ML20151Y197
| ML20151Y197 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 08/17/1988 |
| From: | Howe A, Jordan M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20151Y130 | List: |
| References | |
| 50-373-88-01OL, 50-373-88-1OL, 50-374-88-01OL, 50-374-88-1OL, NUDOCS 8808260272 | |
| Download: ML20151Y197 (200) | |
See also: IR 05000373/1988001
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U. S. NUCLEAR-REGULATORY' COMMISSION
REGION III
Reports No. 50-373/374-88-01(0L)
Docket Nos. 50-373/50-374
Licenses No. NPF-11/NPF-18
Licensee:
LaSalle County Station Units 1 and 2
Examination Dates:
April 25-29, 1988
Chief Exaininer:
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Allen G. How '
nior 0 rations
Dat(/
Engineer, Re
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Approved By:
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Michae[/ ordan, Chief, Operator
Date
Licensin Section No. 1
SUMMARY: Written and operating replacement examinations were. administered
to five senior reactor operator (SRO) candidates and to three reactor operator
(RO) candidates. All candidates passed these examinations.
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8808260272 880818
'7'
ADOCK 05000373
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DETAILS
1.
Examiner
A. Howe, Chief Examiner
T. Fish
S.. Hare
R. Miller
M. Sullivan
2.
Exit Meetina
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At the conclusion of the site visit, the examiners met with facility
representatives.
The following personnel attended this e.It meeting.
Facility Representatives
W. Huntington, Services Superintendent
J. Renwick, Production Superintendent
P. Manning, Assistant Superintendent Technical Services
S. Harmon, Operations Training Group Leader
T. Shaffer, Training Supervisor
R. Raguse, BWR Supervisor, Production Training
T. Hammerich, Technical Staff Supervisor
R. Weidner, Simulator Operator, Production Training
M. Okopny, Lead Licensing Instructor
J. Borm, Quality Assurance
M. Harper, Quality Assurance
J. Settles, Regulatory Ar,surance
NRC Representatives
A. Howe, Chief Examiner, Region I
T. Fish, Examiner, Region I
S. Hare, Examiner, Region III
R. Kopriva, Resident Inspector
R. Miller, Examiner, donalysts
M. Sullivan, Examiner, Sonalysts
The following items were discussed during the exit meeting:
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a.
No generic training weakness was noted during the administration of
the examination.
However, the following generic strengths were
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observed during the examination administration:
(1) Teamwork and communications.
(2) Use of alarm response procedures and abnormal procedures.
(3) General control board awareness,
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b.
The marginally satisfactory quality of the information submitted by
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the licensee for examination preparation was discussed.
Details are
provided in Attachment 1.
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c.
Simulator fidelity problems slowed the examination and provided
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ambiguities with the actual control panels.
Details are provided in
Attachment 2.
3.
Examination Review
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Responses to licensee's comments, concerning the written SRO and R0
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examinations are provided respectively in Attachment 3 and Attachment 4.
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Attachments:
1.
Quality of Information
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Provided for Examination
Preparation
2.
Simulation Facility Fidelity
Report
3.
SR0 Examination Comments
and Resolutions
4.
R0 Examination Comments
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and Resolutions
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ATTACHMENT 1
QUALITY OF INFORMATION PROVIDED FOR EXAMINATION PREPARATION
1.
System lesson plans were found to be inaccurate and incomplete.
Examples
of those items are provided below:
Item
Example
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Inaccurate
Service Air System lesson plan gave a detailed
description of a mode of operation available but
not implemented at the facility (Question 3.10).
The Simulator Malfunction Book erroneously stated
that a rod overtravel would produce a rod drift.
Incomplete
The logic for the Primary Containment Isolation
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System (PCIS) was not provided in the lesson plan.
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2.
Reference for Question 3.10 was not revised to reflect current
operational conditions, although specifically addressed in the 1986 NRC
Examination Report.
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3.
The following examination preparation materials were not provided to the
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examiners even though they were available and normally provided.
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a.
Lesson plans for the "principals of integrated reactor operations."
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b.
Administrative procedures, such as:
(1) Station and Operations Department organization.
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(2) Handling and shipping of radioactive material.
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(3) Radiation Work Permit procedure to supplement the Radiation
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Protection Standards sent.
(4)
Independent verification.
(5) SCRAM reduction.
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(6) Responsibility for signing records.
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(7) Degraded equipment logs.
(8) Key control.
(9) Defeated annunciators.
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(10) Transport of heavy loads over the fuel pool.
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(11) Housekeeping.
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(12) Fire protection and equipment.
(13) Conduct of surveillance tests.
(14) Summary of all procedures available at the facility,
c.
Operating procedures were incomplete and the specific missing
procedures were too numerous to list.
In many cases, operating
procedures for whole systems were not provided and in others only
one aspect of the operation was provided.
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d.
Surveillance procedures were too few (e.g., only four).
In general,
a surveillance procedure should be provided, if the simulator can be
used to perform the surveillance.
e.
Piping and instrument diagrams; and single line elementary logic
diagrams were not provided to supplement the diagrams provided in
the lesson plans.
In retrospect, submittal of this infotmation may
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have prevented some of the facility examination comments made for
April 1988 Replacement Examination.
4.
Inadequate tabbing made it difficult to locate information in the
following materials:
a.
Reactor theory lesson plans.
b.
Thermodynamics lesson plans,
Abnormal procedyres.
c.
5.
The learning objectives for lesson plans were generic.
Typically,
16 similar learning objectives were provided for each system.
In many
cases, the lesson plan objectives were "nonobjective" (e.g., "There are
no instruments for . . . system.").
Also, the lesson plan objectives
required the student to "discuss" a topic rather than learning specific
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required operator knowledge.
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ATTACHMENT 2
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee:
Commonwealth Edison Company
Post Office Box 767
Chicago, IL 60690
Facility Licensee Docket Nos.:
50-373/50-374
Facility License Nos.:
NPF-11/NPF-18
Operating Tests administered at:
LaSalle Simulator
Operating Tests Given On:
April 27, 28, 29, 1988
During the conduct of the simulator portion of the operating tests
identified above, the following apparent performance and/or human
factors discrepancies were observed. These discrepancies are categorized as
planned malfunction discrepancies and unplanned discrepancies.
Planned malfunction discrepancies:
1.
When a failure, high, of a RBM was inserted, the only effect was the
recorder een moving upscale. An alarm and rod clock should also
occur.
(Reference malfunction No. 20)
2.
When a shaft seizure of both recirculation pumps was inserted (to
simulate effect of a dual recirculation pump trip), flow dropped in
loop A but did not decrease in loop B. (Reference malfunction No. 197)
3.
When a recirculation pump failure to downshift was inserted and
automatic downshift conditions were present, the recirculation pump
downshifted. (Reference malfunction No. 202)
4.
When a RCIC system reduced capacity was inserted, the malfunction
produces a turbine trip when the system is secured. When RCIC was
started for a test and the malfunction was inserted a trip again
occurred. (Reference malfunction No. 71)
5.
When a failure of jetpump #10 was inserted and a reactor scram
occurred, the simulator lost fidelity causing reactor vessel pressure
to exceed 4000 psig almost instantly. When this occurred, the
scenario was stopped prematurely and the candidates escorted from the
simulator. (Reference malfunction No.199)
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6.
When an APRM C downscale malfunction was inserted, both the upscale
and downscale lights were lit at the backpanel. (Reference
malfunction No. 13)
7.
When re3ctor feed pump B flow transmitter fails high was inserted
then removed (as if the instrument had been repaired), the panel flow
indication returned to nermal but the input to the feedwater flow
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control system remained fansed, thus preventing the candidates from
placing feedwater control an automatic.
(Reference malfunction
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No. 139)
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Unplanned discrepancies:
1.
The reactor operator placed the mode switch to SHUT 00dN to cause a
scram. The reactor did not scram. Later the training staff found a
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wire to the mode switch broken which was temporarily repaired.
2.
When a scenario was begun, a rod block was active for no apparent
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reason. This delayed the examination since rod movement was planned
as a part of the scenario.
3.
RHR was placed in suppression pool cooling in preparation for a test.
No heat was being added to the suppression pool, yet high temperature
conditions were recorded on the Div 1 instruments and recorder. This
interfered with the progress of the scenario since the candidates
spent time investigating the cause of the high temperature condition.
4.
The RBM B recorder pen failed downscale.
5.
At full power conditions, the core plate d/p recorder read 0.
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6.
During a scenario when the generator was to be synchronized to the
grid, the GEN #1 PROT RELAY TRIP alarm was in for no reason, Later it
alarm cleared for no reason. This alarm delayed the scenario since
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the candidates were trying to find its cause,
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7.
The C APRM drifted between 20% and 40% when actually at 40% power.
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The fidelity of the simulator is considered poor because of the significant
number of simulator performance discrepancies relative to the sample size.
This is of concern because of the adverse effects experienced during the
conduct of these examinations and the potential impact on future examinations.
Poor simulator fidelity can also adversely affect the quality and
effectiveness of replacement and requalification training. The licensee should
review the performcnce of the simulator in order to improve fidelity.
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ATTACHMENT 3
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SENIOR REACTOR OPERATOR (SRO)
EXAMINATION COMMENTS AND RESOLUTIONS
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General observations based on examination performance are:
1.
SR0s demonstrated a good knowledge in the following areas:
a.
Response of reactivity coefficients to change core reactivity during
olant transients.
(Question 5.01)
b.
Physical arrangement of the Main Steam Line Radiation Monitors and
the Rod Sequence Centrol System.
(Question 7.04)
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c.
Methods to obtain suppression pool temperatures and reasons why
computer points are used when pool level is low.
(Question 7.04)
d.
Operator actions per LaSr;ie up9 rating Abnormal (LOA) Procedures.
(Question 7.08, 7.09, and ,'.10)
e.
Guidelines for working hours.
(Question 8.03)
2.
SR0s demonstrated a poor knowledge in the following areas:
a.
Reasons why vessel and pressure change in response to a
recirculation pump trip and how the critical power ratio c.'1anges as
a result of a CI.ange in T0 Circulation flow.
(Question 5.04
and 5.06)
b.
High Pressure Core Spray (HPCS) pump suction valve response to an
initiation signal when the condensate storage tank (CST) suction
and the suppression pool valves are closed.
(Question 6.02)
c.
Ability to state the required checks to verify that an emergency
diesel generator is properly operating in response to a loss of
coolant accident (LOCA) signal without associated bus undervoltage.
(Question 7.05)
d.
Responsibilities of the Station Director, which I.iay not be delegated
per the Generating Station Emergency Plan (GSEPO)and the bases for
average power range monitor (APRM) f1 a biased rod block setpoints.
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(Question 8.05 and 8.0ii
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Comments resulting in significant changes to the master answer key, or
comments "not accepted" by the NRC, are listed and explained below.
Comments
made that were insignificant in nature and resolved to the satisfaction of
both the examiner and the licensee during the post exam review are not listed
(i.e., typographical errors, relative acceptable terms, minor setpoint
changes, etc.).
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LASALLE SRO EXAMINATION OF APRIL 1988
5.02
Facility Comment:
The answer key for this question only considers the change
in voids.
It should be noted that the situation described
in this question is a complicated result of two factors -
voids and fuel temperatures.
The doppler coefficient
actually becomes less negative as the fuel temperature
increases.
This is to say that at 40% power the fuel
temperature is at a lower temperature than at 50% power.
What the graph, Figure 50, in the reference cited below
shows is that as void percent increases, the doppler
coefficient becomes more negative and that as fuel
temperature increases the doppler coefficient becomes less
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negative. Without precise numbers the actual change in
the doppler coefficient (e.g., the delta k/k per 1 F of
fuel temperature change) will be hard to predict.
Proposed Resolution:
Either approach to answering this
question should be accepted for full credit.as long as the
examinee shows adequate undarstanding of the theory
concepts involved.
This question should be written so
that it is limited to one factor (e.g., just discuss
doppler coefficient changes with respect to fuel
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temperature or with respect to void fraction changes).
Reference:
Reactor Theory Lesson Plan, Rev. 2, P.168.
NRC Resolution:
Comment is accepted.
Answer key is modified.
5.04a
Facility Comment:
(1) The answer to part a of ;his question really tells
the pressure effects which will be seen on a longer
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basis than one may interpret the question to ask.
Immediately after the trip there may be no noticeable
effects on the pressure in the reactor.
The effects
listed occur as the tripped pump coasts dcwn and would
not be noticeable until several seconds have elapsed.
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(2) The answer key currently has a lot of details such as
void action, power reduction, steam reduction, steam
flow reduction, frictional head loss changes, EHC
response, the location of the pressure sensor for the
EHC (electro hydraulic control) system, etc.
The
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examinee does not need to go into this much detail in
order to show basis understanding of what happens and
why.
Resolution:
(1) If immediate affects are assumed and
the examinee responds in the fashion
mentioned in (1) above, he should be given
full credit.
(2) Full credit should be given without all
the mentioned details being written down.
References:
Answer Key Question 5.04a; General pump
characteristics.
NRC Resplution:
Comment is partially accepted.
The candidate does not
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need to provide all the details of the answer key but
should be able to discuss the basic concept as to
why reactor pressure changes.
Answer key modification
is not required.
5.04c
Facility Comment:
The answer to part c of thi! yetstion centers on reverse
flow through the idle loop after the pump trip.
This
reverse flow will not be an immediate action due to the
pump coast down after it trips.
The flow in the active
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loop will increase due to the fact that the two pumps are
effectively in parallel feeding a common header.
In this
situation when one running pump's flow is reduced the
other pumps flow increases due to reduced backpressure.
Resolution:
This answer shoula also be acceptable for
full credit.
NRC Resolution:
Comment is accepted.
Tne facility's comment of reduced back
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pressure applies to both the pump coast down period and to
the short cycling of the core due to reverse flow through
the idle loop jet pumps.
Answer key is clarified.
5.05a
Facility Comment:
Our plant is designed with an automatic recirculation flow
control valve runback which occurs if less than
two feedpumps are running and water level is low (at the
low alarm point).
As the question is written, thi- is a
viable alternate answer.
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Resolution:
Flow control valve runback should also be
accepted for full credit.
Reference:
LaSalle Systems Chapter 6 Recirculation Flow
Control System Rev. 2.
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NRC Resolution:
Comment is accepted.
Since the candidntes would not be
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able to determine whether RPV level is above or
below 31.5 inches at 7 seconds on Figure 2, the
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FCV runback will be accepted as an alterr. ate aaswer
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for full credit.
Answer key is modified.
5.05c
Facility Comment:
For part c of this question another accurate way to answer
would be to discuss main turbine performance after a
In the scenario set up for this question, the
turbine will not trip immediately.
Instead, the turbine
will continue to draw off steam until it trips on reverse
power.
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Resolution:
This answer also should be acceptable for
full credit.
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NRC Resolution:
Comment is t.ccepted.
Answer key is clarified.
5.05e
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Facility Comment:
On the figure given to u ' examinees for this question
(Figure 2] the stable steam load shown is closer to
35% than our 25% bypass capability.
If an examinee
introduces a steam line break or other possible failure
into his answer, the answer should be acceptable for
full credit.
References:
Figure related to Question 5.05; and LaSalle
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Systems Chapter 21, Main Steam (shows 25% bypass
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capability), Rev
4.
NRC Resolution:
Should Figure 2 mislead a candidate to suspect total steam
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flow is higher than that for normal bypass flow following
a reactor scram, no penalty will be imposed for attempting
to explain this assumption.
Answer key modification is not
required.
5.06a(2)
Facility Comment:
In part (1), the answer key includes mention of
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increased "heat removal rate at the clad surface" when
inlet subcooling exists.
The issue here is that colder
water takes more heat energy to reach the onset of
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transition boiling, and, therefore, the critical power
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will increase.
An answer which demonstrates a good
understanding of the concept here will not necessarily
have to discuss clad heat removal rate.
Resolution:
Mention of clad heat removal rate should not
be required for full credit.
NRC Resolution:
Comment is accepted.
The answer key was intended to
address the concept that more heat / energy can be removed
from the clad surface or that more heat / energy can be
absorbed by the coolant.
Answer key is clarified.
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5.06b
Facility Comment:
The referenced TP0 does not require the operator to
discuss how the critical power ratio (CPR) changes with
various parameters (see Attachment 5).
In addition to
this, the Knowledge and Abilities (K/A) Catalog
(NUREG 1123) reference for this question applies only to
the first part of this question.
Nowhere in the K/A
Catriog are the operators required to deal with the
complicated issue of the change of CPR with flow,
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pressure, etc. The question also references the GE BWR
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Series on Heat Transfer and Fluid Flow Chapter 9.
We do
not use this specific book, but have it for reference.
However, there is nothing in this book or in the material
which we teach from which would answer this question.
The
issue is complicated by the fact that CPR equals criticul
power divided by the actual bundle power; and both the
numerator and the denominator vary with changes in
recirculation flow.
Thus, a generalization such as is
given in the answer may not always be true depending on
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power levels, rod patterns, peaki-a factors, etc.
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Resolution:
On the basis of the-,
.41derations, this
portion of the question should be aropped and the point
totals adjusted accordingly,
References:
Core Thermal Lesson Plan Objectives; and
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K/A Catalog, pp. 6.2-13, and 6.2-14.
NRC Resolution:
Comment is not accepted.
1.
The Terminal Performance Objective Number 5.f requires
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the operator to discuss the function of K corrections
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to the MCPR LCO.
In order to discuss the function of
K , the operator must understand the concept that for
ibereasedrecirculationflowbundlepowerincreases
faster than critical power.
(Refer to the Lesson
Plan on Core Thermal Hydraulics Page 37.)
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2.
The BWR K & A Catalog requires the operator to
define CPR (K/A No. 293009 Kl.18) and to explain
the basis of the limiting condition for
CPR (K/A No. 293009 K1.19 and Kl.27).
To
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Answer 5.06b the operator must define CPR and then
explain the basis for raising the operating
limit MCPR such that safety limit MCPR is not
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violated by a recirculation flow increase.
(Again,
refer to Lesson Plan on Core Thermal Hydraulics
Page 37.)
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In summary, the operator must understand how CPR changes
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with core flow in order to understand operating limit
MCPR and to enable compliance with Technical
Specifications 2.1 (Safety Limits) and 3.2.3 (Minimum
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Critical Power Ratio).
Answer key not modified.
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5.08
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Facility Comment:
This question has two problems that make answering it
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-difficult.
The first is that the question does not make
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it clear what the answer should look like.
Should the
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examinee list 1 through 8 and match one or more letters
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next to each number as his answer? Or should he list J
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through d and then use one (or more) numbers as his answers?
If a matching question is not in two parallel columns, it
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is very common for the matches to be listed at the top.
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The reverse is true in'this question.
To make the
examinee know exactly what is expected of him, the format
should clearly indicate what is to be done.
For example,
to elicit the response desired on the answer key, the
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question should have a line in front of each lettered item
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such as "
a."
This would reduce confusion on how to
set up the answer conceivably, an examinee could answer
in the format of:
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1.
c(d).
2.
d.
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3.
c.
4.
a.
5.
b.
6.
b.
7.
b.
8.
a.
To do this, the answers to 1, 7, and 8 are strained at
best (for example, MCPR deals with OTB not stable film
boiling), but the second problem with the question makes
things worse, not better.
The examinee is tolo to choose
the best answer for each and that "a" through "d" nay have
more than one answer.
The examinee who has started to
setup his answer as shown above will then choose the best
one from a to d and feel free to use any letter as often
as he needs.
These two problems are additive.
At least
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one student asked for clarification on this question
because he was unsure of what was wanted.
No clarification
was made to 'he whole class.
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Resolution:
If an examinee did answer in this way, then
the question should be graded in a fashion so that he/she
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gets full credit for right answers and is not penalized
for attempts to find a "best" answer just because the
question is written in a confusing manner.
Reference:
Answer Key for Question 5.08.
NRC Resolution 1
Comment will be considered on a "case-by-case" basis such
that a candidate is not unduly penalized by the
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construction of the question.
Answer key modification
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is not required.
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5.09b
Facility Name:
There is a difference between the answer key and the
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question. The question asks for reactor pressure
"increases", but the answer key says "decreases".
The
answer key should be corrected to say "increases".
This
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affects the "available NPSH" answer which should say
"increases".
The second column for Required NPSH" is
not changed.
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Recommended resolution:
Change the answer key to read
"reactor pressure increases".
Also, the first answer to
part b should be "increases", not decreases.
Reference:
Answer key to 5.09b.
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NRC Resolution:
Comment is accepted.
Answer key modified to correct
typographical errors.
5.11b
Facility Comment:
There are other equally acceptable answers to caplain how
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control rod density affects the void coefficitnt.
If the
examinee shows adequate understanding of the physics of
the core without mentioning, for example, changing the
size of the core, he should not be penalized.
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Resolution: Accept d{scussions of the CRD density affects
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on the void coefficient which do not talk about core size
for full credit.
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NRC Resolution:
Comment is accepted.
The intent of the question is to
ascertain whether the operator understands the concepts
associated with the thermal neutron physics of the core,
not the ability to memorize specific phrases.
Answer key
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is not modified.
6.01b
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Facility Comment:
The response provided in the answer key is incorrect,
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Referring to LaSalle Systems Description, Chapter 14,
Page 28, an APRM flow unit failing upscale will result
in a R00 BLOCK due to an upscala trip (108%) or a
comparator trip (10% aflow).
Resolution:
Change the answer to RODBLOCK instead of
NO ACTION.
References:
LaSalle Systems Description, Chapter 14,
Page 28,
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NRC Resolution:
Comment is accepted.
Answer key is modified.
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6.02a
Facility Comment:
The valve line up given in condition #2, CST suction valve
shut and suppression pool suction valve shut, is an
abnormal lineup and would not be allowed due to procedural
requirements.
Therefore, the trainee should not be required
to memorize an interlock arrangement for an improper valve
lineup.
Resolution:
Delete condition #2 of Question 6.02, part a
with an appropriate reduction in point value.
NRC Resolution:
Comment is not accepted.
The intent of the question is to
examine the candidates understanding of the HPCS automatic
initiation interlocks.
Although this valve lineup is
abnormal, it may exist while swapping the HPCS suction
path between the CST and the suppression pool.
Furthermore
the operator is required to be able to monitor the
automatic operation of the HPCS system valves.
(BWR K & A Catalng 209002 K/A No. A3.01).
Question retained; answer key modification is not required.
6.04b
Facility Comment:
No correct alternative is provided.
The correct answer
would be that level will initially decrease then return to
its initial value of 36 inches.
This occurs due to the
response of the dynamic compensator.
Resolution:
Delete part b of this question with an
appropriate reduction in point value.
Reference:
LaSalle Systems Description, Chapter 31,
Page 28.
NRC Resolation:
Conment is partially accepted.
Choice Number 1 or choice
Number 2 will be accepted as a correct answer since the
selection of either will demonstrate the candidate's
knowledge of the initial response of the feedwater level
control system for a steam flow transmitter failure
downscale.
A comment will be made on this question to
prevent future use with incomplete choices.
Answer key
is modified.
6.08a
Facility Comment:
Since the question does not specify with or without
operator action, depressing the substitute position select
pushbutton would be a correct answer.
Resolution:
Add "Depressing the substitute position
select pushbutton" as an acceptabic answer.
References:
LaSalle Systems Description, Chapter 19,
Page 12.
8
i
- -
-
-
-
i
-
.
.
.
NRC Resolution:
Comment is accepted. Answer key is modified.
6.10a
Facility Comment:
Answer #6, SBGT auto start pushbutton depressed, is
incorrect.
This will only start the affected SBGT train;
it will not result in a Raactor Ventilation (VR) System
isolation.
This'is not identified as an isolation in
either reference cited below.
Resolution:
Change the number of required responses to
"four (4) of the five ISOLATION SIGNALS..." with an
applicable reduction in point value.
References:
LaSalle Systems Desc"iption, Chapter 49,
PP. 19-20, and Chapter 60, P. 16.
NRC Resolution:
Comment is not accepted.
The answer to LaSalle Examination
Question Bank Number 01-020226 (Section 6 Page 32) states
that depressing the SBGT auto start push buttons will
provide an isolation signal to the reactor Building
Ventilation System.
No logic diagrams were provided to
show that this action is incorrect.
Answer key not
modified.
6.10b
Facility Comment:
The point value seems too heavily weighted i.e., its worth
0.25 points to identify a parameter which will cause a
VR isolation, yet for identifying the one parameter that
will isolate VR, and not start SBGTS, 0.5 points is given.
The point value doesn't seem to reflect the importance of
test item.
'
Resolution:
Reduce point value of part b.
NRC Resolution:
Although the answer to part b is a single signal, it is
.
the one Reactor Building Ventilation System isolation for
i
which the operator must be aware that SBGr will not
automatically start and therefore impacts the operator's
ability to maintain / verify secondary containment
per Technica* Specification 4.6.5.1 due to a decrease /
loss of secondary containment to outside air differential
pressure.
Answer key is not modified.
6.11b
Facility Comment:
Two conditions are described in the reference cited which
can result in a SELECT ERROR.
The first condition:
" ...whenever the operator selects a rod that is not in the
currently latched group..." corresponds to alturnative 2
of the question.
The second condition: "... selects a rod
other than the one responsible for an insert or withdrawal
block...," corresponds to alternative 3 of the question.
Therefore, alternative 2 or 3 is a correct response.
9
_
-
-
.
..
[
-
,
,
,
.
Resolution:
Allow alternatives 2 and/or 3 to be correct
l
responses.
References:
LaSalle Systems Description, Chapter 18,
Page 16.
,
NRC Resolution:
Comment is not accepted.
