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Exam Rept 50-220/85-14 on 850910-12.Exam Results:All Candidates Passed.Util 851004 Comments & Recommendations to NRC Re Reactor & Senior Reactor Operator Exams on 850910 Encl
	ML20136J512
Person / Time
Site:	Nine Mile Point
Issue date:	12/31/1985
From:	Crescenzo F, Keller R, Kister H, Lange D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20136J510	List: IR 05000220/1985014; ML20136J505; ML20136J522;
References
	50-220-85-14, NUDOCS 8601130438
	Download: ML20136J512 (75)
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 85-14 OL FACILITY DOCKET NO. 50-220

. FACILITY LICENSE NO. OPR-63 LICENSEE: Niagara Mohawk 300 Erie Boulevard West Syracuse, New York 13202 FACILITY: Nine Mile Point, Unit 1 EXAMINATION DATES: September 10-12, 1985 CHIEF EXAMINER: hAD   u OS C Crescenzo, R tor Engineer Tsaminer) ,date m4['   /2 2't!f{
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Reviewed By: vu - D. Lange, Lfdd Re Engineer (Examiner) / date Reviewed By:

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R. Kellbr,~ Prir$e' cts 6ection 1C

     /Zl)/![Ib date Approved By: _

H(.Kister> Chief,ProjectsBranchN [ //fF

     ' dite SUMMARY:
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Operator Licensing examinations were conducted at Nine Mile Point Unit 1 during the period September 10-12, 1985. Two Senior Operator upgrade candidates and one Senior Operator Instant candidate were administered both written, oral and simulator examinations. One Instructor Certification Candidate and one Reactor Operator Candidate were administered written - examinations only. All candidates passed the examination '

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1 , l REPORT DETAILS ! , _ TYPE OF EXAMS: Replacement X i ' EXAM RESULTS: I R0 l SR0 l Inst Cert I _ l Pass / Fail l Pass / Fail l Pass / Fail l I I I l Written Exam I 1/0 l 3/0 l 1/0 l l l l l l l l l l I I Oral Exam I N/A I 3/0 l N/A l l l l l l l 1 1 I I I Simulator Exam l N/A I 3/0 I N/A l

I I I I I I I I I I l Overall I 1/0 1 3/0 1 1/0 l l l l l l CHIEF EXAMINER AT SITE: Frank Crescenzo OTHER EXAMINERS: Tom Morgan (EG&G Idaho) Personnel Present at Exit Interview: . NRC Personnel Frank Crescenzo

NRC Contractor Personnel i Tom Morgan EG&G Idaho Facility Personnel John Aldrich, Operations Superintendent, Unit 1 Rick Zo111tsch, Nuclear Training Superintendent Randy Stefried, Nuclear Training Assistant Superintedent Don Straka, Nuclear Training Supervisor, Unit 1 Summary of NRC Comments made at Exit Interview: I a) Given such a small number of candidates for operating examinations, i no generic strengths or deficiencies were note .

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b) It was noted that some miror discrepancies existed between the simulator and the plant. In general the simulator performed wel c) In some areas the training material provided to the examiners did not provide enough detail for the generation of non-superficial questions. This was anostly a concern with instrumentation and control system d) The training and plant staff were helpful and cooperative throughout tne examination perio . Summary of facility comments and commitments made at exit interview: a) The staff was aware of the simulator / plant discrepancies noted in 4.b and is planning to correct them in the futur b) The training material provided to the examiners is in some cases a guide for the instructors who cover much greater detail in the classroo Attachments: 1. Written Examination (s) and Answer Key (s) (SR0/R0) 2. Facility Comments on Written Examinations 3. NRC resolutions of Facility Comments on Written Examinations

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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION

FACILITY: _H1HE_ MILE _EQ1HI - REACTOR TYPE: _RWR-QE1 _____ ___ DATE ADMINISTERED:_&iLaRLla______--- EXAMINER: JWQRQAM&_I.____ _____ APPLICANT: . ____ _ IMEIRECT1Qka_IQ_&EELICANIl Use separate paper for 'the answers. Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each categor/ and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % or APPLICANT'S CATEGORY __YALUE_ _IQIAL ___ACORE___ _YALUE__ _______------__CAIEQQRI - _____ p 2, 5'O 23 % 11.99-_ _31299 ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, TLUIDS, AND THERMODYNAMICS pl. l1 _11.QQ__ _&i:22 ___________ ________ PLANT SYSTEMS DESION, CONTROL, AND INSTRUMENTATION

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_11.QQ__ _& trit ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL g3 cc 94. 0 $ _al.ER__ _4tzee ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 48,63 ita.aQ__ laa.QQ ___________ ________ TOTALS FINAL ORADE _________________% All work done on this examination is my own. I have neither , givon nor received aid, l I

___________________________________ l APPLICANT'S SIONATURE _

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QUESTION 5.01 (2.00) The reactor is exactly critical high in the source rang A control rod is withdrawn one notch.

st ~~~ a. Describe what happens to indicated neutron level and why? Continue your description until a steady state condition is reached. Assume no further operator action and no other parameters are change (1.0) b. Describe how reactor period would respond and why? (1.0) QUESTION 5.02 (3.00) With regard to moderator temperature coefficient answer the following questions: o. Per degree change in the moderator temperature. WHEN is MORE reectivity added, at 50 F or 200 F? Explain your choic (1.5) HOW and WHY does the core age affect the coefficient? (1.5) QUESTION 5.03 (2.00,) Fill in the blanks with one of the given choices in the paragraph below describing the inverse power response to rod movement "As a shallow rod is inserted during power operation, the tal_____ __ (INCREASED, DECREASED) void formation propagates all the way up to the top of the core, causing a reactivity (b]__________ (INCREASE, DECREASE) which more than offsets the reactivity (cl__________ (ADDITION, SUBTRACTION) due to the insertion of the control rod. The not effect is a small reactivity (d1__________ (INCREASE, DECREASC), resulting in a small to)__ __

 (INCREASE, DECREASE), in reactor power." (5 0 0.4 eal  (2.0)
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QUESTION .S.04 (2.00) Assume the reactor is operating at 100% power and TWO recirculation pumps tri Indicate how each listed indicated parameter would FIRST change (Increase or Decrease) and briefly explain WHY the change occur Reactor power , 1 Reactor water level Feedwater flow (2.0) QUESTION 5.05 (2.00) The Reactor has been operating at 95% power for several days. An operator RAPIDLY reduces reactor power to 60% by reducing the speed of the recirculation pumps. During the next 2-3 MINUTES the operator notices that the reactor power slowly increases approximately 3%. EXPLAIN the cause of this effec (2.0) QUESTION 5.06 (3.00) The reactor is subcritical with a Keff of .95 a SRM countrate of 200 cps. The control rods are withdrawn and the new countrate is 400 cps, How much reactivity was added? (2.0) What would be the status of the reector if the same amount of reactivity, determined in a.. was added again? (1.0) e /

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. _ /, o t-y ~ sj QUESTION 5.07 48-544 c. Exp ;in hai h ;;::: tr the 'u:!

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incte : ;;r!; in ;cre li 'i.0) 4. As i ii . expun ua w increases, one limit h; gin: te incr::ee .i a dec - F+e.ing rate and ihwn dwus...... Li.i ihnwe cnanges in the fuel that can==. t h e % .". 7 L;;G R iamii to change in tnas manne (1.5) QUESTION 5.08 (3.00) Indicate HOW each of the coefficients are effected (Increase. Decrease or Remain the samel by each of the three parameters listed? Consider each parameter separatel a. Rod Worth (delta K/K/ Bank) by: 1. Moderator temperature INCREASES Voids DECREASE 3. Fuel temperature INCREASES (3 0 0.33 eal Alpha Doppler (delta K/K/ F fuel) by: 1. Core age INCREASES 2. Fuel temperature DECREASES 3. Voids DECREASE (3 0 0.33 eal Alpha Voids (delta K/K/ % voids) by: 1. Fuel temperature INCREASES 2. Core age INCREASES 3. Control Rod Density INCREASES [3 0 0.33 eal QUESTION 5.09 (1.50) In the main condenser, circulating water flow rate is approximately 20 times that of the steam flow rate. Why are these flow rates different?

(Consider thermodynamic principles in your answer)   (1.5)

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QUESTION 5.10 (2.00) Assuming the reactor has been operating at full power for an extended period of time when a scram occur (Assume time in core life is EOC) During a restart eight hours after the scrams How will rod worth be affected? (overall core) (1.0) How will Radial and Axial flux distribution be affected? (1.0) QUESTION 5.11 (2.00) The reactor has been operating at 75% power for several days when power is increased to 100% power by recirculation flo With no further operator action, HOW and WHY will reactor power vary over the next several hours?

(Take your discuss to when reactivities have stabilized.) (2.0)

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A J LAMI_11&IEM1_DE11GE _QQMIRQL _&HQ_1MEIRUMERI&I1QE PAGE 6 l

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* QUESTION 6.01 (2.50)

Answer the following questions regarding the Main Steam Syste . WHAT are two (2) functions of the flow restrictors? (1.0) 2. What control and protection features use the output from the restrictors? (0.5) b. When in Startup at 750 psig, WHAT are the five (5) parameters, if exceeded, will cause the MSIV's to automatically isolate?

 [Setpoints not required)   (1.0)
' QUESTION 6.02 (3.00)

When a scram signal occurs at power, describe IN DETAIL how the Control Rod Drive and its associated Hydraulic Control Unit function to insert the control rod. As a MINIMUM in your answer include chich components open, close, energine, deenergize, and motive force for the entire rod trave QUESTION 6.03 (2.00) The three Core Spray isolation valves and the CS test valve are interlocked togethe Describe these 3 interlock (0.75) Describe the operation of the core spray sparger break detection syste Include in your answer WHERE pressure is physically sensed and WHAT delta pressures are sense (1.25) OUESTION 6.04 (2.50) For each of the IRM (Intermediate Range Monitoring) range changes listed below provide the following: 1. The indicated level on the new range and Any automatic actions initiated as a result of the indicated level on the new rang Switching from range setting 5, reading 25, up to range setting (1.0) Switching from range 6. reading 39, down to range (1.53 (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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* QUESTION 6.05 (2.00)

For each of the following, state whether a ROD BLOCK, HALF-SCRAM, FULL SCRAM, or NO PROTECTIVE ACTION is generated for that conditio NOTE: IF two or more actions are generated, i.e. rod block and a half-scram, state the most severe, i.e. half-scra Assume NO oper-otor actions, APRM 11 Downscale, Mode Switch in RUN (0.5) (4 LPRM inputs to APRM 15. Mode Switch in STARTUP (0.5) Both Flow Conv. Units Upscale (>107% flow), Mode Switch in RUN (0.5) APRM 12 and 16 Upscale, Mode Switch in STARTUP (0.5) QUESTION 6.06 (3.00) By referring to the Recirculation MG logic disgram (figure 1).

DESCRIBE in detail, what the 41bx relay does during the startin sequence for a recirculation pum (Assume all conditions necessary for starting are met.)

