IR 05000336/1986010
| ML20214T729 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/15/1986 |
| From: | Coe D, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214T724 | List: |
| References | |
| 50-336-86-10OL, NUDOCS 8609300389 | |
| Download: ML20214T729 (138) | |
Text
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION
REGION I
EXAMINATION REPORT NO. 86-10(OL) FACILITY DOCKET NO.
50-336 FACILITY LICENSE NO. DPR-65 Licensee: Northeast Nuclear Energy Co.
P. O. Box 270 Hartford, Connecticut 06141 FACILITY: Millstone 2 EXAMINATION DATES: July 7-11, 1986 CHIEF EXAMI.ER: - Dougla's Coe, Mead ~ ctor Engineer date (Examiner REVIEWED BY: ) / 7I/% Roti
, Chief, Projects Branch #1 date APPROVED BY: f3 b HdrDi(Qr, Chief, Projects Branch #1 dat'e / SUMMARY: Eight Senior Reactor Operator (SRO) and seven Reactor Operator (RO) licensing examinations were administered.
Seven SRO and five R0 licenses were issued. Two candidates failed the simulator examination and one candidate failed the written examination.
8609300389 860922 PDR ADOCK 05000336 V PDR
REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: l R0 l SR0 l 1 Pass / Fail l Pass / Fail l I I I Written Exam I 6/1
7/0 l I I I Oral Exam l 7/0
7/1
I i i Simulator Exam I 6/1 l 7/1 l l
I Overall l 5/2 l 7/1 l CHIEF EXAMINER AT SITE: D. Coe, NRC OTHER EXAMINERS: R. M. Keller, NRC J. D. Smith, PNL (contractor) J. W. Upton, PNL (contractor) 1.
Summary of generic strengths or deficiencies noted on operating exams: Senior operator candidates were noted to be weak in the following areas:
' a.
Use of plant paging system to inform plant personnel of significant changes in plant status which may require coordinated actions outside the control room, such as for a reactor trip, or high airborne activity.
b.
Interpretation of Technical Specifications and actions required by them.
No generic deficiencies were noted for operator candidates.
2.
Summary of generic strengths or deficiencies noted from grading of written exams: Operator candidates were noted to be weak in the following areas: a.
Understanding the inability of safety injection alone to prevent core damage with a failed open primary safety and maximum decay heat.
l . _ . _. .. _- -_ . _ -. -. __ -
-
b.
The purposes / functions of the HPSI system during a Main Steam Line Break.
c.
The reason for limiting the current to Bus 24E from Unit 1 RSST.
d.
Operation of Diesel Generator control transfer switches.
e.
The basis for the maximum time allowed for AFW auto-initiation.
Senior operator candidates were noted to be weak in the following areas: a.
Effect of raising Tave on ASI.
b.
Regulating control rod interlocks.
c.
Depressurizing the RCS with no pressurizer spray flow available.
d.
Preferred source of RCS makeup for shutdown from outside the control room.
e.
Basis for maintaining RCS pressure below 265 psig when placing CVCS in service during a plant startup.
3.
Personnel Present at Exit Interview: NRC PERSONNEL , D. H. Coe, Chief Examiner G. Grant, Resident Inspector Facility Personnel W. Romberg, Station Superintendent H. Haynes, Manager, Operator Training J. Heg, Unit 2 Operations Assistant M. Wilson, Supervisor-Unit 2 Operator Training J. Becker, Assistant Supervisor-Unit 2 Operator Training R. Spurr, Senior Simulator Instructor G. Bender, Senior Instructor 4.
Summary of NRC Comments made at exit interview: The examiner expressed the opinion that overall candidates performance during the operating examination was above average.
The following NRC concern regarding the simulator was raised: An ESAS activation followed by a Loss of Normal Power will cause the - EDG's to tie on to the emergency bases without any load shedding or sequencing of running ESAS equipment. The licensee stated that
their investigation of the simulator's response showed that the simulator did not react as would the actual plant.
This was due to a simulator programming error.
Due to the potential safety concern that arises if the plant response were the same as the simulator, licensee should determine that actual plant response would provide for Emergency Diesel Generator load sequencing under the same conditions.
The following NRC recommendations regarding training material were made: Some system descriptions need to be updated to include current plant configuration. Many of these are difficult to read due to unlabeled or mis-referenced figures and diagrams.
In general, the material provided was well indexed. The NRC noted that an upgrading effort is currently in progress.
The following NRC-concerns were raised regarding the facility's procedures: a.
Note-2 of the Minimum Required SI delivery curve (in the E0P's) is confusing and caused misinterpretation during the examina-tion.
This will be reviewed by NRC Headquarters.
b.
A0P 2558, Emergency Boration, Step 4.8 wording is not consistent with E0P 2525 page 3 Note. This refers to the number of minutes of required boration for stuck CEA's. This has been corrected by the licensee's Change 2 to sevision 1 of E0P 2525.
c.
E0P's 2532, 2534, and 2537 Fig. 4.1, are missing a "<" in the topmost decision box causing one candidate to initially mis-diagnose a primary leak inside containment as a steam break inside containment. This has been corrected by the licensee's Change 1 to Revisien 2 of E0P's 2532 and 2534, and Change 2 to Revision 2 of E0P 2537.
d.
A typographical error exists in OP2204, Rev. 7, Page 6 (first caution) which results in an incomplete reference to Technical Specifications. This has been corrected by the licensee's Change 6 to Revision 7 of OP-2204.
Attachments: 1.
R0 Written En nination and Answer Key 2.
SRO Written Examination and Answer Key 3.
Facility provi 'ed comments on written examir.ations 4.
Resolutions of Facility comments on written examinations -
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _. ATTACH /G Wr IMASTER U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MILLSTONE 2 _________________________ REACTOR TYPE: PWR-CE _________________________ DATE ADMINISTERED: 86/07/07 _________________________ EXAMINER: COE,0.
_________________________ APPLICANT: _________________________ INSTRUCTIONS TO APPLICANT: ___-______________________ Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each cctegory and a final grade of at least 80%. Examination papers will be picked up sin (6) hours after the examination starts.
, % OF l CATEGORY % OF APPLICANT'S CATEGORT l VALUE TOTAL SCORE VALUE CATEGORY ________ ______ ___________ ________ ___________________________________ _I'5 00_I_____'S
I_1__ l ___________ ________ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
, _
__ _
___________ ________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _'5 00_'_1____ _['5 00 ___________ ________ 3.
INSTRUMENTS AND CONTROLS _i__ _'5.00'_______['5.00 ___________ ________ 4.
PROCEDURES - NORMAL, ABNORMAL, _ ____ 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 aid.
5PPE5C55TI5~55555TURE~~~~~~~~~~~~~~ t
_ - _ _ _ - _ _ _ _ _ _ _ - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ l.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isEss557sisics-sE3i isEssFEE Es5 FEUi5 FE5s ____________________________________________ QUESTION 1.01 (.50) Which one of the following descriptions best supports the reason why Xenon reactivity increases sharply after a trip following 1000. hrs.
of operation at 100% power ? (0.5) A) Xenon decays less rapidly due to a reduction in the neutron flux.. B) Iodine half-life is much shorter than Xenon half-life.
C) Iodine production is greatly reduced and Xenon producticn is greatly increased due to the reduction in neutron flux.
D) Due to reduced neutron absorption, Iodine concentration increases, and Xenon decays directly from Iodine, thus Xenon increases.
! OUESTION 1.02 (.50) i Which one of the following conditions would cause a 1/M curve to predict [ criticality earlier than it will actually occur (conservatively)? A.
Source located too near detector.
D.
Source located too far from detector.
C.
Initial fuel assemblies loaded near the detector.
D.
Initial fuel assemblies loaded near the source.
QUESTION 1.03 (2,50) Two identical reactors are taken critical using continous rod withdrawal.
Reactor A has a rod speed of 30 inches per minute and reactor B has a rod speed of 15 inches per minute.
a.
Which reactor will have the highest source range counts at criticality? Why? (1.5) b.
How will 10E-4% critical rod heights compare in the two reactors? Explain briefly.
(1.0) (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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. . . - .. .~ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isiss557sisiCi-sEEi isissFEs Es5 FEUi6 FE5s ____________________________________________ QUESTION 1.04 (2.00) Indicate at which time in core life (BOL or EOL) the following accidents are more - severe (result in a longer time spent at higher power).. Briefly explain why by considering the effect of BOTH HTC and FTC and ' indicate which is dominant, a.
Main steam line break E1.03 b.
Total loss of coolant flow C1.03 GUESTION 1.05 (3.25) During a reactor startup and after the reactor is critical the Wide Range Los Channel indication is observed to increase from 10E-5% to 10E-4% power in 90 seconds with no rod motion.
, If the effective delayed neutron fraction is 0.005 and assuming an ,~ average neutron precursor decay constant of 0.08 sec-1, how much reactivity was added after criticality? ,
QUESTION 1.06 (3.00) Compare the CALCULATED Estimated Critical Position (ECP) for a
startup to-be performed 4 hours after a trip from 100% power, to the ACTUAL control rod position if the following events / conditions occurred. Consider each independently. Limit your answer to HIGHER than, LOWER than, or SAME.as the ECP.
a.
An inadvertant RCS dilution has been occuring for the last 4 hours.[0.63 b. The startup is delayed until 8 hours after the trip.
CO.63 c.
The steam dump pressure setpoint is increased to a value just below the Steam Generator saftey valve setpoint.
E0.63 d. Pressurizer pressure is lowered by 100 psia.
[0.63 e.
All Steam Generator levels are being raised by 5% as the ECP is reached.
[0.63 q d (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isEER557sssiCE-sEAi iEALEFEE As5 FE5i5 FE50 ____________________________________________ QUESTION 1.07 (2.50) The reactor is in the process of being started up. Power is at 10E-4% and CEA's are in manual group control.
A malfunctioning steam generator safety valve lifts, causing an increase in steam flow to about 6% of total plant capacity.
With no operator action explain HOW (increase, decrease, or remain the same) and WHY each af the following parameters will change immediately following the malfunction. Calculations are not necesr.ary.
d55R m e. a oc3oiM (_ M TC, a.
ria in steam pressure E0.83 b.
Primary Tave CO.8J c.
Reactor power CO.9] OUESTION 1.08 (1.00) a.
What would pressurizer relief valves discharge temperature be if quench tank pressure is 5 PSIG, there is a steam bubble in the pressurizer and RCS pressure is 2200 PSIA? [0.5] b.
If quench tank pressure is 5 PSIG and RCS pressure is 1050 PSIA, will pressuriner relief valve discharge temperature be GREATER THAN, EQUAL TO, or LESS THAN that in part a? [0.53 (***** CATEGORY 01 CONTINUED ON NEXT PAGE xxx**)
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5' --- isEEs557sEsiEs-REsi isssEFEE Es5 FEUi5 FE5s
.---
i QUESTION 1.09' (2.50) Answer the following questions assuming that a reactor trip has just occured from extended 100% power operation and that all plant systems function as designed.
Assume the trip was an inadvertant manual trip.
a.
In which direction should pressuriner pressure change (increase or decrease) and to what 100 psi range of values should it change before starting to return to normal? E0.53 ' b.
After stable shutdown conditions are achieved following the reactor trip, a primar.y safety valve sticks'open.
Reactor Coolant Pumps are tripped according to procedure and primary pressure drops to 1000 psia.
1) Would you expect ACTUAL pressuriner level to increase, decrease, or remain constant? Why? EO.53 2) If natural circulation conditions were established, and CET's read 585F, which of'the three types of natural. circulation conditions most probably exist in the core? CO.53 3) If natural circulation flow were not attained, would full safety injection and charging flow be sufficient to prevent core damage? Why or why not? [1.03 00ESTION 1.10 (2.25) There are THREE primary parameters that affect DNB and can be controlled by the reactor operator.
Do not consider core flow or flux distribution.
Briefly explain how each of-these parameters affect the margin to DNB.
QUESTION 1.11 (2.00) Would fuel center line temperature INCREASE, DECREASE, or REMAIN THE SAME in each of the following situations? BRIEFLY EXPLAIN WHY.
a.
Power decreases with constant Tave.
[0.53 b.
Tave increases with constant power.
CO.53 c. Core age increases with constant power.
CO.53 d.
Pr'essurizer pressure increases with constant power.
[0.53 (***** CATECORY 01 CONTINUED ON NEXT PAGE *****)
. .1. _ PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isEER557sARiEs-RE47 iEAssFEE As5 FEUi5 FE5s ____________________________________________ 00ESTION 1.12 (3.00) The plant is in a Natural Circulation Mode of core cooling.
As the fission product heat decays away, describe how and why you would expect the ' followins RCS parameters to change. (Assume that S/G pressure is being maintained at 900 psia).
a.
Teold 'EO.753 6.
Thot E0.753 c.
Core delta T E0.753 d.
Loop. transit time E0.753 (****x END OF CATEGORY 01 *xxx*)
2+ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
QUESTION 2.01 (1.00) Describe how the Vessel Seal Leak Detection System is designed to monitor leakage.
QUESTION 2.02 (1.00) , What. feature (s) ensure that the RWST cannot be siphoned below the minimum Tech Spec level if. a rupture were to occur in the RWST recirculation system piping? Consider ruptures in both the suction and discharge piping of the recirculation pump.
- GUESTION 2.03 (3.00) Answer the following questions regarding the design of the Engineered Safety Features System: a.
The design purpose of the HPSI system during a Design Basis Large Break LOCA is to (PREVENT / MITIGATE) damage to fuel pin cladding.
[ Choose one.
C0.53 ' b.
In addition to core cooling and inventory maintenance, what other purpose does the HPSI system serve during a hain Steam Line Break?' [1.03 c.
Describe the paths that reactor decay heat thermal energy would take immediately following a post-LOCA Sump Recirculation Actuation Signal.
Start with the core and continue to the ultimate heat ' sink.
E1.03 " d.
List two means of removing Hydrogen from containment following a LOCA.
For each method, indicate the name of the component or system which is used.
CO.53 i
(***** CATEGORY 02 CONTINUED ON NEXT PAGE
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________ QUESTION 2.04 (3.50) Answer the following questions regarding Baron Precipitation Cnntrol (BPC): a.
When is BPC used and why is it necessary (two reasons)? [1.0] b.
Sketch one-line diagrams of both the normal (Prefered) and alternate BPC flow paths showing components (other than instrumentation) and power operated valves directly in the flow path.
Valve designations are not required, but indicate if a valve has a motor, air, or other power operator.
Also, indicate the location of containment penetrations.
[2.5J QUESTION 2.05 (1.50) Answer the following questions regarding the Containment Air Recirculation and Cooling System (CARCS) and its associated support systems.
the -'KCS cA o after an SIAS? Assume a.
What automatic actions should occur to an initial normal lineup.
[0.6J b.
If 100% is defined as the heat removal capacity necessary to limit containment pressure to less than design pressure following an accident, what is the total heat removal capacity, in percent, of all CAR fans? E0.4J c.
What instrument system uses the Containment Air Recirculation and Cooling System as a means of normal cooling? E0.53 QUESTION 2.06 (1.00) Answer the following questions regarding the Emergency Diesel Generator (EDG) System a.
What minimum number of EDG starts can be provided bv the starting air flasks assuming they are initially full and are not recharged? E0.5J b.
How soon should the EDG be at rated speed and voltage following an auto start? E0.5J (xxxxx CATEGORY 02 CONTINUED ON NEXT PAGE 2x***)
__ 2.
P.LANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
, _______________________________________________________ QUESTION 2.07 (2.50) Answer the following questions regarding the operation of Bus 24E.
a.. List three-loads'which are supplied directly from Bus 24E (other " than other busses).
[0.6] b.
Describe the Reactor Safety function.of the Kirk Key interlock associated with Bus 24E.
CO.93-c.
If Bus 24E is supplied from its backup source, a limit is placed on i the load current supplied by that source. What is the reason for this limit? [1.03 QUESTION 2.08 (3.00) Answer the following questions regarding a loss of instrument air (assume normal, at power initial conditions): . a.
How would a complete loss of instrument air outside containment immediately affect the following components / systems. Choose ONE of the following for each component / system.
fail open/ flow maximum fail closed / flow stopped fail as is/ flow cannot change no effect/ system functions normally hain Feedwater Regulating Valves Pressurizer spray valves S/G feed pump control-Letdown Atmospheric Dump Valves AFW flow control valves EDG service water supply valves , hSIVs CO.25 each] I b.
Describe two means of interconnecting the IA system with backup sources of air cressure. Indicate automatic setpoints, if any. E1.0] ' GUESTION 2.09 (2.00) List three TYPES of ESF related components which are cooled by RBCCW immediately following an SIAS, and state IF and HOW any ESF function could be affected.upon a loss of RBCCW to each. Consider each component type separately.
(***** CATECORY 02 CONTINUED ON NEXT PAGE
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________ QUESTION 2.10 (1.00) What design feature of the clean liquid rad waste system ensures that waste additions are not made to a monitor tank which is being discharged? QUESTION 2.11.
(2.00) The discharge of both LPSI pumps passes through valve 2-SI-306 "SDC flow control valve * during shutdown cooling.
a.
What is-the potentially adverse consequence of this valve being closed during power operation? [0.53 b. What three PHYSICAL precautions are taken to ensure it REMAINS open while the plant is at power? [1.53 OUESTION 2.12 (3.50) Use Figure 2-1 (CVCS line diagram) as an answer page to indicate the following: a.