The selection of a rod which
will result in an insert or a withdraw error is not the
same as the selection of a rod in a group that is not
currently latched.
The operator may select a rod in
the latched group and insert or withdraw it beyond group
,
limits resulting in an error display without a select
error alarm.
Answer key is not modified.
.t
1
7.01
Facility Comment:
Though not a direct entry condition, LGA-ATWS-04 is
executed concurrently with LGA-ATVS-01 any time ATWS-01 is
entered.
ATWS-01 is Power Control.
ATWS-04 is Level
Control.
One of the functions of ATWS-04 is to control
power by controlling RPV level.
If an examinee
understands that whenever he is in LGA-ATWS-01 he is also
in LGA-ATWS-04 he should not be penalized.
If the examinee states that LGA-ATWS-01 has been entered
,
in the first part of the question, he should receive
credit for that part of the entry condition for LGA-ATWS-03.
i
He has already demonstrated that he knows that the ATWS-01
procedure has been entered.
Once an operator knows that
ATWS-01 has been entered he knows to exit all non
-
ATWS procedures and to enter the ATWS procedures.
He now
j
looks for containment parameters that will require entry
into the ATWS-02 or in this case ATWS-03 (high pool
temp).
,
I
Also the answer key for the first part of the question
3
states that ATWS-01 was entered due to pressure of
s
1080 psig and failure to scram.
This tends to be slightly
confusing since the entry conditions to ATWS-01 are worded
'
is such that "Boron has been injected into the RPV to
shutdown the Reactor
]
AND
!
Any of the following:
a.
level below +12.5"
i
j
b.
pressure above 1043 psig
!
c.
drywell pressure above 1.69 psig"
The other condition is the one that states "A" condition
which requires a scram
.
10
,
-
-
-_.
--
_
~. - --
__ .,
_
. . - -
-,
.-
-
p
-
.
AND
Either of the following:
a.
Reactor is critical
b.
power can't be determined"
Therefore, the way the plant conditions were set up, it
may be interpreted as follows.
Entry into ATWS-01 was
made only due to condition requiring a scram because the
pressure part does not come into play until after boron
has been injected.
If this interpretation is taken, then
the 1080 psig would not be listed as an entry condition.
Resolution:
Do not deduct any credit if LGA-04, ATWS Level
'
Control, is included as a procedure that is catered.
It's executed concurrently with LGA-ATWS-01.
If the examinee mentions that LGA-ATWS-01, Power
-
Control has been entered, then he should not be
deducted any points in the second part of the
question if he answers that "LGA-ATWS-03 has been
4
entered because of high suppression, pool temperature
(above 100'F)."
Knowledge that LGA-ATWS-01 has
already been demonstrated in the first part of the
question.
'
'
No credit should be deducted for not saying 1080 psig
as an entry condition, since 1080 psig is not an
i
entry until after boron has been injected.
1
i
References:
LGA-ATWS-01 Steps B.1, 8.2 and C.7; and
j
l
LGA-ATWS Flowchart.
t
NRC Resolution:
Comment is partially accepted.
The high reactor pressure
,
condition will be deleted from the answer key since boron
L
j
injection is not provided in the initial conditions.
No
credit will be deducted if the candidate states a procedure
that is entered by the direction of LGA-ATWS-01, such as
LGA-ATWS-04.
However, statement by the candidate that
LGA-ATWS-01 has been entered is not sufficient to
i
demonstrate that the candidate knows he must enter
LGA-ATWS-03 whenever LGA-ATWS-01 is entered.
Answer key
is modified.
,
i
'
l
I
!
!
11
1
- - - . .
,
. - -
- ,
,
.--
,-
- - , -
--
.
-
_
.
.
.
.
.
7.02a
Facility Comment:
No learning objective exists for stating from memory the
specific level or D/W temperature at which an onscale
reading may exist on the upset and shutdown level
'
instruments when actual level is at or below the lower
instrument tap.
To prevent the necessity of looking up
the value or committing the specific values to memory,
,
red tags are mounted in the control room adjacent to the
affected level indicators.
This was done as part of the
,
human factors review for E0P Verification several years ago.
4
Resolution:
Delete required value specific information
for 7.02a.
If 7.02a. is not deleted, accept reasonable
answer on a parameter level i.e. ,
"hot drywell reading
low on scale."
Reference:
LaSalle Control Room placards adjacent to
Upset and Shutdown level indicators.
NRC Resolution 1
Comment is partially accepted.
The senior reactor
operator should be aware o) the general conditions which
can affect indications for major plant parameters.
In
i
addition, the general precautions are not listed on the
E0P flow charts and are not integrated into the E0Ps.
Answer key modification is not required.
7.02c
Facility Comment:
Other reasonable answers should also be acceptable, such
,
as "to prevent turbine damage" since this directly related
,
to the insufficient lube oil flow.
Note that the General
Precautions in the LGA's do not state any reason, they
just say don't operate below 2100 RPM.
The other
'
clarifications are out of the LGA lesson plan,
i
Resolution:
Accept reasonable statements about bearing
damage, turbine damage, etc., for full credit.
i
!
References:
LGA General Precautions; and LGA Lesson Plan
p. 4.
NRC Resolution:
Answers provided by the candidates need not agree
"word-for-word" with the answer key.
The questien was
intended to determine that the candidate is aware of the
adverse effects of operating RCIC below 2100 rpm.
Answer
key modification is not required.
7.04b
Facility Comment:
Step F.5.b of LOP-CM-03 includes the Remote Shutdown Panel
temperature indicator as an additional method to
accurately determine Suppression Pool temperature during
i
low level conditions.
j
i
1
12
w
_
_
.._.
. ..
. _ _ _
f
.
.
!
.
Resolution:
Also accept Remote Shutdown Panel Supp. Pool
l
Temperature as one of the acceptable answers.
t
i
Reference:
LOP-CM-03, Suppression Pool Bulk Temperature
.
'
Determination.
NRC Resolution:
Comment is accepted.
Answer key modified.
t
j
I
i
7.05a
i
Facility Comment:
The answer key implies that af ter starting "... prevents
i
j
reverse power tripping of the Diesel Generator" that
further clarification is needed for full credit.
The
further clarification was stated as ";..due to large load
i
,
changes on the grid."
t
Resolution:
The statement about "...due to large load
l
changes..." should not be required for full credit.
If
l
the candidate states that reverse power is the concern,
j
j
this is adequate for full credit.
NRC Resolution
Answers provided by the candidate need not agree
f
1
"word-for-word" with the answer key.
Answer key
i
modification is not required.
7.05b
i
Facility Comtrent:
The question does not specify the indications are only
i
from the control room.
There are numerous other parameter /
indications which are checked locally and should be
included in the answer key as acceptable answers.
It is
especially important to check the 0/G because many of the
o
trips are now bypassed and it is running unloaded,
i
i
i
Resolution:
Accept any of the local parameters that are
!
listed in the procedure for full credit (i.e., ventilation
'
!
starts, fuel oil makeup, etc.).
It should also be noted
?
)
that 0/G Operating parameters include LOCA indication for
I
i
the following:
.
-
'
Cooling Water Flow
.
Oil Pressure
!
Fuel Oil Pressure
t
l
Cooling Water Temperature
Generator Winding Temperature
Cylinder Temperatures
Oil Temperature
Engine Running Light (in Control Room)
]
See LOS-DG-Mi, Attachment A, for a list of local readings
,
taken during surveillances.
References:
LOP-0G-02 pp. 15, 16, 17; LCS-0G-M1, Att. A.
w
'
I
1
13
1
- -
-
-
-- --
- -
---
- a
l
.
-
.
.
NRC Resolution:
Comment is not accepted.
The question did not ask the
candidate to memorize the parameters monitored during a
surveillance procedure.
The question specifically stated
that four thecks were required in accordance with
LOP-DG-02 to verify proper operation following a
LOCA without an under voltage condition.
The candidates
should know what is required to verify the automatic
initiation of an emergency system.
Answer key is not
modified.
7.06b
Facility Comment:
Answer key stresses thermal shock to the SDC return
nozzles.
Another equally acceptable answer would be
thermal shock to the RHR heat exchanger.
Resolution:
Accept thermal shock of the RHR heat
exchanger as an acceptable alternate answer for full credit.
Reference:
LOP-R4-07, 50C System Startup and Operation,
Page 3 and 4.
NRC Resolution:
Comment is accepted.
Answer key is modified.
7.06c
l
Facility Comment:
Bottom Head drain flow greater than 25 gpm is an alternate
acceptable answer.
l
Resolution:
Accept alte4nate answer of Bottom head drain
i
flow greater than 25 gpm
i
References:
LOP-RR-04, P. 3, Step D.6.
l
NRC Resolution:
Comments is accepted.
Answer key is modified.
7.09a
Facility Comment:
Also accept as one of the conditions that could cause an
i
automatic scram 'he APRM scram at 118% OR at .66WR + 51%
l
(clipped at 113.5%).
-
Resolution:
Accept APRM scram as an alternate answer for
full credit since it is also listed in the same LOA.
'
Reference:
LOA-EH-01 P. 2, B.4.
l
NRC Resolution:
Comment is accepted.
Answer key is modified,
7.09b
i
Facility Comment:
Although not included in the procedure as a way to control
'
pressure with bypass valves available, the "bypass jack"
would also be an acceptable answer.
With use of the
bypass jack push buttons on the EHC console, bypass valves
can be manually operated to control pressure.
Scenario's
can be postulated where the B/P valves when the load Set
would be ineffective.
14
-
-
-
.
.
-
_ _ _ _ _ _ _ _ .
._ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _
__ __________________________ _ __________ _
- _ _ _ _
.
.
.
.
.
'
Resolution:
Accept "use of bypass jack" for full credit,
although not included in the procedures use of the bypass
jack is an acceptable method to control pressure and this
statement also shows that the candidate has a working
knowledge of the EHC system.
References:
EHC simplified logic orawing.
NRC Resolution:
Comment is not accepted.
The question specifically asked
how the procedure (LOA-EH-01, EHC Pressure Regulation
Malfunction) directs the operator to control turbine load,
not how he can control reactor pressure with the bypass
valves.
In addition, turbine control valve demand will
not be directly controlled by the bypass jack; therefore,
for a failure of the pressure regulator causing a decrease
in reactor pressure (opening of the control valves) the
bypass jack would not assist the operator in controlling
turbine load or reactor pressure.
Thus, the operator must
strictly adhere to the procedure since it directs the
proper methods for directly controlling the turbine.
Answer key modification is not required.
7.10a
Facility Comment:
The procedure and the answer key both say reduce power
to "... about 63% power." The intent of this statement is
to reduce Reactor power to within the capability of the
condensate system without reliance on pump forward heater
drain.
Since this is an approximate number and not a
concrete setpoint then the Reactor power specified should
not be just 63%.
Resolution:
Put an acceptance band of i 5% on the answer
key so that any number between 58% and 68% inclusive is
acceptable for full credit.
Reference:
LOA-HD-01.
NRC Resolution:
In general, for this examination a tolerance band of 10%
of the answer key value will be used to determine the
acceptability of a numerical answer /setpoint.
Answer key
modification is not required.
i
7.11
Facility Comment:
The automatic actions listed in LOA-TG-06 are not
all inclusive of plant response.
Other answers may also
be correct for automatic actions on the indicated turbine
trip.
Ac:ept for full credit any reasonable auto, action
that would occur, even if not included on the answer key
and/or procedure,
i
i
l
4
15
__
_
_.
.
_
-.
--
. .
_
.
>
.
.
>
,
,
j
.
See LOA-TG-04 Pages 2 and 3 for automatic actions that
occur with the exception of Item #4, bypass valves wi!)
!
open.
It is not reasonable to expect the candidates to
i
,
'
memorize specific automatic actions stated in procedures
!
,
since they may or may not occur and do not form a
j
]
consummate list.
!
Resolution: Accept any reasonable automatic actions that
I
j
occur on a turbine trip even though not specifically
addressed in this procedure.
The following are_also some
1
'
other auto actions not listed in sither procedure.
If
.
further documentation is necessary, please contact the
facility on specific items and it will be promptly
,
l
forwarded to you.
~
)
2
Other acceptable auto actions not specifically addressed
!
l
in LOA-TG-04 or LOA-TG-06.
!
Extraction steam spill valves open
i
4
Rectre, pump M/A stations transfer to manual (due to
!
downshift signal)
i
1
1043 psig Reactor Scram (Possible)
l
1076 psig Relief Valves Start to Open (Possible)
l
]
1135 psig ATWS Recirc Pumps Trip Off (Possible)
.
APRM hi flu't scram (Possible)
!
!
Heater drain tank level automatic setdown to 3.2 ft,
i
OCB's open on Generator cross trip
l
Generator field breaker opens
!
.
Generator voltage regulator transfers to manual
t
If in normal A.C. lineup, Bus 151 and 141X will auto.
'
transfer to System Aux transformer
.
'
References:
LOA-TG-04 and 06.
!
l
NRC Resolution:
Comment is accepted.
Since LOA-TG-06 does not list all of
'
3
the significant automatic actions, automatic actions
provided by the candidates will be evaluated on a "case by
<
case" basis.
Answer key is clarified,
l
8.02b
'
Facility Comment:
This questian is worded such that the concept being asked
'
is "identify the difference between a functional test for
a bistable channel as compared to an Analog channel.
,
The wording in the question may be misleading to the
,
l
examinee.
The Technical Specifications (TSs) actually
breaks the definition into two parts, one for Analog
'
l
channels and one for Bistable channels.
The Analog
channel functional test injects a signal into the channel
"as close to the sensor as practicable..." whereas the
,
Bistable channel function test injects a signal "into the
1
sensor...".
First, it is not operationally significant
i
16
j
_-
-
. .
- - . .
- - . - _ . .
.- - - . -
- . - - _ - , - , - _ _ ,-.
__.
-
.-
.
.
f
that the SRO know this.
The actions of a functional test
are governed by an on site review approved LaSalle
Instrument surveillance which is performed by the
Instrument Maintenance Staff (not Operating).
Second,
the K/A numbers referenced do not relate to the question
asked.
Third, there are instances where analog channels do have a
functional test that has signal injected into the sensor.
See LOS-WL-Q1, Lake Blowdown Flow Indicator Channel
Functional Test.
See LOS-WF-Q1, Liquid Radwaste System
Operability and Radwaste Effluent Flowmeter Channel
Functional Test.
Both of these procedures actually send
flow through the instrument and record actual flow sensed
at the sensor and these are considered analog channels.
Resolution:
Facility recommends deleting part b of this
question with the associated reduction in point value.
References:
TS Definitions, LOS-WL-Q1, and LOS-WF-Q1.
NRC Resolution:
Comment is accepted.
Since the question did not specify
that the candidate should answer in accordance with
Technical Specifications and some channel functional tests
i
may insert a simulated signal directly into a sensor
,
channel, the answer key is modified.
i
8.05
Facility Comment:
The e>:act wording in the answer key should not
be necessary for full credit.
If the individual states
')
"authorize people to exceed normal radiation levels" it
should not be necessary to include "beyond 10 CFR 20
limits".
Resolution:
Accept any reasonable alternate wording that
,
states the concept for full credit.
'
NRC Resolution:
Answers provided by the candidates need not agree
"word-for-word" with the answer key.
Answer key
modification is not required.
8.06
Facility Comment:
The following comments are provided for Question 8.06:
a.
Comment:
The radiation release given states "from
the stack..."
The plant conditions given would have
also Caused Standby Gas Treatment (SBGT) to initiate
{
automatically (-50" Rx level or 1,69 psig Drywell
pressure).
The examinees may discuss that with SBGT
running that both the Stack Wide Range Gas Monitor
(WRGM) and the SBGT WRGM must be monitored for
,
,
release rate.
The readings of both WRGM's should be
-
added together to determine plant release for
classification of GSEP.
,
17
_ _ _ _ _ _ _ _
. _ _ . -_ _-___ ______ ___ ___ _______ ____ ___ __ .
_ __
_ _ _ _ _ _ _ _ _ _ .
__ _ _
- _ _ _ _ _ _ _ _ _ _
.
~
.
.
Resolution:
No point reduction should occur for this
discussion or for essumption that the given release
rate includes both WRGM's.
References:
LZP 1200-2, P. 1.
b.
Coment:
Once the General Emergency has been declared
I
and recommended protective actions are being assessed
l
using L7.P-1200-5 Attachment B (Page 4), no reduction
i
in points should occur for stating that initial
l
recomendations should be (S) S) S) NARS Form 9C,
!
0. E, and F.
l
Resolution:
Note 1 allows for the Station Director
'
to either omit this step or make these recommendations
based on current plant stat:" Dd the time from
classification, then make
"EN recommendations as
necessary.
Therefore, it F-
w listic for the
examinee to state the abovi i; initial recomendation
and then make further reconwendations as stated in
the answer key.
References:
LZP-1200-5 Attachment B, Recommend 2d
Protective Actions flowchart,
c.
Comment:
Also full credit should be given for giving
the recommended protective action as "NARS Form E) E)
(
5) 9C, G
H, and F".
The answer key indicates for
'
,
'
"9G" must be further defined as "evacuation of the
entire 2 mile radius around the plant."
Resolution:
This should not be required for full
credit since this is simply copying the NARS Form,
which was given as an attachment to the exam.
If
the examinee states what NARS Form section, by
Alpha-numeric identification, is to be filled out
then this has demonstrated the level of knowledge
required.
If the examinee did not feel it was
necessary or requ1r? to re-state the recommended
actions by copying them from the NARS Form, they
should not be penalized.
The question asks the student to determine the GSEP
i
classification and recommended protective actions.
l
It also asks for the attachments for determining the
protective actions.
But the answer keys required
0.25 pts. . for stating the procedures used to
determine the ASEP classification which was not asked
,
'
for.
18
i
,
.
.,
.
If the student correctly classified the GSEP event,
stated the procedures for determining recommended
protective actions and clarified the symbols for
these protection (i.e.. 9C, G, H, & F, E) E) S)),
but ended up with the wrong protective actions he
should get a majority of the points for the problem
because he demonstrated the ability to use the
procedures.
References:
LZP 1200-2; and LZP 1200-5, Recommended
Protective Action Flowchart.
NRC Resolution:
Comment is pa'.tially accepted.
Reasonable discussions of
the release rate equalling the sum of the SBGT Vent Stack
End the Station Vent Stack will certainly be accepted for
full credit.
Again, answers provided by the candidate
need not agree "word-for-word" with the answer key but
must demonstrate his ability to use procedures to determine
the correct protective actions to be recommended.
Partial
credit has been allotted for applying the correct procedure
to determine the protective actions.
Furthermore,
recommendation of the correct protective actions is the
final measure of whether the candidate has or has not
properly used the procedure.
Answer key is clarified.
8.07a
Facility Comment:
The APRM setpoint basis quoted in the answer key is
a basis for use of the formula (.66W + 72%) T in
determining the setpoint.
This formula _ allows for
adjustment of the rod block setpoint based on changes
in T. T = fraction of rated thermal power (FRTP) divided
by Maximum fraction of. limiting power density (MFLPD).
What this ratio is determining is if there is a peaking
problem somewhere in the core.
If the ratio of FRTP
over MFLPD ic less than one (1), then somewhere in the
core we are operating on an LHGR limit and the flow biased
rod block and scram setpoints should be set more
conservative to protect this local peaking spot.
It is
not imperative that the license candidate know the basis
behind flow biasing, since this is a design / licensing
analysis over which he has no control.
Note that the
K/A reference for this question has an asterisk which also
shows the rating spread was very broad or more than
15% of the raters indicated this knowledge is not
re : Jired.
It is imperative that the candidate know that the ratio
of FRTP over MFLPD be checked in accordance with the
appropriate procedure (LOS-AA-51) and that action be taken
as required by that procedure.
[
19
,
l
!
.
_ . _ - _ _ _ . _ _ _ _
_
-_
.
.
.
.
It also should be noted.that the APRM lesson plan
discusses the fact that APRM setpoints are set to prevent
fuel damage and that the fixed scram setpoint from
exceeding the LSSS value on a failure of the flow control
in the high flow direction.
s
Resolution:
Delete this part of the question with the
appropriate reduction in point value.
If not deleted,
then accept for full credit any reasonable discussion
which talks about preventing fuel damage and/or prevent
from exceeding the LSSS setpoints.
Another reasonable
discussion could include allowing maneuverability of
reactor power with flow while still allowing for
conservative trip setpoints that would prevent fuel'
damage.
References:
LOS-/.-S1; and LaSalle Systems Chapter 14,
pp. 8, 10, 12.
NRC Resolution:
Answer provided by the candidates need not agree
"word-for-word" with the answer key.
Answer key
modification is not required.
8.07b
Facility Comment:
Since it is the concept of what EOC-RPT is designed for,
any reasonable discussion which talks about negative
reactivity insertion and/or improvi,g the MCPR
consequences of Turbine Trip or Load Reject.
Other
discussions may state this recovers the loss of thermal
margin which occurs at the end-of-cycle.
Resolution:
Full credit should be given for the above
mentioned discussions without having to discuss the
physical phenomenon of void feedback adding positive
reactivity faster than control rods adding negative
reactivity.
The question asked for Bases and/or design
reason for this trip which would not include a discussion
on the physical effects of the plant.
References:
Recirc. Lesson Plan, Page 30; Tech Spec Bases
for EOC-RPT.
NRC Resolution:
Answers provided by the candidates need not agree
word-for-word" with the answer key.
Answer key
modification is not required.
)
8.07c
'
Facility Comment:
Also accept reasonable answers that are the same concept.
Discussion on limiting inventory loss is the same as
preventing fuel uncovery or limiting core differential
pressure (due to excessive flow out the steamline break).
Reference:
Main Steam Lesson Plan, p. 12.
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NRC Resolution:
Answers provided by the candidate need not agree
"word for"word" with the answer key.
Answer kcy
modifb. ; ion is not required.
~ J9b
Facility Comment:
The answer key focuses on the fact that individual rods
are selected to monitor LPRM's for oscillations since the
APRM recorders may not respond fast enough to show the
oscillations.
This is only partially correct. Another
major reason for monitoring regional LPRM's is that tne
core could be experiencing regional oscillations which
could be "out of phase" and therefore would be masked on
the APRM's (i.e., if LPRM's are going upscale in one
region of the core and downscale in another region, the
APRM oscillation may be dampened by the out-of phase
in9uts).
Resolution:
Also accept statements that regional LPRM's
are monitored to verify that either there are no Global
oscillation of the entire core or that there are no
regional oscillations in the core.
Either statement is
correct for full credit.
References:
Special Operating Order No. 88-20; LOS-RR-SR1,
Thermal Hydraulic Stability Surveillance; and LOA-RR-09,
Core Instability (new procedure).
NRC Resolution:
Comment is accepted.
Answer key is modified.
8.09d
Facility Comment:
Due to the open ended style of this question, all
reasonsble discussion concerning this concept should be
acceptable.
Discussions may include the regional
out-of phase oscillations discussed earlier, the potential
loss of APRM trips (due to out-of phase oscillations),
potential for exceeding thermal limits, etc.
Resolution:
Accept any of the above discussion for full
credit.
NRC Resolution:
Answers provided by the candidate need not agree
"word-for-word" with the answer key, but should address
the concept of thermal hydraulic instability or its
adverse effects.
Answer key modification is not required.
8.10
Facility Comment:
Any reasonable answer for which a worker would leave a
controlled area.
Since these are numerous it would be
hard to give all exainples (i.e. , when told to by Shift
Supervisor - General work practice, when the mask
and/or breathing becomes difficult - Nuclear General
Employee Training). Also note that the individual may
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state when told to leave by the timekeeper, this is common
terminology for the Rad Chem Tech keeping time at a high
rad or controlled area.
Resolution: Accept alternate, reasonable answers.
NRC Resolution:
Comment is partially accepted.
Timekeeper will be
accepted as an equivclent to Radiation-Chemistry
Department.
Although the answers provided by the
candidates need not agree "word-for-word" with the answer
.
key, they must address the conditions listed in
LRP-1001-1.
Answer key modification is not required.
8.11.c
Facility Comment:
The answer key specifies that "...per
the definition of
operability, all necessary attendaat instrumentation and
controls must be operable...", this particular part of the
answer should not be required for full credit.
If the
examinee states that startup is not allowed due to Tech.
Spec. 3.5.1 and Tech. Spec. 3.0.4 will not allow entry
into a mode while depending on the action statement, this
is adequate.
The question does not ask for the
operability definition, only which Tech Spec prevents
startup.
Resolution:
Do not require the definition for operability
to be required for full credit.
NRC Resolution:
The definition of operability in the answer key was
intended to clarify entry into T.S. 3.5.1 and will not be
required for full credit if the candidate clearly
demonstrates that he understands ADS valves cannot be
considered operable with a partial logic system failure.
Answer key is clarified.
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ATTACHMENT 4
REACTOR OPERATOR (RO)
EXAMINATION COMMENTS AND RESOLUTIONS
General observations based on examination performance are:
'
1.
R0s demonstrated a good knowledge in the following areas:
a.
How xenon affects power changes.
(Question 1.01)
b.
What factors influence the point of criticality.
(Question 1.02)
c.
Ability to calculate Keff.
(Question 1.07)
d.
Bases and operation of the low-low set function.
(Question 2.03)
e.
Relationship between the recirculation system controls and the
average power range monitors (APRMs).
(Question 2.04)
f.
Emergency diesel generator controls.
(Question 3.04)
g.
Operator actions for a stuck open relieve valve.
(Question 4.02)
2.
R0s demonstrated a poor knowledge in the following areas:
a.
Ability to predict changes in required and available net positive
suction head.
(Quastion 4.02)
b.
Bases for main steam isolation valve (MSIV) automatic isolation due
to low pressure.