QUESTION 6.07 (2.75) a. What four (4) conditions are required inorder for the Primary ADS valves to actuate? (2.0) When at power, What are three (3) of the DIRECT indications a MSERV is full open? (0.75) QUESTION 6.08 (2.50) Answer the following questions about the Rod Worth Minimizer's control of rod movement, when the rod selected results in a select orro c. WHAT happens when the rod is withdrawn one notch? (0.5) Assuming the rod has been withdrawn one notch, as in (a.) above, HOW much further can the rod be withdrawn and WHY? (0.5) 1 Assuming the rod has been withdrawn to its maximum limit, l as in part (b.) above, WHAT happens when the rod is inserted ) and HOW far can it be inserted? (1.5) j (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)

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QUESTION 6.09 (2.50) c. An APRM inop alarm has just been received. WHAT are three (3) possible causes for this alarm? (1.5) b. WHAT other alarm (s) were received because of this APRM INOP condition? (1.0) QUESTION 6.10 (2.25) WHAT are the sources of makeup water to the Emergency Cooling Condenser? Include HOW long each will last and in WHAT order they are use (Assume the reactor had been operating at full power for an extended period of time.)

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QUESTION 7.01 (3.00) A loss of Reactor Building Closed Loop Cooling has occurre WHAT are the required immediate operator action steps, per N1-SP-22 Loss _of RBCLC procedure?

(If a seperate procedure is to be performed concurrently, include its immediate action steps.)

qp (Assume RBCLC can not be regained.) (3.0) QUESTION 7.02 (2.50) Answer the following questions regarding the malfunction of the Control Rod Drive' System, Procedure N1-SP-1 c. For the stuck rod and uncoupled rod procedures a caution states,

Do NOT scram the control rod .... . WHAT is the reason for not wanting to scram the control rod?   (0.5) For a stuck rod which has failed to respond to the normal withdrawal command and the reactor is not at normal pressure-3> and temperatur WHAT i(,the maximum the drive water pressure can be adjusted to?    (0.5)

c. When an overheating condition for a CRD exists WHY does it state NOT to attempt to correct CRD temperature alarms by applying repeated drive signals. (0.5) During a ROD DRIFT incident WHEN must a manual scram be initiated? (0.5) o. At WHAT CRD instrument air pressure, as indicated in the in the control room, MUST a manual scram be initiated? (0.5) QUESTION 7.03 (3.00) With the reactor operating at rated power conditions, an inadvertent opening of one (1) Solenold-Actuated Pressure Relief Valve occurs resulting in a blowdown to the toru No scram trip settings are oxceeded as a result of this transien WHAT are the immediate operator action steps as required by the procedure N1-SOP-9?

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QUESTION 7.04 (2.00) According to N1-OP-46. Normal Operation of HPCI System: a. Explain the PURPOSE and the BASES of the following Caution statemen The AC motor operated discharge valve associated with an electrically driven feedwater pump should be in the open position, if that pump is NOT locked out, and available to run. (1.0) If initiation of HPCI was due to turbine or low level scram, WHEN can HPCI be manually overriden or shutdown? (1.0) QUESTION 7.05 (2.50) While performing the Heating and Pressurization portion of the Startup Procedure N1-OP-43, at WHAT pressure are the following performed? c. Start SJAE Open Main Stop Valve #2 slowly to warm valve chest c. Close reactor head vent Start one feedwater pump e. Put #11 or #12 cteanup pump in service [S e 0.5 ea) (2.5) QUESTION 7.06 (2.50) Several procedural steps in the railure of Reactor to Scram procedure N1-SOP-32, state If procedure symptoms still exist... perform that ste WHAT are the five (5) procedure symptoms? f

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QUESTION 7.07 (1.50) When may a "short pre-startup checkoff system" (Form 1) checkoff list be used? C.75) During approach to criticality, when is one notch-step with-drawal required? (.75) QUESTION 7.08 (3.00) 4 What are the five plant conditions which require starting the second reactor water clean-up (RWCU) pump? (1.25) Why is the second RWCU pump started during these conditions? (.75) If the cleanup system is to be returned to service following a low-low level vessel isolation, then Radiochemistry Sampling Procedure N1-SP-12 must be performed with satisfactory results prior to system startu Under what conditions may this require-ment be waived? (1.0) QUESTION 7.09 (2.50) Answer the following questions regarding the shutdown procedure of one recirculation pump with the reactor critical, full pressure and temperature and some power level less than 90%. o. WHY is the bypass valve checked to be fully open on the affected loop? (0.5) WHAT automatic action is expected when the discharge valve is fully closed? (1.0) The last step of the procedure states Crack open pump discharge

valve .... HOW is this step accomplished and WHY is it required? (1.0)

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e QUERTION 7.10 (2.50) In the start up procedure for testing of the Emergency Diesel Generators there is a caution which states At extended light load operation " souping" can be expected to occur with any Diesel Generator. WHAT does the term " Souping" refer to? (0.5) WHY do you want to minimize " Souping"? (0.5) c. WHAT limitations are imposed on the Diesel Generator operations i hrder to minimize " Souping"? (1.5)

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' QUESTION 8.01 (3.00)

The Radiation Protection Procedures describes areas that must be poste Define each of the areas listed below, Restricted Area (1.0) Radiation Area (1.0) Airborne Radioactivity Area (1.0)

QUESTION- 8.02 (3.50) ' What conditions must be met to establish Reactor Building l Integrity? (2.03 When must Reactor Building Integrity be in effect? (1.5) " QUESTION 8.03 (2.00) What are the control room MINIMUM staffing requirements during: 7 Normal Operation, C1.0) 7 Reactor Startu (1.0) i j QUESTION 8.04 (3.50)

.What action (s) must be taken if it is necessary to defeat an annunciator alarm under conditions where there ISN'T an approved procedure? (1.0) List four circumstances where placement of electrical jumpers, changing or removal of leads and the blocking of relays may be performe (2.0) How is it-possible to distinguish between a " control room jumper" and a "special jumper" on the Jumper / Block Log? (0.5)-

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* QUESTION 8.05 (3.00)

With regard given to the Radiation Work Permits, answer the idllokings o. WHAT is the purpose of RWPs? (0.S) b. WHAT six (6) conditions require a RWP? (1.5) c. WHEN is an extended RWP issued and WHAT three (3) approval signatures (by title) must be obtained? (1.0) QUESTION 8.06 (3.00) The following data was taken during a single day of operation: IDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE SHIFT 1 11,700 GAL 2,100 GAL SHIFT 2 12,000 GAL 2,800 GAL SHIFT 3 12.200 GAL 2,200 GAL Were the three (3) Technical Specification limits for leakage violated? JUSTIFY YOUR ANSWER (3 e 1.0 ea) (3.0)

QUESTION 8.07 (2.S0) Unit 1 is at 100% powe Below is the work history of a Reactor Operator from Unit 1 for the past 19 days: Days 1-S : 8 hours / day Day 6 : 6 hours Day 7 : 18 hours Days 8-12: 8 hours / day Day 13 : 15 hours Day 14 : 10 hours Day 15-19: 8 hours / day (NOTE: All work started at 0000 of each day and was consecutive.)

In accordance with Admin Procedure, AP-4 Sec 8.0, Overtime Procedure for Station Personnel. WHAT violations to the guidelines occurred during this work period?

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* QUESTION 8.08 (2.50)

Explain HOW the two (2) mark-up and the hold-out tags are utilize .(Include when each of the three (3) tags are used and for what perpose.)

QUESTION 8.09 (2.00) g}

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ng a startup per_N1-OP-43, IRM 12 and RWM are bypassed to imporp eration, after the reactor is critica mum manning requirement me When preparing to go t it is discovered that APRM 14 is g erradic indica and is bypasse After bypassing the APRM the c

   ' tor ator places the mode switch in RU Have the Tech  Specifications been va  tod?

If YES, cribe which items were violated an required action If , describe which items were considered and WH (2.0)

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- 1 r I C ; From Master M/A o Tech, M 1 E 2 l g l A L h M/A 888I30# (TYPi cal of Five)

41 bx ll a iI Positions Scoop Tube Start 2 to 20% when discharge 1 Dual BV not open 33 Dual Limit DV Limiter h

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Run g 41 bx Start g g,,9 Function

Contacts disc val e

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        ' Scoop  Secop E/P  Tube + Tube Positioner

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EQUATION SHEET

f = ma y = s/t Cycle efficiency = (Net work

out)/(Energy in)

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

[=m     .

KE = 1/2 mv a = (Vf - Vo )/t A = AN A=Ae' g PE = agn Vf = V, + at w = e/t A = &n2/tjjg = 0.693/t1/2 y - 2 t 1/2 8# = U tmM A= n04

 , 3p
     [(el /2) * (*b))

aE = 931 sn -

      .-1:x m = V,yAo  ,
    , ,

Q = mCpat Q = UA I = fo e'# Pwr = Wfah I=I n 10-*/IY' TVL = 1.3/u P = P 10 sur(t) HVL = -0.693/n p = p et lT o SUR = 26.06/T SCR = S/(1 - K,ff) CR x = 5/(1 - K,ffx)

SUR = 25e/1* + (a - o)T CRj (1 - K ,ffj) = CR 2 II ~ "eff2)

. T = ( t*/ s ) + ((a - o '/ Io] ~

M = 1/(i - K,ff) = CR)/G, T = 1/(o - a) M = (1 - K,ffa)/(1 - K,ff)) T = (a - o)/(Io) SOM = (1 - K ,ff)/K ,ff a = (K ,ff-1)/K ,ff = 4K ,ff/K,ff t" = 10# seconos I = 0.1 seconds ~I l o = ((L*/(T K,ff)] + (I,ff/ (1 + If)] Ij dj = I d 2 ,2 7d

P = (r4V)/(3 x 1010) Idjj 22 2 I = sN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) , , Water parameters Miscellaneous Conversions 1 gal. = 8.345 le curie = 3.7 x 1010 dos 1 ga]. = 3.78 liters 1 kg = 2.21 lem 1 ftJ = 7.48 ga I np = 2.54 x 103 Stu/hr Density = 62.4 lbg/ft3 1 mw = 3.41 x 100 Btu /hr Density = 1 gm/cm3 lin = 2.54 cm Heat of vaporization = 970 Stu/lem *F = 9/5'C + 32 Heat of fusion = 144 Stu/lbm 'C = 5/9 ( *F-32) 1 Atm = 14.7 psi = 29.9 in. H STU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.

