Text

Exam Rept 50-572/86-01 on 860114-16.Exam Results:Five of Six Candidates Passed Exams & Issued Senior Instructor Certificates.One Candidate Failed Written Exam.Master Exam & Resolution of Facility Comments Also Encl
	ML20140E909
Person / Time
Site:	05000572
Issue date:	03/24/1986
From:	Dudley M, Kelly R, Kister H; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20140E899	List: IR 05000572/1986001; ML20140E895;
References
	50-572-86-01, 50-572-86-1, NUDOCS 8603280232
	Download: ML20140E909 (62)
	v • d • e

{{#Wiki_filter:< . . . . o U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.: 50-572/86-01 '(0L) FACILITY DOCKET NO.: 50-572 LICENSEE: Westinghouse Electric Corporation P.0 Box 598 Pittsburgh, Pennsylvania 15230 FACILITY: Nuclear Training Services Facility EXAMINATION DATES: January I4-16,1986 //) /[ - 3- % /- Yd CHIEF EXAMINER: Reactor Engineer Examiner Date REVIEWED BY: N/ k8 Chief, Projects Section 1C Date ~ ~ APPROVED BY: J Y [ [o Chief, Pfrbjects Branch No. 1 ' DAte SUMMARY: Six Senior Reactor Operator Instructor Certification candidates. were examined. One of the candidates failed the written examination.

Five Instructor Certifications were issued.

8603280232 860324 l PDR ADOCK 05000572.. V pg '1 . . ,

. . , . REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: l Inst. Cert l l Pass / Fail l I I I I I l Written Exam l 5/1 I l l I I I I l Oral Exam l 6/0 l

1 I I I I l Simulator Exam I 6/0 l l l

I I I l Overall l 5/1 l l l l CHIEF EXAMINER AT SITE: A. Vinnola, EG&G OTHER EXAMINERS: B. Picker, EG&G m - r ie** vew

, . . . ,

1.

Summary of generic strengths or deficiencies noted on oral and simulator exams: Some Reactor Operators and Balance of Plant Operators did not adequately communicate turbine and reactor power levels during normal power changes.

Some candidates were weak on the operations of the Nuclear Excore instrumentation.

2.

Summary of generic strengths or deficiencies noted from grading of written exams: Candidates did not know the additional loads supplied by batteries NK 11 and NK 14.

3.

Comments on availability of, and candidate familiarization with plant reference material in the control room: Candidates used the Instrument Response Book during abnormal casualties instead of the Abnormal Operating Procedures.

4.

Personnel Present at Exit Interview: NRC Contractor Personnel A. Vinnola, EG&G B. Picker, EG&G Facility Personnel K. Ruzich J. Fischer J. Jones 5.

Summary of NRC Comments made at exit interview: The examiner noted that candidates did not refer to the Abnormal Operating Procedures during the simulator examination, but rather used the Instru-ment Response Book. Also, there was poor communications between operators ' concerning turbine and reactor power levels during normal plant power changes.

The examiner noted that.there was a generic weakness among the candidates concerning the manipulations of the Excore Nuclear Instrumentation system during normal and abnormal system operations.

6.

Facility comments and the resolution of facility comments on the written examination are contained in Attachment 2.

The changes to the written examination as a result of facility comments are also contained in Attachment 2.

l

r .. . . .

Attachments: l '. Written Examination and Answer Key (SRO) 2.

Facility Comments on Written Examinations Made After Exam Review i . _ _

,htYMbmenif/ ,-

'

U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SNUPPS-2 _.

' REACTOR TYPE: PWR-WEC4 , DATE ADMINISTERED: 86/01/14 . EXAMINER: PICKERuB. APPLICANT: YlhJl-L \\e W l INSTRUCTIONS TO APPLICANT: Use separate paper for the answers.

Write answers on one side only.

Staple question sheet on top of the answer sheets.

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

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS M.4L 2fh40 25.00 6.

PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _R5 00 _ZL_QQ 7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 24. 2f _M _ZE.00 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ~ 98. 7f 100-00 100.00 TOTALS FINAL GRADE % All work done on this examination is my own. I have neither given nor received aid.

APPLICANT'S SIGNATURE .- - _ _ _ - - . _ _- . . . _. _ _.. _. _ _, _

__ ~~ 5.

THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE

JE' ERMODYNAMICS o - .. QUESTION 5.01 (3.75) A reactor whose reactivity information is given in figures 5.1 - 5.14 attached has been operating at 100% power for four weeks when a reactor , trip occurs.

Assume that the reactor is near BOL, and the boron concen-centration is 900 ppe.

Just prior to the trip all rods are out except for Bank D which is 180 steps withdrawn.

Assume boron worth is 10.0 pcm/ ppa, Beff = 0.007, and lambda =0.08 and all rods insert on the trip.

a.

What will Keff be immediately after the reactor trip (assume Tave is at T no load)? State all assumptions, and show all calcula-tions.

(1.2) b.

On a restart, 48 hours after the trip you wish to go critical at 125 steps on Bank D.

What is the required boron concentration? State all assumptions and sources of numbers and show all calcula-tions.

Assume the change in Samarium concentration is negligible.

(1.55) c.

If the reactor is just critical at 10 E-8 amps on the IR range with Bank D at 125 steps, what new Bank D position would be required to obtain a stable startup rate of 0.5 DPM? Show all work.

(1.0) QUESTION 5.02 (2.00) Explain why a dropped rod, when operating at power, could be worth approximately 200 pcm and the same rod stuck, on a reactor trip, could be worth 1000 pcm.

QUESTION 5.03 (2.00) During a reactor startup, the operator stops rod pull no. 10 at 180 steps on Bank C.

The source range monitor (SRM) count rate levels off at 3000 CPS.

The initial SRM count rate was 150 CPS at zero steps withdrawn on Control Bank A with Keff = 0.94.

a.

Calculate the subcritical multiplication factor for this control rod position.

(0.5)

b.

Determine, by calculation, whether the reactor is suberitical, critical or supercritical at this control rod position.

Show all work.

(1.5) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) , i- - ,.- -. - , . . _-. ._ __--_ -_ _ _._.,

~' 5.

THEORY OF__ NUCLEAR POWER _ PLANT OPERATION. FLUIDS _LLND PAGE 3 l

THERMODYNAMICS , .. QUESTION 5.04 (2.50) The turbine power is reduced from a steady state EOL load of 75% to a new steady state load of 50%. Rods are in manual, Boron is at 150 ppm, and turbine is in auto.

ASSUMING: NO operator intervention and No reactor trips.

BOW and WHY will the following parameters change? a.

Main Steam pressure (0.75) b.

Main steam flow rate, Ms (0.5) c.

Main steam temperature, Ts (0.5) d.

Tave (0.75) QUESTION 5.05 (1.50) Answer the following questions pertaining to xenon instability or xenon oscillation.

a.

What is the hazard resulting from xenon oscillations? (0.5) b.

At what neutron flux level (high or low) are xenon oscillations more likely to occur? (0.5) c.

What major change over core life would cause xenon oscillations to be less severe at EOL as compared to BOL? (0.5) QUESTION 5.06 (2.00)

., g

g During the power operation, the pressurizer PORY is cracked open such that the entire pressure drop down to the PRT pressure of 20 psia is considered to occur across the valve.

Determine the quality of the steam exiting the valve.

Show all work.

. (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) -. -- _ _ _ .-- -- __ - - .

5.

THEORY 0F NUCLEAR _ POWER PLANT OPERATION. FLUIDS. AND PAGE

THERMODYNAMICS

.. QUESTION 5.07 (2.00) { ' Refer to Fig. 5.15 to answer the following questions: a.

What protective device is needed on the discharge side of pump no. 3 and is not needed on the discharge sides of pumps no. 1 and no. 27 (0.5) b.

