ML20215N033

Exam Rept 50-321/OL-86-02 on 860903-05.Exam Results:All Four Reactor Operators Passed.Related Info Encl

ML20215N033
Person / Time
Site:	Hatch
Issue date:	10/27/1986
From:	Brockman K, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20215N022	List: ML20215N021 ML20215N033
References
50-321-OL-86-02, 50-321-OL-86-2, NUDOCS 8611040192
Download: ML20215N033 (69)

Text

o - 1 ENCLOSURE 1 EXAMINATION REPORT 321/0L-86-02 Facility Licensee: Georgia Power Company Facility Name: Edwin I. Hatch Facility Docket No.: 50-321 Written and operating examinations were administered at the Edwin I. Hatch Nuclear Plant.near Baxley, Georgia.

Chief Examiner: Ide "

$dn E. Brockma'n, Rector Engineer

/* /17 /7 4 Date Signed Approved by: '/ d e # /c/22/u John F. Munro,4cting Section Chief Date Signed Summary:

Examinations on September 3-5, 1986 Written and oral examinations were administered to 3 candidates, all of whom passed; _ simulator examinations were administered to 4 candidates, all of whom passed.

Based on the results described above 4 of 4 R0s passed.

8611040192 861027 PDR ADOCK 05000321 V PDR

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REPORT DETAILS

1. Facility Employees Contacted:

• T. V. Greene Deputy General Manager-

• C. T. Moore.. Training Manager

• R. S.' Grantham,.0perations Training Supervisor S. Hans, Simulator Instructor B. Smith, Simulator Instructor

• Attended Exit Meeting

-2. Examiners:

• K. Brockman, RII

'J. Hanek, EG&G M. Bishop, EG&G

• Chief Examiner i 3. Examination Review Meeting

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At the1 conclusion of the written examination, the examiners provided ,

. R. S._ Grantham, with a copy of the . written examination and answer key -

for review. 'The comments made by the facility reviewers-'are included as Enclosure 3 to this report, and the NRC Resolutions to.these comments are-listed below.

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a. R0 Exam (1) Question 1.14

.NRC Resolution:

Facility comment is valid.

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Answer key 1.14e will be changed'to either 2-3 or 2-3' .

-(2) Question 2.16 NRC Resolution:

-Agreed. The -question does not elicit specific knowledge, but, instead, allows the candidate to address almost any related topic for full credit. Question 2.16 and corresponding answer will txt deleted._ Category 2 point value will be reduced by 1.00 point and the overall total reduced by 1- 00 point.

2 (3) Question 2.17 NRC Resolution:

Comment acknowledged. This question is consistent with the requirements of ES-202; however, the primary responsibility of the operator is to confirm that the automatic actions occur. The high point value assigned to this question is inappropriate with-regards to the knowledge level required. Question point value will be reduced to 1.50 points (6 at 0.25 each) with a corre-sponding reduction in Category 2 and overall total point values.

(4) Question 3,03 NRC Resolution:

Comment acknowledged. Previously supplied reference material was not accurate. Updated reference material supplied by the facility describes -the -current operation of the RWM as revised by Process Computer Software Change 85-2, which was approved for'both units.

Answer key 3.03b will be revised per the utility recommendation.

Question point value will be reduced to 1.50 with a corresponding reduction in category 3 and overall total point values.

(5) Question 3.08 NRC Resolution:

Facility comment is valid. Because it is technically inaccurate, the question and corresponding answers will be deleted. Category 3 point value will be reduced by 2.00 points with a corresponding decrease in overall total point value.

(6) Question 3.13 NRC Resolution:

Comment acknowledged. 34AB-0PS-008-2 does supplement the panel identifications with a written description. In . the case of components #2, "CRD Pump .1A", differentiating between panels H21-P176 and 1H21-P175 (two different 4160V buses) is considered beyond the scope of R0 knowledge. However, components #1 and #3, "RCIC Pump Discharge (F013) and "RCIC Trip and Throttle Valve" are on the Remote Shutdown Panel #1 and the identifier C82-P001 is adequate. Question 3.13.b.2 and the corresponding _ answer will be deleted. Category 3 points value will be reduced by 0.25 points with a corresponding decrease in overall . total point value.

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f-3 (7) ' Question 4.07 NRC Resolution:

Comment not accepted. While formally declaring an alarm to be a

" nuisance" is an SRO responsibility, the R0 must be knowledgeable of when and why alarms can be declared a nuisance and how they impact him as an operator. The questions is relevant and appro-priate.

.(8) Question 4.08 NRC Resolution Facility comment is valid. The question is specific with regards to the federal radiation limits. (assume NRC-4 completed) The answer key reference of 1250 MREM is in error and will be corrected to "3000 MREM, not to exceed 5 (n-18) accumulated lifetime dose."

Neither 1250 MREM nor 1000 MREM will be accepted as a correct answer.

4. Exit Meeting At the conclusion of the site visit the examiners met with representa-tives of the plant staff to discuss the conduct of the examination.

There were no generic weaknesses noted during the oral examination.

The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted and appreciated.

The licensee did not identify as proprietary any of the material provided to or. reviewed by the examiners.

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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _ HATCH _1gg___ _________

REACTOR TYPE: _gWR-@E8_________________

DATE. ADMINISTERED: __8_6_/_09_/0_3________________

ISTRCRU EXAMINER: _@PENgER 2 _M. _ _

CANDIDATE: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

INgIBygIJgNg_Ig_g@Np19@IE1 Ue separate'- paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at lacst 80%. Examination papers will be picked up six (6) hours after the examination starts.

% OF CATEGORY  % OF CANDIDATE'S CATEGORY

__YBLUE_ _Igl@L ___gggBE___ _y@LUE__ ______________g8IEGQRY _______ _______ _

29,9 9

_g33@@__ _gI.Zg ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID-FLOW 2.L so 24 41 i _E$sgg__ _222sg ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 22.so 23.37 l

_EE1;g__ _;j;ig ___________ ________ 3. INSTRUMENTS AND CONTROLS l 17. 1 7 l _2hz2E__ _EEs2I- ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL,

! EMERGENCY AND RADIOLOGICAL L

CONTROL

,  %.25 l

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_1_0_1_,2_5__

Totals

! Final Grade All work _done on this examination is my own. I haverneither given j nor received aid.

Candidate's Signature

h NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examle:ation the f ollowing rules apply:

1. Cheating on-the examination means an. automatic denial of your application and could result in more severe penalties.

2. Restroom trips are to be limited and only one candidate at a time may leave. -You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

3 .. Use black ink or dark pencil gnly to facilitate legible reproductions.

4. Print your name in the blank provided on the cover sheet of the examination.

5. Fill in the date on the cover sheet of the examination (if necessary).

6. Use only the paper provided for answers.

7. Print'your nam.e in the upper right-hand corner of the first page of each tection of.the answer sheet.

8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne side of the paper, and write "Last Page" on the last. answer sheet.

9. Number each answer as to category and number, for example, 1.4, 6.3.

10. Skip at least three lin'es between each answer.

11. Separate answer sheets from pad and place finished answer sheets face down on your-desk or table.

12. .Use abbteviations only if they are commonly used in facility litetatute.

13. The point value for each question-is indicated in parentheses after the question and can~be used as a guide for the depth of answer required.

14. Show -all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.

15. Partial credit may be given. Therefore, ANSWER ALL' PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

16.- If parts of the examination are not clear as to intent, ask questions of the e:1aminer only.

l 17. You must. sign the statement on the cover sheet that indicates that the l

work is your own'and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.

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10. When you complete your examination, you shall

.a.. Assemble your examination as follows:.

i_ (1) Exam questions on top.

(2)' Exam aids - f.igures, tables, etc.

(3) Answer pages including figures which are part of the answer.

b. Turn in your copy of the examination and all pages used to answer the examination questions.

c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.

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d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.

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'i- fBINCIPLgg_g8_NUgLgAR_ POWER _ PLANT _gPERATIgNz' PAGE 2 IME$dQDyN@diggg_bg@I_JB@NSEEB_@ND_ FLUID _FLgW QUESTION 1.01 (1.00)

.R2 activity is defined'as wh'ich of the following?

a. The ratio _of the number of neutrons at some point in this generati~on to the. number _of neutrons at the same point in the p'revious. generation.

b. The fractional change in neutron population per generation.

c. LThe factor by which neutron population changes per genera-tion.

d. The rate of change of reactor power in neutrons per second~.

QUESTION 1.02 (2.00)

.a. STATE the design feature in the reactor-vessel which ensures proper flow distribution through the core fuel bund 1'es. (0.5)

.b. ' EXPLAIN how the distribution would react during.a power increase by rod pull if this design feature were.NOT PRESENT.

INCLUDE IN YOUR RESPONSE THE REASON (S) FOR THIS REACTION. (1.5) l

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'l __EBINQlE(E@_QE_NQC6E@B_EQMEg_C(@NI_QEEB@Il@Nt PAGE 3 ISEBdQQIN@dlC@t_ME@I_IB@N@EEg_@NQ_E6Qlp_E6QM QUESTION 1.03 (1.00)

Using the boiling heat transfer curve below, the area identified as (1) represnts which area? Chose from a, b, c, or d.

a. natural convection

b. partial film boiling

c. film boiling

d. nucleate boiling I

108 - .

DN --- __

~ 104 -

t T=

4 m

E 10! -

E 8

=

1(f -

i i . . .

10 100 1000~ 10,000 I

Temperature DiMerence (T, - T,)(*F) l I

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P 81Ng]E6gg_gE_NUCLEgg_EgWg8_E69NI_OPERATIgN 1 PAGE 4 IBEBd9DYNedigS 1_dggI_IggNgEgB_OND_ FLUID _FLgW QUESTION 1.04 (1.00)

At end of core life, which of the below modes of heat transfer best dcscribes the transfer of heat from the fuel to the cladding?

a. conduction

b. radiation

c. convection QUESTION 1.05 (1.00)

The reactor is operating and the average bundle power is 10 kw/ft. The critical power for the reactor is 20 kw/ft. What is the critical power ratio?