(Question 2.08)
c.
Ability to predict turbine control valve positica due to a downscale
failure of the maximum combined flow setpoint.
(Question 3.02)
d.
Understanding of why indicated reactor vessel level changes due to
various conditions,
(Question 3.03)
Comments resulting in ssignificant changes to the master answer key, or.
comments "not accepted" by the NRC, are listed and explained below.
Comments
made that were insignificant in nature and resolved to the satisfaction of
both the examiner and the licensee during the post exam review are not listed
(i.e., typographical errors, relative acceptable terms, minor setpoint
changes, etc.).
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LASALLE R0 EXAMINATION OF APRIL 1988
1.03b
Facility Comment:
Part "b" of this question is the same as part "b" of
'
Question 5.06.
Please see that question for our comments
on this portion of Question 1.03.
NRC Resolution:
See NRC resolution to SR0 Question 5.06.
1.04b and c
Facility Comment:
Part "b" and "c" of this Question are tied together in
that the answer to "b" determines the answer to "c".
If
the examinee does not understand the recirculation flow
control system and shows this in his answer to "b",-then
he could be penalized for an answer for "c" which does not
match the answer key.
For example, the examinee thinks
that recirculation flow will remain the same in the master
manual and carries this through to part "c", then he might
say power will decrease because fuel temperature increasing
will cause doppler to increase.
In this example, the
examinee could be penalized for having both parts "b"
and "c" incorrect even though he does understand the
theory and just missed the systems part of the question.
There is another issue that could arise with this question
as we have operational difficulties with the recirculation
system which has more than once required us to "lock up"
the flow control valves.
With the valves "locked up" so
that they will not drift, the signals from the master
controller would not change the valve position.
Conceivably, an examinee might introduce the "real world"
into his/he. answer to this question and give appropriate
answers of no recirc. flow change and a power decrease.
This question is both a systems ques'. ion (part "b") and
two theory questions (parts "a" and "c").
Weighting the
systems part equal with the other two parts in the middle
of the theory section seems inappropriate as.this could
punish an examinee for systems weaknesses in a section
designed to test his/her theory knowledge.
Resolution:
(a)
The examinee who misses part "b"
should still get full credit for
part "c", if his reasoning for
part "c" is sound, but it was just
based upon his wrong answer for
part "b".
(b)
If the "real world" situation of
locked up recirculation valves is
brought up in part
"b",
full credit
should be given for the answer of no
flow change and a power decrease.
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(c)
The point distribution for the
respective parts of this question
should be changed to weigh more
heavily on the theory portion ("a"
and "c") with less on the systems
part of this question (say "a" and
"c" 1.0 point each and "b" worth
only 0.25 point or some other
similar redistribution).
NRC Resolution:
Comment is accepted:
The answer key has been changed to read
a.
Fuel temperature would INCREASE (0.5) to get the
needed delta T to transfer the heat to the coolant.
The corrosion layer will require some delta T across
it to transfer heat (0.5),
b.
' Reactor Recirculation Flow would INCREASE (0.125) to
add positive reactivity to compensate for the
negative reactivity effect of the fuel heat up (0.125).
c.
Core Thermal Power REMAINS THE SAME (0.5) since the
total amount of heat transferred to the coolant
remains constant (0.5).
1.05a
Facility Comment:
If the examinee keys in on the words "differential rod
worth", he/she will probably give an answer based on the
typical GE differential rod worth curve.
The answer to
question 1.05a would then be rod worth increases at first
and then decreases (see Attachment 1).
This answer
also follows the information given in LGP-1-1 (see
Attachment 2).
Both of these show differential rod worth
peaking in value at notch positions lower than 20.
Thus,
the examinee could answer on the basis of these sources
that rod worth would first increase and then decrease.
(Note that the attachments provided are both for startup
conditions.
At 50% power there would be a similar pattern
for differential rod worth with a flatter peak at about
the same spot.
We could not find a picture from GE
showing this, but the nuclear engineers at the station
have told us this from their experience with rod pulls a
power).
Resolution:
The answer key should be changed to allow the
examinee to say that differential rod worth would first
increase and then decrease.
References:
Rx Theory L. P., p. 19?; and LGP-1-1, p. 6.
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NRC Resolution:
Comment is not accepted.
The question asks for changes in rod worth when a rod is
at notch 02 versus when the rod is at notch 20.
It does
not deal with changes in rod worth as the rod is moved.
Also, the proposed answer does not give relative values
for the amount of rod worth increase or decrease.
So it
is unclear, from the proposed answer, whether rod worth
increases, decreases, or remains the same notch 02 to
notch 20.
The overall effect is a rod worth increase.
1.08a
Facility Comment:
(1) The answer to part "a" of this question has converted
the 553 seconds minutes.
This is not asked for in the
question.
(2) There is no specified band of required
accuracy for the answer.
It would seem reasonable to
accept answers ranging from 550 to 555 seconds.
It snould
be noted that some examinees may convert the IRM readings
into % power (using 100 on range 10 = 40% power as their
starting point).
This makes the problem more difficult,
requires more calculations and introduces the potential
for a diversity of answers within this band.
In the
answer key itself it should also be noted that the In of
250 is 5.5214609 which would make the answer 552 seconds,
not 553 seconds.
Resolution:
(1) Full credit should be given for leaving
the answer in seconds.
(2) The grading should allow a range of
answers from 550 to 555 seconds.
NRC Resolution:
Comment is accepted.
The answer was given in several
different forms, all of which are acceptable.
The answer key has been changed to read:
"552 +/ - 3 seconds"
1.09
Facility Comment:
The exam question asks for an explanation of why
peripheral rods change in value, but the exam key
allots 0.5 point for talking about central control rod
worth changes.
Resolution:
Due to the question specifically leading the
examinee to peripheral rod worths, full credits should be
given for answers which adequately answer the question
without discussing central control rods.
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NRC Resolution:
Comment is accepted.
The answer key is revised to read:
Peripheral control rod worth increases (1.0) because the
xenon peak in the center of the core forces the flux to
the periphery of the core (0.5), so the worth of the
peripheral rods, which is determined by the (local flux /
core average flux) 2 increases.
This could lead to a very
large reactivity addition when a peripheral rod is
withdrawn (0.5).
2.04b
F :ility Comment:
This question is not written to elicit the response
required by the answer key. . If the Flux Centro 11er output
signal reaches 106%, then the Loop and Servo Controller
input is also 106%.
Therefore, depending on which 106%
Abnormal Signal Relay is set more conservative a variety
of system responses could occur,
i.e., Flow Control Valve
lockup, Master, Flux, or Loop Controllers transfer to
manual.
(See System Description Chapter 6 figures and
electrical drawings).
Resolution:
Reasonable responses indicating the Candidate
understands the operation of Flow control circuitry, that
discuss the picking up of a 106% Abnormal Signal Relay
outside of the Flux Controller should be accepted as full
credit answers.
References:
LaSalle Systems Chapter 6, Figures 6-1
to 6-6 and prints 1E-1-4205 BX to CF.
NRC Resolution
Comment is accepted.
The answer key was modified by adding:
Flow control valve lockup
Flux controller shifts to manual
Master controller shifts to manual
2.05a
Facility Comment:
This is a standard "list" type question that is routinely
used at LaSalle for Initial and Requalification exams.
However, the SBLC system description and the Facility's
answer key have always treated all seven (?) items as
"positive" effects.
Resolution:
Reasonable responses that discuss "reduced
neutron leakage from hot to cold" and i' sufficient
reactivity to ensure 3% Shutdown Margin" should also be
considered as acceptable alternate answers.
References:
LaSalle Systems Chapter 10, p. 4; and Exam
Bank Question #01-021202.
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NRC Resolution:
Comment is accepted.
The answer key is revised to read:
decay of Xenon
elimination of voids (a)
increased water density (a).
reduced fuel temperature (a)
reduced neutron leakage
sufficient reactivity to ensure 3% shutdown margin
(4 of 6 required @ 0.5 each)
2.06a
Facility Comment:
The RCIC pump has the capability of taking a suction on
the A/B RHR heat exchangers during the Steam Condensing
mode of operation.
Resolution:
Also consider RHR as an acceptable alternate
answer for alternate water supply to the RCIC put..p.
References:
LaSalle Systems Chapter 41, P. 32.
NRC Resolution:
Comment is accepted.
RHR heat exchangers will be accepted
as an alternate water supply for RCIC.
2.06c
Facility Comment:
The facility does not require operators to memorize valve
numbers at LaSalle.
Resolution:
Also accept any reasonable noun name/
description of these valves,
i.e., "test return valves tc
the CY tank".
NRC Resolution:
Comment is accepted.
Valve numbers are considered optional.
2.08a.1
Facility Comment:
Due to the much publicized problems with Static-0-Ring
(SOR) Switches that the Facility has experienced
(setpoints drifting in the non-conservative direction),
'
the Facility's MSL hi-flow SOR switches have been
recalibrated to temporary more conservative setpoints
(Special Op Order #88-15).
These setpoints are
conservative to the effect that power level has been
restricted to prevent inadvertent MSIV closures during
normal operation.
Knowing this information, a candidate could reasonably
expect to see an MSIV isolation if he considers actual
steam flow setpoints versus design setpoints.
Resolution:
Reasonable responses stating that an MSIV
closure occurs should be considered full credit answers if
SOR switches are mentioned.
References:
Special Operating Order 88-15.
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NRC Rt olution:
Comment is accepted.
The examiner was not aware of.
Special Operating Order 88-15 when the examination was
written, because this was not sent with the other material
for preparing the examination.
This information is not in
the system lesson plan.
2.08a.2
Facility Comment:
The Facility does not require candidates to meoorize hi
flow isolation logic.
Each main steam line has
four (4) hi flow Dp switches with two discrete sensing
lines.
Isolation logic is one out of two, twice,
Without
piping, instrumentation, and logic diagrams any attempt to
answer this question would be a gi'ess.
Resolution:
This question should be deleted or credit
should be given for reasonable discussion that states that
one or more Dp switches will see hi flow.
References:
P&ID and Logic Prints for Main Steam Line
Isolat;ons.
NRC Re ,
ion:
Commen
's not accepted.
The logic for this isolation is
exact 1
e same as the logic for RPS (one out of two
taken t
e,.both channels using the same set of pressure
taps [A .use one tap, and C+D use another tap]).
The
operator should have no problems understanding this
'
logic.
Also, the high KA value (3.8) justifies this
question.
2.08a.3
Facility Comment:
Ouring the last Unit 2 outage the Group I isolation low
level setpoint was changed from -En to -129 inches.
See
LaSalle Operating Procedure LOP-PC-03.
Resolution:
Reasonable responses that indicate no action
occurs (isolation) should also be accepted as full credit
answers if U-2 is referenced.
Reference:
LOP-PC-03, pp. 12 and 13.
NRC Resolution:
Comnent is accepted.
The modification was not in the
system lesson plan.
The answer key has been changed to
accept, "no action occurs on Unit 2."
2.08b
Facility Comment:
The answer key for this question focuses on the failure of
the pressure regulator only.
The MSL pressure isolation
also provides protection and would isolate in the event of
a line break.
Resolution:
Reasonable responses regarding MSL break
protection should also be accepted for full credit, for
example,
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to:
conserve vessel inventory,
limit rad release to environs,
prevent excess Dp across core internals
NRC Resolution:
Comment is not accepted.
The pressure regulation system
(EHC) should be able to control the pressure transient on
a main steam line break up to the point where the MSIVs
will isolate on high flow.
Therefcre on a main steam line
break you should only see minor changes in pressure in the
reactor, so the low pressure setpoint~is not a protection
for the. main steam line break.
Conserving vessel inventory,
and limiting radiation releases to the environment are
both bases for the MSIV closure times.
Preventing excess
dp across the core internals is one of the design bases
for the st7am line flow restrictors.
These are
not appropriate answers to the question which was asked,
and will not be allowed.
2.09b
Facility Comment:
Two of the answers listed in the answer key are incorrect.
During the exam review this was discussed with the exam
writer.
He was under the impression that the S-50 reed
switch (overtravel) actuated the Rod Drift alarm.
An uncoupled rod will only actuate the rod drift alarm it
the rod is not selected for movement and it travels past
an odd reed switch (between 0 and 48).
The Simulator
malfunction book was incorrect regarding this.
The console
Operators Malfunction book was checked and found to be
corrected (handwritten).
This is not a controlled
document and should only be used as a scenario guidelina
for Simulator drills and not for examination grading.
Resolution:
Rod drift alarm and annunciator should be
deleted from the answer key.
Reasonable responses
regarding an observed change in drive flow (4 GPM)
to stall flow (1-2 GPM), or a loss of full out indication
on the Full Core Display should also be accepted as full
credit answers.
References:
LaSalle Systems Chapter 17, pp. 18-21.
NRC Resolution:
Comment is accepted.
The answer key has been modified as
follows:
(0.2)
Lose "48" position indication
(0.1)
Drive flow changes to stall flow
(0.1)
Loss of full out indication
(full core disp.
(0.1)
or any other reasonable answer
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2.10
Facility Comment:
See 6.02 for comment on this question.
NRC Resolution:
See NRC Resolution for SR0 Question 6.02.
-.05a
Facility Comment:
This question and answer has obviously been altered and
accidentally misworded.
The material in the answer key
referring to a low pressure pump permissive is extraneous
and could not be expected from the question that is asked.
The "why" in the question is also extraneous and will
confuse the candidates.
Arming and depressing the manual
initiation buttons will complete the logic and cause
syster initiation to occur.
The logic scheme is far too
complex to memorize.
It is unreasonable to expect a
response of greater detail ther. to state that "the
required logic is complete, therefore, initiation occurs."
Resolution:
(a)
Delete extraneous answer material.
(b)
Delete "why" from question in question
bank.
(c)
Accept reasonable responses that
indicate the candidate understands
initiation will occur.
NRC Resolution:
The comment regarding question a'.tering and miswording is
neither explained nor supported and therefore will not be
addressed.
Regarding the comment for asking "why", it is
important to know the cause effect relationships and
physical connections between ADS and LPCI and ADS and
core spray (Ref. KA 218000 Kl. 01-4. 0/4.1 and Kl. 02-4. 0/4.1).
However the wording of the question was not adequately
explicit to elicit the answer.
Therefore the answer key
is modified to read: yes (1.0)
3.05d
Facility Comment:
An examinee could reasonably consider the local equipment /
valve manipulations which are required to use the
emergency pressurization station enough to "hamper" the
,
operation of ADS.
Using this logic it is reasonable for
the candidate to censider the nitrogen accu:Aulators at
each SRV as a better explanation of why ADS operation is
not hampered without the nitrogen bottles.
Resolution:
Consider the nitrogen accumulators as an
acceptable alternate answer for full credit.
Reference:
LaSalle Systems Chapter 37, Figure 37-3.
NRC Resolution:
Comment is accepted.
Answer key is revised to include
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3.06
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Facility Comment:
See Question 6.11 Comment.
I
NRC Resolution:
See NRC Resolution for SRO Question 6.11.
3.08c
Facility Comment:
The high voltage low IN0P will generate a rod block signal
in addition to the RPS scram signal.
Resolution:
IN0P should be considered an acceptable
alternate answer for this question.
NRC Resolution:
Comment is accepted.
The answer key is revised to read:
Rod Block (0.25)
due to IRM downscale or high voltage low (0.25)
3.09a
Facility Comment:
The level setpoint setdown circuitry does not reduce the
setpoint by half as stated in the answer key.
The setpoint
is reduced to 18 inches regardless of the setpint tape
setting.
Resolution:
Change answer key to reflect a setpoint of
18 inches.
NRC Resolution:
Comment is accepted..
The answer key is revised to read:
The setpoint setdown circuitry reduces the operator
selected setpoint to 18" when a low level trip occurs.
(0.5)
3.10
Facility Comments:
This question is the same question as 3.09 on the NRC exam
from June 3, 1986.
The comments which were made and
accepted then apply on this exam, also.
The facility
believes that this question is not operationally
significant.
Due to vendor recommendations, operationally
the staticn air compressors are never operated in the
modulate plus two modes of operation.
They are left in
the modulate mode.
On the 1986 exam, the resolution was
to modify the answer key to give credit for the discussion
of surges and loading concerns.
Resolution:
Accept the same modification of the answer
key to give credit for the discussion of surges and
loading concerns.
NRC Resolution:
Comment is accepted.
The answer key has been modified to
give credit for discussion of surges and loading concerns.
It is important to note that the facility provided system
l
17;. son plan (dated March 1987) clearly describe the
modulate 2 step mode of operation and the conditions when
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it "should be placed" in this mode.
The operational
concerns mentioned in the facility comment are not reflected
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in the lesson plan.
Also the facility did not Evide the
system operating procedures which could air.s nas1*,een
checked to verify applicability of the question.
There is
also a concern that based on the history of this question
and comments that.the facility'is not properly revising
its lesson plans to reflect current plant operations.
4.01a
Facility Comments:
This question is a repeat of Question 4.04 on the NRC exam
given a LaSalle on June 3, 1986.
Our comments here are
the same as the ones which were accepted for that exam.
On both tests the question . incorrectly refers to "LOA-GP,
General Precautions".
The LGAs have the General Precautions
and this could confuse the examinees.
Thus, in part a
of 4.01 full credit should be'given if the examinee's
answer discusses that this should only be done when
conditions are stable and under control, or when continued
operation would worsen plant conditions.
Also, this
action should only be taken after review and approval by
the SR0 immediately available.
This is the criteria
discussed in LAP-1600-2, Conduct of Operations.
Resolution:
Either the answer already in the key from the
LGA General Precautions or the answer from the LAP cited
above should be accepted for full credit.
References:
LAP-1600-2, Rev. 31; and NRC resolution to
Question 4.01 of the June 3, 1986 R0 exam.
NRC Resolution:
Comment is accepted.
The answer key is modified to give
full credit for either answering from the LGA or from the
LAP.
4.01b
Facility Comment:
In part b of 4.01, the examinee could continue to answer
from the LAP which requires monitoring relevant parameters
by a licensed operator to assure safe operation of the
plant while the system or component is in manual control.
Resolution:
The examinee should be given full credit for
either answering from the LGA's as shown in the answer key
i
or for using the LAP.
References:
LGA Precaution No. 11 and LAP-1600-2.
NRC Resolution:
Comment is accepted.
The answer key has been modified to
give full credit for either answering from the LGA or from
the LAP.
11
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4.03a
Facility Comments:
Reference cited identifies sufficient warming of the LP
turbines as the basis for not doing the overspeed trip
test.
,
Resoluticn:
Allow warming of the LP turbines to be an
acceptable answer for full credit.
Reference:
LOP-TG-02, P. 6, Limitation and Action #10.
NRC Resolution:
Comment is accepted.
The answer hey has been modified to
read:
to ensure proper rotor warming (0.5)
also accept:
to ensure proper LP turbine warming
4.03c
facility Comments:
The responses provided in the answer key are not found in
LOP-TG-CR; although, these are acceptable reasons for
reducing power.
Additionally, power is reduced for other
reasons.
These reasons are tied to various procedures.
In LOA-CW-01, discussion Item #2, power is reduced to
place the plant in shutdown or hot standby.
LOA-1(2)PM03J-B511 has power reduced to stabilize vacuum.
As a good operating practice, power is reduced to minimize
the severity on the plosit is viii a puiential turbine trip
and scram.
Resolution:
Accept the following responses as acceptable
answers for full credit:
improve or stabilize vacuum.
'
-
-
reduce probability of tuibine damage from
-
overspeeding or high back pressure.
reduce rate of vacuum loss to allow time for
-
corrective action or shutdown.
reduce severity of potential turbine trip or scram.
-
NRC Resolution:
Comment is accepted.
The answer ke/ is revised to
read:
i
'
improve or stabilize vacuum
reduce probability of turbine damage from
overspeeding or high backp essure
reduce rate of vacuum loss to allow time for
corrective action or shutdown
reduce severity of potential turbine trip or scram
(2 of 4 required @ 0.5 each)
4.05
Facility Comment:
Refer to temments for SR0 Questions 7.06.
NRC Resolution:
See NRC Resolution for SRO Question 7.06.
12
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4.07a
Facility Comments:
As discussed in reference cited, the 45% power limit,is
based on two concerns: core instability and potential
reactor scrams.
The explanation part of the answer key
is weighted more toward the scram concern than the
instability concern (which is ignored except in
part b of this answer key).
This means that the examinee
would lose points for discussing the instability concern,
but for only discussing the scram concern.
Resolution:
Redistribute point.value assignment to
equalize the points between discussion of the potential
scram concern and the core instability concern.
Reference:
LOA-FW-01, P. 4.
NRC Resolution:
Comment is accepted.
The answer key has been modified as
follows:
Limit ,f 45% of rated core flow -or- 49 x 10E6 lbm/hr
Recirculation flow (0.5)
Rapid flow biased setpoint decrease and/or core flow
instability (0.75) together with
The APRM signal input to the thermal power monitor
being time delayed (0.25)
reduces the margin to APRM scrams during core flow
reductions (0.5)
4.07b
Facility Comment:
The same information for answering this question may have
been provided in part a.
This may lead to difficulty in
,
answering part b.
Resolution:
Allow credit if asoiding a reactor scram or
I
core instability is provided in answering part a.
'
NRC Resolution:
Credit will be given in part b if the flow instabilities
j
were discussed in part a.
4.08
Facility Comment:
Refer to comments for SRO Question 7.04.
NRC Resolution:
See NRC Resolution for SR0 Question 7.04.
4.10
Facility Comment:
Refer to comments for SR0 Ques, tion 8.10.
NRC Resolution:
See NRC Resolution for SR0 Question 8.10.
13
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6
U.
S.
NUCLEAR REGULATORY COMMISSION
'
SENIOR REACTOR OPERATOR LICENSE EXAMINATION
FACILITY:
_LAgALLE_1&g_____________
- :1 (: ' "' rb
!
'e
REACTOR TYPE:
,
_gWR-ggg_________________
t-
.
DATE ADMINISTERED: _gggg4fg6________________
EXAMINER:
_ _N_R C__ _ _R_E _G_I O_N_ _I _ _ _ _ _ _ _ _ _ _
_
_
CANDIDATE:
_________________________
,
INgIguCIJgNg_Ig_CeNDJgelgi
j
Use
separate
paper for the answers.
Write answers on one side only.
Staple question sheet
cn top of the answer
sheets.
Points for each
question are indicated in parentheses after the question.
The passing
grade requires at least 70% in each category
and a final
grade of at
least 90%.
Examination papers will be picked
up six (6)
hours after
the examination starts.
.
% OF
CATEGORY
% OF
CANDIDATE'S
CATEGORY
__VeLue_ _Igreg
___SCOgE___
_vetuE__ ______________Ce1Eqqay_____________
,'
/
.2E 99__ _2Eagg
________ 5.
THEORY OF NUCLEAR POWER PLANT
l
___________
OPERATION, FLUIDS, AND
-
THERMODYNAMICS
_29199__ _2Eagg
________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
___________
AND INSTRUMENTATION
_2E 99__ _2E 99
________ 7.
PROCEDURES - NORMAL, ABNORMAL,
___________
EMERGENCY AND RADIOLOGICAL
CONTROL
!
I
_2E 99__
2E 99
______.._S.
ADMINISTRATIVE PROCEDURES,
___________
CONDITIONS, AND LIMITATIONS
199:99__
________%
Totals
___________
Final Grade
4
1
All work done on this examination is my own.
I have neither given
nor received aid.
___--_----_________________________
Candidate's Signature
- m o n-
a r s
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
'
During the administration of this examination the f ollowing rules apply:
1.
Cheating on the examination means an automatic denial of your application
and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may
,
leave.
You must avoid all contacts with anyone outside the examination
room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil gnly to facilitate legible reproductions.
4.
Print your name in tne blank provided on the cover sheet of the
examination.
'
5.
Fill in the date on the cover sheet of the examination (if necessary) .
)
6.
Use only the paper provided for answers.
<
7.
Print your name in the upper right-hand corner of the first page of each
section of the answer sheet.
1
i
8.
Cor.secutively number each answer sheet, write "End of Category __" as
appropriate, start each category on a ngw page, write gnly gn gng sidg
of the paper, and write "Last Page" on the last answer sheet.
.
9.
Number each answer as to category and number, for example,
1.4,
6.3.
10. Skip at
least th gg lines between each answer.
t
11. Separate answer sheets from pad and place finished answer sheets face
down on your desk or table.
12. Use abbreviations only if they are commonly used in facility litetatute.
'
13. The point value for each question is indicated in parentheses after the
question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer
to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of
the examinet only.
17. You must sign the statement on the cover sheet that indicates that the
work is your own and you have not received or been given assistance in
completing the examination.
This must be done after the examination has
been completed.
.
,
s
- * 1
'n
18. When you complete your examination, you shall:
,
a.
Assemble your examination as follows:
'
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
I
b.
Turn in your copy of the examination and all pages used to answer
the examination questions.
c.
Turn in all scrap paper and the balance of the paper that you did
not use for answering the questionn.
d.
Leave the examination area, as defined by the examiner.
If after
1eaving, you are found in this area while the examination is still
l
in progress, your license may be denied or revoked.
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IBEBd99XN8dJgS
.
QUESTION
5.01
(2.50)
For each of the following events, WHICH COEFFICIENT of reactivity will
act FIRST to change core reactivity and WILL the reactivity added by
the coefficient be POSITIVE or NEGATIVE.
a.
Control rod drop at power
(0.50)
6.
SRV opening at power
(0.50)
c.
Loss of shutdown cooling (when shutdown)
(0.50)
d.
Main turbine trips while at 30% power
(0.50)
e.