What will be the total system flow rate (GPM) and the pump discharge header pressure with all three pumps operating? EXPLAIN.

(1.5) QUESTION 5.08 (1.50) Will the comparison of thermal neutron flux for a reactor operating at full power near BOL to the same reactor operating at full power near EOL be HIGHER, LOWER, or the SAME? EXPLAIN YOUR CHOICE.

(1.5) (,ce s i d a r Ut -u s 4) QUESTION 5.09 (3.00) a.

What four primary system parameters do you, as the operator, monitor to assure that DNBR is greater than 1.37 (2.0) b.

What is the significance of maintaining DNBR >= 1.3? (1.0) QUESTION. 5.10 (1.00) What is the most significant type of heat transfer (conduction, convection, or radiation) taking place under each of the following conditions? Consider each condition separately.

a.

Nucleate boiling.

b.

Accident condition in which coolant is boiled and converted to steam in the reactor vessel.

c.

Heat from fission thru the fuel rod.

d.

Decay heat removal by natural circulation.

(1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) . . - - - - - - -

i 5.

THEORY'OF NUGLEAR_PQWER PLANT OPER& TION _ FLUIDS._uAND PAGE

t THERMODYNAMICS

.. QUESTION 5.11 (2.25) HOW and WHY is the Doppler Power Coefficient affected by each ~ f the following?_ (Answer each independently) o a.

Buildup of fission gasses in the fuel to clad gap.

(0.75) b.

Fuel densification.

(0.75) c.

Clad creep.

(0.75) QUESTION 5.12 (1.50) a.

What TWO reactivity factors are utilized to accomplish a STRETCHOUT on a reactor core? EXPLAIN.

(1.0) b.

What is ONE disadvantage with allowing a core to complete a STRETCHOUT? (0.5) , (***** END OF CATEGORY 05 *****) . e-h- - -- - vu - y- -- e

6.

PLANT SYSTEMS DESIGN CONTROb _6ND INSTR M N_TATION PAGE

u t

.. . QUESTION 6.01 (1.00) What is the reason for the interlocks that prevent closing of the letdown isolation valves from the main control board unless all letdown orifice isolation valves are closed? QUESTION 6.02 (1.00) If during a daily load follow operation, boron concentration changes are required and the changes are within the capabilities of either the BTRS or the reactor makeup system, which system would you use? WHY? QUESTION 6.03 (2.00) An attempt to open the RCS to RHR Pump A suction isolation valves, PV-8702A and HV-8701A, is made and they fail to open.

List FOUR interlocks, any one of which could be preventing the valves from opening.

QUESTION 6.04 (3.00)

List the FOUR permissive / control interlocks associated with the control circuitry of the steam dump system and give the purpose for each.

QUESTION 6.05 (2.00) What conditions will initiate an AUXILIARY FEEDWATER ACTUATION SIGNAL (AFAS) for: a.

A motor driven AFW pump? b.

The turbine driven AFW pump? Supply set points and coincidences where they are applicable.

QUESTION 6.06 (1.50) Give THREE advantages that are obtained by controlling the speed of the feed pumps and thus the differential pressure between the steam system and the discharge of the feed pumps.

(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) .- . __.

. . __

_ _ _ _ - _ - 6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

'

.., QUESTION 6.07 (1.50) What are the SIX sources of input data to each microprocessor of the Subcooling monitor?- QUESTION 6.08 (2.00) Following a reactor trip, the recorder trace of one of the intermediate range neutron detectors levels out at 10 E-9 amps: a.

Is the detector likely to be overcompensated or undercompensated? (0.5) b.

What is the compensating voltage used for? (1.0) What additional operator action will be required in the above c.

case to complete the shutdown? (0.5) QUESTION 0.09 (1.00) What does the intermediate range rod block do and what is its purpose? QUESTION 6.10 (3.00) Answer the following questions pertaining to rod insertion limits: a.

Provide TWO reasons / basis that explain why the limits are power dependent.

(1.0) b.

What TWO accident conditions are considered in the determination of the minimum rod insertion limits? (1.0) c.

Why does the minimum shutdown margin requirement change over core life? (1.0) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) ., -., - - - - - - _. _ - . - -

. ' 6.

PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE

.. , , i QUESTION 6.11 (2.50) a.

What are three different sets of conditions that will initiate the load shedding and sequencing system for a train of safeguards power? (1.5) b.

What TWO conditions are required to actually start the Shutdown Sequencer? (1.0) QUESTION 6.12 (2.00) What are the FOUR additional loads supplied by NK11 and NK14 batteries, such that they are required to have additional capacity? QUESTION 6.13 (2.50) Using the attached drawing, SNP-OPS-25, Reactor Core Safety Limits vs. Protection Boundary, indicate on your answer sheet WHAT protection is associated with each line, 1-5.

(2.5) (***** END OF CATEGORY 06 *****) --- .-.

" 7.

PROCEDURES - NORMALuABNQRMAL. met!ERGENCY AND PAGE

RADIOLOGICAL CONTROL , -. . QUESTION 7.01 (3 00) List the FOUR IMMEDIATE ACTION steps of FR-S.1, RESPONSE TO NUCLFAR POWER GENERATION /ATWS.

Also provide the " Response not obtained" actions for the FIRST TWO steps only.

QUESTION 7.02 (1.50) List the TWO independent conditions that would require stopping of all RCP's according to procedure E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

QUESTION 7.03 (2.25) What are the THREE overall objectives of E-3, STEAM GENERATOR TUBE RUPTURE, according to the ERG Background Document? QUESTION 7.04 (2.25) One of the IMMEDIATE ACTION steps in ECA-0.0, LOSS OF ALL AC POWER, states " Check if RCS is isolated".

What THREE things does the operator check to satisfy this step? QUESTION 7.05 (2.00) a.

What is the difference in operator action required for SI Actuation and SI Reinitiation criteria found on the E procedure FOLDOUT PAGES? (1.0) b.

What TWO conclusions, concerning plant status, can be drawn if either of the SI Actuation criterion on the FOLDOUT PAGE is

0'$b*es *fa carvleelades '. l. !$crNtox esceetlef Qhad daltane) ca.n he,

' e ~rn4c. concernmg -the occadi P ad sfo:fw5 Aelb l2., IJhec/ mutf 1b. opciido,- l Jo in 9encra.1 hrms don-the Crbion is egceedee(., QUESTION 7.06 (1.00) What is the reason for the following caution in GEN-N-02, PRIMARY PLANT HEATUP-HOT SHUTDOWN TO HOT STANDBY.

" Ensure that the steam pressure is above 585 psig prior to raising the RCS pressure above 1970 psig"? (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) . - . _ -. _.

.- -

- - . .. 7.

PROCEDURES - NORMAL. ABNORl h EMERGENCY AND PAGE

" R_ADIOLOGICAL CONTRO_L_ -- , ,

. . QUESTION 7.07 (1.25) What recourse does an individual worker, who suspects violations of 10 CFR regulations (especially those applicable to radiation protection), have available to him? QUESTION 7.08 (3.25) - ( Pr.., -b e -mr4 pa,.hr S A 20 year old radiation worker has a lifetime accumulated dose of ,6 rem and his form NRC-4 is maintained current by the utility.

During the present quarter, he has already received 1000 mrea.

a.

How much more dose may he legally receive during this quarter.

(1.75) b.

What dose can he legally receive per 10CFR20 in-the next quarter, assuming that he received the maximum legal dose in this quarter? (1.5) I I QUESTION 7.09 (3.00) ! In accordance with 10 CFR 20, certain occurrences involving licensed material require notification of the NRC.

Indicate by number whether occurrences a.

through d. requires: 1.

Immediate notification followed by a written report within 30 days.

or 2.

