IR 05000247/1985028
| ML20151T596 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/28/1986 |
| From: | Coe D, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151T582 | List: |
| References | |
| 50-247-85-28, NUDOCS 8602100316 | |
| Download: ML20151T596 (48) | |
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x U. S. NUCLEAR REGULATORY ~ COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-247/85-28(0L)
FACILITY DOCKET NO. 50-247 FACILITY LICENSE N0. DPR-26 LICENSEE: Consolidated Edison Company of New York, In Buchanan, New York 10511
. FACILITY: Indian Point 2 EXAMINATION DATES: December 10-13, 1985 CHIEF EXAMINER: ret fM u W /- M'<7b D . ' H . Co e ,' R ctor Engineer (Examiner) Date REVIEWED BY: h)
R.~ M. Kel fer, Chief, Projects Section IC'
l/17/ /6 Date APPRENED BY: /!-177d>
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R. S' Kister, Chief, Projects Branch No. 1 Date'
SUMMARY: . Written exams were administered to 5 SR0's and one Instructor Certification candidate (1 Upgrade and 4 Instant). Oral and simulator exams were administered to four SRO's (Instant) and one ,
Instructor Certification. All candidates successfully passed all portions of the examinatio '
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B602100316 860205 PDR ADOCK 05000247-G PDR OFFICIAL RECORD COPY OL EX 50-247/85-28 - 0003. ,
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REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS:
1 SRO I Inst. Cer l l Pass / Fail l Pass / Fail l l l l l 1 I I IWritten Exam I 5/0 1 1/0 l I I I I I I I I 10ral Exam I 4/0 1 1/0 l I I I I I I I I ISimulator Examl 4/0 l 1/0 l I I I I I I I l l0verall I 5/0 1 1/0 l l l l l CHIEF EXAMINER AT SITE: D. H. Coe, NRC OTHER EXAMINERS: D. G. Ruscitto, NRC
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i 0FFICIAL RECORD COPY OL EX 50-247/85-28 - 0004. ,
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O O-s 3 Summary of Generic Deficiencies Noted from Written Exam
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R0 Exam - Not give SRO Exam - (6 candidates).
The following were minor areas'of weakness:
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ATWS procedure (FR-S.1) methods for emergency boratio Basis for Technical Specification limits on RCS pressure, flow rate, and Tav Knowledge of the contents of OAD-5,~ Procedure Adherence and Us Proper sequence of err.ergency plan implementing procedure (IP-1001)
steps during a declared aler Understanding of the dilution factor (Xu/Q). Interface with Plant Staff During Exam Period The training staff were helpful in providing the necessary administrative support for the examination. 'thile the simulator exams were in progress,
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several individuals entered the simulator room without first getting permission from the Chief Examiner. Distractions of this nature must be strictly avoided during future examination . Personnel Present at Exit Interview NRC Personnel D. Coe, Chief Examiner L. Rossbach, Senior Resident P. Kelley, Resident Facility Personnel-J. Basile, General Manager - Operations A. Giorgio, Training Manager ,
-D. Koutouzis, Nuclear Training M. Mueller, Simulator Instructor
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T. Mansell, Simulator Instructor W. Kriebel, Simulator Instructor Summary of NRC Comments Made at Exit Interview In accordance with present regional policy, no preliminary results were given. There were no generic knowledge or training deficiencies noted during the oral or simulator exam FFICIAL RECORD COPY OL EX 50-247/85-28 - 0005. ,
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01/17/86
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' O s 4 Comments were made regarding simulator performance. Several simulator malfunctions disrupted the continuity and/or accurate simulation of the scenarios the examiners were running. This distracts candidates from the
"real" casualty and makes evaluation mare difficult for examiners due to unexpected changes in the scenario. Examples of this included electrical loads dropped from an energized bus as if the bus had lost power, pumps that start or stop when they should not, and pumps that cannot be turned on or off when they should. The simulator froze twice during a single scenario and required more than one minute each time to restart. Although a small number of simulation difficulties are expected during NRC exams since new and different scenarios are always run, the number of problems being experienced at the IP2 simulator is approaching that level which will require excessive NRC examiner time spent pre-running and modifying scenarios. This time must necessarily be spent on the same day the scenario is used for exam purposes, lengthening each exam da The NRC is presently formulating minimum standards for facility simulators that may be used for NRC examinations. Until these standards are approved, the IP2 simulator may be considered adequate for exam purposes unless the Chief Examiner onsite deems otherwise due to excessive malfunction . Open Items Which Were Closed 85-16-03 RMS is not modeled where it is needed to confirm LOCA RMS model responded properly during a LOCA conducted as part of this examinatio Initial Condition #1 is for solid conditions onl I.C. #1 is now available with a nitrogen bubble-in the pressurize No remote shutdown or operations procedures are permanently located at remote shutdown station Procedures are now located at the alternate safe shutdown panel in the PA Remote S/G and Pressurizer level indications were not labeled at the remote shutdown panel. All indications read in inches of water D/P with no available conversion tabl ,
Indications are now labeled and a conversion chart is poste No dedicated phone system is available for the operators of the several remote shutdown stations. Presently, walkie-talkies are the only instantaneous direct communication method available, and these must be brought from the guard house when neede OFFICIAL RECORD COPY OL EX 50-247/85-28 - 0006. /21/86
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=,- 5 As per Mr. Murray Selman's letter dated December 30, 1985 to Mr. Richard Starostecki, Director, Division of Reactor Projects, Region I, dedicated walkie-talkies for the exclusive -
use of operators evacuating the-control room were placed just outside the control room. An in plant wireless' communication system is still being acquired but the dedicated walkie-talkies satisfy the intent of the open item. The effective use of the portable walkie-talkies was demonstrated during the ASSS,-
Appendix R, walkthrough conducted for an NRC inspection team on September 16, 198 Attachments:
1. Written Examination and Answer Key (SRO)
2. ' Facility Comments and Resolutions
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OFFICIAL RECORD COPY OL EX 50-247/85-28.- 0007. ,
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11 S. NilCLEAR RFGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENGE EXAMINATION FACILITY: INDIAN POINT 2
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RFAC10R lYPE: PHR-HEC 4
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DATF ADHlNISTERED: 05/12/10
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FXAMINER: BARBEF, _________________________
APPLICAN1: _________________________
INSTRUCTIONS TO APPLICANT:
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lise separate paper for the answer Write answers on one side onl Strple question sheet on top of the answer sheets, Points for each question are indicated in parentheses arter f. h e question. The passing stede requires at least 70% in each catosory and a final grade of at 1cest 80%. Examination papers will be picked up six (6) hours after thn examination starts,
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CA1EGORY
________ ______ _________-_ ____-___ ___-_-___---___--_----_---__-_-----
25 0 00
___1__0 __ '_1__
__'5 ___________ ________ 5 THFORY OF NilCLEAR POWER PLANT OPERATION, FLUIDS, AND THERH0 DYNAMICS 25 0 50 PL ANT SYSTEMS DESIGN, CON 1RUL,
___1_0___ _'1_1_0_ ___________ ________ 4 AND INSTRUMENTATION 25 0 25 0 PROCEDURES - NORMAL, ABNORMAL,
___1_0___ ___1_0_ ___________ ________ EMERGENCY AND RADIOLOGICA CONTROL 25 0 50 ADMINISTRA11VE PROCEDURES,
___I_0___ _'I_I_0_ ___________ ________ CONDITIONS, AND LIMITATIONS 100.00 100.00 f0TALS
________ ______ ________-__ ________
FINAL GRADF _________________%
All work done on this examination is my own. I have neither given nor received ai PPLIC5UiI5~5EGU5TURE~~~~~~~~~~~~~~
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. + THEORY OF NUCLEAR POWER PLANT 0PERATION, FLU 1DS, AND PAGE 2 J
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QUESTION 5.01 (3.00)
', s. Provide a definition of DNB. (1.0)
b.EWould DNB be more'likely, less likely, or neither if the following parameters were decreased while at power (and prior to any trip)?
