes of the valve interlock which pemits two of the four steam lines to be closed without scra:mning the Reactor. (1.0) c) List Three (3) functions of the Main Steamline Flow restrictors.

(1.5)

Answer:

a) Five (5). (0.5) b) To provide flexibility for testing (0.5) and to provide for possible malfunction of steam line isolation valves during plant operation. (0.5).

c) 1) Restricts discharge

2) Protects vessel internals from large ap.

3) Provides signal for MSIY closure.

O Provide steam flow measurement to FWCS.

e 5) Steam flow indication on 9-5 panel (0.5 for each of three correct answers)

Ref. F-OP-1 Pg. 4.

6.5 i

t

6.6 You suspect that a relief valve is stuck open. List five (5) instrument indications in the control room, including back panels, that you would use to confirm that a relief

' valve was stuck open. (2.5) [

Answer:

1) Acoustic monitor (alarms).

2) Tailpiece temperature (alarm).

3) Generator (electrical output).

4) Torus temperature.

5) RPV pressure and water oscillations.

6) Relief valve solenoid indicating light (showing energized state).

(0.5 per correct answer, any 5).

Ref. FSP-06 pg. 2 of 3.

7) Steam flow feedflow mismatch and level decrease.

8) Plant efficiency decrease.

9) (Other acceptable answers.)

6.6

s 6.7 Concerning the Recirculation System (Operating Procedure F-0P-27).

List the two (2) conditions that will cause a recirc runback, the percent of pump speed the runback is set

. for and explain the reason for'the runback. Al so l

which of the runbacks seal in? (2.0) l i

l Answer:

Feedflow less than 20% (0.25) or discharge valve not full open (0.25) causes a recirc runback, to 26% recirc (0.5) to protect against cavitation, prevent axial thrust on pump with discharge valve not full open (0.5).

Reactor Vessel level less than 196.5" without both feedwater pumps running (0.5), limit is 44% (0.5) to get you into the range of one feedpump to prevent low level scram.

The 44% runback.

Ref. F-0P-27 pg. 4,15, and 13 l

I F

6.7

6.8 The reactor water cleanup system (F-0P-28) is operating in the blowdown mode and discharging to the condenser. Indications are as foTlo7s:

Cleanup inlet flow = 120 GPM Cleanup flow to condenser = 50 GPM Cleanup flow to reactor is = 0 GPM (leak in pump room)

Non Regen outlet = 225*F Using these indications list three (3) auto actions that should have occurred in the reactor water clean up system (include the parameter (s) which should have caused each action. Setpoints are required. (3.0)

Answer:

a) System should have isolated, inboard and outboared; isolation valves should have shut (0.5) due to high nonregenerative Hx outlet (0.5) (140*F) (0.5). Leak should have caused high temperature isloation in the pump room (0.5) at T ambient +40*F (0,5), RWCU pumps should have tripped due to closing of any isolation valve (0.5).

Ref. F-0P-28 pgs. 3 and 8.

f I

6.8

6.9 Concerning neutron monitoring instrumentation:

.a) What is required regarding LPmi inputs to an APRM, for the APRM to be considered operable? (1.0) b) If the requirements of part "a" were not met, would you be required to manually trip the circuitry? Explain why.

. (1.5)

Answer:

a) ~To consider an APRM operable there must be at least two (2)

LPRM's per level (0.5) in addition to a total of at least 11 LPRM inputs must be operable. (0.5) b) Yes,(0.5). If the total number of LPRM's falls below 11, an APRM inop. signal would automatically be generated by the circuitry but if the 2 LPRMS per level is not met, the circuitry will not auto generate an inop. This must be manually initiated. (1.0).

Ref. F-0P-16 pg. 9 End of Section 6 6.9

-6 7.0 Procedures - Normal, Abnormal, Emergency and Radiological Control (25.0) 7.1 In accordance with F-AOP-1 " Reactor Scram":

a) Following a reactor scram from full power, if reactor vessel is-not isolated, how is reactor wate'r level and pressure to be maintained? With the Reactor vessel isolated, how is the reactor water level and pressure to be maintained? (1.5) b) If instrument nitrogen is lost to the drywell instrumentation what must be done? (0.75)

Answer:

a) Level-Feedwater pumps if possible, RCIC and/or HPCI. (0.75)

Pressure - Bypass valves Level - RCIC and/or HPCI. (0.75)

Pressure - steam condensing mode of RHR, HPCI, RCIC and/or safety relief valves.

b) Line drywell instrumentation up to instrument air, only if instrument nitrogen can not be restored. (0.75)

Ref. NRC Exam bank RRP0000203 F-A0P-1 pg. 3 and 2.