Which valve (s) receive a reposition signal from SIAS? Use 'S-S' for shuts or 'S-O' for opens. [0.93 b. Which valve (s) receive a reposition signal from CIAS? Use 'C-S' for shuts or 'C-O*'for opens.
[0.53 c._Which major systems connect to the CVCS at points A through G? [2.13 (xxxxx' END OF CATEGORY 02
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INSTRUMENTS AND CONTROLS PAGE.11 ____________________________ ? QUESTION 3.01 (3.00) , Answer the following questions regarding the Control Element Drive and Position Indication System:
.a.
What regulating control rod interlocks limits are in effect when in hANUAL INDIVIDUAL control? [1.53 b.
What system supplies rod position signals to the upper and lower CEA stops - (UCS, LCS)? CO.53 c. What are the instrumentation signals / conditions that could provide . a ' Dropped Rod' annunciator? (Two required) [1.03 i
QUESTION 3.02 .(1.75) 't With Channel X selected for pressuriner level control, level instrument LT 110X fails low.
What automatic actions and indication / alarms will result, including the effect on actual pressurizer. level? . QUESTION 3.03 (1.50) What three (3) conditions will actuate the individual channel (A,B,C, or D) power Trip Test Interlock? . -0UESTION 3.04 (1.50) What.three (3) conditions must exist in order for the turbine by pass i v,1ves to respond to signals generated by the Reactor Regulating System? (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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3.
INSTRuhENTS AND CONTROLS PAGE
____________________________ GUESTION 3.05 (2.75) For each of the three modes of diesel control as determined by the two j key operated switches, indicate l 1. Both switch positions l 2.
The key location (which switch, if any, the key is in) 3.
Which location (either ' local" at the diesel or ' remote' in the control room) has start /stop control of the diesel.
Do NOT consider the manual actuation of the overspeed trip or the local emergency control transfer button in your answer.
Format your answer as shown below: Mode 1 Mode 2 Mode 3 local switch position ______________:_________________:_______________ remote switch position ______________:_________________ _______________ key location
______________:_________________:_______________ start /stop control ______________;_________________;_______________ GUESTION 3.06 (3.00) For each of the following sets of conditions (a through d), which of the ESAS signals listed below, if any, should be actuated within a short time after the condition arises.
SIAS SRAS hSI CSAS EBFAS CPVIS CIAS AEAS or none i a.
A main steamline rupture (large break) occurs outside of containment from 100% power.
[1.23 //ps-/ r-co m of NS i t/, ' b.
The reactor has been shutdown and is undergoing a cooldown.
High activity levels exist in the RCS.
Pressurizer pressure is 1550 psia and the SIAS block has been activated.
A large break LOCA occurs at the pressurl:er surge line.
[1.23 c.
High radiation levels exist in the refuel pool area (approximately 125 mr/hr), and EBFAS is hANUALLY actuated.
Which signal (s) above will have FINAL priority? [0.33 d.
A containment gaseous activity monitor (channel D) fails high.
[0.33 (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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'3.
INSTRUMENTS AND CONTROLS PAGE
_________,__________________ QUESTION 3.07 (3.00) ' Answer the following questions regarding the Reactor Protection System a.
Which reactor trip inhibits are automatically removed above 10E-4% power? [1.53 6.
Which reactor trips are automatically bypassed below 15% power? [1.03 c.
Which reactor power signal is used to bypass the wrips below 15% power? CO.53 OUESTION 3.08 (1.50) How is the thermal power signal developed (what inputs are used and
what operations are performed on the inputs)?
i QUESTION-3.09 (2.00) Describe how and why the Feedwater Control System (FWCS) reacts to the following two occurances.
Assume a normal full power plant condition with the FWCS in MANUAL, and feed flow, steam flow, and level transmitter selector switches on BOTH.
a.
Turbine Trip C1.0] b.
The A S/G Alternate level transmitter fails high.
C1.0]
QUESTION 3.10 (3.00) , List each of the three positions of both the Mode Selection and Auto-Permissive switches associated with the control of each AFW pump and-flow ! control valve.
Briefly describe the function of each position.
Six positions total.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
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INSTRUMENTS AND CONTROLS PAGE
____________________________ l 00ESTION 3.11 (2.00) Which'of the following monitor channels have automatic-actions (other than indication and alarm) associated with them? Briefly describe the automatic actions.
, l a.
Reactor Building Closed Cooling Water Monitor b. Radwaste Vent Monitor - Gaseous I c.
Condensate recovery tank monitor d.
Clean radwaste monitor ! (***** END OF CATEGORY 03 *****) -.
> i 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ A5i5t55isAt E5i4TR5L -
R ____________________ QUESTION 4.01 (2.00) A caution in the Emergency Diesel. Operating procedure (OP 2346A) states "If a LNP is initiated, do not reset undervoltage relays at the ESAS Cabinet (s) until immediately prior to paralleling the Diesel Generator with the RSST during restoration per EOP 2528 (Electrical Emergency).' Briefly e:: plain the reason for this caution.
QUESTION 4.02 (1.50) Under what three conditions of RCP seal failure does OP2301C (RCP Operation) direct that the reactor be tripped, if critical? QUESTION 4.03 (3.00's What are five of the seven immediate actions required if a LOCA occurs during a normal cooldown after SIAS is blocked? Be specific.
QUESTION 4.04 (2.00) During a quench-tank filling operation: a.
Why should quench-tank pressure be maintained less than 50 psig? E1.03 b.
Overfilling the quench tank should be avoided.
Why? [1.03 00ESTION 4.05 (1.00) In the System Operating Procedure, ' Plant Heatup,' OP 2201, at the step where the Volume Control System is placed in service, there is a caution which states, 'Haintain RCS pressure below 265 psia by operation of the back pressure control volves.' What is the reason for this CAUTION statement? (xxxxx CATEGORY 04 CONTINUED ON NEXT PAGE *****)
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PROCEDURES - NORMAL, ABNORMAL,'ENERGENCY AND PAGE
~~~~ RAU5ULUU5 CAL CUUTRUL ~
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____________________ QUESTION 4.06 (3.50) For each of~the situations below indicate wh' ether the plant should be tripped immediately.
For situations which do not require an IMMEDIATE trip, specify at what point a reactor trip, if any, is required assuming the given condition continues to deteriorate.
For situations which do not require ANY reactor trip, specify what general > response would be required (shutdown, borate / dilute, raise / lower load).
Consider each situation separately.
a.
Two CEAs drop during a startup at 10E-3% power.
[0.53
6.
One CEA drops and pressurizer level decreases to and stablizes at 16%. Reactor power is 90%. [0.53
c.
A PDIL condition causes an alarm.
Reactor power is 50%. [0.53 d.
Iristrument Air press drops to 75 psig. Reactor power is 30%. E0.53 i e.
RCP seal temperature increases to 195F.
Reactor power is 90%. [0.53 f. Inadvertant ECCS initiation. Reactor power is 80%. [0.53 g.
Pressuricer level is 40% and slowly decreasing with maximum ' charging flow.
The cause cannot be found.
Letdown is isolated and
reactor power is 90%. [0.53 i QUESTION 4.07 (4.00) . For EACH of the below conditions list FOUR parameters used in the immediate action steps of EDP 2525 - Standard Post Trip Actions to determine if the conditions ~ exists.
a.
PORVs and Pressuriner safeties are NOT open.
[2.03 f b.
Noraal containment conditions.
[2.03 i QUESTION 4.08 (3.00) LIST the six (6) safety functions that are'in the safety function ], status check in the order of importance.
1 (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) ! , t ! . . .. . . . . . . . . . . . . . .
4.
PROCEDURES ' NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ RA5i5E55icAE 55sTR5L
____________________ QUESTION 4.09 (1.00) During a reactor shutdown from outside the control room, AOP 2551 Shutdown From Outside The Control Room directs that MSIVs are to be -losed if plant cooldown exceeds a specified limit.
If the control room is unavailable, how is this accomplished? Switch or component numbers not required.
QUESTION 4.10 (2.00) What casualty was considered'in EACH case below when determining the three minute 25 second time delay prior to AFW auto-initiation to ensure it was a.
sufficiently long, C1.03 , b.
but not too long.
E1.03 For EACH of these two casualties state WHY it was considered.
QUESTION 4.11 (2.00) A Health Physics survey in an area of the plant shows the following results* Beta gamma dose rate 4.0 mrem /hr Mixed (fast and thermal) neutron dose rate '6.0 mrem /hr a.
How should this area be classified radiologically? CO.53 b. How long can a radiation worker who has a complete NRC Form 4 (exposure record) work in this area without exceeding the Hillstone initial administrative exposure limits for beta samma and for neutrons.C1.03 c.
Who's approval is required to exceed the limits in (b) above? E0.53 (***** END OF CATECORY 04
- .*)
(************* END OF EXAMINATION
- )
- _ _ _ _ _ _ _ f a ma v o s/t Cycle efficiency o (Natwrx . , out)/(En:rgy in)
- s = V t + 1/2 ac o = mg o
E = mc KE = 1/2 mv a = (Vf - V )/t A = AN A=Ae**
~ g g PE = mgn Vf = V, + at w = e/t t= &n2/t1/2 = 0.693/tl/2 1/2*f# * E(*1n)(t )] t
- y ,yg [(t1/2) * (*b)3 - aE = 931 am I'= I e' * g Q = mCpat I = I e~" Q = UAat g Pwr = W ah I=I 10'*/UL g n TVL = 1.3/u sur(t) HVL = -0.693/u - P = P 10 p = p e /T t ,
SUR = 26.06/T SCR = S/(1 - X,ff) CR = S/(I - X,ff,) x CR (I - K,ffj) = CR II - keff2) - SUR = 25o/ t* + (a - o)T
Z T = ( t*/o ) + [(a - o )/lo] M = I/(I - X,ff) = CR /CR, j T = t/ ( o - a ) M = (1 - X,ffg)/(I - K,ff)) T = (s - o)/(la) SDM = (1 - K,ff)/K,ff p = (Keff-l) A,ff = aK,ffA,ff t* = 10 seconds T = 0.1 seconds ' - p = ((t*/(T K,ff)] + [i,ff (1 + AT)] / I dj = I d 2,2 2 j P = (c4V)/(3 x 1010) Id gd j
2 = oN R/hr = (0.5 CE)/d (meters) R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions I gal. = 8.345 lbs.
1 curie = 3.7 x 1010dps I ga:. = 3.78 liters 1 kg = 2.21 lem
1 ft* = 7.48 gal.
1 hp = 2.54 x 10 Stu/hr Oensity = 62.4 lbm/ft - 1 nw = 3.41 x 106 Stu/hr
' Oensity = 1 gm/cm3 lin = 2.54 cm , Heat of vaporization = 970 Stu/lbm 'F = 9/5'C + 32 Heat of fusion = 144 Stu/lbe 'C = 5/9 (*F-32) ' 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-lbf 1 ft. H O = 0.4335 lbf/in.2
- ._ _ . - _ _ - . - - __ __ _ _ _ -. .
l 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isEEs557sisics-sE5i iEAssFEE 395 FEUi5 FE5s ____________________________________________ ANSWERS -- HILLSTONE 2-86/07/07-COE,0.
ANSWER 1.01 (.50) B (0.5) REFERENCE Reactor Theory Lesson 11 (NET 3-10.2) Enab. Obj. 2 SF1 CRDS/000 K5.33 ANSWER 1.02 (.50) D Or-S REFERENCE Reactor Theory Lesson 12 Enab. Obj. 2,3 (NET 3-12.2,12.3) SF1 CRDS/010 K5.16 f%sw+lcN hpov-o N V4h%ek w & & Fen.cf{,},If7 p27 ANSWER 1.03 (2.50) a.
B so.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 (more neutron generations 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 care, not on neutron level.
(0.5') REFERENCE Reactor Theory Lesson ? Enab. Obj. 2 (NET 3-12.1) SF1 CRDS/000 A1.06
. . . _ _. .
i 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
~~~~iU5R5567UdE5555~55dT TRdU5FER hs5 FLUi5 FL5s ~ - ____________________________________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,D.
ANSWER 1.04 (2.00) , { ~ a.
EOL - MTC is more negative-and thus imparts greater positive reactivity ' from the drop in coolant temperature.
[0.53 FTC becomes more negative (due to Pu-240 buildup) and would insert more negative reactivity from the power increase, but is less , ! in magnitude.
E0.53 b. BOL - MTC is less negative (or positive) and thus imparts less negative . ! reactivity from the coolant heat-up which tends to hold power up.
CO.53 i FTC is less negative and thus imparts less positive reactivity l as power drops, but is less in magnitude.
[0.53 REFERENCE . Reactor Theory Lesson 10 Enab. 0bj. 3,4 ' Lesson 8 Enab. Obj. 5,6 (NET 3-9.1,9.2,9.3) SF1 CRDS/000 K5.26, K5.48
ANSWER 1.05 (3.25) sur(t.
P = Po(10) [0.63 j-4 -5 1.5 sur
10 (10) CO.63 = 0.666 OPh C0.33 solve for sure sur = T = 26.06/sur 26.06/0.666 = 39.13 see CO.63 = rho = beta eff/(1 + lamba*bar X T) = 0.005/(1 + 0.08(39.13)) CO.83 12.1 X 10E-4 delta k/k CO.353 = 0.001210 = REFERENCE
Reactor Theory Lesson 7 Enab. Obj. Ic, 6 Lesson 8 Enab. Obj. 6 SF1 CRDS/000 K5.47 (NET 3-6.4 to 3-6.6) l l h ! J l l J
a
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1.
. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
~~~ isEEs557sisiEE-sEAi iEEssFEs Es5 FE0i5 FE5p ____________________________________________ ANSWERS -- HILLSTONE 2-86/07/07-COE,D.
l ANSWER 1.06 (3.00) a. LOWER b. HIGHER l-c.
HIGHER d.
SAME e.
LOWER [0.6 each] (3.0) REFERENCE Reactor Theory Les in 12 Enab. Obj. 106, OP-2208 SF1 CRDS/OOO A2.07: A4.03 ANSWER 1.07 (2.50) a.
Steam pressure decreases [0.43 SC pressure decreases as demand is increased due to the removal of energy. CO.43 b.
Tave will decrease. CO.43 due to energy removal via.S/G E0.43 c.
Power will increase CO.43 due to the positive reactivity added by HTC.
CO.53 REFERENCE Reactor Theory Lesson-10 Enab Obj ic, 2b-Lesson 2121J Enab Obj 10 SF5 HRSS/000 K5.08, A2.05 SF1 CRDS/000 K5.29 ANSWER 1.08 (1.00) a.
228 F [0.53 b. greater than C0.53 REFERENCE Lesson 2121G Enab Obj 5,
pp 16,17 Steam Tables
L.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
~~- isEEs557sERICE-RE3i iEAssFEE Es5 FEUi5 FE50 ____________________________________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
SF3 EPE/008 EK3.02 ANSWER 1.09 (2.50) a.
Decrease to 1900 - 2000 psia b.
1) Increase, due to HPSI and charging flow.
CO.53 _
- ;,.
,q .,. c ;; z _ Refhy belli[0.53 2 ) _r., - -+- . n9 3) No. [0.53 Flow out the safety is not sufficient to r.3 m o v e the initial decay heat produced by the core.
With no other heat removal, core damage will result.
[0.5] REFERENCE a.
Lesson 2121J Enab Obj 11e SF3 EPE/009 PWG 110 b.
Mitigating Core Damage Book 1 pp 38-50 and Steam Tables SF3 EPE/009 EA2.04, EK1.01, EK3.22, EA2.39 ANSWER 1.10 (2.25) 1) Reactor power E0.253. Increasing reactor power results in increased heat flun and operating closer to DNB.
[0.53 2) Tcold E0.25]. If pressure is held constant and Teold is decreased, subcooling will increase.. Therefore the heat flun required to' reach DNB will increase and the margin to DNB will increase. [0.5] 3) Pressurl:er pressure EO.253.
If Teold is held constant and pressure increased, subcooling increases and margin to ONB increases. [0.5] REFERENCE Lesson 2121I Enab Obj 4 SF9 RPS/000 MS.01 MP2 question 708 I
~1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
--- isEss557sAsiEs-REAi isissFEE As5 FEUi5 FE5E
......--
ANSWERS -- HILLSTONE 2-86/07/07-COE,0.
ANSWER 1.11 (2.00) a.
Decreases, smaller delta T required to transfer energy to.RCS.
[0.53 b.
Increase, center line temperature responds to RCS temperature in order to maintain constant delta T across cladding.
[0.53 c. Decrease, fuel swelling and clad creep reduce clad gap which reduces delta T_across gap and lowers center line temperature.
[0.53 d.
No change, pressure has little effect on heat transfer in subcooled fluids.
[0.53 REFERENCE Lesson 2121I Enab Obj 6 CAF part c MP2 question 785 SF4 RCS/000 K5.01 ANSWER 1.12 (3.00) a.
Teoid util remain constant [0.25] Since it follows S/G saturation temperature.
[0.53 b.
That will decreaseEO.253 since less fission product heat is being produced than is bcing removed by the steam generators.
[0.53 c.
Core delta T will decreaseE0.253 since the amount of decay heat is decreasing.
[0.53 d. Loop transit time will increase E0.253 since the driving head for flow (core delta'T) is decreasing.
[0.53 REFERENCE Lesson 2121J Enab Obj 4,5 MP2 question 603 SF4 EPE/015 EK1.01 SF4 RCP/000 K5.04 l ,
\\
.. -..- -. ._ -_. . -. - - _.. t e r 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________ ANSWERS'-- MILLSTONE 2-84/07/07-COE,0.