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1) RCS Pressure 2) RCS Flow 3) Tave 4) Total Steam Flow
. A DNBR of'1 3 was chosen as the limit for the IP-2 Safety Limit curves
- - depicting the maximum' allowable combination of reactor power, RCS temp-4 eraturer pressure, and flow. This value provides for operation with a
. margin from actual DNB (DNBR =.1.0). What made this margin i 'necessary? (1.0)
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QUESTION 5.02 (2.25)
Answer the following questions using the provided IP-2 data. Show all work and state all assumptions ~.
, Tave= 547 F MTC= -4 pcm/ F All rods out Rod control is in manual-Boron worth =100 ppm per % delta rho Doppler only Power Coeff= -9 pcm per % power a. Assuming no operator actione calculate the new Tavn if power is increas-ed from 90 to 100%. Reactivity coefficients are constant with power and temperature. (1.5)
b..How large a Baron addition or dilution would be necessary to maintain a constant Tave for the power escalation of part a? (Full credit given correct process). (0.75)
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GUESTION 5.03' (3.00)
s.The IP-2 RCc is designed to allow and promote natural circulation. What RCS design 'Jature ensures that natural circulation (NC) will occur?(1.0)
b. If, during stable NC cooldown, AFW flow is sudden 1v stopped, how will NC flow initially repond and why? (1.0)
c. If a NC cooldown and concurrent depressurization is conducted too rapidly, a void could occur in the vessel head region. Specifically, what causes this to occur? (1.0)
QUESTION 5.04 (2.50)
s. Which one of the following descriptions best supports the reason why Xenon reactivity increases sharply after a trip from 1000 hrs. at 100%
power ? (0.5)
1) Xen:.., decays less rapidly due to a reduction in the neutron flu ) Iodine half-life is much shorter than Xenon half-lif ) Iodine production is greatly reduced and Xenon production is greatly ~
increased due to the reduction in neutron flu ) Due to reduced neutron absorption, Iodine concentration increases, and Xenon decays directly from Iodine, thus Xenon increases, b. Give two. reasons why Sm-149 is not as much of a concern to an operator after a reactor trip as is X (1.0) A Xe oscillation in a reactor core might be produced by certain types of rod motion. -How would the Xe oscillation resulting from the following.
, two cases be different? Explai (1.0)
I. A turbine runback occurs with rods in auto. Rods drive 60 step II. Rods are driven 60 steps starting from the same position as in Case I, but slowly over 4 days time.
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GUESTION 5.05 (2.50)
. What is the subcooling margin (SCM) of the plant if the following conditions exist ? (1.0)
Th =580 F Ppzr=2185 psis Tave=550 F Psg =850 psis Te =520 F b. If plant power is raised from 50 to 100%, how will SCM change (increase, decreaserstay the same) ? Why ? (0.75)
c. Which of the follouing would give a smaller SCH? Assume identical RCS pressures. Briefly explain wh ) SCM during a controlled natural circulation cooldown immediately follouing a reactor trip from loss of flo ) SCM from continued operation at 5% power 3) SCM produced when all RCP's are operated at normal no-load temperature after extended shutdow QUESTION 5.06 (1.50)
. How does Deta Bar change (increase, decrease, stay the same), if at all, if temperature is raised 5 F in a short period of time ? Why ? (0.75)
b. Which condition would result in a higher startup rate; a rod ejection accident at BOL or EOL ? Explain. (0.75)
QUESTION 5.07 (4.00)
a. If core power is increased from 50% to 100%, how, if at all, will differential rod worth change (increase, decrease, stay the same) for the following 3 cases ? Why ?
1) Rod position and boron held constant, and temperature allowed to-decreas (1.0)
2) Doron constant, Bank D at 150 steps and is then fully withdrawn, temperature remains constan (1.0)
3) Rod position constant, boron dilution used, temperature constant (1.0) Indicate whether the follouing statement is True or False and explain your answer:
The differential rod worth of Bank D rods at the moment of criticality during a startup will be the same as when the rods are at the identical height during power range operation (1,0)
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. TilEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
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QUESTION 5.08 (2.25)
For the followins 3 cases, state whether the criticality of the reactor can be determined. If criticality can be determined, state so. If not, state the action that could be taken to determine criticality. (2 25)
Case 1 Case 2 Case 3 Rod motion in progress Yes No No Boron dilution in progress No No Yes SR level status (CPS) 5+E05 4+E04 3+E05 increasing increasing increa SUR status (DPM) + + + oscillating oscillating oscil slightly slightly positive OUESTION 5.09 (2.50)
Two idcntical reactors are taken critical usins continous rod withdrawa Reactor A has a rod speed of 48 steps per minute and reactor B has a rod speed of 24 steps per minut Which reactor will have the highest source range counts at criticality? Why? (1.5)
-8 Hou'will 10 critical rod heights compare in the two reactors?
Explain briefl (1,0)
GUESTION 5.10 (1.50)
Assume that a reactor has tripped and that no heat is removed from the primary coolant AFTER Tave stabilizes at its no load value. Assume the following parameters are constant and calculate the length of time it will take for pressurizer safeties to lift if primary pressure is maintained at exactly saturated conditions and no boiling occur Mass of water in the RCS = 214,000 lbm Cp = 1.2 BTU /lbm-degree F RCP heat input to primary = 15 MW Decay heat is 5% of full power (***** END OF CATEGORY 05 *****) . . . - . . .. - -- . . - - .. ,
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i' PLANTJSYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 'PAGE 6
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! QUESTION 6.01- (2.75) Why can't the High Main Steam Flow SI signal provide the initial
. protective action for a Main Steam Line Break upstream of an MSIV. (1.0)
b. Initiating a.High Containment pressure actuation signal requires deenergizing 2 of 3 pressure control devices. Yet, a High-High signal requires 2 sets of 2 of 3 pressure devices to operate. What is the purpose of this redundancy and why is it necessary ? (0.5)
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'c. If an automatic SI signal is received in which only busses 2A and SA are
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-energized will-the motor driven AFW pumps start ? If so, why ? If not, why not ? (0.75)
, d. What 2~ actions must' occur in order to defeat an automatic SI condition I at the IP-2 plant ? (0.5)
QUESTION 6.02 (3.50)
Explain HOW and WHY each of the following rod control components ~is used to modify a power mismatch signal. Include relative gains chosen (highelow), if appropriate. (3.5)
Impulse-Las unit Non-Linear Gain unit Variable Gain unit-QUESTION 6.03 (1.50)
What is the pur:ycse of/ reasons for the following operational limitation a. Stop charging flow prior to stopping letdown flow. (1,0)
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b. Letdown (LD) flow is automatically bypassed to the VCT when LD temperature is 145 F. (0.5)
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v PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 7
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QUESTION 6.04 (2.50)
Answer the following questions regarding the Reactor Coolant Pump a. Wh. * is the most probable cause of a low thermal barrier delta P alarm ? (0.5) What design features of the component cooling water system prevent overpressurization in the event of a leak in the thermal barrier HX ?
(1.5) The.RCP seal package performs various functions, one of these is bypassing the #1 Seal. What is the purpose of this function ? (0.5)
00ESTION 6.05 (2.00)
a. WF.at are the 2 automatic start signals for the turbine driven AFW pump ?
