IR 05000029/1986007
| ML20203M521 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 08/14/1986 |
| From: | Dudley N, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20203M519 | List: |
| References | |
| 50-029-86-07OL, 50-29-86-7OL, NUDOCS 8609040034 | |
| Download: ML20203M521 (100) | |
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l U. S. NUCLEAR REGULATORY COMMISSION REGION I
OPERATOR LICENSING EXAMINATION REPORT l
EXAMINATION REPORT NO. 86-07(0L)
FACILITY DOCKET NO.50-029 FACILITY LICENSE NO. DPR-3 LICENSEE: Yankee Atomi'c Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 I
FACILITY: Yankee Nuclear Power Station EXAMINATION DATES: May 21 to 22, 1986 CHIEF EXAMINER:
b 7-24[
y Noel F. Du
, Lead eactor Engineer Date REVIEWED BY:
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_7[/0/66 Robert M. Kellfr, Chief
~ Date Projec s Section IC APPROVED BY:
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gb M
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Harry B. KQter, Cn ' f
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'Date Projects Branch No. 1 SUMMARY: Two Senior Reactor Operator (SRO) candidates and four Reactor Operator (RO) candidates were administered licensing examinations. One R0 and two SR0 licenses were issued. One R0 failed hiv oral examination due to poor performance on the audit of the startup certification program.
The System Training Manuals were found to be incomplete, inaccurate, and inconsistent with plant operating procedures.
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8609040034 860826 PDR ADOCK 05000029 V
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REPORT DETAILS
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TYPE OF EXAM:
Replacement
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EXAM RESULTS:
J R0 SR0 Pass / Fail Pass / Fail Written Exam 2/2 2/0 Qral Exam 0/1 0/0 Overall 1/3
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2/0 CHIEF EXAMINER AT SITE:
N. Dudley, Lead Reactor Engineer 1.
Deficiencies noted during operating examination:
In accordance with Operator Licensing Examiner Standards, NUREG 1021, Chapter ES-303, E.1.b, an audit of the startup certification program was I
conducted during the operating examination.
The results of the audit indicated deficiencies in the understanding of the plant specific startup procedure. The candidcte freely admitted that he did not know the answers.
to specific questions since he had never been present for a startup.
Since the candidate was knowledgeable in most other aspects of plant operations, it is felt that inadequate training has been conducted on the plant specific startup procedure.
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Deficiencies noted during the examination review and grading of the written examination.
The Systems Training Manual (STM), Chapter 3, page 93, describes the use of a safety injection pump to feed the steam generator through the emergency feedline spool piece. This lineup is not included as an option in the Loss of Emergency Feed Procedure OP-3203.
STM, Chaper 15, page 13, also discusses the use of the safety injection pump to feed the steam generator using the feedwater line even though the lineup is not in accordance with procedures.
Reference is made to Figure 3-36 to show the flow path.
Figure 3-36 shows the steam generator narrow range level transmitter temperature equalizing column.
Reference is also made to a flow diagram in section 2.
Section 2 has not been issued.
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- STM, Chapter 26, pages 6 and 7, discusses in detail the design and opera-tion of a charcoal filter for the post-accident H2 removal system. The filter has been removed from the facility.
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Deficiencies noted in training material.
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The reference material for examination preparation was sent to the Region
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on March 17, 1986. On March 20, 1986, the licensee contacted the region to arrange for submittal of updates to the reference material.
It was agreed to freeze the reference material so that the candidates would not be forced to keep up to date on plant and procedural changes on a daily basis.
It was not realized at the time that the System Training Manual (STM) had not been updated to include the maior modifications made to the plant during the last outage. As a result, the STM Chapter 18, page 3 discussion of the three-way valve (VD-TV-203) to the ventilation stack from the ventilation header of primary components is incorrect since the three-way valve has been comp'letely removed. Also, the STM Chapter 15, page 21 discussion of the narrow range pressurizer level channel density compensation is incorrect since the level channel has been replaced with solid state components.
The Reactor Operator Training Manual was sent to provide information for 12 STM chapters which have not been written. The STM chapters will be
written to correct and improve the information contained in the Reactor Operator Training Manual.
Three of the unwritten STM chapters were not covered by information in the Reactor Operator Training Manual.
These chapters dealt with the reactor vessel and internals, incore instrumen-tation, and emergency diesels.
Facility training documents which were not originally provided to the NRC for examination preparation included:
Core XVII-XVIII Refueling; Pre-Startup Training, Nov-Dec 85 Yankee Nuclear Power Station Safety Analysis Assumptions
Mitigation of Core Damage
Electrical Operating Instructions INPO Approved Learning Objectives for Systems and Procedures AP-7104 Core Operational Limits i
The Core XVII-XVIII Refueling; Pre-Startup Training Document and the Yankee Nuclear Power Station Safety Analysis Assumptions were requested and provided after the initial submittal.
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Learning objectives were provided; however, two different numbering systems vere used.
It appears both INP0 approved learning objectives and older facility learning objectives were provided in the area of thermo-dynamics and reactor theory. Only the older facility objectives
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for systems and procedures appear to have been provided.
Examiners had difficulty referencing examination questions to learning objectives due to the organization and generality of the learning objectives.
As a result of the examination process, the reference material supplied by the licensee has been found to be inadequate, incomplete, inaccurate, and inconsistent.
Numerous documents and manuals, which contained information directly related to plant operations, were not provided by the licensee.
The STM has only 29 of the planned 42 chapters written.
Information is unavailable for the material covered by three of the planned unwritten chapters. The STM has not been kept up-to-date with plant modifications. The STM details plant lineups which are not allowed by plant procedures. As a result, as of March 17, 1986, the candidates did not have a complete accurate set of reference' material that reflected actual plant systems and operations.
Candidates should have a complete and accurate set of reference material from which to study prior to any future licensing examination.
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Personnel Present at Exit Interview NRC Personnel N. Dudley, Lead' Reactor Engineer H. Eichenholz, Senior Resident Inspector Facility Personnel L. Heider, Vice President, Manager of Operations D. Vassar, Plant Operations Manager N. St. Laurent, Plant Superintendent B. Drawbridge, Assistant Plant Superintendent E. Chatfield, Training Manager L. Laffond, Senior Instructor 5.
Summary of NRC Comments at exit interview:
The examiner reviewed the number and type of examinations administered during the week and stated that results would be issued in approximately 30 days.
The examiner stated that difficulty had been encountered in examination preparation due to an inaccurate System Training Manual and the lack of information provided on the initial reference material submittal.
The Training Manager stated that numerous revisions had been prepared for the System Training Manual since March; however, the revisions had not been provided to the NRC due to the agreement to freeze the materia.
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Changes made to written examinations:
Facility comments were considered during the grading of the examination; however, not all comments resulted in changes to the examination answer key. Many comments did not change the questions on this examination, but
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will be considered when writing future examinations.
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Answer Number Change Justification Question 1.07 a DELETE "and assume no Steam driven auxiliary l
operator action."
feedpump must be started manually.
Question 1.08 b DELETE.
The initial condition that rods are in auto-matic was not stated.
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Question 2.02 b DELETE.
Charcoal filter has been removed from the facility.
t 1.06 a Add "[0.3] for correct Question asked for order."
relative magnitudes of coefficients.
1.07 b Add "(Alternate answers Alternate answers are accepted if appropriate possible depending on assumptions were how natural circulation
stated)."
is interrupted.
1.08 a Answer reworded.
Provided clarity.
1.09 Values corrected.
Nomographs had been
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misread.
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2.11 a Add "or #1 + 3 variable Charging pumps have speed charging pumps sufficient capacity to maintain level."
make up for leak.
- 2.11 b + c Points increased to Corrected point values.
[1.0].
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3.'10 a Points increased to Corrected point value.
[1.0].
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l 4.05 Add "PZR heaters Includes additional auto-
energ'ize."
matic a6 tion.
4.06 Add that T is stable Clarifies the expected h
"or decreasing."
response of T n natural h
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circulation.
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1 Answer Number Change Justificatien 4.10 Increased points to Co'rrected point value.
[0.75 each].
Question 5.02 a DELETE.
Reactor not allowed to
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be operated in this condition by Technical
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Specifications.
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Question 5.05 b (1)
Changes " Rods drive is Clarifies amount of rod at" to " Rods drive in."
movement.
-Question 6.03-b + c
~0ELETE.
The reference material
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is inconsistent which made grading of question impossible.
Reference 5.03 Change to " Yankee WEC -
Corrects reference.
Fundamentals of Nuclear Reactor Physics "
5.07 b Delete "and energy" and Clarifies answer.
add "and overcomes the increasing energy loss from the reference leg
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i-break.
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5.08 Change "642.8" to "636" Provides correct value and correct math.
for saturation temperature at 2000 psia.
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l 6.02 Delete "First," cnange Allows credit for set
" valve" to " valves" and point for eithe'r safety add "2560 or."
valve.
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7.01 c Change "all operating Not all MCPS would be-MCPS" to "the operating tripped.
MCPS."
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Reference 7.02 a Change "pg 1" to "pg 2."
Corrects reference.
r Reference 7.07 a Change "pg 3" to "pg 13."
Corretts reference.
Reference 7.07 b Change "pg 13" to "pg 3."
. Corrects reference.
Reference 7.08 a Change "pg 1" to "pg 2."
Corrects reference.
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Answer Number Change Justification 8.10 b Change "2" to "4."
Choice 4 is correct for instrumentation required by Technical Specifications.
Reference 8.10 c Add "co Page 3/4-3-23."
Corrects reference.
Attachments:
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Written Examination and Answer Key R0 2.
Written Examination and Answer Key SRO 3.
Facility Comments on Written Examinations
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U. S. NUCLEAR REGULATORY COMMISSION
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REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
YANKEE ROWE
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REACTOR TYPE:
PWR-WEC4 DATE ADMINISTERED: 86/05/20 EXAMINER:
DUDLEY. N.
APPLICANT:
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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
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question are indicated in parentheses after the question. The passing i
grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE_
CATEGORY 25.00 25.00 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2.
PLANT DESIGN INCLUDING SAFETY
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AND EMERGENCY SYSTEMS 25.00 25.00
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INSTRUMENTS AND CONTROLS 25.00 25.00 4.
PROCEDURES - NORMAL, ABNORMAL,
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EMERGENCY AND RADIGIAGICAL
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100.00 100.00 TOTALS nu.
FINAL GRADE
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All work done on this examination
.*.s my own. I have neither given nor received aid.
APPLICANT'S SIGNATURE
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMTCS. HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.01 (.80)
Select the most correct statement from the following.
A. If two centrifical pumps are in parallel then the combined pump-head will be approximately the sum of the individual pump heads.
B. If two centrifical pumps are in series then the combined power requirements will be approximately equal to the cube of the individual pump power.
C. If two centrifical pumps are in parallel then the combined flow will be approximately equal to the sum of the individual pump flows.
D. If two centrifical pumps are in series the flow of each pump will be approximately equal to the square of the individual pump speed.
QUESTION 1.02 (2.40)
a. Why does nucleate boiling heat transfer remove more heat than non-boiling heat transfer?
b. Why does film boiling remove heat transfer res'ove less heat than non-boiling heat transfer?
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QUESTION 1.03 (2.40)
With all systems in manual and no operator action, what effect will in-
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creasing the circulating water flow through the condenser tubes have on pressurizer level? Justify your answer assuming charging pump speed control is in manual operation and no operator action.
QUESTION 1.04 (2.50)
In Figure 4, explain why EOL differential worth has a peak at about a.
80 inches withdrawn and the BOL differential worth does not.
(1.0)
b.
What is the effect of a dropped rod on reactor power and Tave.
Explain. Assume all systems in manual and no reactor trip.
(1.5)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.05 (2.50)
Describe a xenon oscillation by explaining how it starts and why it oscillates.
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QUESTION 1.06 (2.50)
a. The overall power coefficient is a combination of three coefficients.
List those coefficient in order from the largest to the smallest contributor to the overall power coefficient.
(1.5)
b. Why does the Total Power Defect increase rapidly above 510 MWT7 (1.0)
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QUESTION 1.07 (3.00)
a.
While at 100% power, assume a loss of station power with a concurrent diesel failure results in a total loss of all station power except for batteries. Starting when power is lost, explain how the core will be cooled. Include the final heat sink, f ::
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(1.5)
b.
How will the following parameters be trending if the core is not being cooled following a loss of power. Briefly explain.
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Delta-T across the core (0.5)
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Steam generator level (0.5)
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Steam generater pressure (0.5)
QUESTION 1.08 (3.00)
The reactor is at 85% power with rods in manual and Tave is 515 F.
A steam leak equivalent to 10% load occurs on a steam line.
a.
From a reactivity standpoint, explain how and why reactor power changes.
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.09 (2.90)
Power is to be reduc.ed from 100% to 30%. Boron concentration is 100 ppm and rods are fully withdrawn. What will be the final boron concentration Use at equilibrium conditions if Bank C will be inserted 45 inches.
attached figures, state all assumptions and show all work.
QUESTION 1.10 (3.00)
Under what conditions could the Moderator Temperature Coefficient a.
be positive? Explain.
Will differential boron worth increase, decrease or remain the same b.
for the following conditions.
1.
Increased boron concentration 2.
Increased core life 3.
Increased Tave
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS i*
QUESTION 2.01 (1.50)
Explain why heat exchangers are not required for vapor container depressurization or long term core cooling following a loss of coolant accident.
QUESTION 2.02 (1.50)
The following questions refer to the Post Accident H2 Systen:
a. Explain the operation of sampling valve HV-SOV-1, during a large break loss of coolant accident?
b. "h:t 2: tire ch:213 he * '-- if t'- hi;h t- ;;reture el r fos-the ch:: r:1 filt: i: ::::i :d?
QUESTION 2.03 (2.00)
a. What is the power supply to each of the four main coolant pumps?
b. Why is a main coolant pump prevented from starting unless the cold leg stop valve is closed?
QUESTION 2.04 (2.50)
For each of the following systems indicate to which coolant leg they are connected. Also include the loop or loops in which the connections are made.
a. Pressurizar surge system b. Pressurizar spray system c. Safety injection system f
d. Charging system e. Bleed system
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l QUESTION 2.05 (2.00)
a. Describe what actions, if any, are required to restore Component Cooling Water flow to the Main Coolant Pumps if trip valves CC-TV-205 and CC-TV-208 close due to loss of instrument air.
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b. Describe how emergency cooling water can be supplied to the Main Coolant Pumps on a complete loss of Component Cooling.
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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QUESTION 2.06 (2.50)
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If all three emergency boiler feed pumps become inoperable after a loss j
of feedwater accident:
a. h t additional pumps can be used to feed the generators?
(1.5)
(1.0)
b. h t flow paths must be established?