Notification within 24 hours followed by a written report within 30 days.

a.

A whole body exposure of S rem or more, or the skin of 30 rem or more, or of the extremities of 75 ren or more.

(0.75) b.

Loss of theft of licensed material if the circumstances could potentially result in a substantial hazard to persons in unrestricted areas.

(0.75)! c.

Loss of one working day or more of operation.

(0.75) d.

Damage to property greater than 200,000 dollars.

(0.75) ! l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)

7.

PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE

" RADIOLOGICAL CONTROL .

. -. . QUESTION 7.10 (1.50) a.

How long of a wait period is required before a restart of ' a RCP, if it stopped before reaching full speed? (0.75) b.

What is the basis for the wait time 7 (0.75) QUESTION 7.11 (2.50) During an accident recovery procedure, the operator who has been monitoring the Critical Safety Functions reports the following information.

a.

Suberiticality - green b. Core cooling - orange c.

Integrity - orange d. Beat Sink - red e. Containment - yellow Rank the CSF's according to their importa- .e in numerical order from 1 to 5.

QUESTION 7.12 (1.50) BRIEFLY EXPLAIN why an isotope analysis for iodine must be performed within 2 to 6 hours following a thermal power change in excess of 15 percent in a one hour period.

(***** END OF CATEGORY 07 *****) l -- -. .. .-. . - -

... _ _ _ _ _ _ _ " 8.

ADMIN _ISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIONS PAGE

'

.. . . QUESTION 8.01 (1.50) What are 3 of the 4 actions must be taken or operational limitations observed following a violation of a safety limit?

QUESTION 8.02 (1.50) < ' Temporary changes to procedures may be made if certain provisions are met.

List three such provisions.

! QUESTION 8.03 (1.50) ! ! Unexpected absenses can allow the shift crew composition to be one less than the minunum requirements if certain provisions are met.

What are these provisions?

t f i QUESTION 8.04 (4.00) ! a.

It is 0400 on 8-20-85 and the reactor power is operating at 40%. Considering the Delta I history listed ~below, at what clock and ! calendar time can power be increased above 50%? Assume the l Delta I Technical Specifications have not been violated during ' the given history and power changes shown are step changes.

' Show all work.

(2.5) DATE TIME (leaving band) TIME (reentering band) POWER _______________________________________________________________ 8/19/85 0300 0308

i 8/19/85 1747 1833

, 8/20/85 0248 0400

l l b.

Give the basis for maintaining AFD within Tech. Spec. limits.

(1.5) I: QUESTION 8.05 (3.00) e i What 5 conditions must exist for containment integrity to exist in accordance with the Technical Specification definition? Details such as leak rates, etc., are not required.

.

e (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l i i l ' I .. _ _ - _ - - _. _.

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' 8.

ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE

. . -. . QUESTION 8.06 (2.25) Give the basis for each of the following reactor trips: a.

Loss of primary flow.

(0.75) b.

Pressurizer high water level.

(0.75) c.

Overpower delta T.

(0.75)- QUESTION 8.07 (1.50) What actions must be taken when a Limiting Condition for Operation is not met AND the associated ACTION requirements do not apply? QUESTION 8.08 (2.00) Answer the following questions pertaining to refueling operations: a.

A Tech. Spec. states, "when any reactor vessel head closure bolt is less than fully tensioned or with the head removed, the boron concentration shall be maintained sufficient to ensure the more restrictive of the following reactivity conditions is met".

State the TWO restrictive reactivity conditions.

(1.0) b.

What is the Limiting Condition for Operation for the Source Range monitors? (1.0) QUESTION 8.09 (3.00) In accordance with the Emergency plan, what are 6 of the 7 primary responsibilities of the on-shift Emergency Organization? l i (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) . -. _ _ _

' 8.

ADMINISTRATIVE PROCFDURES. CONDITIONS AND LIMITATIONS ~ PAGE 14 ' m . . _ . . QUESTION 8.10 (2.50) In accordance with the SNUPPS Emergency Plan: a.

Who by job position / title assumes the Fire Brigade Leader . position AND to whom does the Fire Brigade Leader report? (1.0) b.

What duties / functions are performed by the Shift Technical-Advisor.(STA) upon declaration of an emergency? (1.0) c.

Which member of the On-Shift Emergency Organization (by job position / title) has the responsibility for the " Initial" classification of an emergency which occurs during a weekend on backshift? (0.5) ($gl U QUESTION 8.11 .- - aI AccordiMo technical specifications, what is the minimum Quadrnnt Power Tilt Ratio (QPTR) at 90% rated thermal power? (0.75)0 J .. -- ____ _ _ a. E At what power level is QPTR no longer applicable? (0.75) h g What protection does this limit provide? (0.75) l (***** END OF CATEGORY 08 *****) (************* END OF EXAMINATION ***************) ._ -- _ _ .

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

EQUATION SHEET ' - . . f = ma v = s/t Cycle efficiency = (Net work cut)/(Energy in)

w = mg s = V,t + 1/2 at

E = mc~ . XE = 1/2 mv a = (Vf - 1 )/t A = AN A = A e"*

g PE = mgn Vf = V, + at

= e/t x = tn2/t1/2 = 0.693/t1/2 l/2*ff C(tU)(t')]

t y, p D A= [(t)/2) + (t )3'

b cE = 931 sn - m = V,yAo-Ex , . . Q = mCoat I = I e~"* Q = UAa7 g I = I,l'O~* ##' Fwr = W sh f TIL = 1.3/u P = P 10 ""It) HVL = -0.693/u

g P = P e*# o SUR = 25.06/T SCR = 5/(I - K,ff) CR = S/(1 - K,ffx) x I CR (1 - X,ff)) = CR (1 - keff2 SUR = 25a/t* + (a - o)T j

T = (t*/s) + ((5 - oV Io] M = 1/(1 - X,ff) = CR)/CR3 7 = 1/(s - a) M = (1 - Xeffo)/(1 - X,ff 3) T = (a - s)/(Io) SDM = (1 - K,ff)/Keff a = (Keff-1)/X,ff = M,ff/K,ff t' = 10 seconcs I = 0.1 seconds ~I a = ((t*/(T X,ff)] + (T,ff (1 + IT)] / ll*Id Id 2,2 2 P = (I4V)/(3 x 1010) Id gd j

2 I = sN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) , Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lem.

I curie = 3.7 x 1010 dos 1 ga]. = 3.78 liters 1kg=2.21lbm}Stu/hr I hp = 2.54 x 10 1 ft4 = 7.48 gal.

Oensity = 62.4 lbT/ft3 1 mw = 3.41 x 100 5tu/hr Density = 1 gm/c9 lin = 2.54 cm Heat of vaporization = 970 Stu/lem

F = 9/5'C + 32 Heat of fusion a 144 Stu/lem

'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.

1 BTU = 778 ft-lbf l ft. H O = 0.4335 lbf/in.

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80 12 0 160 200 228 240 ROD BANK POSITION STEPS WITHORAWN INTEGRAL ROD WORTH VS. STEPS WITHDRAWN BOL, HZP, CONTROL BANK C Figure 5-4 S N P-SIM - 4 REV-1 Pace 13

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80 12 0 160 200 228 240 l ROD BANK POSITION STEPS WITHDRAWN INTEGR AL ROD WORTH VS. STEPS WITHDRAWN BOL, HZP, CONTROL BANK D Figure 5-5 SNP-SIM-5 REV-I Page 14

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SO gig 80 0 POEPER (% OF FULL POWER DEFECT VS POWER,AT BOL AND EOL, CYCLE I SNP-RF-13 Figure 5-.12 REV-2 Page 22 . . . - , _ _.