ANSWER 2.01 (1.00) . Temperature is monitored in the line that taps off between the two ! concentric vessel 'O' rings.
If the inner 'O' ring leaks, the temperature will increase and an annunciator in the control room will be actuated.
REFERENCE Reactor Vessel construction 5.0. Para. IIB.
HP2 question 245 i SF2 RCS/000 K4.05 ANSWER 2.02 (1.00) Siphon breaker in suction line of recirculation pump is located at 91.4% level.CO.53 Recirculation Pump discharges to top of tank. CO.53 ' REFERENCE
Dws No. 25203-26015 and HPSI S.D.
pg 4 (?) CAF i HP2 question 382 l SF2 ECCS/000 K6.09
SF2 ECCS/020 A1.09 f ANSWER-2.03 (3.00) a.
MITIGATE CO.53 b.
To ensure adequate shutdown margin CO.53 so that a reactor restart
due to cooldown is inhibited CO.53 c.
Core -> Coolant recircing to sump via HPSI/LPSI -> containment spray ~ 'l system -> RBCCW system (SDC HX or CAR fans) -> Service water system -> 1, Long Island Sound (environment) CO.2 each] d. 1) H2 recombiners '2) Purge via Enclosure Blds Filtration SystemCO.25 ea] REFERENCE HPSI S.D.
pg. 1 Ctmt Sprav S.D.
pp. 11-13 and Fig. 7 Emerg Vent S.D.
SF6 CSS /000 PWG 4 SF6 CCS/000 PWG 4 SF2 ECCS/000 PWG 4 and ECCS/030 K4.03 j l SF2 HRPS/000 PWG 4 I
,
! I
$ i . I
-, - ~ _ _. -, - - _ _.., _ _. - - - - - _ _ _ _ .... ,- _.. -. _ _. _ _. _.,. _ - _,.,. _ _ -, _ _. -.. -, _ _
,__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
ANSWER 2.04 (3.50) , { a.
8-12' hours atter initiation of SIAS due to LOCA [0.2] ' 1) Crystallization of boron in fuel region could inhibit heat transfer 2) and coolant flow through core [0.4 each] b.
Normal path [0.8, 0.2 each for 4 valves], [0.2] for pump, [0.23 entot.
Alt. path [0.8, 0.2 each for 4 valves], [0.21 for pump and [0.1] for Regen Heat e>: changer, and [0.2] for entmt location.
l REFERENCE l HPSI S.D.
pg. 13 E0P 2532 Step 3.33 SF3 EPE 011 EK3.13, EK3.08, EA1.11 l l ANSWER 2.05 (1.50) a.
Idle CAR fan starts on low speed Operating CAR fans switch from hi to low speed 10 inch RBCCW valves on cooler outlets open [0.2 each] b.
133% [0.43 c.
Nuclear Instruments [0.5] REFERENCE Cont Ventilation S.O.
pg 5-7 SF6 CCS/000 A3.01, A1.02, K1.02 ANSWER 2.06 (1.00) a.
three [0.5] YEC b EC b.
12 seconds [0.5] c, r 6 20 Sec P f (plus or minus 3 seconds for full credit) REFERENCE Y, 8. /, /, 2. G.1 EDG handout (system description) pp 1-8 SF7 EDG/000 A3.02, A3.04 .
) 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
_______________________________________________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
ANSWER 2.07 (2.50) a.
Swing Service water pump, swing HPSI pump, swing RBCCW pump C0.2 each] b.
Prevents tieing both 4160 emergency busses together through Bus 24E.
p'e eu+> 4xceebeug =inA (wifs en Rs5T hreakw /TC-~2/S J oVc Nssig. . f ar E": 11E te
- 2---
c' : ECC" **- =',r##4-i ~ * c.r-- e- - c c, c o m. -"c .. . y-4 9. t_, ti.o3 p m 1, n m 4, REFERENCE a. Elect Dist (In-house) S.D.
App A pg A-1 b.
FSAR Section 8.2.2.2 O P R 3'f 3 REV 7 CAdo " 7U c.
Elect Dist (In-house) 5.D. pg 3 j SF7 AC/000 K3.01, A2.06, A2.09 ANSWER 2.08 (3.00) a.
MFWRV fail as is p:r spray valses no effect (entmt air reve charged) S/G feed pump control fail as is Letdown system flow stopped ADVs fail closed AFW flow control valves fail oPen EDG service water supply fail open CO.25 each] HSIVs no effect (independent air reciever) b.
1) Auto valve to station air compressor X-ties SA to IA at 85 psis IA pressure CO.53 2) manual X-tie valve to Unit 1 station air via Unit 2 station air CO.53 ) REFERENCE
Instru'nent Air S.D.
AFW S.D.
pg 3 ESAS handout Appendix A pg 10 (EDG SW valves) SF8 IAS/000 K3.02, K4.02 , . . .m.
. .. m.
. . m.
.- . . .. . . m . . . . . -
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE-26 _______________________________________________________ , ANSWERS -- HILLSTONE 2-86/07/07-COErD.
ANSWER 2.09 (2.00) HPSI/LPSI/(CSP) seal coolers CO.23 None, coolers extend seal life and are not required for ESF operation CO.4] ESF room air coolers CO.2] ESF pump motor overheating [0.53 CAR cooling unit CO.23 Rate of energy removal from en'tmt will be slower via only entmt spray [0.53 og ,co,$ect h EF hbj CSPi REFERENCE HPSI S.D. pp 8-9 ldg j{3 fT*
- j#
h LPSI S.D.
pg 7 M CC "~P - CG S.D.
pg 5 RBCCW S.D.
pg 4 Emerg Vent S.D.
pg 1 . Dws 25203-26022 (RBCCW) l SF 10 CCHS/OOO K3.01 SF 10 CCWS/030 PWG 4 ANSWER 2.10 (1.00) I
Inlet and outlet valves are interlocked CO.53 so that the discharge valve I may be opened only if the corresponding inlet valve is shut.CO.5] REFERENCE Radwaste S.D.
pg 6 l SF 11 LRS/000 M4.01
l ANSWER 2.11 (2.00) a.
The inability of the LPSI pumps to inject water into the core following a LOCA.
[0.53 b.i' '.' * -rim-m - - ' r- .. _1-F . ^r+ri",'I' "i ni ;- ::.n , 2) Manual operator on the opposite side of the valve shaft is pinned and locked to the handwheel.
[0.53 3) Handwheel is locked in position. [0.53 g} gI,c /c fsNc( fo "sJ"ges/Nem Ce,d / REFERENCE LPSI S.D.
pg 9 SF2 ECCS/000 K4.10 g Re%4 /use Ock b 2-6J-3d (d. I3 { SF3 EPE 011 PWG 7 l oe.u,o % e, s) r< IA * *yh "J Ses&nn-7 4,8 [n y 3 9 o.s~ u d ]
~ - _ _ _ _ _. _ - _ - ._ 2.
PLANT DESIGt1 INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
---------------------------------
ArlSWERS -- MILLSTONE 2-56/07/07-COE,0.
ANSWER 2.12 (3.50) a.
9 valves as shown [0.1 each] b.
5 valves as shown [0.1 each] c.
A RCS/P:t F RWST l C SDC HX outlet G PMW l ' C LPSI suction, or SDC D RCP seal leakoff E HPSI [0.3 each] REFERENCE CVCS trainin3 du3 SF1 CVCS/000 K1.04, K1.15, K1.19, K1.23 SF1 CVCS/010 K1.01, A2.05 -
I - o g A }) ' ' ' ,y "' ' (R-C) U) , (- U1 r .
I (R-C) .JLJ
4 j M ,r _ - h ' P A Pp
g . a s k h $ ~ n - M[A 99_ -T*"k d,# u c
, e& , - , ,,,, 0 4;IRW(SFPC) 4L4k f
- y h
9 CH ,, [A_' s (S) (R T) p g ' H Oi Note r , - - - LPSI Q , ,, y - - su w _ ' SIT , ,y (S) (R T)' F '
- 1
J '- ' Not'e 01 ( ) & X~ @ V (SFPC) (Warmup Line) B Pp N NP ( wr u i f - (S) (S) ' Q Or I T CH (Pp Disch) g (S C) l, , f - e _ - h = 8 i q A Pp M Xo@ C) ' E i ~ (S , ! %MM ?>om g i @ L y s) g:' b ' ""S' J.g$) . ,
- 41
" " ", t ' , % ;p M A (m 4.l h L CI,[o
, -- [ i h p 4 1.
'A-g ; C ~t m w - h$ F ' F ' e e Q i Containtnent ' i ". !' ' Sumo f'l h - RWST m * TSP I h
). . ow M 8ec.tas: i...,.c.,on.....e. < _ > < > sir $ s.
hj h ' I m J 2. r , a r m a - , 0 (CH) Pp Suct ._ - - - - - - - - - - - - - - - - - - - - - - _ _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _, _ _. __.
_ _ _ _ _
O I I (R C)
(- ,r,, z (R-C)
jga g ru, @Q M c
mn
_ V g 3 - - g 5 z I A Pp
- 4
' P j g dE dL
CS a: d 5<
E8 _ g . ,r z .),) C X-f' ) P) w j A . L,.o'"V = g
,,,, e_ Q,",c*$',"" f d 3 9 <s4 (am !
._
- ~)......
$, L
- =
-- - v , a
[ ' A Pp ' r d.
________ ' LPSI 'r-um _ pum i " g (,)(S)(R-T)' y ' t sit ,,f,,
p r - et t J ,, s Note s1 [T gw ag ' g V (SFPC) -, (Warmup Line) B Pp [ am- ,r v [ .
(S) (S) (S C) l, Ig I b S) , , l ' ' ^ ' ' ,,A PO e t
u - .. r .. g V } 4-- HPSI ] h j ,
@{ phO ) B Pp g , r (.ed+Pgf% h: " _ 9 ,,a- -- e , :0 s -
' (j ' g;,e -- e w 'h I
7 l & - u < c o i
' r , r
g g Containment ,, g Sump ig g, i ${ RWST r h M (4}+ w sootas: j, y g ' - > '
. _. t gyy,y mt** to'oop Ob , _ 'g g " w gr } i, 2. rrom siis 2.3. a 4 a r r a ' , (R)
- (CH) Pp Suct
, . -. .. .. . - - _. - -.,. - -.. - _ _. - - - -. - _ - - -. - - - - - - -....... .
) 1/ - _ V y ' _ E h ._ N i L x Tu
A,T c W r , a S M A iqsh! A s P B , A s [ _ / c l s E _ t A i i ( s -
o, g !: ! - _ i _ - a ) A' , H a - . C _ F I - _ X 9gi _ _ E
- _ , - . N , . - _ O , I )4 - ., N )S W c.
R M F . O A W I ( _ I.
T G P i A Eg F C D < hu, _ I - e _ F
I R - R R F y - U L - h f~ P -
( %,l N S , , ( ' > - - - O -
'+ _ A n V - a } - y ' , ' l _ ,a flp _ g _ - > p[a' '6 T-O _ Q,',,,
M CV
@ ^
'_ ,<,
' - _ . ~ _.
_ - R " v2 )S - _ < > p = A - p , _ _ G - P e _ = E - 'a g ___ D s _ Q.
NpO _ = ( - _ t R = R - - L _ L= - _ - _ - _ ' b @" pjb h ( )L 'i $, C ) EV n ( E W L ( a C h 'm C n B f q R s r "g gy , , 4l !J s ' e '. _ e - _ n k3 "n _ . - - _ s - - y'l s S I O S - E _ , , i . Wi c
i_ _ s b _
'
_ ]a ]5 G.
_ r bh _
[o [o ) _ . a. g
J ) _ _ A _ _ g,- _ e _ c _ t i t 3i _ c
o b f:;!< ub t < f - _- g _- , - p n-p1 o ~ P RJ b i. o e
o s @e % &%M it - i - j - - s 6uia - e p s v e l v a g T, ]4 > o l n a
1 v C Az D i
i, e
' t ~ e e P _ - ic m C R " i 0 b c n u j j
ce ,l I
, _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _. - - - - _ _ _ -. _ _ _ __ __ 3.
INSTRUMENTS AND CONTROLS PAGE
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
j ANSWER 3.01 (3.00) a.
1.
Upper and lower electrical limits
, 2.
CEA withdrawal prohibit ! 3. CEA motion prohibit CO.5 each3 ] b. Computer position indication E0.53 c.
1.
Rod drop from reed switch ] 2.
NI negative rat,e of power change from NI system E0.5 each] REFERENCE ) CEDS S.D.
Logic Diagram i.
MP2 question 704 j SF1 CRDS/000 K1.05, K4.07 ! ANSWER 3.02 (1.75) i 1.
letdown to minimum j 2. both backup charging pumps start j 3. Channel X PZR level Hi/Lo annunciator -
4.
Actual PZR-level increases {%7 S-d o,3c,u[] j 5. PZR level Lo/Lo annunciator ! 6. All heaters de-energine 7.
selected controller (X) output signal to minimum -- N ? c c 51 ! l REFERENCE
Pressurizer level / pressure control S.D.
Section IV pp 22-24 and Fig. 3 j SF2.EPE 028 EA2.01, EA1.02 ! l HP2 question 526 l ! - I ' j ANSWER 3.03 (1.50) I i l 1. Linear power channel summer control switch out of the (A + B)/2 position.
[0.53
2. Linear power channel high voltage bistable tripped CO.53 i 3.
Reactor protection system calibrate and indication panel !
j Delta T power calculator test switch out-of the operate position. E0.53 l ! h or-R Psc iP calibrele_ swNck i
REFERENCE ! NI S.
D.
Linear Channel para L.3 I Y.Zero-cf8W C.-caffbr*fe-swf/chr(2)
SF9 RPS/000 K6.04 on M L w Po m ch.<nele nf oh l 895 L<stm P)n 2380 -1, P.7 l
l OfMC [o, 5] f C, fr.y +efi-S d o b $ L S ( 2- ) W NZ l t,n pm Ch fr A of 00#3] ! i , ! l [W 3 @ 0.Twcl] ! !
, -____ 3.
INSTRUMENTS AND CONTROLS PAGE
____________________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
MP2 question 596 ANSWER 3.04 (1.50) 1.
The turbine has tripped. [0.53 /S" 2.
Condenser Vacuum is greater than-Z: inches
- -:
C.
t 15 : :- ' - ' ' saf.
_ [0.53 'Y'(juick of tw onl ) oR 54 wh[e 3.
Reactor Reg. channel selected is y " CO.53 //' ' REFERENCE Reactor Reg S.D.
pp 11-14 and Fig. 7 MP2 question 716 SF5 SDS/020 K4.01 ANSWER 3.05 (2.75) l Mode 1 Mode 2 Mode 3 j local switch position Maintenance Normal Normal CO.2] E0.23 CO.2] remote switch position Local Local Remote ! [0.23 CO.23 [0.23 key location Local Either or none Remote C0.2] CO.23 E0.23 Start /stop control only at start /stop localEO.23 only at local emers stop remote remote [0.25] EO.25] E0.25] REFERENCE l EDG handout pg 7 SF7 EDG/000 A4.01 ! ! ,
i i
l
,. _ _ _ _ _ _ l l k~ 3.
INSTRUMENTS AND CONTROLS PAGE
l ____________________________ { ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
ANSWER 3.06 (3.00)
a.
MSI CIAS SIAS EDFAS CO.3 each] b.
CIAS EBFAS CSAS CPVIS i:.; c ci: 5745 [o. 24- ] c.
EBFAS CO.3] d.
CPVIS CO.3] ! REFERENCE
ESAS 5.D.
ESAS Logic diagram Emerg Venttistion S.D.
EBFAS/AEAS Logie diagram pg 10 ' i SF2 ECCS/000 K1.02 SF2 ESFAS/000 K1.01
ANSWER 3.07 (3.00) . a.
RC flow l RCP speed ) Thermal Margin / Low pressure CO.5 each] , I b.
Loss of turbine i Linear Power Density [0.5 each] l l c.
Linear safety channel E0.5J j REFERENCE RPS S.D.
pg 3,
i NI S.D.
Power Range Linear Channel PLra J.
SF9 RPS/000 K4.06
i ANSWER 3.00 .(1.50) ) { Average of Thot's} minus {high auctioneered Teolds} C0.5] CO.5] CO.5] REFERENCE I NI S.D.
Power Range Linear Channel para G.
SF9 RPS/000 K4.03 'l ! i
l i l ! \\
- _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ i 3o INSTRUMENTS AND CO'NTROLS PAGE
I l
ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
! l i l ! ! 1' ANSWER 3.09 (2.00) a.
MFWRV ramps shut for 30 see (should fully close) and MFW Bypass valve f ramps to 75% open (5% flow) [0.53 to help match feed flow with the ! reduced steam flow E0.52 , b.
No effect CO.53 since the BOTH position low selects,the level signal to be used.
[0.53 REFERENCE FWCS S.D.
pg 3 and 9 j l.
SF5 HFW/000 A2.ll, K4.02, K4.18 l l l ANSWER 3.10 (3.00) I ! Auto-Permissive 1. Pull-to-lock CO.253 Blocks auto start signal E0.25] ,
2. Reset E0.25] Resets Automatic start function for the pump EO.25] 3. Start Co.25] Starts the AFW pump E0.153 and opens JAwr flow control j valvey E0.13 behk j
Mode Selection Switch (for each flow control valve) ! 1.
Normal E0.253 allows valve to open in auto E0.253 l
2. Override [0.25] allows manual control following auto AFW initiation ! ' l E0.253 3.
Reset CO.25] resets logic to return to " normal' position.