Loss of one high pressure feedwater heater
(extraction steam is isolated)
(0.50)
OUESTION
5.02
(1.00)
-
Reactor power is increased from 40% to 50% causing the VOID fraction.
to INCREASE by 2%.
WILL the DOPPLER coefficient become MORE or LESS NEGATIVE.
EXPLAIN
<
YOUR ANSWER.
(1.005
'
QUESTION
5.03
(1.00)
MULTIPLE CHOICE
The Unit 1 reactor trips from full power and squilibrium xenon
conditions.
Four (4) hours later the reactor is brought critical and
power level is maintained on range 5 of the IRMs for the next two (2)
hours.
WHICH ONE of the following statements CORRECTLY describes the control
<
rod motion required to maintain a steady reactor power.
(1.00)
a.
Rods will have to be withdrawn due to xenon build-in.
b.
Rods will have to be rapidly inserted since the critical
reactor will cause a high rate of xenon burnout.
c.
Rods will have to be insertud since xenon will closely follow
its normal decay rate.
d.
Reds will remain approximately as is as xenon establishes
its new equilibrium value for this power level.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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iTHEORY OF NUCLEAR POWER PLANT OPERATION _FLUIpg2_AND
PAGE
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QUESTION
5.04
(3.00)
i
Unit 2 is operating at 85% rated power when recirculetion pump B
trips.
STATE whether the following parameters will INITIALLY INCREASE,
DECREASE, or REMAIN THE SAME and EXPLAIN WHY.
a.
Reactor pressure
(1.00)
6.
Indicated reactor water level (TWO (2) REASONS REQUIRED)
(1.00)
c.
Loop A Jet pump flow
(1.00)
QUESTION
5.05
(2.50)
.
Answer the following questions concerning the response of the Unit 2
reactor plant to a COMPLETE LOSS of ALL FEEDWATER PUMPS.
REFER TO FIG.
2.
- '
INITIAL CONDITIONS:
- The reactor is initially operating at 100% ratad power
- ALL reactor feed pumps are lost at time zero
- NO operator actions are taken
a.
WHY is reactor power DECREASING between t = zero to t = 7 seconds.
(0.50)
b.
WHY did the reactor scram at t = 7 seconds.
(BE SPECIFIC) (0.50)
c.
WHY does reactor pressure DECREASE between t = 7 to t = 30
seconds.
(0.50)
d.
WHY doen reactor water level begin INCREASING after t=28
seconds.
(0.50)
e.
WHY does TOTAL steam flow STABILIZE after t = 40 seconds.
40.50)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
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QUESTION
5.06
(3.00)
a.
STATE whether CRITICAL POWER will INCREASE, D'ECREASE, or REMAIN
THE SAME for each of the following changes.
EXPLAIN.
1.
Increased core inlet subcooling
(1.00)
2.
Gractor pressure increases from 930 psig to 980 psig
(1.00)
b.
STATE whether the CRITICAL POWER RATIO will INCREASE, DECREASE, or
REMAIN THE SAME for an INCREASE in the total recirculation flow
rate.
EXPLAIN.
(1.00)
QUESTION
5.07
(2.00)
The reactor is operating at 75% rated power and the operator is
withdrawing control rods to attain the 100% rod line.
WILL the wi thdra-al of a central control rod from notch 04 to 08 have
- '
a LARGER or SMAL.*R affect than withdrawal of the same rod from notch
36 to 40 on EACH
the f ollowing core parameters?
.
i
a.
Overall core thw mal power
(0.50)
b.
Axial flux distribution
(0.50)
c.
Radial flux distribution
(0.50)
d.
Local power surrounding the rod
(0.50)
,
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(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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QUESTION
5.08
(2.50)
l
MATCH EACH of the following THERMAL LIMITS (a to d) with the
l
statement (s) (1 to B) wnich best describe (s) that thermal limit.
'
(More than one answer may be applicable to a through d.)
(2.50)
a.
LHGR (Linear Heat Generation Rate)
b.
c.
MCPR (Minimum Critical Power Ratio)
d.
1.
Prevents stable film boiling from occurring in 99.9% of the fuel
rods.
2.
The two most limiting transients for this thermal limit are LOAD
REJECT without bypass and a FEEDWATER LEVEL CONTROLLER failure.
~
3.
Will allow a maximum of 0.1% of the fuel bundles to experience
transition boiling.
4.
Prevents fuel clad cracking due to excessive stress being exerted
on the fuel clad by the fuel pellet.
5.
Prevents failure of the fuel clad due to the reduced capability to
remove heat from the core following a Loss of Coolant Accident
6.
Prevents the fuel clad from reaching high temperatures which may
result in a Zirconium Water Reaction.
7.
Prevents the fuel cladding from exceeding 2200 deg. F during a
FEEDWATER LEVEL CONTROLLER failure.
B.
Prevents the plastic strain on the fuel rod from exceeding 1%
f ollowing a Loss of Coolant Accident.
t
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
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QUESTION
5.09
(3.00)
Complete the following table by indicating HOW EACH of the following
conditions will affect AVAILABLE and REQUIRED Net Positive Suction
Head for the reactor recirculation pumps.
(INCREASE, DECREASE, NOT
AFFECT)
CONDITION
AVAILABLE NPSH
REQUIRED NPSH
a.
Feedwater injection
temperature increases
(1.00)
6.
Reactor pressure increases
(1.00)
c.
Recirculation FCV (Flow
Control Valve) is slowly
throttled open 10%
(1.00)
_ ,
QUESTION
5.10
(2.00)
The thermal neutron flux profile changes as reactor power is increased
from criticality to 100% rated power.
Foe each of the following conditions, STATE whether thermal neutron
'
flux is peaked near the TOP or near the BOTTOM of the core.
EXPLAIN.
a.
The reactor is at the Point of Adding Heat
(1.00)
b.
The reactor is at 100% rated power
(1.00)
QUESTION
5.11
(2.50)
For reactor operation at equilibrium full power conditions, the
control rod density INCREASES from the beginning r* the fuel cycle
l
(BOC) until approximately one half into the cycle.
a.
EXPLAIN WHY this control rod density INCREASE occurs.
(1.00)
b.
HOW will the void coefficient of reactivity be affected by this
change in control rod density?
(INCREASE or DECREASE)
EXPLAIN.
(1.50)
(***** END OF CATEGORY 05 *****)
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PAGE
7
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QUESTION
6.01
(2.50)
Unit 2 ia operating at 55% rated power and 55% rated core flow with
two recirculation loops in service.
For EACH of the following conditions, WILL a SCRAM, HALF-SCRAM, ROD
BLOCK, or NO ACTION be generated? (For conditions that produce more
than one action, state the more severe action, i.e half-scram is more
severe than a rod block.)
a.
(0.50)
b.
APRM flow unit B fails UPSCALE.
(0.50)
c.
Inboard MSIV A and outboard MSIV C slowly drift closed.
(0.50)
d.
APRM B indicates 80% power.
(0.50)
e.
Turbine Stop Valves on steam lines A and D fail closed.
(0.50)
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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.
QUESTION
6.02
(3.00)
'
a.
For EACH of the following HPCS initial valve lineup conditions
indicate the FINAL position of the given valves following an
AUTOMATIC HPCS initiation:
1.
CST suction valve open, Suppression Pool suction valve shut.
(0.50)
2.
CST suction valve shut, Suppression Pool suction valve shut.
(0.50)
3.
CST suction valve shut, Suppression Pool suction valve open.
(0.50)
l
4.
HPCS full flow test dow stream stop valve (COli) open.
(0.50)
)
b.
MULTIPLE CHOICE
Following an AUTOMATIC HPCS INITIATION, CHOOSE the ONE statement
which correctly describes the condition which must be satisfied to
'
allow the operator to reset the HPCS initiation logic and return
the HPCS system to the standby lineup
(1.00)
^
1.
Whenever Drywell pressure has decreased to 1.2 psig and the
operator depresses the initiation reset pushbutton.
l
2.
Whenever reactor water level has increased above the low
water level initiation setpoint the operator depresses the
initiation reset pushbutton.
3.
Whenever the Drywell high pressure initiation signal is
present, the operator murt increase reactor water level above
Level 8 and then depress the initiation reset pushbutton.
4.
Whenever the Drywell pressure and the low water Invel
initiation signals are present, the operator must manually
close the injection valve (FOO4) and then depress the
initiation reset pushbutton.
(***** CATEGORY 06 CONTINUED ON NEXT FAGE *****)
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9
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QUESTION
6.03
(1.00)
MULTIPLE CHOICE
CHOOSE the ONE statement which correctly describes the Main Steam Line
Radiation Monitoring System.
(1.00)
a.
There are four (4) radiation monitoring channels, any two of these
charnels in the trippsd condition will cause an MSIV inclation.
b.
The two (2) channel select switches for the main steam line
radiation monitors allow the operator to select which radiation
monitor in each division will provide an RPS trip signal.
c.
There IS one radiation detector assigned to each steam line so that
it only monitors the radiation of that one steam line.
d.
The main steam line radiation detectors are physically arranged
such that significant increases in radiation levels can be
detected with any number of steam lines in service.
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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10
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QUESTION
6.04
(2.00)
Consider EACH of the f ollowing plant conditions SEPARATELY.
NOTE: NO OPERATOR ACTION IS TAKEN.
a.
Unit 1 is operating at 100% power, two turbine driven feedpumps in
service in 3-element control, when the B FEEDWATER flow
transmitter fails UPSCALE.
(1.00)
h.
Unit 2 is operating at 70% power, two turbine driven feedpumps in
service in 3-element controls. when ONE (1) STEAM flow transmitter
output signal to the Feedwater Level Control System FAILS to ZERO.
l
'
(1.00)
i
SELECT THE CORRECT Feedwater Level Control System / Plant RESPONSE
from the following CHOICES for parts a and b.
1.
Reactor water level decreases and stabilizes at A lower
~~
level.
2.
Reactor water level decreases and initiates a reactor
3.
Reactor water level increases and stabilizes at a higher
level.
4.
Reactor water level increases and initiates a turbine
trip.
U
Y Y A'
gg:
Q [ & w & ,
- l OL
2
A) Wo
(ha 4c
gm 4 sw~) lauf.
wm
mm
S
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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11
2
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QUESTION
6.05
(3.50)
Unit 2 is operating at 100% rated power when one of the jet pump
risers FAILS.
a.
WILL the f ollowing Control Room indications INCREASE, DECREASE, or
REMAIN THE SAME7
1.
Core flow
(0.50)
2.
Core differential pressure
(0.50)
3.
Reactor power
(0.50)
6.
FILL IN THE BLANK
The flow through the failed jet pump will be in the _________
(FORWARD or REVERSC) direction and will be _________ (HIGHER THAN,
EQUAL TO, or LOWER THAN) normal flow through the (intact) jet
pump.
(1.00)
!
c.
WHAT are the TWO (2) ADVERSE EFFECTS of a failed jet pump which
require the reactor to be shutdown?
(1.00)
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GUESTION
6.06
(3.50)
!
Answer the f ollowing questions concerning the RCIC (Reactor Core
Isolation Cooling) system.
a.
STATE whether EACH of the following valves is normally OPEN or
CLOSED when RCIC is in the STANDBY LINEUP 7
(1.00)
1.
Turbine Steam Supply Stop valve (F045)
2.
Turbine Governor valve
3.
Minimum Flow valve (F019)
4.
Cooling Water to the Lube Oil Cooler stop valve (F046)
!
b.
RCIC has AUTOMATICALLY INITIATEO.
For EACH of the situations
r
'
listed below, STATE whether FINAL RCIC injection into the reactor
wills
CONTINUE AUTOMATICALLY (no operator action),
REINITIATE AUTOMATICALLY,
)
REQUIRE CONTROL ROOM operator action, OR
REQUIRE LOCAL operator action.
-
1.
The RCIC Gland Exhauster VACUUM PUMP FAILS.
(0.50)
2.
A 125% overspeed trip is received due to low control oil
'
-
pressure.
Control oil pressure is then returned to normal.
(0.50)
3.
After decreasing to 25 psig, the reactor pressure increases to
110 psig.
(0.50)
4.
After increasing to 60 inches, reactor
v ssel water level
e
DECREASES to 50 inches.
(0.50)
c.
TRUE OR FALSE
.
The outboard steam isolation valve (FOOB) will CLOSE whenever
l
RCIC is in the STANDBY LINEUP AND the operator depresses the
MANUAL STEAM ISOLATICN BUTTON.
(0.50)
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QUESTION
6.07
(2.50)
a.
ADS ( Automatic Depressurization System) has automatically
initiated and the ADS valves have OPENED.
WILL the following
operator ACTIONS or CONDITIONS cause the ADS valves to CLOCE.
(YES or NO)
1.
The Drywell pressure is reduced to 1.3 psig and the operator
depresses the Drywell pressure reset button.
(0.50)
2.
The operator secures ALL low pressure ECCS (Emergency Core
Cooling) pumps.
(0.50)
3.
ADS logic channels A and C lose power.
(0.50)
4.
The operator depresses the low level reset pushbutton.
(0.50)
b.
TRUE OR FALSE
The ADS logic must sense pressere at the discharge of the LPCS or
the LPCI pumps in order for the operator to initiate ADS by arming
and depressing the MANUAL INITIATION BUTTONS.
(0.50)
-
QUESTION
6.08
(2.25)
A reactor plant startup is in progress on Unit 1.
WHAT is the CONDITION under which a reed switch failure at notch
a.
16 on a rod in the selected RSCS group will NOT result in a RSCS
rod block?
(0.50)
b.
WHAT information is displayed on the Operator Display Panel when
the "Amber Display Control" button is in the FREE RODS position?
{
BE SPECIFIC.
(0.75)
'
c.
WHEN will the RSCS rod blocks be automatically BYPASSED and WHAT
PARAMETER is monitored to provide this bypass?
(1.00)
OUESTION
6.09
(1.00)
A reactor startup is in progress and IRM C indicates 35 on range 8.
WHAT will occur if the operator inadvertently changes the IRM C Range
Switch to range 77
(SHOW ANY NECESSARY CALCULATIONS.)
(1.00)
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QUESTION
6.10
(1.75)
i
The Reactor Building Ventilation System has automatically isolated.
m.
WHAT are FIVE (5) of the six ISOLATION SIGNALS which could have
caused the itolation?
(SETPOINTS NOT REQUIRED.)
(1.25)
b.
STATE the ONE (1) Reactor Building Ventilation System isolation
SIGNAL which will NOT initiate the SBGT (Standby Gas Treatment)
System.
(0.50)
QUESTION
6.11
(2.00)
A reactor startup is in progress on Unit 2 and the operator is
withdrawing rods to attain criticality.
The f ollowing errors are being displayed by the Rod Worth
a.
Minimizer (RWM):
withdraw error
insert error
insert error
-
STATE the ACTION that must be taken by the operator to clear the
control cod block.
(1.00)
b.
MULTIPLE CHOICE
CHOOSE the ONE condition which will cause the RWM SELECT GRROR
light to be lit.
(1.00)
1.
WHENEVER one insert error exists and a red other than the red
causing the insert error is selected,
h
2.
WHENEVER the operator selects a control rod which will result
in an insert or a withdraw error.
,
3.
ANYTIME a rod block has been initiated by the RWM and the rod
selected is not one of the rods causing the block.
4.
AFTER the operator has withdrawn or inserted a rod which is
NOT in the presently latched RWM group.
)
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OUESTION
7.01
(2.00)
Initially the Unit 2 reactor was operating at 90% power when the
outboard MSIVs (Main Steam Isolation Valves) failed CLOSED.
The conditions following the MSIV closure are as follows:
- Reactor pressure is 10SO psig
- Drywell pressure in 1.55 psig
- Reactor recirculation pumps have tripped
- Reactor power is 70%
- Drywell temperature is 120 deg. F and in reasing
- Suppression Pool temperature in 10S deg. F and increasing
Suppression Pool level is + 1 inch increasing very slowly
-
Select tbn LaSalle General Abnormal Procedures which must be entered
AND STATE the CONDITION (S) which require (s) entry into each of the
Abnormal Procedures selected.
(Refer to the list of General Abnormal
Procedures below.)
LGA-ATWS-01 ATWS POWER CONTROL
i
LGA-ATWS-02 ATWS SECONDARY CONTAINMCNT CONTROL
LGA-ATWS-03 ATWS PRIMARY CONTAINMEN7 CONTROL
-
i
LGA-ATWS-04 ATWS LEVEL CONTROL
LGA-01
LEVEL / PRESSURE CONTROL
'
LGA-02
SECONDARY CONTAINMENT CONTROL
,
LGA-03
PRIMARY CONTAINMENT CONTROL
LGA-04
LEVEL RESTORATION
,
'
LGA-05
RPV FLOODING
(S r i a n '
ltn/,
- .
% % gavn pu neJeu, wivs EfA $du / MdTNQ
a~M ~ wg Q oitC,in V & efy m a % s<o.
q~
q .
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16
B09196991986_99 NIB 96
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QUESTION
7.02
(3.00)
Answer the following questions concerning the General Precautions for
i
the LaSalle General Abnormal Procedures.
The operator is cautioned that under accident situations, actual
a.
RPV water level may be anywhere below -10 inches on TWO (2)
RPV level instrument ranges.
STATE the CONDITIONS which will result in incorrect level
indication on the f ollowing level instruments.
i
1.
Shutdown Range
(0.50)
2.
Upset Range
(0.50)
b.
STATE the TWO (2) CONDITIONS which require the operator to close
the ADS SRVs (Automatic Depressurization System Safety Relief
Valves) following an automatic initiation of ADS.
(1.00)
c.
STATE the TWO (2) REASONS that the operator is cautioned AGAINST
- '
operating RCIC at less than 2100 rpm.
(1.00)
,
QUESTION
7.03
(3.00)
Answer the following questions concerning the opt.rator actions
directed by the LaSalle General Abnormal Procedures,
WHY is the operator directed NOT TO VENT the Drywell by LGA-03,
a.
Primary Containment Control when the Drywell p r e'.sur e is LESS THAN
60 psig AND the Drywell temperature is GREATER THAN 212 deg. F7
(1.00)
b.
WHY is the operator directed by LGA-04, Level Restoration, to wait
until level is -275 inches before opening an SRV when the
following renditions exists
- NO systems are aligned /available for injection to the RPV'
with at least one pump running
- RPV water level is FALLING
- RPV pressur e is GREATER THAN 57 psig?
(1.00)
c.
WHY does LGA-ATWS-01, ATWS Power Control, direct the operator to
runback the reactor recirculation pump FCVs (Flow Control Valves)
to MINIMUM PRIOR to tripping the recirculation pumps when the main
turbine is on the line?
(1.00)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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. PROCEDURES - NORMAL _QBNQRMALx_EMERGENQY_ANQ
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17
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6891969 GIG 86_QQNIBQL
QUESTION
7.04
(2.00)
Answer the following questions concerning the determination of the
bulk temperature of the Suppression Pool in accordance with LOP-CM-03,
Suppression Pool Bulk Temperature Determination.
a.
Why is the operator directed to determine the Suppression Pool
Temperature by the value print of the computer points L122 or L123
if Suppression Pool Level is LESS THAN 698 feet 11 inches (-8
inches).
(1.00)
b.
STATE TWO (2) additional METHODS for determining the Suppression
Pool bulk temperature.
(ALL ECCS PUMPS ARE OPERATING)
(1.00)
QUESTION
7.05
(2.00)
Answer the following questions concerning the Emergency Diesel
Generator in accordanco with LOP-DG-02, Startup of the Diesel
Generator.
..
a.
WHY does this procedure caution the operator to mairitain GREATER
THAN 200 KVAR when operating the Diesel Generator in parallel with
the grid.
(1.00)
b.
STATE the FOUR (4) CHECKS the operator must perform in order to
VERIFY that Diesel Generator 1A is operating properly following a
LOCA without an undervoltage on the 142Y bus.
(1.00)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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. PROCEDURES _ _NgRMAL _ABNgRMAL _gMERggNgY_gND
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18
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3
689196991986_G9NIBg6
-
QUESTION
7.06
(3.00)
The reactor is in Shutdown and the operator is preparing to place RHR
(Residual Heat Removal) into the Shutdown Cooling (SDC) mode.
Answer
the following questions concerning core cooling in accordance with
LOP-RH-07, Shutdown Cooling System Startup and Operation.
a.
WHY does this procedure CAUTION the operator to ensure that RP'/
level is at or above 40 inches as indicsted on the Shutdown Range
prior to starting an RHR pump in the Shutdown Cooling (SDC) mode
with no other forced flow through the vessel.
(1.00)
b.
WHY is the operator CAUTIONED to slowly cut-in the RHR heat
exchanger upon startup of an RHR loop in the Shutdown Cooling
Modo.
(1.00)
c.
STATE TWO (2) of the three CRITERI4 which will ensure that core
cooling flow is sufficient to PREVENT temperature STRATIFICATION
in the RPV.
(1.00)
_
QUESTION
7.07
(1.00)
WHY is the operator CAUTIONED by LOP-RI-02, Starting and Operating thw
RCIC System procedure, to trip the Main Turbing prior to injecting
with RCIC (Reactor Core Isolation Cooling).
(1.00)
i
QUESTION
7.08
(2.00)
The reactor is operating at 55% power when a complete loss of ALL
OFF-5ITE power occurs AND ALL of the AC BUSSES remain Deenergized.
STATE FOUR (4) of the five IMMEDIATE OPERATOR ACTIONS per LOA-AP-OB,
Total Loss of AC Power.
(2.00)
b
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BODI96991996_99 NIB 96
GUESTION
7.09
(2.50)
The reactor is operating at 100% rated thermal power and the Main
Turbine is at 100% load when the EHC Pressure Regulator MALFUNCTIONS.
Answer the f ollowing questions in accordance with LOA-RH-01, EHC
i
Pressure Regulation Malf uncti on.
'
a.
STATE THREE (3) CONDITIONS which will require a MANUAL SCRAM OR
cause an AUTOMATIC reactor SCRAM.
(1.50)
b.
HOW is the operator directed to control TURBINE LOAD if the Bypass
!
Valves are AVAILABLE7
(0.50)
c.
How is the operator directed to control STEAM FLOW if the operator
determines that the Bypass Valves have FAILED OPEN7
(0.50)
3
1
QUESTION
7.10
(2.00)
!
The Unit 2 reactor plant is operating at 90% rated power when all of
the operating Heater Drain Pumps TRIP.
- '
LOA-HD-01, Loss of Pumped Forward Heater Drain Flow procedure,
directs the operator to REDUCE reactor power.
STATE how the operator should reduce power and to STATE the power
a.
level which he in directed to immediately attain.
(0.50)
b.
STATE the THREE (3) PARAMETERS that the operator is maintaining by
,
!
reducing reactor power.
(INCLUDE APPLICABLE SETPOINTS.)
(1.50)
i
QUESTION
7.11
(2.50)
A turbine trip from 670 MWe has occurred and the Bypass Valves
indicate CLOSED.
,
WHAT are FIVE (5) of the seven AUTOMATIC ACTIONS that the operator is
l
required to verify by LOA-TG-06, Turbine Trip with Failure of Bypass
l
System.
(INCLUDE THE SETPOINTS AT WHICH THE AUTOMATIC ACTIONS OCCUR.)
(2.50)
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20
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QUESTION
8.01
(2.00)
STATE whether EACH of the following situations IS or IS NOT considered
to be a CORE ALTERATION in accordance with Unit i Technical
Specifications.
a.
The removal of LPRM (Local Power Range Monitoring) strings from
the core when irradiated fuel is loaded into the reactor vessel.
(0.50)
6.
The installation of neutron sources into the reactor vessel prior
to leading f uel bundles into the core.
(No fuel is loaded into
the vessel.)
(0.50)
,
c.
The withdrawal and insertion of a control rod for the purpose of
timing the control rod while all fuel is removed from the reactor
vessel.
(0.50)
d.
The removal of the steam dryers from the reactor vessel in
preparation for removal of the fuel bundles.
(0.50)
a .
QUESTION
B.02
(1.50)
TRUE OR FALSE
a.
A channel check of SRM B may be perf ormed by comparing its coutit
rate to the count rate indicated by the other 6RMs.
(0.50)
l
b.
A channel functional test of an analog trip syatem is performed
I
]
by inserting a simulated signal into the channel sensor to verify
the operability of the trip functions.
(0.50)
!
c.
An Offgas radiation monitor channel calibration may be performed
by adjusting the channel output so that it indicates the same as
the radiation monitoring channel which has just passed a channel
functional test.
(0.50)
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21
,
QUESTICN
B.03
(1.50)
'
STATE whether EACH of the following situations WILL or WILL NOT EXCEED
the guidelines of LAP-100-17, Overtime Guideline for Personnel that
Perform Safety Related Functions.
a.
The shift operator works 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> a day for two consecutive days
with an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> break after the first shift.
(0.50)
b.
The Shift Control Room Engineer works 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> a day for 4
consecutive days and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> a day for the following 3 days.
,
(0.50)
r.
The shift operator is scheduled to work 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> this shif t and is
expected to return and relieve for his next shif t in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and
then work a 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> shift.
(0.50)
QUESTION
8.04
(1.50)
Answer the following questions in accordance with LAP-2OO-3, Shift
-
Change.
WHO can relieve the Shift Control Room Engineer if another
qualified LSCS Shift Control Room Engineer in NOT available AND
'
'
WHAT CONDITION MUST be met?
(1.50)
,
QUESTION
O.05
(2.00)
In accordance with the General Stations Emergency Plan, STA1E the
FOUR (4) RESPONSIBILITIES of the Station Director which CANNOT be
delegated.
(2.00)
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22
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OUESTION
8.06
(3.50)
A LOCA (Loss of Coolant Accident) has occurred at Unit 2.