QUESTION 1.06 (1.00)

Which of the below best expresses the total peaking factor.

a. TPF = APF + RPF + LPF

b. TPF = APF + RPF LPF

c. TPF = APF X RFP X LPF l d. TPF =

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l APF + RFP + LFF l

QUESTION 1.07 (1.00) l Following a scram with some of the rods stuck out, the plant is cooling i down at 100 F/ hour and K eff = 0.99. Considering only the effects of the cooldown, WHEN will the reactor THEORETICALLY go RECRITICAL?

(Assure all assumptions are listed with your answer!)

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'1:__FBJUCIPLEg_gF_ NUCLEAR _PQWER_ PLANT _gPERATIgN1 PAGE 5 IbsBM9DXN@MICg2_dE@I_IB@NgEER AND_ FLUID ___FLgW QUESTION 1.08 (2.00)

State how each of the below conditions will effect the void coefficient.

(Limit the answer to INCREASE, DECREASE, or REMAIN.THE SAME.)

'a. moderator temperature increases

b. core void fraction increases

c. fuel temperature increases

d. core age increases QUESTION 1.09 (1.00)

The value of Beta (eff) changes'over core life because: (choose one)

a. As lambda (eff) decreases, beta must also decrease.

b. There is an increased percentage of fissioning from Pu-239, which has a smaller delayed neutron fraction than U-238 and.U-235.

c. As U-235 is fissioned and effectively used up, there are fewer fast fissions which result in fewer delayed neutrons.

d. The microscopic cross sections for the isotopes producing delayed neutrons progressively become smaller with core age.

QUESTION 1.10 (1.50)

State how each of the below listed conditions will effect. control rod worth? (Limit the answer to INCREASE, DECREASE, or REMAINS THE SAME.)

a. increasing moderator temperature

b. increasing the percent voids

c. increasing the fuel temperature

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' 1. PRINCIPLES OF NUCLEAR POWER PLANT' OPERATIONt PAGE 6 ISEBdggyN@dlCgt_ME@I_IB@NgEE6_@ND FLuig_ELgW QUESTION 1.11 (3.00)

During a startup, the reactor is critical at 3.0 x 10E3 counts.

A control rod is notched out resulting in a do'ubling time of 85 sec.

a. What is the reactor period?

b. If the heating power of the unit is 8.8 X 10E5 counts, how long will it take to reach this power level?

QUESTION 1.12 (1.00)

Briefly define speed droop with respect to a Diesel Generator.

QUESTION 1.13 (1.00)

Ucing the attached Figure One, Identify in which two. regions generator operation is prevented because of Generator Interlocks and/or Trips.

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'le.__EBINglELgg_gE_NUCLg@g_Pgyg8_ PLANT _gEg8@Ilgh PAGE 7

. IHgBMODYNAMICg1_HgAT_TRANgFgR_ANp_FLUI_p_FLgW OUESTION 1.14 (2.50)

Utilizing the Rankine Steam Cycle diagram below, identify the line segment which represents each of the processes listed a. through e. (e.g.

1-2).

a. isentro'pic steam expansion (ideal turbine)

'b. isentropic compression (ideal pump) .

c. constant pressure heat reamoval (condenser)

d. constant pressure heat addition (boiler)

e. isentropic steam expansion (non-ideal turbine) o CRITICAL POINT P = 1015 p sia

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1. % I 7 \

P 1 1 r\ / CONDENSER W / = 28ln. Hg VACUUM

'p

1. q PUMP WORK IN \

3, , (3' -

s*/ = 2in. Hg ABSOLUTE h ,/ A CL i 2 l

N I

l ENTROPY (Btu /lb *R)

QUESTION 1.15 (1.00)

Why will U-238 not fission by absorption of a thermal neutron?

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'1. PRINCIPLE @_QF NUCLEAR _PgWgR_ PLANT _QPERATION t PAGE 8

_ IggR$ggyN@dIggt_Ug61_IRANSFER AND__ FLUID FLOW QUESTION 1.16 ( 1 '. 00 )

Concerning control rod worths during a reactor startup from 100% PEAK XENON versus a startup under XENON-FREE conditions, which statement is correct?

a. BOTH central and peripheral control rod worth will be LOWER regardless of core XENON conditions. _

b. CENTRAL control rod worth will be HIGHER during the-PEAK XENON startup than during the XENON-FREE startup.

c. BOTH central and peripheral control rod worth will be the SAME regardless of core Xenon conditions.

d. PERIPHERAL control rod worth will be HIGHER during the PEAK XENON startup than during the XENON-FREE startup.

QUESTION 1.17 (2.00)

A significant amount of excess reactivity must be loaded into a core at BOL. For each-of the following, LIST the approximate value of K-excess which must be. loaded to'avercome that negative reactivity component at rated-equilibrium conditions.

a. Moderator temp _ increase

b. Void fraction increase

c. Samarium buildup l

d. Xenon buildup l

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e d25 PLANI _DgglgN_JyCLUDING_ SAFETY _AND EMERGENCY 1 _SYSTEMg PAGE' 9 QUESTION 2.01- (1.00)

The Full-Core Display on the center panel has a BLUE scram light for each control rod. DESCRIBE what is DIRECTLY indicated when this light is-illuminated.

-QUESTION 2.22 ( '. 50 ) -

'Which one of the following is.the initiation signal for the diesel driven fire pumps? (1.0)

1. . Low Fire Water Header. Pressure

2. Low Air Pressure (Dry Pipe Systems)

'3. High. Fire Water Header Flow Rate

4. Smoke Detector Actuation QUESTION 2.03 (2.00)

TRUE or FALSE Answer each of the following as they describe Low-Low Set (LLS) logic,las applied at-Plant Hatch?

a. -Lowers LOTH the opening and closing setpoints of the LLS valves.

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.b. Cont'rols'the operation-of all relief-valves, excepting E and H.

c. Is activated by a position switch which confirms any SRV opening.

d. Is applicable to Unit 2 ONLY.

i QUESTION 2.04 (1.50)

L j a.- How is the-integrity of the Core Spray piping between the i RPV and the Core Shroud monitored? (1.0)

b. How would the indication (s) change if the integrity was LOST. (0.5) l

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• 2 L__[L@NT_DEgIgN_INCLUDIN@_gAFETY_AND_ EMERGENCY gY@TEMg PAGE- 10 QUESTION 2.05 (2.00)

List four-of the five conditions required to assure Primary Containment integrity (Other than drywell and pressue suppression chamber Intact.)

QUESTION 2.06 (2.00)

Answer the below questions about the RCIC and support systems.

a. The barometric condenser vacuum pump discharges to what system component?

~b. What is the power supply to the barometric condenser condensate pump? (Bus number / Identification for BOTH UNITS)

c. The flow element for the governor valve is between what components in the discharge line.of the RCIC pump? { choose from the below selections)

1. Injection Valve MO-13 and Pump Discharge Valve _MD-12

2. Pump Discharge Valve MD-12 and Discharge Check Valve #14

3. Discharge Check Valve #14 and the RCIC Pump Discharge

4. Injection Valve MD-13 and feedwater line "B" (Unit 2)

d. The lube oil cooling water returns to where?

OUEST;ON 2.07 (1.50)

The Unit 2 CRD Accumulator charging line has five restricting orifices.

a. What is the purpose (s) of these orfices? (0.5) i b. How is this orificing provided for in Unit 1? (1.0) l l

l QUESTION 2.09 ( 2 '. 00 )

l What are the maximum and minimum injection times for Standby Liquid l

Control System AND what are the bases for these times?

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?i__PLgNI_pgglgN_INCLUp1NG_ggEgIy_gNp_gdggggNCY_gY@Igdg PAGE 11 QUESTION 2.09 .(1.00)

Once the Standby Liquid Control System has been initiated, the entire contents of the tank must be injected. WHAT is the basis for this esquirement?

' QUESTION 2.10 (1.50)

Briefly describe how a FISSION CHAMBER is capable of detecting IONIZING RADIATION.

' QUESTION 2.11 (1.00)

List four of the design methods used to reduce.the senstivity of'the IRM's as compaired to the SRM's.

QUESTION 2.12 (1.00)

The main steam line restrictors in conjunction with the MSIV'S are designed to prevent what event and during what accident conditions?

-QUESTION 2.13 (1.00)

What are the bases for the minimum and maximum closing: time of the Main Steam Isolation Valves? Limit your answer to the nuclear fuel

. considerations.

QUESTION 2.14 (1.00)

The Unit i Startup Level Control Valve (F-113) bypasses which of the below listed components? (Choose the most inclusive answer!)

a. RFP "A"

b. RFP "B" -

c. Both RFP's "A" & "B"

d. Both RFP's "A" & "B" and the 7A & 7B heaters l

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'2___p te NI_ pg@lg N_] ngl UplNg_ gg Fgly_gNp_gdg RggNgy_gypIgdg PAGE 12 QUESTION 2.15 (1.00)

The motor cooling water supply for the RHR Service water pumps is supplied by RHR Service Water. The tap off of this supply is ...

(Choose one)

_a. between the pump and discharge isolation valve (F012).

b. between the discharge. isolation valve.(F012) and the stainers.

c. between the strainers and the reactor building.

d. Inside of the reactor building.

OUEST!OM 2.11 (1.SS) 3E' - _ _

What plant desion fna+e c; miivw the use of common electrical buses for r-- -

r a e !_ RC I compenentc?

\.5D -

QUESTION 2.17 C.00)

List in the order of occurance, starting with the quickest after the initiating event, the below list of ECCS operating events.

The initiating event is a design basis loss-of-coolant accident and normal auxiliary power is lost.

a. CS pumps at rated flow and CS injection valves full open which completes CS startup.