l 7.1

- -- - -- -- - -. - . - - - i

6 -

7.2 With regard to the Reactor Building Closed Loop Cooling System (RBCLC):

a) Your. are cautioned against large temperature swings in-the RBCLC. What are the two-(2) purposes of this caution (i.e., what adverse effect could a rapid temperature increase have and what adverse affect could a large temperature decrease have)? (2.0) b) During the winter, if the temperature of RBCLC can not be held at or above 75'F,- with TCV-101 fully open and two coolers in operation. What action should you take? -(1.0)

Answer:

a) A rapid increase in temperature could raise drywell pressure significantly. (1.0) decrease could cause drywell to torus differential to go out of specification low (?_1.7 psig). (1.0) b) Throttle service water flow and if still can't reduce temperature, remove one of the operating Hx from operation. (1.0)

Ref. F-0P-40 pg. 5 7.2 1

'7.3 A severe fire causes the NCO to leave the control room before he can shut the plant down. List the _four (4) methods, in order of preference, that the NC0 is supposed to do to scram the Reactor. (2.5)

Answer:

1) Manually tripping the turbine at the front standard.

2) De-energizing the RPS from the distribution panels in the relay room (panel 5-6A, RPS A; Panel 5-6B, RPSB).

3) Opening the RPS MG set supply or output breakers in the electric bay.

4) Isolating and venting instrument air to the scram valves.

Underlined items not required; 0.5 points per right answer, 0.5 points for proper sequence).

Ref. F-AOP-43 pg. 2 of 4.

7.3

7.4 According to F-A0P-12 (Loss of-Instrument Air):

a) Identify four (4) automatic actions that should have.

occurred as instrument air header pressure decreased to 80 psig. (2.0) b) Does a loss of instrument air require a manual scram?

If not - why and if so under what conditions? (1.0)

Answer:

a) Loading of operating air compressors at 110 psig decreasing.

Start of standby air compressors at 100 psig decressing.

Closure c 39-FCV-110, service air isolation, at 95 psig decreasing.

Closing of 39-FCV-111, breathing air isolation at 85 psig decreasing. '

(0.5 per correct answer, set points not required).

b) Yes - If pressure continues-to decrease and corrective actions cannot be accomplished immediately, or if an automatic scram occurs. (1.0)

Ref. USNRC Exam bank RRP0000201 F-A0P-12 pg. 2.

7.4

• I 7.5 During rapid RPV depressurization to below 500 psig:

a) What level instrumentation will provide an accurate measure of vessel level? (1.0) b) What level instrumentation is not reliable under this condition and why? (1.0)

Answer; a) Gemac and Fuel Zone.

(0.5 for each correct answer) b) Wide Range Yarway; Flashing in the reference leg.

(0.5 for each correct answer).

Ref. F-EOP-1 pg. 15.

7.5

7.6 According to E0P cautions (F-EOP-1) under what condition (s) can-an ECCS. system be secured or placed in manual mode? (1.5)

Answer:

If by at least two independent int' cations (0.5)

1) Misoperation in automat.; mode is confirmed (0.5) or

2) Adequate core cooling is assured (0.5)

Ref. F-EOP-1 pg. 21.

7.6

7.7 According .to E0P cautions (F-E0P-1) under what three (3) conditions would you as an SRO, be permitted to exceed the cooldown rate permitted by technical specifications? (1.5)

Answer:

1) Conserve RPV water inventory. ,

2)' Protect primary containment.

3) Limit radioactivity release to the environment.

(0.5 per answer).

Ref. F-EOP-1 pg.- 27.

7.7

rm- j 7.8 According to F-EOP-2 (RPV control) and F-EOP-4 (Primary Containment Control):

a) List the RPV Control (F-EOP-2) entry conditions (including setpoints). (1.5) b) List the entry conditions for Primary Containment Tontrol (F-EOP-4) including setpoints. (1.25)

Answer:

a) 1) - RPV water level below 177 in.

2) RPV pressure above 1045 psig.

3) .Drywell pressure above 2.7 psig.

4) When MSIY isolation is required.

5) Whenever a reactor scram is required and reactor power is above 2.5% or cannot be determined.

6) When directed by another emergency operating procedure.

(0.25 points per answer).

Ref. F-EOP-2.

Answer:

b) 1) Suppression pool water average temperature >95'F.

2) Drywell average temperature >135'F.

3) Drywell pressure >2.7 psig.

4) Suppression pool water level >0.0 inches.

5) Suppression pool water level <-1.5 inches.

(0.25 for each correct answer).

Ref. F-EOP-4 pg. 3.

7.8 ,.

.o 6

7.9 How is Boron injected upon the failure of SBLC valves to actuate? (0.5)

Answer:

CRD (0.5)

Ref. F-EOP-3 pg. 35 or F-A0P-37.

7.9

.l.

.g.