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- L__IEEQELQLEHQLEAE_EQER_EL&EI_QEERAT.lOH&_EkulDim_&MQ PAGE 16 THEEMQQ1MEMLQ1 ,- ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, WSTERC:PY ANSWER 5.01 (2.00) a. Neutron level would start and continue to increase until the point of adding heat is reached.'. As the coolant heats up, negative react-ivity is added and power turn Power would stabilise.at the point of adding hea (1.0) Period would take a prompt jump due to the production of prompt neutronsiUImmediately after the prompt jump, the rate of power change decreases to a rate controlled by delayed neutrons until the reactivity is no longer being increased. Then a prompt drop would occur as the rate of reactivity addition drops to zero. A stable period would continue until the negative reactivity is inserted by temperature of the coolant increasin Stabilize at infinit (1.0) REFERENCE NMP1 Reactor Theory Module i part 10, 11, & 12 ANSWER 5.02 (3.00) F (0.53 The moderator density change per degree T, at the higher temperature, is greater [1.0 (1.5) As core age increases alpha T decreases, (less negative) [0.5 Control rods are withdrawn to compensate for fuel burnup (long term rod withdrawal). The Moderator to fuel ratio increases such that the plant is less undermoderated (1.0 (1.5) e REFERENCE NMP1 Reactor Theory Module 1 part 12 ANSWER 5.03 (2.00) Decrease Increas Subtractio Increas Increas [S e 0.4 eal (2.0) REFERENCE NMP1 Reactor Theory Module 1 part 14

l - 1-_ THEQR1_QE_MuQLEAR_EQWER_ELARI_QEER&I1QH _EkH1QEm_&MQ PAGE 17 ) IEEEMQQ1 MEW 1QE j ,. ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, . ANSWER 5.04 (2.00) A. Decrease (0.25) due to increased void content in the core as flow decreases (0.25). (0.5) B. Increase (0.25) due to increased voiding in the core (0.25) and recirc pump no longer taking suction on the annulus (0.25). (0.75) c. Decrease'(0.25) due to steam flow decrease (0.25) and level increase (0.25). (0.75) REFERENCE NMP1, Simulator Malfunctions Cause and Effects RR01 and RR03 ANSWER 5.05 (2.00) The reactor is now producing less steam to go to the turbine. There will be less extraction steam and reheater drain steam going to the feedwater heater.(1.0) Therefore less feedwater heating will occur resulting in colder feedwater entering the vessel (.5) which will cause reactor power to increase about 3% from the positive reactivity addition (alpha m).t.5) (2.0) REFERENCE NMP1, Simulator Malfunctions Causes and Effects MS10

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- 1- IEEQEl_QE_MuQLE&E_EQWER_ELAMI_QEER&I1QE._ELu1QE._&MQ PAGE 18 IEERMQQIE&M1QE .. ' ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. . ANSWER 5.06 (3.00) CR1 (1-Keff1) = CR2 (1-Keff2) [0.75) 200 (1 .95) = 400 (1-Keff2) 200 (1 .95)/400 -1 = -Keff2

 .975 = -Keff2  (0.251 delta p = Keff2-1/Keff2 - Keff1-1/Keffi (0.753 delta p = .975-1/.975 - .95-1/.95 delta p = ( .0256) - ( .0526)

delta p = .027 (0.251 (2.0) Part b. will be graded independently of part delta p = Keff3-1/Keff3 - Keff2-1/Keff2 (0.75)

 .027 = Keff3-l/Keff3 - .975-1/.975
 .027 = 1-1/Keff3 - ( .0256)
 .0014 = l-1/Keff3
 .9986 = -1/Keff3
 .9986 Keff3 = -1  10.25]

Keff3 = 1.0014 super critical (will accept critical) (1.0) REFERENCE NMP1 Reactor Theory module 1 part 8

l . : )F ANSWER 5.07 M ,445 !ri'imity at low exposure e developing in t (0.25 As power d, thes p aaa ratchet open and eventually bring the pell c . th the clad [0.251. This increases the a ility to transfer heat I . the limit is o increase because less heat is stored in the fuel Lu.A51-44-El , BT Fission gas build- a

   ::u inv ...: e==ad_st{ess on the cla (0.5) Reduce transfer rate due to fission g_   (0.5) al peaking factor decrease .5)

REFERENCE MMP1 Thermodynamics, Heat Transfer and Fluid Flow Module X

. 1- THEQEl_QE_MEQLEaE_EQWER_ELAMI_QEER&I1OM _ ELE 1QEm_&HQ PAGE 19 IHERMQQ1M& MICE .. ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. . ANSWER 5.08 (3.00) n.1. increase c.2. increase c.3. remains the same , b.1. increase b.2. Increase b.3. decrease c.1. increase c.2. decrease c.3. increase (9 0 0.33 eal REFERENCE NMP1 Reactor Theory, Module 1, part 12, 13. & 14 ANSWER 5.09 (1.50) Circulating water is maintained subcooled while the steam undergoes o change in phase.t0.5) The heat removal required to condense the steam (i.e. latent heat of condensation) accounts for the large difference in flow rates.[1.01 (1.5) REFERENCE NMP1 Heat Transfer and Fluid Flow Module IX part 2 & 3 ANSWER 5.10 (2.00) Increased worth of the peripheral rods and decreased worth of those on the core cente (1.0) Axial flux distribution could be severly top peake Radial flux distribution will peak in the peripheral region and be substantially lower in the core cente (1.0) REFERENCE Nine Mile Point Reactor Theory Module I part 14 __

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, ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. .

ANSWER 5.11 (2.00) Reactor power would increase above 100% because of the removal rate (by burnout) exceeds the production rate of xenon (0.53. These rates are different because iodine production increase immediately, while xenon production increases only after iodine starts to decay [0.5 When todine starts decaying and the production of xenon is increased reactor power will decrease and continue to decrease until oquilibrium xenon is reached (0.51. Power will be less than 100% chen stabill ed [0.5 (2.0) REFERENCE NMP1 Reactor Theory Module i Chapter 16 and figure 16-6

- L__ELARI_11ETEME_DEElGE._QQMIRQL&_ARQ_lMETEUMEMIAllQR PAGE 21 3NSWERS -- NINE MILE POINT -85/09/10-MORGAN, * .

ANSWER 6.01 (2.50) c. 1. The restrictors protect the fuel barrier by limiting the loss of water from the reactor vessel before the MSIV closure (0.251 in case of a main steam line rupture outside the primary containment (0.25 The restrictors also serve as flow elements for the main steam flow instrumentatton (0.51 ,g (1.0) 2.,,T)9_ J e s t r i c t o r s instrumentation is used in the pr.=:r; e n.;iasent

!Ir!:~tien 10.251 and reactor water level control system (0.25)  (0.5) . Reactor vessel low-low water level (5 ")

2. Main steam line break (105 psid)(120% flow) 3. Area high temperature (200 F) 4. High Radiation in the main steam line tunnel (5XNormal) 5. Lo-lo-lo Condenser Vacuum (7 ") [5 0 0.2 eal (1.0) REFERENCE NMP1 Simulator System Manual, Main Steam, Ch 21, pg 2 & 6 ANSWER 6.03 (3.00) A scram signal deenergizes the scram pilot valves [0.51, venting air from the scram inlet and outlet valves, allowing them to opent0.5 This vents water from the overpiston area of the CRD to the SDVt0.51 and applies HCU accumulator water to the underpiston area of the CRDt0.5 This dp provides the initial motive force for the rod [0.5 As accumulator pressure drops below reactor pressure, a ball check valve in the CRD opens to apply reactor pressure to the CRD to complete the scram stroket0.5 (3.0) REFERENCE NMP1 Simulator Systems Manual, CRDH, Ch Sa, pg 6 & 7 ._ __ . _ _ _ . _ _ - . _ _ _ _ _ _

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1-_Ek&MI_111 TEM 1_REELGE._CQEIRQL._&MR_1MEIRUMEMIAHQE PAGE 22 ANbWERS -- NINE MILE POINT -85/09/10-MORGAN, >

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ANSWER 6.03 (2.00) . The inside and outside inlet isolation valves must be closed before the test valve can be opene . The inside inlet valves and the test valve must be closed be-fore the outside inlet valve can be opene . The outside inlet and test valve must be closed before the inside inlet valves can be opene l (3 0 0.25 eal (0.75) Differential pressure is sensed between the core spray injection line (0.251 and the instrumentation pressure tap which measures above core plate pressure (0.25 A break in the CS piping outside the throud would cause the dp to increase due to the added pressure 1 l drop 10.251 across the steam seperators (0.2G1 and steam dryers (0.25 l (Outer pipe of SBLC injection line acceptable for low side) (1.25)

' REFERENCE

; NMP1, Simulator System Manual, Core Spray, Ch 17 pg 3 and figure 17-2 ANSWER 6.04  (2.50) .5 on range 7 (0.51 No automatic action, downscale at 2% [0.5 (5% of full scale)    (1.0) on range 5 10.51 IRM high rod block (0.51 and IRM high-high half scram (0.5 (1.5)

REFERENCE , NMP1, Simulator System Manual, IRM, Ch 9b, pg 6 & 11 . ANSWER 6.05 (2.00) ,

rod block or half-scram if concurrent hi IRM assumed half-scram

. rod block i full scram     to G 0.5 eel  (2.0)

i I REFERENCE NMP1, Simulator System Manual, APRM, Ch 9d, pg 1 & 5 of June 1985 rev ! between pages 4 and 10

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1- ELARI_EIETEME_REE1QEm_QQEIRQL _hEQ_LEEIRUMEEIAI1QE PAGE 23 j i ANSWERS -- NINE MILE POINT -85/09/10-MORGA .

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ANSWER 6.06 (3.00) By placing the control switch in the start position remote contacts close to energize a relay (41bx) and energize a time delay rela Energining the (41bx) relay blocks the generator tacometer speed signal [0.751 and signal from the dual limiter into inserts.the(40-50% the blind controllert0:5 speed After the time delay relay times out, the generator field breaker closes [01' This opens a contact which causes relay (41bx) to be deenergizedt'0<51)) The (41bx) relay than removes the dual limiter #1 from the starting circuitt0.'51 and substitutes dual limiter 02 with a 20% signal to the blind controller [0l5 (3.0) REFERENCE NMP1, Simulator Systems Manual, Ch 4, pg 4-10, and figure 4-7 ANSWER 6.07 (2.75) . Power is available to,the Core Spray Pumps and Ads Logic,W / / Lo-Lo-Lo Level of -10 inches High Drywell Pressure of 3.5 psig see timer has timed out (4 0 0.5 eal (2.0) . Open light Alarm Discharge piping high temp alarm 4 Acoustic monitor alarm (3 of 4 0 0.25 eal (0.75) REFERENCE NMP1, Simulator System Manual. ADS, Ch 15, pg 1, 4, & 6 ANSWER 6.08 (2.50) A withdrawal error and withdrawal block occu (0,5) Zero notches because of the withdrawal block that is impose (0.5) The first notch in clears the withdrawal errofand block (0.5 The next notch causes an insert error (0.51. The rod can be driven in to the 00 position (0.5 (1.5) REFERENCE NMP1, Simulator System Manual. RWM, Ch 6, pg 11 & 12 ,

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- 1- E.k&EI_11EIl2iE_QE11GMmOQRIRQL._ANQ_1MEIRUMEMIAI1QR PAGE 24 ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, * .