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b. Reactor low-low-low- water level reached. Second signal to start LPCI and CS: Auto-depressurization sequence begins.

c. LPCI pump A, B, and D at rated flow and LPCI injection valves wide open, completing LPCI startup.

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( d. HPCI injection valve open and pump at design flow.

e. LPCI pump C at-rated flow and LPCI injection valves wide open j f. Diesel generators provide emergency power i

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21__Eh@NI_DgglgN__IblCLUDING_@AFETY AND_ EMERGENCY _@Y@TEME PAGE 13 QUESTION 2.18 . (1.00)

WHY is-the Rod Worth Miniminer not required at high (greater than '30%)

power?

QUESTION 2.19 (1.00)

What is the basis for a higher chloride limit on reactor coolant during'

'high power-operation?

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~'21__lNgIgud[NIg_AND_CQNTROLS PAGE 14 QUESTION 3.01 (2.00)

List the four ranges of. reactor vessel level indication provided in the control room:AND the value in inches available from each~ range.

OUESTION 3.02 (2.00)

The SPDS Primary Display provides information concerning the ADS, among other parameters,

e. - DESCRIBE how the ADS Valves are differentiated from the other SRV's. (0.5)

b. STATE the significance of the following COLORS, as they are used for indication of ADS Valve status: (1.0)

1. -Green

'2. Orange

c. LIST the two~(2)~ time periods.over which TREND Displays are-given. (0.5)

OUESTION 3.03 (1.75)

Concerning thelRod Worth Minimi:er (RWM) .on Unit 2:

a. When a select error occurs on the.RWM, STATE whether.the operator can still move the rod. (YES or NO) ASSUME THE RWM IS NOTJBYPASSED

, and NO ROD BLOCKS ~ EXIST PRIOR TO SELECTING THE ROD. (0.5) 4

b. EXPLAIN the bases for ycur decision in part (a). Consider in your explanation both an attempted insert and withdraw action.. ( 1.- 25 )

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g. f1NgTRUMENT@_AND_ CONTROL @

PAGE 15 QUESTION 3.04 (1.00)

Given: Unit 2 in control of D/G "B" D/G "B" Mode Switch in TEST (Surveillancc being performed)

Electrical distribution NORMAL (Full Power Lineup)

D/G "B" is at rated speed and voltage, but not synchronized, when power is lost to 4160. volt Bus 2F. Which of the following accurately dsscribes the system operation?

a. Bus 2F can be powered by D/G "B" when the operator takes the Output Breaker Switch to CLOSE and has the SYNC SCOPE activated.

b. Bus 2F will be powered by D/G "B" automatically, after 12 seconds; appropriate loads will be picked up sequentially.

c. Bus 2F can not be powered by D/G "B" while it is in the TEST mode, given these conditions.

d. Bus 2F can be powered by D/G "B" when the operator resets tNe Lockout Relay, activates the SYNC SCOPE, and takes the Output Breaker to CLOSE.

QUESTION 3.05 (1.00)

Briefly describe consequences of closing the SUT breaker while the UAT breaker is closed and the " Cutout Switch" (panel _651) is not in the Cutout position. (Include final position of both breakers ar.d the status of the electrical bus.)

QUESTION 3.06 (2.00)

With regard to the Off-gas Radiation Monitoring System:

a. LIST the three (3) combinations of radiation instrument trip signals causing an_Off-gas System auto-isolation. (1.5)

b. LIST the Off gas System valve (s) which CLOSE on an auto-isolation. (0.5)

QUESTION 3.07 (1.50)

The Main Turbine first stage pressure switches provide permissives and/or control signals for several plant functions. LIST three (3) of these permissives/ control functions.

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91__lygIRUdEMI@_AND_ CONTROL @ PAGE 16

-OUESTIOM T.99 '2.99?

For the following situations, select the correct Feedwater Conte System / plant response from the list (a through e).

Azsume NO operator actions are taken.

1. The plant is operating at 100% power, in dP c rol, when one Feed-water Flow Detector fails, to indicate Zer low (i . e. , Fails Down).

2. The plant is operating at 100% pow in dp control, when one Steam-Flow Detector fails, to indicat ero Flow (i . e. , Fails Down).

a. Reactor water level d eases and stabilizes at a lower level.

b. Reactor water el decreases and initiates a reactor scram.

c. Peactor ater level increases and stabilices at a higher level.

d. actor water level increases and initiates a turbine trip.

c. Non: cf tFr 252/ _

QUESTION 3.09 ( .50)

True er Falso Manual isolation of the HPCI system by the pushbutton on panel P603 can only be accomplished if an auto initiation signal is present.

QUESTION 3.10 (2.00)

List the four automatic isolation signals and their respective j setpoints for the HPCI System.

l QUESTION 3.11 (1.75)

a. What is~the purpose (s) of the 13 minute timer in the Unit 2 ADS system?

b. Resetting the 13 minute timer once the 2 minute timer has timed out will have what effect on the ADS system operation?

l c. Failure of the RHR pumps during Automatic initiation of the ADS system will produce what effect? (concern your answer to the ADS system only) l 1

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

q __ldjl6MMEyl@_ANQ_ggyIB06@ PAGE 17 QUESTION 3.12 (1.50)

Creifly discuss the reason (s) for the LEVEL B transmitter being the normal colected device for reactor vessel level ~ control.

\ .15 QUESTION 3.13 '1.59'

a. The Unit 1 Remote Shutdown Panel (s) has the capability of operating:

(0.75)

1. Which RHR pump (s)?

2. Which SRV's with ADS capability? ,

3. Which CRD pump (s)?

b. From which of the Unit 1 Panels (lef t column) could the following information (right column) be found? .(0. 75)

PANEL NUMBER CHOICE INFORMATION CHOICE 1

c. C82-P001 1. RCIC PUMP DISCHARGE (F013)

b. CO2-P002 42. CnD PUMP 1A l>E'E'NED

c. H21-P176 3. - RCIC TRIP AND THROTTLE VALVE

d. 1H21-P175 QUESTION 3.14 (1.50)

List- the ef f ect a . lass of 120VAC to ATTS cabinets (panel-2H11-P609 or 2H11-P611) will have on the below systems,

a. Secondary Containment.

b.- PCIS Group II and V

c. PCIS Group I OUESTION 3.15 (1.00)

List the two MSIV-LCS interlocks required for the system to be operable.

Consider only the Outboard Interlocks.

(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)

9___INgIBUMgNIg_@Ng_CgdIBgL@ PAGE 18 t

t QUESTION 3.16- (2.00)

Consering the Main Turbine Generator:

a. - List the initiating signals and'setpoints-which' actuate the " GEN'ERATOR PROTECTION CIRCUIT ENERGIZED" ,

b. If a. runback has commenced, the turbine will_ trip if what current limitations are not met? (include time limits) s e

I i

k l

s f

l-I l

t I

l (***** END OF CATEGORY 03 *****)

l t

I

I1__QBOggDU65@_;_ NORMAL t_ ABNORMAL t EMERGENCY _AND PAGE 19 R_ A_ D__I O_L_O_G_ _I C_ A_L_ _C_O_N_ T R_O_L_

OUESTION 4.01 (2.00) 34AB-OPS-009-2, " Inability to Shutdown With Control Rods",

ctates :

... if at any time, either condition ____(a.1) _ or____ (a.2) ____

exists,and either __ (b.1) .,___ or ____ ( b.2) ____, and if it is obvious that the reactor cannot be shutdown and, in.the judgment of the Shift Supervisor, or in his absence, a licensed operator, a hazard exists to the environs, personnel, or the plant, utilize the standby liquid control system per 34SO-C41-001-2."

a. LIST conditions (a.1) and (a.2).

b. LIST conditions.(b.1) and (b.2).

QUESTION 4.02 (1.00)

~34AB-OPS-002-2, " Pipe Break Inside Primary Containment", lists numerous conditions indicative of a break. Which one of the following supports the suspicion of a SMALL break inside con-

.tainment.

a. Decrease in reactor water level; pressure and/or temperature in-crease in Drywell; airborne activity increase in Drywell; increased DWFDS operating frequency.

b. Decrease in reactor pressure; pressure and/or temperature increase in Drywell; generator load decrease; DWFDS high level.

c. Reactor Scram from low water level; pressure and/or temperature increase in Drywell; generator load decrease; increased DWFDS operating frequency.

d. Reactor Scram from high Drywel.1 pressure; generator load decrease; airborne activity increase in Drywell; DWFDS high level.

QUESTION 4.03 (2.00)

A pipe break occurs and results in a decreasing reactor water level and decreasing reactor pressure. LIST the four (4) criteria which would require the MANUAL initiation of ADS, given auto-initiation has not occured.

(***** CATEGORY 04 CONfINUED ON NEXT PAGE *****)

3:__PBggggyBEg_;_NgBM861 ABNORMAL2 _EMERggNgY_ANg . PAGE '23 B8DI96991986_99NIBg6 QUESTION 4.04 (1.00) 34GO-SUV-001-OS, " Control and Surveillance of Locked Valves",

provides guideance concerning the proper way to confirm and v rify a locked valve position. DESCRIBE the proper way to CONFIRM Locked Valve. position and Locking Device integrity.

QUESTION 4.05 (2.00)

STATE which Emergency Classification is appropriate for the following d=finitions.

a. Events are.in progress'or have occurred which involve actual or potential substantial degradation of the level of. safety of the plant.

b. Events are in progress or have occurred which indicate a potential degradation of the_ level of safety of the plant,

c. Events are in progress or have occurred which involve actual or imminent s.ubstantial core degradation or melting with the potential for loss of containment integrity.

d. Events are in progress or have occurred which involve

~

an actual or likely major failure of plant functions needed for protection of the public.

I

~

QUESTION 4.06 (1.00)

A plant startup is in progress and condenser vacuum is being established in~accordance with 34GO-OPS-001-2, " Plant Startup".