CO.253 ' REFERENCE AFW S.D.
pp 5-6 , I SFS AFW/000 K4.02, K4.06 ANSWER 3.11 (2.00) a.
none E0.53-b.
none E0.53 j c. -s+ to.53 Mr alan si,;5+s cRT dw%e & Aus% Fulkge Td 1c> L ! d. High alarm closes two discharge isolation valves to stop discharge flow EO.53 REFERENCE
Radiation Monitoring System Lesson Plan Table 7.5-5 i SF9 ARH/000 K1.01 Agj-d dye.
Sfe<g Mikk & j SF9 PRM/000 Kl.01 OP a n M Su+ 9. C p3 3 gg,Q,
m a as,03,o2c A,Swc,A,k
! l l
..,.. - - - ~ _. -. _ - - -... - -,, _.. - - - _.- --..... ...-.- ... . -. - - - - _ _ _ _ _ _. _ _ _
_ _ - _ _ _ - _. _ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ 565dLd55bdL EUNTRUL ' ~~~~~~~~~~~~~~~~~~~~~~~~ R ____________________ ANSWERS -- HILLSTONE 2-86/07/07-COE,0.
ANSWER 4.01 (2.00) If the U/V relays are reset when the EDGs are operating in a LNP condition and'a SIAS subsequently occurs [1.03, all ESF loads required by SIAS initiation will be energized simultaneously rather than sequentially and overload the respective EDG.
E1.03 REFERENCE OP2346A ps 6 SF7 EDG/000 K4.11 ANSWER 4.02 (1.50) 1) One seal is failed CO.53 AND 2) delta P across either of the other two seals is less than or equal to 500 psi CO.5] AND 3) any doubt exists as to the failure of the third seal.
REFE i E lyNy --h * 00230 0' I' ' O b' d OP23010 precaution 6.1.9
,;g} g j 7 y j 7o F - (pt<- of 23OlC 8,/T) f SF4 RCPS/000 A2.02, PWG 10 c 3} gge,44 gW c /,<)< Vr:fv4_ c[cSc5-- ANSWER 4.03 (3.00) Manually initiate HPSI CO.63 Q 4-c P2 3o/ C 6, '/- ) g,y} ! Open any closed SIT isolation valves CO.6] j Close any open leak check drain valves [0.6]
Ensure a LPSI. suction flowpath is aligned from the RWST CO.63
Check open 2-SI-306 CO.63 , Manually initiate LPSI E0.63 { Close SIT test header stop valve (2-FT,o3) [0.6] Eany 53 i I REFERENCE OP2207 pg 9 step 5.7.1 Y gg gcYy by d
SF3 EPE/011 EA1.04 , h Ac.M; 52M
9 ekks ana/w alqns k/ & c a d . l w b.
.
! l ! _ _. - -. _ _ _. _. _ -., _ _.. _ _ _ _ _ _ _ _ _ -. _ _ _ _
. ._- ___________ ___-___ _ _ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
-
~~-----------
R 6iUL55iEAL E5 sir 5L ~~~~ ____________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
AMSWER 4.04 (2.00) a.
In order to avoid possible distortion (or rupture) of the rupture disk on the quench tank.
[1.03 6.
Overfilling should-be avoided in order to prevent water from entering int o the waste gas header.
E1.03 REFERENCE OP2301A Sec 6.1 and 6.3 SF3 PZR PCS/000 PWG 7 ANSWER 4.05 (1.00) j To prevent the shutdown cooling system from isolating CO.53 and injection tank outlet valves from auto opening. [0.53 . REFERENCE OP 2201 Sect 5.1 CAUTION pg 9 SF2 CVCS/020 K6.03, K6.02 , SF2 ECCS/020 A3.02 SF4 RHRS/000 K4.07 , ' ! l i ,
I l~ L:.... .__
. _ _ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ EDist55i5AL 55sTR5t-~~~~~~~~~-------~~~~~-- - R ____________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
ANSWER 4.06 (3.50) a.
No trip C0.25] commence normal shutdown E0.25] b.
trip E0.53 c.
no trip CO.25] emergency borate E0.25] d.
Trip [0.5J -
- - -- --,.-
_ _ _; +. -- , ; _, c_ ,enr r e.., m p E' g ._ _s _ f.
no trip E0.25] monitor plant parameters and restore systems and reduce load if necessary CO.25] 3+ tri P CO.53 REFERENCE a.
AOP 2556 pg 2 SF1 EPE/003 EA1.06 and PWG 10 b.
AOP 2556 pg 2 i c.
AOP 2558 pg 2 SF1 EPE/024 EK3.01 d.
AOP 2563 pg 2 SFB EPE/065 EA 2.06 ej : SF10 EPE/026 EK3.03 ,CI ;;u , f.
ADP 2571 pg 2 SF2 ECCS/050 A2.01
AOP 2568 pg 3 SF2 RCS/020 PWG 10 O P3301 C hed 7 Se b h 2 lS~ ANSWER 4.07 (4.00) ! i a.
1.
OT level ! 2.
OT pressure l 3.
OT temperature
4.
Acoustic monitor ! b.
1.
containment pressure 2.
containment temperature 3.
containment radiation l 4.
containment sump level CO.5 each] REFERENCE E0P 2525 steps 3.3d, 3.5 SF1 EPE/007 PWC 11 ! l
! . l I j
_ - _ _ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
j - -~~~~~~~~~~~-~~~-~~~~~~~ ~~~~ d65UL5GiEdL E5sTR5L R ____________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,0.
, , i l ANSWER 4.08 (3.00)
1.
Reactivity control ,
2.
RCS inventory control ! 3.
RCS pressure control l 4.
RCS heat removal 5.
Containment integrity 6.
Vital avv.iliary( be g/ ace [,K W N 'I4y' Y"ca
,
C+0.53 each REFERENCE , j Millstone 2: EMERGENCY OPERATING PROCEDURE, E0P 2526, OPS form
2526-1, pp.
3-5.
I SF1 EPE/007 EK3.01, PWG 10 ll I ANSWER 4.09 (1.00) I By opening the 125vde supply breaker to the solenoid operated air ' valves.-21?_.;,g -: : T.
to f__' ch+ ' ' - - __ REFERENCE AOP 2551 pg 5 step 4.12 SF8 EPE/068 PWG 6 ANSWER 4.10 (2.00) a.
Sufficiently long so that even if an affected S/G is fed during , a DBA MSLB E0.53 an unacceptable return to power does not occur. [0.53- ' ' '. Not too long so that following a complete loss of main feed [0.53, o S/G's are not allowed to go dry EO.253 AND PORV's are not challenged CO.253.
REFERENCE AFW S.D.
pg 10 and Appendi:: A AOP 255/ lpg 3 NOTE SF5 EPE/054 EK3.04, EK3.05 -- - - _ _.. _ _ _ - - _ _. _ _ _.... _ _ _ _ __ _ _ _ _ _ _ _.. _j
. - - _ _ _- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _. _.
_ _ - _ _ _ _ _ _ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~~ d65ULUU55AL 5UUTRUt ~
~~----~~------
R ____________________ ANSWERS -- MILLSTONE 2-86/07/07-COE,D.
SF5 EPE/040 EK1.05, EK1.06 ANSWER 4.11 (2.00) a.
radiation area (> 0.5 mrem /hr) [0.25] neutron radiation area (> 2.5 mrem /hr) CO.253 6.
1000 mrem /(4 mrem /hr) = 250 hours for beta samma.
250 mrem /(6 mrem /hr) = 41.6 hours for neutrons (thus n's are limiting) E1.03 c.
HP supervisor / designee CO.5] l REFERENCE ! a.
SHP 4906 pp. 2,3 I b. and c.
SHP 4902 pp. 9-11
PWG 115 ! i ! t i f , t ! ! .. . .. .. . .. .....
" " " " MASTER - . . _.. . , U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION - MILLSTONE 2 Facility: PWR-CE Reactor Type: Date Administered: 86/07/07 Examiner: SMITH, J.
ANSWER KEY Candidate: INSTRUCTIONS TO CANDIDATE: Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.. Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers wi.ll be picked up six (6) hours after 'the examination starts.
Category % of Candidate's % of Value Total Score Cat. Value Category
25Ao 5. Theory of Nuclear Power Plant Operation, Fluids and Thermodynamics
25Ao 6. Plant System Design, Control and Instrumentation
- l3.S 23 90
,26'
7. Procedures - Normal, Abnormal, Emergency, and Radiological Control
25.4 8. Administrative Procedures, l Conditions, and Limitations , M'i8 f TOTALS Final Grade % , ! All work done on this examination is my own; I have neither given nor received aid.
i Canaidate's Signature l , f .
.
, - - - - - - - - - - l . .
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: , t 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
. 2.
Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.- Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
Consecutively number each answer sheet, write "End of Category _de of " as 8.
appropriate, start each category on a new page, write only one si the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example,1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in parenthesis after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer l to mathematical problems whether indicated in the question or not.
15.
Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
l 16.
If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
( . _ __ . . 18. When you complete your examination, you shall: , a.
Assemble your examination as follows: (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
, (3) Answer pages including figures which are a part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
! c.
Turn in all scrap paper and the balance of paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
I ! . .-.-.m - --wam- .---,__--__-.-,._.4 %_ ,__ m _,r,-,,.m,,_,,,_.., _- _.__--_.-_v,e., --,-,,, - --__ -. -,,,,.,,,,.. - - - -. --
, _____ . 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
. THERMODYNAMICS- , QUESTION 5.01 (.50) ' WHICH one of the statements below most correctly describes the transfomation of the energy from fission events into heat energy? (0.5)' (a.) Th? energy released as kinetic energy of fission fragments provides less than 50% of the energy released per fission event.
(b.) Fissioning of the isotope U-238 provides more than 50% of the themal energy generated in the core.
(c.) About 200 Mev of energy is released per fission event (neglecting neutrinos) of which about 15 Mev is released after a delay time.
(d.) All of the fission-event kinetic energy is absorbed in the coolant.
QUESTION 5.02 (.50) WHICH three (3) of the statements below are correct for a nuclear reactor of the Millstone Unit 2 type? (0.5) I (a.) The product of.the macroscopic cross section (SIGMA) and the neutron flux (phi) gives the neutron. reaction rate i (interactions per em**3 per sec).
(b.) If the themal neutron flux is doubled in one hour, the thermal power produced in the nuclear reactor is doubled.
(c.) The neutron microscopic cross section (sigma) for a certain element varies with neutron energy and is dependent on the isotope of the element.
(d.) The thermal-neutron microscopic fission cross section for U-238 is larger than that for U-235.
. (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) .-.- - - -
- - - _ _ - _ - _ _ - _ _ _ - _______ ___ _ '
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS } - l QUESTION 5.03 (3.50) ' A reactor startup is in progress.
You are given the following ,, information.
Shutdown groups A and B fully withdrawn
RCS is at No Load Temperature and Pressure
K(eff) = 0.95
Countrate = 50 CPS and stable.
- a.
HOW MUCH negative reactivity is in the core? (SHOW all calculations.)
(1.0) The control room operator begins regulating rod withdrawal.
Later, the operator notices that the countrate has increased to 100 CPS.
b.
HOW MUCH negative reactivity is present in the core? (SHOW all calculations.)
(1.5) WHAT change, if any, has been made in the value of the SHUTDOWN c.
MARGIN, as defined in the Technical Specifications, as a result of the rod withdrawal? EXPLAIN.
(1.0) . QUESTION 5.04 (3.00) The reactor has a startup rate (SUR) of 0.4 decades per minute (DPM) at BOL.
HOW long will it take after passing 25 watts to reach a.
4 megawatts? (1.0) L b.
If the same amount of reactivity as was added in part (a.) is added to the reactor at EOL, WHAT would be the resultant SUR7 - (STATE assumptions and SHOW calculations.)
(2.0) i
QUESTION 5.05 (1.50) - ! I The discharge valve of a centrifugal pump is opened for additional I flow.
INDICATE the change in the following parameters.
(Answer INCREASES, DECREASES or REMAINS THE SAME.)
(1.5) l a.
Discharge pressure b.
Pump power i c.
NPSH (***** CATEGORY 05CONTINUEDONNEXTPAGE*****) _- -_- _.-.-_- - _- -.. - -. - _ _ _ _. . _ - -. _ _ -. -
] . 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
- THERMODYNAMICS QUESTION 5.06 (1.50) - HOW would the following errors in the ECP calculation, used to predict ". criticality during a startup, affect the critical rod position? Answer with either HIGHER THAN EXPECTED or LOWER THAN EXPECTED.
a.
Actual RCS boron concentration was higher than assumed.
(0.5) b.
Actual core burnup was less than assumed.
(0.5) c.
Actual RCS T(avg) during the critical approach was 10 degrees F less than assumed (ASSUME EOC).
(0.5) QUESTION 5.07 (1.00) Cooling water is being supplied to a heat exchanger at the rate of 1000 lbm/hr.
It undergoes a temperature increase of 35 degrees F in the heat exchanger. The water that is being cooled is entering the heat exchanger at a temperature of 120 degrees F at 600 lbm/hr.
WHAT is the outlet temperature of the water that is being cooled? (1.0) QUESTION 5.08 (2.00) EXPLAIN the effect of delayed neutrons on reactor control during a power increase.
(2.0) I QUESTION 5.09 (3.00) During end of cycle operation with the reactor at 100% power, all rods out, boron concentration at 10 ppm in the RCS, and T(avg) 10 deg F less than Treference.
EXPLAIN HOW and WHY reducing the power to 95% would affect ASI for the following means of reducing power.
- 1.
Control rod insertion (1.0) . 2.
Boron addition (1.0) 3.
Raising T(avg) (1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) _ _. . . _ _ _ _ _ -. _ _.
- _. - _.. -. . - _ -. _ _ - .
.___ ____ ____ ______ . 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
. THERMODYNAMICS QUESTION 5.10 (1.50) Assume the following conditions for a Steam Generator: .' i feedwater temperature = 400 degrees F j
i feedwater flowrate = 1.0 x 10**7 lbm/hr
steam pressure = 1000 psia
HOW much themal energy in MW is being transferred to the secondary fluid in the Steam Generator? (SHOWyour calculations.)
(1.5) QUESTION 5.11 (.50) Suberitical multiplication can be described as the process by which: (CHOOSE the correct answer).
(0.5) (a.) Delayed neutrons and their fissions in the fuel maintain power constant with Keff greater than 1.
(b.) Delayed neutron precursors generate source neutrons with Keff , less than 1.
(c.) Source neutron: are utilized to increase power with Keff slightly I greater than one.
(d.) Source neutrons and their fissions in the fuel maintain power ' constant with Keff less than 1.
i QUESTION 5.12 (1.00) GIVE two (2) reasons why, at BOL, the nuclear instrumentation countrate increases as reactor coolant temperature increases.
(ASSUME no rod movement or RCS boron concentration changes.)
(1.0) . , (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) } .-. _ _ ._ - - _ _ _ -. __ . _. - - - _ - _ - _ - _ - - - _ _ _ . . _.
._ . _ _ _ _ - _ - _ . 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
. THERMODYNAMICS QUESTION 5.13 (2.00) ' A PORV on the pressurizer is leaking.
Primary pressure is 2250 psia and the PRt pressure is 20 psig.
}' Qatuk%Y a.
WHAT is the temperature in the tailpipe? (1.0) b.
WHAT is the condition of the fluid in the ta11 pipe (supersaturated, saturated, or % moisture)? (STATE alcumptions and SHOW calculations or values from Hollier chart.)
(1.0) QUESTION 5.14 (1.50) SELECT the letters of the most. correct responses from those listed below. The responses are in answer to the statement, " Water hammer can be caused by...." (1.5) (a.) operating a pump with too high of a net positive section head.
(b.) operating a pump near the peak of its efficiency curve.
(c.) suddenly closing a valve in a pipe containing flowing liquid or gas.
(d.) starting a feed pump when the feed ring is filled with steam.
. QUESTION 5.15 (2.00) Toward the end of cycle, action must be taken to maintain Keff = 1 as the fuel burns up. These actions (d. through d.) are listed, STATE whether power and RCS temperature can be maintained at their 100% values or are allowed to decrease.
(Each action below requires two (2) responses: one (1) for power and one (1) for temperature.)
-(2.0) a.
Dilution of RCS boron concentration.
. b.
Deboration of ion exchanger placed in service.
c.
Turbine control valves are gradually opened to the full-open position.
d.
Coast-down is performed.
(***** END OF CATEGORY 05 *****) { .
- _ _ _ _ _ _ _ _ _ _ , 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
' QUESTION 6.01 (1.00) When making a load change, the procedure directs you to switch . from channel Y of the reactor regulating system on C04 to - channel X when T(avg) is between 556 F and 559 F and switched . ' back to channel Y when outside of this temperature band.
EXPLAIN WHY this is done.
(1.0) I , QUESTION 6.02 (1.00) DESCRIBE the design feature that assures an unrestricted flow path for the containment air. recirculation units during a loss of coolant incident.
(1.0) , QUESTION 6.03 (1.50) WHAT plant and ESAS conditions would be present if, at the ESAS actuation cabinets, the SIAS block modules are tripped (green lights) but the block lights on the SIAS, CIAS and EBFAS actuation
modules are not lit? (1.5) . j QUESTION 6.04 (1.00) The comparator averager provides an input to which of the following? (1.0) a.
power ratio calculator b.
TM/LP setpoint calculation c.. axial ' offset calculation d.