Include coincidences, if appropriate. (1.0)
b. What automatic action occurs if flow from a motor driven AFW pump drops to 25 spm ? (0.5)
c. If flow is reestablished at a maximum of 45 spm, will any further automatic action take place ? Why or why not ? E>tplain (0.5)
QUESTION 6.06 (1.75) What'is the chemical compound used in the spray additive tank ? Give 2 reasons why it is used. Be specific. (1.25) An automatic signal initiates containment spray and the reactor operator resets the signal that controls the spray pumps and valves after the signal clears. What if anything, will happen if a subsequent valid signal is recieved after the pumps are stopped and the valves returned to normal?(0.5)
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 8
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QUESTION 6.07 (3.75) List 2 conditions that will cause a diesel generator to auto st' ar Be specific. (0.5) What 3 Emergency Engine Shutdown trips are inhibited (blocked) if an SI signal is present ? (0.75) If Diesel' Generator 21 was the only AC supply for Instrument Bus 21, what would be the status of the following inverter indications a short time after the diesel trips. Indicate the correct response and explain your answer. (2.5)
DC input voltage (V2) higher /same/ lower Inverter output voltage (V1) higher /same/ lower Inverter supplying load light (PL1) on/off In synch light (PL2) on/off-Alternate source available light (PLS) on/off QUESTION 6.08 (2.25)
, a List, in sequence, the design features that operate to mitigate (lessen)
the effect'of a pressure increase from normal operating pressure to the RCS pressure safety limit. (1.0) For a pressurizer level and pressure decrease during a LOCA, what automatic actions will function to protect equipment and prevent the loss of pressurizer level? (0.75) An insurge into the pressurizer causes a pressure increase and a variable high level alarm at 5% above program level. So then, why do all pressurizer heaters come on in this condition ? Wouldn't this cause a larger pressure excursion ? Explain. (0.5)
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6 '. PLANT SYSTEMS DESIGN,' CONTROL.,-AND INSTRUMENTATION PAGE 9
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QUESTION '6.09 (3.00)
Answer-the.followins concerning the Nuclear Instrumentation System Explain the response of the reactor protection. system upon a
. simultaneous loss or failure of both IR channels (where both channel-outputs 90 to zero) for each of the following two cases. Briefly explain the differences, if-an (1)- The reactor is in the middle of the IR durin3 a startu (2) 'The reactor is at 20 % steady state power.'E1.0 each] Assume that while at 100% power one IR channel fails HIG If a reactor trip occurs while in this condition, what additional actionfmust be taken durin3 the emergency procedures to ensure
'the proper operation; of -the NI system? Why? [1.03 QUESTION 15.10 (2.00)
In the. event of the loss of, normal, AND alternate, AND emergency sources ~
. of power to the four IP2 480vac busses, there exists a method of supplying
backup electrical' power to certain IP2 components from a normally
.de-energized sourc What is the location of the breakers which will feed this backup-power to these components? (0,5)
b. List six types of components to which this backup power source may
- be directly connecte (1.5)
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QUESTION 7.01 (2.50) Which one of the following plant conditions by itself will automatically initiate SI ? (0.5)
1) Pressurizer pressure of 1872 psig 2) Pressurizer level of 4%
3) Steam pressure of 585 psis 4) Containment pressure of 3 psis 5) S/G 1evel of 18%
b. Which one of the following plant conditions by itself will automatically trip the reactor ? (0,5)~
1) Pressurizer pressure of 2335 psig 2) Steam pressure of 600 psis 3) Pressurizer level of 4%
4) S/G 1evel of 18%
5) Containment pressure of 1 psis c. TRUE/ FALSE If IP2 equipment and instrumentation is operating as designed and P-7 issatisfied, a turbine trip will always cause a reactor trip. (0.5) TRUE/ FALSE Adverse. containment conditions allow manual SI termination at a lower pressurizer level due to excessive localized heating of the reference les (0.5)
e. Which one of the following conditions represents the maximum allowable EDG loading during a LOCA. (0.5)
1) A load of 1750 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> 2) A load of 1950 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> 3) A continuous load of 1750 kW 4) A continuous load of 1950 kW QUESTION 7.02 (1.75)
c. Per the ATHS procedure (FR-S.1), what 2 conditions require emergency boration of the RCS ? (0.5)
b. During an ATHS, if emergency boration can't be established and subsequent attempts to establish normal boration fail, the operator must perform 3 manual actions. What are they ? (0.75)
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~ QUESTION 7.03 (2.00)
a. The IP2 Tech. Specs.-allow.a rod misalignment of "+" or *" 12 steps when bank demand is less than or equal to 210 steps and "+' 17 and
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12 steps when above 210 steps. Why is the tolerance band wider at higher rod heights ? (0.5)
. b. Per SDP 15.5, Dropped Rod and Ha1 positioned Rod Verification, what 2 indications are used to verify the existence of a potentially misaligned rod.-(0.5)
c. During the realignment of a dropped rod certain rod control design features ensure only the affected rod is moved. What are they and what operator actions are necessary to ensure their operation ? (1.0)
QUESTION 7.04 (2.00) gg-I.3 Resfunse to Answer the following questions concerning IP2 Procedure A ^^ 8 Voids in Reactor Vessel, If venting is performed to remove voids in accordance with this procedure, a maximum venting time must be computed. Why? (0.5)
t . chment 3 of this procedure provides an alternate method to esti '
RCS voi out RVLIS. Under what specific plant conditi is alternate method i to be used? (0.5) An Indirect Symptom / Indication s during spraying operations a rapid increase in pressuri' el indica e eous RV head void.
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O However, a simila anse might be observed for spra tion with a RV steam . ow would the on-shift SRO differentiate between a
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l steam void in these circumstances ? (1.0)
Q gy yk Voi$ 3 M OS A 0 SW 7 k2 awf g&S- a g-e IA5 dA -lo sbw A4-. Y voEA ? 0. y) O sf & mA B nos <d & !~ JL<_ cd"IU' ys's, i
wat nu a vSca n z t w A veM ? co. s6 (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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QUESTION 7.05 (2.00)
Answer the following questions concerning the IP2 Plant Heatup and Startup Procedure Prior to reactor startup, a minimum temperature is imposed based on RCS boron concentration. What is the purpose of this limit ? Why is it necessary ? (1.0)
b. An additional limitation is imposed on the minimum permissible '
combination of RCS temperature and pressure. What is the purpose of this limit ? (0.5)
c. A minimum condition for criticality requires a minimum temperature be reached before startup can commence. If nuclear heat can't be used because this limit is not met, then how's the heatup accomplished ?