Which one of the following is the proper sequence for component /

subsystem startups.

a. Steam Packing Exhaust ~er, Steam Seal Header, Mechanical Vacuum Pump, Steam Jet Air Ejector,

b. Steam Seal Header, Steam Packing Exhauster, Mechanical Vacuum Pump, Steam Jet Air Ejector.

c. Mechanical Vacuum Pump, Steam Packing'Exhauster, Steam Seal Header, Steam Jet Air Ejector.

d. Steam Packing Exhauster, Mechanical Vacuum Pump, Steam Seal Header, Steam Jet Air Ejector.

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

E_;==PROCgDUggg_;_NgBM862_@pNgBM@Lx_gMgBGENCY_8ND PAGE 21 B8D196991986_99NIBg6 QUESTION 4.07 (1.00)

List the criteria required for an annunciator alarm to be' determined a

" Nuisance Alarm". Cref. 31GO-OPS-008-OS]

QUESTION 4.08 (1.50)

a. List your quarterly exposure limit to whole body radiation.

(assume NRC -4 completed) b.~ List your quarterly. exposure limit to the extremities to radiation.

c. List your quarterly exposure limit to the skin of the whole body to radiation.

QUESTION 4.09- (2.00)

List four pieces of information which must be entered onto a RED TAG

'in order for it to be properly hung. (Information which an operator must-write on the tag.)

QUESTION 4.10 (3.50)

List seven of the eight immediate operator actions steps for a reactor scram with the MSIV's OPEN in accordance with 34AR-OPS-001.

An acti'on: step may consist of more than one action item.

QUESTION 4.11 (1.25)

-a. List the minimum SRM core monitoring requirements per the Technical Specification during normal core alterations. (Include the-number of SRM's required and any location requirements.) (0.75) b.. What is the' minimum allowable indicated value on SRM'S during normal core alterations? (0.5)

OUESTION 4.12 -( 1. 00 )

List the automatic actions which occur in the feedwater and recirc systems i

due to a loss of 120/240 valt vital AC distribution cabinet 2A.

(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)

3;__PRgCEDURES; _ NORMAL _ABNgRMAL 1 _ EMERGENCY 2 __AND PAGE 22

_R@gIgLgGICAL_CgNTRgL QUESTION 4.13 (2.00)

Concerning.the Scram Pilot Valve' Air Header

e. Per 34AB-OPS-020-2, at what value of scram pilot valve header air

-presure is a manual scram required?

b. Why is operator. action necessary when this minimum air pressure

.is reached?-

QUESTION 4.14 (2.00)

List the four indications for a loss of RPIS.

QUESTION 4.15 (2.00) 42FH-ENG-010-2, " Control Rod Movement", provides numerous STANDARD PRACTICES which apply when rods are being moved for the purpose of changing power level,

a. With LHGR > 8 Kw/ft, STATE the period of time ycu must wait between successive notch withdrawals of the same rod. (0.5)

b. With HIGH POWER and HIGH FLOW conditions, STATE which type of rod - SHALLOW, DEEP, CENTRAL, or EDGE - should NOT be with-drawn nor inserted. (1.0)

c. With power < 30%, STATE the verification which is utilized when latching the first rod-in any group. (0.5)

OUESTION 4.16 (1.00) l 34SO-B31-002-2, " Reactor Recirculation Pump Speed Changes",

l~ cautions the. operator not to start an idle pump unless the other pump is operating at less than 50% loop speed.

~

l EXPLAIN the reason for this procedural CAUTION.

l l

r I

1

(***** END OF CATEGORY 04 *****)

l i

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

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EQUATION SHEET f=ma v = s/t 2 Cycle efficiency = " " '

w = ms s = v,t + hac E = mc -

a = (vf - v )/t KE = my 2 v =v + at A = AN A = A,e

-At f o PE = mgh w - e/c A = in 2/tg = 0.693/tg W = v4P q(eff) = (tu)(es )

t AE = 931Am .

( , )

Q=$ CAT I - I .o EX ~

, P ,

Q = UAAT I - I

~

UX

'Pur = W g i' I-I to -X/ M g

y.p toSM(t). TVL = 1.3/u y-p .t/T -

HVL & 0.693/u o

'SUR = 26.06/T ~

T = 1.44 DT SCR = S/(1 - K,gg)

SUR = 26 fA{c) g CR x = S/(1 - K,ggx)

I ~

eff)1

• II ~ Kaff)2 T = '(t*/o ) + [(f

• p)/A g,og I 2 T = 1*/ (o - D

, M " IIII ~ Eeff) = CR /CR0 g T = (I - p)/ A,gg o g , Cg _ g eff)0 /(1 - K,gg)g p = (K,gg-1)/K,gg = R ,ff/Keff SDM = (1 - K,gg)/K,gg .

~

p= ( L*/TK,'gg -) + [I/(1 + A,gfT )] 1* = 1 x 10 seconds P = E$V/(3 x 1010) x aff = A 0.1 seconds I = No Idiy=Id22 4 WATER PARAMETERS Id g =Id2 1 gal. = 8.345 lbm. R/hr = (0.5 CE)/d (meters)

I gal. = 3.78 liters R/hr = 6 CE/d (feet) -

1 ft = 7.'48 gal. MISCELLANEOUS CONVERSIONS .

3 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 10 Density = 1 gm/cm i kg = 2.21 lbm Heat of vagori:stion = 970 Ecu/lbm 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lba 6 1 Mw = 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in. I'g. 1 Btu = 778 f t-lbf 1 ft. H 2O = 0.4333 lbf/in 1 ' inch = 2.54 cm F'= 9/5 C + 32 "C = 5/9 (*F - 32)

1:__ERINQlPLES_Qg_ NUCLEAR _PQWER_ PLANT _QPERATIQN s PAGE '23 o Id[RdQQYN@ dig @t_ME@I__TRAN@FER_AND_ FLUID FLQW

_ ANSWERS -- HATCH 1&2- -86/09/03-SPENCER, M.

ANSWER 1.01 (1.00) b REFERENCE DPC, Fundamentals of Nuclear Reactor Engineering, p. 96 EIH: L-RQ-602 ANSWER 1.02 (2.00)

a. Orificed Fuel Support Pieces (Core Orificing) (0.5)

b. As power increases, the flow to the central (higher powered) bundles would decrease CO.253; flow to the peripheral (lower powered) bundles would increase CO.253. This is due to'the increased two-phase flow resistance developed in the higher powered bundles where there is greater boiling CO.53 and the resultant " restriction" to flow,.thus sending it preferentially through the lower powered bundles CO.53.

REFERENCE-General Electric NEDE 24810, September 1983 GGNS: Thermal Limits LP, p 28 EIHz. GPNT, Vol-V, Chapter 2.2.J

, ANSWER 1.03 (1.00) l C

l REFERENCE l

EI HATCH'- Heat Transfer and Thermal Limits, Chapter 10.2, page 10.2-29 l

l ANSWER 1.04. ' (1.00) a.

REFERENCE EI HATCH - Heat Transfer and Thermal Limits, Chapter 10.2, page 10.2-7 l

l

"11____PRINg1E6gg_gE_NugLg@B_EgWg8_E68NT _ OPgB@IlgN2 PAGE 24 ISEBd9DyN8diggi_dg81_IggNgEgB_@ND FLUID;FLgW ANS'WERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 1.05 (1.00) 2 (CPR = CP/ ACTUAL POWER)

. REFERENCE EI HATCH - Heat Transfer and Thermal Limits, Chapter 10.2, page 10.2-10 ANSWER- '1.06 (1.00) c.

REFERENCE EI HATCH - Heat Transfer and Thermal Limits, Chapter 10.2, page 10.2-15 ANSWER 1.07 (1.00)

DELTA K / K = (Keff-1) / K eff

= .99 -1 / .99

= 0.0101 (0.25)

DELTA K / K = MTC ( RATE X TIME )

TIME = (DELTA K / K) / MTC X RATE TIME = 0.0101 / -1X10E-4 X 100 TIME = 1.01 HOUR (0.75)

REFERENCE EI HATCH - General Electric, Reactor Theory, Chapter 4, page 4-12b GE ANSWER 1.08 (2.00)

a. remain the same

b. increase

c. increase

d. decrease (4 3 0.5 ca) l l

[

1- _ PRINCIPLES;OF_ NUCLEAR _PQWER_ PLANT _QPERATIQNi PAGE 25

.- .IHE85QQyN@dlCS _HE@l_IB@NSEEB_@NQ_E6UlQ_ELQW t

ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

REFERENCE EI HATCH - General, Reactor Theory, Chapter 4, page 4-70 (GE)

ANSWER 1.09 (1.00) b REFERENCE EI HATCH- General Electric, Reactor Theory, Chapter 3, page 3-73 GE ANSWER 1.10. (1.50)

a. Increase

b. Decreases

c. Remains the Same (3 G 0.5 ea)

REFERENCE EI HATCH - General Electric, Reactor Theory, Chapter 5, pages 5-12a, 5-13a, and5-14a. GE ANSWER 1.11 (3.00)

a. T= 1.44 X doubling time CO.5'ea]

T= 1.44 X 85 T= 122.4 seconds [ 0.25 for math & 0.25 for units] (1.5)

(time / period)

b. P final = P initial e [0.25]

(time / 122.4) 8.8 X 10E5 = 3 X 10E3

• c In(293.3) = time / 122.4 time = 5.68 X 122.4 [0.38 ea3 C1/2 for method, 1/2 for math]

time = 695.23 seconds tanswer dependent on part i period] (1.5)

la__C81NQlPLES OF_ NUCLEAR _ POWER PLANT _gPERATION t PAGE 26-c 1HERMODYNAMICEt_ HEAT _TRAN@FER_AND_ FLU _I_D FLgW ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

REFERENCE EI HATCH - General Electric, Reactor Theory, Chapter 5, page 5-11b GE ANSWER 1.12 (1.00)

Speed droop is the percent of frequency drop E0.34] between the no load

[0.333.and full load CO.33] conditions of an ac generator.