-99W calculation PNB QUESTION 6.05 (1.50) f LIST two (2) purposes of the steam bypass spray valves which open - t independently on temperature switch signals located near the main steam bypass lines? (1.5) f (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) < _ _ _ _ .. _ __- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ ._
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - __ _ ___- _ __ _ _ _ - ____ _________ ____________ _ -_.
- . 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE Bo , QUESTION 6.06 (2.00) LIST four (4) signals that will cause the feedwater regulating valves to " lockup" (solenoids de-energize).
(2.0), , . QUESTION 6.07 (4.00) a.
NAME two (2) of the four (4) turbine battery system loads (do not include inverters or distribution panels).
(1.0) b.
Using Figure 6.07, DRAW all the connections and components that tie Bus 24C to Bus 201A (125 VDC (D01)) and DRAW.the loads that are connected to Bus 201A. DO NOT include breakers.
(3.0) QUESTION 6.08 (2.50) i a.
WHAT three (3) regulating control rod interlocks / limits are in l effect when the control is in MANUAL INDIVIDUAL 7 (1.5) b.
WHAT are the two (2) instrumentation signals / conditions that could provide a DROPPED ROD annunciator? (1.0) QUESTION 6.09 (3.00) COMPLETE the following for the pressurizer pressure control program (control /alam/ event) when deviating away from the normal set point.
(3.0) 2350 psia A.
- 2300 psia B.
- i 2275 psia C.
2250 psia [ D.
Nomal setpoint 2225 psia E.
_ 2200 psia F.
2100 psia [ G.
- (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
. .
, I l - i
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$ b s i pot A tV 20t A 2V porn tV ' VsI AL Vil At Vil A1.
20s tg 2V Q 825 VOC Vtf AL n > 125 VOC 125 VOC 125 VOC i y
INSI PANEt . m ' tNSI PANEt INST l'AffEt INSI PANE L CD ' , O
(DV 10) M EDV 30) (DV 201 lDV 40) t-
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' DIST PANEt un H , $ (U t 33 (D 12) (D 2 t) tO 221 g
m
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2 VII#L VII AL YII AL 2 VtAC 4-4 120VAC Y'I#L H 120 VAC 120 VAC 5 20 VAC @ INST PANEt fNSI PANE t tNSI PANE t INSI PANEt T
(VA 10) - IVA 30) (VA 203 (VA 403
-' -B S2 --88I , nijs 209 D ntJS 20tp , ! k 125 VllC - t 25 vt1C - B St TUnelNE DAff Q
Tun 0lNE DAf f
$ - B 8I . 9 32 v ~ 8 48A T q note iI . 1"0 88 $ m .
' , Note t Adminestrat.ve Mer Contret $ ,
- -
FIGURE 6.07 (CUESTION) .. b l
i -
.
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ , 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
, QUESTION 6.10 (2.50) Eight (8) to twelve (12) hours after the initiation of a SIAS signal , due to a LOCA the operator will initiate boron precipitation control.
' HDW is this accomplished if the LPSI Systems are UNAVAILABLE 7 INCLUDE the flow path.
(2.5)*
QUESTION 6.11 (2.50) LIST the five (5) design criteria for the Emergency Core Cooling System.
(2.5) QUESTION 6.12 (2.50) , For the conditions listed below, indicate the ESAS channels that should actuate.
Consider the conditions in each part of this
question as seperate from the conditions in the other parts.
a.
S/G #1 pressure 800 psia S/G #2 pressure 500 psia S/G #1 level 68%, narrow range S/G #2 level 45%, narrow range Pressurized pressure 1800 psia Pressurizer level 20%, narrow Spent-fuel pool area rad.
range ' monitors 150 mr/hr (1.0) b.
S/G #1 pressure 900 psia S/G #2 pressure 900 psia S/G f1 level 68%, narrow range S/G #2 level 68%, narrow range Pressurizer pressure 1500 psia Pressurizer level 33%
RWST level 8% Containment pressure 30 psig , spent-fuel building radiation sensor channels (all) reading 100 mr/hr.
(1.5) i l . .
. (***** END OF CATEGORY 06 *****) .. _ _ _ _ _ _ _ _ _ - _ - - _ - _. -.. _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _
. . 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10 . RADIOLOGICAL CONTROL QUESTION 7.01 (2.00) ', During the refueling mode with the vessel head removed and core alternations in progress, a chemistry sample indicates the boron - concentration in the primary system is 1620 ppm.
' a.
WHAT are your actions? (1.5) b.
HOW soon do you have to perform these actions? (0.5) QUESTION 7.02 (2.00) LIST four (4) plant conditions that would indicate that a loss of primary coolant has occurred.
(2.0) QUESTION 7.03 (1.00) WHAT operation is required if the RCS boron concentration is to _ be changed by greater than 50 ppm? (1.0) , ! QU /.u4 %(1.50) pELETED ~ LIST three (3) symptoms for which emergene is mandatory in accordance with AOP 2558, " Emergency Boration."
(1.5) QUESTION 7.05 (3.00) LIST the six (6) safety functions that are in the safety function status check in the order of importance.
(3.0) ' - , QUESTION 7.06 (1.00) WHAT are the maximum ADMINISTRATIVE limits for quarterly and i annual NEUTRON exposure? (1.0) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) . -. - _ _ - - _ -. - _ _ - - _. - -_ . _ - . _.-
.. _ _ _ _ _ _ _ - _ _ . 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11 . RADIOLOGICAL CONTROL ! QUESTION 7.07 (.50) Answer TRUE or FALSE.
', The primary purnose of the Combined Intermediate Valves is to . maintain turbine speed after an Rx trip to prolong RCP coast l down.
(0.5) QUESTION 7.08 (1.50)
During a reactor trip recovery, the procedure cautions the secondary plant operator to not overreact to a low steam generator level and
to add feedwater slowly. ~This is done to avoid WHAT three (3)
conditions? (1.5) '
QUESTION 7.09 (3.00) ' Assume during a natural circulation cooldown auxiliary spray becomes
inoperable. GIVE three (3) alternate methods of depressurizing the RCS (1- ^*da" af pref:rence).
(3.0) ! I t QUESTION 7.10 (2.00) LIST the four (4) conditions or indications that require you to emergency borate per AOP 2558 - Emergency Boration.
(2.0) l QUESTION 7.11 (.50) WHAT is~ the maximum exposure a visitor can receive during a visit? (0.5).
. QUESTION 7.12 (2.00) During a natural circulation cooldown of the RCS the potential for voiding the reactor vessel head exists. GIVE two (2) control room indications of voiding.
(2.0) (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) -. -, -, - -..-_--- --- - -. _.- . - -... - - _ - - _ _ _ _ _ - - . _ - - - -. . - _ _ - - -.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - _ - - -. _____ _, . 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12 . RADIOLOGICAL CONTROL < QUESTION 7.13 (1.00) When shutting down from outside the control room and with automatic '. i boration unavailable - i a.
WHAT is the source of makeup to the RCS? (0.5) b.
WHY is that source selected? (0.5) i QUESTION 7.14 (2.00) While placing the volume control system in service during a plant i heatup, you are cautioned to maintain the RCS pressure below.265 psia by operation of the back pressure control valves.
EXPLAIN WHY this is required.
(2.0) QUESTION 7.15 (1.50) WHAT three (3) safety precautions should be taken when manually opening or closing a 480 V, 4160 V or 6900 V b eaker locally? (1.5) i , QUESTION 7.16 (.50) WHAT depressurization event ha: made it necessary to trip all reactor coolant pumps at 1600 psia? , (a.
Main steam rupture at 100% power.
(b.
RCS cold leg small break.
(c.
Main steam rupture at 5% power.
(d.
RCS hot leg small break.
(0.5) ' l . (***** END OF CATEGORY 07 *****) . _ _ _.. _ __ __-__. _ _ _ _ __- _...._._._________ _ _ ___ ___ _.
_ _ _ _ _ _ _ _ _ _ _ _ - - _. _ _ _ _
_ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - - -. ' ' 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 13 , l QUESTION B.01 (3.00) Complete the following table.
(3.0).
Mode Reactivity, K(eff) % Power T(avg) 1. Power Operation > or = 0.99 )5 ) or = 300 F 2. Startup ) or = 0.99 3. Hot Standby ( 0.99 4. Hot Shutdown ( 0.99 5. Cold Shutdown 6. Refueling QUESTION 8.02 (1.50) ANSWER this QUESTION by " filling-in-the-blanks".
{ In Mode 1 at Unit 2, the LCO for the Auxiliary Feedwater System ' requires that at least three (3) independent Steam-Generator Auxiliary Feedwater Pumps and associated flow paths shall be operable. The bases for this LCO is that the Auxiliary Feedwater System ensures that (1.5) . QUESTION 8.03 (1.50) ANSWER the parts of this QUESTION by " filling-in-the-blanks".
' a.
At Millstone Unit 2, the Technical Specifications specify an allowable peak linear heat rate for operation in Mode 1.
At BOL, this limit is (13.6,15.6, or 17.6) kW/ft, while at EOL 1t is (13.6,15.6, or 17.6) kW/ft.
(1.0) b.
One method of determining that the limit in (a.) is being met requires verification that the ASI is within its limits.
If ASI was 0.0, then operation at a maximum of % of rated - thermal power would be allowed.
(0.5) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) ..
. 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE I4 , QUESTION 8.04 (2.00) Assume a Safety Limit Violation has occurred at Unit 2.
According to the Unit 2 Technical Specifications, WHAT two (2) actions must , , be taken within I hour?. (2.0)- QUESTION 8.05 (2.00) DEFINE IDENTIFIED LEAKAGE by LISTING the two (2) types.
(2.0) QUESTION 8.06 (2.00) WHAT four (4) conditions must exist for a safety injection tank to be considered operable in Modes 1, 2, and 37 (2.0) QUESTION 8.07 (1.00) Answer TRUE or FALSE.
a.
WHEN classifying an event, the Shift Supervisor should use only the conditions as they exist at the time.
(0.5) b.
After the Shift Supervisor is relieved from his initial function as acting director of station emergency operations he assumes the position of manager of control room operations.
(0.5) QUESTION 8.08 (2.00) In regard to safety tagging, WHAT two (2) actions must be taken if the Operator in Attendance, who is responsible for valve and/or breakers must leave the job site for more than one (1) hour? (2.0) . QUESTION 8.09 (2.50) Related to procedures, DEFINE an INTENT CHANGE and give three (3) examples.
(2.5) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) .. ._.
._ . .-
__ . 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS ~ PAGE 15 - , r QUESTION 8.10 (1.50) ' WHAT is required for the approval and implementation of a non- . intent change to a procedure ~l (1.5) ' QUESTION 8.11 (2.00) Several types of non-emergency incidents are classified as GENERAL INTEREST EVENTS (echo). LIST four (4) of these types of events.
(2.0) QUESTION 8.12 (2.00) DEFINE or DESCRIBE the following terms as they apply to the Technical Specifications: a.
Operable-Operability (1.0) - b.
Limiting Conditions for Operation (1.0)
QUESTION 8.13 (1.00) a.
Liquid waste discharges beyond the quarterly limit of 1.5 Ci organ /5 Ci whole body is permitted with the concurrence of (position).
Je&}-- (1,o) uuse soioe fer ligid dir9mes are made assuming a dilution
u.
' flow of cubic feet / min as per Aaminisi. tiv: Procedure ACP 6.03.
D ELETED f+rk 6, QUESTION 8.14 (.50) The local power density trip is intended to: (CHOOSEone) (0.5) (a. prevent exceeding 15.6 kW/ft linear heat rate (b. prevent fuel centerline melt - (c. prevent exceeding a clad surface temperature of 2200 deg F (d. prevent exceeding a minimum DNBR of 1.3 (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) . - _ _-- - - - - _
. . B.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE I6 . QUESTION 8.I5 (.50) ANSWER TRUE or FALSE.
, . The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 degrees F.
(0.5) . I ! . l (***** END OF CATEGORY 08 *****) l l (************* END OF EXAMINATION ***************) ! l l -.. .. - .. .. .. -.. _ _ _ _.. _ _ .. _. .. _ _. _.,,.. _.... - _. _ _ -. - -. _ - -
l . ' 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17 THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 5.01 (.50) - (c.)
[+0.5] REFERENCE 1.
Millstone 2: Hot License School, " Reactor Theory and Operating Characteristics," Lesson #2 - Enabling Objectives.
ANSWER 5.02 (.50) (a.), (b.), and (c.)
[+0.5] REFERENCE 1.
Villstone 2: Hot License School, " Reactor Theory and Operating Characteristics," Lesson #3 and #4 - Enabling Objectives.
, ANSWER 5.03 _(3.50)
a.
rho = (Keff-1)/Keff = (0.95-1)/0.95 = -5.26% delta k/k [+1.0] b.
CR2/CR1 = (1-K1)/(1-K2) K2 = 0.975 [+0.5] rho = (Keff-1)/Keff = (0.975-1)/0.975 = -2.56% delta k/k [+1.0] c.
NONE,' assuming the highest worth rod is not in group 1 or 2.
[+1.0] REFERENCE 1.
Millstone 2: Lesson Plan 2116-000111.
2.
Millstone 2: Hot License School, Reactor Theory and Operating Characteristics, Lesson #7, Enabling Objectives.
! -._ _ _ .- .. .--. . ____ - -- -_ . . ' 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE.18 THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 5.04 (3.00) - a.
P = Po 10**(SUR)(t) 4 x 10**6 = 25 x 10**(0.4)(delta T) delta T = 13 min [+1.0j b.
Assune Beta (eff) = 0.006 at BOL = 0.005 at EOL landa = 0.08 t = 26/SUR = 26/0.04 = 65 sec rho = Beta (eff)/(1+1amda*t) = 0.006/1+(0.08x65) = 0.096% delta k/k [+1.0] EOL T = (beta - rho)/(lamda* rho) = (0.005-0.00096)/(0.08x0.00096) =-55.2 ;cc - 57.2,(6 SUR = 26 = & 4M DPM [+1.0] o.19 5 REFERENCE 1.
Millstone 2: Lesson Plan 2116-000126.
2.
Millstone 2: Hot License School, Reactor Theory and Operating Characteristics, Lesson #7, Enabling Objectives.
ANSWER 5.05 (1.50) a.
Decreases b.
Increases , ! c.
Decreases [+0.5] each REFERENCE 1.
Millstone 2: Lesson Plan 2121-000619.
2.
Generic: Fluid Mechanics and Hydraulic, Giles, McGraw-Hill.
l t - _ _, - - - - .- - _ - -,. - - -, - _ -. _ , _, - - - -.
. . ~ ' PAGE 19 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, Af!D THERMODYNAMICS 86/07/07-SMITH, J.
ANSWERS -- MILLSTONE 2 - ANSWER 5.06 (1.50) - a.
Higher than expected
- +0.5; b.
Lower than expected , 0. 5, + c.
Lower than expected +0.5 REFERENCE 1.
Millstone 2: Lesson Plan 2116-000465.
2.
Millstone 2: Hot License School, Reactor Theory and 0perating Characteristics, Lesson #9, Enabling Objectives.
ANSWER 5.07 (1.00) Heat transferred to the cooling water Q = m*cp* delta T = 1000 lbm x 1 BTU /lbm-deg F x 35 deg F Q = 35000 BTU /hr [+0.5] Therefore, delta T = (35000 BTU /hr)/(600 lbm/hr x 1 BTU /hr/deg F) 120 deg F - 58.3 deg F = 61.7 deg F outlet temp.
[+0.5] REFERENCE ., 1.
Millstone 2: Lesson Plan 2121-000837.
l l l l
I I -., _. - - - -. _ __ -..-......._-.-._____....,--..,,,,,4.- ,,-. ,.y ,,
. . ' 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 5.08 (2.00)
The presence of delayed neutrons drastically increases the average neutron generating time.
[+0.5] Delayed - 12 - c.
- +0.33;
, 0.33, Prompt = 10** 4 ;;cm + 0.33 Average = 0.0? ret + , The rate at which the neutron population can increase is limited by the average neutron generation time.
The average neutron generation ti.ae determines the number of generations that can occur in a given time period.
[+0.5] REFERENCE 1.
Millstone 2: Lesson Plan 2116-000112.
2.
Millstone 2: Hot License School, Reactor Theory and Operating Characteristics, lesson #8, Enabling Objectives.
ANSWER 5.09 (3.00) 1.
Adding negative reactivity to the top of the core causes the ASI to become more positive (power is driven to the bottom of the core). [+1.0] 2.
A change in T H as power decreases is greater than the change y clua_. e pesi//ve_ in T c.
With a -MTC, less negative reactivity is inserted in g.,.fY g,44, . the top of the core than the bottom.* Also, the MTC is more negative at the temperatures at the top of the core. [+1.0] 3.
4ai;ing T( vg) adda negative reectiv;iy uniformly across the enra cand ng the a.SI nei iv dmwge#[+1.0] AS.T Ac 4.~d yYe=Ny e e REFERENCE 1.
Millstone 2: Lesson Plan 2116-000726.
oP no4 ced C%< School, Reactor Theory and Operating
Millstone 2: Hot License 2.
Characteristics, Lesson #11, Enabling Objectives.
+ As S/c & & k < W T' W A A A C ' % sk n k A m. D A L &W k a, a s.
M r r~ A ~. % '~'# #
- ;M n w - A h. Bd M TC L M
^~'3 s4-A 4<g cf A e + L J # W w *L'-9 N' $'*k
_ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ . 5.
THEORY OF NUCLEAR POWER PLANT' OPERATION, FLUIDS, AND PAGE 21
THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-5MITH, J.