(0.5)
GUESTION 7.06 (2.75)
c. What does the SHS's signature on the Job Specific RHP signify? Be specifi (0.75) The IP2 plant is, in cold shutdown and 2 operators are venting the RCS side of the regenerative H.X. Pre-job planning indicated that radiation levels during venting should jump from 300 mR/hr to 2 R/hr. During the actual venting radiation levels jump to 4R/hr, but the operators attribute it to fission product gases which they believe will decay off shortly. What actions are you, the SHS, going to take ? Why ? (1.0) As a part of the Radioactive Haste Reduction program, bass of waste are color coded as to type and content. Match the contents of each bag with its colo (1.0)
___ Contaminated tools 1) Orange Contaminated trash 2) Yellow
~~[ Contaminated anti-c's
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3) Green
___ Clean trash 4) Red
___ Clean tools 5) Clear 6) White (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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GUESTION 7.07 (2.75) a. You are the SRO monitoring Fuel Handling in the "u 1 Suildins. You notice precariously that the-bridge on the top operator cornermoving of :#E _f,fugl a assembly (FA) sets it th- Jts :3: ;;hs. In which direction, if at all, will the dillon load cell change (increase /
decrease / stay _the same).? What is the potential consequence, if any, of the bridge operator's action ? .^. :v;; 0111 n lead :11 licit cr- not ir c==t, 0.75) A ssa ~< yo% slee k ce lalc. g e cue.c] cock k W if5 kwt eeen re c) e d .
b. If a high radiation alarm is sensed by R-Kl4 Fuel Building ARM ),
3% automatic actions are performed. List them. (1.0)
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c. Containment Fue1 Handling bridge and trolley movement interlocks are bypassed and an inattentive bridge operator slams the fuel handling masts into the cavity side wall at full speed. Your as the SRO supervising fuel handling, anticipate that the FA he was carrying was severly damaged. What actions should you take ? (1.0)
DUESTION 7.08 (3.00) One cause for concern during a Steam Generator (S/G) tube rupture is that the leak will BYPASS one of the three Soundaries that keep fission products from reaching the public. What are these 3 boundaries and which one is BYPASSED during a S/G tube leak ? (1.0)
b.-What are 4 methods that_can identify which S/G has a tube rupture per E-3, S/G Tube Rupture ? Setpoints are not required. (1 0)
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c. If a fuel failure accompanied a steam generator tube rupture, what might be the consequences of isolating the affected steam generator significantly above 541 F ? (1,0)
l QUESTION 7.09 (2.25) A control room atmospheric condition exists that requires evacuatio If plant power is 100%, wh6t immediate action should the SRO tako? (1.0)
b. Who (title),'by what method, and where is S/G feeding being contro13ed from outside the control room?. Assume normal heat removal c&pabilitie (0.75)
c. If adequate SDM doesn't exist, what action should the SRO take or have taken? Individual steps are not necessar (0.5)
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. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 14
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QUESTION #5" " *b " " O' O 7.10
/ % of all Ac Poser 7#
(1.50)
c.Whatmethodisusedtokeepthe.corecooledwhen)IndianPoint-Station loses normal and emergency AC power ? (1.0)
- EcA of-M Ao ::;l cr.dAlt b. The errp'- r-tery actions of Loss Caergm..my AC require letdown be isolated cr:d terrer atr trintained ' ^^? c b>; Inter: C'.
What is the purpose of &hece actionk ? (0.5)
+kis QUESTION 7.11' (1.50)
The following critical safety functions are listed in alphabetical orde Re-order them in their correct order of priorit c. Containment Core cooling c. Heat sink d. Integrity e. Inventory f. Subcriticality QUESTION 7.12 (1.00)
Several hours after a lorse LOCA has occurred, what criteria is used f to determine whether high or low head recirculation'must be established? Numbers are not require (1.0)
.
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15
QUESTION 8.01 (2.50)
c. Briefly explain why IP2 was allowed to hydrostatically test the RCS to 3110 psig when '.he RCS pressure safety limit is 2735 psia. (1.0) IP2 Tech Spec. 2.1 requires the limitation of RCS Tave, pressurizer pressure and RCS flow rate .o specific and independent value What is the basis for these limits?
Time limits are not required. (0.5) If the IP2 total nuclear peaking factor was found to contain extra unnecessary conservatisms and the next cycle's Reload Core Design eliminated them, how, if at all, would peaking factor change ?
(increase / decrease / stay the same) Briefly explain your answer. (1.0)
QUESTION 0.02 (2.00)
Answer the following questions concerning IP2 facility staffing.~ Assume plant power is 100% unless otherwise specifie a. Due to unforseer, circumstances only one Radiological technician is available to relieve the offsoing shift. Is this permissible by Tech Specs ?
(0.5)
b. What is the minimum number of fire brigade members ? What is the minimum number of operators necessary for safe shutdown ? Is it permissible to use an individual required for safe shutdown as a member of the fire brigade ? (1.0)
I '
c. Por Tech Specs, select the one set of crew manning that best represents the minimum requirement for reactor startup. An STA is on-shif Do not consider the SHS in your answe (0.5)
1) 2 SR0s, 1 RO in the control room, 1 NPO 2) 2 SR0s, 2 R0s in the control room, 2 NP0s .
3) 1 SRO, 2 R0s in the control room, 2 NP0s 4) 1 SRO, 1 RO in the control room, 1 NPO QUESTION 8.03 (2.00) Why are the NSIVs required to be closed within 5 seconds of receipt of a close signal ? 2 reasons. (1.0) What Tech Spec actions are required if a CST discharge is found shut ? {
(1.0)
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16
________________________________________________________ QUESTION 8.04 (3.00)
For each of the 3 cases below identify which, if anyt Tech Spec operational loakage limits are violated. Justify your answe c. 21 S/G 1eakage is 0.2 spm; leakage through RHR/RCS valves has increased from 2.0 to 3.0 spm; Primary PORV leakage is 3.0 spm; leakage from the packing of the manual valve upstream of the Letdown Isolation valve (LCV-459) is 2.5 spm; unidentified leakage is 0.4 spm. (1.0)
b. 23 S/G 1eakage is 0.15 spm; leakage through the RHR/RCS valves has increased from 0.5 to 3.5 spm; Primary PORV leakage is 2.5 spmi leakage from a downstream weld on the body of the manual valve upstream of LCV-459 is 0.4 spm; unidentified leakage is 0.2 spm. (1.0) S/G 1eakage is 0.25 spm; leakage through RHR/RCS valves has increased from 3.5 to 4.1 spmi PORV leakage is 5.1 spm; leakage from an upstream weld in the body of the manual valve upstream of LCV-459 is 0.2 spm; unidentified leakage is 0.2 spm. (1.0)
GUESTION 8.05 (3.00)
Answer the following questions concerning OAD-5, Procedure Adherence and Us a. What type of procedure (SDPrADI,etc.) requires that its steps be followed in their proper sequence ? What action is required befora deviating from a step sequence ? (0.75)
b. Under what circumstances is a written procedure document not required to be in active use ? What requirements exist under these circumstances ? (0.75) AD-5 specifically allows operators to depart from the intent of procedures, Tech Specs and other license requirements under certain circumstances. In one case prior approval is needed, in the other case only notification is necessary. Describe each case includin3 either the necessary notifications, or in the case where prior approval is needed, what lowest authority (by title) may give such approval? (1.5)
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 'PAGE 17
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GUESTION 8.06 (3.50)
Answer ~the following questions concerning the IP2 Emergency Pla a.~Per IP-1013, which one of the following sets of factors best represents the considerations of the SMS (ED) when making shelter and evacuation recommendations to the counties and N.Y. state. (0.5)
1) Plant conditions, Risk / Benefit to Public, Event class 2) Plant conditions, Duration of Release, Evacuation time 3) Release rattss, Duration of Release, Event class 4) Release rates, Risk / Benefit to Public, Evacuation time b. You, as the SMS have declared an alert due to plant conditions at IP Place the following EP actions in their proper sequence as you implement IP-1001 during off-normal hour (1.0)