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8 . ~. e ~ , ~ ~ , - a, ~, , , , XENON WORTH (PCM) . XENON INTEGRAL WORTH CURVES (FOLLOWING PLANT STARTUP AT BOL) SN P - R F - 15 REV-2 Page 34 Figure 5-13 d .-- _._. -.,.. -- . - -.. . _ , . -, _, _ - - _ _ ._- - _., -. - _ -,, - -. -,. -. _ -.,, -

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2, - . . (nod) AllAllov3M NONEX XENON REACTIVITY D,EFECT AFTER SHUTDOWN ' FROM INDICATED POWER - SNP-RF-16 REV-2 Figure 5-14 Page 35 . _ _ _ -. -.--7 -,. _. _ - _,. _ _ _ _. ., -._- _, - ,,. ..,.--,-,y,., ..-- -

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a

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No Te so ?> ,20 To soo

7 '*7 ) Fl O W R A ~I~ C. h 1 l Figure 5-15 i i ! ._.

_. --

I -. . . .- . UNACCEPTABLE OPER ATION "1" 660- \\ \\ "5" \\ ,, a., ,,3 " \\ . % 640-w , % % 620- ' @ N N / - 600- , s

- %

' ,, 2 N ' ' seo- - , / ' f f ACCEPTABLE OPERATION I

560-I l ? o to 4o

so too izo ,,, PERCENT POWER FIGURE SNP-OPS-25: REACTOR CORE SAFETY LIMITS vs. BOUNDARY OF PROTECTION (REY. 1) ,

e ) I l 1 - - . . _ -.. _. - .. _ _-._. __ _

T ' h.

THEORY _OF NUCLEAR _ POWER PLANT OPERATION. FLUIDS. AND PAGE

T_HERMODYNAMICS , . ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

i ANSWER 5.01 (3.75)

NOTE: Values from graphs should be +/- 10% a.

ROD WORTHS S/D Banks A-E + Control Banks - Bank D (228-180) = Total 3085 PCM [0.1] + 2788.6 PCM [0.1] - 22.5 PCM [0.1] = 5851.1 PCM [0.3] POWER DEFECT = -1565 PCM-[0.1] RESULTS: (-5851.1) + 1565 = -4286.1 PCM [0.25] reactivity after trip.

Keff =

1 = = 0.9589 [0.25] 1 - reac.

1 - (-0.042861) (1.2) b.

BORON CHANGE due to critial rod position difference.

-SB& PCM Bank D @ 125 steps [.1] - 00-5 PCM [0.1] Bank D 9 180 steps =

/34~ -399 PCM to be diluted out [0.25] =/af XENON CHANGE

550 I'CM [0. 25]

2000 I'CM C equil.[h CM sfter 40 heas = 2"' ^ A"um 'Y"I

3cropcm o

POWER DEFECT +1565 PCM [0.1] RESULTS 73w t e,r W 376o +1565 PCM +' 550 PCM - 307 PCM = -1608. PCM net change [0.25] 3 76 o aihr 5 T 3 74. c, + PCM_ = +160 ppm to be added [0.25]. Therefore 10 PCM/ ppm gSS:e3 pg, o

,; 7 6 900 ppm + 18a ppm = -1980 ppm [0.25] (1.55)

c.

Beta 0.007 REAC.

= = =.00136 or (lambda * period) + 1 (0.08 * 52) N +1 REAC.

= 136 PCM [0.5]

~: /275 __ Therefore with Bank D 9 125 steps which equals -145 PCM remaining; _ i all ",ai D re'ir =ust Lu cc;;12tely withi aw.. B. S L (1.0) , Trt'.s no+ poss;bic. cc.s') REFERENCE SNUPPS, Phase 1, Volume A, Chapter 6, Page 6-19 and 6-32; Chapter 7, Pages 7-12 through 7-22 and page 7-51; Phase I, Volume B, Chapter 3, Page 3-41 and Chapter 4, Page 4-11; Chapters 5, 6 and

. .- _ _ _ _ _ _ _ _ _ _ ._ '5.

THE_QRY OF NUCLEAR _PQWER PLANT OPERATION. FLUI_DS. AND PAGE

TI_IERMODYNAMICS . . I ANSWERS -- SNUPPS-2-86/01/14-PlCKER, B.

SNUPPS Figures SNP-SIM-5 thru 6; SNP-SIM-11 thru 15; SNP-RF-13, 15, 16 ~ ANSWER 5.02 (2.00) (When a rod is stuck out with all other rods inserted,) that rod is exposed to a much higher flux than the flux in the rest of the core.

(Because rod worth is a function of the relative flux difference between that adjacent to the rod and the core average, the rod could be worth about 1000 pcm, which is much more than normal.)

(1.0) (If a rod is dropped, while the rest of the rods remain out, the opposite to the above happens.)

The flux is depressed adjacent to the dropped rod relative to the flux in the rest of the core and (so the dropped rod could be worth about 200 pcm).

(1.0) REFERENCE SNUPPS Phase 1, Volume B, Chapter 6, Page 6-15 b '

ANSWER 5.03 (2.00) '

}#2= 333,3 Y=

- = CC# (0.5) M -GR40/CR NO~,: 5000 1 ',0 = a.

f b.

Calculate ff after rod pull no. 10 W h d u b,l, r-EW? CB*%-M / > keff 90/1 - keff 10 M W C d r rt d-7 g g " NQ ' ~ < gLM C,X c, ~ 20 = 1 - 0.94/1 - keff 10 ' L'_ W h ~ ~ .,. I 1 - keff 10 = 1 - 0.94/20 = 0.06/20 = 0.003 A: - I - krA Keff Q10 = 1 - 0.003 = 0.997 b di = /-/n (1.0) <.997 Since Keff %10 is less than 1, the reactor is suboritical.

(0.5) REFERENCE o SNUPPS Phase 1. Volume A, Chapter 8, Page 8-16, hfacfear {chr-Plat (sies, cly 8, m 6 &# N k y klnb w b e

r 5.

ILIEQBY _OF NUCLEAR POWER __ PLANT OPERATION. FLUIDSn AND PAGE

' THERMODYNAMICS , ' ' ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

M ole. : auEt**M *

a.

The main steam pressure will be higher [0.25]. The turbine control valve will assume a new position in the closed direction thus increasing the resistance to steam flow, creating a higher back pressure which causes the main steam pressure to rise [0.5]. (0.75) b.. The main steam flow rate, Ms, will be lower [0.25]. Closing down on control valve will reduce steam flow [0.25]. (0.5) c.

Ts will be higher [0.25]. Since we are operating under saturated conditions as steam pressure rises the steam temperature must also increase [0.25] (0.5) d.

Tavg is higher [0.25]. Tave must be higher to add negative reactivity to help balance out the positive reactivity added ~ due to a lower Xenon concentration at the new power level and pos. reactivity from the Doppler Coeff[0.5]. (0.75) REFERENCE SNUPPS Thermal, Volume II, Chapter 12 . ANSWER 5.05 (1.50) a.

A local increase in the neutron flux means that heat is produced at a higher rate than is provided in the reactor accident analysis.

(There is a possibility of localized damage resulting from some of the fuel elements becoming overheated.)

(0.5) b.

High.

(0.5) c.

The higher negative moderator temperature coefficient of reactivity at EOL would more effectively counteract the reactivity changes due to changes in Xenon concentration.

(0.5) REFERENCE SNUPPS Phase I, Volume B, Chapter 4, Page 4-28

- . _. - - - _ _ _ _ S.

THEORY OF NUCLEAR _ POWER PLANT OPERATIONmEWIDS. AND PAGE

^ THERMODYNAMICS , AN$WERS -- SNUPPS-2-86/01/14-PICKER, B.

ANSWER 5.06 (2.00) / / / 7.7 'it26 A [1.0] From Steam Tables: 196.27 + (

9 0.1)

= - - 196.27 960.1 g, / 960.1 X" = 96rS% steam or M moisture. [1.0] 95 9'l*4 OR 4.83 */- If values are taken from Mollier Diagram 4% is acceptable.