ANSWER 5.10 (1.50) ' Q = m*(h(steam) - h(feedwater)) [+0.5] = 1.0 x 10**7 lbm/hr (1192 BTU /lbm - 375.1 BTU /lbm) = 817 x 10**7 BTU /hr [+0.5] = 817 x 10**7 lbm/hr/3.41 x 10**6 BTU /hr MW = 2396 MW [+0.5] REFERENCE 1.
Generic: ACADEMIC PROGRAM FOR NUCLEAR POWER PLANT PERSONNEL, Volume III, " Nuclear Power Plant Technology," 1973, General Physics Corporation, pp. 2-141 through 2-148.
ANSWER 5.11 (.50) (d.)
[+0.5] REFERENCE 1.
Millstone 2: Lesson Plan 2116-000121.
2.
Millstone 2: Hot License School, Reactor Theory and Operating Characteristics, Lesson #12, Enabling Objectives.
ANSWER 5.12 (1.00) a.
Neutrons travel further, more leakage.
[+0.5] b.
Lower boron density, less boron absorbtion.
[+0.5] C' A f * M'** b O ^ [" *"M C' REFERENCE 1.
Millstone 2: Lesson Plan 2116-000674.
2.
Millstone 2: Hot License School, Reactor Theory and Operating Characteristics, Lesson #9, Enabling Objectives.
.. - - _ __ .- .- .-. ... . - - -. - _ _- - _ _ -
. .
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 5.13 (2.00) - a.
From Mollier chart (constant enthalpy) ( sat. line [+0.5] therefore temperature in pipe = T(sat) for 20 psig (35 psia) = 259 F [+0.5] b.
hg 2250 = 1117.7 constant enthalpy process [+0.5] hg 35 psia = 1166.5 hf 35 psia = 227.5 xey = 1 calc 227.5x + 1166.5y = 1117.7 5% moisture [+0.5] REFERENCE 1.
Generic: APPLIED THERMODYNAMICS, Faires, MacMillan.
ANSWER 5.14 (1.50) The answer is both (c.) and (d.). [+1.5 for both, +0.5 for (c.)
only, +0.5 for (d.) only] REFERENCE 1.
Millstone 2: NRC Reference Data, Book 16, Exam Bank.
ANSWER 5.15 (2.00) Power Temperature a.
maintained [+0.25] maintained [+0.25] b.
maintained [+0.25] maintained [+0.25] l c.
maintained [+0.25] decrease [+0.25] d.
decrease [+0.25] decrease [+0.25] ' REFERENCE 1.
Hillstone 2: Lesson Plan 2116/7-000195.
l 2.
Millstone 2: Hot License School, Reactor Theory and i
- - _ _ _ . . 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23
THERMODYNAMICS ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
Operating Characteristics, Lesson #9, Enabling Objectives.
. . -__ _ .. . - _-_ _ _ .-. - -. -_ .. .. ._
-, .
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 6.01 (1.00) . Done to prevent the electronic noise generated by the quick open relay (4-7) of channel Y from causing inadvertent and undesirable equipment and instrument response. [+1.0] REFERENCE 1.
Millstone 2: OPERATING PROCEDURE, OP 2204, p. 7 2.
Millstone 2: SYSTEM DESCRIPTION, "Rx REGULATING," Section IV A, p. 10.
ANSWER 6.02 (1.00) To assure an unrestricted flow path, the ductwork system which is located on the containment air recirculation fan discharge is separated from the fans by fusible link plates located immediately downstream of each fan. [+1.0] REFERENCE 1.
Millstone 2: SYSTEM DESCRIPTION, " Containment Ventilation," Rev.
1, p. 4.
. ANSWER 6.03 (1.50) Pressurizer pressure is less than or equal to 1750 psia [+0.5] with SIAS block permissive present, [+0.5] but the block buttons on CO-1 have not been pushed.
[+0.5] REFERENCE 1.
Millstone 2: Lesson Plan 2384-000370.
2.
Millstone 2: SYSTEM DESCRIPTION, "ESAS," Book 9, No. 4.
ANSWER 6.04 (1.00) a.
[+1.0] power ratio calculator (which computes set point).
-_.
. . .- _ - _ - - _ - - _ -.
. . ' 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
. REFERENCE 1.
Millstone 2: Lesson Plan 238-1-000432.
. . ANSWER 6.05 (1.50) The purpose of this spray is to prevent possible cavitation at the condensate pump suction [+0.75] and loss of vacuum durin when the main steam dump / bypass valves are open. [+0.75]g periods REFERENCE 1.
Millstone 2: SYSTEM DESCRIPTION, " Condensate System," Rev.1, Section 5.0, p. 7.
ANSWER 6.06 (2.00) 1.
Low instrument air supply header pressure (60 psig).
2.
Low controller output air pressure (3 psig).
3.
Loss of control power.
4.
Low controller electrical signal output.
[+0.5] each REFERENCE 1.
Millstone 2: SYSTEM DESCRIPTION, "Feedwater System," Section IIIC, p. _ . . . ' 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 a ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 6.07 (4.00) . a.
Main turbine energency bearing oil pump, emergency hydrogen seal oil pump, 2 SGFP turbine emergency oil pump.
t Any two (2) [+0.5] each.
' 0.4' ' 0.4' 2-INV-1 b.
Charger 201A + + ' 0.1' ' 0.1' 2-INV-3 Charger 201C + + Battery 201A l+0.4l VIP 1 Static Switch l+0.4l 2-INV-5 +0.4 VIP 1 Static Switch +0.1 ' 0.4 Transformer 24 to 22 l+0.3l Bus 22E + , Reg 120 VAC inst panel [+0.2] [+3.0] maximum REFERENCE 1.
Millstone 2: Book 9, System 9, "125 VDC and AC," Figures 1 and 8.
ANSWER 6.08 (2.50) a.
1.
Upper and lower electrical limit [+0.5] 2.
CEA withdrawal prohibit [+0.5] o,- cus P 3.
CEA motion prohibit [+0.5] er cM r . b.
1.
Rod drop from Reed switch [+0.5]. 2.
NI negative rate of power change from NI system.
[+0.5] I REFERENCE 1.
Millstone 2: Lesson Plan 2302.
2.
Millstone 2: SYSTEM DESCRIPTION, " Reactor Regulating," Book 9, No. 5.
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- NVE 84 If I'
WViHIf R .gyg pq g g n i N TAL l VIP Il (VIP 31 (VIP 2) (VIP 41 0t82 $ N TAL N V IAL N NW q 125 VDC r --7 r--
825 VOC t25 VDC r-7 e-
F T __-jg.
DISI PANEt DIST PANEL I~L _F ' 3 _-if OfSi PANEL I g _i e DeSi PANEt , e7 rs
eT * j t_U . o s s , i t.
-.4 ._ _.
s _ _a _ 10 11) SI AflC ST ATIC 10 123 (D 2 tl . Switch SWIICH STATIC STAilC (D 22) Z , SWITCH SWITCH g
2 VtAC l 2-VI AC 3 2 VtAC 2 2 VtAC 4 VIT AL Vli AL VIT AL Vli AL 120 VAC 120 VAC 120 VAC 5 20 VAC ! sNST PANEt eNSI PANEL INSI PANEL INSI PANEL (VA 101 (VA 303 gvA 20) (VA 40) . {{^ ,,,c,,I-' f3Jn-.62 m . 52 . ! 0uS201D a my s - - -[ - Dus'20 t D N n 2 sNv s - - n
125 VDC - .Nvt fits n
- C ' g 125 VDC -
.Nvg tes t es & j - IVIP Si nEG 120 VAC <> B 32 (VIP Ol REG 120 VAC-M4A y g g B St IURDINE DAEI - 88i O TURotNE DATI ,' i INSI PANEL g nog,3 g INSI PANEL Inole il NON VII AL NON VIT AL , ' Note I Admenestratwo May Control - 120 V A t: 125 VilC ONE LINE DRAWING l
, . ' 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 6.09 (3.00) , A.
spray valves fully open, high pressure alarm - B.
spray valves start to open C.
proportional heaters at minimum D.
normal setpoint (given) E.
proportional heater at maximum F.
backup heater on G.
low' pressure alarm [+0.5] each REFERENCE 1.
Millstone 2: SYSTEM DESCRIPTION, " Pressurizer Level and Pressure Control," Figure 11.
ANSWER 6.10 (2.50) If the LPSI Systems are unavailable, the Facility 1 [+0.5] HPSI train can be used to inject water into the Pressurizer via the CVCS cross-tie and the Auxiliary spray valve. [+1.0] This established reverse flow in the core region such that water from the core is flushed out the cold leg break. [+1.0] REFERENCE 1.
Millstone 2: SYSTEM DESCRIPTION, "High Pressure Safety Injection," Section IV B2, p. 13.
ANSWER 6.11 (2.50) 1.
Limit fuel cladding temperature to less than 2200 deg F clad [+0.5] i ! 2.
Limit fuel clad oxidation to less than 17% oxidation thickness [+0.5] 3.
Limit H 2 production from Zr/H2O reaction to less than 1% of i all Zr in core [+0.5] ' Maintain coolable core geometry][+0.5] 4.
Provide long term cooling [+0.5 5.
! REFERENCE i l 1.
Millstone 2: Lesson Plan 2306-9-000381.
_ - -. . _ - - _ _ _ - -____ - . -. - - _ - - _ _ _ _ .. _ - - . _.
_. _ - . . '
6.
PLANT SYSTEMS DESIGN CONTROL, AND INSTRUMENTATION PAGE 28 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
i 2.
Millstone 2: SYSTEM DESCRIPTION, "HPSI, LPSI," Book 8, Nos. 5 and 6.
ANSWER 6.12 (2.50) - a.
AEAS MSI M ri ',T, each ' o....,,2 b.
SIAS CIAS EBAS CSAS ACA5 - SRAS to.3 EG.25] each REFERENCE i 1.
Millstone 2: Book 9, SYSTEM DESCRIPTION #4, "ESAS," pp. 3-7.
oP 2:3 4 c-p 7 wh 4, /_ , l l l . , r -,n-. -- --- -, ,, -. - - -. - -, .-----,_ -.,.,,,,- -,, _.- -,..------,, 7-y, - --,,w-- -,--,---,----~e,- n - -
, ' 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 7.01 (2.00)
a.
1.
Suspend all core alterations [+0.5] 2.
Restore boron concentration >1720 ppm or K(eff) (0.95 [+0.5] 3.
Do not increase reactivity in the core [+0.5] b.
15 min.
[+0.5] REFERENCE 1.
Millstone 2: Lesson Plan 2303.
2.
Millstone 2: TECHNICAL SPECIFICATION 3/4.9.1, " Boron Concentration."
ANSWER 7.02 (2.00) 1.
abnormal decrease in pressurizer level S~,,hereasy cadetshM(M5SM4 2.
decreasing pressurizer pressure 6,ikemsig g w ( M A b 3.
high containment radiation
- M W 7. kk M b A
3' 4.
unbalanced charging and letdown flows [+0.5] each 4 of 7 g a,Q, REFERENCE 1.
Millstone 2: EMERGENCY OPERATING PROCEDURE, E0P 2533, Rev. 2, Section 2.a. p. 2.
P Ea r 1S n % ~ 2 i 3 2-l ANSWER 7.03 (1.00) . ! Operate the sprays to equalize the boron concentration in the ' pressurizer. [+1.0] l REFERENCE i ' 1.
Millstone 2: OPERATING PROCEDURE, OP 2201, Rev. 14, Section 4.19, p. 6.
! l . . - - - _. - - -, .- - - _ _ _ _ ,. -- - -,___~,. __-_,__... --.- - -... -. -. _. _
.- , . ' 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 2-86/07/07-SMITH,J.
ANS .7,04 (1.50) ' 1.
Two or regulating or shutdown CEAs have not dropped into the core fo ing a reactor trip signal.
4-2.
Unanticipated react ooldown has been initiated.
3.
PDIL alarm is anunciated.
4.- When the plant is in the shutdown o he refueling condition
(NI or audible count rate), an unexpla increase in reactivity occurs.
Any three (3) [+0.5] each, +1.5 maximum.
REFERENCE I G L F 6' D . 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, " Emergency ! Boration," AOP 2558, Rev. O, Section 2, ENTRY CONDITIONS, e p. 2.
, i ANSWER 7.05 (3.00) 1.
Reactivity control 2.
RCS inventory control . 3.
RCS pressure control ! l 4.
RCS heat removal l 5.
Containment integrity ( $ [,c. plu c$ a ~ c e ll "s 4 [ " ce ) 6.
Vital auxiliary [+0.5] each l REFERENCE 1.
Millstone 2: EMERGENCY OPERATING PROCEDURE, E0P 2526, OPS fom 2526-1, pp. 3-5.
l l !
. - _ _ -. .. . . .-_ . ' 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 2-86/07/07-SMITH,J.
ANSWER 7.06 (1.00) ' Quarterly - 300 mrem [+0.5] Annual - 500 mrem [+0.5] ' REFERENCE 1.
Millstone 2: SHP 4902, Rev. 8, Section 8.1.2.11, p. 10.
ANSWER 7.07 (.50) FALSE [+0.5] (To prevent turbine overspeed following an Rx trip.)
REFERENCE 1.
Millstone 2: Lesson Plan 2323A.
l ANSWER 7.08 (1.50) 1.
Excessive pressurizer level and pressure transient.
[+0.5] 2.
Excessive cooldown rate.
[+0.5] 3.
Overfilling steam generators.
[+0.5] REFERENCE 1.
Millstone 2: EMERGENCY OPERATING PROCEDURE, E0P 2526, Rev. 1, Section 3.5, p. 5.
ANSWER 7.09 (3.00) K'depressurizetheRCS.
Fill and drain the pressurizer to cooldown and thereby [+1.0] K Depressurize the pressuri:2r by ambient cooling. [+1.0] % Open a PORV as needed to reduce RCS pressure.
[+1.0]
.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND-PAGE 32 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
REFERENCE 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, A0P 2553, Rev. 1,
Section 4.6, p.
6.- . ANSWER 7.10 (2.00) 1.
2 or more CEAs not inserted after an Rx trip 2.
unanticipated RCS cooldown 3.
unexplained increase in reactivity when in shutdown or refueling 4.
PDIL alarm is annunicated [+0.5] each REFERENCE 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, A0P 2558.
ANSWER 7.11 (.50) 300 mrem [+0.5]. REFERENCE 1.
Millstone 2: SHP 4902, Rev. 8, Section 3.2.1.2, p. 16.
ANSWER 7.12 (2.00) 1.
Pressurizer level increases greater than expected while using auxiliary spray.
[+1.0] 2.
Pressurizer level decreases while operating a charging pump or a HPSI pump. [+1.0] REFERENCE 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, A0P 2553, Rev. 1, Section 3.2a/b, p. 2. d E o P 2 F34- /b-E $9 3,2o
- 3 NstQ DJ MA k A~ p,pM h NM YY$
$$ w Ja A%& +.. C M vu%eA A-<l M As4 M- /do2 (rcc 4;p/y
.
7. ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND .PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 7.13 (1.00)
a.
Makeup to the RCS must be from the RWST [+0.5] b.
To insure that the boron concentration is greater than or equal to that of the RCS. [+0.5] REFERENCE 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, A0P 2551, Rev. 2, Section 4.20h, p. 7.
ANSWER 7.14 (2.00) To prevent the shutdown cooling system from isolating [+1.0] and the safety injection tank outlet valves from auto opening. [+1.0] REFERENCE 1.
Millstone 2: OPERATING PROCEDURE, OP 2201, Rev. 14, Section 5.1.3, p. 9.
ANSWER 7.15 (1.50) Stand to the side of the compartment, [+0.5] extend only[+one arm across the front of it [+0.5] and turn your face away.
0.5) REFERENCE 1.
Millstone 2: ABNORMAL OPERATING PROCEDURE, A0P 2551, Rev. 2, Section 3.1/3.2, p. 2.
ANSWER 7.16 (.50) (d.)
[+0.5] REFERENCE 1.
Millstone 2: Lesson Plan 2532-000443.
2.
Millstone 2: E0P 2532, " Loss of Primary Coolant."
r . 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 34 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 8.01 (3.00) , Mode Reactivity, K(eff) % Power T(avg) 1. Power Operation > or = 0.99 >5 ) or = 300 F (given) (given) (given) (given) 2. Startup > or = 0.99 ( or = 5 > or = 300 F (given) (given) 3. Hot Standby ( 0.99
> or = 300 F (given) (given) 4. Hot Shutdown ( 0.99
300 F > T(avg)) (given) (given) 200 F 5. Cold Shutdown ( 0.98
( or = 200 F (given) 6. Refueling ( or = 0.95
( or = 140 F (given) [+0.25] each REFERENCE 1.
Millstone 2: TECHNICAL SPECIFICATIONS, Definitions, Table 1.1, pp. 1-7.
ANSWER 8.02 (1.50) the RCS can be cooled-down to less than 300 degrees F [+0.5] from normal operating conditions [+0.5] in the event of a total loss of off-site power [+0.5]. REFERENCE 1.
Millstone 2: SAFETY TECHNICAL SPECIFICATIONS, Basas, Auxiliary Feedwater System, Section 3/4.7.1.2, p. B 3/4 7-2.
ANSWER 8.03 (1.50) ' 0. 5' a.
15.6 + ' 0.5l 15.6 + , [+0.5] or-IAW f:.3 2-1Ls cc 3,2-2 h j 5LSgec. I 2.2 b.
100%
REFERENCE 1.
Millstone 2: SAFETY TECHNICAL SPECIFICATIONS, Power Distribution
, .