1) Prepare a list of personnel who have called in 2) Prepare a message for beeper transmittal 3) Assign a CCR communicator 4) Authenticates original emergency phone call 5) Assign personnel to perform minimum job function per form 40 c. How would an increase in each of the following factors affect the thyroid dose (increase, decrease, no change) at 2 miles from the IP2
. site boundry? Briefly explain why. Consider each separately. (2.0)
1) I-131 concentration 2) Noble Gas concentration 3) Wind speed
4) Dilution factor (Xu/0)
,
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18
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GUESTION 8.07 (2.00)
L Answer the following questions concerning Tech Spec Instrumentatio o. Give two reasons for isolatin3 main feedwater lines upon actuation of the Safety Injection System? (0.5) Tech Spec Table 3-1 specifies the setting limits for ESF instrumentation In which reactor operating condition is the table applicabl (Select the best answer) (0.5)
1) Not in cold shutdown 2) Not in hot shutdown 3) Cold shutdown 4) Hot shutdown According to Tech Specs, the only single item required before an Instrument Tech removes the cover plate on the rear of the ESF panel is (0.5)
1) Tave < 350 F 2) No ESF instrumentation out-of-service 3) Safety Injection bypassed 4) Watch Supervisor permission 5) A signed off work order d. Which one of the following lists best represents the operational safety instrumentation covered by Tech Specs Sect. 3.5. (0.5)
1) Pzr pressure, Containment pressure, Turbine trip (overspeed)
2) Pzt pressure, Radiation monitoring, AFH flow 3) Pzt level, Containment pressure, AFW flow 4) Pzr level, Radiation monitoring, Turbine trip (overspeed)
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GUESTION 8.08 (2.50)
Answer the following questions concerning OAD-19, Tasout Lo3 Identify (by name or color and sizelnermal or small) the type of tag to be used in each of the following situations. (1.0)
1) Tags on the suction and discharge valves of a pump to be removed for bench testing 2) Tags on a pump only allowing it to be run during emergency situations 3) Tags on a CCR charging flow gauge 00S due to maintenance 4) Tags on a breaker supplying power to a motor operated valve to be removed for rebuildin b. Who (title) normally prepares tagouts at IP2 ? Who (title) independently verifies the tagout ? (0.5)
c. What 2 evolutions / operations require an independent check of the placement and restoration (removal) of stop tags. (0.5) Per the Caution in OAD-19, what action is necessary if an independent verifier finds a valve stuck on its open seat when it's supposed to be closed. (Select the best answer) (0.5)
1) Close the valve and notify the on-watch RO 2) Inform the on-watch RO 3) Inform the on-watch SWS 4) Close the valve and notify the on-watch SHS
'00ESTION 8.09 (3.00) If containment integrity is not met at Hot Full Powere then what 2 options does the SRO have for corrective action ? Time periods are not required. (1.0)
b. When is containment integrity required ? (0.5)
c. For each of the following situations, indicate (yeseno) if containment integrity is preserved. If it is violated, explain what would be necessary restore it. (1.5)
1) The equipment door is properly closed. Maintenance is repairing a crimped closed penetration pressurization supply line to the doo ) The inside door of the personnel air lock is properly closed. The outside door is wedged'open with a 2 x 4 to allow for a detailed inspection of the knife (door) seal ) The outside containment purge exhaust valve has the operator removed but the valve is close (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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t ADMINISTRATIVE ~ PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20
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____' ______________________________________________________ QUESTION 8.10 (1.50)
Diesel Generator 21's operability loa'd test, which is required every 31 days,'is scheduled for today. The last three tests were completed 36, 48, and 102 days-ago'respectively. The plant is at 100% power. Are
'
Technical Specifications being met? Esplain.why or why not.
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(***** END OF CATEGORY 08 *****)
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secon Conversions I curie = 3.7 x 108'dps I kg = 2.21bs .
1 gal = 3.78 liters -
1 gm/cm' = 62.4 lbs/ft'
1 ft' = 7.48 gal 1 in = 2.54 ca 1 yr = 2.15 x 10' se I gal. = 8.3453 lb MW = 3.41 X 10' BTU /HR y
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 21
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 5.01 (3.00)
a. DNB is the point at which a small increase in heat flux results in a
'larse increase in clad temperature or a significant reduction in heat transfer coefficient.(A. drawing of the heat transfer curve (Odot v delta T) would be acceptable if the above concept is included) (1.0)
b. 1) more 2) more 3) less 4) less (0.25 each)
c. The curves that were plotted were a direct result of a computer model of predicted DNB conditions. Since a computer model was-chosen there exists a reasonable uncertainty with a reasonable'probabil-ity that DNB may exist if no tolerance is allowed "4.e. DNBR=1.0) (1.0)
REFERENCE IP-2rThermo., pg. 9-33,9-37 IP-2rThermo., pg. 9-37,38,40 IP-2rTech. Specs., Sect. 2 Basis ANSWER 5.02 (2.25)
a.(For the plant to remain critical the positive reactivity inserted by MTC siust counter the negative reactivity due to Dopple Void reactivity is assumed to be negligible.)
(10% pwr)(-9.0 pcm/ %pwr)= -90 pcm (0.5)
(-90 pcm)/(-4 pcm/ F)= 22.5 F decrease (0.5)
Tave=524.5 F (0.5)
b. (100 ppm / % delta rho)(.090 % delta tho)= 9 ppm (0.5)
dilution (0.25)
No double jeopardy to part REFERENCE t IP2rThermo., pg.5-70,71
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 5.03 (3.00)
n. The heat sink must be at a higher elevation than the heat source. (1.0)
b. Initially, NC flow uill be reduced (0.2) due to less cold feed water entering the S/G causing less heat removal (0.2) which causes higher RCS temperatures in the U-tubes and reduces the thermal drivin3 head (temperature difference between U-tubes and core) (0.6)
c. The head contains a large mass of stored energy which can't be adequately cooled during.a rapid NC cooldown. (1.0)
REFERENCE IP2, Thermo., pg 9-34 IP2, ES-0.2, pg. 7 ANSWER 5.04 (2.50)
a. 2 (0.5)
b. Sm-149 has a smaller absorption cross section and therefore less reactivity worth than Xe (0.5), and does not deca lik (0.5).
(witt not JA posi+,ve y awayrenfledy )e Xe c. Case I would be a more noticeable Xe transient (0.2) because the local power changes occured rapidly with respect to Xenon's ability to maintain equilibrium with local powe (0.8)
REFERENCE IP2, Reactor Theory, pg. 6-14,20,21,23,24
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARDER, S.
ANSWER 5.05 (2.50) From the C-E Stm Tables, Tsat for 2200 psia = 649.5 F (0.5)
SCM= Tsat-Th= 649.5-580= 69.5 F (0.5) decrease (0.25)
Th increases as unit delta T increases with power (0.5) (0.25)
' Core delta T during natural circulation cooldown will approach full load delta That is greater than in the other 2 case (0.5)
REFERENCE IP2, Thermo, Sect. 2 IP2rVol. 1, pg. 14, RCS Sys. Descr.
ANSWER 5.06 (1.50) Stay the same (0.25)
Beta Bar's magnitude is strictly dependent on the concentrations of U-235, U-238, Pu-239, and Pu-241 and they change only with life and not with~ temperature. (0.5) EOL (0.25)
Since the beta of the core is smaller a larger SUR would result for'a given reactivity insertion. (0.5)
REFERENCE IP2, Reactor Theory eps. 4-14 to 4-18
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-. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 5 ~. 0 7 (4.00) ) decrease (0.3)10wer temperature reduces the' rods " sphere of influence" by reducing the number of neutrons the rod sees. (0.7)
2) Decrease (0.3)
Rods are moved from a high flux region to a low flux regioni rod worth decreases. (0.7)
3) Increase (0.3)
Decrease in boron increases the rods " sphere of influence" by increasing the number of neutrons the rod see (0.7) True (0.2) DRH is essentially independent of power as long as the relative neutron flux (local flux divided by total core flux) remains constant (0.8)
(will accept False as correct if accompanied by an explanation of temperature, Doppler, and Xe effects that change relative flux from HZP to power operation.)