DETERMINE the emergency event CLASSIFICATION and STATE the
protuctive actions that should be recommended to Of f site Agencies.
(STATE THE ATTACHMENTS USED FOR DETERMINATION OF PROTECTIVE ACTIONS
,
AND SHOW ALL WORK ON ATTACHMENTS.)
Plant conditions are as follows:
No PROJECTED dose rates are available
Reactor water level is -161 inches
'
Drywell pressure is 40 psig
Radiation release rate from the stack is 5.0 EB uci/sec
Wind speed is 10 mph
Wind direction is 65 to 72 degrees
NOTE:
SELECTED EMERGENCY IMPLEMENTATION PROCEDURES HAVE BEEN
ATTACHED FOR REFERENCE (LZP-1200'S).
, -
OUESTION
B.07
(3.00)
1
WHAT is the Technical Specification BASIS and/or Design Reason for the
following Technical Specification requirements?
The APRM flow biased red block setpoint.
(1.00)
a.
(TWO (2) REASONS REQUIRED)
)
b.
The End of Cycle Recirculation pump trip.
(1.00)
(DESCRIBE ADVERSE AFFECT THAT IS PREVENTED.)
)
'
The MAXIMUM CLOSURE TIME of the MSIVs.
(1.00)
c.
(TWO (2) REASONS REQUIRED.)
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OUESTION
B.08
(2.50)
,
In accordance with LAP-200-1, Operating Department Organization,
during Abnormal Operations the Shift Engineer shall establish himself
in the Control Room at the command authority responsible for the
operation of the plant.
a.
DEFINE Abnormal Operations.
(1.00)
b.
WHAT is the MAXIMUM TIME that the Shift Engineer has to establish
command of the Control Room.
(0.50)
c.
WHAT are the TWO (2) CONDITIONS which must be satisfied for the
Shift Engineer to relinquish his command of the Control Room?
(1.00)
QUESTION
8.09
(3.00)
'
The Unit 2 reactor plant is operating on the 90% rod line at 70% of
rated ctre thermal power when Reactor Recirculation Pump B trips.
- '
Following the recirculation pump trip, reactor power indicatad by
APRMs is 45% and indicated core is flow is 38 E6 lbm/hr.
a.
WHAT IMMEDIATE CORRECTIVE ACTION is required by the Special
-
Operating Orders?
(0.50)
b.
HOW does the Special Operating Order direct the operator to
monitor core power level?
EXPLAIN WHY.
(1.00)
c.
WHICH ONE of the fo11owi3g Technical Specification regions of
operation has been entarud?
(0.75)
1.
Surveillance Region Restricted Zone
2.
Surveillance Region Allowable Ione
3.
Unrestricted Zone
,
d.
WHY does Technical Specifications place such stringent
requirements on, plant operation in the SURVEILLANCE Region? (0.75)
,
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OUESTION
8.10
(1.50)
Answer the f ollowing questions concerning radiological controls at the
LaSalle Nuclear Station.
LIST SIX (6) of the eight CONDITIONS which require a worker to
LEAVE a Controlled Area per the Radiation Protection Standards
procedure, LRP-1000-1.
(1.50)
!
QUESTION
8.11
(3.00)
Consider the Technical Specification requirements for the following
situation.
INITIAL CONDITIONS:
- The reactor is in Cold Shutdown.
- Preparation for a reactor startup is in progress.
- ADS /SRV related equipment malfunctions exist (a to c listed
below).
~~
STATE whether Technical Specifications WILL ALLOW or WILL PROHIBIT a
reactor startup and pressurization to 920 psig for each of the
following equipment malfunctions concerning the SRVs.
STATE AND
EXPLAIN ALL APPLICABLE TECHNICAL SPECIFICATIONS.
(NOTE:
Consider EACH of the following SEPARATELY)
a.
B21-F013E ADS solenoid is removed and B21-F013R ADS solenoid
failed to energize during the ADS Logic Functional Test.
(1.00)
b.
B21-F013K control switch is stuck in the closed position.
(B21-F013K i s not an ADS valve, but performs the LLS relief function)
G rccol Clear:fd d %.
SRV M i- F*c t ~b k c/5 is 54ack 644 * catd.
(1.00)
-
ye tI Uen et n</ V A e . Va/Ve t's C/e.resl.
c.
ADS Trip System B failed the Logic Functional Test.
(1.00)
- NOTE:
TECHNICAL SPECIFICATIONS ARE ATTACHED FOR REFERENCE *
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25
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ISEBD99Y_N9819@
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
ANSWER
5.01
(2.50)
a.
Doppler (0.25), negative (0.25)
b.
Void (0.25), negative (0.25)
c.
Moderator temperature (or fuel temperature)
(0.25), negative (0.25)
d.
Void (0.25), positive (0.25)
e.
Moderator temperature (0.25), posi ti ve (0.25)
REFERENCE
LESSON PLAN ON REACTOR PHYSICS PG 120 - 172.
LEARNING OBJECTIVE NO. 16, 17, AND 18.
GE BWR SERIES ON REACTOR THEORY
3.8
4.1
3.4
3.5
3.7
...K/A VALUE
239002A106
295005K101
295014K203
295014K206
295021K201
...(KA*S)
ANSWER
5.02
(1.00)
-
The Doppler Coefficient becomes more negative (0.50).
As the void fraction increases, the density of the moderator decreases
so that the neutron slowing down tame and length become longer, thus-
resonant absorption increases due to neutrons to spending more time yi
the resonant energy, spectrum. @(0.50)M M +) Q( d b
NelMcN
D
c4
J
bouw At.se
w .4 ew h
&u
REFERENCE
M
4
LESSON PLAN ON REACTOR PHYSICb PG 168.
LEARNING OBJECTIVE NO. 19.C.
GE BWR SERIES ON REACTOR THEORY.
2.6
...K/A VALUE
292004K111
...(KA*S)
ANSWER
5.03
(1.00)
a
REFERENCE
LESSON PLAN ON REACTOR PHYSICS PG 216 - 226.
LEARNING OBJECTIVE NO. 21.
GE BWR SERIES ON REACTOR THEORY
3.2
...K/A VALUE
292006K107
...(KA*S)
!
.
-
.
.
.
Ex _IHEQ6Y_QE_Nyq(EQS_EQWEB_E(QNI QEE8611Qdt_E(9,1Q@t_QND
PAGE
26
.
IHggdggyN@dlQS
1
ANSWERS -- LASAL.LE 1&2
-88/04/26-NRC - REGION I
l
ANSWER
5.04
(3.00)
l
a.
Reactor pressure decruases (0.50).
Increased voiding reduces
reactor power cousing less steam to be generated.
As the steam
flow through the steam lines decreases the frictional headlosses
decrease.
(0.25) Since EHC maintains a relatively constant
pressure at the equalizing header, reactor pressure decreases.
1
(0.25)
l
b.
Indicated level increases (0.50) due to increased voiding in the
core (causing increased flow resistance) (0.25) and due to a lower
suction flow being taken on the annulus. (0.25)
e
c.
Loop A Jet pump flow increases (0.50) due to the revers 9 flow
throuph Loop B jet pumps. (0.50) (04 dw /c A4deece/ 4%ch g oce<.s/
kW wM bel/W YJ un fttCo)
REFERENCE
LESSON PLAN ON EHC CHAPT 26.
LEARNING OBJECTIVE NO. 14.A.
- '
LESSON PLAN ON RWLC CHAPT 31.
LEARNING OBJECTIVE NO. 14.A.
)
LESSON PLAN ON RECIRC CHAPT 5.
LEARNING OBJECTIVE NO.
3.B.
LOA-RR-06 PG 1.
3.9
3.7
3.6
...K/A VALUE
202001K303
202001K304
295001K301
...(KA'S)
i
i
l
i
l
--
---_-.
. - ._.
- _-
-
.___---
_.
...
.
, Et,_ISEQBy_QE_NyC(EQB_EQWEB E(QUI _QEEBQIlQUz_E(ylDSg_QND
PAGE
27
.
IMEQdQDINQUICS
ANSWERS -- LASALLE 1&2
-80/04/26-NRC - REGION I
Q
tQ,4
h y&ggg ga &4 g & qqf
gd/,y/,f"j,g])
a.
Feed flow is rapidly dropping to zerol therefore, core inlet
subcooling is decreasing; adding negative reactivity.
b.
Reactor scram on Low reactor level (12.5 inches)
c.
BPVs are fully open to reduce pressure to 920 psig and are
removing more heat f rom the reactor than is being(gen
ted
(decay heat or residual heat from stored en r,gy) b
g4
a
Wo"-
d.
RCIC and HPCS begin injecting.
W
6
Turbine Bypass Valves are open to control reactor pressure,
e.
(and the main turbine has tripped).
(5 required, 0.50 ea.)
REFERENCE
LESSON PLAN ON HPCS CHAPT 36 PG 14, LEARNING OBJECTIVE NO.
9.A.
LESSON PLAN ON RCIC CHAPT 41 PG 22, LEARNING OBJECTIVE NO.
9.A.
LESSON PLAN ON EHC CHAPT 26 PG B AND 16, LEARNING OBJECTIVE NO.
6.A.
LESSON PLAN ON APRM CHAPT 14 PG 26, LEARNING OBJECTIVE NO. 9.
GE BWR SERIES ON REACTOR THEORY.
-
3.9
4.1
4.1
4.4
...K/A VALUE
295031K202
295031K204
295031K207
295031K211
...(KA*S)
'
ANSWER
5.06
(3.00)
a.
1.
Increase (0.50)
Increasing the subcooling increases the heat
i
removal rate at the clad surfaces therefore, the bundle
power required to cause transition poi 1}ng at the c19d sur,f ac
will increase
Flvu .A42f w
.
d4 ti M 4A4
& h C & d. (0.50)/f40? Yo
'
/
_
a t*?s-k l W 1449pressureincreasestheam&a)
'
/&
2.
Decrease (0.50)
As reactor
@nt of
l
heat which must be added to the coolant to cause vaporization
'
decreases; therefore, the bundle power required to cause the
onset of transition boiling at the clad surface decreases.
(0.50)
b.
The Critical Power Ratio, CP/AP, will decraase (0.50) because an
increase in core flow results in a larger increase in the actual
power of a bundle than the increase in critical power of the
I
bundle. (0.50)
REFERENCE
LESSON PLAN ON CORE THERMAL HYDRAULICS PG 29 - 32, 36 - 30.
l
LEARNING OBJECTIVE NO. 6.C AND 5.F.
GE BWR SUR!ES ON HEAT TRANSFER AND FLUID FLOW SECTION 9.
l
3.7
3.6
3.3
3.2
...K/A VALUE
<
1
4
- - - -
.
.
,-
._
h__IBEQBY_QE_M9G6EeB_EQWEEEbeNI_QEESQIlgdi_E(ylQ5_6BQ
PAGE
28
t
.
.
ISEBdQQ1Ned1GE
-
ANSWERS -- LASALLE 1&2
-80/04/26-NRC - REGION I
i
1
293009K118
293009K119
293009K122
293009K124
...(KA*S)
!
ANSWER
5.07
(2.00)
m.
Larger
b.
Smaller
c.
Larger
d.
Smaller
'
(4 raquired, 0.50 ea.)
REFERENCE
LESSON PLAN ON REACTOR PHYSICS REVIEW PG 102.
LEARNING OBJECTIVE NO. 19D.
GE BWR SERIES ON REACTOR THEORY.
2.9
...K/A VALUE
'
292005K112
...(KA*S)
- '
ANSWER
5.08
(2.50)
a.
4
b.
5, 6
.
c.
3
)
d.
2
(0.50 ma.)
l
REFERENCE
LESSON PLAN ON CORE THERMAL HYDRAULICS SECTION ON CORE THERMAL LIMITS
i
PG 22.
LEARNING OBJECTIVE NO.
4, 5.F, AND 6.
-
3.6
3.4
3.6
3.5
3.6
...K/A VALUE
i
293009K107
293009K108
293009K111
293009K112
293009K119
...(KA'S)
I
i
r
!
.
f
I
f
!
i
i
+
f
._
..
-
. _ . .
_
_
.--
.._
.
.
Qu__IBEQBY_QE_UUQ6 EBB _EQWEB_ELBNI_9EEBBI19Bi_EbW19Ei_8NQ
PAGE
29
-
.
'
16E659910601gS
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
5.09
(3.00)
CONDITION
AVAILABLE NPSH
REQUIRED NPSH
a.
Feedwater injection
Decreases
Not affected
temperature increases
(0.50)
(0.50)
/
la
b.
Reactor pressure
creases ,Dedreases
Not affected
(0.50)
(0.50)
c.
Recirculation FCV (Flow
Decreases
Increases
Control Valve) in slowly
(0.50)
(0.50)
throttled open 10%
REFERENCE
LESSON PLAN ON LASALLE FLUID FLOW AND HEAT TRANSFER PG 64 - 70.
LEARNING OBJECTIVE NO. 13.
3.3
3.4
2.8
...K/A VALUE
202001K607
202001K609
293006K110
...(KA'S)
- -
s
ANSWER
5.10
(2.00)
'
!
a.
During low power operation /high rod density the flux profile is
TOP PEAKED (0.50) necause there is little void production (0.25)
and the contrcl rods are significantly inserted into the core
depressing the flux in the lower region. (0.25)
b.
During high power operation / low rod density the flux profile is
BOTTOM PEAKED (0.50) due to the large void f raction in the upper
area of the core (0.25) and the rod density in the lower
region of the core has been significantly reduced. (0.25)
,
1
REFERENCE
i
!
LESSON PLAN ON REACTOR PHYSICS PG 134.
LEARNING OBJECTIVE NO. 19A.
3.4
3.3
3.8
3.2
...K/A VALUE
201003K503
292005K110
292000K118
292008K119
...(KA*S)
,
i
..
.
.
E___ISEQBY_QE_NQQLg68 EQWE6_E66dl,Q&EEQIlQUg_E(ylQQg,8ND
PAGE
30
.
.
ISEBd991NQdl(@
-
ANSWERS -- LASALLE 1&2
~BB/04/26-NRC - REGION I
,
ANSWER
5.11
(2.50)
Early in the cycle burnable poisons are depleted faster than the
a.
fuel thereby requiring rods to be inserted to reduce core
reactivity (or go maintain power no greater than 10,0*/. rated).
'
(1.00) [ N6ft ! ru. .23 9
u:L% 6 M/ a
- s g4
b.
The void coefficient INCREASES (0.50) because rod insertion has
reduced the ef f ective size of the core and more thermal neutrons
will leak into/be absorbed by the control rods (1.00)
REFERENCE
LESSON PLAN ON REACTOR PHYSICS PG 144 AND 204.
LEARNING OBJECTIVE NO. 17.D AND 23.C.
3.7
3.0
...K/A VALUE
,
201003K507
29002K503
...(KA*S)
L
~
0
1
i
l
1
. - _ .
_
-
-
. _ .
--- .
,
_. _
ks_ELQUI_EYEIEdE_QEllGMs_QQUIBQLs_QUQ_lMQIBQd(W1611QG
PAGE
31
-
,-
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
6.01
(2.50)
i
Half-scraga Rod 0$$c :.,
a.
6.
M
,
_ _ . . , , , , ,
c.
Half-scram
,
d.
Rod block
(0.66 x 55% + 42 = 78.3%)
l
e.
No action
~
t
(5 required, 0.50 ea.)
REFERENCE
LESSON PLAN ON RPS CHAPT 20 FIG 20-9 AND FIG 20-10.
LEARNING OBJECTIVE NO.
3, 4, 6, AND 11.
LESSON PLAN ON APRM CHAPT 14 PG 48.
LEARNING OBJECTIVE NO. 14 AND 15.
'
3.8
3.7
4.1
3.7
3.7
...K/A VALUE
212000K305
215005K401
239001K127
245000K104
245000K307
)
...(KA*S)
i
.
-
ANSWER
6.02
(3.00)
a.
1.
CST suction remains open and the Suppression Pool suction
,
remains shut.
!
l
2.
CST suction valve opens and the Suppression Pool suction
remains shut.
r
3.
CST suction remains shut and the Suppression Pool suction
remains open.
4.
The HPCS test valve closes.
(4 required, 0.50 ea.)
b.
2
(1.00)
REFERENCE
LESSON PLAN ON HPCS CHAPT 36 PG 20.
LEARNING OBJECTIVE NO. 6.A.2 AND 6.A.4.
1
LOP-HP-04 PG 2.
I
3.7
3.3
3.8
3.3
...K/A VALUE
209002A101
209002A108
209002A201
209002A301
...(KA*S)
1
. -
-
-
,
.__
.
,
,. kLa_EkeNI EYRIENE_QEEIGNi_GQUIBQLs_QSQ_INEIBWNENI6IlQN
PAGE
32
ANSWERS -- LASALLE 1&2
-38/04/26-NRC - REGION I
l
-
F
h
ANSWER
6.03
(1.00)
f
d.
>
,
REFERENCE
l
LESSON PLAN ON PROCESS RAD CHAPT 72 PG 7, B, AND 9.
LEARNING DBJECTIVE NO. 2.A AND 3.
i
3.8
3.6
...K/A VALUE
!
239001K401
272000K101
...(KA'S)
i
ANSWER
6.04
(2.00)
a.
2
(1.00)
(OR E)
b.
1
(1.00)
}
REFERENCE
4
'
LESSON PLAN ON RWLC CHAPT 31 PG 28 AND 30.
LEARNING OBJECTIVE NO. 12.
-
3.8
3.4
3.4
3.8
...K/A VALUE
259002A101
259002A201
259002A202
259002K301
...(KA'S)
1
-
,
ANSWER
6.05
(3.50)
'
a.
1.
Increase
!
1
2.
Decrease
I
3.
Decrease
j
(3 required, 0.50 ea.)
1
'
b.
Reversed (0.50), Higher (0.50)
i
f
c.
1.
Increased blowdown area in the event of a
desi gn-b asi s-acci d ent (LOCA)
f
l
2.
Reduces the capability of care reflood (above 2/3 core height)
j
(2 required, 0.50 ea.)
n
REFERENCE
4
LESSON PLAN ON VESSEL INTERNALS CHAPT 2 PG 32.
LEARNING OBJECTIVE NO. 3.E AND 5.B.
,
!
TECH SPEC DASES PG 4-1.
SIMULATOR MALFUNCTION NO. 199.
.
3.9
3.7
3.7
...K/A VALUE
j
202001A201
202001G010
202001K601
...(KA*S)
i
_
_ __ ___ _.
..
.
-.
-
. . . . _ , _
-
-
.-
.- _ _- , - _
__
._
ta Ek6dl EYEIEDE EEElldt EQU1696A 6dE idgIByd(UIBIlQN
PAGE
33
.
.
,
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
9
r
ANSWER
6.06
(3.50)
A
a.
1.
Closed
,
2
Open
3.
Closed
4.
Closed
(4 required, 0.25 ea.)
b.
1.
Continue automatically
2.
Requires local operator action
3.
Requirem control room operator action
.
4.
Requires control room operator action
i
(4 required, 0.50 ea.)
c.
False
(0.50)
REFERENCE
LESSON PLAN ON RCIC CHAPT 41 PG 22, 42, AND 44.
- -
LEARNING OBJECTIVE NO.
3.A,
6.D, AND 9.
3.7
3.7
3.0
3.5
3.3
...K/A VALUE
217000A201
217000A202
217000A209
217000A302
217000K402
...(KA*S)
-
ANSWER
6.07
(2.50)
a.
1.
No
2.
Na
3.
No
4
Yes
j
(4 required, 0.50 ea.)
1
b.
Falso
(0.50)
i
REFERENCE
l
i
LESSON PLAN ON ADS CHAPT 37 PG 28, 35, FIG 37-2 AND FID 37-4.
l
LEARNING OBJECTIVE NO.
6.A,
9, AND 10.
'
4.0
4.0
3.8
4.1
3.6
...K/A VALUE
219000K402
210000K403
218000K501
21BOOOK601
218000K606
...(KA*S)
,
W
--
- -
- .
-
. -
. _
,
.
6sa_EbeUI_EXEIEdE_QEElGWz_Cgulag6t_suD_1gEIByugglellgu
PA~E
34
,.
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
6.08
(2.25)
PK(d.deo MC'
8
When red densit,y Uisrester tha
7%
(
.50),
a.
- )
0 pack W v
a-
1
,
b.
In the RODS FREE position the amber (0.125) LEDs for all the rods
assigned to the selected RSCS group (0.25) which are allowed to
move (0.25) in the direction selected by the SELECT button will be
lit. (0.125)
c.
20% rated power (0.50)
as sensed by turbine first stage pressure. (0.50)
REFERENCE
LESSON PLAN ON RSCS CHAPT 19 PG 5,
11, AND 12.
I
LEARNING OBJECTIVE NO. 2.B, 2.E, 3, 4 AND 6.
3.6
3.2
3.3
3.4
...K/A VALUE
,
201004A201
201004K402
201004K404
201004K604
...(KA*S)
'
i
- '
ANSWER
6.09
(1.00)
1
IRM C will read 35 on range 7 (0.25)
IRM rod block setpoint is 100/125 scale (0.25)
The rod block on range 7 occurs at 100/125 x 40 = 34.6
(0.25)
i
IRM C initiates a rod block (since 35 > setpoint of 34.6) (0.25)
REFERENCE
LESSON PLAN ON !RMS CHAPT 12 PG 26.
LEARNING OBJECTIVE NO. 6.A.2 AND 9.
I
3.3
3.7
...K/A VALUE
215003A401
215003K401
...(KA*S)
a
,
i
l
1
,
J
-
-
-
.
.
.
.
l
,. Isa.&k6dl_IIElEUE_EEElEdt GQdl6Qbt 6dE 1dElRQM(NIQIlQN
PAGE
35
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
,
'
i
i
l
ANSWER
6.10
(1.75)
a.
1.
D/W pressure (1.69 psig)
2.
Low reactor level
(- 50 inches)
,
3.
High radiation building exhaust duct (10 mr/hr)
j
4.
High radiation fuel pool exhaust duct (10 mr/hr)
5.
Steam tunnel high differential pressure
6.
BBGT auto start pushbuttons depressed
j
(5 required. 0.25 ea.)
b.
Steam tunnel high differential pressure
(0.50)
,
REFERENCE
'
LESSON PLAN ON PCIS CHAPT 49 PG 19 AND 20.
LEARNING OBJECTIVE NO.
1.
LESSON PLAN ON R.B. VENT CHAPT 60 PG 9 AND 16.
LEARNING OBJECTIVE NO.
9.
3.6
3.8
4.0
3.8
3.5
...K/A VALUE
261000K101
261000K401
29001A301
29001G007
29001kl01
- '
...(KA*S)
.
ANSWER
6.11
(2.00)
I
a.
Rod 32-35 must be inserted (1.00)
b.
3
(1.00)
REFERENCE
LESSON PLAN ON RWM CHAPT 18 PG 14, 15, 16, AND 29.
LEARNING CDJECTIVE NO.
6.B.
3.3
3.5
3.5
3.4
...M/A VALUE
201006A205
201006K401
201006K402
201006K403
...(KA*S)
l
.
,
.
Zs..BBQGERWEES. .UQBdebt.etNQBdekt EdEEEEUGY.QUE
PAGE
33
.
,
'
'
BeQ196QQ1G86.GQUIB96
,
1
ANSWERS -- LASALLE 1&2
~68/04/26-NRC - REGION I
'
>
I
l
ANSWER
7.01
(2.00)
LGA-ATWS-01 ATWS POWER CONTROL is entered (0.50) due to:
- 1.
PSV pr: ru : 21 1090 pe69--44rGM- 4M
g, , 5 0
2.
failure t
scram (Beactor p
PRIMARYCqNTAINMENTCONTROLisente,edas
(mat '. WCLl crek
IO S'O Ps
n apwn etw
LGA-ATWS-03 ATW
r
(0.50) due tot
,
1.
A Suppression Pool temperature of 100 deg. F (0.25) and
2
LGA~R14U-c/t/srerfJLGA-ATWS-01 has b en e/nteredlc(0.25)L
lb]ste :.
eur r
at
,
i
REFERENCE
LGA-ATWS-01 AND -03.
LESSON PLAN ON LGA.
LEARNING OBJECTIVE NO.
6.
4.7
...K/A VALUE
295037G011
...(KA*S)
- '
ANSWER
7.02
(3.00)
a.
1.
Whenever indicated level is below +151 inches on the shutdown
range level indicator (0.50).
2.
Whenever D/W temperature is above 100 deg. F (0.50)
,
,
b.
1.
When directed by an LGA caution to stay above 57 psig (0.50).
2.
Misoperation in the automatic mode (0.25) is confirmed by at
3
i
least two independent indications (0.25)
c.
1.
To prevent bearing damage (due to a lack of lubrication by the
j
shaft driven oil pump) (0.50)
2.
To prevent damage to the exhaust check valve (0.50)
(Also accepts
to prevent intermittent exhaust steam flow and
water hammer in the exhaust line.)
,
REFERENCE
'
LGA-GP
LESSON PLAN ON LGA PG 4 AND 5.
LEARNING OBJECTIVE NO. 1 and 4.
4.7
...K/A VALUE
294001A116
...tKA'G)
.
-
-
-
-
-
-
-
-
.
. s
.
1
.
\\
-
I
' Zz__P8QCEDUBES_;_NQRM@62_6BNQRM@6t_EdE8GENQY_@ND
PAGE
37
-
l
'
88 Dig 6QGIC86_CQNIBQ6
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
-
\\
l
ANSWER
7.03
(3.00)
a.