ANSWER 6.09 (2.50) . Less than four operable LPRM inputs 2. Channel switch not in operate 3. Module unplugge (3 0 0.5 eal (1.5) b. 1. 1/2 scram alarm . 2. Rod withdrawal block alarm (2 0 0.5 eal (1.0) ev UMA~ REFERENCE NMP1, Simulator System Manual, APRM, Ch 9d, pg 4 & 5 ANSWER 6.10 (2.25) 1. Elevated (40,000 gal) make-up storage tanks supply continuous cooling for eight hour . (200,000 gal) Condensate storage tanks supply continuous cooling for approximately "" '-"-- 4 0 k* - 3. Raw water (via fire pumps), indefinite

  [ order 0.25 ea, supply 0.25 ea, cooling time 0.25 eal (3 0 0.75 eal (2.25)

REFERENCE NMP1, Simulator System Ma.nual. Emergency Cooling System, Ch 14, pg 2 __ _ _ - _ _ _ _ . . -

- L P.RQQEQEREE - EQEM&km_&REQRMAL&_EMERGEEC1_&EQ PAGE 25 R&D1QLQQlO&L_CQEIRQL ,. ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. . ANSWER 7.01 (3.00) 1. Try to restart RBCLC pump (0.25) Initiate a manual reactor scram a. Shift mode selector switch to refuel b. Check all rods fully inserted c. Observe power level decreasing d. Check auto bus transfer e. Verify generator has tripped f. Monitor Reactor vessel water level g. Fully insert IRM and SRM detectors and follow power down as it decays of !? e 0.25 ea) (1.75) h. Control reactor pressure and remove decay heat (step h is the same as steps 4.5, & 6) 3. Manually trip the reactor recirculation pumps (0.25) 4. Place-emergency condensers in service (0.25) 5. Close Main Steam isolation Valves (0.25) 6. Start idle control rod drive pump (0.25) REFERENCE NMP1, Loss of RBCLC and Reactor Scram procedur N1-SP-16 pg 2 & 3 N1-SP-22 pg 2 & 3 ANSWER 7.02 (2.50) Further mechanical damage may occu (0,5) Reactor pressure plus 400 psi (0.5) c. This will only temporarily cool the drive with the effect of putting undesirable temperature cycles on the CRD (0.5) d. If three or more control rods simultaneously drift (0.5) o. 60 psig (0.5) REFERENCE NMP1, CRD Malfunction N1-SOP-15, pg 2,3,4,7,9 &10 _ _ _ _ _ - - _ . - _ ._ _ _ __ _ _ _ _ __ _ _ _ _ _ _ .

1.__EEQQEQUEER_ _KQEMAL _&EEQEM&L&_EMERGEMQ1_&MQ PAGE 26 l R&QLQLQQ1 CAL _CQEIRQL l l ,. ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. ' I l l ANSWER 7.03 (3.00) 1. Verify Automatic station response (0.51 2. Verify that reactor vessel water level (0.251 and reactor pressure remain in a near normal range (0.25 ; 3. Leaving the control swith in the auto position (0.25), each reset button should be depressed to close the affected valve [0.25 . Check computer printout (0.t253({p Valve Monitoring System panel (0125] to confirm and identify open relief va l ve .[p.ss ] If the reset fails to close the affected valve, pull the fuses for the relief valve (0.5 . The Containment Spray System shall be placed in service in the suppression chamber water cooling mode (0.5 (3.0) REFERENCE NMP1, Solenoid Actuated Pressure Relief Valve Opening, N1-SOP-9, pg 3 ANSWER 7.04 (2.00) o. To insure a discharge path is available to the feedwater pump [ 0 .15 1 because with limited off-site power the valves cannot be opened as power board 151 would be de-energized ( 0 ,'5 3 . (1.0) When all redundant or corroborating instrumentation are checked, or there is sufficient operational or instrumented evidence to show that the system is not performing its intended function 10.51 and/or continued operation will prolong or produce an unsafe condition (0.5 (1.0) REFERENCE NMP1, HPCI Operating Procedure, N1-OP-46, pg 12 ANSWER 7.05 (2.50) psig psig psig psig go o. 60-100 psig (Will except +/- 18%1 (5 0 0.5 eal (2.5)

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L EEQQEQUEEE - EQRMAL _ARMQRMAL _EMERGEMQY._&MQ PAGE 27 EAQ1QLQQ1Q&L_QQMIRQL .- ANSWERS -- NINE HILE'P0ilNT -85/09/10-MORGAN. . REFERENCE NMP1, Startup PIocedure, N1-OP-43, pg 7,8,9,10 & 13

._

ANSWER 7.06 (2.50) Procedure symptoms: This procedure applies if, The reactor has failed to scram (0.31, or if following a reactor ' scram, one (1) of the following conditions exist: Two (2) or more adjacent control rods fail to insert below 06 position (0.5) OR Thirty (30) or more control rods fail to insert below the 06 position (0.5) OR The resultant rod pattern resembles one of the patterns detailed on the core map diagrams in the procedure (0.S).

OR Reactor power is increasing as indicated by neutron instrumentation (0.S *2.5) REFERENCE NMP1, Failure of Reactor to Scram, N1-SOP-32, pg 2 ANSWER 7.07 (1.50) If a startup is within 24 hours of shutdown (scram) (.75)

(Statement concerning maintenance acceptable as part of answer) In the power regions of the fuel (i.e. notches 00-30) (.75)

REFERENCE NMP1, Startup and Shutdown, N1-OP-43, pg. 2 & S

1

- 11 ERQCEDERE1_ _RQRE km_AREQRMAL _EMERGEECY._&EQ PAGE 28 RAQLQLQQLCAL_COMIRQL ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, T.

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ANSWER 7.08 (3.00) Reactor startup, Shutdown, Scram, Power change, or rod swa (1.25) Minimise radistion level buildup in reactor recirculation piping by removing iv. soluble materials before they plate-ou Also helps reduce the magnitude of the thermal cycling at the feedwater nossles during periods of low feedwater flo (Either acceptable)(.75) Waived by SSS if he has determined, by evaluation of station instrumentation and operating conditions, that no core uncovering has occurred during the vessel isolation even (1.0) REFERENCE NMP1, Reactor Cleanup System. N1-OP-3, pg. 4 ANSWER 7.09 (2.50) c. This will prevent premature tripping of the pump as the discharge valve is furt er d (Q.5 m (0.5) b. At about 20% _; rt)ngc ogs

  && the discharge valve (0.51, speed will automatically drop to 20% [0.5).     (1.0)

c. The valve is given approximately a 15 see open signal (0.51 this will allow valve disc and stem to warm up (0.5 (1.0) REFERENCE NMP1, NSSS, N1-OP-1, pg 15 ANSWER 7.10 (2.50) c. The term " Souping" refers to thM accumulation of tube oil in the engine exhaust system @ue to light load operatio (0.5) b. Depending upon the amount of " Souping" that has taken place, an exhaust fire could result when the engine is suddenly loade (0.5) c. After 4.5 hours of operation for any load less than 40% [0.5) or after 8 hours of operation at idle speed (440-560 rpms) [0.5 The diesel engine must be run at or above 40% of rated load for a minimum of 30 minutes and until inspection shows that the exhaust stack is clean [0.5 (1.5)

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' 1.__E1QCEDUREE - EQRM&Lm&REQEM&LmDERGEEC1_&EQ PAGE 29 RAQ1QLQQ1 CAL _CQEIRQL .. ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, l

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REFERENCE NMP1, Emergency Diesel Generators, N1-OP-45, pg 12

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- R.- ANLEl&IRAIRE_P_EQQEQQEEEmQQERLILQEEdEQ_LIMLIAILQEE PAGE 30 ANSWERS -- NINE MILE POINT -85/09/10-MORGAN. T.

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ANSWER 8.01 (3.00) Any area where access is controlled for the purposes of protec-tion of individuals from exposure to radiation and radioactive materials. (Limits from 10CFR20 pt.105b acceptable) (1.0) Area, accessible to personnel, where there is radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 5 mr, or in 5 consecutive days a dose in excess of 100 m (1.0) Any area in which airborne radioactive materials exist in concen-trations in excess of 25% of values listed in 10CFR2 (1.0) REFERENCE NMP1, Access and Radiological, S-RP, pg 4 & 6 ANSWER 8.02 (3.50) The reactor building is closed and the following conditions are met: (0.5) At least one door in each access opening is closed (0.5) SBOT system is operable (0.5) All reactor bull. ding ventilation auto isolation valves are o pe r a b l e @]) a r e secured in the closed position (0.5) Must be in effect in refueling [0.51 and power operating conditions (0.51 and whenever irradiated fuel or the fuel cask is being handled in the buildingt0.5 (1.5) REFERENCE MMP1, Technical Specifications 1.12, pg 4, 3.4.0, pg 165

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i L * * 1MLEIRAI1YE_EEQQEQU.EEEmQQEnlIlQEE. AND LIMLIAllQEE PAGE 31 ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, T.

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ANSWER 8.03 (2.00) Licensed SRO (.25) 2 Licensed RO (.25) 2 Unlicessed Operators (.25) 1 STA (.25) Licensed SRO - C.25) 3 Licensed RO C.25) 2 Unlicensed Operators (.25) 1 STA (.25) REFERENCE NMP1, Technical Specifications Figure, 6.2.1, pg 250 ANSWER 8.04 (3.50) The SSS shall write an explanation on the annunciator relay log over his signatur A copy of annun lator log shall be attached to a temporary procedure change notic Original copy goes to Station Superintendent on first buisness day!- A copy is attached to the procedure or jumper / block lo Enteries made in both SSS and Control Room Log Book Also the defeated annunciator will be identified by a marker attached to the annunciator windo (1.0) . In accordance with approved maintenance procedures In accordance with approved test or surveillance procedures In accordance with approved modifications To facilitate the conduct of tests and checks To preserve the safety, function, and/or integrity of the , station or system l For non-routine activities during refueling outage (4 0 0.5 ea) ( .0) l 1 Serial numbers on special jumpers are distinct from those on l

control room jumper (0.5) REFERENCE NMP1, Control of Equipment, Placement of Jumpers or Blocks or Lifting of Leads, AP-3.3.2, pg 1, 2, 4 & 5

. -R ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LLWLI&lLQME PAGE 32 ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, T.

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ANSWER 8.05 (3.00) c. To control activities in radiological environment (0.5) . Contamination levels >10,000 dpm/100 cm2 2. Airborne radioactivity requiring respiratory equipment 3. Neutron radiation exposure 4. Use of vacuum cleaners or portable HEPA units in restricted areas 5. High radiation ~ area entries 6. Unknown condition in area or system being entered 7. Maintance in radiation or high radiation area [6 0 0.251 (1.5) For certain-routine or repetitive work (0.25) Station Superintendent, Radiation Protection Supervisor and an appropriate member of supervision of the group to perform the wor (30 0.25 eal (0.75) REFERENCE NMP1, Radiation Work Permit Procedure, S-RP-2, pg i & 14 ANSWER 8.06 (3.00) LIMIT: Unidentified Leakage 5 GPM X.60 min /hr X 24 hr/ day = 7,200 gal / day (0.51 Unidentified increase of Leakage 2 GPM X 60 min /hr X8 hr/ shift = 960 gal / shift (0.51 Total (Identified plus Unidentified) Leakage 25 GPM X 60 min /hr X~24 hr/ day = 36,000 gal / day 10.51 (1.5) ACTUAL: Unidentified leakage was 7,100 gal / day [0.333 Maximum unidentified increase was 700 gal / shift (0.331 Total leakage was 43,000 gal / day [0.341 (1.0)

.The total leakage limit was exceeded.'   (0.5)

W A'% Mq %M (L ' N W) REFERENCE NMP1, Technical Specification 3.2.5.a. pg 89

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. 8._ ADM1MLEIR&TLYE_ERQQEQQREEm_QQEQ1IlQME _&MQ_LIMLI&IlQE PAGE 33 ANSWERS -- NINE MILE POINT -85/09/10-MORGAN, T.