(2.0) REFERENCE ~ Volume II, Chapter 10, p.

10-71; Westinghouse, Thermal Science, ASME steam tables ANSWER 5.07 (2.00) a.

A pressure relief valve.

(0.5) b.

By adding the flow rates from pump 2 and 3 together and plotting the resultant curve at various head pressures-a total flow of 62 gpm [0.5] is found at 82 psig [0.S]. Whereas pumN will not contribute any flow because the system head pressure is above its shut-off head [0.S]. (1.5) REFERENCE Westinghouse, Thermal Science, Volume II, Chapter 10 ANSWER 5.08 (1.50) Thermal Flux at BOL is LOWER _(or HIGHER at EOL) (0.75) p-(RR=Macroscopicabsorp. infuel*ThermalFlux) -{ 0. 5 ) since the amount of fuel decreases the thermal flux must increas to maintain a constant RR.

O,70)( G. 5 )- REFERENCE SNUPPS Phase I, Vol. A, Chapter 4, pp 31-36

_ _ _ _ "- 5.

THEQBY OE_HUCLEAB_PQWER_ELANT_OPEBATIQN mFLUIDSu6ND PAGis

T,HERMODYNAM_ICS , ~" ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

ANSWER 5.09 (3.00) a.

SYSTEM PARAMETERS 1.

Primary coolant temperatures.

2.

Primary system pressure.

3.

Primary coolant flow rate.

4.

Reactor power level.

[0.5 each] (2.0) b.

There is a 95% confidence that 95% of the fuel rods have not . t-exceeded the heat flux resulting in DNB.( pnM qd cer - (1.0) , M M *Moj k isu ew prob 4c ) REFERENCE Westinghouse, Thermal Science, Vol. II, Chapter 13, Page 13-23 ANSWER 5.10 (1.00) a.

Convection b.

Radiation / convection (large Delta T) c.

Conduction , d.

Convection (natural) [0.25 each] (1.0) REFERENCE WTHP Chapter 13 SNUPPS Thermal Science, Volume I, Chapter 3, pp 8, 36, 59-71, 75-90 ANSWER 5.11 (2.25) a.

Fission gasses pollute the Helium gas causing a reduction in gap thermal conductivity [0.5]. This results in increased fuel temp-erature change for a given power change, causing an increase in the.agnitude of the coefficient [0.25]. (0.75) b.

Fuel densification causes a decrease in the fuel pellet dimension resulting in an increase in the fuel to clad gap dimension and higher fuel temperature [0.5]. This causes an increase in the magnitude of the coefficient [0.25]. (0.75) c.

Clad creep effectively shrinks the clad into closer contact with the fuel, increasing the gap thermal conductivity [0.5]. This results in a fuel temperature decrease and a lower value for the coefficient [0.25]. (0.75) ! - - - _ _ _ _ _ -. . - _ __ - _ _.. _ _. -- - . - _ _.

5.

THEORY OF_ NUCLEAR _ POWER _ PLANT OPERATION, FLUIDSuMD PAGE

THERMODXNMICS . ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

' REFERENCE Reactor Control for Large PWRs, Chapter 2, p.

2-40 to 49.

ANSWER 5.12 (1.50) , a.

MTC through lower Tave.

> D 2 @ $0.5 each] (1.0) FTC through lowering & power.

Daxe v"Y k Ye

b.

Shorter life in subsequent fuel cycles.

Undesirable flux distribution in subsequent cycles [one req.] (0.5) REFERENCE SNUPPS Phase I, Vol. B, Chapter 2, p.

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

r

6.

PLANT __SYSIEMS_DESI_GNuCONTROh_6ND INSIBUtgNTATION PAGE

' ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

.., ANSWER 6.01 (1.00) The reason for these interlocks is to insure that the regenerative heat exchanger is always at reactor coolant system pressure to prevent steam flashing and possible damage to its tubes.

REFERENCE SNUPPS Phase II, Volume II, Chapter 1, Page 1-7 ANSWER 6.02 (1.00) Use the BTRS [0.5] because the quantity of radioactive wastes that must be reprocessed is reduced [0.5]. (1.0) REFERENCE SNUPPS Phase II, Volue II, Chapter 3, Page 3-3 ANSWER 6.03 (2.00) (ValveHV-8811)containmentsumptoRHRPumpAsuction, 1.

is open.

(0.5) (ValveHV-8812 RWST to RHR Pump A suction, is open.

(0.5) 2.

(ValveHV-8804)RHRPumpAdischargetothecentrifugalcharging 3.

pumps suctions and safety injection Pump A suction, is open.

(0.5) 4.

RCS pressure is greater than 383 psig.

(0.5) REFERENCE SNUPPS Phase II, Volume II, Chapter 4, Page 4-15 l

6.

PLANT SYSTEMS DESIGN uCONTROL uAND INSTRUMENTATION PAGE

' ANSNERS -- SNUPPS-2-86/01/14-PICKER, B.

... ANSWER 6.04 (3.00) 1.

P-4 (reactor trip interlock) [0.25] - shif ts -the Tavg mode from the load rejection functiop to the plant trip function [0 'E]. (0.75) Ace s+em Naps Co 7sj 2.

C-9 (condenser available interlock) [0.25] - Protects the condenser from overpressurization by blocking steam dump actuation during periods of insufficient vacuum [0.5]. (0.75) 3.

C-7 (loss of load interlock) [0.25] - arms the steam dumps for operation following a load rejection (of greater than 10 percent in120 seconds)[0.5]. (0.75) 4.

P-12 (low-low Tavg interlock) [0.25] - closes all dump valves when Tavgdecreases(to550F)preventinganuncontrolledcooldownfrom occurring [0.5]. (0.75) REFERENCE SNUPPS Phase II, Volume III, Chapter 4, Page 4-6

ANSWER 6.05 (2.00) S/ Glow-lowlevel(onMCBganelRLO18)CO Manual actuation ( C. 25 f2-a.

1.

y [D,2Y,0 32. 3 percent E4ktT on 2/4 2.

Trip of both m(airt feed pumps.[,2)gn any S/G Ihrig,TJ 6*d MP detectors LDA (h26T--. 3.

s I c.2) s. 'B/oc.i.

O S/G low-low lev]el [0.2] 32.3 percent [0.1] on 2/4 b.

1.

detectors [0.1] on 2/4 S/G's [0,1] (0.5) 2.

Manual actuation on MCB panel RLO18.

(0.25) 3.,Undervoltage on either 4.16 engineered safety feature pus (NB01 or NB02).

(0.25)

REFERENCE SNUPPS Phase II, Volume III, Chapter 5, Page 5-13 . . - - - - - -

- . _ - - - - - - ... ' 6.

PLANT _ SYSTEMS _ DESIGN. CONTROL. AND INSTRUMENTATI_Otl PAGE

- ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

.,o , ANSWER 6.06 (1.50) ' 1.

Feedwater. control valve wear at low feeding rates is reduced by allowing the valve to be further open.

(0.5) , 2.

Efficiency is improved by reducing pump power requirements.

(0.5) 3.

Feedwater control valves are maintained in a more linear operating range.

(0.5) i REFERENCE SNUPPS Phase II, Volume III, Chapter 6, Page 6-17 i

ANSWER 6.07 (1.50) ! 1.

(Two) wide range hot' leg temperature RTD's.

(0.25) 2.

(Two) wide range cold leg temperature RTD's.

(0.25) , 3.

(One) RCS wide range pressure transmitter.

(0.25) 4.

(Twenty-five) core thermocouples.

(0.25) 5.

(Three) thermocouple reference junction box temperature RTD's.

(0.25) 6.