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 35 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
Limits, Section 3/4.2.1, p. 3/4 2-1.
> ANSWER 8.04 (2.00) - 1.
The NRC Ope.otie..a Cc.ter-shall be notified by telephone.
2.
Unit placed in hot standby.
[+1.0] each REFERENCE 1.
Millstone 2: TECHNICAL SPECIFICATIONS, Section 6.7, p. 6-15/16.
ANSWER 8.05 (2.00) 1.
Leakage (except CONTROLLED LEAKAGE) into closed systems from . ' pump seals or valve packing [+that are captured and conducted to a sump or collecting tank 1.0] 2.
Leakage into containment atmosphere from specifically known sources but that is NOT PRESSURE BOUNDARY LEAKAGE [+1.0] REFERENCE 1.
Millstone 2: TECHNICAL SPECIFICATIONS, Definitions, p.1-3.
! l ANSWER 8.06 (2.00)
1.
Isolation valve open and power removed from the valve operator.
2.
Between 1080 and 1190 cubic feet of borated water.
3.
Minimum boron concentration of 1720 ppm.
4.
Nitrogen cover-pressure between 200 and 250 psig.
.[+0.5] each l I REFERENCE ! 1.
Millstone 2: TECHNICAL SPECIFICATIONS, Section 3.5.1, p. 3/4 5-1.
' l ,
.
' 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 36 ANSWERS -- MILLSTONE 2-86/07/07 SMITH, J.
ANSWER 8.07 (1.00) , ' 0.5' a.
FALSE.
+ b.
TRUE.
l+0.5 REFERENCE 1.
Millstone 2: EPIP 4010.
ANSWER 8.08 (2.00) 1.
the job must stop [+1.0] 2.
the valves and/or breakers must be tagged as listed on tag log sheet (SF210) [+1.0] REFERENCE 1.
Millstone 2: Administrative Procedure, ACP-QA-2.06A.
ANSWER 8.09 (2.50) An intent change is one which includes a change in the basic method of the procedure, a change which could endanger plant personnel or equipment, or a change to any acceptable criteria. [+1.0] Examples: 1.
A change that results in an alteration of the purpose, objective or applicability of the procedure.
2.
A change that alters acceptance criteria, setpoints or limits.
3.
A change that alters QC hold points.
l 4.- A change involving the application or performance of a special , l process.
5.
A significant addition or deletion of procedure steps which is not consistent with the original purpose of applicability of the procedure.
Any three (3) [+0.5] each, +1.5 maximum i)NIlcemp h~ spedht % fltS whit ' " " chees.
,_.
. . 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 37 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
. REFERENCE 1.
Millstone 2: ACP-QA-3.02, Rev. 35, Section 4.8, p. 5.
. . ANSWER 8.10 (1.50) The concurrence of TWO [+0.5] LICENSED SENIOR [+0.5] Reactor Operators (SR0's) from the PARTICULAR UNIT [+0.5] involved.
REFERENCE 1.
Millstone 2: ACP-QA3.02, Rev. 35, Section 6.8.1.3, p. 23.
ANSWER 8.11 (2.00) GENERAL INTEREST EVENT 1.
Toxic material release 2.
H= :rdeu: ws:tc or toxic waste spill 3.
Personnel emergency 4.
Atmosphere immediately hazardous to life h m gpre 470l-4, 5.
%mb threat , " Any four (4) [+0.5] each.
, REFERENCE 1.
Millstone 2: ACP-1.15, Rev. 9, p. 4.
I AEPIP V7o/- +
r, ' .
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 38 ANSWERS -- MILLSTONE 2-86/07/07-SMITH, J.
ANSWER 8.12 (2.00) , a.
A system, subsystem, train, component, or device shall be operable or have operability when it is capable of performing , its specified function (s) and when all necessary attendent instrumentation, controls, normal and emergency electrical power sources, cooling, or seal water, lubrication, or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are alsocapableofperformingtheirrelatedsupportfunction(s).
(The essence of the above for full credit.) [+1.0] b.
Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.
[+1.0] REFERENCE
1.
Millstone 2: Lesson Plan 2001-000350.
2.
10 CFR 50.36.
3.
Millstone 2: TECHNICAL SPECIFICATION, " Definitions," p.1.6.
ANSWER 8.13 (1.00) I stath/ uni /stationservicessuperintendent-[:0.5PG,o) j a.
REFERENCE l 1.
Millstone 2: Administrative Procedures, ACP 6.03.
i i l ANSWER 8.14 (.50) j b.
[+0.5] REFERENCE , 1.
Millstone 2: Lesson Plan 2380-1-000537.
2.
Millstone 2: TECHNICAL SPECIFICATION, LIMITING SAFETY SYSTEM SETTINGS BASES, 2.2.1, "LOCA Power Density-High," , l p. 32-6.
l l l l
. - ,. .
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39 ANSWERS -- MILLSTONE 2-86/07/07-SMITH,J.
ANSWER 8.15 (.50) . TRUE.
[+0.5] . REFERENCE 1.
Millstone 2: TECHNICAL SPECIFICATION BASES, 3/4.2.1, " Linear Heat Rate," p. B 3/4.2-1.
I i !
< A .- . ..___________________..._________...._____. ____ ______________ ____________ EQUATION FORMULA AND PARAMETER SHEET .._-_ ____.__________....___.__ ____________________ ______________...__ . . Where my = m2 (density)3(velocity)3(area)1 = (density)2(velocity)2(area)2 - _______________...___._-____..........__.____......_.......__.......____.. KE="{2 PE = mgh PE + KE +P V y y = PE +KE +P Y22 where V = specific,, y y
2 volume P = pressure , .__..___...__..__.__..._.__-__.._......__....._.____......_._...____.... Q=be(T-Tin) Q = UA (T,y,-Tstm) Q = m(h -h ) p out i2 ___.....___________.............. ____________..____..._____.._________... P = P 10(SUR)(t) p, p,t/T SUR = 26.06 T= = I# ~ #I " o o T p p A,ff ..______________......__________________._.....____..............___...__ delta K = (X,ff-1) CR (1-X,ff1) = CR (1-Keff2).
CR=S/(1-K,ff)
2 '
(1-Ketf1) (1 eft) x IOM , ,3 M = (1-Keff2).
K A'ff = 0.08 sec SDM = eff . ......_____ _.... _______________________..... _.........._____________ __._ ' y = A e-(decay constant)x(t) decay constant = in (2) " 0.693 A t t g 1/2 1/2 _________ ________________....________ _________________.._._________. ___ Water Parameters Miscellaneous Conversions
I gallon = 8.345 lbs 1 Curie = 3.7 x 10 dps I gallon = 3.78 liters I kg = 2.21 lbs
3 1 ft = 7.48 gallons I hp = 2.54 x 10 Btu /hr I
6 Density =62.4lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 Dm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 . Heat of Fus. ion = 144 Btu /lbm 1 inch = 2.54 centimeters -
! 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec
i l 1 ft H O = 0.4335 lbf/in.2
2
.......__......_______...___..._______________......______..____.________
1 . . -. _ _ _ _ _. _ _ _ _ _ ___ ,..,____._..m__,,____ _. _,. _ _ _,,. - _.. _. _, - ... _ _,,, _ ,, _.. _. _ _ _,. - _ _, _ _., _ _ _ -.... _ _ _ - _
ATTACHMENT 3 OPERATOR LICENSING' EXAMINATION . MILLSTONE 2 i July 7-11, 1986 Facility comments on the Written Examinations.
, , s , ,, Question Comment
- .
.. . T-
4 ' (See attached Sheets) ' < , - i \\ n
(] 1.02 'Ihe answer key gives only D. as the correct answer. Choice B.
'O is also correct (see attached reference sources). When the detector is too close to the source, it sees a large proportion of source neutrons and only a relatively small number of fission neutrons. As keff is increased, the increase in fission neutrons has a relatively small effect on detector reading. If the detector is far away frce the source, it sees more fission neutrons than source neutrons so an increase in the fission neutron population (caused by an increase in keff) has a relatively large effect on detector reading. This makes the 1/M plot, for a wide separation between source and detector, conservative.
, Since the question tells the student to choose only one answer, full credit should be given for either B. or D.
Reference: 1) Reactor Theory Iesson Plan 10-12 2) Manual of ::xper#---'- '-- "- "----~'--- \\ Reactor Facility, 1975 3). Experimental Reactor Physics, Profio,1976 4) Nuclear Engineering Handbook, Etherington,1958 i ! ( l I I i l l 0-1- ._ - - - _ _ - - _ -. -. _. . -- . _ _ - _ - _ - _ _ _ _ - - - - - - - _ -
_ _ - - - _ _.
. - _ . _. _ -... ._ _ _ __
1.09, Part b., Number 1) i The answer key gives the reason for the increase in PZR level j as being due to HPSI and charging flow. It can also be due to
void formation in the RCS (see attached references). Credit should also be' given for RCS voiding.
Reference: 1) Millstone 2, FSAR, Section 14.5 , 2) CEN-ll7, Inadequate Core Cooling - A Response to NRC IE Bulletin 79-06C, Item 5 For Combustion Engineering Nuclear Steam Supply l Systems, 1979
3) CEN-199, Effects of Vessel Head Voiding during , Transients and Accidents in CE NSSS's,1982
! 4) CEN-ll4, Amendment 1-NP, Review of Small Break I I Translents in CE NSSS's, 1979 i !,O ! !
i i l
i i i ! ! !O-2-t ! ,
-- .. - _ - - - . _ _ - . - -. f 1.09, Part b., Nunber 2)
J For a primary pressure of 1000 psia, the saturation tenperature
is 544.6 F.
If the CETs read 585 F, this indicates superheated conditions in the core which inplies scme amount of core uncovery. W e type of Natural Circulation which would exist in this condition would be Reflux Boiling. We answer key says 'No Phase. %e answer key should be changed to Reflux ! Boiling.
Reference: 1) CEN-152, Rev. 02, CE Bnergency Procedure , Q2idelines
2) CEN-ll7, Inadequate Core Cooling - A Response to NRC IE Bulletin 79-06C, Item 5 For CE i i NSSS's, 1979 " l O '
i
1 i
-3- ' - ! - - - - - -. -. - - - - - - - - -.. -. -.. - _ _ - _ - - _ _ - _ _,, _ . -. -.. -. _ -. _., - -. - _.. - _ - -..
. 2.06b The key answer of 12 plus or minus 3 seconds is a typical start O time for the D/G start time as described in the reference material. The time required by the Technical Specifications is less than or equal to 20 seconds.
The key should include an answer of less than or equal to 20 seconds or the typical time of 12 + 3 seconds.
Reference: T.S. 4.8.1.1.2.a.2 O t i
l
i } , I i
, i ! l l ! . . ' t-4-i { -..- -
2.07c %e question asks what is the reason for the limit on the ' backup source supply to 24E. We reason for the limit is the design rating of the breaker 15G - 21S and its associated bus.
his limit is above the capacity necessaty to supply one ECCS train. We fact that the bus has sufficient capacity to carry one ECCS train following an accident is not the reason for the limit.
We key answer should reflect that the reason for the limit is the capacity of the breaker and associated bus.
Reference OP 2343 Rev. 7 Caution af ter Step 7.23 In-House Electrical Distribution System Description, Page 3 ,
l ! i I
l l l ! ! !
l i ' -s- !
_ _ _. - - _ _ _ _ _ -. _ _ _ _. _ _. _. _.. _. _ _ _ _ _. -. _ _... _ _ _ _ _.. _.. _ _ _ _. _ _ _ _ _ _ i , i i 2.09 Also acceptable for the affect of loss of RBCCW to the CAR ! coolers would be no ccupromise due to the redundancy provided i ) by the contairinent Spray System.
,
I l Referencer Contairunent Spray System Description i f i ! I i ! t.
'
iO I ! i l
l ! i e ! l l t I \\ ' i ! ,
l ! l-6-t .' l i'.-.-
_. _ _ _ _.. _. _. - ....... - _ - ___ _ _ _ _ _.__ . _ _ _ - _ _. _ _ i i e $ ' i ! 2.10 Additional correct information may be supplied.
l
< , i In addition to the information stated in the key answer, the } ! discharge valve will close if the inlet valve is 'ciiened.
I l I i , Same reference as the key
/ i . i , & f i !
i j l '
I f i i
l8 . I i , , ! i I I ! < i f
i
\\ -7- ! ' - \\ \\ l i - ! l . -- . --
_ _ ._ . _ _.
- - - _ _. _ _ 2.llb More information may be supplied to describe how HV-306 is V electrically disabled than those given by the key answer. ne operating procedure for Shutdown Cooling Systen describes the actions taken to place 2-SI-306 in its blocked open position.
Any 3 of those 5 should be acceptable.
%e key answer should accept any three of the following f 1. Place the keylock switch to the "SI" position t 2. Remove the fuseblock for 2-SI-306 3. Isolate the air supply 4. We manual operator on the oppocite side of the valve shaf t is pinned and locked to the handwheel 5. %e handwheel is chained and locked in position Reference OP2310 Rev 8 Section 7.6.8 O ,v i . I l i , j-8-i
!
3.03 More conditions exist that will actuate a Power Trip Test Interlock than those listed by the key answer.
'Ihe following switches, when out of the operate position, result in a PITI.
1) A+B/2 switch (RPSCIP) 2) delta POWER CAlfUIA'IOR TEST switch (RPSCIP) 3) 2ERO-OPERATE-CALIBRATE switches (2) on NI Linear Power Channels 4) TRIP TEST switches (2) on NI LPCs.
In addition, a PITI is also initiated by tripping the LPC IW bistable.
'Ihe full credit should be given for any three of the above conditions.
i Reference RPS Imsson Plan 2380-1, Page 7.
' ) t i i
1 - , I l ! i !O , f-9-
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3.04 h setpoint for the needed condenser vacuum is 15" Hg. The selection of Channel Y affects the quick open characteristic - only, the valves will still function in response to the area l demand signal fran Reactor Reg.
In addition to the answer given, the Tavg controller nust be in auto for the B,C, & D valves to respond. Additional correct information is that 'A' valve will respond to a Reactor Reg input if that input is greater than the PIC signal.
W key answer should include: 1. Tave controller must be in AUID for the B,C, & D valves to respond to the area demand signal.
2. Condenser Vacutan is greater than 15" Hg 3. Tutbine Trip 4. Channel Y selected for quick open signal Reference: Reactor Hoc. System Description, Pages 10 14 O-10-
_ _ _ _ _ _. _ _ _ _ _ _. _ _ _ _ _ _ _ _.. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ ,
i i
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3.05 he D/G can always be tripped locally by the emergency stop a . pushbutton of the overspeed trip. %is information may be provided in addition to the information in the key answer.
t ! Reference: D/G System Description, Page 6 j i >
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a + ! 3.06 A SIAS will occur at a contairinent pressure less than 4.75
j ps ig. Only the low RCS pressure initiation of SIAS was . blocked.
j l s j 'Ihe key answer should include SIAS
! References ESAS System Description, Page 4 i.
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3.10 Auto-Nrmissive h start position starts the respective pung and opens both ' flow control valves. This information is not perfectly clear - in the system description referenced.
W key answer should accept that both valves are opened by the start position of the Auto - Permissive switch.
Reference: AEW System Description, Page 5 & 6 O O-13-
- - - .. -.. - . .. - -.= _ . .
3.ll.C High alarm shifts Condensate recovery tank discharge from
returning to Aux. Steam Feedwater Surge Tank (Turbine Bldg.) to aerated waste system. %e line to aerated waste is isolated by ' / a locked closed valve due to high tenperature fluid in Condensate Recovery Tank i
ne key answer should accept that a high rad alarm autanatically shif ts the flowpath from the Aux. Steam Feedwater Surge Tank to the Aerated Waste System.
We net effect of this shif t is the isolation of flow due to , the aligment of the rest of the system.
i Reference: 25203-26026 Aux. Steam & Condensate P&ID I OP 2331 Rev. O Section 8.5, Page 23 .
, , ! i !O-14-l . -.. .. -. -. . - .... - . - -. - - -
-._- = __ =. . -. - - _ -.. _ _ _ _ _ - __- - - l , ! , l 4.03 Wese actions are provided as specific guidance that applies only during the SIAS blocked phase of a plant cooldown. %ey i s i are included on Step 5.7.1 of the plant cooldown normal operating procedure. %ey are not considered "imediate
actions" as are the steps of EDP 2525. W ey are not expected to be available from memory. %erefore, any action that essults in safety injection flow should be accepted, such as a s . manual actuation of SIAS.
i j Reference: OP 2207 Rev. 14 Step 5.7.1 ' i
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{ -15-i ) l -. -. _.. _ _ _ _ _ _.,. - _ _. _ _ _ _, _.. _., _ _ _ ~ - _ _... ... . . . _ _,. -, ~. -
4.06.e High RCP seal tenperature response is provided in two Millstone p)
4 V 2 operating procedures. he answer key refers to the information presented in AOP 2564, Ioss of RBCCW. We information provided in OP 2301C, Reactor Coolant Pung operation, includes the "RCP lower seal tenperature hi" alarm response to initiate a controlled shutdown if tenperature stabilizes at 170 F and to trip the reactor if greater than 170 P.
Se question does not specify if the cause of the high seal tenperature is related to a loss of RBCCW or not, knowledge of which is necessary to select the correct procedure and response.
We answer key should reflect in the absence of other problems, that the reactor and pung should be tripped when seal l tenperatures exceed 170 F.