REFERENCE EI HATCH - General Electric, Electrical Science, Chapter 7, page 7-90. GE ANSWER 1.13 (1.00) 4

a. Region II

b. Region III REFERENCE EI HATCH - General Electric, Electrical Science, Chapter 7, page 7-90. GE ANSWER 1.14 (2.50)

a. 2-3 i

b. 4-1 f

c. 3-4 t

d. 1-2

s. 2-3 -45 0 0. " es-t-ye w 'l- Y (5@O,5eM REFERENCE EI HATCH - Genera 1' Electric, Heat Transfer and Fluid Flow, Chapter 5, GE page 26a.

i I

-'1 PRINgIPLES_OF_ NUCLEAR POWER _ PLANT _QPERATICNt - PAGE 27

. ISEBdQDYN@ digs t_HE61_IB@NSEEB_@ND_ELUID_E(QW ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 1.15 (1.00)

The energy added to the nucleus CO.53 by.a thermal neutron will not overcome the critical energy of the atomEO.53. ( mass defect: binding Gnergy)

REFERENCE EI HATCH - Basic' Atomic and Nuclear Physics, Chapter 6, page 5.

ANSWER 1.16 (1.00)

.d REFERENCE SSM BOOK 2, CH 2-A, SEC 13.7, PG 161 DRESDEN LESSON PLAN BOOK FOUR CHAPTER 12, PAGE 35 EI HATCH - Reactor Theory, Chapter 5 ANSWER 1.17 (2.00)

a. 4.77% (+- .48%)

b. 3.8% (+- .38%)

c. 1.0% (+- .10%)

d. 3.0% (+- .30%) [4 0 0.5 each] (2.0)

REFERENCE Dresden Reactor Physics Lesson Plan Book 4, Ch. 12, PP. 26-32, Figures 58-62 GGNS: OP-NP-513, OP-NP-514 EIH: L-RQ-604; L-RQ-605 i

l r

l l

i

91__gL@NI_DEDIGN INCLUDING _ SAFETY _AND_ EMERGENCY _ SYSTEMS PAGE 28 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 2.01 (1.00)

Both the inlet (0.33) and outlet scram valves (0.33) for that rod'are open (0.33).-

REFERENCE BFNP: LP#29,P.6 EIH: GPNT, Vol VII, Chapter 9.2.1, Chapter 9.3 ANSWER 2.02 ('.50)

b. 1 (1.0)

REFERENCE GGNS: OP-P64-501, p7 EIH: Tech. Spec. 4.7.6.1.f.4 ANSWER 2.03 (2.00)

e. true

b. false

c. false

d. false REFERENCE EIH: GPNT, Vol V, Chapter 5.1..II.B; L-RQ-703 ANSWER 2.04 (1.50)

a. The D/P between the bottom of the core (above the core plate) .and

[ the spray sparger pipe just outside of the RPV (downstream of the j injection valve) is monitored. (0.5) l b. If the integrity of the core spray header is lost, the D/P will INCREASE. (0.5) f REFERENCE l

EIH: GPNT, Vol VI, Chapter 8.2; L-RQ-739 l

l l

l

'2.__eLeNI_pggl@N_INCLUp1NQ_g86EIX_8Np_gdggggNCY_@Y@IEd@ PAGE 29 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 2.05 (2.00)

1. 'All penetrations required to be closed during accident conditions are either capable of being closed by an operabel containment automatic isolation valve system, or are closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position. (except as'provided in TS. section 3.6.3.1)

2. All equipment hatches are closed and sealed.

3. Each containment leakage rates are within the limits of T.S.

4.- The sealing mechanism associated with each penetration is operable.

5. Each containment air lock is operable.

(any 4 9 0.5 ea)

REFERENCE EI HATCH . Lesson Plan Vol 5,' Chapter 3, page 3.1-28 Rev i ANSWER 2.06 (2.00)

a. Torus

b. Unit i = R24-5021 CO.253 Unit 2 = 2R24-5021 CO.25]

c. _3.

d. barometric condenser (" barometric" required for full credit)

(4 3 0.5 ea = 2.0)

REFERENCE EI HATCH - Lesson Plan Vol. 5, Chapter 4.5, Figure 4.5(1) Rev. 1 L-RQ-737 ANSWER 2.07 (1.50)

c. Prevents CRD pumps from attaining "run out" conditions during a scram. (0.5)

b. Unit 1 has four installed orifices CO.53 and the charging header isolation valve is throttled to act as the fifth orfice.CO.53 y -\ -m m g - .m - -<< _ . _ , ,

n - - - - - ,

Isl_ChgNI_Q@@l@y_lygbyQly@_@@Egly_AND_ EMERGENCY SYSTEMS PAGE 32 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

a REFERENCE EI HATCH - Lesson Plan Vol. 5, Chapter 4.2, Rev. 1

-ANSWER 2.08 (2.00)

Max = 125 minutes (0.5)

Min = 50 minutes (0.5)

Bases: Max = Assures the baron gets into the reactor considerably quicker than the cooldown rate. (0.5)

Min = Assures there is sufficient mixing so the baron does not retirculate through the core in uneven concentrations (0.5)

(Precludes power " chugging").

REFERENCE EI HATCH - Lesson Plan Book, LP 11.1 ANSWER 2.09 (1.00)

Low concentration of baron [0.53 can cause the moderator coefficient to b@ positive. [0.53 REFERENCE EI HATCH - Lesson Plan Book, LP 11.1 ANSWER 2.10 (1.50)

When-a pulse of ionizing radiation enters the detector, a definite number of ion-pairs are formed in the argon gas space surrounding the center electrode E0.53. Under the influence of the electric field caused by the applied potential, the negative ions will migrate toward the electrode and the positive ions'towared the cylinder wall E0.53.

This difference in potential is then electonically measured as radiation E0.53 REFERENCE EI HATCH - Lesson Plan Vol. 7, Chapter 9.1, page 9.1.1-5, Rev. 1

$21_gLANT DE@IgN; INCLUDING _ SAFETY _AND_gMERggNCY_gYgTgMg PAGE 31 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 2.11 (1.00)

1. lower argon.' pressure.

2. Iower operating voltage

3. lower uranium: content

4. closer spacing between cathode and anode (4 S 0.25 ea.)

REFERENCE EI HATCH - Lesson Plan Vol. 7, Chapter 9.1, page 9.12-3, Rev. 1.

ANSWER 2.12 (1.00)

R actor vessel water level does not fall below the top of the core [0.343 following a steam line rupture outside of the primary containment [0.333 bsfore the main steam line isolation valves close CO.333 REFERENCE

-EI HATCH - Lesson Plan Vol. 6, Chapter 5.1, page 5.1-6, Rev. O ANSWER 2.13 (1.00)

Min = ensures the pressure rise within the reactor will not.cause a significant increase in fuel cladding temperatures. (Limit power increases due to pressure spike.) (0.5)

Max = assures the fuel barrier is protected against loss of cooling if the MSIV closure. takes the max. time (ie. the steaming rate greater than the-feed rate and could uncover core}. (0.5)

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 5.1, page 5.1-9, Rev. O ANSWER 2.14 (1.00) c.

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 5.3, page 5.3-20 i

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

$22_gL9NT DEgIgN_ INCLUDING SAFETY AND EMERGENCY _gYgTEMg PAGE 32 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 2.15 (1.00) c.

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 6.5, page 5-6 m,moem ,4 , ,. mm 1ELc7 tt -

The Core Sprav Aye+rr C.0a; and LPCI~ operation E0.333 are arranged

.-__r rndently CT_7^' t 2::: pli:S th: 25jectiv ef ad:quet: crc: : cling.

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 8.8,.page 1-1.

\ .50 ANSWER 2.17 '!.02)

, 1 =B l

2=F 3=D 4=A 5=E o.25 6=C (6 G-er6-ea)

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 8.0, page 0-16 ANSWER 2.18 (1.00)

It has been determined that all control rods worth, regardless of rod pattern, are low enough that they will not result in significant fuel dcmage CO.53 if a rod drop accident occurs [0.53 (i . e. h </= 280 cal /gm).

i REFERENCE EI HATCH - Technical Specifications, page 3.3-15 (UNIT 1)

, '21',_E6@NI_pggigN_INC69Q1Ng_g8EgIy_8NQ_ EMERGENCY _ SYSTEMg PAGE 33 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 2.19 (1.00)

During high steaming rates, the oxygen content is lower than that at

-low steaming rates.

REFERENCE EI HATCH - Technical Specifications, page 3.6-18.

1 l

)

I

'3di_1 DEI 6gMgyIg_8dg_ggdIBg6g PAGE 34 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 3.01 (2.00)

1. Normal control range O to +60

2. Emergency shutdown range -150 to +60

.3. Shutdown vessel flooding range -17 to +383 4., Post accident flooding range. -317 to -17 (0.25 for range and 0.25 for value = 2.0)

REFERENCE EI HATCH - Reactor Vessel Instrumentation, Chapter 2.3, page 2.3-2 ANSWER 3.02 (2.00)

a. ADS Valves are in Double Boxes (0.5)

b. (1) No Demand (0.25) or Demand Met (0.25)

(2) ADS Armed (0.5)

c. 60 Minutes (0.25) and 6 minutes (0.25)

REFERENCE EIH: L-RQ-752 (1986); DCR-157

) . 'W

> ANSWER 3.03 '1.75L

c. -Yes [ Apply in accordance with part (b)] (0.5)

b. It can be moved out one notch (0.25) before a withdraw error will block further movement. (0.25) If ti,e ,wJ -o, ...m -ted. 1

-

E will move as far as the operator wants (0  ?"' u,1ums it is the third insert error. (0.9M) If it-rE the third insert error, then it would in r m en1y : : ctch 'S.25) D b -Am\ o n e- mk A (o 1*M bbe-

% m w b _ doc wAL We E bRu mom =~tLO'N REFERENCE l

EIH: GPNT, Volume VII, Chapter 9.2; L-RO-721 i

(

l

hi'_1NSIBydENIg_ANp_CgNI69Lg. PAGE 35 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 3.04 (1.00)

C REFERENCE

.EIH: GPNT, Vol VI, Chapter 7.2; EIH Simulator; L-RQ-742 ANSWER 3.05 (1.00)

.The closed breaker will trip open CO.33] and the breaker being closed will trip open CO.33] resulting in a de-engergized bus. [0.343 REFERENCE '

EI HATCH - Lesson Plan #27.1, page 13, lesson obj #13.