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l QUESTION 2.07 (2.50)
Trace the two paths by which power is distributed from.the 115 KV bus to 120 vac #1 vital bus. Include any voltage changes and list all buses and MCCs in the flowpath.
QUESTION 2.08 (2.00)
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i a. h t function does the bubble in the pressurizer perform during (0.6)
normal plant operations?
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b. h t function does the bubble in the pressurizer perform during (0.5)
design transients?
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c. ht would happen on a rapid insurge if there was no surge (0.8)
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deflector?
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QUESTION 2.09 (2.50)
a. h t type of accident requires a Non Return Valve (NRV)
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closure.
c b. Explain how a NRV is closed once an automatic closure signal is
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(1.5)
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(":52" UATEOMY U2 GCLDialTli) ON RER PAGE N)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
QUESTION 2.10 (3.00)
a. Under what accident is the accumulntor designed to deliver water to the primary?
(1.0)
b. Describe how an accumulator delivers water to the primary upon receipt of a Safety Injection Actuation Signal. Include any automatic actions which will terminate the delivery of water.
(2.0)
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i QUESTION 2.11 (3.00)
n. How will the charging system respond to maintain pressurizer level if ble6d flow is 20 gpm, main coolant leakage is 3 gpm s.nd no operator action is taken?
b. hw is radioactive water prevented from leaking to the atmosphere from the seals of the purification cooling and drain pump?
c. How is the temperature of the water passing to the ion exchangers controlled?
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INSTRUMENTS AND CONTROM PAGE
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l QUESTION 3.01 (1.50)
c. How would the plant respond to an uncontrolled dilution if the rod control system was in automatic and no operator action was taken?
b. Explain how rod motion is affected when the single step rod circuit is activated. Include what causes the single step rod circuit to be activated.
QUESTION 3.02 (2.50)
What effect, if any, will a loss of power to the Pressurizer Narrow Range Pressure Channel have on:
a. High and low pressurizar pressure alarms b. Prossurizer spray (PR-MOV-191) valve operation c. Backup heater ' operation d. Power operated relief valve operation e. Charging pump operation QUESTION 3.03 (2.00)
Indicate what automatic actions, if any, will oc.:ur for the following Radiation Monitoring System alams.
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a. High level on the particulate monitor on the primary vent stack during venting of the Vapor Container.
b. High level alarm on the steam generator blowdown tank effluent
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monitor.
c. High level alarm on the liquid radwaste affluent monitor during
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a radwaste discalurge.
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d. High level on the spent fuel pit area radiation monitor with the ventilation system in normal operations.
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INSTRUMENTS AND CONTROLS PAGE
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I QUESTION 3.04 (2.00)
For each of the following situations indicate:
1. Whether the rod operation selector switch should be positioned to Auto, Manual, or All IN; AND 2. Whether the rod group selector switch should be positioned to A, B, C, D, Off, or Spare.
a. Normal power plant operations at 100% power.
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b. Recovery of a dropped rod in group B while at 70% power.
c. Rod withdrawal at the start of a normal reactor startup.
QUESTION 3.05 (2.50)
What are five (5) of the six plant parameters which can be monitored from the Safe Shutdown System (SSS) panel in the SSS Building?.
QUESTION 3.06 (2.50)
Explain how the Feedwater Regulating System will respond, in automatic, to a failure of the steam pressure compensating signal if the signal fails high. Include a brief two or three sentence description of the signal processing. Indicate whether the final steam generator water level
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will be higher than, lower than, or the same as the initial water level.
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l QUESTION 3.07 (2.50)
Will the plant scram in respense to a failure of a NR pressuriser level instrument low, if the plant is at 80% power steady state and no cperator action is taken? Include all automatic equipment responses cnd follow the transient to the point where the plant is stable cr a scram has occurred.
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INSTRUMENTS AND CONTROLS PAGE 10
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QUESTION 3.08 (3.00)
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Indicate whether or not the following coincident instrument failures will result in a reactor scram. Consider each set of failures ccparately and justify your answer.
a. One intermediate range power NI and one power range NI both fail high when both coincidence switches for the scram logic are set to the coincidence position.
b. The narrow range level channel on steam generators #1 and #2 both fail low.
c. Main coolant loop 1 pressure channel fails high and main coolant i
loop 2 pressure channel fails low.
d. The over current relay on phase A for the main coolant pump in loop 1 fails high and the under current relay on phase B
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for the main coolant-pump in loop 2 fails low.
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QUESTION 3.09 (3.50)
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Indicate the immediate (within 10 sec.) response of the following components to a Safety Injection Signal. Assume normal electrical ptwer is available.
a. Low Pressure Safety Injection Pumps recirculation valve (CS-MOV-532)
b. Low Pressure Safety Injection Pumps
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c. Service Water Pumps d. CCW supply valve to the Vapor Container (TV-208)
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e. Nitrogen Pressurization valva (SOV-46) to the Accumulator f. Breaker BT 1A if the diesel generator is loaded on bus for load test
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3. Makeup valve for the charging system (CH-MOV-523)
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QUESTION. 3.10 (3.00)
c. A continuous rod withdrawal accident occurs from low in the source
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range. Will an undercompensated or overcompensated IR detector allow c "M*W a greater peak power to be reached? Briefly explain why?
(1.0)
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b. Using figure 6-1 trovided, explain the operation of the relays and contacts necessary to block the at power reactor scrams. Assume initial turbine load is 20 MWe and decreasing.
(2.0)
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PROCEDURES - NORMAL ABNORMAL EMERGENCY AND PAGE 11 RADIOLOGICAL CONTROL
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QUESTION 4.01 (1.50)
o. What is the Yankee ADMINISTRATIVE CONTROL limit for WEEKLY Whole Body Exposure?
b. What is the Yankee ADMINISTRATIVE CONTROL limit for QUARTERLY Whole Body Expo.=ure?
c. Whose approval must be obtained prior to exceeding the Yankee exposure limits?
QUESTION 4.02 (1.50)
Explain why an operator should trim back to two Safety Injection trains when the Safety Injection Tank level reaches 19 feet.
QUESTION 4.03 (2.50)
What steps must be taken if a reactor scram cannot be verified after cn emergency shutdown from power?
QUESTION 4.04 (2.00)
a. What action should be taken if after an automatic Safety Injection the main coolant pressure is steady at 1550 psig, the high pressure safety injection (HPSI) header pressure is steady at 1550 psig with the HPSI pumps running, and safety injection flow is sero?
b. Explain why feedwater should not be supplied to a steam generator with a pressure of below 300 psig after a Safety Injection Actuation has occurred.
QUESTION 4.05 (2.00)
What automatic actions will occur on a loss of #1 120 VAC Vital Bus?
QUESTION 4.06 (2.50)
What are the FOUR indications used to verify that natural circulation has been established after a loss of offsite power?
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL QUESTION 4.07 (2.00)
Pricr to exceeding 300 F during a plant startup a dedicated operstor 10 requirtd to be stationed at the main control board. What is the colo responsibility of the operator and how does he perform this rocponsibility?
QUESTION 4.08 (2.00)
>
,
What should be the position or cor.dition of the following components cr parameters just prior to initiating shutdown cooling? Assume main
'
coolant temperature is between 300 F and 330 F, and main coolant procsure is 280 psig.
d c. Steam generator levels b. The small hogger, Nash vacuum priming pump, and air ejector c. The six ECCS pump-motor breakers d. The accumulator fill valve (SI-V-54)
o. The safety injection actuation channel switches QUESTION 4.09 (3.00)
a. What are five syptoms of a steam generator tube rupture besides decreasing pressurizer level and pressure?
b. List THREE of the five objectives of a Steam Generator Tube Rupture, cs presented in OP-3107.
QUESTION 4.10 (3.00)
What actions, if any, should be taken if the following indications are procent?
.
Reactor Power is 90%.
The 1500 spe low flow alarm for component cooling is activated.
er *
%-
"#
Component cooling water flow indication is zero.
.
The main coolant pump bearing temperatures are 140 F and increasing.
The 75 spm low component cooling water flow to the main coolant pumps is actuated.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- -.
.
__
,
-
..
I
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13
.
RADIOLOGICAL CONTROL QUESTION 4.11 (3.00)
A suming initial reactor power is 100%, indicate for each of the cituations listed below whether to:
1. Maintain the plant at power 2. Reduce power to a lower level 3. Conduct a shutdown 4. Manually scram the reactor a. A call is received in the main control room stating that a bomb has been place in the plant and will explode in two hours.
.
b. The service water low pressure auto start alarm is present and the service water header pressure is 40 psig and cannot be increased.
c. A fuel element failure exists and the coolant activity level cannot be reduced below Technical Specification limits using the purification system.
d. Neutron channel indications become erratic due to low level in the Neutron Shield Tank.
e. It becomes impossible to use the normal charging line to the coolant loops.
f. A rod drop occurs without causing a scram.
.
.
,.
.
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.
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(***** END OF CATEGORY 04 *****)
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FIGURE S STEAM PRESSURE AT STEAM GENERATOR OUTLET VS. NSSS POWER
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-
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Water Parameters
. Miscellaneous Convers' ions # *
vege,-
.
1 gal. = 8.345 lbm.
I cur'ie = 3.7 x 1010dps
.
I gal. = 3.78 liters 1 kg = 2.21 lbm
1 fte = 7.48 gal.
I hp = 2.54 x 10 Btu /hr Density = 62.4 lbm/ft3 1 m = 3.41 x 106 Stu/hr
-
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'F = 9/5'C + 32 Heat of fusion = 144 Btu /lbm
- C = 5/9 ('F-32)
1 Atm = 14.7 psi =.29.9 in. Hg.
.
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 14
'
,
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
-
ANSWERS -- YANKEE R0WE-86/05/20-DUDLEY, N.
ANSWER 1.01 (.80)
<
C. [0.8)
REFERENCE Thermal Hydraulic II pgs. 10-32 to 10-48
___.......__....________.............___________....... __.......____......
kppendixpgA-9 2.9
,
ANSWER 1.02 (2.40)
o. ' Nucleate boiling creates turbulent flow which promotes more mixing (0.6). The coolant picks up latent heat of vaporization and carries it to cooler parts of the channel (0.6).
b.
In film boiling, a film of steam costs the clad surface and forms an insulating layer (0.6) which greatly reduces the heat transfer coefficient (0.6).
REFERENCE Thermal Hydraulic I pas. 3-72, 3-73, 13-17, 13-20
-
.........___..__.........._______.........__..................._____.......
3.4 003 000 K 5.01 3.3
................_____............___........___.............__.............
OBJ: 200510
.
i i
ANSWER 1.03 (2.40)
I More steam will be drawn from the SG's due to increased condenser vacuum More heat will be extracted from the RCS due to increased steam flow
,
!
or feedflow.
RCS moderator becomes more dense due to reduced temperature
,
,
j Pressuriser level will decrease (0.6 each)
,
REFERENCE Thermal Hydraulic II pg. 9-10
...............__......__..__..............................................
I 3.5 039 000 A 1.05 3.2
........__.........__...............__.......___...__.....................
OBJ: 200514
'
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.
i
.
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE 15
-
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- YANZEE ROWE-86/05/20-DUDLEY, N.
ANSWER 1.04 (2.50)
Neutron flux is shif ted to the top of the core at EOL due to fuel a.
depletion at the bottom of the core (0.5) and since rod worth is proportional to flux, rod worth will increase in the top of the core. (0.5)
b.
Reactor power will decrease and return to equilibrium power. (0.5)
Tave will decrease. (0.5)
The negative reactivity inserted by the dropped rod would be countered by positive reactivity inserted by MTC. (0.5)
REFERENCE Reactor Core Control pas. 6-14, 8-20, 3-3
.
....__ __....__....___..___....___.___....._______...___.._________.._____.
3.1 001 000 K 5.05 3.5
'
K 5.10 3.9
...__.........____....__. _____________.._______......________.......___..
OBJ: 200916 ANSWER 1.05 (2.50)
Inward rod motion followed by rod withdrawal [0.5] would cause a flux
'
chift to the top of the core will burnout Xe in the top of the core cnd produce I-135 while in the bottom of the core Xe is building in from I-135 decay as I-135 production is less [0.8]. As Xe builds into the top of the core from I-135 and as Xe decays away in the bottom of the core, the flux will shift towards the bottom of the core [0.7]. After Xe con-
,
centration changes in the top and bottom of the core, the flux will shift
-
cgain [0.5].
,
REFERENCE Reactor Core Control pas. 4-28, 4-29
,
...__...... _________....________...__....... ___...__..._________.........
3.1 001 000 K 5.38 3.5 co; =
a Ek s
,
-
- -
- - - _
=.
-.
.
-
.. - _
_
.
'
.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 16
.
i THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
!
ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
l l
ANSWER 1.06 (2.50)
c.
Doppler only power coefficient (0.s]
Moderator only power coefficient;, [0.k on1 g r e g ficient [0.9]
Vo
'
b.
Programmed Tave begins to increase at 510 MWT [0.5]
..
l a larger % of boron is removed from the core at higher temperatures
[0.5]
i REFERENCE R: actor Core Control pas. 3-29 to 3-41
.... ______.......______...___.______.........____ ________.... ___________
'
3.1 001 000 K 5.49 3.4
,
.___...__...._____________________.........__________..___......__........
OBJ: 200913 i
,
j ANSWER 1.07 (3.00)
c. Natural circulation flow removes heat to steam generators due to
change in density of water [0.75]
Water in SG changes to steam and is vented to atmosphere. [0.75]
b 1.,.7.ncrease [0.25] Thot increases as Tcold remains constant (0.25]
2. Decrease [0.25] There is no f 'd flow to steam generator [0.25]
.
3. Decrease (0.25] No heat input to maintain temperature (0.25]
{ALTIMWATE AWS WEAS ACsEPTr0 IF APMo Ms Air M$$ung.rcH.$ WC4! $7droe)
,
RErIRENCE i
Thermal Hydrualic II pas. 14-26, 14-27
...............___......____...___............_______............_______...
3.4 000 015 EK 1.01 4.4 3.7 000 055 EK 1.02 4.1
.
..... ______......___..___.....__...__ ___.........._________.........___..
OBJ: 200514
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.
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,
s 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17
.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 1.08 (3.00)
Reactor power will increase (0.5) to match secondary power, N pos-c.
itive reactivity inserted by MTC (0.5) wh4elp will be countered by ritt negative reactivity from power defect. : ;;r : in::r ::: (0,5).
h.