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ANSWER 8.07 (2.50) 1. Days 7 & 8 the operator worked more than 24 hours in a 48 hour perio (Days 7& 8 the operator had less than 8 hours off between work period '. Day 7 the operator worked mor than 16 hours in a 24 hour period 4. Days 7-13 the operator worked more than 72 hours in a 7 day perio . Days 13 & 14 the operator worked more than 24 hours in a 48 hour perio [5 9 0.5 eal (2.5) REFERENCE NMP1, Administration of Operations, AP-4, pg 10 and Technical Specifications, 6.2.2.H. pg 247 ANSWER 8.08 (2.50) The red and blue mark-up tags are used to insure positive control over the system or components that are the subject of the maintance and/or testing. (0.51 The red tags allows no operation of the equipment. [0.51 The blue tag allows the controller of the mark-up to operate the oquipment. [0.51 The yellow hold out tag is used on equipment or system which are not tagged for maintance or testing but is operable according to standing operating procedures er upon which, for any reason, temporary special limitations on operations have been placed. [1.01 (2.5)

[[ Red and Blue tag usage needs to be verified at the facility.11 REFERENCE NMP1, Control of. Equipment Mark-Ups, AP-3.3.1, pg 2 ANSWER 8.09 C 2 . 0 0 1-

0.51 Tech Spec 3.6.2.a table a (and g), min er of-downsea trips channels is not oth the startup cnd run modes ( .6.2.a ( es the co ods to be inserted [0.51

. . .a (7) no control rods withdra     (2.0)

REFERENCE NMP1, .6.2.a pg 188, 193, 217 and 3.1.1.3.b pg 29 Simulator Systems Manual Ch 9d, figure 9d-6

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U. S. NUCLEAR REGULATORY COMMISSION-REACTOR OPERATOR LICENSE EXAMINATION FACILITY: NINE MILE POINT _________________________ REACTOR TYPE: BWR-GE2 _________________________ ' DATE ADMINISTERED: 85/09/09 _________________________ EXAMINER: HOWE, APPLICANT: _ __ __ __ _ INSTRUCTIONS TO APPLICANT: __________________________ Use separate paper for the answers. Write answers on one side onl Stople-question sheet on top of the answer sheet 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 start % OF L' CATEGORY  % OF APPLICANT'S CATEGORY VALUE  TOTAL  SCORE VALUE    CATEGORY

________ ______ ___________ ________ ___________________________________

-25.00  25.00

________ ______ ___________ ________ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW _ I __._ _1 ___________ ________ PLANT. DESIGN INCLUDING SAFETY AND EMERGENCY. SYSTEMS ' 23.00 25.00 INSTRUMENTS AND CONTROLS 25.00 25.00 PROCEDURES - NORMAL, ABNORMAL, ________ ______ ___________ ________ 4 EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ___________ ________ FINAL GRADE _________________%

:All work done on-this examination is my own. I have neither given nor received'si ~~~~~~~~~~~~~~