Venting the D/W under these conditions may remove noncondensibles
from the primary containment which would result in an implosion of
the D/W caused by a rapid collapse of the steam in the D/W (with a
lack of noncondensibles in the D/W) when the procedure LATER
directs i.mitiation of the D/W sprays. (1.00)
b.
A level of -275 inches is the (Minimum-Zero-Injection) level at
which core cooling is no longer assured by the heat transf er f rom
the clad to the coolant; thus, opening an SRV at this point
INCREASES the steam flow through the fuel assemblies to absorb
more heat from the fuel ensemblies.
(Opening an SRV before level
drops to -275 inches results in less efficient steam cooling.)
(1.00)
c.
The recire FCVs are runback to minimum prio,
'c tripping the
.
recirc pumps in order to provent tripping the main turbine due to
the swell in RPV level.
(This prevents the removal / loss of a
- '
major heat tink during an ATWS.) (1.00)
REFERENCE
LESSON PLAN ON LGA PG 6,
20, 28, 29, AND 37.
LEARNING OBJECTIVE NO.
9.A,
10.B, AND 12.A.
3. 4
3.9
4.2
4.2
4.2
...K/A VALUE
293OO9 GOO 7
295024 GOO 7
295024K101
295037K209
295037K301
...(K4'S)
i
l
l
l
._.,
-
-_.
-
- . _ .
..
_ _ . . _ _ _ _ _ , _ _ _ . - . . _ _ _ _ _ _ __ _
.
.
.'
7. PROCEDURES - NORMAL 1_ABNQRMALz_EMERGENQ1_AND
PAGE
30
.
B8 Dig 6QQlC66_CQNIB06
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
7.04
(2.00)
a.
When the Suppression Pool level is<
-8 inches most of the RTDs
which measure the Suppression Pool temperature are no longer
covered with water and will not read the correct relative
temperature of the Suppression Poal.
(1.00)
6.
1.
RHR temperature recorders on P601 if the RHR system is in
service.
2.
Contact pyrometer on the suction piping of any ECCS pump
taking suction on the suppression pool.
3.
If HPCS is running, by placing a temperature gage in the
temperature well of the HPCS pump suction.
j
4.
If RCIC is running, by checking the local RCIC pump discharge
tempera ure indicator.
-
_
.
[ 4t2 @irnd,M- %
NW mms an
kM
S'
(2 requ
0.50 ea.)
MMm
.)
\\
REFERENCE
'
- '
LOP-CM-03 PG 3,
6,
AND 7.
LEARNING OBJECTIVE NO.
3.9
3.9
...K/A VALUE
295030A202
295030 GOO 7
...(KA*S)
.
ANSWER
7.05
(2.00)
a.
Maintaining > 200 KVAR while in paralle
with the grid prevents
reverse power tripping of the Diesel Generator due to large load
changes on the grid. (1.00)
b.
1.
Frequency 60 hz
(+/- 0.5 hz)
2.
Voltage 4160 volts
(+/-
150 volts)
3.
Diesel Generator Cooling water pump starts (1DG01P)
4.
Diesel Generator output breaker did not close
1
(4 required, 0.25 ea.)
REFERENCE
LESSON PLAN ON D/G CHAPT 47.
LEARNING OBJECTIVE NO.
6.C.
LOP-DG-02 FG 3 AND 15,
3.1
3.6
3.1
3.2
3.6
...K/A VALUE
264000A109
264000G010
264000201
264000301
264000306
...(KA'S)
-
-
- - -
_ . . - -
. .
.
. . .
.
?
7-.
PROCEDURES - NORMAL _ ABNORMAL _E_('!E_B@ENCY_AND
PAGE
39
2
g
,
BOD 1969GICB6_CQNIBQ6
-
ANSW4RS -- LASALLE 1&2
-88/04/26-NRC - REGION I
1
,
ANSWER
7.06
(3.00)
l
i
a.
Ensuring vessel level is at or above 40 inches prevents the l evel
fluctuations in the downcomer caused by the pump startup frore
'
resulting in RPS/PCIS initiations at 12.5 inches.
(1.00)
b.
Slowly cutting in the lHR heat exchanger prevents thermally
stressing the SDC return nozzles
(because the RHR heat e>tc}W/
2nger
byhd4
cannot tse pr
wa
e ).
O)
-
'
-
c.
1
st one recirc pump
s opera ing
n the loop thi.c is not
aligned for SDC.
2.
With no recire pumps running, SDC flow > 6000 gpm
(+/- 500 gpm)
3.
RPV level is above +578 inches (+50 inches on S/D Range)
(+/- 5 inches)
4- maiG 1+4% Ju)M/ss >ar
,
(2 required, 0.50 ea.)
REFERENCE
_-
8dD 4.o&ff- 64 g 3.
LOP-RH-07 PG 2,
3,
4, AND 5.
3.4
3.3
3.6
a.4
3.2
3.2
...K/A VALUE
.
205000A105
205000 GOO 7
205000K102
290002G010
290002K603
290002K611
...(KA'S)
ANSWER
7.07
(1.00)
Tripping the main turbine prior to injecting with RCIC prevents
carryover from the RC.ip spray from damaging the turbine.
(1.00)
REFERENCE
LOP-RI-02 PG 2 AND 4.
3.5
3.1
...K/A VALUE
217000G010
245000K502
...(KA*S)
- _ _
_
- _ _ . _
. _ _ _ _ _
_
_ _ . .
_ _ _ _ . . _ _ _ _ _ .
_ _ _
_
-
_
____
_ _ _ _ _ _ _ _ _ _ _ _ _
. _ _ _
_
.
,
Z:. _PRQCEDUBES_;,, i i@6z_@BNQ80@61_EME8GENCy,_@ND
PAGE
40
)
-
.
!
'
BBDig6QGLC@L,je g896
i
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
l
<
l
ANSWER
7.08
(2.00)
1
1.
Verify reactor scram.
2.
Initiate RCIC.
.fo
{
3.
Veri
SRVs open (0.25)Amaintain pressure between 900 and 1000
psig
(0.25)
1
4.
Attet t to start the Diesel Generators.
l
5.
Notify the Enift Supervisor.
(4 required, 0.50 ea.)
REFERENCE
LOA-AP-OB PG
1.
4.1
...K/A VALUE
295003G010
...(KA*S)
ANSWER
7.09
(2.50)
'
- . -
a.
1.
If reactor power is approaching a safety setting.
,
2.
If reactor pressure reaches 1043 psig
(+/-
10 psig)
l
3.
If a MSIV isolation closure occurs. p
4.
If reactor pressure -- tu-bire 1 eu
trolled
6,
/?ff.P1 W
(. //fW Cr* * bblJ h5l[can ot ble c //.7< S*$').
f
&. '?F W ' h"O "E O Ja a M
'
.
b.
By use of the Load Set (or thu Load Limit) (0.50)
c.
By use of the Max Combined Flow Limit (0.50)
REFE
LOA-
01 PG 2.
3.8
3.8
3.8
3.7
...K/A VALUE
295007A105
295007G010
295025A102
295025G010
...(KA'S)
1
-
,
4
.
,
ZL_E8QQEDUggS_ _NQ8d@6t_QBNQBd@61_EdE8GENQY_@ND
PAGE
41
-
BBDiQ6QGlg@6_QQNIBQ6
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
]
ANSWER
7.10
(2.00)
a.
Reduce power by reducing reactor retire flow (0.25) to about 63*/.
power
(0.25)
b.
1.
Reactor level above the low level al arm (0.25) 31.5 inches
(0.25)
2.
Condensate polisher differential pressure (0.25) < 60 psig
(0.25)
3.
Reactor feedpump suction pressure above the low suction
pressure trip point. (0.50)
REFERENCE
LOA-HD-01 PG 1 AND 2.
3.1
3.4
3.2
...K/A VALUE
295009A103
295009 GOO 7
295009K203
...(KA'S)
1
-
ANSWER
7.11
(2.50)
1.
Reactor scram (0.25) on turbine stop (control) valve closure.
(0.25)
,
2.
RPT downshifts recirc pumps to slow speed (0.25) on 12.5 inches
reactor water level (0.25)
3.
SRVs open (0.25) when reactor pressure is above 1076 psig.
(+/- 50
psig) (0.25)
4.
Low Low Set (LLS) actuates (0.25) with two or more SRVs
open. (0.25)
5.
Main generator trips (0.25) on turbine trip (or electrical fault)
(0.25).
6.
Reactor feed pumps trip (0.25) on Reactor Water Level B.
(0.25)
7.
Recirc
ump
trip (0.95) at 11,3
psig reactor pressure. (0.25)
b
N ufdM lt2N
+
t
,a
REFERENCE
LOA-TG-06 PG 2. p g
g
J.
3.3
3.6
3.6
3.6
3.9
3.3
3.7
...K/A VALUE
295005A101
295005A102
295005A105
295005G010
295005K201
295005K203
295005K207
...(KA'S)
1
i
6
)
.
_
._.
_
_
_ - - _ . - _
- .,_ --
.
.
, .' 0 __8DdlN1SIg81[VE_PgggEDUBES _ggNDLIlgNS _@ND_ Lid [I@IlgNg
PAGE
42
-
2
z
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
8.01
(2.00)
a.
IS a core alteration
b.
IS NOT a core alteration
c.
IS NOT a core alteration
d.
IS NOT a core alteration
(4 required, 0.50)
REFERENCE
TECH SPEC DEFINITIONS.
4.2
4.1
3.9
...K/A VALUE
290002 GOO 5
290002G011
95023GO11
...(KA'S)
ANSWER
B.02
(1.50)
a.
True
5.
"21. gr.Wou2L
- '
c.
False
(3 required, 0.50 ea.)
.
REFERENCE
TECHNICAL SPECIFICATION DEFINITIONS.
3.3
3.4
3.3
3.3
...K/A VALUE
216000G010
216000G012
272OOOG010
272OOOG012
...(KA'S)
l
ANSWER
8.03
(1.50)
a.
WILL EXCEED
b.
WILL EXCEED
1
c.
WILL NOT EXCEED
i
(3 required, 0.50 ea.)
REFERENCE
'
LAP-100-17 PG 2 AND TECH SPEC PG 6-2.
3.7
...K/A VALUE
294001A101
...(KA*S)
, , _ _ . _ . - _ . - - . .
-
_ - .
_ . .
, . - .
. ._. . . _ . . .
....__,_....- , .
.
~
.
, ' 0 t_8DdlNISI66IlVE_PRQQEDUBESz_CQNDillONS _6ND_LidlI6110N@
PAGE
43
.
2
.
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
]
i
l
ANSWER
8.04
(1.50)
An individual holding a valid SRO license (Shift Foreman
designated as plant SRO) (0.50) AND another individual qualified
as an LSCS Shift Technical Advisor (0.50) is WITHIN 10 minutes of
the Control Room. (0.50)
REFERENCE
LAP-2OO-3 PG 4 AND 5.
3.7
...K/A VALUE
290701A103
...(KA*S)
ANSWER
8.05
(2.00)
1.
The final decision to declare an emergency condition.
2.
The decision to notify and rncommend protective actions to offsite
authorities (when the Manager of Emergency Operations or Corpurate
Command Center Director have not been contacted or are not
prepared to make an informed decision).
-
3.
The decision to authorize personnel exposure beyond 10CFR2O
limits under emergency conditions.
4.
The decision to request help from the Department of Energy.
-
(4 required, 0.50 ea.)
REFERENCE
LZP-1110-1 PG 2.
4.7
...K/A VALUE
294001A116
...(KA*S)
ANSWER
8.06
(3.50)
i
General Emergency (0.50)
(5.0 EB uti.sec at 10 mph) LZP-12OO-1 and
LZP-12OO-2 (0.25)
'
EVec4&f ggacugte O to 2 mile
(0.50) radius (0.25)
_, , 1 t . , %2 to 5 mile (0.50) downwind sectors L, M, N (0.25)
Shelter 5 to 10 mile (0.50) downwind sectors L,
M, N (0.25)
Using Attachment B of LZP-12OO-5 for determining the protective
actions (0.25) and NARS FORM for determining the downwind sectors.
& EMk h fu $4E;Ift.y) h ,,
(REFERENCE 0 R.M A144
/
LZP-12OO-1 ATTACHMENT B PG 14.
LZP-12OO-5 AND LZP-1210-2.
4.7
4.3
4.5
4.5
...K/A VALUE
i
---
,_
.-,
-
_
-
. .
.
.,283_890lNJgIBOIJVE_PBgCEgUSEg2_CgNg]IJgN32_ sng _LidlIBIIgyg
PAGE
44
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
294001A116
295038A201
295038GO11
295039K301
...(KA*S)
ANSWER
8.07
(3.00)
a.
To ensure that MCPR does not be:ome less than the fuel cladding
safety limit (0.50) or that greater than
1*/. plastic strain does
not occur in the degraded situation. (0.50)
b.
Recirculation pumps are tripped to reduce core flow in order to
reduce the void collapse (0.50) (during two of the most limiting
pressurization events (Turbine trip / Load reject)) so as to prevent
the positive reactivity added by the void collapse from exceeding
the negative reactivity added by the control rod scram. (0.50)
c.
To contain fission products (0.50) and
to ensure the core is not uncovered f ollowing steam line breaks.
(0.50)
REFERENCE
TECH SPECS BASIS PG 2-3,
3-3,
AND 4-5.
4.1
3.8
4.0
...K/A VALUE
202OO1 GOO 6
202OO1K505
215005 GOO 6
...(KA'S)
~
.
ANSWER
8.08
(2.50)
A scram from power operations, transients with a potential for
a.
scram, or events which result in a violation of an LCO.
(0.50 for discussion of scram and 0.50 for discussion of LCO)
b.
10 minutes (0.50)
c.
The transient has stabilized (0.50) or
he has been properly relieved (0.5)
REFERENCE
LAP-2OO-1 AND LAP-2OO-5 PG 1 AND EB-01-020538.
4.7
...K/A VALUE
294001A116
...(KA'S)
--
- -
_ _ .
-.
.
__
.
.
-
. .
_ _ _ _ . _
-
.
. + .
.
1
.,.
8.x_ ADMINISTRATIVE PROCEDURES
CONDITIONS _AND LIMITATIONS
PAGE
45
2
x
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWER
8.09
(3.00)
a.
Immediately insert the cram rods. (0.50)
b.
The operator monitors power by selecting a designated set of
control rods for observing all LPRMs in the core (0.50), because
the APRM recorders may not respond fast enough to indicate a tual
.htteuu2p
8
power chan
(0.50). [d240
,
nei/4 b ~gesmp da .
&J g )Q]
fegffg
%y
'
'
'
>
a
c.
1.
(0.75) (J
&
Qpff
f
d.
The Surveillance Region is the thermal hydraulic instability
,
region in which operation of the plant may result in uncontrolled
oscillations of core flow and power.
(0.75)
REFERENCE
TECH SPECS 3/4.4.1 AND ASSOCIATED BASES.
3.6
3.9
3.4
3., 7
3.3
...K/A VALUE
3.6
3.7
3.6
202OO1K301
202OO1K303
202OO2A201
202OO2 GOO 6
202OO2G010
295001 GOO 7
295001G010
295001K305
...(KA*S)
-
,
ANSWER
8.10
(1.50)
'
1.
When instructed to do so by the Radiation-Chemistry
Department.
j
2.
Upon failure (or suspected failure) of personal protective
equipment.
3.
Unexpected deterioration of radiological conditions.
4.
In the event tat the workers current accumulated dose
equivalent status becomes uncertain or his dose equivalent is
equal to the that authorized for the job.
5.
The "assembly" siren sounds.
j
6.
Completion of work assignment.
7.
Injury.
8.
Unexpected radiation monitor alarm and the area dose rate is
unknown.
(6 required, 0.25 ea.)
REFERENCE
LRP-1000-1 PG 7 AND 8.
3.8
...K/A VALUE
294001K103
...(KA*S)
. ..
--
-
_- .
-
. - -
-
,
__--__
-
-
1
l
.
l
, , , 8 .
ADMINISTRATIVE PROCEDURE @z_CQNDITigNS _AND_LIMITATl@N@
PAGE
46
,
z
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
ANSWER
8.11
(3.00)
a.
Startup is not allowed (0.50) because T.S.
3.5.1 requires 6
operable ADS valves for mode 2 above 122 psig and per T.S.
3.0.4 a
mode cannot be entered dependent upon an action statement.
(T.S.
3.5.1.a action statement applies for one of the six ADS valves
being inoperable.)
6.
Startup is allowed (0.50) because the safety function of the SRV
is not inhibited as required by T.S.
3.4.2.
(0.50)
c.
Startup is not allowed (0.50) because T.S. 3.5.1 requires 6 ADS
valves to be operable for mode 2 above 122 psig
(per the
definition of operability all
Scessary attendant instrumentation
and controls must be operable
of which an entire logic division
is inop) AND because T.S. 3.3.3 requires ADS actuation instrumentation
to be operable in mode 2 and per T.S. 3.0.4 a mode cannot be entered
dependent upon an action statement (T.S. 3.3.3.c action statement
applies for one inoperable trip system).
(0.50)
REFERENCE
-
LESSON PLAN ON ADS CHAPT 37.
LEARNING OBJECTIVE NO. 15.
LESSON PLAN ON MAIN STEAM CHAPT 21.
LEARNING OBJECTIVE NO. 15.B.
TECH SPECS 3.4.2, 3.5.1,
3.0.4,
AND DEFINITIONS.
4.3
4.3
4.4
4.2
...K/A VALUE
21BOOOGOO5
218005G011
239002 GOO 5
239002G011
...(KA*S)
l
- --
_ _ - - - _ _ _ .
f ..
TEST CROSS REFERENCE
PAGE
1
)
,
.
.
QUESTION
VALUE
REFERENCE
.
05.01
2.50
RICOOOO985
05.02
1.00
RICOOOO986
05.03
1.00
RICOOOO987
05.04
3.00
RICOOOO988
05.05
2.50
RICOOOO989
05.06
3.00
RICOOOO990
05.07
2.00
RICOOOO991
05.08
2.50
RICOOOO992
05.09
3.00
RICOOOO993
05.10
2.00
RICOOOO994
05.11
2.50
RICOOOO995
.
25.00
06.01
2.50
RICOOO1007
06.02
3.00
RICOOO1008
06.03
1.00
RICOOO1009
06.04
2.00
RICOOO1010
06.05
3.50
RICOOO1011
06.06
3.50
RICOOO1012
06.07
2.50
RICOOO1013
06.08
2.25
RICOOO1014
06.09
1.00
RICOOO1015
06.10
1.75
RICOOO1016
1
_.
06.11
2.00
RICOOO1017
-_-_-_
25.00
.
07.01
2.00
RICOOOO996
07.02
3.00
RICOOOO997
07.03
3.00
RICOOOO998
07.04
2.00
RICOOOO999
07.05
2.00
RICOOO1000
07.06
3.00
RICOOO1001
07.07
1.00
RICOOO1002
07.08
2.00
RICOOO1003
07.09
2.50
RICOOO1004
07.10
2.00
RICOOO1005
07.11
2.50
RICOOO1006
\\
25.00
08.01
2.00
RICOOO1018
08.02
1.50
RICOOO1019
08.03
1.50
RICOOO1020
1
08.04
1.50
RICOOO1021
1
08.05
2.00
RICOOO1022
08.06
3.50
RICOOO1023
08.07
3.00
RICOOO1024
08.08
2.50
RICOOO1025
__._
..
_
,
_
-
.
a
.
s*
TEST CROSS REFERENCE
FAGE
2
-
.
p
,
.
QUESTION
VALUE
REFERENCE
00.09
3.00
RICOOO1026
08.10
1.50
-RIC0001027
08.11
3.OO
RICOOO1020
---_--
25.00
--
-- GuuB --tuuB
100.00
DOCKET NO
373
._
.
i
,
w--
-
, - .
m-
. .-.
,,-,m--c.----r
-
e.-_
_ - , , -
..---,----,-,.-~-,,_..s,
. , , , _ , - . . , - , - , - - - -
..,.-,.--_..--r-.&-.---,-,-.v
-.n,--.-.5.
- -
,-..w--
.
-
-
-
-
.
3 .
[d j '
, .
- .
U.
S.
NUCLEAR REGULATORY COMMISSION
,
REACTOR OPERATOR LICENSE EXAMINATION
FACILITY:
LASALLE 1&2
REACTOR TYPE:
BWR-GE5
DATE ADMINISTERED: 88/04/26
i
EXAMINER:
HRC - REGION I
__
CANDIDATE:
.
INSTRUCTIONS TO CANDIDATE:
Use
separate
paper for the answers.
Write answers on one side only.
Staple question sheet
on top of the answer
sheets.
Points for each
question are indicated in parentheses after the question.
The passing
grade requires at least 70X in each category
and a final
grade of at
least 80%.
Examination papers will be picked
up six (6)
hours after
the examination starts.
X OF
i
CATEGORY
X OF
CANDIDATE'S
CATEGORY
VALUE
TOTAL
SCORE
VALUE
CATEGORY
25.00
25.00
1.
PRINCIPLES OF NUCLEAR POWER
PLANT OPERATION, THERMODYNAMJCS,
l
HEAT TRANSFER AND FLUID FLOW
j
I
25.00
25.00
2.
PLANT DESIGN INCLUDING SAFETY
i
AND EMERGENCY SYSTEMS
1
25.00
25.00
3.
INSTRUMENTS AND CONTROLS
l
25.00
25.00
4.
PROCEDURES - NORMAL, ABNORMAL,
EMERGENCY AND RADIOLOGICAL
CONTROL
100.00
X
Totals
Final Grade
All work done on this examination is my own.
I have neither given
nor received aid.
""~
Candidate's Signature
1
1
.
-
---e-
-
.
.
. .
. . .
-
- . .
.
J P
'
.
,.
,
s
j
[ps;L ; AL~ L C4
g
.
s
'
'
.,
s,
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules apply:
1.
Cheating on the examination means an automatic denial of your application
and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may
leave.
You must avoid all contacts with anyone outside the examination
room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the
examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each
section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as
appropriate, start each category on a new page, write only on one side
of the paper, and write "Last Page" on the last answer sheet.
.
9.
Number each answer as to category and number, for example,
1. 4,
6. 3.
10. Skip at least three lines between each answer.
,
11. Separate answer sheets from pad and place finished answer sheets face
-
down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
'
13. The point value for each question is indicated in parentheses after the
j
question 3: ' can be used as a guide for the depth of answer required.
14.
SL ow all calculations, methods, or assumptions used to obtain an answer
to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16.
If parts of the examination are not clear as to intent, ask questions of
the examiner only.
17. You must sign the statement on the cover sheet that indicates that the
work is your own and you have not received or been given assistance in
completing the examination.
This must be done after the examination has
been completed.
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18. When you complete yrsur examination, you shall:
.
a.
Assemble your examination as follows:
(1)
Exam questions on top.
(2)- Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer
the examination questions,
c.
Turn in all scrap paper and the balance of the paper that you did
not use for answering the questions.
d.
Leave the axamination area, as defined by the examiner.
If after
,
leaving, you are found in this area while the examination is still
in progress, your license may be denied or revoked.
_
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1.'
PRXNCIPLES OF NUCLEAR POUER PLANT OPERATION,
PAGE
2
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOM
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QUESTION
1. 01
(2.00)
Reactor power is decreased from 90X to 50% in one hour by decreasing
Recirculation flow. No control rods are moved and no further change
in Recirculation flow is made (Recirc Pumps in Individual Manual).
a.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the decrease WHAT is reactor power (LESS THAM,
GREATER THAN or EQUAL TO 50% POWER)?
EXPLAIN WHY.
(1.00)
b.
60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> after the decrease WHAT is reactor power (LESS THAN,
GREATER THAN or EQUAL TO 50X POWER)? EXPLAIN WHY.
(1.00)
l
QUESTION
1.02
(2.00)
List four (4) reactor conditions or characteristics which influence
the point of criticality.
(2.0)
~
QUESTION
1.03
(3.00)
a.
STATE whether CRITICAL POWER will INCREASE, DECREASE, or REMAIM
THE SAME for each of the following changes.
EXPLAIN.
1.
Increased core inlet subcooling
(1.00)
2.
Reactor pressure increases from 930 psig to 980 psig
(1.00)
b.
STATE whether the CRITICAL POWER RATIO will INCREASE, DECREASE, or
REMAId THE SAME for an INCREASE in the total recirculation flow
rate.
EXPLAIN.
(1.00)
.
(aeeau CATEGORY 01 CONTINUED ON HEXT PAGE ===ne)
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PRINCIPLES OF NUCLEAR POMER PLANT OPERATION,
PAGE
3
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
,
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QUESTION
1.04
(2.25)
Corrosion products build up on the outside of the fuel cladding.
What will happen to the following parameters (Increase, Decrease,
Remain the Same) as the corrosion products build up on the cladding?
EXPLA1H YOUR ANSWER.
(Assume that the Recirc Flow Control is in
Haster Manual and the operator takes no action)
(i.oo)
a.
Fuel centerline temperature.
( G 44-)
LM' 1,
b.
Reactor Recirculation Flow
s-.
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l. 00
c.
Core Thermal Power
GUESTION
1.05
(3.00)
STATE HOW DIFFERENTIAL ROD VORTH CHANGES (increases, decreases,
remains the same) for each of the conditions listed below when operating
at SOX power. EXPLAIN YOUR ANSWER.
,
a.
A rod is withdrawn from notch 02 to notch 20.
(1.0)
i
b.
Localized voiding of a region not previously volded.
(1.0)
'
c.
Control rod density increases
(1.0)
QUESTION
1.06
(3.00)
The Residual Heat Removal pumps are being used in Shutdown Cooling Mode.
HOW will AVAILABLE and REQUIRED Het Positive Suction Head for the Residual
Heat Removal pumps be affected by each of the following changes
(INCREASE, DECREASE, or NOT AFFECTED) ?