" ^ (T :: - T:t-)
SS%
".^. (515 - 572) - S5/05 " ^ (T::: - 550)
(0.5)
f 510 prir ec-erpe-9e te ':59 F f--
et--- tebic) (0.5)
7-e - Siv v (n si REFERENCE Reactor Core Control pgs. 3-5, 3-9 Thermal Hydrualic II pgs. 12-8, 12-9
_____ ______... _______.._____.. ___...___________...........__________ ___
3.5 039 000 K 5.08 3.6 A 2.05 3.3
_______________........ ________.....___.... ________.......__ _____.....
OBJ: 200911 200506
ANSWER 1.09 (2.90)
100 Reds J 900 - 1200 = -44008 pcm (0.6)
Defect fs09PJ - 225 = MSM0e pcm (0.6)
X3 3tS0* - M00' = t050 pcm (0.6)
JICC - 1910 to 20 400 pcm for boron i i13 Arsume 10 pcm/ ppm: 400 pcm/10pem/ ppm = 40 ppa (0.5)
ll45
//y.5 Final boron concentration: 100 + 40 = MG ppm (0.6)
-
1/1 Rly REFERENCE Reactor Core Control pgs. 9-21 to 9-25 Data Reference Manual, Sec C o
................................___......__.__.____'........ __...........
.
3.1 001 000 K 5.09 3.5
........... __________....__....... ______......_______........_____......
OBJ: 200917 200919 200913 l
_ _
_
..
..
-- -
-
- " - ~'
f
.
,,1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 18
.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 1.10 (3.00)
a.
High boron concentrations [0.3] and high RCS temperature (0.3]
at BOL [0.2]
The expansion of the moderator due to temperature increase would remove boron form the core which would add positive reactivity. [0.7]
b.
1.
Decrease [0.5] (because of more competition for thermal neutrons with other boron atoms.)
2.
Decrease [0.5] (because of more competition for thermal neutrons with fission products.)
3.
Decrease [0.5] (because less boron must be added to increase boron concentration by one ppm)
,
REFERENCE Reactor Core Control pgs. 5-13 to 5-16
..........____.____________________________________..___.____ __.._________
'
3.1 001 010 K 5.29 2.9 K 5.21 3.4
___.....____________________________________________.____.....____...__...
OBJ: 200911
-
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l l
_. _ _ _ _ _ _ _ _ _ _ _ _ _
._
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 19 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
i ANSWER 2.01 (1.50)
Heat removal from the steam and water in the vapor container is
,
cecomplished by conduction through the steel of the vapor container
'
'
ball.
[1.5]
RussaMNCE STH: Chap 19 Emergency Core Cooling, p4
______.__.............______...______._.._____________.....________._.....
KA: p 3.6-13, K40.04, 3.7
_________........._.._____..............._________.. _______.... ___.....__
OBJ:
110302 ANSWER 2.02 (1.50)
e. HV-SOV-1 will be energized by a VC isolation signal which will close valvs.
[0.4]
Key switch can override isolation signal to open valve. [0.35]
5.
Ertr$lich re vic- -ir fler (12 SCFu) +'-r ;' filter. [0.75] s
-
REFERENCE STH: Chap 26 Post Accident H2, p 4, 7
.......... ___..._________________________...........___________.....__
KA:
p3.6-24, A4.01, 4.0
%
...... _______.....___________........______________________...__________
OBJ:
102802
-
ANSWER 2.03 (2.00)
.-
.
c.
PUMP NO.
POWER SUPPLY 4.
l
2400v Bus 3
2400v Bus 1
2400v Bus 1
- "V #@'
.
2400v Bus 2
[0.25 aa]
I
,,,
'
b. To prevent a coldwater accident.
[1.00]
REFERENCE STM: Chap 14 Main Coolant, p 14
__________.........___.. ____......_______....___________......._____...
KA: p 3.4-1, K2.01, 3.3 p 3.2-1, kl.08, 4.5
...._......__.._____________....________...______... _.... __________..___
OBJ:
100301
_
_ _ _ _ _ _ _ _ _ _ _ _ _
-
_
_
.
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 20 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
100236 i
100202 ANSWER 2.04 (2.50)
,
c.
Loop 1 hot leg [0.5]
b.
Loop 2 cold leg [0.5]
c.
All four loops (Loops 1,2,3 & 4)
cold leg [0.5]
d.
Loop 4 hot leg [0.5]
o.
Loop 1 cold leg [0.5]
REFERENCE STM Chap 13 Main Coolant, p 38, Fig. 13-20
............... _____________________________....__________.._______......
KA: p 3.2-1, K1.08, 4.5
................ ___.........__............. ______________............
OBJ:
100233 100234 ANSWER 2.05 (2.00)
a. manually position 3 way emergency valve to lineup b4ckup N2. [1.0)
b. Via a hose connection from the firemain or service water on the inlet header.
[1.0]
REFERENCE
~ '
STM: Chap 20 Component Cooling, p 18, 19
,
OP-3002
'
OP-3115
........ ________._______________...........___.... ____.. ______________
KA:
p 3.2-3, K6.02, 3.6
. ___......._____._.................... ________________________________
OBJ:
100811 100805 I
_
. -.
.
-.
.
.
.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 21 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 2.06 (2.50)
c. Main Coolant System charging pumps [0.75]
Safety Injection System pumps [0.75]
Safe Shutdown System secondary makeup pump b. Charging pump via the emergency feedline spool piece [9 53' O.3]
Charging pumps via the blowdown cross connect (0 2]
Safety injection pumps via the blowdown cross connect.[0.5)
REFERENCE STM: Chap 3 Condensate and Feedwater, p 89, 93 OP-3203 Rev. 13, p2
....__......._______________________________________________......._____.
KA:
3.4-13, K1.01, 4.2
_-_.._............__.......__________._.....___.________....__..__...____
OBJ:
106111 ANSWER 2.07
'(2.50)
Normal 115 KV/ 2400 V transformer [0.1]
2400 V/ 480 V transformer
[0.1]
Bus 5-2
[0.1]
MCC 1 Bus 1
[0.1]
'
Battery Charger 480 vac/ 125 vde
[0.4]
Battey Dist Switch Board # 1
[0.1]
Inverter 125 vdc/ 120 vac
[0.4]
'
,_~~
- 1 Vital Bus Bypass power supply
,_
,
115 Kv/ 2400 V transformer
[0.1]
,
2400 V/ 480 V transformer
[0.1]
~
480 v Bus 6-3
[0.1]
Emerg Bus 1
[0.1]
Emerg MCC 3
[0.1]
Emerg MCC 5
[0.1]
my, 480 v/ 120 vac transformer (0.5]
Inverter Switch
[0.1]
Vital Bus #1 REFERENCE
,
STM: Chap 40 Vital Bus, p 2
.. ______........__.________........._____________________....._....
KA: p 3.7-2, K4.10, 3.1
____..._________________________.____.____..__________________________.
Lesson Plan 10062X
.
-Dv67 6 9 WW e-
.
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
.
l l
ANSWER 2.08 (2.00)
a. Accommodates pressure surges. [0.6]
b. Prevents water discharge throught the safety valves and the solenoid operated reitef valve.,[0.6]
b. Colder water could shoot to surface of the water and start condensing the steam bubble on a rapid insurge.
[0.8]
REFERENCE STM: Chap 14 Pressure Control and Relief, p2
..... ______........________.. ___________________. _______________. _____
KA: p 3.3-3, Generic 4, 3.6
__________..___. _______________________________________......____________
OBJ: 100444 100474
'
101001 ANSWER 2.09 (2.50)
a. Steam Line Break [1.0]
e
b. Hydraulic oil is vented from under poston [0.8)
N2 on top of the piston forces the NRV shut [0.7]
'
REFERENCE STM: Chap 33 Reactor Protection, p 23
. -,
ROTH: Steam Generator and Distribution, p 3-5, 3-7
. t,
.__....._._.......____...._______....___........__......____...._______._
,
KA: p 3.5-2, K4.06, 3.3
______.___...... _______________....____.______. _... _____..________...
'
OBJ:
103925
.x,
?
--
...
__
_ _ _ _ _ -
..
.
.
l I
.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 23 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 2.10 (3.00)
'
c. During a large loss of coolant a'ecident [0.5] coincident with a loss of outside electric power [0.5].
b. SOV's open to admit N2 to accumulator. [0.3]
Accumulator is pressurized with N2 (to about 475 psig) [0.5]
Accumulator will deliver water through check valves when main coolant pressure decreases (to 460 psig). [0.5]
Level switches will close accumulator outlet valve after (700 cubic feet) of water has been injected.
[0.5]
Nitrogen will be vented off. [0.2]
REFERENCE SIN: Chap 19 Emergency' Core Cooling, p3
______________..______...__________......_________...________________...
KA: p 3.2-13, K4.04, 3.0
_________________________..._______......... ___________....._..___.
OBJ:
100631
.
ANSWER 2.11 (3.00)
n. Charging pumps will continue to run until the high level (HL) auto-stop
.is reached. (0.5) Then they will cycle on and off as level drifts above and below the HL setpoint. (0.5) n
- I<3 VnR1AOLE SPEE O Cl/Agir# PUMPS MAIND9M LML b. Double mechanical seals are pressurized with domineralied water. [&:fSt 1.0 c. Temperature is controlled by a TGV on the CCW line to the low
. pressure surge tank cooler. [ed64 fO REFERENCE STM: Chap 15 Charging and Volume Control, p 10 Chap 16 Purification, p 5, 6
_______________________________________..______________________________..
KA:
3.2-21, K4.02, 3.3 eM 3.1-16, K4.03, 2.8
.,N
.. ________________________..________ ________________________________..
OBJ: 100473
.!
l
__
_
.
.
3.
INSTRUMENTS AND CONTROLS PAGE 24
. ANSWERS -- YANKEE ROWE
'-86/05/20-DUDLEY, N.
ANSWER 3.01 (1.50)
c. An uncontrolled dilution would cause a Tave increase Increase in Tave will cause inward rod motion Rod motion begins and stops when Tave decreases to less than setpoint.
b. Above 130 MWe, rods can only be pulled one step at a time. To pull rods again 'he manual rod control switch must be allowed to spring return to c
off and then shifted back to the rods out position. (0.75)
REFERENCE o. Yankee STM, pg 34-11 b. pg 34-16 c. pg 34-16,34-17 ANSWER 3.02 (2.50)
c. Both alarm [0.5]
b. Will not open automatically [0.5]
c. will not turn on of off automatically [0.5)
d. Will not open automatically [0.5]
o. No effect [0.5]
REFERENCE
.-
SIN: Chap 15 Charging and Volume Control, p 18 C:re XVIII Pre-Start-up Training Manual:
EDCR 84-307, p 99
__________________________________________________________________________
K+A: p 3.2-21; 011-K4.03 2.6
'
__________________________________________________________________________
_
OBJ:
101002
..
f
- f.,h" s
l
-,
g
.,-
. rQy w s
.
, - - _ -. - -. _ _ -. _ -
... -. -.
-
. -
.
. _.
.
.
3.
INSTRUMENTS AND CONTROLS PAGE 25 ANSWERS -- YANKEE R0WE-86/03/20-DUDLEY, N.
ANSWER 3.03 (2.00)
c. No automatic action [0.5]
b. Tank discharge valve shuts [0.5]
c. Closure of effluent flow control valve [0.5]
d. No Automatic action [0.5]
REFERENCE FSAR: Chap 215 Radiation Monitoring System, p 215:3 Chap 413 Fuel Handling Accident, p 215:4
_____________........___..______...______________....____________ _______.
KA: p 3.11-1, K1.01, 3.4 p 3.11-14, K6.10, 2.5 p 3.11-27, EA1.02, 3.1 p 3.11-30, EA2.05, 3.6
......_________________________________..__________________________________
OBJ: 107302 107202 ANSWER 3.04 (2.00)
c. Auto [0.3] C [0.3]
l b. Manual [0.4] B [0.3]
c. Manual [0.4] D [0.3]
REFERENCE STH: Chap 34; Rod Control, p 10, 11, 21
... _______........__.... __.....__...._____........._._____..........
-- K+A: 3.1-5:A4.03 4.0
..........._........_____....___.....___...______......... __________
-- OBJ:
100112 100113
~k~" *
ANSWER 3.05 (2.50)
,
PZR pressure
.
PZR level Csre Exit Therocouples M:in Coolant Tc Scurce Range NI S/G 1evel
[any 5 at 0.5 each]
_ -.. _.
_
_
_
._
_
.
.
.
-
3.
INSTRUMENTS AND CONTROLS PAGE 26 ANSWERS -- YANKEE R0WE-86/05/20-DUDLEY, N.
REFERENCE EDCR 84-308 AN3WER 3.06 (2.50)
High indicated steam press, causes lower indicated steam flow. [0.5]
Tais results in steam flow - feed flow mismatch which would attempt to lower feed flow. [0.5]
Level error would attempt to increase feed flow. [0.5]
Level error trims steam flow which would increase feedflow. [0.5]
Final level should be equal to initial level. [0.5]
REFERENCE STM: Chap 3; Condansate and Feedwater, p 61, 63
. _____________________________________.._________...._.._...___...____
KA: 3.5-38:A2.11 3.0
________...____________..._____________. ______......_____....... ____...
OBJ:
105612 ANSWER 3.07 (2.50)
Heaters deenergize (0.4]
B11ed valve (CH-LCV-222) closes [0.5)
Charging pump speed increases to maximum [0.4]
Charging pumps with control switches set to auto will start [0.4]
Eventual scram on high pressura or PZR high level due to filling
,_
the PZR. [0.8]
REFERENCE STM Chap. 15; Charging and Volume Control, p 21-24
K+A: 3.2-2.3:A2.11 3.4 Lenrning Objective: 100442 d O.
.
.. - -
- - - - - - -
...
.......
..
.
.
INSTRUMENTS AND CONTROLS PAGE 27
.
ANSWmS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 3.08 (3.00)
a. Scram [0.25]
Any two of the 6 High Power bistables will cause a trip. [0.5]
b. Scram [0.25]
Any 2/4 levels on SG will cause trip. [0.5]
c. No scram [0.25] High and low pressure trips provide separate signals
'
therefore 2/3 coincidence is not met. [0.5]
d. Scram (0.25]
Each failure will indicate MCP failure and meet the 2/4 logic for scram. [0.5]
REFERENCE STH: Chap $1; Excore Instruments, p 31-39 Chep 33; Reactor Protection, p 19, 34, 36 K+A: 3.1-44:EA2.04 4.4 Learning Objectives: 101205 101203 101213 ANSWER 3.09 (3.50)
a. Closes or is prevented from opening b. 3 LPSI pumps start
,,,
c. No response d. No response e. No response (opens after 12 sec. due to timer)
'
f. Opens g. No responsa
"
[0.5 each]
,
REFERENCE
"
STM: Chap 19; Emergency Core Cooling p 13, 16, 19, 55, 60, 62A, 66 K+A: 3.2-17:A2.01 3.9 Learning Objective:
100616
i
.
3.
INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 3.10 (3.00)
.
c.Undercompensated(0.h The effective threshold of the IR high SUR scram protection is raised which allows a greater SUR and hence a higher peak power will be reached prior to reactor scram. (0.4 6'
is,
b. At 16 MWe and decreasing contact K3-P closes (0.5)
At 14 MWe and decreasing contact K5-P ' closes (0.5)
Depressing the Trip Bypass PB energizes Ki-P & K2-P relays (0.5)
The Ki-P contact closes sealing in the K1-P & K2-P relays which block the at power scrams. (0.5)
REFERENCE Yankee S*IN, pg 31-26 Yankee STM, pg 33-10, RO LP obj. 101218 & 101211
_........._..______________________._._________________..... ________
OBJ:
101203
,
- .v-e s
%
e Y
'
c
v"4%f me. * Tk sic
- - -
- -.
--
.
,
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29 RADIOLOGICAL CONTROL ANSWERS -- YANKEE R0WE-86/05/20-DUDLEY, N.
ANSWER 4.01 (1.50)
c. 300 mram/ cal, wk. [0.50]
,
'
b. 2500 mram/ cal. qtr. with completed Form-4 [0.50]
c. Plant Health Physicist or his designated ~ alternate. [0.50]
REFERENCE
,
'
Radiation Protection Manual pp.
10,11
_________._______.._____.________...________......___________..._________
NUREG 1021, ES-202, p 2 KA:
3.11-31 EK1.02 2.5
...__________________________......________________________________________
OBJ: 300110 ANSWER 4.02 (1.50)
To reduce the flow which will lower the pressure drop due to friction in the pump suction line and incresse the Net Positive Suction Head for the pumps which remain running. [1.5]
REFERENCE STM: Chap 19 Emergency Core Cooling OP-3051, p 2
______.........____________..______..._____________________________.______
KA: p 3.2-11, K6.03, 3.6
........______....____...__._________.___________________.__..___________
OBJ:
100616 ANSWER 4.03 (2.50)
Depress the Manual Scram Button (s) [0.6] and depress the Manual Turbine Throttle Trip. [0.6]
o,m r If rods still do not move begin an immediate emergency boration
+
per OP-3105 [0.5] then manually open the scram breakers in the Switchgear Room [0.4] and open circuit No. 12 Control Rod Drive Power Supply on No. 2 Battery switchboard. [0.4]
,
l
!
REFERENCE OP-3000; Attachment A, p. 1; Attachment B, p. 1; Attachment C, p. 1
..._____..________________......___________.......___.____________..__.
KA: p 3.1-50, Generic 11, 4.5
....___.......______._________.____.._________..._________....._______
OBJ: 300323 l
_
--
_
- - -
__
_
- - _. _
_,__. _..._ _..
-,_
.
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
ANSWER 4.04 (2.00)
a. Shutdown one of the three trains of SI. [1.0]
'
b. Adding cold feedwater at a high rate could compound the accidentCO.7]
or further decrease MC pressure due to cooldown. [h&]
0,3 REFERENCE OP 3051, p. 1 OP 3051; Attachment B, p. 4
__...___________.._________________..__________.___________.._______..____
KA: p 3.3-12, EK3.12, 4.4 p 3.5-28, A2.04, 2.9
,
p 3.3-8, EK3.24, 4.1
____....___________....___________________..___________________..._________
OBJ: 300317 ANSWER 4.05 (2.00)
Reactor Scram [0.5]
Turbine Trip [0.5)
Boiler Feed Pump auto trip [0.5] 3
' " ~
Condensate recire valve opens [0.9)
P 2. R HE ATER S ENERG11E [0,2]
REFERENCE OP-3250, p. 1.
_...._... _____..._____________________.....______..._________.........
,,
KA: p 3.7-27, EA2.19, 4.0 m.
__......_....____..____.____.._-_______....._-_.._....._____..._______...._
OBJ:
300323
-
ANSWER 4.06 (2.50)
wp Delta T less than full power [0.6]
'#
Tc following SG Tsat, and constant or decreasing [0.7]
Th stable [0.6] og Deceremvs.
Th matched with core exit thermocouples (0.6]
REFERENCE OP-3054, Rev 4, p2
.......________________ __________.________...___...._______________. ___
KA: p 3.3-10, EK2.37, 4.2
.._____..........________..___________________.___...___________________
OBJ: 300326
.
._ _
_
__
.
.
.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- YANKEE R0WE-86/05/20-DUDLEY, N.
ANSWER 4.07 (2.00)
Prevent LTOP events-[0.8)
Monitors: main coolant system pressure [0.3]
status of Pzr heaters [0.3]
status of SI pumps and valve lineup [0.3]
'
Maintain 400 psig margin between plant and NDT upper limit curve. [0.3]
REFERENCE OP-2100 Rev 31, p 4, 13
__...__________________________..___________.______...___. ___________
KA: p3.3-1, K4.03, 3.8
______________ __....___________________.._______________.___________.._
OBJ:
300206 ANSWER 4.08 (2.00)
a. -10" b. either small hogger or vacuum priming pump operating, air ejector secure c. All breakers racked out and locked d. Locked closed
,
e. Auto-cutout
[0.4 each]
REFERENCE
.
OP-2105 Rev 26, p 7, 8
.,.
OP-2653 Rav 18, p 2, 3
,
,
__......_________...______...... ________________...____________________
KA: p 3.5-22, A4.03, 1.8 p 3.2-12, A4.01, 4.1
..._.___......___________________________.. _____________________________
OBJ:
300206
, 4%
I
.__
. - - -
'.
4.
PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL ANSWERS -- YANKEE R0WE-86/05/20-DUDLEY, N.
ANSWER 4.09 (3.00)
c.
High radiation on the air ejector radiation monitor.
Level increase on faulted steam generator.
Feedwater flow decrease to faulted steam generator.
High radiation on main steam line radiation monitors.
Possible initiation of safety injection.
[0.3 each]
b.
1. To minimize the release of radioactive material.
2. To establish capability to supply feedwater to all steam generators and to isolate feedwater to the faulted steam generator.
3. To maintain the ability to remove the necessary decay heat from the reactor through the intact steam generators via the steam dump valves.
4.Tomaintainthemaincoolantsy{teminasubcooledstate during the recovery.
5. To prevent flooding the faulted steam generator.
[any 3 @ 0.5 each]
REFERENCE OP-3107, p. 1
___.._____.___________...____.. ______________._________________..____
KA: p 3.4, A4.08, 4.1
__________________....___________._______________..____..._______________
OBJ: 300212 300326
.n ANSWER 4.10 (3.00)
,
l Scram reactor and initiate OP-3000, " Emergency Shutdown frcm Power"
.
Initiate OP-3300, " Classification of Emergencies."
+
After opening the exciter field breaker, stop all main coolant pumps
initiate OP-3054, " Natural Circulation."
erv9Fu
<
-
If loss of component cooling flow is due to loss of A.C. supply,
'
initiate OP-3251," Loss of A.C. Supply."
[ 2. 00,- 0. 5 0' ;;i-]
)
[if @ 0*15 vsk
!
REFERENCE OP-3115, pp. 1-2.
.___________...____________________.___________________________.______
KA: p 3.10, Generic 11, 3.9
____............________________.__...__._________________________...
OBJ: 300323
. _ -
.
.
4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- YANKEE ROWE-86/05/20-DUDLEY, N.
.
ANSWER 4.11 (3.00)
c. Maintain plant at power b. Scram c. Shutdown
..
,
d. Scram e. Shutdown f. Reduce power REFERENCE OP-3004 Rev 7, p 2 OP-3009 Rev 7, p 2 OP-3100 Rev 9, p 2 OP-3108 Rev 5, p 1 OP-3114 Rev 9, p 2 OP-3118 Rev 7, p 2
___________________ _______________..______.______...._. ________. ___
KA: 3.1-44, Generic 10, 3.9
_...___.__________._....__________________. ___.... ________..........___
OBJ: 300317
_
,M
_,
/
/
- _ _.
.-.
.
- - - _
. -.. _.
.
.
__
__ -
_ _ _ _ _ _ _
1)10 (f WE
\\
M/OS/20
.
TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE
_
6.d-Ofg jf)-
.......-....--
..........
01.01
.80 DUD 0001460 /Vu 01.02 2.40 DUD 0001459 #)f 01.03 2.40 DUD 0001461 JV/
01.04 2.50 DUD 0001463 RV3 01.0 2.50 DUD 0001466 #46 01.0 2.50 DUD 0001467 JV7 01.0 3.00 DUD 0001458 43f 01.0 3.00 DUD 0001462 dvA 01.09/
2.90 DUD 0001464 4 VV y.
01.10 3.00 DUD 0001465 py5
..___.
25.00 02.01 1.50 DUD 0001427 30 02.02I 1.50 DUD 0001477 #5Y 02.03 2.00 DUD 0001425 J/5 02.04 2.50 DUD 0001428 #8t 02.05l 2.00 DUD 0001429 dif 02.063 2.50 DUD 0001430 ###
DUD 0001431 >#/)
02.07 2.50 DUD 0001433 D 02.08 /
2.00 02.09'
2.50 DUD 0001434 2"/
02.10 3.00 DUD 0001426 p.<6 02.11)
3.00 DUD 0001476 3 53
______
25.00
~
03.01 1.50 DUD 0001473 J52 03.02 2.50 DUD 0001449 53" 03.03 2.00 DUD 0001450 23/
03.04 2.00 DUD 0001451 151 03.05 2.50 DUD 0001452
>8}
s
.
03.06 2.50 DUD 0001453 A W
.,,
03.07 2.50 DUD 0001454 #35 (j.,.
.
-
03.08 3.00 DUD 0001455 13'
.,," 7 03.09 3.50 DUD 0001456 /17 03.1 3.00 DUD 0001478 >II
.....
25.00 t
..
04.01 1.50 DUD 0001424 J/V 04.02 1.50 DUD 0001432 zit 04.03 2.50 DUD 0001436 1z6 04.04 2.00 DUD 0001438 tz7 04.05 2.00 DUD 0001439 t?T 04.0 2.50 DUD 0001469
/ uf 04.07 2.00 DUD 0001470 /I#
04.08/
2.00 DVD0001471 # 5/
04.09J 3.00 DUD 0001435 z*5 04.10d 3.00 DUD 0001440 279
-
-.
.
-.
-
_. _
- -
--~ -
g
-
_
.
s'
.
.
V TEST CROSS REFERENCE PAGE
QUESTION VALUE REFERENCE Afi _4
.--
________ ______
_. _______
,
04.11 3.00 DUD 0001468 #d
______
25.00
______
______
f 100.00
.
'I
,
<.y>
i
!
!
- J
%
--
,
a.,
.-
-<
-
Rituhmed 2ppsrER
.
.
EXMM U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
YANKEE R0WE
~
REACTOR TYPE:
PWR-WEC4 DATE ADMINISTERED: 86/05/20 EXAMINER:
BARBER, S.
APPLICANT:
.
INSTRUCTIONS TO APPLICANT:
UI;c esparate paper for the' answers. Write answers on one side only.
Stcplo question sheet on top of the answer sheets.
Points for each que2 tion are indicated in parentheses after the question. The passing grcda requires at least 70% in each category and a final grade of at
,
lcr::t 80%. Examination papers will be picked up six (6) hours after the cxamination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL i
CONTROL
'
25.00 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS j
100.00 100.00 TOTALS FINAL GRADE
%
411 work done on this examination is my own. I have neither givin nor received aid.
-
APPLICANT'S SIGNATURE l
.
.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
.
THERMODYNAMICS QUESTION 5.01 (.50)
Which one of the following descriptions best supports the reason why Xenon reactivity increases sharply after a trip from 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> at 100% power?
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 production 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.
QUESTION 5.02-(2.00)
Answer the following questions concerning coefficients of reactivity.
et Me" 2nd thy deer the value cf MTC ch2ng: 2 tr ;cr:ture in incrc:::d
,
An => nuerraderated cere ? (0.75)
.
b. Which initial temperature,58 F or 557 F, will give the largest change in the magnitude of MTC as boron concentration is lowered from 1000 to 500 ppm 7 BRIEFLY EXPLAIN WHY, (0.75)
c. WHY does power defect become more negative as the core ages 7 (0.5)
QUESTION 5.03 (2.00)
Consider two separate startups for physics testing. Assume the only difference between the two is that during startup #1 rod speed is twice as during startup #2. Qualitatively COMPARE and BRIEFLY EXPLAIN the difference between the two startups with regard to the following:
(The time of rod pulls and time between rod pulls are the same.)
a.
Critical rod height
.
(0.5)
b.
Time to criticality (0.5)
Power level at criticailty (0.5)
c.
d.
Startup Rate at criticality (0.5)
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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS QUESTION 5.04 (3.00)
c. If core power is increased from 50% to 100%, how, if at all, will differential rod worth change (increase, decrease, stay the same) for the following 3 cases ? Justify your answer.
.
1) Rod position and boron held constant, and temperature allowed to decrease.
(1.00)
2) Boron constant, group C is at the core midpoint and is then fully withdrawn, temperature remains constant.
(1.00)
3) Rod position constant, boron dilution used, temperature constant (1.00)
QUESTION 5.05 (2.00)
c. Give two reasons why Sm-149 is not as much of a concern to an operator after a reactor trip as is Xenon. (1.0)
b. A Xenon oscillation in a reactor core might be produced by certain types of rod motion. How would the Xenon oscillation resulting from the following two cases be different? Explain. (1.0)
1. A turbine runback occurs with rods in auto.
Rods drive JA/
ts-ett 60 steps.
2. Rods are driven 60 steps starting from the same position initial pcsition as in Case 1, but are slowly driven in over four days.
QUESTION 5.06 (2.00)
-9 A reactor at BOL is critical at 10 amps. Rods are withdrawn at 30 inches psr minute for 10 seconds in sequence.
a. Calculate the resultant SUR immediately after the rod pull.
Assume differential rod worth is 20 pcm per inch, effective delayed neutron fraction is 0.006 and the effective precursor decay constant is 0.1 sec-1.
(1.0)
b. Explain how and why the SUR would change if the reactivity change was conducted at EOL 7 No calculations are necessary, (1,0)
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS QUESTION 5.07 (1.50)
c. During natural circulation cooldown, pressurizer level suddenly increases after.the initiation of pressurizer spray. What is the most
'
probable cause for what is occuring ? (0.5)
b. Assume a small LOCA results from the rupture of a pressurizer level transmitter sensing line. Which would cause a more rapid depressurization: a rupture in the reference (upper) line or in variable (lower) sensing line ? Justify your answer.