(Two) par narrow range pressure transmitters.

(0.25) .- REFERENCE ! SNUPPS Phase II, Volume IV, Chapter 1, Page 1-15 ANSWER 6.08 (2.00) ' a.

Undercompensated.

(0.5) i b.

The compensating voltage cancels the effects of gamma activity within the detector such that the detector output current is representative of neutron activity only.

(1.0) c.

The operator should manually reenergize the source range instruments.

(0.5) REFERENCE l SNUPPS Phase II, Volume IV, Chpater 2, Pages 2-11, 2-22 and 2-24 !

, . . .. . . . .

_ _ ' 6.

PLANT SYSTEMS DESIGN,._CQNTRQhi_AND INST _RUMENTATION PAGE

AN$WERS -- SNUPPS-2-86/01/14-PICKER, B.

ANSWER 6.09 (1.00) This rod stop prevents outward rod motion in both the manual and automatic rod control [0.5]. It. limits the possiblity of exceeding the intermediate range high level and power range low level reactor tripsetpoints(of25 percent)priortotheirbeingblocked[0.5]. (1.0) REFERENCE SNUPPS Phase II, Volume IV, Chapter 2, Page 2-24 ANSWER 6.10 (3.00) Shutdown margin changes with reactor power.b*c"--A k'N a.

Te p eride cacepteblu modiel flu Atatributic='. [0. 5 c c c u } --. (1.0) Po * be.4 ec A. b.

A Steam Rupture accident.

i A Rod Ejection accident.

[0.5 each] (1.0) c.

The moderator temperature coefficient of reactivity becomes more necative over core life.

(1.0) REFERENCE SNUPPS Phase II, Volume IV, Chapter 3, Page 3-14 ANSWER 6.11 (2.50) a.

UV on Safeguards bus SIS or CSAS UV on Safeguards with SIS or CSAS [0.5 each] (1.5) b.

D/G output breaker shut [0.5] and preferred power supply breakers open [0.5]. (1.0) REFERENCE SNUPPS Phase II, Vol.

I, Chap.

3, pp 13-15 -.- . - _ . _. . .

-

6.

PLANT SYS_IEMS_DESIGHuCONTBQLuAND INST _BUMENTATION PAGE

ANS,WERS -- SNUPPS-2-86/01/14-PICKER, B.

,,, ANSWER 6.12 (2.00) 1.

D/G field flashing.

2.

Breaker control.

3.

MCB power 4.

Emergency lighting.

[0.5 each] (2.0) REFERENCE SNUPPS Phase II, Vol. I, Chap.

5, p.

co ANSWER 6.13 (2.50) crr 1.

.9P DELTA T 2.

S/G SAFETIES 3.

2235 ;;it!- c 4.

OP DELTA T . o 5.

NUCLEAR OVERPOWER [0.5 each] (2.5) REFERENCE SNUPPS Phase II, Vol. IV, Chap.

5, p.

5-5; FIG. SNP-OPS-25 l i i -. - - - - - - - - - - - - - - .-. .

7.

PROCEDURES - NOBMAL. ABNQBMALuBMERGENCY AND-PAGE

' - ,BADIOLOGICAL COliTBQL . ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

' ANSWER 7.01 (3.00) 1.

Verify reactor trip [0.3] Response not obtained - Manually trip reactor. [0.3] If reactor will not trip, then manually deenergize the rod drive MG sets (PG HIS-16 and PG HIS-18). [0.3] If MG sets will not deenergize, then manually insert control rods. [0.3] (1.2) 2.

Verify turbine trip. [0.3] Response not obtained - Manually trip turbine. [0.3) If turbine wi]1 not trip, then manually run back turbine. [0.3] If turbine cannot be run back, then close main steam line isolation and bypass valves. [0.3] (1.2) 3.

Check AFW pumps running.

(0.3) 4.

Initiate Emergency Boration.

(0.3) REFERENCE FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION /ATWS ANSWER 7.02 (1.50), we. Ecc.S ~PW Awem t least one charging pump % r SI pump running) [0.75] k 1. A4 2.

RCS pressure (trip parameter) less than 1320 psig [0.75]., (1.5) No 9oWM Sc.duchd Qo< ph w ke<O.

REFERENCE E-1, LOSS OF REACTOR OR SECONDARY COOLANT ANSWER 7.03 (2.25) 1.

Limit the release of Radioactive effluents.

2.

Restore Reactor Coolant inventory.

3.

Stop primary to secondary leakage.

[0.75 each] (2.25) REFERENCE E-3, STEAM GENERATOR TUBE RUPTURE, ERG Background E-3, p.

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

-

Iu_PBQQEDUBES - HQBMAL1_ASNQBMAht_EMERGENGLAND PAGE

BADIQLOGIQAL CONTROL , AN$WERS -- SNUPPS-2-86/01/14-PICKER, B.

' ' ' ANSWER 7.04 (2.25) 1.

PRZR PORV's - CLOSED (BB HIS-455A, 456B).

(0.75) 2.

Letdown isolation valves - CLOSED (BG HIS-459, 460).

(0.75) 3.

Excess letdown isolation valves - CLOSED (BG HIS-8153, 8154).

(0.75) REFERENCE ECA-0.0, LOSS OF ALL AC POWER ANSWER 7.05 (2.00) a.

SI actuation requires SI to be actuated, whereas reinitiation says to operate pumps as necessary.

(1.0) b.

Control of plant is lost and SI is necessary [0.5 each] (1.0) REFERENCE Westinghouse ERG Executive Manual, Generic Issues, FOLDOUT, p.

I ANSWER 7.06 (1.00) To prevent a steam line low pressure safety injection initiation.

(1.0) REFERENCE SNUPPS GEN-N-02, Page 5 l ANSWER 7.07 (1.25) He may request an NRC inspection.

REFERENCE SNUPPS Phase I, Volume C, Chapter 4, Page 4-9 ! _.

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

7.

PROQEDUREfi_- _NQRMAbuABNQRMAhugMERGENQY AND PAGE

BADIOLOGIQR_QQNTRQb

. ~ ~ ~ ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

ANSWER 7.08 (3.25) a.

He can receive an additional 2000 mrem.

(5(N-18) = 5(20-18) = 10 rem. max allowed. [0.5] 6 rem. + 3 rem. = 9 rem total which is < 10 rem. allowed [0.5] and he has 1000 mrer he may receive 3 - 1 = 2 rem. [0.75].) (1.75) b.

He may receive only 000 mrem [0.5]. 10 is max. since he 1.ceived 9, he may only receive 10 - 9 = 1 (1000 mrem) [1.0] '(1.5) REFERENCE SNUPPS Phase 1, Volume C, Chapter 4, Page 4-14 to CRL 7 o. l o t ANSWER 7.09 (3.00) a.

2.

b.

1.

c.

2.

d.

1.

[0.75 each] (3.0) REFERENCE SNUPPS Phase 1, Volume U, Chapter 4, Page 4-22 ANSWER 7.10 (1.50) a.

must sit idle for 30 minutes.

(0.75) b.

Damage to motor windings.

(0.75) REFERENCE Selected System Procedures, Volume 2, Procedure BB-N-02.1, p.

. - - - -

~~ 7.

EBQQEQQBES - NQBM6huABNQBtMLt_MENGLAND PAGE

BADIOLOGICAL CONTROL . ... . ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.

I ANSWER 7.11 (2.50) 1. d 2. b 3. c 4. e 5. a [0.5 each for correct order] (2.5) REFERENCE PI 1F-0 Background Info.

1-5 H P/ O -R w,f y / h' y [44 P "d4 $ Y Gavnclin fom. A *- E* & " . . . b)6sTtNGHeux 0.uW'c groY.P J ' F-o crrhed wha % S s 67CEE3 t ANSWER 7.12 (1.50) (Power changes of this magnitude cause large variations in fuel pin pressure.)