REFERENCE IP2, Reactor Theory, pg. 7-19 to 7-21 and fig. 7.1 ps. 7-23,7-25 ANSWER 5.00 (2.25)
1) Criticality is uncertain, (0.25); Stop rod motion (0.5)
2) Reactor is critical (0.75)
3) Criticality is uncertain, (0.25); Stop baron dilution (0.5)
REFERENCE IP2, POP 1.2, step 3.17
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSHER 5.09 (2.50)
a. D (0.25) because A will reach critical rod height sooner than B (0.25)
and the closer Keff is to one, the longer it takes to reach a new steady state neutron level (mo're neutron senerations are required to achieve the larger population)(0.5), thus B will allow its neutron population more time to achieve a higher suberitical level than A (0.5).
b. Same (0.5), critical rod height is dependent only upon the reactivity characteristics of the core, not on neutron leve (0.5)
REFERENCE IP2 Rx Theory Text pp. 4-48 to 4-56 and pg. 3-22 ANSHER 5.10 (1.50)
5% decay heat = (0.05) 2758HH = 137.9 MH E0.13 Safeties lift at 2485 psis = 2500 psia E0.1]
Tsat at 2500 psia = 668 F E0.33 delta T required = 668 - 547 = 121 F CO.13
(1.2 Btu /lbm-F)(214,000lbm)(121F) = 3.10 X 10 Btu E0.33 137.9 MH + 15 MH = 153 HW (56.896 BTU / min /KW)(1000KH/HH)
= 8.7 X 10 Btu / min EO.33 7 6 3.10 X 10 /8 7 X 10 = 3.5 minutes E0.33 REFERENCE CE steam tables Tech Specs (Pzt safety setpoint and thermal power rating)
FSAR Vol 2 (prima'ry mass)
Thermo text ps 1-10 and 1-31 ( definition of Cp)
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. PLANT SYSTEMS DESIGN, CONTROLr AND INSTRUMENTATION PAGE 26
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSWER 6.01 (2.75) The affected non-return MS check valve would close(0.5) and the High MS flow SI signal must sense high MS flow on 2 MS lines.(0.5)
b. To minimize the' possibility of generating a false signal since initiation of containment spray (NaOH) would cause unnecessary corrosion (0.5) No (0.25)
Busses 6A and 3A must receive an SI signal and energize before the motor driven pumps start. (0.5)
d. Safeguards actuation sequence finished ( 0. 25 e a .K 2 r ey'c()
SI manual reset PB's depressed 2 e.w d e. +be- .t Lous pas s vr< pnhin REFERENCE Ac Autt xcys o. IP2, Vol. 10, pg. 9 footnoter ESF Sys. Desc b. pg. 12 c. pg. 22,23 d. p , M $ D 2 8 ANSWER 6.02 (3.50)
Icpulse-Las unit On a load changer it senses the rate of change of the difference of the 2 inputs.Its output reflects the magnitude cf the rate of change of Pimp and NI powe (Steady state differences result in zero output) (0.5)
Initiates (Anticipates) a fast response to a change in load (0.5)
Non Linear Gain unit: Converts the output of the Impulse-Las unit to a temperature error.(0.5)
Low gain is used for small (<1%) mismatch and higher gain is used for larger mismatches (0.5)
Initiates rod motion more quickly with a larger change in power. (0.5)
Variable Gain unit Selects a higher gain at a lower power and vice vers (0.5)
Reactivity insertions at a high power have larger effect than those at low power. (0.5)
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, REFERENC e. IP2r Vol. 15, pg. 84-87, Core Design and Control Sys. Desc ANSWER 6.03 (1.50) Prevents thermal shocking of RC pipins due charsins flow. (1.0)
b. Protects demin resin (0.5)
REFERENCE
' IP2rVo , ps. 5, RCS Sys. Desc b. ps. 11 ANSWER 6.04 (2.50)
a. Loss of seal injection (0.5)
b. Check valve upstream and motor operated valve downsteam with connecting pipins are rated for RCS pressure. _Ll. M ( o. r )
High flow is sensed on a flow element which closes the motor operated outlet valve. (0.5) Ac he f valv e,. (0. r) Establishs the additional lower radial bearins cooling flow needed at low RCS pressure. (0.5)
REFERENCE a. IP2, Vo , pg. 33, RCS Sys. Desc b. Ps. 46,47 c. ps. 50
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ANSHERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 6.05 (2.00) Low-Low level (0.25) on 2 of 4 S/G's (0.25)
Loss of outside power concurrent with a Main TG trip (provided an SI signal doesn't exist) (0.5)
b. Recire flow is established.( SOV-1321 opens to open the recire valve)
(0.25)
A timer is started.( Trips pump if 55 spm is not established w/in time period) (0.25) Yes, if the timer times out and flow is less than 55 spm , the motor driven AFH pump will trip. (0.5)
REFERENCE IP2, Vol. 21, pg. 70, FH Sys. Desc b.ac. pg. 71 ANSWER 6.06 (1.75) NaOH (0.25)
Limits offsite thyroid doses (to less than 10CFR 100 guidelines) by trapping iodine (inorganic) into a liquid phase. (0.5)
Increases sump pH (to about 8.5)'to minimize general corrosion due to boric aci (0.5) Spray pumps will start r valves will stroke to thef- SI position (discharge valves open). Spray is initiated (0.br REFERENCE a. & IP2, Vol. 10.2, pg. 5,16, Containment Spray Sys. Descr.
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ! ANSWER ~6.07 (3.75)
oR a. UV on busses 2 A r 3 A r 5 A ;wdT 6 A . (0.25)
'
SI signal (0.25)
.
l b.'Emersency stop button (0.25 ea.)
Overcurrent relay.(51V)
Reverse power relay (32) lower (0.1 ea.) Loss of EDG causes a loss of the alternate AC feed
' same ( MCC 26A)and the battery charger output to' Battery on 21. Battery voltage will decrease due to' loadin off Inverter output voltage remains the same since the off it will put out 118 VAC for any DC input between 105 VDC and.140 VDC. The inverter is still carrying the load from its DC supply so PL1 is on. .Since the alternate feed is deenersized the In-synch light is
,
off. PL5 is off since MCC 26A is deenergize ( 5-items 0 0.4 ea.)
REFERENCE IP2r Vol. 27.1, ps. 71-73, Electrical Sys. Descr. and Fis. 1 ps. 96 a.ab. IP2, Vol. 27.3, pg. 37, Diesel Generator Sys. Descr.
ANSWER 6.08 (2.25)
- Sprays initiate ( 2260.to 2310 psis ) (0.25-ea.)
Primary.PORVs open ( 2335 psis-)
High pressure reactor trip ( 2385 psis )
Safety valves open ( 2485 psis )
b. All heaters off (18%) (0.25 ea.)