7.10 According to F-A0P-24 (Stuck Control Rod):

a) List the actions and any specific precautions / limitations this procedure directs you to do to move or free a stuck rod. (1.0) b)- If you are unsucessful in freeing the control rod what -

two (2) actions must be taken. (1.0)

Answer:

a) Increase drive water pressure.(0.25) If R xis <650 psig limit 1 drive water AP .to <600 psid.(0.25) Attempt to purge.

air- from drive (0.25) (can only be done above 20% power with RSCS out of service).

Vent drive piping.(0.25) b) Electrically disarm directional control valves (0.5) rearrange Control- Rod Pattern or shut reactor down.(0.5)

Ref. F-A0P-24 7.10 i

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

i i

7.11 According to F-A0P-31 (Loss of Condenser' Vacuum) what actions can be taken to minimize the rate at which vacuum is decreasing? (2.0)

Answer:

Trip the recombiner.

Reduce reactor power.

Place spare air ejectors inservice..

Shut the vacuum drag valve.

(0.5 for each correct answer).

Ref.'F-A00-31 pg. 3.

7.11

- _ . . -- .-~. -

7.12 In the event that a fuel assembly is dropped during fuel handling, list the actions you would take to mitigate the consequences of the dropped assembly. (2.0)

Answer:

Cease operation of the refueling equipment; refuel floor shall be immediately evacuated.

The Control Room shall be notified by the licensed operator.

The Control Room Operator shall sound the evacuation alarm and evacuate the Drywell and Reactor Building.

(plus any other reasonable actions)

Ref. F-EOP-31 pg. 2 or F-A0P-44 End of Section 7 7.12

8.0 Administration Procedure, Conditions, and Limitations (25.0) 8.1 According to Radiation Protection Procedures:

a) When are extended RWP's used and what is the maximum allowable time an extended RWP can be issued for? (1.5) b) Under what two (2) conditions would you not use an extended RWP? (0.5)

Answer:

a) For certain routine repetitive functions throughout the plant, (1.0) 1 year (0,5) b) Where neutron radiation (0.5) is present and where airborn radiation (0.5) signs are posted. (1.0)

(alsoacceptablg: unknown conditions, contaminated level

>500,000 DPM/cm ,10R/hr Beta-Gamma, maintenance in radiation area)

Ref. Radiation Protection Procedures pg.14 and 15, 29.

8.1

8.2 Define or explain the following term.

Secondary Containment Integrity (2.0 4

Answer: Reactor Building is in tact (0.5).

At least one door in each access opening is closed (0.5).

The Standby Gas Treatment System is operable (0.5).

All automatic ventilation system isolation valves are operable or secured in the isolated position (0.5).

Ref. Tech Specs pg. 5.

Y 8.2

m 8.3 According to Shift Relief and Log Keeping (0DS0-4):

0D50-4 states that each normally accessible pump will a) be checked for abnormal conditions. List four (4) of these abnormal pump conditions that are to be checked. (1.0) b)- The off going shift supervisor or assistant shift supervisor shall prepare Section "C" of the shift turnover checklist. What is included in Section "C" of the shift turnover checklist? (Include four (4) items that must be entered). (2.0)

Answer; a) Excessive packing leak off.

Excessive vibration.

Proper oil levels in bearings.

Proper suction and discharge pressure and flow.

Exc-issive bearing temperatures.

Lubrication.

-(0.25 for each correct answer up to four (4) correct answers).

b) Technical Specification items (1.0)

Brief description, action statement, technical specification no., date and time of expiration, actual time, date.

(0.25 for each correct up to four (4) correct)

Ref. 0050-4 pgs. 5, 6, and 15.

8.3

_.-____.c _

6 8.4 Which of the following events must be reported to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the event has been discovered? (2.5) a) Receipt of Special Nuclear Material.

b) Thef t of Source Material .

c) Changes in Security Plan made due to attempted theft of Special Nuclear Material.

d) Attempted theft of Special Nuclear Materials.

-)

e Unaccounted for SNM (special nuclear materials) exceeding applicable limits.

f) Report of radioactive contamination on package of radioactive materials.

g) Severe accident involving licensed materials.

h) Loss of licensed materials.

1) Licensed operator becomes disabled due to a auto injury.

j) Personnel exposure of a terminated employee.

Answer:

b, d, f, g, and h.

(0.5 for each correct answer).

Ref. Rules of Practice PS0 #6 pg. 21.

8.4

)

=

I

)

1 1

8.5 Fitzpatrick is operating at 90% power. You are to be the Shif t l Supervisor From 12:00 midnight - 8:00 a.m. shift. During shift change you are informed that during your shift the pressure suppression - Reactor building suppression chamber vacuum breaker instrumentation surveillence is to be performed. You are further told it has been five (5) menths since this surveillence test was last performed. The instrument technicians performing the surveillance tests inform you that one of the vacuum breaker's pressure switch is in-operable and the vacuum breaker pressure switch has a as-found setpoint of 1.0 psid. They also inform you that the in op pressure switen must be replaced and it will take two (2) weeks to get a new switch.

XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX X NOTE: USE THE ATTACHED SECTION OF THE TECHNIC'. SPECIFICATIONS X X TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE SECTIONS X X OF THE T.S. YOU USE TO DEVELOP YOUR ANSWER. X XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXYXXX a) What continued plant operation situation exists due to the above conditions? (1.5) b) What action must be taken as a result of this surveillence test? (1.5)

Answer:

a) You are in violation of Technical Specifications on the maximum surveillance interval. This is a reportable occurance. Section 4.7.4, pg.177. (1.5) b) You are above the setpoint permitted by Technical Specifica-tions on the 1.0 psig. Orderly shutdown shall be initiated and the Reactor shall be in cold shutdown within 24 hrs.

Section 3.7.8.a. pg.180a. (1.5)

Ref. Technical Specifications pgs.177. and 180a.

8.5

8.6 Define or, explain the following term.

Primary Containment Integrity (2.5)

Answer; a) Primary containment integrity means that the drywell and pressure suppression chamber are intact (0.5) and all of the following conditions are satisfied:

1. All manual containment isolation valves on lines connected to the Reactor Coolant System or containment which are not required to be open during plant accident conditions are closed. These valves may be opended to perform necessary operation activities. (0.5)

2. At least one dcor in each -irlock is closed and sealed. (0,5)

3. All automatic containment isolation valves are operable or de-activated in the isolated position. (0.5)

4. All blind flanges and manways are closed. (0.5)

Ref. Technical Specifications 4, 5 8.6

. l 8.7 Concerning plant staffing requirements.

a) What is the minimum operations shift crew composition during full power operation, and during refuel or cold shutdown? (1.0) b) Does a shift technical advisor have to be in the control room at all times? (Explain) . (1.5)

Answer:

a) Full Power Cold Shutdown / refuel 2 SR0 (licensed) 1 SRO (licensed) 2 SRO (licensed) 1 R0 (licensed) 2 0 (unlicensed) 1 0 (unlicensed)

STA (0.5 for full power correct, 0.5 for cold shutdown correct) b) No. (0.5) He is required to be on site and readily available to the control room except during the cold shutdown or refuel mode. (1.0)

Ref. Technical Specifications 247, 247a, 0D50-1.

8.7

8.8 In accordance with Technical Specifications the reactor was scrammed due to suppression temperature >110*F.

The reactor is now in hot shutdown suppression pool cooling is on and the suppression pool water temperature = 98'F. (2.25)

Can you place the reactor mode switch into startup? Explain your answer fully.

Answer:

No, (0.5) you must be below 95'F in suppression pool (0.75).

Technical Specifications prohibits you from entering into an operational condition while relying on an action statement. (1.0)

Ref. Technical Specifications 3.0.D pg. 30a; 3.7A pg.166.

I 8.8 t

4 8.9 Fitzpatrick is in the process of starting up,- (mode switch in startup), with all systems and components normal except that "A" IRM is inoperable and bypassed. The "C" IRM now loses power and declared inoperable. Can Fitzpatrick continue startup in this condition? (Explain). (1.25)

Answer:

a) .Yes (0.5). Place the RPS A channel in the tripped position within one (1) hour. (0.75)

Ref. Technical Specifications Table 3.1-1 pg. 42, 3.0.C pg. 30.

l 8.9

1 8.10 The technical specification limit on reactor coolant chloride concentration is different at low steaming rates (less than 100,000 lb/hr) than at high steaming rates (greater than 100,000 lb/hr).

a) Which condition (low steaming or high steaming) permits the highest reactor coolant chloride concentration? (0.5) b) Explain the basis for your answer in part a. (1.0)

Answer; a) High steaming rate (0,5) b) At low steaming rates a more restrictive limit is established to assure chloride-oxygen contamination are kept within limits at steaming rates above 100,000 lb/hr boiling causes deoeration of the reactor water thus maintaining oxygen concentration at low levels. (1.0)

Ref. Technical Specification 3.6C pg.140, Bases pg.149.

Figure 4.6.1 pg. 164 i

G

~

8.10

\

s:

8.11 Define or explain the following terns:

a) LCO (0.5) b) Core alteration (0.5) c) Startup/ Hot Standby (0.5) d) operable (0.5)

Answer; a) The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility.

(0.5) b) The act of moving any component in the region above the core support plate, below the upper grid and within the shroud.

(0.5) c) The reactor mode switch is in the Startup or Hot Standby mode position, the reactor coolant is greater than 212 degrees F and the reactor pressure is less than 1005 psig (0.5) d) System operable when it is capable of performing its intended function.

(0.5)

Ref. Technical Specifications Pg. 3,1, 2, 4.

End of Section 8 End of Test I

8.11 lL