Any clad defects would enable fission gases to escape [0.75]. An iodine analysis would therefore show a fuel pin defect [0.75]. (high amount of fission products in the reactor coolant) (1.5) REFERENCE GEN-N-06; page_1 i _ _ _. - - _ _ _ - ~(*********CAF******** FOR FURTHER DETAILS ON BACKGROUND FOR PRECAUTIONl _, ( . ,4 - - - -, -,., -, - - ,

8.

ADMINISTBATIVE_PROCEDLJBES. CONDITIONS _AND_LIMLIAT_I_QtiS PAGE

2

ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.
.. . ANSWER 8.01 (1.50) 1.
The NRC operations center shall be notified within one hour.
2.
A safety limit violation report shall be prepared.
3.
The safety limit violation report shall be submitted to the commission, the NSRB, and the vice president - Nuclear within > 14 days of the violation.
, 4.
Critical operation of the unit shall not be resumed until authorized by the commission.
[3 of 4 required 9 0.5 each] (1.5).
REFERENCE T.S., Section 6, Administrative Controls, Page 6-16 ANSWER 8.02 (1.50) 1.
The intent of the original procedure is not. altered.
(0.5) 2.
The change is approved by two members of the plant management-staff, at least one of whom is the operating supervisor, holding an SRO license on the affected unit.
(0.b) 3.
The change is documented, reviewed by the ORC, and approved by the plant superintendent within 14 days of implementation.
(0.5) , REFERENCE T.S., Section 6, Administrative Controls, Page 6-17 ANSWER 8.03 (1.50) The maximum period of time that the crew shortage can exist is two hourn [0.75] provided immediate action is taken to restore ' the shiq crew composition to within the minimum requirements [0.75]. (1.5) REFERENCE T.S., Section 6, Administration, Page 6-5 , l I
- 8.
ADMI_NLS_T_BATIVE PRQCED_URESuGQN_QITIONS. AND LIMIT _&TJ.ONS PACE
ANSNERS -- SNUPPS-2-86/01/14-PICKER, B.
... ANSWER 8.04 (4.00) a.
Start back-counting the penalty minutes from 0400 on 8/20/85.
(0.5) At 40% power .(0400 - 0248) x 0.5 = 36 penalty min.
(0.5) This leaves (60 - 36) = 24 penalty minutes available.
(0.5)
At 60% power - (1833 - 24) - 1809 on 8/19/85 (0.5) The power can be increased at any time 24 hours after the above time, or after 1809 +1/-0 on 8/20/85.
(0.5)

b.
The limits on axial flux difference ensure that FQ(Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not

exceeded during either normal operation or in the event of xenon redistribution following power changes.
(1.5)
REFERENCE . Tech. Spec. 3/4.2,. POWER DISTRIBUTION LIMITS Basis for Tech. Spec. 3/4.2, AFD, Page B 3/4 2-1.
i ! ANSWER 8.05 (3.00) 1.
All penetrations that are required to be closed during accident conditions are EITHER capable of being closed by an OPERABLE containment isolation system, [0.5] OR closed by manual valves, , blind flanges or deactivated automatic valves secured in their

closed positions [0.5]. (1.0) ' 2.
All equipment hatches are closed and sealed.
(0.5) ' 3.
Each air lock is OPERABLE.
(0,5) ' 4.
The containment leakage rates are within the Tech. Spec. limits.
(0,5) 5.
The sealing mechanism associated with each penetration is OPERABLE.
(0.5) i ! REFERENCE

SNUPPS Phase II, Volume 1, Chapter 2, Page 2-6; SNUPPS Technical Specification Definitions, Page 1-2 l .

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

8.
ADMINISTRATIVE PROCEDUREL_gONDITIONS. AND LIMITATIONS PAGE
ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.
,. ANSWER 8.06 (2.25) a.
Provide core protection to prevent DNB.
b.
To prevent water relief through the pressurizer safety valves.
c.
Provides assurance of fuel integrity, limits the required range for overtemperature protection, and provides a backup to the high neutron flux trip.
[0.75 each] (2.25) l REFERENCE Bases for LIMITING SAFETY SYSTEM SETTINGS, Pages B2-5 thru B2-8 ANSWER 8.07 (1.50) Within one hour [0.).51 action shall be initiated to place the unit in Co.1 a HODE in whi~ch the specification does not apply [0.751 (1.5) Ct. ev) REFERENCE SNUPPS T.S., 3.0.3, p.
3/4 0-1

ANSWER 8.08 (2.00) a.
1.
Either a Keff of 0.95 or less, [0.5] or i 2.
A boron concentration of greater than 2000 ppe [0.5] (1.0) b.
As a minimum, two source range monitors (installed or portable) shall be operating [0.33], each with continuous visual indication in the control room [0.33] and one with audible indication in the containment and control room [0.34]. (1.0) REFERENCE T.S. 3/4.9.1, Boron Concentration, Page 3/4 9-1 T.S. 3/4.9.2, Instrumentation, Page 3/4 9-2 t i i . l - -,., -,, - -,,,, -,, -.. - - - -. - - - - - - - -, - - - - -..,,, -, ., - - , --
- 8e ADMINISTRATIVE _ PROCEDURES CONDITIONS. AND__LItf1TATIONS PAGE
& ' ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.
C ANSWER 8.09 (3.00) 1.
Initial classification and declaration of emergencies.
2.
Request technical and operational support.
3.
Initial coordination of on-site emergency actions / response.
4.
Notification of and communications with off-site organizations.
5.
Making protective action recommendations to off-site authorities.
6.
Initiating the activation of the other emergency organizations.
7.
Requesting local off-site support (i.e., firefighting, ambulance, law enforcement).
[6 of 7 required @ 0.5 each] .(3.0) REFERENCE Emergency Response Plan, Chapter 5, Page 5-8 ANSWER 8.10 (2.50) a.
Operating Supervisor [0.S], Shift Supervisor (Emergency Coordinator). [0.S] (1.0) b.
(Report the Control Room) and be a technical advisor to the S.S.
[0.15] Monitor plant instrumentation [0.3] and the Safety Parameter Display System (SPDS) [0.3] and provide recommendations to the SS concerning reactor safety. [0.25] (1.0) c.
Shif t Supervisor.
(0.5) REFERENCE SNUPPS Emergency Response Plan, Chapter 5, pages 5-7 to 5-12
-
_ _.__ _ 8_1.__AQULMLSIB&ILYE_E8QGEQUREGi._GQMQLT.LQHE _AHQ_LLMLIAILQNS PAGE
r
. , of ANSWERS -- SNUPPS-2-86/01/14-PICKER, B.
ANSWER 8.11 f t. Lib) d.ht k75.) i .. ' &, di.
50% or less (0.75) g c.
Radial power distribution (heat generation rate).
(G.75) REFERENCE TS; 3/4 2-15 and 2-17; B 3/4 2-5.
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.no .. R ATE. (G P M) Flo wFigure 5-15 ANSWER _ . _ _ _ _ . .. _ _ _ _ .. .. - . .
,T /Vir716 hine a f al