Letdown isolated (18%)
Low pressure SI o r- c La gh <} pM 5pu he /A c t-M 4<- -
c. A.larse insuse into the pressurizer causes it to become.subcooled. The heaters come on to ensure the ensuing.outsurge will be capable of ,
. flashing to' steam. (0.5)
REFERENCE l' IP2r Vol. 1r ps. 82, RCS Sys. Desc b. Es. 71 .
l . c '* P3 69-70
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSWER 6.09 (3.00) (1)- The reactor will trip on SR high flux [0.53 due to P-6 dropping out.and reenergizing SR Nuclear Instruments (NI). [0.53 (2) The reactor will not trip E0.53 due to P-10 which prevents reenergization of the SR NI even if P-6 drops out. [0.53 The Defeat P-6 Pushbuttons.must be depressed [0.53 in order to allow the SR NI to reenergize. CO.53 REFERENCE IP2 SD 13 pg 60,61 Fig 13,27,32 IP2 SD 28 pg 22 23 ANSHER- 6.10 (2.00)
a. IP1 Superheater Blds. (0.5)
b. Service water pumps Charging pumps SI pumps '
RHR pumps CCW pumps AFW pumps MCC 27 (any 6 for 0.25 each)
REFERENCE IP2 SD 27.1 pp. 86-91 and Fig. 32-36
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. PROCEDURES -~ NORMAL, ABNORMAL, EMERGENCY AND PAGE 31
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 7.01 (2.50)
- A b. 4 c. JRtf( FA L S E (,P-7- it san 5[/c) wken few u- 4 108) FALSE
. REFERENCE a. IP2,E-Orps. 1, Reactor Trip /SI
.b ~. pg. 2 ~ IP2,E-0,ps. 1, Reactor Trip /SI IP2,Vol. 28,ps. 35, Unit Protection-Sys. Desc d.' IP2,ES-1.1,ps. 5,SI termination o. IP2,E-1,ps. 8, Loss of Reactor / Secondary coolant
,
ANSWER 7.02 (1.75)
a. PR > 5%. (0.25 ea.)
IR SUR not negative b. Start 1 SI pump (0.25 en.)
Block low pressure SI as necessary
.
'Depressurize the RCS to 1400 psis REFERENCE a. IP2, FR-S.1, pg. 3, ATWS EDP b. pg. 4 i
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.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSWER 7.03 (2.00)
a. Rod worth is insufficient to significantly, affect core peaking factors at higher core height. (0.5) (u.hll a cc e pt R P I .w e-c e W <7 b. Thermocouples (0.25 ea.) h p<-)h/
cr-edd-Incore detectors Lift coil disconnect switches (0.5)
Unaffected rods' switches are opened (0.5)
REFERENCE a.-IP2, Tech. Spec. 3.10.5, Rod Misalignment IP2, SDP 15.5, pg. 1 IP2, A 16.1.1, pg.3, Dropped or Hisaligned Rod AOP ANEkER 7.04 (2.00) To prevent dangerous Hydrogen concentrations from building up inside containmen (0.5)
b. L' r. d c r Od;cr;; c c r, t c i r.c c r. t c a r d i t : c r; c . '^
.5} O Re (>m'<-in N M ( ) .L+e-e-f a RCP (c. SD
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..se r. 3 rigg ygig exiet _ ir:3 yill der ecruize th" cyrte : but # 12 c h i r;3 i r- the head uill h eld r ecture er .
S i r:c e ne # 12 r 51 :; 3 crey c *er 32cerer veld -> cigni#icant r-essvae-der erce ure'd be ebce-"ad de-ing cr-'ying er=*s+4nne - 't ni REFERENCE g.I,3 Y4+ -(b t R d [.co- a Fed- IP2, , . p ;.; . 7-9- c i d s i r. "" S e e d Afo c[, 2 p3 12 ,ff-ep f, ([
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. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33
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ANSWERS -- INDIAN POINT 2
~
-85/12/09-BARBER, ' ANSWER 7 05 (2.00)
'
. a. Ensures MTC is not positive (0.5)
Prevents unnecessary power excursions due to temperature increases during heatup. (0.5)
, b.; Prevents exceeding NDT limits' ( Fracture toughness concern ) (0.5)
,
c. RCPs provide the necessary RCS heat input ~(0.5)
REFERENCE
,
a. IP2, POP 1~1, . pg . :10, Plant Heatup procedure
,
b '. ; I P 2 , POP 1.2, pg.2, . Reactor startup procedure j .c. IP2r Tech. Specs.'3.1.c
ANSWER 7.06 (2.75) That plant conditions are stable and are not planned to change (0.25)
at the job site (0.25) thus changing radiological. conditions (0.25)..
- . b. Stop work and notify radiation' protection (0.5)
Radiological conditions differ than those expected during pre-job planning. (0.5)
c. 4,1,2,3,5 (0.2 ea.)
1EFERENCE a, IP2, SAO-302, pg. 7 b.'IP2r SAO-303r.pg. 8,'ALARA Program, Sect. 4.10 c. IP2r Rad Worker Retraining Program, Sect.-18 i.
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND~ PAGE 34
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 7.07 (2.75)
'
a. Decrease (0.25) p, gpj hel a55Mffef The gripper geuld dicen320e # ce the #uel 2ccembly caucing it te #211 cnte the ' ( It would be in a position other than assumed for fuel handling accidents') (0.5)
b. Cuppl, "n ctep VC. pqe, Sa[fI y (wkst valses clogc ' " S ^
.25cc.)[o,_T3)
Sui r 1y d; .7:: =1 == = t/c presSm n }lz? f,ke. a lxc elege, (o,33)
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UD WC'bW .ON m GMs C c. Isolate containment ventilation (Any 4 0 0.25 ea.)
Suspend all fuel handling operations Evacuate containment Monitor R-14. If~ release cccurs refer to EP Document book for classification REFERENCE a. IP2r Vol. 17, ps. 21, Fuel handlins Sys. Desc b. ps.10 c. IP2r A 17.1, ps.1, Irradiated Fuel Damage in the Reactor Cavity ANSWER 7.08 (3.00)
a. clad (0.25 ea.)
RCS containment Containment is bypasse b. 1) Unexpected rise in S/G NR level (0.25 ea.)
2) High S/G sample radiation level 3) High steam line radiation 4) High S/G blowdown radiation i c. Since the S/G is a. saturated system, isolation of the S/G at a saturation temperature with a pressure > than the steam dump and se PORVg setpoint would cause an unnecessary release to the public. (1.0)
REFERENCE or to h }I a.& b.r IP2, E-3, Steam Generator Tube Rupture c. ps. 5
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- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND' 'PAGE 35
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ANSWERS --' INDIAN POINT 2 -85/12/09-BARBERr ANSWER 7.09 (2.25) Trip the reactor (by either tripping the turbine; tripping the reactor lo'cally or. Prior to leaving the CCR), announce CCR evacuation over PA, take n' h^^'", .it;I ' /;r SRO lose Des
. CCP c r. d ^C"';. p /a r
/g M e. N ws f o ), ree o 3 2 g c)3
~".2 / adi Proceed to the local pressurizer level and pressure control station (0.4) ~
Fest
, b.VRO (0.25)gs M'* gfg 3 ,.f - 97- yy
"cr:i' ':- i..3- S/G 1evel -t c c r. : :; r t-- A F W h m a l ii i, 17, ; n s p r o p e r level (tr6-66%)
!. (0.25) at the-Aux Feed pump Blds. local S/G level control panel (0.25). Emergency borate (0.5)
REFERENCE' A 2 /, /.</ pg . j23 IP2r ^-i ^ C ; . i . _1 _ C _ ._,. ...miity ;- i t .- um;;ce "curr ^vn rie
.
b. pg .ilf
' p ANSWER 7.10 (1.50)
i The steam driven AFW pump provides feed (0.5) and the sec. PORVs provide steam relief to promote natural circulation. . (0.5)
b.. Minimize RCS inventory loss (and shrinkage due to cooldown) (0.5)
REFERENCE g~
a. IP2r A-4-B, ps. 4r Loss of Normal and Emergency Power 4 b.!Ps. 7 ANSWER 7.11- (1.50) Suberiticality b. Core cooling c. Heat sink d. Integrity
! a. Containment f. Inventory (0.25 each)
REFERENCE
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IP2, ERG Status Trees
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PAGE 36
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- RADIOLOGICAL CONTROL-
,_____________'_______
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' ANSWERS -- INDIAN POINT /12/09-BARBER, S.