EPPL5CEUYI5 555UdiURE ! I

O

.. .? PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,  PAGE 2
~~~~~  ~ ~

T55R566 555C5~~55dT TRI 5EER d 6~EL656 _______________________________________"EL6E _____

GUESTION' 1.01 (1.50) Censider.the core delayed neutron fraction a. What is the change in the fraction over core life?

-     (0.50)
;b. What is-the reason for the change?   (0.50)

c.-What effect does'the change in this fraction have on reactor response to reactivity changes? (0.50) GUESTION 1.02> (2.25)

.o.- 'How^does feedwater heating improve the efficiency of the power plant?    (1.0) If the highest pressure feed heater'is removed from service (extraction steam isolated), whet happens to kilowatt output of the generator and why?   (1.25)

QUESTION 1.03- (1.50)

.How'does the doppler coefficient of reactivity change with an increase 1in core void fraction? Why?   (1.50)

QUESTION 1.04 (2.50) A. Give three (3) advantages to allowing coolant to boil in-the core (1.5) B.:Why'is-operation beyond the point of Onset of Transition Boiling avoided ? (1.0)

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T ERE66 dU5C5,~E5dT TRdU5EER d 6~ FLU 56~FLUU ____________________________________________ QUESTION 1.05 (3.00) Censider.two control rods. Both rods are at notch position 1 Rcd A is located near the center of the core and rod B is located at the core edg a. After a long_ shutdown the reactor is started up and brought to 50%-power. To add'the greatest reactivity for a one notch withdrawl, which rod would you choose and why? (1.50)- b. The. reactor scrams after operating at high power for a long time and has now been started up and raised to 50% powe ,To add the most reactivity for a one notch withdrawl, which rod would you choose and why? (1.50) GUESTION 1.06 (2.50) For the power history below, sketch a curve of core xenon concentration versus tim Assume xenon concentration starts at 50% equilibriu PERCENT 100 - REACTOR 50' POWER 0- . . . . , . . . . . . . . . . . . O 20 40 60 90 100 120 140 160 180 TIME IN HOURS DUESTION 1.07 (2.00) A variable speed centrifugal pump is running at 1800 RPM with a capacity of 670 GPh and requires 70 HP while producing a head of 47 Ft-Lbm/Lbf. The pump speed is increased to raise the capacity to 1340 GPh. What is the new-pump speed, head, and brake HP ? (2.00)

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0UESTION- 1.08 (1 50)

Given: LReactor_ pressure at time t= 399.4 psis Reactor-pressure at time t+ 1 hr= 980.5 psis a.-What is the heatup rate? (Show calculations) (1.00) b. Is this rate acceptable for your plant? (0.50) .. QUESTI0d '1;09 (2.00) D3 fine condensate depression and briefly explain why EXCESSIVE

. condensate : depression is undesirable in the main condenser .
      (2.00)

GUESTION' 1 10 (2.25) Nine Mile 1 Point. Unit.1 is operated within three (3) specified Thermal Limits. List each'of-the three limits and explain the specific heat trans-

'far related problem that the limit protects agains (2.25)

GUESTION L 1.11 (2.00) Your reactor.has just scrammed from extended full power operatio Ten'(10) hours 2ater cooldown is complete, and the SDh is determined to_be-1% dk/k, since all rods did not inser EXPLAIN the changes to theLSDH AND any possible adverse co.nsequences for the next 20 hours. (2.00)

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1. - PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 ________________________________________ w________RMODYNAMICS, THE HEAT TRANSFER AND FLUID FLOW ___..________________________________________

. QUESTION- 1.12  '(2.00)

For the following events listed below , identify which reactivity coef-ficient would first cause power to change. Indicate in your answer the direction of changer (more or less negative), _for each coefficien .: Reactor feed' pump tri . A single safety relief valve lift . Turbine stop valves close , (no scram).

. 4. Control rod drop acciden (0.50 for each cor rect ans.)

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'0UESTION 2.01 (2.00) Two' sets.of vacuum breaker, valves are provided on the primary containment. -Briefly describe the relief path of each vaccum breaker. set and give the relief setpoint (1.50) Why are these vacuum breakers required?  (0.50)

QUESTION 2.02 (1.00) The Standby Liquid Control System (SLC) has a minimum required storage tank temperature of 65 F. How is solution temperature maintained end why? GUESTION 2.03 (2.50) List five (5) of the seven (7) signals that will cause the main steam isolation valves to shut. Also give the setpoints and bypasse (2.50) GUESTION 2.04 (1.50) Concerning the Control Rod Drive Hydraulic system a. With reactor pressure at 500 psig, what should be CRD drive water pressure? (0.50) b. How is the CRD drive water pressure maintained when driving in a control rod? (1.00) QUESTION 2.05 (2.00) Consider the Automatic Depressurization System (ADS): a. What are the ADS system power supplies? (Note different parts of the system may have different power sources.) (1.00) b. In addition to ADS Electromatic Relief valve position, how would an oper,a_ tor determine which ADS valve is lifting? (1.00)

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QUESTION 2.06 (1.50) Ragarding the RPS backup scram valves: 0. What is their function? .(1.00) b. What is their power supply? (0.25) c. Are.they energized.or'de-energi=ed during a scram? (0.25) QUESTION 2.07 (2.50)

-In regard to the Core Spray System o. What is the rated capacity of the Core Spray pumps ?  (0.5)

b. What. signals will cause an automatic initiation of the system? INCLUDE.SETPOINTS (1.0) What-is the purpose of having parallel isolation valves installed in the system? (1.0) 00ESTION 2.08 (3.00) Answer.the following with regard to the Emergency Cooling System a.-What are two sources of makeup to the Emergency Condenser Hakeup Tank? (0.5) What are the initiation signals for the system? Include setpoint (1.0) c. How may tne automatic initiation feature be overridden?

 .
      (.5)

d. After system initiation and pressure is < 1000 psis, How is the cooldown rate controlled? (1.0)

QUESTION 2.09 (3.00) Hcw is the integrity of ECCS piping inside the reactor vessel verified during normal operation. In your answer includei SENSING POINTS, SPECIFIC

' SYSTEM (S) WHOSE PIPING IS VERIFIED, WHY IT IS VERIFIED and the response
~ of the instrumentation to a loss of integrit (3.0)
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-GUESTION 2.10 (2.50)

Concerning Refueling Operations; c. List four(4) methods available to verify proper fuel bundle orientatio (1.0) b. Consider the REFUEL INTERLOCK alarm located on the ROD BLOCK MONITOR PANEL. List two (2) conditions, including interlocks, that this alarm could be indicating? (1.0) Under normal operations, prior to fuel handling , Procedure N1-OP-34 Refueling Procedure has a prerequisite which states, 'The Fuel Pool key lock switch on the 'G' panel shall be placed to the Refuel position when handling fuel or irradiated fuel casks.' What is the purpose of doing this? (0.5) GUESTION 2.11 (1.50) What are three (3)nses for the Recirculation MG Set Tachometer-Generator output? (1.5) 0UESTION 2.12 (2.00) Concerning the Generator Stator Cooling Water Systemi a. What three (3) conditions will cause a Turbine Governor Runback 5 (Setpoints are required) (0.75) b. . Will an automatic Reactor Scram occur on a Governor runback trip signal? If yes, from what? If not, how could a subsequent scram be prevented? (0.5) What is the importance of regulating flow within this system to maintain pressure between 22-28 psi? (0.75)

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QUESTION 3.01 (1.00) With regard to the Reactor Recirculation ~ Pumps during pump startup the MG set motor breakers will not close unless certain valves and relays are in their required startup positions. For the list given

'balow, provide the required positions / condition (1.00)

1). suction valve 2) . discharge valve 3) discharge bypass valve 4).86 relays GUESTION 3.02 (1.50) Consider a total loss of instrument air, how will the following valves fail? (1.50) o. MSIV's b. Feedwater FCV-c. Atmosphere to Torus vacuum reliefs d. Make up valve to condenser e. Scram inlet and outlet valves f. . Emergency condenser condensate return valve GUESTION 3.03 (2.25) Concerning the Hi/Lo'Lo/Lo Rosemont vessel level instruments: a. What is their indicating range and where is their instrument zero relative to TAF (Top of Active Fuel)? (1.00) b. List five trip functions which occur at 5' level on these instrument (1.25) GUESTION 3.04 (3.00) ConsirJer the RWM (Rod Worth Minimi er): o. What is the purpose of the system? (Include in your answer, the accident it is designed to mitigate.) (1.00) b. List three systems which provide input signals to the RW (1.00) c. What must occur for the RWM to generate an insert block? (1.00)

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QUESTION 3.05 (2.00) Concerning the IRMs o.~What are the power. supplies to the system? (0.50)- b What is the setpoint of the IRM scram and when is it automatically bypassed? (1.00) c..What two conditions will bypass all rod blocks from an IRM

 . channel?    (0.50)

00ESTION 3.06 (1.50) A caution in N1-OP-1 states that " ... at least two recirculation loop cuttion valves and their associated discharge valves will be in the full open position except when the reactor is flooded to above the main steam no :les ...'. Enplain the basis for this caution statemen (1.50) GUESTION 3.07 (3.00) What trips are associated with the APRM flow units ? (Include in your answer any setpoints, what operational constraints occur as a result of each trip, and what problem (s) each trip could indicate) DUESTION 3.08 (1.00) iE What is the reason for providing local alarms and local indication on some area radiation monitors (ARMS)? (1.0) QUESTION 3.09 (3.00) Concerning the recirculation pump seal assembly, describe what indications / alarms would be seen in the control room given the following system failure a+ Failure of no. 1 seal onl (1.00) b. Failure of no. 2 seal onl (1.00) c. Failure of both seal (1.00) .

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QUESTION 3.10 (2.00) Indicate wether the following statements concerning the control circuitry fer the emergency cooling system are TRUE or FALS (0.50 each) c. While attempting to close the DC powered Steam Isolation Valve, 60 ft-lbs of torque is applied before the valve is fully closed. The closing circo.t will deenergize and the valve will remain mid positione b. Loss of RPS Bus 11 will NOT cause condensate return valve til to ope c. Following an auto initiation, once the initiating parameter has cleared the condensate return valve will remain open and system operation will continu d. While attempting to open the AC powered Steam Isolation Valver 60 ft-lbs of torque is applied when the valve is mid positione The opening circuit will deenergize and the valve will remain mid-positione QUESTION 3.11 (3.00) A. Will a normal transfer of an RPS Bus fron. 4ts normal to its emergency power supply cause any pr otective action? WHY or WHY NOT? (1.0) B. Will a shift of a REACTOR TRIP Bus from its normal to its emergency power supply cause any protective action? WHY or WHY NOT? (1.0) C. Which five (5) motor generator sets at NMP may be driven by AC or DC motors? (1.0) QUESTION 3.12 (1.25) For the following valves associated with the PRIMARY CONTAINMENT / COOLANT ISOLATION SYSTEh, give the fail position on a loss of motive power or control signa Reactor water cleanup leaving reactor Shutdown cooling leaving reactor Main steam warmup Core spray pump suction Containment spray system drywell and suppression chamber common supply (0.25 each)

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QUESTION 13 13 ( .50)

Answer.the following'TRUE or' FALSE: 0."The-control switch for the recirculation pump cooling (RBCLC) wate isolation valve IV-70-92 has three positions OPEN, AUTO, and CLOS There is'NO control function associated with the AUTO position of this switc (0.25) b.s At' low RBCLC system heat loads, service water flow is regulated to maintain proper RBCLC system temperatur (0.25)

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____________________ QUESTION 4 01 (2.00) Roferring to procedure N1-SOP-15 when a control rod is stuck during pewer operations and cannot be freed, what actions should be taken to prevent further rod movement and why? Can the operator attempt to ccram the rod? (2.00) QUESTION 4.02 (3.00) According to Procedure N1-SOP-3,Feedwater Halfunction(Decreasing FW Flow); o. What immediate actions would you take if feedwater flow rapidly decreas-ed due to a loss of the Shaft Feedwater Pum (1.50) b. Due to the above transient RX. Vessel level is decreasing at a very rapid. rate. At what Vessel level would you take action to depressurice the vessel ? (0.50) Is-it necessary to close the MSIVes during this transient ?(Explain)(.4) d. List three (3) conditions that could cause HPCI to automatically init-iate as a result of this transien (0.60) QUESTION 4.03 (1.50) Consider Operating Procedure N1-OP-43 c. During startup, vessel level may remain at the level of the flange. Why would level be maintained here and what are the required positions of the hSIV's ? (1.0) When should vessel level be lowered in the startup ? (0.5) QUESTION 4.04 (3.00) The reactor is at 75% power and you receive indication that

.both service water pumps have tripped on overloa In accordance with NI-50P-14, what are siu>(<6) of the seven immediate