ANSWER 3.06 (2.00)

a. 2 upscale Hi Hi Hi_ radiation trips .

1 upscale Hi Hi Hi radiation trip and i downscale/inop trip 2 downscale/inop trips - 1 from each channel (0.5 each)

b. Off gas System Outlet and Drain valves, OR Discharge valve to the stack, cooler condenser and moisture separator drain valves, and holdup line drain valve. (0.5)

REFERENCE EIH: GPNT, Vol. VI, Chapter 6.8-21; Vol. VII, Chapter 9.7-1-7; L-RQ-708; L-RQ-746 ANSWER 3.07 (1.50)

1. Recirculation Pump RPT Breaker Permissive

2. RSCS Bypass

3. TSV, 10% Closure Scram Bypasss

4. TCV, Fast Closure Scram Bypass (3 3 0.5 each)

REFERENCE EIH GPNT, Vol V, Chapter 4.1-9; GPNT, Vol VI, Chapter 5.5-10; GPNT, Vol VII, Chapters 9.2.2-11 and 9.2.3-5

Edi_ld}I69dEMIE_@yQ,ggyI6QL@ PAGE 36 ANSWERS - . HATCH 1&2 -86/09/03-SPENCER, M.

w e,.,c o

, mo

,, mms

,-.~~<

M LETEb

1. e.

2. a.

REFERENCE BFNP: LP#12, p 24; T T #20; OI-57, p 53; RQ 85/01/02 EIH: ' Lesson .5 pp 10-12 BSEP- 26; HD 17-2/3-B, Section 3.2 mmdC. Or C30 5~1 ANSWER 3.09 ( .50)

True REFERENCE EI HATCH - Lesson Plan 5.1, page 20 ANSWER 3.10 (2.00)

1. Low Steam line pressure 100 psig +/- 10#

2. High HPCI steam line flow 300 % +/- 10%

180" D press water +/- 10"

3. High temperature in the HPCI Ambient + 90 F +/- 10 F steam line spaces

4. Turbine exhaust diaphragm high press 10 psig +/- 1 psig (8 9 0.25 ea)

REFERENCE EI HATCH - Lesson Plan Vol. 6, Chapter 8.3, page 8.3-8

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

L.32' INgIggMgNIg @Ng CgNIggL@

PAGE 37-ANSWERS ---HATCH 1&2 -86/09/03-SPENCER, M.

ANSWER 3.11 (1.75)

a. Ensures adequate core cooling for steam line breaks outside containment [0.25] which.are not isolated [0.25] and which do not result in a High Drywell pressure signal CO.25]

b. None ( will not close the ADS valves ). (0.5)

c. None ( CS pump pressure maintains the 2 minute timer contacts ) (0.5)

REFERENCE EI HATCH - Lesson Plan 38.1, page 5.

ANSWER 3.12 (1.50)

The RFP and Main Turbine trips (level) are 2 out of 3 trips. [0.253 Trip "A" & "C" utilize a commmon reference leg. (LTN004A & LTN004C}EO.253 If a fault in the common leg occurred, the indicated level would dccrease, E0.25] the reactor feedpumps increase in speed, E0.253 and the actual vessel icvel could increase [0.25] to the steam line connections on the vessel because the trips are out of service. [0.253 REFERENCE EI HATCH - Lesson Plan Vol 7. Chapter 9.5, page 9.5-12 L .15 ANSWER 3.13 'i.52)

a. 1. "B"

2. NONE

3. "A" & "B" (no part. credit for only one) (3 0 0.25 ea = 0.75)

6. 1 = a. (C82-P001) 2  :. 0i21 rip DuM%

2 3'= a. (C82-P001) J3'O 0.25 ea)

~

REFERENCE EI HATCH - Lesson Plan 52.1, pages 3 and 12.

J w- , ..,,,__~-_.... -

.7--, . - - , - m-..,..._ - , . - . _ , , _ , _ , . , , , , . , , - . , y - --

g g _'__ INSTRUMENTS _AND_CgNTRgLS PAGE 38 ANSWERS -- HATCH'i&2 -86/09/03-SPENCER,.M.

_ ANSWER 3.14 (1.50)

a. - One SBGT startsCO.251 All R/B supply and exhaust vent fans trip CO.253 and the inboard or outboard valves close CO.253.

b. Half a group isolation (inboard or outboard valves close. ) (0.25)

c. Half a group I isolation' signal CO.253 NO valve action CO.253 REFERENCE EI HATCH - Lesson Plan'# 44.3, page 10, lesson obj # 7.

ANSWER 3.15 (1.00)

1. Reactor pressure and MSL pressue between the outboard MSIV's and turbine stop and bypass valves must be less than 35 psig (+/- 5 psig). (0.5)

2. Dilution air must be available (0.5)

REFERENCE EI HATCH - Lesson Plan #49.1, page 7.

-ANSWER 3.16 (2.00)

c. 1. Low stator cooling water inlet pressure CO.25] 3 26.36 psig

(+/- 10%) CO.253

2. High stator water outlet temperature from generator CO.253 3 81 C (+/- 10%) CO.253

b. 1. Current less than 15000 amps (+/- 10%) CO.25]

within 2 minutes (+/- 1 min) CO.253

2. Current less tha 4500 amps (+/- 10%) CO.253 within 3.5 minutes (+/- 1 min) CO.253 REFERENCE EI' HATCH - Lesson Plan #17.1, page 10, lesson obj #14

'41 ' PROCEDURES - NORMAL _ABNQRMAL t 1_gMgRggNCY_AND PAGE 39

.- RAgIg69@lC8(=CONTRgL ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

' ANSWER 4.01 '(2. 00) c) 1. Five.(5), or more adjacent control rods not inserted below the 06 position (0.5)

2. Thirty (30), or more total control rods not inserted below the 06 position. (0.5) b) 1. RPV level cannot be maintained (0.5)

2. Suppression Pool water temperature cannot be maintained below 110'deg F (0.5)

REFERENCE EIH: 34AB-OPS-009-2 ANSWER 4.02 (1.00)

'd REFERENCE EIH: 34AB-OPS-002-2 ANSWER 4.03 (2.00)

1. Inability of High Pressure Systems (HPCI/RC .C/CRD) to restore level.

2. One or more Low Pressure Systems running (LPCI, CS, or 1 CP & CBP)

3. Reactor-Pressure greater than 350 psig (+/- 25 psig).

4. Reactor Level decreasing to 121.5" (+/- 5.5 ").

REFERENCE EIH: 3450-B21-001-2; 34AB-OPS-002-2 i

l ANSWER 4.04 (1.00)

Turn the handwheel'in the DESIRED POSITION direction (1/4 turn max) CO.53; Verify the locking devico integrity and proper installation by attempting to misposition the valve [0.53 REFERENCE EIH: 34GO-SUV-001-09 1-

'fE*_E699E996EE_2_d96U66t_6109606ht EMERGENgY_AND 'PAGE 43 669196QGig@6_ggNI6g6 ANSWERS -- HATCH 1&2- -86/09/03-SPENCER, M.

-y ANSWER 4.05 (2.00)

o. Alert CO.5J

b. (Notification of) Unusual Event CO.53

c. General Emergency CO.53

d. Site Area Emergency CO.53 REFERENCE EIH: GET Handbook, pp 57, 58, 60, 61; 10AC-MGR-006-OS ANSWER 4.06 (1.00) a REFERENCE BFNP BF-OI-66, pp 5-7 EIH: 34GO-OPS-001-2 ANSWER 4.07 (1.00)

1. Be alarming repeatedly or be incapable of acknowledging and resetting CO.53

2. . silencing of the annunciator must not have any

, detrimental effect on plant operations CO.53 l REFERENCE EI HATCH - 31GO-OPS-008-OS, page 4 ANSWER 4.08 (1.50)

c. ._,n.=_ . ^ _ . . . ... ,'

__ 3,000 m"h m (015) "cI b ** *

b. 18,750 mrem (0.5)

c. 7,500 mrem (0.5) i

E.

'i._&BQCEQUBES'- NQRMAL _ABNQRMAL t _t EMERGENCY AND PAGE. 41 8891gLQGlC@L_CQNIBQL 4

ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

REFERENCE EI HATCH - 60AC-HPX01-0 ANSWER 4.09 (2.00)

c. Clearance Number

b. Sequential Tag Number

c. Component Identification

d. Required Position of Component (4.9 0.5 ea)

REFERENCE EI HATCH 10AC-OPS-001-05 ANSWER 4.10 (3.50)

1. Place the Mode Switch to SHUTDOWN.

2. Depress all four Manual Scram Pushbuttons.

3. Confirm Flux is decreasing.

4. Check that green FULL IN lights are lit for operable rods. Manually insert any rods not FULL IN.

5 Depress the Main Turbine Trip button and check that generator PCBS and exciter field ACB trips after driving steam is depleted

<6 . After initial level transient.is over observe that Reactor Level stabilizes between +32 and +42 inches, by use of multiple indications.

7.. Maintain Reactor Level between +32 and +42 inches utilizing startup FW configuration. If required, transfer RFPT speed controller (s) to MANUAL.