NPSH
AVAIL.
REQUIRED
a.
Reactor Water temperature increases
b.
Reactor Water level decreases
c.
RHR System flowrate decreases
,
(ma==n CATEGORY 01 CONTINUED ON NEXT PAGE asuom)
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PRINCIPLES OF NUCLEAR POSER PLANT OPERATION,
PAGE
4
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THEPHODVNAMXCS, HEAT TRANSFER AND FLUID FLOU
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QUESTION
1.07
(2.25)
a.
A reactor has a K effective of 0.91 and an initial countrate of
(1.5)
1000 cps on the source range monitors. Reactivity is added until
the count rate is stable at 6000 cps. WHAT is the new value of
K effective?
(SHOW ALL WORK)
b.
Concerning subcritical multiplication, answer TRUE or FALSE.
(0.75)
,
As Keff approaches 1.00 a larger change in neutron level occurs
'
for a given change in Keff.
QUESTION
1.08
(2.50)
j
During a reactor startup, the reactor is currently at 20 on IRH Range 3
on a 100 second period. The Point of Adding Heat is assumed to be
at 50 on IRH Range
8.
a.
How long will it be until the Point of Adding Heat is reached?
(1.5)
(SHOW ALL WORK)
'
b.
Assuming that the operator takes no actions, WHAT HAPPENS to
(1.0)
reactor period and reactor power once the Point of Adding Heat
is reached. EXPLAIN YOUR ANSWER.
,
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OUESTION
1.09
(2.00)
i
Why should an operator be extremely cautious when withdrawing
(2.0)
peripheral Control Rods when starting up the reactor 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
after a scram from 100% power?
(Your answer should explain WHY peripheral Control Rod Worth changes
due to
1.
fission product pcison behavior, and
,
2.
changes in the core flux profile)
QUESTION
1.10
(3.00)
For each of the following events, WHICH COEFFICIENT of reactivity will
act FIRST to change core reactivity and WILL the reactivity added by
the coefficient be POSITIVE or HEGATIVE.
a.
Control rod drop at power
(0.75)
,
!
b.
SRV opening at power
(0.75)
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c.
Loss of shutdown cooling (when shutdown)
(0.75)
d.
Hain turbine trips while at 30X power
(0.75)
(eaaae END OF CATEGORY 01 aanaa)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY ,3YSTEMS
PAGE
5
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QUESTION
2.01
(1.50)
WHAT are three (3) of the five (5) AUTOKATIC ACTIONS which ALWAYS
(1.5)
occur as a direct result of a Main Generator Lockout Relay
(86 Device) actuation on Unit 1?
l
GUESTION
2.02
(2.00)
LIST four (4) of the five (5) signals which will automatically
cause a recirculation pump to downshift from fast speed.
(2.0)
(Include Setpoints)
GUESTION
2.03
(2.00)
Concerning the relief valve LOW-LOW SET (LLS) function:
a.
WHAT is the purpose of the relief valve LOW-LOW SET function?
(0.5)
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b.
WHAT condition actuates LLS?
(0.5)
c.
HOW is the function of the relief valves affected by the
(0.5)
actuation of the LLS?
,
d.
The LLS actuates (only on MANUAL, only on AUTOMATIC, or
(0,5)
on EITHER MANUAL OR AUTOMATIC) operation of the relief valves.
(CHOOSE ONE)
j
GUESTION
2.04
(2.50)
For the following questions assume that the Recirculation System
I
is in Master Manual,
a.
HOW will the Recirculation System respond to a downscale failure (1.0)
of
"C"
APRM with the reactor at 60X power and on the
100'. rod line?
/
b.
What automatic action occurs when the flux controller signal
(0.5)
reaches iO6X?
c.
What is the alternate APRM input to the flux controller?
(0,5)
d.
How is the flux input to the flux controller switched from the
(0.5)
normal input to the alternate input?
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(ouaa* CATEGORY O2 CONTINUED ON NEXT PAGE anaau)
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2. .
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
6
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QUESTION
2.05
(3.00)
Concerning the Standby Liquid Control System (SBLC):
a.
In addition to decreased control rod worth, VHAT are four (4)
(2.0)
positive reactivity effects that the SBLC system is designed
to overcome?
b.
During normal power operations, HOW WILL the Reactor Water
(1.0)
Cleanup System respond to a Standby Liquid Control System
initiation? WHY?
QUESTION
2.06
(3.00)
Concerning the RCIC system.
a.
What are the normal and alternate water supplies for the RCIC
( 0. 5 )
pump?
b.
What signal will cause the alternate water supply valve to open? (0,5)
-
c.
When the alternate water supply valve reaches the full open
(1.0)
position, what three (3) valves will receive a close signal?
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d.
What are two (2) of the three (3) adverse consequencec of
( 1, 0 )
J
operating the RCIC turbine below 2100 rpm?
(nausa CATEGORY O2 CONTINUED ON NEXT PAGE unumm)
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
7
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QUESTION
2.07
(3.50)
Concerning the Emergency Diesels.
a.
For each condition listed below, state whether the
"O"
Diesel will
supply power to
1) bus 141Y
2) bus 241Y
3) neither bus 141Y nor bus 241Y
Consider each condition separately.
1)
-140" Reactor water level on Unit 1
(0,5)
2) 2 psig Drywell pressure on Unit 2,
ten (10) minutes later
(0,5)
there is an undervoltage on bus 141Y
3) An undervoltage condition occurs on both units simultaneously (0,5)
and then two (2) minutes later Unit 2 LPCI automatically
initiates.
4) During a surveillance with DG
"O"
aligned to 141",
an
(0.5)
]
undervoltage on Unit 2 occurs.
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b.
What are five (5) of the seven (7) diesel generator output
(1.5)
breaker close permissives which must be met before the DG "1A"
,
output breaker will close on its bus after an automatic start?
QUESTION
2.08
(3.00)
a.
The reactor is at 100% power with all MSIVs open. State the response
of ALL the MSIVs during each of the following conditions.
(Consider each condition separately. Assume no operator actions)
1.
the inboard MSIV on the
"C"
steam line slowly drifts
(0,5)
fully closed
2.
the
"B"
steam line flow sensor LOW pressure tap breaks off
(0,5)
exposing the dp cells to drywell pressure (assume a small
break, therefore constant drywell pressure)
3.
All feedwater pumps trip and level drops to
-60"
(0,5)
b. WHAT are the three (3) reasons for an automatic closure of the
(1.5)
MSIVs when t, team line pressure drops to less than 854 psig while
the mode switch is in RUN?
(manen CATEGORY O2 CONTINUED ON NEXT PAGE anonn)
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P.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
8
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QUESTION
2.09
(2.50)
a.
With the mode switch in STARTUP, UNIT 2 will scram if CRD
(2.0)
header pressure drops to less than 1157 psig for more than
10 seconds?
What is the reason for this scram?
Your answer should address:
(1) what is the basis for this scram
(2) what is the basis for the 1157 psig setpoint
(3) why this scram is bypassed when the mode switch is in RUN.
b.
According to LGP-i-1,
you should perform a coupling check after
(0,5)
withdrawing a control rod to position 48 by attempting to withdraw
the rod. WHAT FOUR (4) INDICATIONS would you receive if the rod were
UNCOUPLED?
OUESTION
2.10
(2.00)
-
For EACH of the following HPCS initial valve lineup conditions
indicate the FINAL position of the given valves following an
AUTOMATIC HPCS initiation:
,
i
a.
CST suction valve open. Suppression Pool suction valve shut.
(0.5)
b.
CST suction valve shut, Suppression Pool suction valve shut.
(0.5)
c.
CST suction valve shut, Suppression Pool suction valve open.
(0.5)
d.
HPCS full flow test downstream stop valve (FOli) open.
(0.5)
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(aaa*a END OF CATEGORY O2 ****m)
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3.
INSTRUMENTS AND CONTROLS
PAGE
9
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QUESTION
3.01
(3.00)
Concerning jet pump flow indication:
a.
HOW does the DIFFERENTIAL PRESSURE relate to FLOW through
(0.5)
the jet pump?
b.
If one sensing point is on the jet pump, WHAT is the other
(0,5)
sensing point used by the
differential pressure transmitter
to determine the differential pressure across the jet pump?
c.
During single loop operation, HOW are the individual loop
(1.0)
flows AUTOMATICALLY processed to calculate INDICATED total
core flow? WHY?
d.
WHAT three signals are used from each recirculation loop to
(1.0)
determine if the single loop total core flow indication is
used?
QUESTION
3.02
(3.00)
~
With the Unit operating at 75% power, an electrical fault causes
the Haximum Combined Flow Setpoint of the EHC system to drop to
minimum.
.
HOW WILL EACH OF the following RESPOND after the fault? VHY?
(Consider response through ONE HIHUTE after the fault.
Assume NO OPERATOR ACTION.)
ATTACHED FIGURE, EHC LOGIC, IS PROVIDED FOR REFERENCE
a.
Turbine control valve position
(1.0)
b.
Bypass valve ponition
(1.0)
c.
Reactor power
( 1. 0 )
(aaa*a CATEGORY 03 CONTINUED OH HEXT PAGE aueaa)
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3. ' INSTRUMENTS AND CONTROLS
PAGE
10
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QUESTION
3.03
(3.00)
STATE what will INITIALLY happen to the indicated reactor water
I
level (increase, decrease, remain the same) for each of the following
conditions.
a.
the equalizing valve leaks
(0.5)
b.
a rapid temperature increase occurs in the Reactor Building
(0.5)
near the transmitter
1
c.
the reference leg isolation valve packing glands leak
(0.5)
d.
a rapid decrease in vessel pressure occurs
( 0. 5 ) .
e.
a rapid increase in drywell temperature occurs
(0.5)
f.
the reactor scrams
(0.5)
GUESTION
3.04
(2.00)
When paralleling a diesel generator with the 4KV bus:
a.
WHAT do you use the governor control for
(1.0)
1.
before the output breaker is closed?
,
2.
after the output breaker is closed?
,
b.
WHAT do you use the voltage regulator adjust for
(1.0)
1.
before the output breaker is closed?
2.
after the output breaker is closed?
,
(moeaa CATEGORY 03 CONTINUED ON NEXT PAGE anewn)
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INSTRUMENTS AND CONTROLS
PAGE
11
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QUESTION
3.05
(2.50)
ANSWER the following questions concerning ADS initiating logic:
a.
With Unit i at 100% power, the channel A and channel
C manual initiation push buttons are rotated and
depressed.
WILL the ADS function occur?
WHY?
(1.0)
b.
If ADS is manually initiated, WILL the ADS SRV opening be
delayed by the 105 second timer?
(Yes/No)
(0.5)
c.
If an ADS blowdown is in progress, with all initiation
signals still present, WILL depressing the ADS logic reset
pushbutton switches reinitialize the 105 sec. timer and
close all ADS valves.
(Yes/No)
(0,5)
i
d.
EXPLAIN HOW a loss of BOTH Drywell Pneumatic Air (<160 psig)
AND the nitrogen bottles will NOT hamper the operation of
the ADS.
(0.5)
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(ouano CATEGORY 03 COV.TINUED ON NEXT PAGE enaea)
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INSTRUMENTS AND CONTROLS
PAGE
12
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OUESTION
3.06
(2.00)
A reactor startup is in progress on Unit 2 and the operator is
withdrawing rods to attain criticality,
a.
The following errors are being displayed by the Rod Worth
Minim 12er (RWM):
withdraw error
insert error
insert error
STATE the ACTION that must be taken by the operator to clear the
control rod block.
(1.00)
b.
MULTIPLE CHOICE
CHOOSE the ONE condition which will cause the RWM SELECT ERROR
light to be lit.
(1.00)
t
'
1.
WHENEVER one insert error exists and a rod other than the rod
causing the insert error is selected.
2.
WHENEVER the operator selects a control rod which will result
in an insert or a withdrav error.
..
3.
ANYTIME a rod block has been initiated by the RWM and the rod
selected is not one of the rods causing the block.
4.
AFTER the operator has withdrawn or inserted a rod which is
,
NOT in the presently latched RWM group.
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QUESTION
3.07
(2.00)
STATE whether each of the following Reactor Protection System (RPS)
scrams CAN or CAN NOT be bypassed? And if a scram CAN be bypassed.
I
DESCRIBE HOW lt can be bypassed.
(2.0)
I
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a.
APRM high flux or power
b.
MSIV closure
c.
Manual
d.
Turbine control valve fast closure
e.
Main steam line high rad
,
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(veeam
EGORY 03 CONTINUED ON NEXT PAGE aaaan)
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INSTRUMENTS AND CONTROLS
PAGE
13
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QUESTION
3.08
(3.00)
Concerning the Intermediate Range Monitoring System (IRMs):
a.
WHAT are the two (2) scram signals generated by the IRMs?
. ( 1, 0 )
b. WHAT are the two (2) conditions when the IRM scrams are
(1.0)
bypassed?
c.
If the mode switch is in STARTUP, what two (2) automatic actions (1. 0)
would occur if IRM A suffered a total loss of 24 VDC?
WHY would each of these actions occur?
I
GUESTION
3.09
(2.50)
Concerning the FWLCS Setpoint Setdown function:
,
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a.
EXPLAIN the function of Setpoint Setdown and HOW it is
accomplished?
(1.0)
1
b.
WHAT would be the resul t on water level ( INCR EASE,
_-
DECREASE, or NO CHANGE) if, the Setpoint Setdown logic
j
initiated sporadically (K11 contact energized) at a full
,
power condition.
WHY?
(Assume 3-element control.)
(1.5)
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(aanaa CATEGORY 03 CONTINUED ON NEXT PAGE aaswe)
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QUESTION
3.30
( 2. OO's
'
The Instrument and Service Air systems receive air from a common
set of three (3) air compressors,
a.
The Unit i station air compressor is lined up in the
'ON',
' Modulate' mode of operation.
If system demand is less
tha,n 60X capacity, VHAT action, in reference to mode of
operation, must be taken and WHY?
(1.0)
b.
If Instrument Air were completely lost, in WHAT position
would each of the following valves fall?
(1.0)
'
a'
1.
scram inlet valve
2.
reactor water cleanup filter /demin. inlet and outlet.
valves
3.
outboard MSIV valves
4.
Turbine Building Closed Cooling Water Temperature
Control Valve
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(manaa END OF CATEGORY 03 maeea)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
PAGE
15
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RADIOLOGICAL CONTROL
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GUESTION
4.01
(2.50)
According to Procedure LOA-GP, General Precautions:
a.
WHAT precautions must be considered PRIOR TO placing an ECCS
system in manual?
(1.5)
b.
WHAT precautions must be taken WHILE an ECCS system is in
manual?
WHY?
(1.0)
OVESTION
4.02
(2.50)
The plant is operating at power when an SRV inadvertently opens.
As per LOA-NB-02, The Stuck Open Safety Relief Valve, the operator
cycles the SRV control switch from AUTO to OPEN and back to AUTO.
a.
If this action does NOT close the SRV, WHAT other method
can be performed in an attempt to close the s' tuck open
,
valve?
(0.5)
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b.
WHAT are two (2) control room indications the operator
would have if the valve closed?
(0.5)
c.
WHAT three (3) conditions woul d require the operator to
manually SCRAH the plant if the SRV remained open?
(1.5)
QUESTION
4.03
(3.00)
Procedure LOP-TG-02 Turbine Generator Startup, has several precautions.
EXPLAIN the reason for each of the following precautions:
a.
WHY should the turbine overspeed trip test not be done
(0,5)
unless the turbine has been operated for at least three
days with a minimum load..ng of iOX?
b.
WHY sh'uld the speed of the turbine be allowed to decrease
(0.5)
to 1200 rpm before the vacuum breaker is opened?
.
c.
Il condenser vacuum is slowly decreasing, why is the operator
( 1, 0 )
directed to reduce reactor power before manually tripping the
turbine?
(Give two reasons)
d.
Durit;g a startup, WHY is the operator cautioned not to operate
( 1. 0 )
the turbine below 800 rpm for greater than 5 minutes?
(an===
CATEGORY 04 CONTINUED ON NEXT PAGE ammaa)
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PROCEDURES
NORMAL, ABHORMAL, EHERGENCY AND
PAGE
16
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RADIOLOGICAL CONTROL
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QUESTION
4.04
(3.00)
A precaution in LOP-RT-02 (Reactor Water Cleanup - Startup)
states that if moderator temperature is less than 300 degrees F
both trains of heat exchangers should be lined up and only one
pump should be operating if reactor water level is normal
a.
What two (2) adverse conditions does this precaution prevent?
(1.0)
b.
What are five (5) of the seven (7) signals which wil)
(2.0)
automatically close BOTH FOO1 and FOO4 (Cleanup Isolation
Valves) ?
(SETPOINTS NOT REGUIRED)
GUESTION
4.05
(3.00)
The reactor is Shutdown and the operator is preparing to place RHP
(Residual Heat Removal) into the Shutdown Cooling (SDC) mode.
Answer
the following questions concerning core cooling in accordance with
LOP-RH-07
Shutdown Cooling System Startup and Operation.
'
a.
WHY does this procedure CAUTION the operator to ensure that RPV
level is at or above 40 inches as indicated on the Shutdown Range
prior to starting an RHR pump in the Shutdown Cooling (SDC) mode,
with no other forced flow through the vessel.
( i . ') )
b.
WHY is the operator CAUTIONED to slowly cut-in the RHR heat
exchanger upon startup of an RHR loop in the Shutdown Cooling
Mode.
(1.0)
c.
STATE TWO (2) of the three CRITERIA which will ensure that core
c^oling flow is sufficient to PREVENT temperature STRATIFICATION
in the KPV.
(1.0)
,
,
l
GUESTION
4.06
(3.00)
Per procedure LOA RX-Oi, "Control Room Evacuation", LIST SIX (6)
(3.0)
of the operator IMMEDIATE ACTIONS performed in the control room
prior to evacuation.
1
l
(naana CATEGORY 04 CONTINUED ON NEXT PAGE aanaa)
i
_ _ _ _ _ _
__ _ ...
__ . ..-_
, _ _ _ , _ _ . _ _ _ _ _ _ _ _
b.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AMD
PAGE '17
.
RADIOLOGICAL CONTROL
-
.,
GUESTION
4.07
(2.50)
One immediate action for Procedure LOA-FW-Oi, "Loss of Feed Water
Heater (s)",
requires core flow be decreased approximately 5 x 10 E6
lbm/hr for each 10 degrees F decrease in feed water temperature.
The immediate action also states a minimum core flow limit.
a.
LIST the minimum core flow limit per LOA-FW-Oi, and EXPLAIN WHY
(2.0)
the limit exists.
b.
STATE what is prevented by maintaining core flow above the limit.
(0,5)
QUESTION
4.08
(2.00)
Answer the following questions concerning the determination of the
bulk temperature of the Suppression Pool in accordance with LOP-CM-03,
Suppression pool Bulk Temperature Determination.
a.
Why is the operator directed to determine the Suppression
(1.0)
Pool Temperature by the value print of the computer points
L122 or L123 if Suppression Pool Level is LESS THAN
698 feet 11 inches (-8 inches).
.
b.
STATE TWO (2) additional METHODS for determining the
(1.0)
Suppression Pool bulk temperature.
(Assume all ECCS equipment is operating)
QUESTION
4.09
(2.00)
,
,
The reactor is operating at 55X power when a complete loss of ALL
OFF-SITE power occurs AND ALL of the AC BUSSES remain Deenergized.
,
STATE FOUR (4) IMMEDIATE OPERATOR ACTIONS per LOA-AP-08,
Total Loss of AC Power.
(2.0)
,
QUESTION
4.10
(1.50)
,
Answer the following questions concernir.g rad 401ogical controls at the
,
LaSalle Nuclear Station.
,
LIST SIX (6) of the eight CONDITIONS which require a worker to
"
LEAVE a Controlled Area per the Radiation Protectio? Standards
procedure, LRP-1000-1.
(1.50)
1
!
(amana END OF CATEGORY 04
====m)
I
'
(.....aeaeeeaa END OF EXAMINATION naeen=****u**=*)
>
_ , , _ . - , _ - _ _ . _ , _ _ , - _ _ _ _ - . _ _ -
_.
_.
_
,
,
.-. -
.
.
. - .
, p ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
PAGE
18
i
,
..
THERPODYNAMICS, HEAT TRANSFER AND FLUID FLOW
,
ANSWERS -- LASALLE 182
-88/04/26-NRC - REGION I
i
APSWER
1. 01
(2.00)
A.
Less than 50% (0.5)
Xenon concentration is higher than just after the power reduction
(0.5)
B.
Greater than 50X (0.5)
Xenon concentration is lower than just after the power reduction
,
(0.5)
REFERENCE
TPO 2ib
Reactor Theory text
292OO6K1.14
292OO6K114
...(KA*S)
ANSWER
1.02
(2.00)
1.
Xenon concentration
~
2.
Moderator Temperature
3.
Control rod position
4.
Order of rod withdrawa
5.
Core Exposure
(4 of 5 required @ 0.5 each)
REFERENCE
TPO 19.c
EWN8tNP Phyxies Review
,
'
LaSalle: LGP 1-1,
p.
6.
292OO8K401
...(KA'S)
i
l
l
'
i
i
1
!
1
i
i
i
e
.
. .
.
.
.
.t.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
PAGE
19
-
,,
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
.
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
ANSWER
1.03
(3.00)
a.
1.
Increase (0.50)
Since the incoming water is colder, more heat
can be added to the coolant before OTB occurs, therefore
the power at which transition boiling occurs will
increase.
(0.50)
2.
Decrease (0.50)
As pressure increases the amount of
heat required f or vaporization decreases;
therefore, the bundle power required to cause
transition boiling decreases.
(0.50)
b.
The Critical Power Ratio, CP/AP, will decrease (0.50) because an
increase in core flow results in a larger increase in the actual
power of a bundle than the increase in critical power of the
bundle. (0.50)
REFERENCE
LESSON PLAN ON CORE THERHAL HYDRAULICS
LEARNING OBJECTIVE NO.
5
_
GE BWR SERIES ON HEAT TRANSFER AND FLUID FLOV~SECTION 9.
293OO9K122
293OO9K124
...(KA'S)
,
ANSWER
1.04
(2.25)
a.
Fuel temperature would INCREASE (0,5) to get the needed delta
T to transfer the heat to the coolant.
The corrosion layer
will require some delta T across it to tra
,f, r; heat O&rG6),(d'I)
I to add positive
b.
Reactor Recirculation flow would INCREASE
reactivity to compensate for the negative reactivity effect of
the f ue l he a t up 44 AL) ( s.s t s)
c.
Core thermal power REMAINS THE SAME (0.5) since the total
amount of heat transfered to the coolant remains constant '^ 25!
i
(o.f)
REFERENCE
1.
LaSalle: Fluid Flov and Heat Transfer, pp. 76 and 78, TFO:II.B.S.
290002K506
...(KA*S)
I
i
i
i
,
-
.
.
.
.
.
.
,
_
L.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
PAGE
20
.
,
'
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOU
+
,
,
ANSWERS -- LASALLE i&2
-88/04/26-NRC - REGION I
i
ANSWER
1.05
(3.00)
a.
Rod worth increases,
(0.5) due to higher flux.
(0,5)
b.
Rod worth decreases, (0,5) due to decrease in thermal neutrons.
(0,5)
'
c.
Rod worth decreases,
(0.5) due to shadowing or increased
competition for thermal neutrons (0.5)
REFERENCE
'
1.
LaSalle: Reacter Physics, pp.
184, 188, 190, and 198, TPO:19.C.
292OO5K107
292OO5KiO9
...(KA'S)
.
ANSVER
1. 06
(3.00)
Avail.
Required
,
a.
DECREASE
REMAIN THE SAME
>
_
b.
DECREASE
REMAIN THE SAME
c.
INCREASE
DECREASE
(0.5 pts each)
REFERENCE
LaSalle: Fluid Flow and Heat Transfer, pp 64-70, TPO #13
202OO1K101
202OO1KiO3
202OOiK105
202OO1K122
...(KA'S)
,
I
'
l
1
!
!
'
i
'
,
- .
.
-
.
,
'
-
t.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
PAGE
21
.,
-
THERHODYNAMICS, HEAT TRANSFER AND FLUID FLOW
,
ANSWERS -- LASALLE 182
-88/04/26-NRC - REGION I
!
'
,
i
ANSWER
1.07
(2.25)
a.
CR1
1-Keff2
(0.5)
- ------------
CR2
1-Keffi
f
CRi(1-Keffi)
1-Keff2
CR2
CR1(1-Keffi)
Keff2
1
'
-
CR2
Keff2
1 - (1000m(1-0.91))/6000
(0,5)
--
Keff2
0.985 +/- 0.005
(0,5)
b.
TRUE
(0.75)
'
REFERENCE
LaSalle: Question and Answer Profile, Physics Review
TPO 15.b
292OO3K102
...(KA*S)
ANSWER
1.08
(2.50)
t
a.
P1: 50 on Range 8 : 500 on Range 6 : 5000 on Range 4+ Range 3
(0.5)
PO dt/T)
(0.5)
Pt :
20 dt/100)
(0.25)
5000 :
In(5000/20) : t/tOO
100 In(J50) : t
i
,
CFL t 3 ess
t
444 seconds : 9 min. 13 sec : 9.2 min
(0.25)
l
b.
period decreases to infinity and reactor power stabilizes
(0,5)
'
due to moderater temperature increasing which adds negative
- ( 0. 5 )
reactivity
j
i
!
i
1
_
. _ _
. .
_
. - -
-
,
.t .
PRINCIPLES OF NUCLEAR POMER PLANT OPERATION.