(1.0)
QUESTION 5.08 (2.50)
a. What is the subcooling margin (SCM) of the plant if the following conditions exist ? (1.0)
Th =525 F Ppzr=1985 psig Tave=515 F Psg =685 psig Tc =505 F b. If plant power is lowered from 75% to 50%, how will SCM change (increase, decrease, stay the same) 7,WHY 7 (0.75)
c. Which one of the following would give a smaller SCM7 Assume constant and normal Main Coolant (MC) and Steam Generator (SG) pressures for each case. Briefly explain your answer.(0.75)
1) SCM during a controlled natural circulation cooldown immediately following a reactor trip from loss of flow.
2) SCM from continued operation at 5% power 3) SCM produced when all RCP's are operated at normal no-load temperature after extended shutdown.
%
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS QUESTION 5.09 (3.00)
a.
While at 100% power, asseme a loss of station power with a concurrent diesel failure results in a total loss of all station power except for batteries. Starting when power is lost, explain how the core will be cooled.
Include the final heat sink and assume no operator action.
(1.5)
b.
How will the following parameters be trending if the core is not being cooled following a loss of power.
Briefly explain.
1.
Delta-T across the core (0.5)
2.
Steam generator level (0.5)
3.
Steam generator pressure (0.5)
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QUESTION 5.10 (2.00)
Assuming you are operating at 85% power indicate how the following changes in plant conditions would affect DNBR (increase, decrease, remain constant). Consider each case separately.
1) The operator withdraws contrcl rods without changing turbine load.
2) Steam Generator safety fails open.
3) Pressurizer heaters are inadvertantly left on.
4) Reactor coolant pump speed decreases.
QUESTION 5.11 (2.00)
The plant is operating at 20% power when the "A" S/G Main Steam (MS) non-return valve fails shut due to a malfunction. Using the listed initial conditions, CALCULATE the new steady state values for the parameters shown.
Assume no operator action, all control systems in manual, and a reactor ceram does not occur.
State all assumptions and show all work.
Initial conditions: Tavg = 515 F Tstm = 500 F Delta T =
8F
-
a.
Turbine power (0.5)
b.
Tavg (for the affected loop) (0.5)
-
Tavg (for the non-affected loops) (1.0)
c.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS QUESTION 5.12 (2.50)
a.
Erittle fracture of your reactor's pressure vessel can occur at stresses well below its yield stress. TWO conditions must be present besides high yield stress. What are these TWO conditions?
b.
How do heatup/cooldown rate limits on the reactor coolant system reduce the probability of brittle fracture?
(0.5)
c.
Why does the concern about brittle fracture of the reactor pressure vessel increase as the plant ages? Include in your answer the specific material PROPERTY that is affected.
(1.0)
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(***** END OF CATEGORY 05 *****)
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
QUESTION 6.01 (2.50)
Antwer the following questions regarding the Vapor Container (VC)
ctmospheric control system.
c. What are the high and low limits on VC pressure? Why must they be maintained ? (1.5)
b. How is a manual VC isolation different from an automatic isolation 7 (0.5)
c. Both VC isolation reset pushbuttons have been depressed in an attempt to operate the Component Cooling Water inlet and outlet valves and the valves won't open. Assuming the test-normal-bypass switches are inoperable, what must be done to successfully open these valves ? (0.5)
QUESTION 6.02 (2.50)
a. List 4 separate and distinct design features that lessen the effects of an uncontrolled MC pressure increase. Setpoints are required.
Assume initial pressure is 2030 psig. (2.0)
b. How will the plant respond if the pressurizer spray valve fails open with all the heaters energized at 100% power. Assume all systems are in automatic and operator action does not occur. Describe the transient until stable plant conditions are reached or until the plant scrams.
(0,5)
QUESTION 6.03 (2.25)
a. All three charging pumps auto-started on low level in response to a small MC leak (3 gpm). If bleed flow is 20 gpm, then how will the charging system continue to respond if no operator action is taken ?
(1.0)
5. Deceribe th-Che g4a
=a d val'-- raa * ea l flaup=*h if tha high pr===nra
_r-fety injectier eyeter ic uced te preride eten: ;cncreter ::crgency
= feed. Include =cjer pump cnd eclve etetuc. (0.75)
-er-Describe-hee charging-flow is maint einad ukaa *ha high pu==nva==fary un}ect4on-syste= is used-to-feed-the-steam-generatcrs. '0.5)
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
QUESTION 6.04 (3.00)
c. List 3 plant conditions that will cause the MS non-return valves to close automatically. Include nominal setpoints. (1.5)
b. How would the MS non-return valves respond to an automatic closure signal if all of the nitrogen in their accumulators had been bled off 7 Briefly explain why ? (0.75)
'
c. How would the plant respond if a MS non-return valve was inadvertantly closed 7 Assume the valve's trip relay failed to energize on closure.
Explain your answer. (0.75)
QUESTION 6.05 (2.75)
c. A continuous rod withdrawal accident occurs from low in the source range. Will an undercompensated or overcompensated IR detector allow a
<
greater peak power to be reached ? Briefly explain why? (0.75)
b. Using figure 6-1 provided, explain the cperation of the relays and contacts necessary to block the At-Power reactor scrams. Assume initial turbine load is 20 MWe and decreasing. (2.0)
QUESTION 6.06 (2.00)
Indicate what automatic actions, if any, will occur for the following Rrdiation Monitoring System alarms, n. High level on the particulate monitor on the primary vent stack during venting of the Vapor Container.
b. High level alarm on the steam generator blowdown tank effluent monitor.
c. High level alarm on the liquid radwaste effluent monitor during a radwaste discharge.
d. High level on the spent fuel pit area radiation monitor with the ventilation system in normal operations.
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PLANT SYSTEMS' DESIGN, CONTROL, AND INSTRUMENTATION PAGE
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QUESTION 6.07 (2.75)
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c. Haw would a loss of power to the cam motors directly affect the rod
'
position indication system? Briefly explain why. Assume rods are in cuto at 100% power. (1.0)
b. How, if at all, would the plant and. rod control system respond to an
,
uncontrolled-dilution while in automatic at hot full power ? Include cny applicable setpoints. (1.0)
!
c. At what power level is the single step rods out circuit effective ?
Explain how rod motion is affected when this circuit is in use?(0.75)
,
- QUESTION 6.08 (1.50)
Explain how the No. 1 480V emergency bus UV protection system would respond
lto a complete loss of power.
QUESTION 6.09 (3.00)
a. List the eight separate, distinct functions that are performed by the l
SIAS-A relay when it is tripped. (2.0)
~b.' List two ECCS design features which ensure large amounts of nitrogen are
'
not injected into the main coolant system following a cold break 7 (1.0)
,
- . QUESTION 6.10 (2.75)
i a. How does the fire protection system pressure maintenance system respond
-
l to under AND overpressure conditions ? Assume the pressure changes are due to changes in level. (1.0)
b. A fire occurs in one of the battery rooms. Explain the sequence of
- -
avents that is necessary to actuate the halon system. (0.75)
s c. How is the actuation of a battery room's halon system different.than i
that of the switchgear room? (1.0)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10 RADIOLOGICAL CONTROL
)UESTION 7.01 (3.50)
Answer the following questions concerning OP-2130, Start Up and Cut-In of an Ic31cted. Loop.
h. Wh:t are the three different methods (system lineups) that may be used to provide pressure control to an isolated loop ? (1.5)
b. Stcrting with the listed valves closed, describe the necessary sequence of volve operations (open, shut) to cut in an isolated loop. Consider only the valves shown. Assume all required plant conditions have been estchlished. (1.0)
Loop bypass valve Loop Tc stop valve Lorp Th stop valve c. Wh:t actions must be taken if an increase in neutron flux level occurs whilo cutting in an isolated loop ? (1.0)
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)UESTION 7.02 (2.00)
Antwar the following questions concerning OP-3117, Refueling Accidents, s. Communications are temporarily lost between the refueling crew and the control room when improper fuel loading causes an inadvertant criticality. List the first two audible indications that would be rsesived in the VC7 Why would these be the first warnings received ?
(1.0)
b. Wh2t immediate operator action is taken to establish reactivity control during a Criticality Accident in the VC. (0.5)
c. Which*one of the symptoms shown would you NOT expect to see if a NEW fual assembly (FA) was dropped while moving it. Assume it was above the w ter when it fell. (Choose,the best answer) (0.5)
1. Visible damage to the FA 2. Gts bubble emission from the FA 3. Plant accident area radiation monitor high alarm (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 11
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RADIOLOGICAL CONTROL
,
QUESTION 7.03 (2.25)
Answer the following questions concerning OP-2103, Reactor Startup and Shutdown.
a. What are the plant indications necessary to call the reactor critical when diluting to criticality ? (1.0)
b. Are the scram breakers tested on every reactor startup 7 If so, how ?
If not, why not ? (0.75)
c. What restriction (s) is/are imposed upon a reactor startup if there are personnel working in the loops or shield tank cavity ?_(0.5)
QUESTION 7.04 (2.50)
a.-List the three automatic actions that occur when power is lost to No. 1 Vital Bus,per OP-3250. (1.5)
b. How is the operation of the Feed and Bleed and pressurizer related valves affected on a loss of No. 1 Vital Bus ? Include the valves that are affected. (1.0)
QUESTION 7.05 (3.00)
In order to maintain the plant at 50% power, work must be performed inside the containment in a radiation field of 110 MREM /HR gamma and 1.2 REM /HR n2utron. The maintenance man selected is 28 years old and has a lifetime exposure through last quarter of 33 REM on his NRC Form 4.
In addition, he hns accumulated 1.8 REM so far this quarter. Assume all the necessary cdministrative approvals will be granted.
a.
Calculate the length of time the man may work in this area without exceeding his 10CFR20 limit? Show all work and state all assumptions.
(2.0)
b.
How would the staytime change if his entry was to save a life? (1.0)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL QUESTION 7.06 (2.75)
Answer the following questions concerning OP-3000, Emergency Shutdown from Power.
c. List the steps of OP-3000 that are specifically written to shutdown the reactor in the event of an ATWS. (1.25)
b. List the actions that occur or are required at 1650 psig and 1200 psig.
Indicate whether they are manual or automatic. (1.0)
-c. True or False After a reactor scram all main coolant pumps should be restarted as soon as possible and left running unless a bona fide loss of main coolant pressure caused automatic initiation of SI. (0.5)
QUESTION 7.07 (1.50)
Answer the following questions concerning operation of the Feedwater system.
a. What automatic action would result if a scram occurred and a standby BFP control switch was set in AUT07 What adverse consequence could result because of this action? Assume the turbine lockout to BFP's is inoperable (1.0)
b. True or False Steam flow should be established to a feedwater heater just prior to establishing condensate or feedwater flow. (0.5)
QUESTION 7.08 (2.75)
n. If a Loss of Main Coolant has occurred, unc'
vt 2 conditions is it permissible to trim back SI flow to two t5,in.
6.tning 7 (1.0)
b. If an inadequate core cooling condition resulted in incore thermocouple temperatures exceeding 700 F, where will the operator be able to read these temperatures and what is their upper bound ? (0.75)
c. What is the purpose of initiating hot leg injection 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA 7 If, hot leg injection is not initiated what potentially adverse consequence might result ? (1.0)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL
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QUESTION 7.09 (3.00)
An:ver the following concerning Main Coolant Pump (MCP) operations:
c.
When must a MCP be stopped on a loss of Component Cooling Water? (1.0)
b.
Why is the temperature limit on the MCP bearings different for a loss of Component Cooling Water? How is the limit changed? (1.0)
c.
Why must a MCP be thoroughly vented, prior to running, after a shutdown when the Main Coolant system has been depressurized ? (1.0)
QUESTION 7.10 (1.75)
An:ver the following concerning hydrazine addition to the Main Coolant System:
e.
Why is the initial charge of hydrazine 2.5 times the theoretical amount required for oxygen scavenging? [1.0)
b.
Under what temperature conditions may hydrazine be used for oxygen
'
scavenging of the Main Coolant System? [0.75]
.
a ethm
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 14
.
QUESTION 8.01 (2.00)
An:ver the following questions concerning facility staffing per Tech Specs.
A;cume plant power is 100% unless otherwise specified.
c. Due to unforseen circumstances only one Radiological technician is cvailable to relieve the offgoing shift. Is this permissible by Tech Specs ? (0.5)
b..What is the minimum number of fire brigade members ? What is the minimum number of licensed operators necessary for safe shutdown ? Is it permissible to use an individual required for safe shutdown as a member of the fire brigade ? (1.0)
c. Per Tech Specs and AP-2001, choose the one set of c>rew manning that best represents the minimum requirement for reactor startup after a refueling outage. An STA is on shift in each case. Select the best answer. (0,5)
1) 2 SR0s, 1 RO in' the control room, 2 A0 2) 2 SR0s, 2 R0s in the control room, 3 A0s 3) 1 SRO, 2 Ros in the control room, 3 A0s 4) 1 SRO, 1 RO in the control room, 2 A0s QUESTION 8.02 (3.00)
For the leakage conditions shown below, indicate whether you could continue to operate indefinitely or must shut down within specified time requirements (Time requirements need not be specified). ' Consider each condition separately. Assume no other concurrent leakage. Include in your cnt;wer the limit the leakage is being evaluated againnt-h c.
0.8 GPM from each of four different valve packing glands. (0.75)
b.
0.1 GPM from a loop RTD weld. (0.75)
.
c.
1.1 GPM unknown leakage. (0.75)
d.
3.0 GPM through the body of a pressurizer safety valve. (0.75)
QUESTION 8.03 (2.00)
What four actions or reports must be completed following a loss of fesdwater transient which results in'RCS pres sure exceeding 2735 psig.
Include time limits for actions to be ccmplet ed in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less.
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NISTRATIVE PROCEDURES, CONDITIONS, JJO LIMITATIONS PAGE 15
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8.
auni QUESTION 8.04 (2.25)
Answer the following co'ncerning Shutdown Margin (SDM);
a.
Wh;c action, if any, should be taken at 25% power if it is determined that SDM = 5.0% Delta K/K ? Setpoints are required. (1.25)
b.
List four factors that are considered when calculating SDM? (1.0)
.
QUESTION S.05 (1.50)
A Diesel Generator's operability load test, which is required every 31 days, is scheduled for today. The last three tests were completed 36, 68, and 102 days ago respectively. The plant is at 100% power. Are Technical Specifications being met? Explain why or why not.
,
QUESTION 8.06 (2.25)
a. What three signatures are required on a Radiation Work Permit (RWP)
before work is allowed to begin ? (0.75)
b. List five circumstances that require the issuance of an RWP. (1.5)
(***** CATEGORY 08 CONTINUED ON -NEXT PAGE *****)
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16
QUESTION 8.07 (3.00)
Classify each of the following occurences using the EAL Classification guides provided.
Identify what factor was used to determine the final classification.
c. A natural circulation cooldown is in progress due to a recent loss of offsite power. Multiple malfunctions in the main and emergency feedwater systems will not allow the feeding of any steam generator.