8. Maintain reactor water level below steam lines. (7 9 0.5 ea)

REFERENCE EI HATCH - 34AR-OPS-001 ANSWER 4.11 (1.25)

c. Two SRM's shall be operable CO.253; one in the core quadrant where fuel or control rods are being moved CO.253 and one in an adjacent quadrant CO.253.

b. 3 cps (0.50)

REFERENCE l EI HATCH - Technical Specifications, page 3.10-2 I

l I

k.

'4Ii_E899E998E9_!_NgBdB61_BgdgBDB61_gdgBggdgy_ANg PAGE 42 889196991986_99NIgg6 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER,.M.

-ANSWER 4.12 (1.00)

1. Reactor. feedwater pump (s) speed decrease to low speed stop. (0.5)

2. Reactor recirculation system runback to speed limiter number one.-(0.5)

REFERENCE EILHATCH - 34AB-OPS-015-2 ANSWER 4.13 (2.00)

a. Less than 50 psig ( +/- 10%) (1.0)

b. To prevent a random insertion of indiv jual control rods. (1.0)

REFERENCE EI HATCH - 34AB-OPS-020-2 ANSWER 4.14 (2.00)

1. Loss of four rod display

2. Rod drift indicator lights for all rods.

3. Select rod block

4. RWM Rod Block annunciator if RWM in service (4 3 0.5 ea = 2.0)

REFERENCE EI HATCH - 34AB-OPS-024-2 ANSWER 4.15 (2.00)

c. 2 Minutes (0.5)

b. Shallow (1.0)

c. " Print Notch Error" function (of RWM) (0.5)

'011_tB99EgyBg!_ _Ng65@6g_6BNQByg62_gdEgggNgy_9NQ .PAGE 43

, B99196991986_99NIBQ6 ANSWERS -- HATCH 1&2 -86/09/03-SPENCER, M.

REFERENCE EIH: 42FH-ENG-010-1/2 ANSWER 4.16 (1.00)

To preclude establishing excessive stressus in the Jet pumps due to flow reversals - OR - to prevent excessive radial bcaring loads (on the recirc pumps)

REFERENCE EIH: 3450-B31-002-2; GPNT, Vol V, Chapter 4.1

l

' ' '

• TEST CROSS REFERENCE , PAGE 1 DUEDTION .VALUE REFERENCE 01.01 1.00 JFHOOOOOO1 01.02 2.00 JFH0000002 01.03 1.00 JFH0000003 G1.04 1.00 JFH0000004 01.05 1.00 JFH0000005 01.06 1.00 JFH0000006 01.07 1.00 JFH0000007 C1.08 2.00 JFH0000000 01.09 1.00 JFH0000009 01.10 1.50 JFH0000010 01.11 3.00 JFH0000011 01.12 1.00 JFH0000012 01.13 1.00 JFH0000013 01.14 2.50 JFH0000014 G1.15 1.00 JFH0000015 01.16 1.00 JFH0000016 01.17 2.00 JFH0000017 24.00 C2.01 1.00 JFH0000018 C2.02 .50 JFH0000019 G2.03 2.00 JFH0000020 C2.04 1.50 JFH0000021 02.05 2.00 JFH0000022 02.06 2.00 JFH0000023 C2.07 1.50 JFH0000024

• 02.00 2.00 JFH0000025 C2.09 1.00 JFH0000026 02.10 1.50 JFH0000027 C2.11 1.00 JFH0000028 02.12 1.00 JFH0000029 02.13 1.00 JFH0000030 02.14 1.00 JFH0000031 02.15 1.00 JFH0000032 02.16 1.00 JFH0000033 02.17 3.00 JFH0000034 02.18 1.00 JFH0000035 02.19 1.00 JFH0000036 26.00 C3,01 2.00 JFH0000037 ,

G3.02 2.00 JFH0000038 C3.03 1.75 JFH0000039 03.04 1.00 JFH0000040 03.05 1.00 JFH0000041 03.06 2.00 JFH0000042 03.07 1.50 JFH0000043 03.08 2.00 JFH0000044 M

e

5 TEST CROSS REFERENCE PAGE 2

' QUESTION VALUE . REFERENCE 03.09 .50 JFH0000045 03.13 2.00 JFH0000046 03.11 1.75 JFH0000047 C3.12 1.50 JFH0000048 C3.13 1.50 JFH0000049 C3.14 1.50 JFH0000050 C3.15 1.00 JFH0000051 03.16 2.00 JFH0000052 25.00 04.01 2.00 JFH0000053 C4.02 1.00 JFH0000054 04.03 2.00 JFH0000055 C4.04 1.00 JFH0000056 04.05 2.00 JFH0000057 C4.06 1.00 JFH0000058 04.07 1.00 JFH0000059 C4.08 1.50 JFH0000060 04.09 2.00 JFH0000061 C4.10 3.50 JFH0000062 04.11 1.25 JFH0000063 04.12 1.00 JFH0000064 04.13 2.00 JFH0000065 C4.14 2.00 JFH0000066 04.15 2.00 JFH0000067 C4.16 1.00 JFH0000068 26.25 101.25

m

• l

. s ENCLOSURE 3

, r. e 1

• Plant E. I. Hatch Reactor Operator License Eu. mination Comments Requiring Written Response September 3, 1986 1.14 Utility Comment:

The stem on part "e" has a conflict in wording. A real (non-ideal) turbine does not experience isentropic expansion of the steam passing through it.

Recommendation:

Since'the stem addresses both a real and ideal turbine process, the key should accept either of the following for full credit on part "e", " 2-3, or 2-3' ".

Reference:

EIH GE Heat Transfer and Fluid Flow Text, Ch. 5, pg. 5-26.

2.16 Utility Comment The question's wording is not specific enough to elicit the desired response that CS is redundant to LFCI. As such, the question may draw a wide variety of GUESSES ranging from LPCI inverters to diesel generators.

From the question, it appears that a response dealing with electrical power supplies is being asked. Additionally, the question is nor specific enough to relate it to a learning objective based on the task analysis.

Recommendation:

It is very strongly recommended that this question be deleted.

Reference:

EIH Lesson Plan Vol. 6, Ch. 8.3,.pg.'8.3-1, and the Plant Hatch Operations Job Task Analysis.

page 1 of 4

)

Plant E. I. Hatch Reactor Operator License Examination Coments Requiring Written Response September 3, 1986 2.17 Utility Comment The apparent intent of the question is valid in that it requires an integrated plant knowledge. However, the question is inaapropriate for the following reasons: To receive full credit for this answer, a detailed knowledge of both the thennal hydraulic response of a degraded core and an intimate knowledge of system interrelat'ionships in regards to component responses to actuating signals is required. The actual sequence is based on computer analysis, which the student would not be expected to have memorized. This material is presented in the text as supporting details, not NEED 1D ENOW material. An operator's responsibilities require him to confirm automatic actions, which would take place in this instance after the entire sequence had occurred. He most likely would not have sufficient time to observe the' actual sequence of events.

Recommendation:

It is recomended that this questhn be deleted, however, as a MINIMlbi, it is very strongly recomended that the point value of this question be reduced.

Reference:

Per key, and Plant Hatch Operations Job Task Analysis.

3.03 Utility Coment The question's' answer key for part "b" is technically inaccurate based on current Rk}! operations. An insert rod block occurs after a single insert error.

Recomendation Change key to reflect actual operation of the RM1 as follows,"It can be moved out one notch (0.25) before a withdraw error will block further movement. (0.25) It can be inserted.one notch (0.25) before an insert error will block further movement.(0.25)"

Reference:

Unit I and 2 Process Computer Software Change Request number 85-2, OrN 85-8.

page 2 of 4

a v,.- ,

.- Plant E. I. Hatch Reactor Operator License Examination Comments Requiring Written Response September 3, 1986

. 3.08 Utility Comment Both parts "a" and '.'b" are . technically inaccurate. The' plant cannot operate in dp mode at 100% power because the SU level control valve will only pass approximately 20% feedwater flow.

Recomendation

. % e answer key needs to-be changed to reflect the " correct" response of "e. None of the above." for both part "a" and "b". Alternately the question may be considered for deletion.

Per the key.

3.13 Utility Comment h e panel number choices in part "b" do not give noun descriptors. At Plant Hatch, equipnent is typically not referred to solely by MPL numbers. For this reason, operators are not expected to memorize plant components by panel MPL ntsnbers. For example, had the choices been listed as "a. C82-P001, RSDP on 158' SW Rx Bldg" this would have been an appropriate testing item.

Recomendation Delete part "b".

Reference:

Per: key.

l l

page 3 of 4 t  ;

r

-.. Plant E. I. Hatch Reactor Operator License Examination Consnents Requiring Written Response September 3,.1986 4.07 Utility Comment The procedure that applies in .this instance is an administrative procedure that is the responsibility of supervision. Specifically, the decision to declare an annunciator a nuisance is made jointly by the shift supervisor (SS) , and the operations supervisor on shift (OSOS) per 7.4.1.1 of 3100-OPS-008-0S, " Control of Annunciators".

Reconunendation It is recommended that this question should be deleted based on the above criteria. It is also recommended that in the future this question be used solely on SRO exams.

Reference:

Procedure 3100-OPS-008-OS 4.08 Utility Comment The question does not address whether or not the federal or plant Hatch

'administratiive limits are required for the answer. The procedure addresses both limits. Therefore, the answer key on part "a" is not

. complete. The federal limit for a quarter is 1250 mrem, however with the fonn 4, exposure can be extended to 3000 mrem provided total lifetime exposure does not exceed 5 rem ( age-18) per 60AC-HPX-01. The Plant IIntch administrative dose limit is 1000 mrem for a quarter. This can be extended by satisfying various administrative requirements to the federal limit of 3000 mrem.

Recommerxlation Revise key for part "a",_to allow full credit for either 1250 mrem, or 3000 mrom provided 5(n-18) is not exceeded. or 1000 mrem administrative limit.