If the cause of the tenperature increase was due to a loss of RBCCW, then a higher limit of , i 250 F is inposed and the reactor and pump may continue to O nnerate.
Reference: AOP 2564 Rev 0 precaution 3.1 OP 2301C Rev 7 Section 8.15 O-16-
. _ - _ _. _ _ -.. - _ _ _ - _ -. - _- ... . -. -. . -. _. . _ - - _ _ .
' Q 4.08 %e hierarchy of safety functions is best depicted by the figure used in support of the EDP 2525 lesson plan.
It
indicates that the vital auxiliary safety function supports the
rest and that Pressure and Inventory Control Fall between -
l Reactivity and Heat Removal. We format of the,0ps Form 2526-1 does not allow this hierarchy to be presented. his is , ! additional correct information that may be included with the answer.
i References EOP 2525 Issson Plan Visual Aids
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s 4.09 The answer need not contain the phrase " allowing the MSIV to fail shut."
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4.11a 'Ihe values in parenthesis in the key answer need not be included for full credit as they were not asked for in the question.
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_. _ _ _ _ _ _ _ _ _ _. _. _ _ _. _ _ _ _ _... - - - _ _ _ _ _ _. _. _ _.. _. _ _ _ _ _ _ _..... _ _ _ _ _ _. _. _ _ _
' . i j 5.04, Part b.
!
! %ere is a math error in the answer key. S e reactor period at j DOL ccmes out to 52.6 sec for the numbers used, not 55.2 sec as i j stated in the answer key, his means that the EOL SUR for the amount of reactivity added is 0.495 DFM not 0.471 DEM as stated in the answer key.
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, h 5.08 'Ihe question can adequately be answered without mentioning the values for neutron generation times. 'Ihe answer key should be ! changed to reflect this.
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'5.09, Part 2.
'Ihe answer key does not fully answer the question asked. ne answer key should read: A change is T as power decreases is greater than the change in H T.
With a - MIC, less negative reactivity is inserted in the c top of the core than at the bottom due to the positive reactivity feedback frcm MIC. Also, MIC is nore negative at the tenperatures at the top of the core. 'Ihis causes ASI to becane nore negative (power is driven to the top of the core).
Reference: 1) OP-2204 Icad Changes, Form 2204-1
- 2) WCAP-10860, Figure 5.3 O
- Note: 'Ihis reference material is Propietary data and should not be made public.
. O-22-
(N 5.09, Part 3 % 2e answer to this question does not explain why raising Tavg adds negative reactivity uniformly across the core.
'1% effects nust be considered: the change in core delta T and the difference in the value of MIC between the top and the bottom of the core.
As SG steam flow is decreased, T increases. He hotter T c.
c entering the core reduces reactor power. At the lower reactor power there is a smaller delta T across the core. We net result is that T cmass nom &an T. But MIC is note c H negative at the top of the core than at the bottan so the effects are approximately off-setting and ASI does not significantly change.
Reference: 1) OP-2204 f. cad Changes, Form 2204-1
- 2) WCAP-10860, Figure 5.3
- Note: his reference material is Proprietary data and should not be made public.
r Ov-23- .-- -. . . .. . - _ _ __.
.. - - - _
d.
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5.12 An additional, correct answer to this question is that the O., nuclear instrumentation countrate increases due to a power increase caused by a positive MIC. Since the question specifles BOL, it is possible that a positive MIC exists. The MIC curves for Millstone 2 are attached.
A power increase due to a positive MTC ahould be added to the answer key as one acceptable answer.
Reference: Figure 5.1 and 5.13 of WCAP-10860 Note: This reference material is Proprietary data and should not be made public.
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_ _ 5.14 of the four choices given as possible answers to this question, O c. is undoubtedly correct. Choice d., however, may or may not be correct based on an inportant student assunption: whether or not feed flow reaches the feed ring.
Both the Main Feedwater Puaps and the Auxiliary Feedwater Punps at Millstone 2 are started with the flow control valves closed so starting a feedwater punp does not, in this case, supply water to the feedring.
Since choice d. requires that the student make an unstated assunption, we believe that full credit should be given for answering either c. alone, or both c. and d.
Reference: 1) OP 2321-2 Feedwater System Prestart Checklist 2) OP 2322 Auxiliary Feedwater System O
l
-25- _ - _ _ - _ -. _- _ _. _ _ - _ _ _ _ . . -. .. .. - _. _ . .. - . . -
_ __ 6.08 "CWP" is the sane as CEA withdrawal prohibit.
O "CMI" is the same as CEA notion prohibit.
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6.09a Additional correct information is that a signal is generated to ~
, ' open backup heater breakers.
, t i ! Reference: Pressurizer Ievel Control System Description j i Figure 11 t i r J
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6.12.a 2e key lists AEAS twice. he point value should be 0.5 for MSI and 0.5 for AEAS
b.
AEAS will not actuate with EBFAS present. his infonnation is provided in section 4.3 of OP 2314G and should be reflected in the key answer.
-
References: OP 2314G Rev 7 Section 4.2 ' ESAS system description, page. 6.
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^s 7.02 The key answer lists the four of the five entry conditions from > a EDP 2532 Loss of Primary Coolant. Increasing containment pressure was not included. Additional symptcms exist beyond those listed as EDP entry conditions. These include high normal containment sunp level and high radiation in the auxiliary building.
c l The key answer should accept the additional answers of: . ! o increasing containment pressure o increasing normal contairnent sung level o high radiation in the auxiliary building i Reference: Sinulator Briefing Guide RO 2-85-1-29A(B) EOP 2532 Rev 2, Page 2 O .
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7.06 The question asks what are the maximum AININISTRATIVE limits for quarterly and annual neutron exposure. SHP 4902, External s Radiation Exposure Control and Dosimetry Issue, states that initial administrative limits are 250 mrem neutron / calender quarter and 450 mrem neutron / calender year. These exposures can be upgraded to a limit of 300 mrem neutron / calender quarter and 500 mrem neutron / calender year with the approval of the Health Physics ~ Supervisor / Designee. The annual exposure limit , may be exceeded with the approval of the HP Supervisor / Designee, Station Superintendent / Designee and the NUSCo RAB Manager / Designee.
The key answer should accept that there is no explicit maxinum limit on annual neutron exposure.
Reference: SHP 4902 Rev 8 Section 8.1.2.11 . ' O-31- - i ! ... . _ - ._ - -_ - _ .. -. - - - - . . - -
_7.09 The guidance presented in the EOPs provides the use of PORVs as J the contingency action to the use of aux. spray. This is due to the need to provide guidance for positive action that will provide the desired results during an emergency condition, such as on S/G Tbbe Rupture. The additional techniques of pressurizer fill and drain and ambient cooling are desirable techniques for depressurization without aux. sprrq when there is no urgency to cooldown. They are desirable over PORV use only when there is no urgency to cooldown. As the question does not-state the reason for the natural circulation cooldown, the order of preference should not be a significant portion of the point value.
Reference: AOP 2553 Rev 1 Step 4.6 EOP 2534 Rev 2 Step 3.9 Steam Generator Ibbe Rupture Additionally, the procedure would be referenced to determining m the actions to be taken if aux. spray was lost during a natural circulation cooldown.
The answer key should be changed to delete priority frm the point valves.
Consideration should also be given to lowering the point value of the question.
i l [ O-32-t . .. - - . . _ -__._ - _ _ - - - _. - - - -. _ _... _ _ - _ _ _. __ _
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i 1- ! 7.11 Additional correct information is the female visitor limit of
4 125 mrenVquarter.
' i , Reference: SHP 4902 Rev 8 Step 8.2.1.2 1- . ! ! r
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_ _ _ ._- -_ , t G 7.12 The answer information reflects the AOP 2553 guidance only.
The EOP guidance has been revised to reflect the addition of the ICC system. This adds the indications of voiding of reactor vessel level less than 100% and saturation condition at the vessel head.
In addition the revisions clarified the pressurizer level response indications.
, The key answer should also accept the following:
o Pressurizer level increases slower than expected for existing HPSI and charging flow (CO3) 2E.
' o Unheated thermocouples in upper head indicated saturated conditions (ICC display) SE.
o Reactor vessel level less than 100% (ICC display) i > ''*" " " ' ' ' "'*" ' O ~ f I L i O-34- ! l . - -- - .- - . _ - - - . _ _ _ _ -. - . _ _ _ _. . _ _. - _ _
__ . _ _ _ _.
.- - - - -- ._ 7.13a he use of the term "autmatic boration" out of context of the , steps of the procedure can be misleading as the availability of the Boric Acid Storage Tanks is not clear to the individual answering the question.
The flow path of the gravity feed valves and the Boric Acid pumps are manually controlled by the two steps prior to the note containing the RNST information. Autmatic boration is a term typically used to describe the operation of the blending conponents of the CVCS which can function automatically to maintain NCr level and adjust RCS boron concentration during , normal control room operations.
With this under consideration, the answer should also accept { " Boric Acid Storage Tanks."
i.
Reference: AOP 2551 Rev 3 Section 8.18 & 4.19 l
1
O > -35- -- . - - . -. . . - - .. _.. .. - _ .. - .. -... _ _ _ _ - - -. - _
, - - ~. -= -- -.. - - _____ --. ._-. - - .. _ - - _ _ - - - . _ - - _ 8.03b' Capliance with the IRR ICO (T.S.3.2.1) can be verified by O' using either the incore or encore monitoring systems. When the excore monitoring system is used, ASI is monitored by referring to one of two Figures, 3.2-2a g 3.2-2b.
(Refer to T.S.
.! 3.2.2).
With ASI = 0.0, the use of figure 3.2-2a results in an allowed power of 100%; 3.2-2b results in a lower value.
The key answer should accept for full credit either "100%" g , "in accordance with the appropriate figure."
Reference: T.S. 3.2.1 l T.S. 3.2.2
i O
i
! , l i O . -36- ' . .. . - .. . . . -. _ _ _ _ _, _ _ _ _ _ _.. _
. - _ _. . - . - - .. . -. _. - -_ - _ _. ._ _ 8.04 The words " Operations Center" should not be required for full credit.
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...._ - - -...-._. _ __ - -- -- ___...--_--. =- - _ _ . -.. - _ . _..
l 8.09 The question asks to define an intent change and give three (3) ! exmples. The exmples given in the answer key are fran those listed af ter the definition in ACP-CA-3.02 and are general in ,
nature.
Reference: ACP 3.02 Rev 38 Section 4.8
!, Specific exmples should be acceptable answers if they fall within the general exmples.
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j 8.10 ACP-OA-3.02 also states the additional correct information that
at least one of the individuals shall be the on duty Shift {
Supervisor.
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i t i Reference: ACP-QA-3.02 Rev 38 Section 6.8.1.3 ! [ } l
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8.11 EPIP Form 4701-4 identifies many non-emergency incidents that . f - are in addition to those listed by the key answer. These
j should be included as acceptable answers.
!- t l ! f Reference: EPIP Form 4701-4 Pages 1-11 i . l !
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- ... _ _ _. _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.. _ _. _. _ _ _ _ _ _ _ _. _ _ _ _ __ _ i i ! l 8.13a 'Ihe key answer is " station / unit / station services .
superintendent." When multiple individuals are separated by a } virgule, only one is required. Full credit should be given for l
any one of the superintendents.
1
Reference: ACP 6.03, Section 6.2.3 -
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- 8.13b The asstaned dilution flow is 1000 f t /sec, not 1000 ft/ min, according to the letter attached to ACP 6.03 Rev 5.
This specific information is not expected of Senior licensed operators as they typically deal with more unit specific paraneters and values. The 1000 f t /sec flowrate 'is approximately equivalent to 2 circulating water punps per unit, the floret associated with each varies frm unit to unit, making its calculation unlikely.
It is reccmnended that the question be deleted.
Reference: ACP 6.03 Rev 5 Attachment A.
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ATTACHMENT 4 OPERATOR LICENSING EXAMINATION MILLSTONE 2 JULY 7-11, 1986 NRC RESOLUTIONS TO FACILITY COMMENTS ON WRITTEN EXAM Question Resolution 1.02 Accepted in accordance with references provided.
1.09.b.1 Not accepted. References provided do not support the conclusion that RCS voiding by itself, will cause an increase in pressurizer level. Original NRC reference to part b (pg. 45) indicates voiding may cause pressurizer level increases only during aux spray, which was not assumed in the question.
1.09.b.2 Accepted in accordance with references provided.
2.06b Will accept < = 20 sec in accordance with T/S 4.8.1.1.2.a.2.
2.07c Accepted.
In accordance with OP2343 Rev. 7 Caution 7.23.
2.09 Will consider during grading.
2.10 Will consider during) reding.
2.11b Accepted in accordance with OP2310 Rev. 8 Section 7.6.8.
3.03 Accepted. Based on reference which was not provided prior to the exam.
3.04 Will consider during grading.' 3.05 Will consider during grading.
3.06 Accepted in accordance with ESAS System Description page 4.
3.10 Accepted in accordance with AFW System Description pp. 5 & 6.
3.11c Accepted in accordance with references provided.
4.03 Will accept any answer which 1) Activates SIAS and 2) checks flowpaths.
4.06e Accepted in accordance with references provided.
4.08 Will consider during grading.
4.09 Will consider during gradin ~ - - _ _ . -
Question Resolution 4.11a Comment correctly describes the inte.1ded meaning of parenthesis in the answer key, except for equations.
5.04b Comment accepted. Answer key changed.
T = (beta - rho)/lamda* rho) = (0.005-0.00096)/(0.08x0.00096) = 52.6 sec SUR = 26/55.2 = 0.495 DPM 5.08 Comment accepted.
Values removed from the answer key.
Delayed Prompt Average 5.09(2) Comment accepted. Answer key changed.
and (3) 2.
A change in T H as power decreases is greater than the change in T c.
With a -MTC, less negative reactivity is inserted in the top of the core than the bottom due to positive reactivity feedback. Also, the MTC is more negative at the temperatures at the top of the core.
3.
As SG steam flow is decreased, T c increases. The hotter T c entering the core reduces reactor power. At the lower reactor power there is a smaller delta T across the core.
The net result is that T c increases more than T H.
But MTC is more negative at the top of the core than at the bottom so the effects are approximately off-setting and ASI does not significantly change.
5.12 Comment accepted. Answer key changed.
c.
A power increase due to a positive MTC.
5.14 Comment not accepted. No change to answer key.
The logical assumption is that feedwater will enter the feedring.
6.08 Comment accepted. Answer key changed.
' 2.
CWP CEA withdrawal prohibit 3.
CMI CEA motion prohibit _ - -- . - -. - - - _.. --.
__ _ -.. _.. . __ ___ _ - _ - _. _ - - - - _ -. - __ _ _ - i
6.09a Comment not accepted. No change to answer key, l Additional correct information is not required.
, 6.12a and b Comment accepted. Answer key changed, a.
AEAS MSI (+0.5) each b.
SIAS CIAS
EBAS CSAS SRAS ' ! (+0.3) each ~ 7.02 Comment accepted. Answer key changed.
5.
increasing containment pressure 6.
increasing normal containment sump level i ' 7.
high radiation in the auxiliary building 7.04 Deleted since question was duplicate of question 7.10.
7.05 Comment accepted. Answer key changed.
1.
Reactivity control 2.
Vital auxiliary , i 3.
RCS inventory control 4.
RCS pressure control [ 5.
RCS heat removal t l 6.
Containment integrity . 7.06 Comment not accepted. No change to answer key.
The answer in key was given in the Millstone Health Physics
Manual as the maximum exposure.
7.09 Comment accepted.
Exam and answer key changed.
Question - Assume during a natural circulation cooldown . auxiliary spray becomes inoperable. GIVE three (3) alternate methods of depressurizing the RCS.
, - -,,.--r,---- _ - - - - - - - -, -,,,,, - - - - -... - --- ,-e ,, - -. - - - - - - - - - - - - - -. - - - - - -. - -, - -,,,,,, , -
Answer - Fill and drain the pressurizer to cooldown and thereby depressurize the RCS.
Depressurize the pressurizer by ambient cooling.
Open a PORV as needed to reduce RCS pressure.
7.11 Comment not accepted. No change to answer key.
Additional correct inforaation is not required.
7.12 Comment accepted. Answer key changed.
3.
Pressurizer level increases slower than expected for existing HPSI and charging flow (C03).
, 4.
Unheated thermocouples in upper head indicated saturated conditions (ICC display).
5.
Reactor vessel level indicates less than 100% (ICC display) 7.13a Comment not accepted. No change to answer key.
' Millstone 2 A0P 2551, Rev. 2, Section 4.20H cautions the operator that " makeup to the RCS must be from the RWST".
8.03b Comment accepted. Answer key changed, . b.
100% (=0.5) or in accordance with Fig. 3.2-2a or 3.2-2b, Tech. Spec. 3.2.2 i
8.04 Comment accepted. Answer key changed.
1.
The NRC shall be notified by telephone.
l 8.09 Comment accepted. Answer key changed.
l Accept specific answer if they fall within the above general examples.
8.10 Comment not accepted. No change to answer key.
Additional correct information is not required.
i i l
-
8.11 Comment accepted. Answer key changed.
Full credit given for non-emergency incidents listed on EPIP 4701-4, Rev. 4 where " State of Ct. Incident Class" is " echo *".
- Unless reported as higher classification (i.e. ALPHA, BRAVO, CHARLIE-ONE, CHARLIE-TWO, DELTA-ONE, DELTA-TWO)
8.13a Comment not accepted. No change to answer key.
Answer key already applied this reasoning for the grading.
8.13b Deleted since question was misleading. }}