-(0.75)
.
b. A continuous rod withdrawal accident causes a reactor. trip on overpower. MC pressure peaked at 2285 psig by strip chart indication j
and is returning to normal. The pressurizer SORV improperly opened for approximately 10 seconds and appears to have properly reseated. (0.75)
,
c.'An accident in the containment results in a broken leg to a worker.
i
,
The worker is still in his contaminated anti-C clothing.
Due to the a~ccident the worker received an exposure of 2 rem.
He is presently on his way to the hospital. (0.75)
d. Following operations at 100% power for six months a steam generator tube rupture occurs which automatically starts all charging pumps. They
,
continue to run until the plant trips on low MC pressure. SI has failed to auto actuate and one hc non-return valve is still open. (0.75)
QUESTION 8.08 (2.00)
Par Technical Specifications, are the following components operable based
_ on the below information in mode 1. (YES/NO only-no explanation required):
a. The safety injection tank level is at 27.0 feet of 2350 ppm boron at a temperature of 125 F. (0.5)
b. Two separate and independent diesel generators are an adequate amount of
'
fuel oil in their separate day tanks and in the fuel oil storage system.
(0.5)
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c. Specific activity of the primary coolant is 1.2 microcuries/g I-131.
(0,5)
d. The MS non-return valves are tested and stroke from full open to full closed in 30 seconds. (0.5)
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 17 QUESTION 8.09 (3.00)
Answer the following questions concerning Yankee Adminstrative procedures.
a. List (by title) the 2 individuals whose approval is needed to startup after a scram. (1.0)
b. Who is normally required to sign the logs (except the Operating Log)
after entries are made 7 Who signs for entries made in the Operating Log 7 When is this signature required to be made 7 (1.0)
c. List'4 of the 6 log books that the Supervisory Control Room Operator must review prior to watch relief. (1.0)
QUESTION 8.10 (2.50)
.
Answer the following questions concerning Tech Spec Instrumentation.
a. Tech Specs allows the bypassing of certain scrams if the reactor is shutdown. Which one of the following is NOT one of these scrams. Assume the reactor trip breakers are closed and the rod drive system is capable of withdrawing rods. (0.75)
1) High MC pressure 2) Low MC p'ressure 3) High pressurizer water level 4) PR NI flux low setpoint
-
5) MS line isolation trip logic b. Which one of the following lists best represents the reactor protective system instrumentation required to be operable per Tech Specs. (0.75)
1) MCP current, VC pressure, Turbine trip 2) MCP current, High pressurizer level, VC temperature 3) Steam generator DP, VC pressure, VC temperature 4) Steam generator DP, High pressurizer level, Turbine trip c. List 3 of the 4 conditions that must be met to satisfy the operability requirements for the incore detection system. (1.0)
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 QUESTION 8.11 (1.50)
An:ver the following concerning the tagging of plant equipment:
a.
Who(by job title) is the Local Control Authority for tagging a Diesel Generator out of service? [0.75]
b.
Who may perform locally controlled plant tagging? [0.75]
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
AGE
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ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
.
ANSWER 5.01 (.50)
B CO.52 REFERENCE Yankee WEC-C' ore Control for large PWR's pg 4-24 to 4-39 ANSWER 5.02 (2.00)
c_
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beccccc 1ccc pc;itre
~ 0. 2 5 ';
Ir:c r e e s i ^; te peret'-e #e-eet-e-e be-eted 'ete-c 'a t e# the cere. Thic cred;ces ther-el e b e c r r t i e r-e r:d edd_ reeiti"e *eee+i"it" te +he re e-(0.5)
b. 557 F (0.25)
The change in density at higher temperature is much greater than at lower temperature. So more boron atoms enter or leave the core at higher temperatures which causes a larger reactivity change for a given temper-atore change. (0.5)
c. Lowering bcron concentration over life makes MTC more negative which causes its contricution to pouer defect to increase. Hence power, defect increases. (0.5)
REFERENCE a.
(Ankee WEC-Core Cont,rol for Large PWF. pg 3-14 b.
3-19 3-20 c.
3-25
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
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ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
ANSWER 5.03 (2.00)
'
s.
Critical rod height:
same for both [0.23 Rod height is dependent upon the amount of reactivity needed for criticality CO.33 (0.5)
b..
Time to criticality shorter for startuP 41 CO.23 The reactivity addition rate is greater, thus a shorter time CO.33 (0.5)
c.
Power level at criticality:
lower for startup 41 CO.23 The higher rod speeds do not allow for neutron population buildup as great as in startup 42 CO.33 (0.5)
d.
SUR at criticality *
Higher for startup 41 [0.23 This is because of the higher reactivity addition rates [0.33 (0.5)
REFERENCE Yankee WEC-Cor: C rt :1 for L;r3:
pgs 8-12 to 8-54 o"o
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F w us u niAL.1 W & cst M Ach,nq hi YRLS
ANSWER 5.04 (3.00)
<
4_&
1) cecrease(0.3)1ouer temperature reouces the roos ' sphere of influence'
bv recucing the number of neutrons tne rod sees.-(0.7)
2) Decrease (0.3)
Rods are moved from a high flue region to a low fluy regioni rod worth decreases since worth is proportional to flun.
(0.7>
3) Increase (0.3)
Decrease in coron increases the rods ' sphere of influence" by increasing the number of neutrons the rod sees.
(0.7)
REFERENCE
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a. Yankee WEC-Core Control for Large PWRs, pg 6-14 to 6-16 b.pg e-22 ANSWER 5.05 (2.00)
a.
Sm-149 has a smaller absorption cross section and therefore less reactivity worth than Xe CO.53, and does not decay like Xe. [0.53.
b. Case 1 would be a more noticeable Xe transient CO.23 because the local power changes occurred rapidly with respect to Xenon ability to maintain' equilibrium with local power. CO.83
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THEORY OF NUCLEAR POWER' PLANT OPERATION, FLUIDS, AND PAGE
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ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
,
REFERENCE.
.,
Yankee WEC-Core control for Large PWR's, pg 4-24 to 4-39 ANSWER 5.06 (2.00)
a.
In 10 sec, rho
=(20pcm/ inch)(30 inch / min)(1/6 min)=100 pcm (0.5)
SUR= 26'(lambda-eff)(tho)/(beta-eff minus rho)
.
SUR = 26 (1E-4)/5E-3 = 0.52 DPh (0.5)
b.
At EOL, beta effective is less (0.6) which would give a higher SUR and shorter time to increase power.
CO.43 REFERENCE Yankee WEC-Fundamentals of Nuclear Physics, pg 7-54 ANSWER 5.07 (1 50)
a. The formation of a steam void in the RV head causes an insurge into the the pressurizer.
(0.5)
6.
The variable leg rupture would be more rapid. (0.3)
The rate of mass m
--y loss is greater in the pressuriner uster space., (0.7)
Ako evtecces Tnt u.Hau n eersy an FRM ret estrnr.sst scs-Nca h.
REFERENCE ranKee WEC-Tt.erm. Hva. frinciples and Applic. to the FWR II, pg 14-26
e e
_, -
v-,,, - - -
,-e..w.,g,_. _ _
_
.
.
.
.
So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
!.
TUER5665UAb5C5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
______________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
'
ANSWER 5 08 (2.50)
a..From the C-E Stm Tables, G36
Tsat for 2000 psia = 402.0 F (0.5)
III T s a t - T h = G,l,iS-525= 117.0 F (0.5)
SCH=
b '. increase (0.25)
Th decreases as unit delta T decreases with power (0.5)
c.~1 (0.25)
Core delta T during natural circulation cooldown will approach full load delta T.
Thot is greater than in the other 2 cases.
(0.5)
REFERENCE Yankee WEC-Therm. Hyd. Principles and Applic. to the PWR II, pg 12-20, A-21
,
o
' ANSWER 5.09 (3.00)
a.
Natural circulation flow removes heat to steam generators (due to L
change in censttv of water) [0.753 W a r. e r in steam generators caanse to steam and is vented to atmosphere
'
(0.75)
,
b.
1.
Increase E0 25]
Thot increases as Teolo remains constant [0 ' 25]
.
2.
Decrease [0.;5)
Tnere is no feed flcu to steam generator C0.25]
3. Decrease CO.253 No heat input to maintain temperature EO.25]
REFERENCE Thermal Hvdroalic II pgs.-14-26, 14-27
___.i_______________________________________________________________________
3.4 000 015 EK 1.01 4.4 3.7 000 055 EK 1.02 4.1 ANSWER 5.10 (2.00)
1) DNBR decreases
!
2) DNBR decreases 3) DNBR increases 4) DNBR decrea,ses 4 items (0.5) ea.
o
!
l
..
.
.
-. - - -. - - - _ -.. _.. -,. - - - - -
-..
- -
.
. - -. _ -.. -.,.... -. - - - - - _ -. ---
.
.
.
.
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,' AND PAGE
gggggg gggggg--------------------------------------
______________.
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
REFERENCE Yunkee'WEC-There. Hyd. Principles and Applic. to the PWR II, pg 13-23,13-24 ANSWER 5.11-(2.00)
a.
-Turbine power - stays constant at 20% power (0.5)
-
b.
Final Tave equals in or 519 F (0.5)
c.
total reactor power has not changed; however, the power-that each of the non-affected loops must transfer.to the
,
-5/G has increased by a factor of 1/3 to compensate for.A loop.
.
.
Grx = m Cp (Th - Tc)
for each loop
^1/3
--->
->
v, initial Delta T was 8 F, and must increase by factor 1/3
--> final Delta T equals 10.7 F (0.7)
final Tavg = Th - Delta T/2 = 519 - 10.7/2
= 513.6 F (0.3)
REFERENCE (snvee WEC-Therm'. Hvd. Principles and Applic. to the PWR'II, pg 12-18-AN5WER 5.12 (I.50)
a.
1; Presence of a flaw (or crack of sufficient size). [0.53 2) Low temperature [0.53.
(1.0)
b. Reduces the thermal stress. (Reduced DT across the RV wall reduces total / thermal / tensile stress.)
(0.5)
c. Neutron e:<posure (integrated) CO.5] makes the material more i
brittle (raises NDT-) (Reduces ductility.)
E0.53 (1.0)
i
'
REFERENCE-Yankee WEC-Therm. Hyd. Principles and Applic. to the PWR II,pg 13-58 to 62
.
.
\\
t i
-
.
--
-.- -
-
-
-
--
.
.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- YANHEE ROWE-86/05/20-BARBER, S.
"
,
ANSWER 6.01 (2,50)
a. 0.75 psis (0.25) Ensures air outleakage can be measured. (0.5)
3.0 psis (0.25) Limits VC pressure to less than design pressure (34.5 psis) during a LOCA. (0.5)
b. The manual actuation won't close the tH5 non-return valves. (0.5)
c. The lockout relays (LOR 1A/B & LOR 2A/B) will have to be reset. (0.5)
REFERENCE
,
a.
Yankee ROTM, Sec.
5, pg. 9-13 b.
Yankce STn, pg 19-66 c.
Yankee STM, pg 19-65,66 ANSWER 6.02 (2.50)
a. Reactor scram 2300 psis Design features (0.3)
'
Spray valve opens 2300 psis Setpoints (0.2)
Solenoid relief valve opens 2400 psig
,
rir:t code saf ety valves opens,2485 psig uten.
c. Reactor util scram on low pressure. (0 5)
REFERENCE a.
rankee STM, pg 14-8 14-15 OP-3311 99 3 b.
'r s n k e e Sim. pg 14-9
.
-
.
s
,
w
%
J e
.
- - -.
.-,r---,,,,....,.,-y-,--.,~-m-r,r.-r.,-,-_,r_,%-.r~.---*--,-,
w e y-r.
---,,.---,,.m-+,~m
,-----<---,----,y-
..,. ~ -,
-
.
.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
ANSWER 6.03 (2.25)
a. Charging pumps will continue to run until the high level (HL) auto-stop is reached. (0.5) Then they will cycle on and off as level drifts above and below the HL setpoint. (0.5)
5.
I r- ' " - c v ? t b ~- het le; i ': j e c t i c "- c e " ": e c t i e r-
-C a c;
,7c: g thc c h a r 5 r. 3 dicchar;c "cade-i c c 12 t t er-
'clve (Cu_wnn_.,7s i
_n.....
- - - x -.- 4 r...,
e m - + m.. + x.,,...- x.
- --
4 m_.-
,._,r..,,,3-o m, o m_,-
+m 4_ x. m
,
_m, mr
., <
+x.
-
- -.;
, ---
.
-a
-
.
--
.
._
-
_.
.--.m.
,,,, _ c u,_ a. n_ n. _ e,.3
. m.
,..3.,,._
., u.. g,,, - _
-
wm,i.
.. -.. _ -
-_
.
.
.- _
.,,..s, m,,.., _.,
m c.
., s _
.4 tu
,
_
-
.-
n.
. b. n..m-.e
- h.MC'
-.
n n
.
.
.
=c. Icc12tc tbc chcr;ir:; r " - c u c t i e r-2 r-i dicche ge div.cic- ': 1vec. (0.5)
3. c. m e.4, r u - M n u _ e.c c i n 4.., un.
r. s -,., 4 r..,
-r... -- + w m...
m.
4. s m e r_ / 3 m m r.
< 4_ 3 *_
_
..,
_.
..
- -
-.
_
_
... -_.
i n_ _cs
.,
_
REFERENCE a.
Yankee STM, pg 15-10
b, pg 15-13
-
c.
pg 15-13 ANSWEF 6.04 (3.00)
s.
High UC pressure 5.0 est:
Condition (0.3)
Low M5 pressure 270(262) psig Eetooir ts (0.Zi
'
M.
The NRV',s won't c'ese. (0.25)
attrogen provices the mctive fccce t. o close toe NRV's sgsinst hvoraulic pressure. (0.5i c.
The reactor would scram on nigh MC pressure (0.5) because of a loss of heat sink. (0.25, REFERENCE a.
Tankee STM, pg 19-58 E0P 3051 & 3201 for setpoints b.
Yankee ROTM, Sec 5 pg 3-7 c.
pg 3-8 i
_ -. _
_
_ _.
- _ _. _,
.
.
.
16.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
ANSWER 6.05 (2.75)
a. Undercompensated (0.25)
The effective threshold of the IR high SllR scram protection is raised which allows a greater SUR and hence a higher peak power will be reached prior to reactor scram. (0.5)
b.
At 16 MWe and decreasing contact K3-P closes (0.5)
At 14 MWe and decreasing contact K5-P closes (0.5)
Depressing the Trip Bypass PB energizes Ki-P & K2-P relays (0.5)
The Ki-P contact' closes sealing in the Ki-P & K2-P relays which block the at power scrams. (0.5)
REFERENCE s. Yankee STM, pg 31-26 b.