,
ANSWER 7.12- (1.00)~
_RHRcirrja-:t-ion flowrate must be above a specified valu (1'0)
.
REFERENCE ~
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-IP2r ES-1.3, pg. 10, Transfer to cold les recire i
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ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSWER 8.01 (2.50)
n. The 2735 psis safety limit was only imposed with FA in the core. The hydro was done without the core loaded. (1.0) To assure that the values assumed in accident analysis are not exceeded durins normal plant operations. (0.5)
c. Decrease (0.25)
The peaking factor incorporates uncertainties which when eliminated would mean that there would be less error in determining the actual incore peak-to-average flux or power distribution. (0.75)
REFERENCE IP2, Tech Specs Sect. 2.1 ANSWER 8.02 (2.00) Yes (0.5) fire brigade (0.25)
4-safe shutdown (0.25)
Nor operators required for safe shutdown must be excluded from the fire brigade (0.5) (0.5)
REFERENCE IP2, Tech Specs Sect.6 ANSWER 8.03 (2.00) Limit RCS cooldown rate (0.5) and reactivity insertion followins a MSLB. (0.5)
pullout b. Place AFW pump in 2;..vol (0.4)
Within time limit (1hr), restore CST lineup or open city water supply to AFW pumps and return-them to auto. (0.6)
REFERENCE IP2, Tech Specs Sect 4 7 basis b. IP2, Tech Specs Sect _
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k b a ADMINISTRATIVE-PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38
__________________________________________________________
ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
..
ANSWER 8.04 (3.00)
a. No criteria are violated (1.0)
b. RHR/RCS leakage increase exceeded margin to 5 spm limit by more than 50%
(0.5)
(3.5-0.5)/(5.0-0.5)= 0.66 or 66% (0.5)
c. O spm through an RCS pressure boundary (0.5)
upstream valve body leakage is unisolable (0.5)
REFERENCE IP2, Tech Spec 3.1. ANSWER 8.05 (3.00) E0P (0.25)
Ensure prerequsite steps are performed prior to deviating from the step sequence (0.5)
b. Routine and repetitive evolutions (0.25)
Precautions, limitations, and prerequisites must be either previewed os memorized prior to the evolution. (0.5)
c. Operators are allowed to deviate from procedures whenever neccessary for the prevention of injury to personnel or to the publici or damase to the facility (0,5)
Chief Ops Engr (Mgr. of Ops), Gen Har, or VP Nuc Pwr to be notified after the event (0.25)
Operators may depart-from Tech Specs or from a licensed condition if no action consistent with these documents provides adequate or equivalent protection of the public health and safety and is immediately apparen (0.5)
SRO approves prior to taking action (0.25)
REFERENCE a. a b. a IP2, OAD-5, pg. 2, Procedure Adherence and Use
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. ADMINISTRATIVE PROCEDURES, CONCITIONS, AND LIMITATIONS PAGE 39
ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, S.
ANSHER 8.06 (3.50)
a, 2 (0.5) ( ustil acc e(' f 3 hs- c *Y NSA W NW L k Y 'IO' ,4,2,1,5 (0.2 ea.) (LJ.*ll a u e p t 2 M 9 8% W#h OM^r SE *
' # I'YCIf f *"O'Y'W #5 N'E 1) Increase, higher concentration of radioactivity will reach the thyroid 2) No effect, noble gases don't concentrate in the thyroid 3) Decrease, higher wind speed disperses the activity throughout the atmosphere 4) Increase, a larger number indicates less dilution in the atmosphere (0.1) Direction (0.4) Explanation REFERENCE a. IP2, E-Plan, IP-1013r pg. 1 b. IP2, E-Plan, IP-1001, ps. 5,6 c. IP2, E-Plan, IP-1007, pg. 5 ANSHER 8.07 (2.00) To prevent excessive cooldown of the RCS (0.25)
Reduces consequesnees of a steam line break (0.25) (0.5)
c. 4 (0.5) (0.5)
REFERENCE IP2,-Tech Specs Sect. ,
.
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m ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 40
ANSWERS -- INDIAN POINT 2 -85/12/09-BARBER, ANSWER 8.08 (2.50) ) Stop.or black with a diagonal white stripe (normal size) (4 0 0.25 ea)
~)
2 Caution or yellow (normal size)
3) Stopesmall size for gauges 4) Stop (normal size) On-watch R0 (0.25 ec.)
SWS Operating orders (0.25 ea)
Tags removed from safety related equipment (0.5)
REFERENCE IP2, 0AD-19, Tagout Los ANSWER 8.09 (3.00) Restore containment integrity (4 hrs) OR (0.5 ea.)
Go to cold shutdown When ever reactor conditions are other than cold shutdown (0.5) ) No, must restore PP supply to the equipment door (0.5 ea.)
2) Yes 3) Yes REFERENCE IP2, Tech Specs, Sect. IP2, Tech Specs, Sect. ANSHER 8.10 (1.50)
No CO.33, each test is within 25% of the required time interval E0.63, but the three consecutive combined test intervals exceed 3.25 of the required interval 00.6 REFERENCE Tech Spec 1.10 pg 1.4
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e ATTACHMENT 2 FACILITY COMMENTS ON WRITTEN EXAMINATION AND RESOLUTIONS 6.01d Comment: Additional answers _are possible since question did not
,
specify whether SI had already been initiate Resolution: Three additional-answers were added based on System Description 2 ,01c Comment: Answer is FALSE since P-7 is considered " satisfied"'
when power is less than- 10%.
Resolution: Accepted. Based on common usage of the term
" satisfied."
7.02b Comment: Question asks for knowledge of steps in a Response Not Obtained column of an E0P. These are not required to be memorized according to Westinghouse Owners Group Emergency Response Guidelines Background Document (Low Pressure Version).E-0, Step 1, Note Resolution: Not accepted. Background Document for FR-S.1, Step 1, Note 1, states: "... since time is a critical factor-during a full power ATWS event, manual actions to initiate emergency boration of the RCS is included as an immediate action." In view of the necessity for timely action in the above case, a competent operator should understand and be able to recall from memory the three basic steps asked for by-this questio ' 7.03a Comment: RPI_ inaccuracy should be accepted as an answe Resolution: Will accept RPI inaccuracy for partial credit in accordance with Technical Specification 3.10-1 .09a_
and b Comment: Procedure A-5-A was replaced and has slightly different operator action ,
Resolution: Answer is changed to reflect current reference, j Procedure A27.1.9, pages 2- .04 Comment: Question is triple Jeopard Resolution: Not accepted. Question tests three different criteria for RCS leakag &
. , , _ ._ , _ . , _ _ _ . . - _ . _ _ _ . - , _ 4
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.t 9 J -Attachment 2 2 8.06a Comment: Answer #3 should be accepted based on the IP-1013 flow char Resolution: Accepte .06b Comment: Procedure requires memorization of steps of an E-Plan implementing procedure and is not required knowledg Resolution: Not accepted. The steps of:the procedure are given in the question. The answer requires their proper-ordering. This requires a knowledge of the procedural intent and how it is to be accomplished, which is-required knowledge. Will accept a reversal of items 3 and 4 in the answer based on the likely possibilit that the steps could be performed in that orde '!
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