.cperator actions for a complete loss of service water flow?  (3.00)
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QUESTION 4.05 (2.50) Concerning procedure N1-OP-14 , Containment Spray System : a. What two (2) signals are required to automatically start the contain-ment spray pump (0.50) b. What action should be taken following a confirmed high radiation alarm on the containment spray raw water system ? (0.50)

c.: The containment spray Raw Water Pumps must be manually started by the control room operator ? TRUE or FALSE . (0.25) d. This procedure directs you not to manually override or shut this system down after an auto initiation unless two conditions are met. What are these two conditions and who is authorized to make this decision ?

     (1.25)

GUESTION 4.06 (2.00) You have determined that a fire of undetermined origin requires icmediate evacuation of the control room. What actions should you try to take prior to evacuating the control room ? (2.00)

-QUESTION 4.07 (1.00)

List four (4) indications, (control room), indicative of an Off Gas explosio (1.00) GUESTION 4.08 (2.00) Why is an operator instructed to ' reduce reactor power to 80% of the original power level with Reactor Recireviation flow' BEFORE removing a feedwater heator string? (1.0) When two condensate booster pumps are required, the preferred linup is with til and 413 running) when one booster pump is required, 411 or $13 should be in service. Why is this pre-ferred? (1.0)

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QUESTION 4.09 (2.00) Dascriber.in general, the four things you would do to reset a high pressure crolant injection (HPCI) initiation, assuming that the initiation signal hos cleare (2.0) 00ESTION 4.10 (2.00) Answer the following questions regarding the shutdown procedure of one recirculation pump with the reactor critical, full pressure cnd temperature and some power level less than 90%. a. WHAT automatic action is expected when the discharge valve is fully closed ? (1.00) b. The lar.t' step of the procedure states ' Crack open pump discharSe valve T...' ' HOW is this step accomplished and why is it required ?

     (1.00)
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GUESTION 4.11 (1.00) What are three (3) conditions which would be defined as radiation

02ergencies-? (1.00) QUESTION 4.12 (3.00) The unit is at rated power. A positive reactivity addition is experienced and it is necessary to conduct an emergency power reduction por NI-OP-4 c. The procedure directs the operator to reduce power by recire flow but it cautions that an APRM scram may occur if the reactivityaddition is_not eventually terminated with control rods. Why is this? (1.00) b, Assume control rods were inserted per the Reactor Analyst instructions; power level has decreased to a stable value and it is desired to return to a higher power. Can the rods inserted to reduce power now be withdrawn? Why or why not? (1.00) c. Could the operator have individually scrammed rods from the scram timing panel to achieve a quicker power reduction? Explai (1.00)

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____________________________________________ ANSWERS -- NINE MILE. POINT -85/09/09-HOWE, ANSWER- 1.01 (1.50) 11. Decrease or at BOL=0.007 and EOL=0.0054 (0.50) b. Due-to.an increase in the power production by Pu-239 which has

.a much smaller delayed neutron fractio (0.50).

c. The reactor'will respond more rapidly to reactivity changes (0.50) REFERENCE Operations Technology Hodule 1, Chap. 9 ANSWER 1.02 (2.25) The' energy recovered in feed heating would otherwise be lost to the main condenser OR less heat is required from the reactor to reach the desired condition (1.0)

.b.- . Kilowatt output from the generator would increase (0.5). Steam that was formerly being extracted now passes through the turbine to the condenser (0.75).

REFERENCE Operations Technology danual, Hodule-9, Part. 6 JCK-151 ANSWER 1.03 - ( 1. 50 ) The coefficient becomes-more negative.(0.50) This is due to the d2 creased average moderator density at higher void fractions cousing an increase in slowing down time; thus the neutrons opend mor e time at resonance energies (resulting in the resonance-oscape-probability decreasing). (1.00) REFERENCE-Operations Technology Manual, Module 1, Chap. 13 _ _ _ - . _--___ _ -

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. ANSWERS -- NINE MILE POINT  -85/09/09-HOWE, ' ANSWER' 1.04 (2.50)

A. 1) Boilin3 heat transfer is more efficient 2) Steam is produced directly by the reactor 3) . Control of void fraction controls reactor power over a limited range 4) The reactor can be operated at a lower pressure (0.5 each 3 required) B. When the Onset of Transition boiling is reached, patches of steam form on the cladding and detach rewetting the cladding. As a result, cladding temperatures fivetuate when these steam patches form'and

 ' detach. These temperature fluctuations will produce cyclic stress in the cladding ultimately resulting in cladding failur (1.0)

REFERENCE Operations Technology Manual, hodule 10, Parts 3,4,and 5 ANSWER 1 05 (3.00) c. Rod A.(0.5) The relative flux at the center of the core is higher than at the edge of the core (due to leakage out of the core).

Since rod worth is dependant on the relative fiv:: (ratio of local to average flux), rods located near the center of the core usually have higher wort (1.50) b. Rod B.(0.5) Upon recovery from a scram, fission product poisons cause a severe flux depression in what was the highest power producing regions. of the core. Effectively, this results in a higher relative flux in the regions of low poison concentratio The result of these shifts in flux distribution is increased worth of the peripheral rods and decreased worth of those in the center of the cor (1.50) REFERENCE' Operations Technology Manual, Hodule 1, Chap. 14

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____________________________________________ ANSWERS:-- NINE MILE POINT -85/09/09-HOWE, ANSWER 1.06~ (2.50) 100% Peak - . .

     .  .

XENON . . CON % E * . ._...

         -
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50% E ss . . .

  ----(0.75)--- .(1.0).----------------(0.75)-----------

0- . . . . . . . . . . . . . . . . . . 20 40 60 80 100 120 140 160 180 TIME IN HOURS REFERENCE Operations Technology Manual, Module in Chap. 15 JCK-154 ANSWER 1.07 (2.00) Capacity =.1340 GPM = (speed /1800 RPM) (670 GPM) Speed = (1340 GPM / 670 GPM).(1800 RPM) = 3600 RPM (1.00)

Head = (3600 RPM / 1800 P,PM) (47 Ft-Lbf/Lbm) = 188 Ft-Lbf/Lbm (0.50)

BHP = (3600 RPM /1800 RPM) (70 HP) = 560 HP (0.50) REFERENCE Gcneral Electrier Thermodynamics, Heat Transfer, and Fluid Flow; Chap. 7 " ANSWER 1 08 (1.50) o Plant operates at saturated conditions. Thus temperatures can be found from steam table Tsat for 414.1 psia = 448 F Tsat for 995.22 psia = 544 F

;j  Heatup Rate = (544 F - 448 F)/ 1 hr = 96 F / hr        (1.00)

b. Yes (heatup rate per TS is 100.F / hr) (0.50)

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{NSWERS~-- NINE MILE POINT  -85/09/09-HOWE, /

REFERENCE' Steam Tables and NMP41 Tech Specs ANSWER l'.09 (2.00) Ccndensate oepression is the temperature difference between the saturation temperature for the existing condenser vacuum and the temperature of the condensate. Excessive condensate d:pression-decreases the operating efficiency of the plant since the subcooled condensate must be reheated in the reactor. ( .for definition) (1.0 for reason).

REFERENCE Operations Technology, Module 9, Part 8 ANSWER 1.10 .(2.25) 1. ECPR- protects against the onset of transition boilin ( .75 ) 2. LHGR- protects against exceeding 1% plastic strain on the clad due to excessive heat generation in the fue ( .75 ) 3.MAPLHGR-. ensures that peal fuel clad temperature will not exceed 2200 degrees F during a DBA-LOC ( .75 ) REFERENCE-Oporations Technology, Module 10, Parts 5, 6 , and 7 ANSWER 1.11 (2.00) Since the reactor was shut down by 1% dk/k as determined at the time of peak. Xenon, then the SDH will decrease as' Xenon decays.[1.03 Since Xcnon (peak) is greater than the 1% dk/k a reactor restart would occur.[1.03 cit. answer: The SDM-does not change.[1.03 The T/S definition lists the SDM for cere in the most reactive condition (i.e. Xenon free).E1.03 REFERENCE Operations Technology, Module le Parts 7 and 16

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____________________________________________ ANSWERS --'NINE MILE POINT -85/09/09-HOWE, ANSWER 1.12- (2.00) 1. Moderator / Void, a. ore negative 2.-Void, more negative l 3. Void, less negative 4. Doppler, more negative (0.50 for each correct answer) t l REFERENCE Operations Technology, Hodule 1, Parts 12 and 13.

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ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER 2.01 (2 00)

     ' Torus to the Drywell    (.5)

Reactor Building Atmosphere to the Torus (.5) 0.5 psid (.5) The primary containment is not designed for a negative pressure differential. (These vaccum breakers ensure that only a minimal D/P will be developed between the drywell and torus, the drywell and the reactor building, and the tcrus and the reactor building.) (.5) REFERENCE Simulator Systems Manual, Chapter 12 ANSWER 2.02 (1.00) An. immersion heater (50kw) maintains solution temperature between 63 F and 75 F. (0.50) Haintaining a minimum solution temperature prevents the solution from becoming saturated and precipitating. (0.50) REFERENCE Simulator Systems Manual Chapt. 8 s ! l L

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ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER 2.03 (2.50)

. SIGNAL   SETPOINT BYPASS
%
,f t c anual vessel isolation  N/A  none v m lo-lo reactor  5 inches none Yj sel level

(f Olo-lo-1,o condensor vaccum

7 inches Hs mode switch in startup or f refuel & Rx press < 600psis alo reactor pressure 850 psis mode switch not in run

' coain steam hi ra X normal , none
. Rhi steam flow  105 psid or 17.9% none of rated flow
' chi temp in steam tunnel  200 F  none (0.20 for each signal)
  (0.10 for each setpoint)
  (0.20 for each bypass)
  (Five of seven from list required)

REFERENCE Simulator Systems Manual Chapts. 11 and 21

~ ANSWER- -2.04 (1.50)

o. Reactor pressure + 260-psig = 500 psig + 260 psig = 760 psis (0.50) b. When driving in, the insert stabalizing valve will deenergi:e and close (0.50). This maintains flow and therefore pressure relatively constant (0.50).

REFERENCE Simulator Systems hanual, Chap. Sa

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ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER 2.05 (2.00) De 125 VDC battery board 11 - relief valves 111,112,113 125 VDC battery board 12 - relief valves 121,122,123 (0.50) 120 VAC from 4100 powerboard 102 - auto init, channel 11 120 VAC from 4260 powerboard 103 - auto init. channel 12 (0.50)

(125 VDC from battery boards and 120 VAC from 4160 powerboards for Core Spray pumps is acceptable) The discharge piping for each relief valve has an acoustic monitor and a temperature element which can be monitored from the aun control room and from the process computer respectivel (1.00)

REFERENCE Simulator Systems Manual, Chapt. 15 ANSWER 2.06 (1.50) c. To isolate and relieve the pressure in the scram valve pilot sit header. This results in a loss of instrument air to all scram valve (1.00) b. Reactor trip buse (0.25) c. De-energize to operat (0.25) REFERENCE NMP I System Simulation Manual chap. 10 pg. 7 ANSWER 2.07 (2.50 CFM at 110 psig (0.5) High drywell pressure - 3.5 psig (0.5) Low-Low level - 5' (0.5) Installed in parallel to provide 100% flow path in case of a failure of one valv (1.0) REFERENCE Simulator Systems Manual Chap.17

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ANSWERS --'NINE MILE POINT -85/09/09-HOWE, ANSWER 2.'08 (3.00) A. CST via Condensate transfer system, Fire Water System (0.5) B. 1080 psig(.33) or low low reactor level of 5 inches (.33) for a period of 10 seconds (.33) (1.0) C. Shut /the steam supply valve (s) (.5) D.~ Alternate opening and closing one condensate return valve (1.0) REFERENCE Simulator Systems hanual, Chap. 14 ANSWE .09 (3.00) A differential pressure sensor is used to confirm the integrity of_the CORE SPRAY piping within the reactor vessel ( between the inside of the vassel and the core shroud).(1.0) To continuously monitor-the integrity of the core spray piping, a Delta P cwitch measures the pressure difference between the two loops, which is offectively the inside of each Core Spray sparager. pipe, just outside of the Rx vessel.(1.00) f the-core spray sparager is intact, this pressure difference will be zer If integrity is lost, this pressure differential will include.the pressure drop across the steam seperator. Alarms at 5 psid in the control room Nr LN n n

As r (1.00) REFERENCE

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Simulator Svstems Manual, Chap. 17 , e

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ANSWERS -- NINE MILE POINT -85/09/09-HOWE, . ANSWER 2.10 (2.50)

. . Fuel assembly serial 4 are readable from the associated control ro . Lugs on the fuel assembly bail handle point at the associated control ro . Channel spacer buttons are above the associated control ro . Channel fastener spring clips are above the associated control ro . Gadolinium rods have longer end plugs which protrude through the-upper tie plat . Overall core symmetr required (0.25 for each correct answer)

b. 1. The mode switch is in refuel with one control 1od withdrawn. (0.25)

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