. ,, . SNUPPS-2 SRO EXAMINATION RESOLUTION OF FACILITY C0fGENTS EXAN DATE 1-14-86 TECHNICAL SECTION 5 Facility Comment: 5.01 a.
An alternate choice for shutdown rod PCM is from our shutdown margin calculation procedure which tells operator to use -4130 PCM.
(Attached) b.
Xe reactivity incorrect. The control rod position accounted for Xe equilibrium.
The only question for Xe is "what amount from last time Keff=1? That amount is +2300 PCM (fg 5-14) which equates to 230 ppm [CB] 1561 PCM [ Power Defect] + 2300 pcm [Xe] + 307 pcm [ Rod] = 4172 pcm = 417.2 ppm + 900 ppm = 1317.2 ppm.
c.
Misread figure 5-5.
Bank D P = 120-130 pcm at 125 steps instead of 135 pcm at 125 steps.
On the border line of being able to get sufficient pcm to get.5 DPM. Answer could easily be "not possible" as well as all the way out depending on graph reading.
NRC Resolution: 5.01 a.
Not accepted because the question stated reactivity valves were given in the attached figures.
b.
Accepted because calculated error was made in determining Xenon reactivity. Answer key was changed to accept 2300 PCM, but the facility's calculation shows the effect of rods as a positive value (+ 307 PRM) but should be negative.
Therefore, the result is: 1565 PCM [ power detect] + 2300 PCM [Xe] -307 PCM [ Rods]'= 3358 PCM = 355.8 PPM boron change = 355.8 + 900 PCM = 1255.8 PPM.
Key corrected.
c.
Accepted because the answer key rod worth is in error. Answer - key was changed to "not possible".
Facility Comment: 5.03 a.
Two different M (subcrit. multiplation)
CR M = 1-Keff Str_ M = CR0
M = 1-Keff = 333.3 See attached for references.
- - . . . . - - -
. . I

. ., . b. Changed by a above.
NRC Resolution: 5.03 a. Accepted because by using the correct theory approach the answer is: and b. 333.3 for M, in part a.
This approach will also give correct answer for part b.
The method used in the key was incorrect and was changed.
Partial Credit will be'given if base count rate is not taken as zero, because if not take as zero 1/M is no longer equal to 1-Keff but the 1/M plot will still accurately predict criticality.
Reference: Westinghouse fundamentals of Nuclear Reactor Physics, Chapter 8, Pages 63, 64.
Facility Comment:

5.04 Agree that the text in Chapter 12 states agreement.with answer key.
but Throughout Thermal Science the heat balance equations are used.
An alternate explanation should be accepted if heat transfer equations or principles are used.
a. Steam flow (m ) when turbine load reduced Q s/g = m ( h) s/g Because Qrx not changed Qs/g so h s/g Because h ins, (momenlarul) . constant h out s/G. Saturated system P s/g , T s/g Q rx - Q s/g = VA (Tavg - Tstm), if Tstm the Tavg must increase to keep Q s/g until Rx power changes.
NRC Resolution: 5.04 a. Will also accept use of heat transfer equations as noted in facility's comment.
Facility Comment: 5.06 The 1126.7 BUT/lbm h for 2250 psia is incorrect.
1126.7 is above 2250 psia.
The interpolation was done in wrong direction.
The h should be 1117.7 BTU /lbm which gives consistent results w/ Mollier diagram results.
.. .'e.' . e .. .. NRC Resolution: 5.06~ Accepted because answer key was incorrect and accepted 1117.7 Btu /lbm; final answer is.4.03%. Facility Comment: 5.09 I know the " horse" answer as stated for the reason 1.3 'is the limit for DNBR.
. But As states in Technical Specifications and Thermal Science Volume II, Chapter 13 page 70.
The ultimate reason is fuel.
overheat or fission product release.
NRC Resolution: 5.09 Accepted because either answer is correct.
Facility Comment: 5.10 b.
If water is present then convection is correct.
Radiation only if no water present (near' fuel rod) Both answers not required NRC' Resolution: 5.10 b.
Agree, because either answer is correct depending on assump-tions.
Facility Comment: 5.11 b.
The wording should be same as part a.
"... In an increase in the fuel clad gap dimension.
This results in an increased fue1~ temperature change for a given.
power change, causing an increase in the magnitude of the coefficient."
The. higher fuel temperature by itself decreases the coefficient.
It is the T/% that makes coefficient increases.
I c.
Same as above.
NRC Resolution: 5.11 b.
Agree, because the reasons have the same effect.
No change and c.
.to the key is necessary.
l a
. ! ' . - .. . Facility Comment:- 5.12 The referenced document does not specifically say MTC or FTC.
We agree! When referenced to other chapters though, Xenon will lower power, also should be accepted.
NRC Resolution: 5.12 Accepted and added Xenon to answer key.
TECHNICAL SECTION 6 6.041. P-4 can also arm steam dumps as another purpose (attached).
NRC Resolution: 6.04 Accepted the comment by adding " arming steam dumps" to key and re-distributed points on.part #1.
Reference: SNUPPS-2 Logic Diagram 7250064 sh. 10.
Facility Comment: -6.05 The referenced document is correct.
But Atttached LOGIC does not distinguish AFAS as the referenced document does. The logic acceps all starts signals as a start for auxiliary feed. This document is most frequently used document and should be accepted as an answer. Any signal as shown starts the designated pumps.
NRC Resolution: 6.05 Accepted because there ar,e two logic diagrams that support five answers.
Facility Comment: 6.10 a. The reference document makes statements about RIL, but I don't see how SDM and radial flux were picked out as the two bases for power correction.
The first paragraph says "...the system... provide a limit that is variable with power...the bases bhind RIL are..." Then list three bases which implies all are bases for the power corrected limit.
All these are acceptable answers.
. d' . . . . . . . . . . . . . . . . . .
. f

. .. . NRC Resolution: 6.10 a. Agree with comment. Answer key changed to accept for full credit: Shutdown margin changes with reactor power because of a change in power detect.
Facility Comment: 6.13 1. OT T vice 0P T (believe key is typo) (attached) 3. " Safety Limit" acceptable; 2235, although that is what reference states, it means nothing.
The line or envelope is a " safety limit".
NRC Resolution: 6.13 1. Accepted, and typographical error corrected.
6.13 3. Accepted, because the question did not solicit the desired answer.
Item 3 was-deleted from the exam and the point value for section ' six reduced to.24.5.
TECHNICAL SECTION 7 Facility Comment: 7.02 "...Two independent conditions..." This is very bad wording.
The answer key list 1 or 2.
The E0P states 1 and 2 which does not make 1 independent of 2.
They are dependent.
The student would seek another independent condition as " loss of CCW" or " Containment Isolated RB."
The student should not be marked incortect for second statement of trip.
NRC Resolutin: 7.02 Accepted. Answer key changed to require one ECCS pump running and RCS Press, less than 1320 psig. Also no points off for any other answer beyond those above.
i Facility Comment: . 7.08 10CFR20.101 says 5(N-18) only applies when wanting to exceed 1.25 per quarter.
In fact, the W reference, Phase 1 Volume C Page 4-14 (attached), says this in the example problem.
Therefore, the student would answer part b with 1.25 REM (1250 m rem).
- - - - - - - - -. - . - - - - - - - - - - .
. 'l

. .. . However-Becaure of experience level of some candidates, the answer would be the 1 REM per key. The last exam at P0TC when returned stated that Region I interprets 10CFR20 to'save 5(N-18) active at all times.
We haven't seen this in writing except on returned examination.
NRC Resolution: 7.08 Not accepted, because the ' question -clearly stated _ the conditions such that the interpretation of 10CFR20.101 by the facility is in-correct. Also the example problem referenced is wrong according to 10CFR20.101.
TECHNICAL SECTION 8 Facility Comment: 8.11 Question asks " minimum" QPTR T.S. states no minimum but states maximum values.
The " minimum" QPTR is 1.
Where numerator - denominator.
Interpret as "the lowest of the Technical Specification Limits" then the answer key is correct.
Either answer is acceptable.
NRC Resolution: 8.11 a. Accepted, because question, as worded, did not properly solicit the answer. This question was deleted and the point value for section 8 reduced to 24.25.
< l
, J }}

v t e Inspection Report - - 1986001
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1986 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle Security Operator Exams Emergency Planning Other

IR 05000572/1986001

Text

=-s

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