Yankee STM, pg 33-10, RO LP obj. 101218 & 101211 ANSWER
.6.06 (2.00)
a.
No automatic action CO.5]
b.
Tank discharge valve shuts CO.53 c.
Closure of effluent flow control valve CO.5]
,
c. No Auton.s ti c action E0.53 REFERENCE FSAP: Chap 2t5 Radiation monitoring Svstem.
p 215:3 Cnap 413 Fuel Hz.r.dl i n g A c c i d e nt.
p 215: 4
_____________________-____________________________________________________
KA: p 3.11-1 K1.01 3.4 p.3.11-14, K6.10 2.5 9 3.11-27. EA1.02, 3.1 p 3.11-30, EA2.05, 3.o
___________________________________________________________________________
OBJ: 107302 107202
.
,mg--
ymm ye ww wems--a
'I
.
.
I I
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
______________________________________________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
-ANSWER 6.07 (2.75)
a. The rod group selection lights for group C would go out. (0.3)
A loss of can motor power will cause a relay (MS) to be deenergi:ed.
which opens a contact (MS) in the power _ supply from the battery to group C indicating ligh'ts. (0.7)
b.-An uncontrolled dilution would cause a Tave increase (0.25). A 2-F inc ease in Tave will cause inward too motion to begin on group C rods.
Rod motion begins and stops when Tave decreases to less than setpoint.
(0.5). The cycle repeats itself if the dilution continues. (0.25)
c.
130 MWe (0 25)
Above 130 MWe, rods can only be pulled one step at a time. To pull rods again the manual rod control switch must be allowed to spring return to off and then shifted-back to the rods out position. (0.5)
,
REFERENCE
,
a. Yankee STM, pg 34-11 b. pg 34-16 c. pg 34-16,34-17
,
ANSWER 6 08 (1 50i Automatically starts No. 1 OG (0.5)
Trips incoming supolv breakers f r o m. 4 8 0'J bus 6-3 (0.5)
When We'.
1 DG comes up to rated voltage its breaker closes automatica11v (0.5)
'
REFERENCE Yankee FSAR. pg 22616
.
%
--
-.
.-
.
.
CONTROL, AND INSTRUMENTATION PAGE
6.'_ PLANT SYSTEMS DESIGN,
___ __________________________________________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
.
U i
ANSWER 6.09 (3.00)
'
a.
-Provides LPSI pump start signal (0.2)
-Provides HPSI pump start signal (0.2)
'
-Supplies an open signal and prevents closing SI-MOV-22, SI-MOV-23,
SI-MOV-24, SI-MOV-25, SI-MOV-4 and SI-MOV-46 (0.6)
'
-Supplies a close signal and prevents opening CS-MOV-532 (0.1)
-Trips the A bus tie breaker to any of the three emergency busses if the DG output breaker and the B bus tie breaker to that bus is closed (0.2)
-Trips the B bus tie breaker to any one of three emergency busses if the OG output breaker and the A bus tie breaker to that bus is closed (0.2)
,.
-Starts a timer to energize SOV-46 which will supply N2 to the
'
SI accumulator (0.3)
-Initiates diverse containment isolation (0.2)
!
b.
Level switches close SI-MOV-1 on low level (0.5) and nitrogen is ventec from the SI accumulator (via the accumulator relief valves)(0.5)
REFERENCE a.
Yankee STM, pg 19-60 b.
Yankee STH, pg 19-60
ANSWER e,10 (2.75)
a.
Under -Pressure maintenance pump starts and compresses the air cushion in tne Fressure Maintenance tant raising its pressure, (0.5)
Over-A solenoid relief.alve operates to reiteve air in the air cushton tank.A mechanical relief backs up the solenoid relief
(0.5)
b.
Two detectors in the area must go into alarm (0.25)
This starts the halon timer (0.25)
<
The timer timing out will cause halon to discharge (0.25)
c.
A given battery room's halon system will cause halon discharge in.both battery room (0.5). The switchgear room's halon system will cause halon
<
discharge in both battery rooms in addition to the switchgear t o o n..
(0.5)
,
REFERENCE I
a.
Yankee FSAR, pg 229'2, ROTH pg 5-5 b.
pg 229.7, ROTH pg 5-16 c.
pg 229:7, ROTH pg 5-15
'
,
a v
-
w--ynw--.
.,--.-m-_~n~.-
y,n.~-.-ym-,,-,
w,-
,-e--
- -,--- _ -,-m -e, m,-e,-+--
n--.-.----,
,~,eym
- -,, -
-
.
-.
-
-.
.
.
J 7.
PROCEDURES - NORMAL, ABNORhAL, EMERGENCY AND PAGE
--- sA5i5E55i5AE 55siEBE------------------------
____________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S..
ANSWER 7.01 (3.50)
-Pressurized by a charging Pump.or purification pump via the isolated s.
,
loop fill header
<
-Connected to another loop via the drain header when that loop is cut-in or Pressurized 1.
-Vented to atmosphere via a hose on the pump vent 3 items (0.5) ea.
b. Open loop bypass Open loop Th stop Open loop Tc stop
.Close loop bypass 4 items (0.25) ea.
Trip hhk operating MCPS (0.5)
c.
Stop the Te stop valve from opening and then close it (0.5)
i REFERENCE s.
Yankee OP-2130, pg 3, step 7 6.
pg 5,6 c.
pg 1 ANSWER 7.02 (2.00)
s.
UC evscustion alarm (0.2) anc increasec sudio countrate sign.al (0.2)
A large increase in neutron flun would occur prior to anv significant increase in gamma f l u::. The high neutron count rate would activate the
,
j VC evacuation a l a r n,.
(0.6'.
b.-Initiate emergency baron injection i one charging Pump at 26 gr.n taking
,
(0.5)
suction from the BAHT)
'
c.
3 (0.5)
'
REFERENCE
a. Yankee OP-3117, pg f 1 b.
pg 2 i
i c.
pg 5,6
+
.
I I
J
.
l
- -.. -.. - - - -., _.. -
- _
_,. - _. _, _. - _. -,,. _ -. - -..... _ _ _ -... _ -,.... _..,,. ~ _ -,.. - _ _,. _ -. _. -. _ _ _. - _
.., _.. - -,, _ -..
>,
.
_.
.--
. _ _. _ -.
_ _ _ _. _
.
-.
,
J 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
~~~k b5bLbb55 E'_UUUTRUL
~
~
~~
____________ _______
ANSWERS -- YANKEE ROWE-96/05/20-BARBER, S.
-
1
ANSWER 7.03 (2.25)
a. SUR cf 0.2 DPM can be maintained (-0.0,
+0.2 DPM) (0.5)
No reactivity addition in progress.(0.5)
6..Yes (0.25)
Depressing manual scram button on the control board and observing BK-1 and BK-2 open indication. (0.5)
c.. Power must be maintained less than 1 E-08 amps. (0.5)
REFERENCE a.
Yankee OP-2103, pg 7 b.
pg 3 c. pg 7-
ANSWER 7.04 (2.50)
'
a.
Reactor seram and turbine. trip BFP auto trip (Condensate recirc valve opens)
,
)
Presturiner hesters energire 2 items (0.5) es.
!
Most be operated t r. manual
<0,.e)
E.leed line tr ip valve (CH-LCV-222)
,
Pressurirer Sprav valve (P5-MOV-191)
Solencio relief valve (FR-50V-90)
Solenoid relief valve isolation (PR-MOV-512)
4 items (0.1) ea.
i
REFERENCE I
a.
Yankee OP-3250, pg 3 i
b.
pg 3
1
,
1
,
i I.
!
!
l
I
)
i
_..
__... _ _ _ _ _ _ _. - _ _ _. - _.... _. _, _, _ _. _ _ _ _, _ _. _ _. _. _ -. _ _,. _ _ _. _ _ _ _.. _ _ _, _ _ -,.. _, - - -. -. _... _ _ _
_
.
.
.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
R 5iBE55iEAE E5UTR5E----------------~~------
~~~~
____________________
ANSWERS -- YANKEE ROWE-86/05/20-BARBER, S.
s ANSWER 7.05 (3.00)
50 REM (0.4)
a.
5(N-18)
=
Total lifetime to date = 33 + 1.8 =34.8 REM (0.4)
Total lifetime available = 50 - 34.8 = 15.2 REM (0.4)
3-1.8 = 1.2 REM (0.4)
Total this quarter available
=
Quarterly Limit is more restrictive than annual limit
+ (1.2 REH/HR) neutton = 1.31 REM /HR 0.11 REh/HR gamma 1.0 HR/1.31 REM /HR = 0,.76 HRS = 46 MIN (0.4)
b.
100 REh one time exposure (0.5)
100 REh/1.31 REH/HR= 76 HRS (0.5)
( Full credit given for any dose rate used from part a.)
REFERENCE 10 CFR 20 Yankee Radiation Protection Manual, pg 8 ANSWER 7.06 (2.75)
a.
-Depress manual scram PS anc manual turbine thrcttle trip (0.5)
-If rods fail to move emergencv oorste (0.25), manually open the scram breakers (0.25) and open circuit No. 12 CR0 power supp1v on No. 2 batterv swi tchocar c (0.25)
-
b.
1650 p e, t g Auto to.2) Safet.v Injection (0.3)
1200 psig nanually (0.2) trip RCPs (0.3)
c.
True (0.5)
REFERENCE a.
'rankee OP-3000, pg i b.
pg 1,2 c.
pg 2 ANSWER 7.07 (1.50)
,
s. Autostart of BFP on a scram (0.5)
Cold water accident (0.5)
b.
False (0.5)
.
e
<
=
.
7.
PROCEDURES -'NORMALr ABNORMAL, EMERGENCY AND PAGE
R b ULUU5CIE~EUUTRb[~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
____________________
ANSWERS -- YANKEE R0WE'
-86/05/20-BARBER, S.
'
REFERENCE c. Yankee OP-2255, pgl3 b. pg #3 f
ANSWER 7.08 (2.75)
a. -If MC pressure remains above the shutoff head of the SI' pumps (1550 psis) (0.25) and SI flow i s verifed to be zero (0.25)
-19 feet in the SIT (0 5)
'
b. Rear of main control board (0.25)
SPDS (0.25)
'
2300 F (+100 F,
-100 F) (0.25)
c. Prevent boren precipitation (0.5)
Core overheating (due to reduction in heat transfer from boron plating)
(0.5)
REFERENCE
{
a. -(ankee OP-3051, pg el
'
b.
OP-3053 cover sheet c. 0F-310c. pg 4 FSAF, pg 411:5 ANSWER T.09 43.00)
,
s.
W i t'h i n three minutes CO.53 OR if the bearing temperature increases to l
200 DEG F CO.53.
b.
To ensure that the main coolant flow continues for at least tuo minutes after the plant scramEO.73. Increased (from 180 F to 200F)
CO.33.
,
c.
Air trapped around the upper radial bearing [0.53 will cause the bearings to overheat CO.53.
-
i REFERENCE a.,
b.,
c.
Sin 13-15
-
-
i i
e i
,m.._,----,,.,._,,,-.,c,--,,
~. -,, _ _, -, - -,,
-vm._,,,..,r_.,
,m
,-,- _-, _..,.-.,.
.-.,_.._-e....
--..,,.-.y
, - _, -.,. _, - - -,
.
..
.
. _.
- _..-
.
N
.
.
'
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
-
--- E 5i5t55iEAt E6sTR5t t
____________________
ANSWERS -- YANKEE ROWE.
-86/05/20-BARBER, S.
<
.
ANSWER 7.10 (1.75)
To compensat'e for absorbed oxygen CO.5] and the thermal a.
decomposition of hydrazine. [0.53 i
b.
Less than 250 DEG F dain Coolant System temperature. CO.75]
.
REFERENCE I
0P 2160, p.1
!
.i
!
T
}.
1
>
d
!
.
',
i
d v
'
>
l
<
a k
-
...
.
..
....
..
..
..__ _ _ _.
_ _ _, _ _,
.
.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
_____________________________________________..___________
ANSWERS -- YANKEE ROWE-86/05/;2-BARCER, S.
ANSWER 8.01 (2.00)
.
m.
Yes (0.5)
b.
5-(fire brigade) (0.25)
2(licensed operators-safe shutdown) (0.25)
No,(oper ator s required for safe shutdown must be e>:cluded f rom the fire brtgade) (0.5)
c.
2 (0.5)
REFERENCE c. Yankee, Tech Specs Sect.6, pg 6-1 b.
pg 6-2 c.
pg 6-5, AP-2001, step te ANSWER 9.02 (3.00)
a.
Continue to operate, Less than 4 GPM identified leakage (does not interfere with leakage detection systems)
b.
Shutdown, No pressure boundary leakage allowed c.
Snutdown. Grester than 1 GF h unidentif 2 ed leavage o.
3hutdown, pressuriner safetv valve is non-isolable. No pressure boundary leakage allowed.
50.25) cirection to.5)
evaluation of applicable limit REFERENCE Yankee T.S.
3.4.5.2 and Sect. 1 definitions ANSWER 8.03 (2.00>
Unit must be placed in at least hot standbv in one hour. (0.5)
NRC is informed (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per Tech Specs and 1 hr for E-plan) (0.5)
Safetv Limit Violation Report shall be prepared. (0.5)
.
Report is submitted to NRC and Upper hanagement (within 14 davs) (0.5)
REFERENCE Yankee Tech Specs, Sect.
6.7, pg 6-13
.
l B.
ADhINISTRATIVE' PROCEDURES, CONDITIONS, AND LIhITATIONS PAGE
__________________________________________________________
ANSWERS -- YANKEE ROWE-86/03/20-DARBER, S.
ANSWER 8.04 (2.25)
,
a.
Initiate and continue to emergency borate (0.5) at greater than 26 3pm of 2200 PPh boric acid (0.25) until SDh is restored. (0.5)
6.
(1)
hCS Baron Concentration (2)
Control Rod Position (3)
Xenon Concentration (4)
Samarium Concentration (5)
Fuel Burnup (6)
MCS T AVE 4 items (0.25) ea.
REFERENCE Yankee TS 3/4.1.1 ANSWER 8.05 (1.50)
No CO.33, each test is within 25% of the required time interval CO.63, but the three consecutive combined test intervals exceed 3.25 of the required interval CO.63.
REFERENCE
-
(ankee T.S.
4.0.2 pg 3/4 0-2 ANSWEh 6.0o (2.25i a. Issuin3 durervisor Healtn Fnysics shift Supervisor 3 items (0.25) ea.
b.-Entry into High Radiation Areas-Entrv into areas posted to require on RWP prior to entry-Entry into Airborne Radioactivity Areas-VC entry-Haintenance or inspection of contaminated or radioactive equipment-Fuel Handling-Situations deemed by Health Physics to require an RWP Any 5 items (0.3)
REFERENCE a.