2..The mode switch is in refuel with the refuel platform loaded and over the core f(0.25) This condition inserts a rod block to prevent control rod withdrawal. (0.25) This places the fuel pool high radiation monitor, on the refueling bridge,on the emergency ventiation circuit (alarms at 1000 mr/hr).(0.50) _ REFERENCE Operations' Technology, Hodule VIII Part 4, and N1-OP-34, Refueling Proe ANSWER 2.11 (1.50) A. 1. MG set speed indication in the control roo . Recirculation flow control (blind-controller input) 3. Generator voltase regulator inpu (.5 each) (1.5) REFERENCE Simulator Systems Manual, Chapt. 4

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ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER- 2.12 (2.00) s. High temperature -> 83 des.C . (0.25) Low pressure - <.17 psis. / ' g #(

.. Low system flow < 442 sp (0.25)   (*

b. NO, (0.25) e Immediately reduce R Recire. flow to minimum, in an 7 6,} attempt to prevent a scram. (0.50) c. System flow.is regulated to maintain inlet pressure low enough to pre-vent water from entering the stator winoings in the event of a lea ( if a leak develops hydrogen will leak into the cooling water). (0.75) REFERENCE NMP. .Ni-OP-44, Gen.' Stator Cooling Water Sys., pg. .1-5 .

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l ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER 3.01 (1.00) 1)'open .2)-closed 3) open 4) reset (0.25 each) REFERENCE Simulator Systems rianual, Chapt. 4 ANSWER 3.02- (1.50) c. Shut b. As is (lock up) c. Open d. Shut e. Open f. Open (0.25 ea) REFERENCE N1-SOP-5, Re , pps. 2- ; ANSWER 3.03 (2.25) o.:0-100' Approximately 86' above TA (1.00) b. Reactor vessel isolation , Containment isolation , Recire.

l pump trip , Core spray initiation (with <365 psig resctor i pressure) , ECS initiation (with 10 see time delay) , Containment i .. spray initiation (with high drywell. pressure). (0.25 each) REFERENCE NMP Simulator Systems Manual cha , ppg. 2, 9.

l ANSWER 3.04 (3.00) 0.-Prevents movement of a control r~od which could allow individual ~ control rods to have greater than acceptable reactivity worth to lessen the severity of a rod drop acciden (1.00) b.-Reactor Level Control (steam flow & feedflow) , RPIS (tod position) , RMCS (rod select information) (1.00) > c..Upon attempting to make a third-insert error (with_two such errors previously made and uncorrected) the RWM will generate an insert bloc (1.00)

-REFERENCE NMP Simulator Systems Hanval chap 6 and NI-OP-37 rev. 8 pg _ _ _
  . _ _ - . . - . _ _ . . _ - _ .

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. ANSWERS -- NINE MILE POINT  -85/09/09-HOWE, <
. ANSWER 3.05 -(2.00)  M h $dN o. Normal 120 VAC and normal 24 VD whenWon (0.50)
' %Loffullscale'onanyrange./C Mode selector switch in RUN and associated APRM not downscal (1.00) ,

c.. Mode selector. switch in RUN or the IRM channel bypasse (0.50).

-REFERENCE

'NMP 1 Simulator Systems Manual, chap. 9 ANSWER 3.06~ (1.50)
.To maintain normal reactor level indication (0.5). Proper vessel level

requires that a flow path through the recirc system be available since _ cost level indicators measure coolant in the shr3vd. (1.0) REFERENCE N1-OP-1 pg. '10 and Simulator Systems Manual pg 4-15 ANSWER 3.07- (3.00).

o flow upscale trip (0.25), '105% of rated total recire flow (0.25),

. rod withdraw b l o c k ( O'. 2 5 ) , indicative of a failure.in the flow
;monitorin3 devices (0.25)
* flow comparter trip (0.25), > 7% deviation from each flow unit output (0.25), rod withdraw block (0.25), indicative of flow unit failures (0.25)
.x, flow unit inop trip (0.25), a) flow unit not in operate (0.125) b) any int'ernal module not' plugged in(0.125), rod. withdraw block (0.25),

indicative of' flow unit-failure (0.25F

' REFERENCE Simulator' Systems Manual, Ch. 9d, pg. Ch.9d-1 ANSWER 3.08 (1.00)

Lecal alarms and indication are provided in some areas to give instant warning tot workers'in the area.that a present level of radiation rate

.hos been exceede (1.0)

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L ANSWERS -- NINE MILE POINT -85/09/09-HOWE, REFERENCE Simulator Systems Hanval pg. 31a-4, and N1-OP-50A p I

    \ ., t t h
; . ANSWER 3.09 (3.00)  y\0 bd c. No. 2 seal pressure would appr 1 seal presst(re. Leakage-throughROincreasesandFS-A(,oachn will alarm at 1.5 spm.)  (1.00)

b. No. 2 seal pressure would drop upon magnitude of the failur Leakage-thru FS B would exceed. Min Flow and alarm hig (1.00) c.-Total. leakage out of the seal system would approach 60 spm as limited.by the breakdown bushing. Both FS A and B would alarm high. Precsvre in both seals would drop dependent on the magnitude of the failur (1.00) REFERENCE Simulator Systems Manual, Figure 4-2 ANSWER .3.10 (2.00) a. FALSE b.TRUE c.TRUE d.TRUE REFERENCE Simulator Systems hanval, pages 14-8 and14- ANSWER 3.1 (3.00) A. No('.25)-This power system has the capability to be synchro- e o r" } nized to its to its emergency supply (.75) prior to trans- ,ha s. i ferring therefore thera is no interuption of powe d t(1.0) OY'l i3'r#g Yes(.25) A half scram will result because the RPS trip busC J s rso supplied by that MG set loses power (.75) CF [1,$,s (1.0)

'C. Continuous power MG sets 162, 172, 167. These are the computer MG set and the two RPS bus MG sets. Also MG sets 161 and 171 L which are the battery charging MG sets.(.2 each)   (1.0)

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ANSWERS --JNINE MILE POINT -85 '09/09-HOWE , REFERENC NI-OP-48. Motor Generator. Sets pg 1-5

! ANSWER 3.12 (1.25)
'
< e. ~ as is b. as is c. closed d.'as}is
.o. open'
 .

REFERENCE Simulator Systems Manual,' Table 11-2, ps. 11-4 to 11-8

~- A N' S WER 3.13 ( .50)

o. TRUEL b'.' FALSE-

. REFERENC Simulator: Systems Manual, pg. 29a-6 and 29a-S
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( * ,, -- o4 . PROCEDURES ~- NORMAL, ABNORMAL, EMERGENCY AND PAGE 31

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RI65UL55fEAE C5ETR6L t ! ____________________ ANSWERS --.NINE'HILE POINT -85/09/09-HOWE, A.

L ANSWER 4.01 (2.00) The' rod should be' declared inop (per tech spec definiion) and no further. attempt should be.made to move the rod. (0.50) The directional control valves should be. electrically disarmed (0.25) cnd the module-valved out of' service (0.25)-(hydraulically icolated).'The control rod should NOT be scrammed as this may

 'ecuse further' mechanical damage-to occur. (1.00)

REFERENCE N1-SOP-15,: Stuck Control _ Rod Immediate Operator Actions

 ' ANSWER 4.02 (3.00) . Shift mode switch to refue Check all rods are fully inserte ._ Observe power' level decreasin . Check for HPCI operation. Ensure that both meter driven feedwater

_ pumps are runnin . Check _-that the emergency condensers are in operatio .' Check that the Core Spray pumps are running snd recirculating back to the toru (0.25 for each correct answer) b., Low-Low-Low-Level (- 10 inches)- (0.50) c. Yes, To conserve coolant inventory. (0.40) . Runout flow of 1.9 x 10(6) or 3800 spm. (0.20) a:2. Turbine Trip (0.20) 3. Low Rx. water level

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 . REFERENCE-N1-SOP-3, and Simulator Scenario Objectives 4 1E2  .
 ,

ANSWER 4.03 (1.50) Jo. To maintain flange tenperature at 125 F (0.5).-The HSIV's will be shut (0.5).

b. After reaching initial criticality and before continuing rod

 .withdrawl. (0.5)
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~ ANSWERS -- NINE' MILE POINT   -85/09/09-HOWE, A '.

REFERENCE-N1-OP-43, .Startup and Shutdown Procedure, Startup Procedure ANSWER 4.04 (3.00) 1. Initiate a manual reactor scram 2. Place Mode Switch to refuel

:3._ Start EM service water pumps 4.: Trip recite pumps 5. Close MSIV's 6. 0).n vacuum breakers. Place turbine on turning gear 7. Pi_.:e'EM condensers in service (.5 each, 6 required)
. REFERENCE

' NI-SOP-14, Loss of' Service Water Cooling, pg 2 ANSWER 4.05 (2.50)

  ~ A combination of lo-lo reactor vessel water level and high drywell pres-sure (3.5 psis.) (0.50)

b.;The raw water-pump.and the containment spray pump in the affected loop should be secured. (0.25) The loop suction and discharge valves should be closed. (0.25)

c. TRUE . (0.25) d. 1. Sufficient evidence shows that the system is not performing its

~ intended function. (0.50)

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

~$hutdown~of the system will be at the direction of the Station Shift Sup (0.25)
,

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

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RI6566U556dL"66ATR6L ____________________ ANSWERS ---NINE MILE POINT -85/09/09-HOWE, ' ANSWER' 4.d6 (2.00)

 ~
 *. Scram the reactor-c Trip 345 KV breakers and tr'ip machine mg Verify' scram
 ;c Verify. turbine trip
 :.c -Initiate :emer3ency . cooling -

n Operate manual isolation' switches - vessel isolation channel #11'and

 ' 412.on the console and verify MSIV and Rx Water C.U. isolation
 * Sound..the fire alarm and identify area (s) if known m' Verify HPCI initiation (0.25 each)
,

REFERENC ~N1-SOP-11, Control Room Evacuation (Fire)

--
' ANSWE .07  -(1.00)
 -1.J0ff gas high ~ temperature: alarm
 '
 . Off gas high pressure alarm-3.~.0ff' gas?interstage blockin's. valves BV 76-12 & BV 76-13 closes; Off gas flow goes to::ero Condenser. vacuum decreases-low vacuum' alarm
. .

5. Reactor: scram E 23' HG. vacuum (Four required 0.25 each) ect es a r s e alti l ardiall- tam th r setpo$ntf-[ eactor f Titia gem ne condender us n cessary

 .. ' 6 . i station  r5 nne l and plant anagement
 ,
  {fo    , v  - . ('.5 each except #6=.25)

'

_ REFERENCE

 . NI- SOP-18, pg 2

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____________________ ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWER 4.08- (2.00)

- c. : This will prevent the other feedwater heater. strings from being overloaded and will preclude possible over power of the nuclear fuel. Also power increase due to increased inlet subcoolin (1.0) This preferred lineup will preclude a system feedwater distur-t  bance'due to the loss of powerboard #101. Also insures HPCI (1.0)

availabilit REFERENCE-N1-OP-16, pg. 17 &l8 ANSWER 4.09 (2.00) Verify a) Feedwater flow on til and #12 is < 1.9 million Ibm /h b) Reactor low level trip is clear (.5) Switch feedwater pump til and #12 H/A stations to manual (.5) Adjust the nanual outputs until the deviation meters on the til

 #12 H/A stations are nulle (.5) Press the 'Feedwater Return to Normal After HPCI' pushbutton on the reactor control consol (.5)
' REFERENCE N1-OP-16, pg. 16 & 17
. ANSWER  4.10 (2.00)

a. At about 20% opening of the discharge valve (0.5), speed will automatically drop to 20%. (0.5) b. The valve is given approximately a 15'second open signal (0.5) this will allow the valve disc and stem to warm up. (0.5) REFERENCE-N1-OP-1, Nuclear-Steam Supply System, pg. 15

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____________________ ANSWERS -- NINE MILE POINT -85/09/09-HOWE, ANSWE .11 (1.00) 1.:High Radiation and/or High Airborne Radioactivity to a local area 2. High Radiation and/or High Airborne Radioactivity to a general area 3. Accidental release of Radioactive liquid from the restricted area (0.33 each) REFERENCE-EPP-1, Radiation Emergencies, page 2 ANSWER 4.12 (3.00) a. As power is reduced, so is the APRM. scram (flow Bias). The distance between the power'line and the APRM scram line does not increase, the are essentially parellel (1.00) b. No. Notify reactor analyst for proper rod withdrawl sequence otherwise high-local power and abnormal flux patterns may occu (1.00)

.c. No. This could lead to abnorms1 flux patterns and possible fuel damag (1.00)

REFERENCE HI-OP-43, pg. 16'

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TEST CROSS REFERENCE PAGE 1 OdESTION VALUE- REFERENC ,--_-____ _--_-- .-----_-_-_ 01.01 1.50 AXA0000003 01.02 ~2 25 AXA0000004

~01.03 1.50 AXA0000005 01.04 H2.50 AXA0000006 01.05 3.00 AXA0000008 01.0 .50 AXA0000009-01.07 2.00 AXA0000010 01.08 1.50 AXA0000011 01.09 2.00 AXA0000016 LO1.10 2.25 .AXA0000017 01.11' 2.00 AXA0000010 01.12 2.00 AXA0000019

______ 25.00-02.01- 2.00 AXA0000007 02.02 1.00 AXA0000012 02.03 ~2'.50 AXA0000014 02.04 1'.50 AXA0000015 02.05 2.00 AXA0000020-02.06 1.50 AXA0000031-02.07 2.50 AXA0000032 02.08 3.00 AXA0000033

>02.09 3.00 AXA0000034 02.10 2.50 AXA0000035 02.11 1.50 AXA0000036 02.12 2.00 AXA0000051*
 -___-_

H25.00 103.01 1.00 AXA0000012 03.02 1.50 AXA0000027

.03.03- 2.25 AXA0000028 03.04 '3.00 AXA0000029 03.05 2.00 AXA0000030 03.06 1.50 AXA0000043 03.07 3.00 AXA0000044 03.08 1.00' AXA0000045 03.09 3.00 AXA0000046 03.10 .2.00 AXA0000047 0 3.11 3.00 AXA0000048 03.12 1.25 AXA0000049-03.13 .50 AXA0000050

______ , 25.00 04.01 2 00 AXA0000021 l 04.02 3.00 AXA0000022 04.03 1.50- AXA0000024 04.04 3.00 AXA0000025 , Y --

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TEST CROSS REFERENCE PAGE 2

. 0tfESTION VALUE ' REFERENCE

________ ______ __________

04.05 2.50 AXA0000026 04.06 2.00 .AXA0000037 04407 1'00

  .~ AXA0000038 04.08: 2.00 AXA0000039 04.09 2.00 AXA0000040
'J04.10 -
 -2.0 AXA0000041 J04.11 1.00 AYo0000042 04.12- 3.00 AXA0000052-
-

______ 25.00 ______ ______ 100.00

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IR 05000220/1985014: Difference between revisions

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v t e Inspection Report - Nine Mile Point - 1985014
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 \|
1985 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 13(0) 14(0) 19(0) 98(0) 8(0) 15(0) 31(0) 41(0) 48(0) Security Operator Exams Emergency Planning Other

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