Reference:

Procedure 60AC-HPX-001-0 l

page 4 of 4 l

Plant E. I. Hatch Reactor Operator License Examination Questions for NRC Attention-September 3, 1986

. 1.08 : Utility Consnent:

Question is uncicar as to intent due to the omission of " magnitude'.', i.e is the change in magnitude of the coefficients the change desired or simply the numerical value. For example, a negative coefficient going

-from a value of -1x10-4 to .5x10-4 could be described as both

• increasing (in the positive direction) and decreasing -(in magnitude).

Suggested resolution:

If apparent that the student.used numerical value vice magnitude for the change, accept the opposite response. For future use, question should.be revised to specify magnitude.

Reference:

Per key, and EIH GE Reactor '1heory Text, Ch. 4, pp. 4-18 to 4-24.

+

6 1.11. Utility Comments

Ihere is a discrepancy between the point value on the. question and the point value breakdown on the key. i Suggested Resolution Change the point value breakdown on the key to total the point value on -

the question or change question point value, as required.

Reference:

-N/A t

4 page 1 of 5 i

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

....t ..-.__m.. , , ~ , - . . _ , . , - , ~ , , , , , -

,,,..,__-_-~,,,.-,-._v.,,,..,,..., .v,-,__ y. ,,

Plant E. I. Ilatch Reactor Operator License Examination Questions for NRC Attention September 3, 1986 1.12 Utility Coment:

This question reference speed droop to the' diesel generator. The key is based solely on a definition of speed droop as it relates to any ac 0 generator. 'Ihe GE Electrical Theory text on page 7-54 and 7-55 discusses the function of speed droop on a diesel generator operating in parallel.

'Ihis is the concept that is stressed by plant Hatch training during the simulator training, and in classroom as a review during the diesel generator module.

Suggested Resolution Allow the following response to provide for full credit. "The speed droop control is used to limit the additional load the diesel generator picks up a result of variances in the grid's loading."

Reference:

Elli GE electrical Theory Text, Ch. 7, -pp. 7-54, 7-55.

-1.17 Utility Coment:

The concept that is addressed in this question is comonly tested at flatch. However our method involves identifying the correct responses when given a list of possible responses. Although the key does allow for some variances around the correct responses, depending on the asstanptions the student made for his response, his answer may fall slightly outside the bounds of the key's answers. For example, if moderator coefficient is assumed to be -1x10-4 K/K/*F, and the temperature rise is assumed to take place from 100F to 500F, this would result in a response of 4%,

wrong per the answer key. Additionally, the references listed do not identify the amount of reactivity that is compensated for by the excess fuel loading.

Recommendation:

Increase the band of acceptabic answers as follows: a. 4 to 5%, b. 3 to 4%, c. 1% +/ .25%, d. 3% +/ .5%.

Reference:

Per Key, and EIll GE Reactor Theory Text, Ch. 4 and 6.

2.02 See Question 1.11.

2.04 See question 1.11 page 2 of 5

Plant E. I. IIntch Reactor Operator License Examination Questions for NRC Attention September 3, 1986 2.05 Utility Coment The Unit is not specified in' the question. The tech spec definitions differ. The key reflects only the thit _2 definition.

Recomendation:

Change key to accept either Units tech spec definition's requirements.

Add the following Unit 1 definition to the key: 1. All non-automatic containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed. These valves may be opened to perform operational activities. 2. At least one door in the personnel airlock is closed and scaled. 3. All automatic containment isolation valves are operable or deactivated in the-isolated position. 4. All blind flanges and manways are closed.

Reference:

Unit I and II Tech Specs, Definition Sections 2.14 Utility Coment The valve designation for the S/U Icvel control valve on Unit 1 is incorrect, Unit I has two valves, the F212, and the F213. Neither of these valves bypass any of the listed components. The RFP bypass valve, F113 is a Unit 2 valve. The Unit 1 RFP bypass valve is the F032 valve.

Additionally, at plant liatch systems are not typically taught by valve numbers, a noun descriptor is used. This question is technically inaccurate. Given the choices for the question, it is believed that most operators would have deduced the desired response since the F113 is the RFP bypass valve on Unit II.

Recomraendation*

No action on this exam, however, revise this question for future.

examinations.

Reference:

Per key, and plant flatch P&ID 11-16303 to 11-16305.

page 3 of 5

Plant E. I. Hatch Reactor Operator License Examination Questions for NRC Attention September 3, 1986 3.06 Utility Coment The answer key is in error for part "a". The trips are Upscale, Downscale, and inop, not downscale/inop. Any combination of these trips will cause an isolation.

Recommendation Change key for part "a" to read, "Any combination of two of the following on the two post treatment radiation monitors, Upscale, Downscale, or Inop.

Reference:

Annunciator Response Procedure: 34AR-602- -2. The lesson plan has a typographical error.

3.10 Utility Comments Answer key is incomplete. The question also uns not specific as to the Unit involved. Credit should be allowed for either units' setpoints, since the question was not specific as to the desired unit.

Recommendation Change answer key to reflect the following, "1. Low steam line pressure

- 100 psig, 2. High HPCI steam line flow - 307% (303%), 3. Turbine exhaust diaphragm high pressure - 20 psig'(20 psig), 4. High temperature leak detection system, n. HPCI pipe penetration room - 169 F (N/A), b.

HPCI emergency area cooler outlet temperature - 169 F (169 F), c.

Suppression Chamber area air temperature 169 F (169 F), d. Suppression Chamber area differential air temperature - 42.5 F (42 F), N0fE: 4c and 4d have a 15 minute time delay associated with them, should not be required for full credit. Setpoints - Unit 2 (Unit 1)

Reference:

Unit I Tech Specs. Table 3.2-2, Nos. 9,10,11,12,13. Unit II Tech Specs, Table 3.3.2-2, No. 4.

page 4 of 5 l

1 l

i

• - Plant E. I. Hatch s

Reactor Operator License Examination

-Questions for NRC Attention September 3, 1986 3.16' Utility Comment-'

Part ."b" answer key needs additional infonnation for its part "1" to be

-completely accurate.

Recomendation Change answer key for part "b.1." to read as follows, " Current less than 15000 amps'(+/- 10%) (0.25) within 2 minutes (+/- 1 min) (0.25), if generator current was greater than 19065 amps initially (not required for full credit.

Reference:

Per Key.

4,03 -Utility Comment lhe question was not specific as to the location of the break, and' the answer key is not completely correct.

Recomendation Reccmsnend changing the answer key to reflect the following, "1. HPCI, RCIC, CRD, or Condensate and FW are unavailable, or unable'to maintain rx unter level., 2. Core spray or RHR is running. 3 and 4 as per key."

j

Reference:

Procedures 34AB-OPS-002-2S, 34AB-OPS-004-2S, 34AB-OPS-005-2S.

,i

, ..i i

3 i

page 5 of 5 i

L L.

t SAFETY EVALUATION

()

(_/ (

The Proposed change to the RWM software reducins the number of allowable insert errors from three to cne is also a change in the function of the RWM as described in the FSAR. The Proposed Chander however, does not. constitute an unreviewed safetw ouestion because:

1) the Probabilitw of occurrence'or the consecuences of an accident or malfunction of couiPaent imPortant to safets PreViouslw evaluated in the FSAR is reduced since the number of allowable insert errors is moved in a more conservative

(. direction than currentlw allowed, M. ,' ii) the possibilitu for an accident or malfunction of a different

+g/ twPe than anw evaluated Previous 19 in the FSAR is not created,

'7 ande g.

• " Y iii) the margin of safets as defined in the basis for anw technical

. sPOCification is not reduced since no technical specification cddresses the number of allowable insert errors in the RWM.

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-= l Rafi Software change to RWM on U2 )

Tho attached software change and safetw evaluation will bring the U2 RWM uP to

'th9 eane level as the U1 RWM. Whenever there are consecutive RWM groups with th9 came rods, an apparent anomalous behavior was observed wherebw the RWM i attscPted to latch uP to the next group even.though there were still two more rods left in the current group. The behavior is well understood (there being no es1 function of the RWM), and this change will eliminate this behavior. It  !

changes the allowable number of insert errors from two to zero. U1 was changed sometime back and no anomalous behavior has been observed since that time.  ;

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< ;c SAFETY EVALUATION The Proposed change to the RWM software reducing the number of allowable insert errors from three to one is also a change in the function of the RWH as described in the FSAR. The Proposed Changer however,

, does not constitute an unreviewed safetw ouestion because!

i) the Probabilits of occurrence or the conseauences of an accident or ac1 function of eauiPment imPortant to safetw Previous 1w evaluated in the FSAR is reduced since the number of allowable insert errors is moved in a more conservative direction than current 1w allowed, 11) the Possibilits for an accident or malfunction of a different

~

twPe than Onw evaluated Previous 19 in the FSAR is not created, and, iii) the margin of safetw as defined in the basis for anw technical specification is not reduced since no technical specification addresses the number of allowable insert errors in the RWM.

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ENCLOSURE 4 Confirined SPDS Training for the following personnel.

E_A_ME 1RQ-533(CR) LRQ-534(SIM) A'ITENDED GRADE SHEET (IRQ 86-1) (IRQ 86-1)

Anderson,J.L. IRQ-85-1K LRQ-86-1 G F Britt,S.R. IRQ-85-1K 1RQ-86-1 G G Dunham,G.W. LRQ-85-1K 1RQ-86-1 F F Eason,R.D. IRQ-85-1K 1RQ-86-1 G F Hayes,L. IRQ-85-1K LRQ-86-1 G G Spell,R.O. LRQ-85-1K 1RQ-86-1 G G Wiggens,A.E. IRQ-85-1K , IRQ-86-1 G F EXAM PREP 86-1 RO-44.2 (CH) 1RQ-534 (SIM)

Whitaker,H.L. 6/16/86 8/15/86 l

j Dowd,J.A. 6/16/86 8/15/86 Harris,S.D 6/16/86 8/15/86 l

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