Yank ee Radiation Protection hanual, pg. 27 b.
pg 29
r
.
,
8.
ADriINISTR ATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS -- YANKEE RGWE-86/05/20-BARBER, S.
ANSWER 8.07 (3.00)
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l a. Site emergency (0.25) loss of functions necessary to achieve hot shutdown. (0.5)
'
b.
Unusual event (0.25) Control rod withdrawal at power due to a malfuntion
,
(0.5)
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c.
Unusual event'(0.25) due to transport of contaminated individual. (0.5)
j d. General Emergency (0.25) LOCA an.d o,en non-return valve and a loss of
ECCS (0.5)
P
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I REFERENCE a.
(ankee EAL Classification guide, pg 7 b.
pg 18
c.
pg 24
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d.
P3 11 ANSWER 8.08 (2.00)
]
s. yes ( 'a l l specs are being met) (0.5)
!
b.
no (three trains are needed) (0.5)
c. no (activity > 1.0 microcuries/ml) (0.5)
d.
yes inon-return valve closeo in greater than 5 seconds) (0.5)
s REFERENCE a.
fankee T.S.
Sect. 3.5.4 b.
Sect. 3.6 1 1
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c.
Sect.
3.7.1.4 d.
Sect. 3.7.1.5
.
ANSWER 9.09 (3.00)
a.
SS and any one of the following* PS,APS,POM 2 items (0,5)ea.
b.
The person making-the entry (0.3)
Supervisor of the operating shift (0.3)
Immediately upon being relieved of shift responsibility (0.4)
'
c. Plant Operating Log Refueling Log Book (when applicable)
Special Order Log Dypass of Safety Function and Jumper Control Log Tagging Order Los Deparmental and Interdepartmental Memos 4 items (0.25) ea.
'
REFERENCE
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a.
Tankee AP-2003, pg 2 6.
AP-2007, pg 2 i
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0.-
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE
__________________________________________________________
ANSWERS'-- YANKEE ROWE-86/05/20-BARBER, S.
c.
AP-2002,-pg 3
ANSWER 8.10 (2.50)
_o.4 (0.75)
b.97 (0.75)
c.
-12 neutron detector thimbles operable-2 operable neutron detector thimbles per quadrant-sufficient operable movable neutron detectors, drive, and readout equipnee nt-10 operable radial position thermocouples 3 items (0.33) ea.
REFERENCE a. Yankee T.S. Table 3.3-2 b. * ed Table 3.3-1 C, eut h pp 3-13 ANSWER 8.11 (1.50)
a.
Shift Supervisor CO.75 each3 b.
On1v those persons suthorized on the Local Control Tagging List REFERENCE AP 0017 p.1
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May 21, 1986 RO/SRO Exam Comments
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GENERAL COMMENTS
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The RO exam was generally written well with each question Referenced to the materials provided, the KA catalog and the Learning objectives provided.
The SRO exam was not as well written.
Many instances of incomplete sentences and question. which were very hard to interpret have been noted in our specific comments for the SRO exam.
Additionally the References for each SRO question are generally related to a single source document with no apparent effort to justify questions through the use of the KA catalog or the Learning
, ;.. objectives provided.
In many cases as noted in our SRO specific coments the references denote wrong page numbers from the Source Documents.
It appears that the SRO exam was developed very quickly without proper thought as to the
,
applicability or importance of questions.
Generically, both the RO and SRO exam required the candidates to think through and write out their responses very quickly.
Some candidates were allowed extra time to finish.
Unless the candidates were told they would be allowed this extra time at the commencement of the exam, they would be forced to rush through each question.
The added stress created may result in the candidates giving answers which are not as complete or as well thought out as they are capable of providing.
The exam should have been developed requiring fewer responses.
The comments provided with this write up represent initial exam review observations by the Training Staff of this facility.
Further comments will be provided to the chief examiner or regional office section chief along with supporting material within five working days as allowed by es 201 attachment 4 of NUREG 1021.
Attachment 1.
RO examination Specific Comments Attachment 2.
,SRO examination Specific Comments Attachment 3.
Supporting Materials
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examination. Specific Comments
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Section I
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Question 1.06 A:
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Our operators have access to a Power Defect curve located in the YAEC Core Data Manual page C-8.
This curve is developed using Moderator and Fuel Temperature only, not Void co-efficient.
We use the Westinghouse Theory books as reference materials, but our objective is to teach the students to discuss and use the Power defect curve.
Answers to this question may assume that we do not use the
,;<. Void co-efficient as modern plants do because we have insignificant voids in our core and do not consider them.
Ref.
YAEC Core Data Manual - page C-8.
Question 1.06B:
.
Below 510 MWT we hold a constant Tave, Theref' ore, the power defect below 510 MWT is comprised of the Doppler Defect only
-
at 510 MWT.
The Turbine control valves should be full open and we then start increasing primary system temperature till we get to 600 MWT.
As we increase Tave, the moderator temperature defect must be added to the Doppler defect to make the power defect curve.
That is why the power defect curve increases rapidly above 510 MWT.
_
Ref.
YAEC Core Data Manual - Pages C-5 - C-8.
Question 1.07:
s States you have lost core cooling.
The reference quoted in the answer Key is for an interruption of natural circulation.
These are two different scenarios.
The answer Key for 1.07 B-2 is also quoted wrong.
It should be " increase" not " decrease".
This question requires an answer that is based on assumptions used in reference material and those of the examiner.
.
Any answer that relates to abnormal conditions in the reactor and steam generator should be accepted as an indication of a loss of natural circulation.
Example -
B-1.
Delta T across core - should increase, but the answer Key is looking at Temperatures in the loops.
If there
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B-3.
Steam pressure in a well insulated steam generator does not decay off rapidly, and will not decrease unless your steaming off.
.
Question 1.08B:
,,s.
If there is a 10% power mis-match between primary and secondary power with the Turbine leading, the reactor must be assumed to come up on power.
How does it do this, by decreasing Moderator Temperature enough to add the positive reactivity necessary to bring reactor power up.
The power increase will be stopped by the negative addition do to doppler.
How far Tave decreases will be dependent on the
!
power mis-match and time of core life.
The answer Key as a final Tave higher than the original.
This is not possible if the moderator must decrease temperature to add position reactivity.
The candidates have been taught that the final steam pressure will be dependent on the final Tave.
This is calculated by using the MTC and the doppler defect to the increase in power.
Figure 5 provided with the question really doesn't fit the question asked.
Ref.
YAEC Core Data Manual - pages C-3 - C-7.
Question #1.09 Answer Key Integral rod worth curve was read in correctly, for 90" to 45" inserts 700 PCM not 1200-1900 - 90"
'
-1200 - 45"
- 700 PCM Power defect was redd incorrectly -
950 PCM at a boran 100 PPM is closer 225 PCM at 180 MWT 30% power
+725 PCM Xeron defect was read incorrectly 3100 PCM at 10 0 r. power 1980 PCM at 30% power
+1120 PCM change
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Section 2 2.02.A.
The Answer Key is correct.
However, the question
.
could be interpreted as asking for how HV-SOV-1 is operated when using the Post Accident Hydrogen Vent System during a Large Break Loss of coolant Accident.. This procedure is described in OP-2658
"g.
" Operation of the Vapor Containment Hydrogen Control Monitoring System and Sampling Systems.
2.02.B.
Deleted during the exam.
2.06.A.
No Comment B.
The safety injection pumps can also feed the Steam Generators through the emergency feedline spool piece.
Ref. STM: Chaper 3 - Page 3-93 2.07 This question could be answered by a one line diagram showing the Major components of figure ?'-7 in STM ch 37 and figure 40-7 in STM ch 40.
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2.10.A.
The question wording is confusing - what's accumulator?
What is primary?
Probably would be interpreted correctly.
B.
The answer Key is not accurate.
SOV's do not admit N2 to the accumulator.
They bleed N2 from the trip valves allowing them to open.
2.llA.
The answer Key is correct if No. 2 changing pump is the only operating pump.
If No. 1 or No. 2 changing pumps are operating Level will be maintained at 120" by the changing pump speed control as described on
.
page 15-21 and page 15-22 of the STM Ch. 15.
Section 3 No Comments
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SRO EXA?i1NATICN SPECIFIC COMMEBTUS
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Question 5.02 A ~
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h is question seems to be inappropriate for this facility in that Tech Specs states that we will never have a positive M.T.C. Ref. Tech. Spec.
3.1.1.4.
Answe2: to 5.02A in the answer key states less positive, Paragraph 3, pages 3-14 Reacter Core control indicates the answer to the question may be nore positive. In any case,the question requires an assumption on the part of the examines to determine where on the H 0/ Fuel ratio curve the question is
refering to because the slope af the line does change.
Reactor Core Control Pages 3-14 + 15
, Figure FND-RF-260
,
Section 5.
Oestion 5.03 We question is written with an incomple.te sentence. What rod soecd in start-up #1 was ccrupared to what rod speed in start-up #2.
We reference should be Fundamentals of thiclear Reactor Physics not Reactor Core Control for Large P.W.R.
The page numbers referen d were correct for the fundamentals book.
Question 5.03D If on a start-up rate the reactor is super critical. We answer key reference doesn't relate the rate at which the reactorisbrought critical to a start-up rate.
_
Section V.
Question 5.08 C This question requires assumptions to be made by th'e examinet. %e reference quoted does not substantiate the answr in the answer key. We examiner made an assumption that holds true for only one plant condition out of a n infinite number of possibilities. On a nonnal natural circulation cool
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down the Delta temperature is dependent on the cool down rate and-previous operating history and time after shut down that the cool down was ccmnenced.
Question 5.09
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the question at least.twice, because he-must make'an assumption on the
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position.of the Test - normal - bypass switch.
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2 e question also did not provide'enough information. he candidate vxuld have to assume a VC pressure df greater than or less than 5 psi. We correct answer will depend on what is assumed, the anser key is incmplete. Cmplete asnwers can be found usinpy!ie same reference.
Question 6.02
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%e answer key is not cmplete. he safety valve which relieves at
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2560 psi is not included. W e refernce provided with the answer key is spfficient.
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Question 6.03 b-Answer key does not describe the charging flow - it describes an
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erroneous feedwater flow -Ref.
Attachment H of 3203 Ioss of Feedwater.
c - The answcr h y should state that normal changing flow is maintained with the SI pumps providing flow to the steam generator blowdown header.
Ref: Loss of feedwater OP 3203, Attachment H Question 6.04 a - Question asks for 3 conditions that cause NRV's to close autmatically.
Answer key lists two, the answer key is correct. %o responses are sufficient for full credit.
Question 7.0L4
'
his is not vital Information to have mmorized. Any relationship to the KA catalog, NUREG 1122, could not be readily found.
Question 7.01C he answer key quotes incorrectly fran the reference. Not all main coolant pumps are tripped.
.
Question 7.02 A In the second part of this question the answer should not address the ganma guard. We ganma guard is not for criticality but for exposure due to other circumstances such as fuel too_close to the surface of the water or fuel damage. % e referance is on page 2 of the procedure not page 1 as indicated in your answer key.
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Question 7.04 A
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Question 7.04 B Point value is wrong in answer key.
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The References are reversed - Part A is page 13 Part B is page 3 7.08.A Reference should be page 2 Not page 1.
7.09.B The question does not make any sense and should be deleted.
7.10.A" This is a good question for a chemist.
There is no reference to the.KA Catalog for this Knowledge.
,<-
There is no reference to our learning objectives.
There is no bases for an operator having to memorize this knowledge.
This question should be deleted.
7.10. B The answer provided is correct, however a more detailed answer is included in the R.O. Manual page 4-18.
8.01.C This question is confusing, because by Tech. Spec.
- 3 is the correct answer by AP-2001 #2 is the correct answer.
The answer Key should accept either answer.
8.03 The answer Key is incomplete.
Additionally the States must be notified in 15 min. for the
"E" Plan.
Also, the answer key goes beyond what is asked by the question by requiring greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limits.
8.04 In the question " Set Points" are asked for.
This is confusing because the candidate is already given the initiating circumstances.
8.06 The answer Key should include the signatures of all people on R.W.P.
Ref. OP-8415 - page 4, Item 4.
8.07 The candidates should have been given all of Attachment
"A" of 3300.
This is a crucial part of
,
correctly classifying the emergency.
This required the candidates to think through 22 pages of EALS.
8.08.A.
The sentence structure of the question may confuse
'the candidate.
B.
It is hard to understand what is being asked for by
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8.09.B.
It is unclear what logs are being asked about by this question.
8.10.A.
This is a poorly worded question and it is requiring that operator to memorize Table of Notations out of Technical Specification.
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B.
Answer Key is wrong, it should be (4).
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Ref.
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Table 3.3-1 Page 3/4 3-2.
C.
Reference in Answer Key should be Page 3/4 3-23 of Tech. Specs.
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'2.11'A This comunent shotild read, If No.1 or ~3are operating '...
6.01 C This conunent has a typo: Should read v.C. pressure of greater or less than 5 psig.
7.04 A This comunent has a typo: Should read same question as
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1.10 A Answer Key should be " low" temperature not "high" temperature.
See attached graph located in YAEC core data manual. The MTC may be positive at lower temperatures but will go negative at higher temperatures.
The reference given in the Answer Key does not addrcss the question asked.
.
See Westinghouse Reactor Core Control for large PWR.
Pages 3-12 through 3-23.
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1.10 B Part 2 The question is confusing. The examinee must make an assumption that boron concentration will, remain constant over core life.
This is the assumption used in the Answer Key which doesn't apply in real life.
In real life you dilute out boron as the core ages; this makes the differential boron worth increase.
.
See the graph on page 5-14, Westinghouse Reactor Core Control for large PWR.
This shows the differential boron worth beccming
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more negative or increasing over core life at a given temperature.
Also, see attached graph of inverse boron worth (IBW) that is located in the YAEC technical data manual.
Yankee assumes that IBW =
Mff dd hm M5
Therefore, differential boron worth =
using W s IBW logic the differential boron worth increases as the core ages.
See also the attached lesson plan, page 2, we use to teach with.
The answer should be " increases" not " decreases" over core life.
2.11 Point values equal 2.5, should equal 3.0.
3.10 Point values equal 2.75, should equal 3.0.
4.10 Point values equal 2.0, should equal 3.0.
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6.07 A The question specifically asks about rod position indication system but the Answer Key refers to group selection lights; two very different things. Rod position indicating lights get power from distribution cabinet
"A" (Ref. STM, pg. 34-6).
Also, the question did not ask what relay would act to cause lights to go out but the Answer Key indicated this is what the examiner wanted.
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We still feel the operators should not have to memorize relay numbers, nor does the NRC according to Examiner Standards.
Refer to ES 202 items 6 & 15.
The SRO examination was very long; requiring 114 responses.
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