PAGE
22
i
.
.,
THERMODYNAHICS. HEAT TRANSFER AND FLUID FLOW
.
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
,
.
.
REFERENCE
REACTOR THEORY REVIEW pp.
113 + 129
TPO 15.b + 16.c
292003KiOB
292OO8K113
...(KA'S)
!
ANSWER
1.09
(2.00)
6,o }
Peripheral control rod worth increases (4-E) 2nd ::rt: 11 :: rte +4
7:d _ rC. d;;r: 2:2:
. 5 ', because the Xenon peak in the center of
'^
s
the core forces the flux to the periphery of the core (0.5),
so
the worth of the peripheral rods, which is determined by
(local flux / core average flux)^2 increases. This could lead to
a very large reactivity addition when a peripheral rod is
wi thdr awn. (0,5)
REFERENCE
'
REACTOR FHYSICS REVIEW pp.
198,226,228
TPO #19.c + #21.c
292OO5K109
292OO6Kio7
292OO6K108
...(KA'S)
_
ANSVER
1.10
(3.00)
a.
Doppler (0,5), negative (0.25)
.
b.
Void (0,5), negative (0.25)
c.
Hoderator temperature (or fuel temperature)
(0,5), negative (0.25)
d.
Void (0.5), positive (0.25)
,
REFERENCE
LESSON PLAN ON REACTOR PHYSICS PG 120 - 172.
LEARNING OBJECTIVE NO.
16,
17
AND 18.
GE BWP SERIES ON REACTOR THEORY
239002A106
295005Kioi
295014K203
295014K206
295021K201
...(KA*S)
,
i
i
.
. . .
..
.
.
,, ). .
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
23
.
ANSVERS
LASALLE 1&2
-88/04/26-NRC - REGION 1
--
ANSWER
2.01
(1.50)
Trip the Main Generator output Breakers (9-10 & 10-11)
Trip the Exciter Field Breaker
Trip the Main Turbine
Trip transformer 141 feeder breakers to SWGR 151 & 152 (6.9 kV)
and SWGR 141x & 142x (4.16 KV)
Auto transfer to enregize SWGR 151, 141x and 142x
(3 required G O.5 each)
REFERENCE
SYSTEM LESSON PLAN #46 TPO #6
SYSTEM LESSON PLAN #44 p 19
245000K406
245000K406
...(KA'S)
ANSVER
2.02
(2.00)
1.
Steam line (or dome) to pump suction temperature difference is
(10.1 degrees delta T.
~
2.
Total feed flow < 30X.
3.
TCV closure with power >30X of rated (EOC-RPT).
I
4.
TSV closure with power >30X of ratad (EOC-RPT).
5.
Reactor water level
<t2.5".
(Also accept RX low level.)
(4 required G O.5 each)
REFERENCE
l
1.
LaSalle: System Description, Chapter 5,
pp. 70, 72 and 80,
!
l
TPO: 9A + 12c.
I
202OO2A101
202OO2K104
202OO2KtO8
202OO2K109
202OO2K406
...(KA'S)
i
I
ANSWER
2.03
(2.00)
a.
To minimize containment fatigue from duty cycles.
(Also accept
l
reduces relief valve cycling.)
(0,5)
b.
LLS logic is armed whenever any two or more of the cafety/ relief
valves are signaled to open.
(0,5)
c.
By changing the open and reclosing pressures at which the valves
associated with the LLS operate.
(0.5)
I
d.
on either manual or automatic
(0,5)
REFERENCE
LaSalle: Question and Answer Profile, Sys 37 Rt
Chapter 01 System Lesson Plan
TPO 6 a.3 + 9a
.
. .
.
2
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
24
,
.,
ANSWERS -- LASALLE 182
-88/04/26-NRC - REG 2OH I
239002K401
239002K402
...(KA'S)
ANSWER
21 . 0 4
(2.50)
a.
The flow control valves will drive open
(O.5)
which causes recirculation flow to increase
(0.5)
b.
Both loop flow controllers shift to manual
(O.5)
AI>. .ecq t FCv te simy , er fin cen trolls < d dh S*"*".*..I
- r
3 3 p.e . e ntev//< ,- si,ff3 f, e.
l
(O.5)
c.
"E"
d.
"C"
is placed in bypass
(0,5)
REFERENCE
REC 1RCULATION FLOW CONTROL SYSTEM LESSON PLAN
TPO 12.a.3
202OO2K403
202003X607
215005K109
...(Ks'S)
>
ANSWER
2.05
(3.00)
~
a.
1.
decay of Xenon
' O. ; j -
c
2.
elimination of voids
' !--
- -
3.
increased water density (moderator cooldown)
(W ,
4.
reduced fuel temperature (reduced doppler)
(W
reduced ntu$ren it.ku
5
sia fhs, gt r. et,1,'f t. .ge.
3 7. SbN
b *<
- 'I"'#'
'b'
)
6
7
.,sgrw
b.
The Reactor Water Cleanup System isolates (FOO4 closes) (O.5) to
prevent removal of the sodium pentaborate (0,5)
.
REFERENCE
Standby Liquid Control System Lesson Plan p.
4 + Fig, 10-6
2i1000G004
21iOOOK105
2iiOOOK407
...(KA'S)
!
.
9
4
.
.
.
.
.
.
.
.
.
.
,
.
}
2.
PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
25
.,
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
ANSWER
2.06
(3.00)
a,
cycled CST
(CY)
Hormal
(0.25)
Suppression Pool er Alternate
(0.25)
AM R Ned erska gs,
b.
low CY tank level
(3'1")
(0.5)
c.
F010
CY Suction Valve
(O.34)
F022
Test Bypass (Throttleable)
(0.33)
F059
RCIC Test Bypass
(0.33)
d.
1.
insufficient lubrication of the turbine bearings
2.
water hammer in the exhaust line (due to check valve operation)
3.
excessive wear and oscillations of the governor valve (due to
low oil pressure)
(2 required @ 0.5 each)
REFERENCE
RCIC SYSTEM LESSON PLAN pp. 8 + 36
217000KtOi
...(KA'S)
ANSWER
2.07
(3.50)
,
a.
1) 3
or
neither
(0.5)
2) 3
or
neither
(0,5)
3) 2
or
241Y
(0.5)
4) i
or
141Y
(0.5)
b.
1.
DG lockout device is not locked out
2.
DG is operating >870 rpm
3.
DG voltage is normal
4.
Feed from the SAT is open
5.
Cross Tie breader to the COMPARISON
"X"
BUS is open
6.
UNIT CROSS-TIE is open
7.
Dead Bus (Undervoltage Device is activated)
(5 of 7 required @ 0.3 each)
REFERENCE
Diesel Generator System Lesson Plan pp. 36-37 + 42-44
TPO #6
264000A210
264000G007
...(KA*S)
_ _ _ _
- 2.'
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
PAGE
26
.,
ANSWERS -- LASALLE 1&2
-88/04/25-NRC - REGION I
.
ANSWER
2.08
(3.00)
I.
<ll fl51Vs clase due fe cen.ervahv <. sding o f Nf fl.W **I e,.p, o r
fn
a.
e all other MSIVs remain open (High flow setpoint is 134%)
(O.5)
2.
all MSIVs isolate (on sensed high flow in the B line)
(0,5)
3.
all MSIVs isolate (on low low water level signal) ,
(0,5)
no esfE.m e ccws o n gnif 3,
b.
1.
The low pressure indicates a pressure regulator malfunction
(O.5)
2.
The pressure drop wil1 decrease the saturation temperature
(O.5)
of the moderator
3.
A rapid change in pressure could cause cooldown limits to be
(0.5)
exceeded
kEFERENCE
PCIS SYSTEM LESSON PLAN pp.
15-17
TPO #1 + #2
239001K401
...(KA'S)
~
ANSWER
2.09
(2.50)
a. At low reactor pressures with the accumulators discharged, reactor
pressure may not be sufficient to scram the rods (O.5), therefore*
a scram is inserted before accumulators can become discharged (O.25)
(1) 1157 psig insures that the accumulator piston is still seated (0. 5)
(sufficient wav.er volume in the accumulator) therefore a normal
scram is assured (0.25)
(2) The trip is bypassed in RUN because reactor pressure wilI be
sufficient to scram the rods (>854 psig) (O.5)
b.
(O.2)
Lose
"48"
position indication
( 0. 1 )
i
a r v e. f l., eL n
fd
,,,,,r(ou.t,,E.,...//f/w
m re e =
(O.1)
.m n . :
_ . . . . _ . =
J.a . fo# c r. 4,f ,Y
/
(o o
- *.y .L m se,.l,lc a ,, w e,
REFERENCE
RPS System Lesson Plan
LGP-1-1
Simulator Malfunction #317
201003A208
201003K402
201003K404
...(KA'S)
- -
,.
._ ,_
__ . - . .
.
-
).
PLANT DESIGN INCLUDING SAFETY AND EHERGENCY SYSTEMS
PAGE
27
'
,,
'
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION I
.
!
ANSWER
2.10
(2.00)
a.
CST suction remains open and the Suppression Pool suction
remains shut.
,
.
b.
CST suction valve opens and the Suppression Pool suction
l
'
remains shut.
c.
. CST suction remains shut and the Suppression Pool suction
remains open,
a
i
d.
The HPCS test valve closes.
(4 required, 0.50 ea.)
1
'
i
REFERENCE
i
LESSON PLAN ON HPCS CHAPT 36 PG 10.
1
LEARNING OBJECTIVE NO.
6.A.2 AND 6.A.4.
!
LOP-HP-04 PG 2.
209002Alot
209002A108
209002A201
209002A301
...(KA'S)
l
_
?
l
l
.
I
i
I
4
,
i
i
I
I
l
i
d
1
I
i
I
n
1
'
5
,
1
!
l
,
.
.
- _ , - , . - . . . - - , - - _ , ,
-
.
-
__
_
_ _
_ . - . _ _
_
.
. -
._
_.
._ _
.
,
'
l ,
- , B ,
INSTRUMENTS AND CONTROLS
PAGE
28
ANSWERS -- LASALI,E 1&2
-88/04/26-NRC - REGION I
t
-
i
ANSWER
3.01
(3.00)
3
'
<~
a.
differential pressure is proportional to the square of
(0,5)
the flow
I
or -
-
i
the square root of the differential pressure is proportional
to the flow
b.
SBLC injection line
- or -
below core plate
(0,5)
i
c.
the idle loop flow is subtracted from the operating loop flow (0.5)
to compensate for reverse flow through the idle loop
(0.5)
d,
breaker 2 position
(0.34)
breaker 3 position
(0.33)
a
,
breaker 4 position
(0.33)
,
.
REFERENCE
.
l
TPO 6. c
4
System Lesson Plan Vessel Instrumentation
_.
202OOiA303
202002G007
202OO2K103
291002K105
...(KA'S)
.
ANSWER
3.02
(3.00)
!
i
l
l
a.
The TCVs will close to 50% flow position
(0,5)
f
l
The TCV low value gate passes a MCF signal of 50X
j
rather than the signal from the pressure controller
(0,5)
!
,
!
b.
The BPVs will remain closed through the transient
(0,5)
l
the MCF summer will send a zero signal to the BPV LVG
(0,5)
'
Reactor power and pressure will rapidly increase following
(0,5)
l
c.
the fault,
i
The reactor will scram on High Flux and/or high pressure
(0.5)
[
4
because of the closure of the TCVs
!
i
)
'
REFERENCE
System Lesson Plan, EHC p.
10 + 46
,
TPO 6.e
12.h
3
j
245000K602
...(KA*S)
,
.
'
l
.
i
1
,
- 3.'
INSTRUMENTS AND CONTROLS
PAGE
29
,
'
ANSWERS -- LASALLE 182
-88/04/26-NRC - REGION I
l
.
ANSVER
3.03
(3.00)
indicated level will initially
i
l
a.
increase
(0.5)
b.
remain the same
(0,5)
c.
increase
(0.5)
d.
increase
(0.5)
e.
Increase
(0.5)
l
f.
decrease
(0.5)
l
l
REFERENCE
'
'
TPO 12
SYSTEM LESSON PLAN REACTOR VESSEL INSTRUMENTATION
'
216000A201
216000A203
216000A207
216000A208
216000A210
216000K324
216000K506
216000K507
216000K512
216000K513
...(KA'S)
i
i
ANSVER
3.04
(2.00)
'
a.
1.
diesel speed (frequency)
(0.5)
2.
load control
(0.5)
'
,
l
b.
1.
voltage control
(0.5)
2.
VAR control
(0.5)
l
'
REFERENCE
TPO 4
1.
LaSalle
System Description, Chapter 47.
2.
LaSalle: Exam Bank,
3-LS-66.
264000A201
264000A304
264000A401
264000 GOO 9
264000K505
...(KA*S)
l
l
ANSWER
.05
(2,50)
l
1. 0 )
a.
Ye.. M
r = 21 i n i t : : t i : r m e r. m . m ire
- r ;;:::r : prp
i
- :: ::
- ^b
"-* '^
V,
!
l
b.
No (0,5)
c.
Yes (0.5)
{
d.
SRV valve pressure can still be supplied from Emergency
also accepf p[hre ca gccamgfgf' erg
l
Pressurization Station
(0,5)
REFERENCE
1.
LaSalle: System Description. Chapter 37
pp. 14-16
TPO: 6.
218000K402
218000K403
218000K404
...(KA*S)
)
\\
1
'
s'
>
INSTRUMENTS AND CONTROLS
PA3E
30
- . ' 1
'
ANSWERS -- LASALLE 182
-88/04/26-NRC - REGION I
.
ANSWER
3.06
(2.00)
,
a.
Rod 32-35 must be inserted (1.00)
..
b.
3
(1.00)
REFERENCE
LESSON PLAN ON RWM CHAPT 18 PG 14, 15,
16
AND 29.
,
LEARNING OBJECTIVE NO.
6. b.
201006A205
201006K401
201006K402
20iOO6K403
...(KA'S)
,
ANSWER
3.07
(2.00)
a.
CAN NOT be bypassed (0.25)
'
b.
CAN be bypassed (0.25), mode switch not in run (0.25)
2
c.
CAN NOT be bypassed
(0.25)
e
d.
CAN be bypassed (0.25), below 30X power (0.25)
l
as sensed by first stage pressure (0.25)
e.
CAN NOT be bypassed (0.25)
,
,
REFERENCE
i
TPO 9
'
1.
LaSalle: System Description, Chapter 20, pp. 13 and 14
TPO:6.
,
212OOOK412
...(KA*S)
,
,
1
ANSWER
3.08
(3.00)
a.
High-High
(120/125)
(0.5)
1
Inoperable (High Voltage Low, Module Unplugged. Switch not in
(0.5)
+
Operate)
i
b.
Hode Switch in RUN
(0,5)
Joystick in Bypass
(0,5)
'
'
(e.if)
l
c.
HALF SCRAH
M.
Due to high voltage low inop
(0.25)
Rod Block
(e.3r)l0 ";-
Due to IRM downscale erk
velfegu l,w
(0.25)
REFERENCE
,
System Lesson Plan
pp.
14 + 26
TPO #9
i
215003K401
215003K402
215003K602
.. (KA*S)
l
!
!
i
!
-
- -
-
- -
.
+
.
- .
3
INSTRUMENTS AND CONTROLS
PAGE
31
,,
ANSWERS -- LASALLE 142
-88/04/26-NRC - REGION I
.
ANSWER
3.09
(2.50)
a.
Setpoint Setdown prevents vessel overfeeding after a scram
(0.5)
The Setpoint Setdown circuitry reduces the operated selected
setpoint hy b = 'A when a low level trip occurs.
(0.5)
Te I S "
b.
Decrease, (1.0) because the level setpoint would be reduced
to 18" regardless of the setpoint tape setting.
(0.5)
REFERENCE
1.
LaSalle: System Description, Chapter St. py- 53, 56, 63, and
64
TPO: 9 + 10
259002K301
259002K404
259002K412
...(KA'S)
ANSWER
3.10
(2.00)
Manually) switch mode selector switch (to the "Modulatt
+ 2 Step"
a.
. . -
pos1*. ton (0.5), because the air operated blow off va1\\e will
not open (0.25) to relieve excessive pressure. (0.25)
OR-
e fe t y t f a r f a ll rre /,'f
handle pressure surges
,
b.
1.
open (0.25)
discuu Am of sgrats %=d fea d[na con ti*4 5
2.
shut (0.25)
Q
d
3.
shut (0.25)
4.
open (0.25)
REFERENCE
1.
LaSalle: System Description, Chapter 68, pp. 15, 18,
19
and
20. TPO:3b, 6a, 14
20100tK603
204000K604
23900iK602
...(KA*S)
.
" -
.
N.
PROCEDURES
NORMAL. ABNORMAL. EMERGENCY AND
PAGE
32
-
RADIOLOGICAL CONTROL
,
'
LASALLE 1&2
-88/04/26-NRC - REGION I
ANSWERS
--
.
0, M e,.d f., bml., .f: sLJl .al L
f
4,
,y ,../t h . n s M /s, cea h m t . p = &
<
ANSVER
4.01
(2.50)
,emldwer666 s if4N a, 4 he r e t A s,s 67
'
540
a.
Do not secure or place an ECCS in MANUAL mode unless, by at
]
least two independent indications
(O.5)
1.
misoperation in AUTOMATIC mode is confirmed (continued
operation would worsen situation)
(O.5)
- OR -
2.
adequate core cooling is assured condition stable
( 0. 5 )
b.
If an ECCS is placed in MANUAL mode, it will not initiate
automatically (0.5).
Make frequent checks of the initiating
i
or controlling parameter
(O. 5) .
(When manual operation is no
i
longer required, restore the system to AUTOMATIC /STANDDY mode
if possible.)
REFERENCE
i
j
1.
LaSal1e: LGA-GP,
p.
2,
Precaution #11.
2030000001
2090010001
209002 GOO 1
2170000001
. . . ( 1: A ' S )
,
'
ANSVER
4.02
(2.50)
a.
Full the fuses for the affected valve.
(0,5)
,
b.
Control switch valve indication following replacement of fuses
,
(0.25) or ta11 pipe temperature.
(O.25) OR any other two (2)
reasonable responses
c.
1.
Four attempts to cycle valve
(0.5)
2.
Pool temperature reaches 110 deg F
(0.5)
'
3.
Two minutes have elapsed
(O.5)
REFERENCE
]
,
1.
LaSalle: LOA-NB-02, pp. 2 and 3.
i
239002A203
...(1:A'S)
i
I
I
1
.
a
'
'
.
i
.
.
.
.
.
.
.
.
.
. .
.
.
.
.
.
.
.
-.
. . . - _
-
.-
'
e, $ .
PROCEDURES - NORMAL, ABNORMAL, EMERGENC'/ AND
PAGE
33
,
RADIOLOGICAL CONTROL
i
ANSVERS -- LASALLE 142
-88/04/26-NRC - REGION I
l
-
'
ANSWER
4.03
(3.00)
,
a.
to insure proper rotor warming
(0,5)
!
,
9l88 acttph to s'asure preter LP t.,61ne were q
i
1
b.
if this is not done there will be severe duty on the last stage
(0.5)
buckets. (prevents overheating of the last stage buckets)
f
'
c.
reducing power will help reduce vacuum, and
0)
sv.
will reduce the probability of damage due to high bjek pressure
P4 4)
1
(and overspeed conditions)~
(2*f 4 reg cad y 0,f %sA)
,
allows tems fe, cerruf W gef en
3 r r c / a sc.s e e,'1f . f 6 4 ,w te,g / A, s ce.m
'
'
d.
operation in this range could lead to high vibrations
10,5)
i
which would not be seen by the Turbine Supervisory Instruments
(0,5)
REFERENCE
,
]
LOP-TG-02
i
2450000010
...(KA'S)
-
ANSWER
4.04
(3.00)
j
a,
cavitation of the RWCU (RT) pumps (due to inadequate NPSH)
- ( 0. 5 )
'
,
Flow oscillations (due to boiling and two phase flow at the
( 0. 5 )
l
system suction line high point)
b.
1.
Reactor Water level low
j
2.
High Ventilation dT from the RT area
'
3.
High Differential Flow
i
4.
High RT area Temp
1
5.
Loss of Isolation Logic Power
1
6.
Loss of RPS bus B or A
'
7.
Manual Isolation Pushbutton
I
J
(5 of 7 required @ 0.4 each)
REFERENCE
l
LOP-RT-02 Reactor Water Cleanup - Startup
204000GOOi
204000G010
204000K404
...(EA'S)
l
j
i
.
'
1
l
- . _ .
._
_
_ _ _ _ _ _ _
._-
- . -
. - _ _ . . _ _ _ , _ _ _ _ _ . _ . , _ _ . _ . _ , _ _ _ . . - - _ _ _ ,.
-
.
.
_
-
- -
-
NORMAL. ABNORMAL. EMERGENCY AND
PAGE
34
y.
PROCEDURES
-
,
'
%
RADIOLOGICAL CONTROL
ANSWERS -- LASALLE 1&2
-88/04/26-NRC - REGION 1
!
a
"
.
,
i
ANSWER
4.05
(3.00)
,
a.
Ensuring vessel level is at or above 40 inches prevents the level
I
fluctuations in the downcomer caused by the pump startup from
resulting in RPS/PCIS initiations at 12.5 inches.
(1.00)
,
b.
The RHR heat exchanger cannot be pre-warmed, so slowly
4
cutting in the RHR heat exchanger prevents thermally stressing the
j
SDC return nozzles.
(1.00)
e
c.
1.
At least one recirc pump is operating in the loop that is not
aligned for SDC.
2.
With no recipe pumps running, S DC f l ow > 6000 dpm
(+/-
500 gpm)
3.
RPV level is above 578 inches (+50 inches on S/D Range)
(+/-
5 inches)
9
D ettm lis.< d d re,'e Clw > 25gpm
(2 required @ 0.50 each)
REFERENCE
LOP-RH-07 PG 2,
3,
4,
AND 5.
205000A105
205000 GOO 7
205000K102
290002G010
290002X603
290002K611
...(KA'S)
.
n
ANSWER
4.06
(3.00)
7
1.
Announce control room evacuation and why.
i
2.
Manually scram the reactor.
!
3.
Place the mode switch in shutdown.
[
4.
Verify that power is decreasing and that all control rods are
e
f
inserted.
5.
Start the Motor Suction and Turning Gear oil Pumps
6.
Trip the Main Turbine.
t
7.
Trip the rectre pumps.
8.
Stop reject of reactor coolant if in progress.
!
l
9.
Confirm no LOCA indications.
10. Verify Bus 141Y (241Y) and 142Y (242Y) are energized.
,
11. Verify that the 250 volt DC and 125 volt DC busses are energized.
!
4
~
12. Verify Bus 143 energized.
13. Place the HPCS diesel select switch in the local position.
(6 required at 0.5 each)
<
REFERENCE
LaSalle Procedure LOA-RX-01.
295000G010
...(KA*S)
4
4
j
I'
. -
-
-
- - -
-
-- - - - - -
-
J
-
.-
NORMAL. ABNORMAL, EMERGENCY AND
PAGE
35
,
J.'
PROCEDURES
-
3
RADIOLOGICAL CONTROL
.
..
'
ANSWERS -- LASALLE 182
-88/04/26-NRC
REGION I
-
.
l
ANSVER
4.07
(2.50)
,
a
a.
Limit of 45X of rated core flow
-or-
49 x 10E6 lbm/hr Recirculation
li
flow.
(0,5) Rapid f1Aw biased setpoint decreases and/or core flow
'
,
]
instability (Gr&)"p'Ns the APR p ggpal input to the thermal power
s
reduces the margin to APRM
monitor being time displayed,
,
scrams during core flow reductions.
(0.5)
(alternate wording accepted)
'
b.
The 45X of rated flow limit is to avoid a reactor scram
-or-
Core flow instability.
(0.5)
REFERENCE
LaSalle
LOA-FW-01,
215005K407
...(KA*S)
t
I
ANSWER
4.08
(2.00)
a.
When the Suppression Pool level is <
-8 inches most of the RTDs
,
~
which measure the Suppression Pool temperature are no longer
covered with water and will not read the correct relative
,
j
temperature of the Suppression Pool.
(1.00)
'
l
b.
1.
RHR temperature recorders on P601 if the RHR system is in
serv 1Ce.
2.
Contact pyrometer on the sdction piping of any ECCS pump
,
taking suction on the suppression pool.
3.
If HPCS is running, by placing a temperature gage in the
temperature well of the HPCS pump suction,
4.
If RCIC is running, by checking the local RCIC pump discharge
temperature indicator.
-
s
Re
ts $/O Peasi ts,pors t,,s in dL+,e
'
,
(2 required, 0.50 ea.)
i
~
REFEREMCE
LOP-CM-03 PG 3
6
AND 7.
295030A202
295030 GOO 7
...(KA*S)
s
j
l
l
1
I
i
)
.-
-
-
7_
-
-
.
- . ' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
PAGE
36
_ , ,.i
RADIOLOGICAL CONTROL
'
m
ANSWERS -- LASALLE ta2
-88/04/26-NRC - REGION I
.
ANSWER
4.09
(2.00)
1.
Verify reactor scram.
,
'
2.
Initiate RCIC.
3.
Verify SRVs open (0.25) maintainpressure(between900and1000)
psig. (0.25)
4.
Attempt to start the Diesel Generators.
5.
Notify the Shift Supervisor.
(4 required, 0.50 ea.)
1
REFERENCE
{
LOA-AP-OS PG
1.
I
295003G010
...(KA*S)
l
!
ANSWER
4.10
(1.50)
1.
When instructed to do so by the Radiation-Chemistry
'
-
Department.
2.
Upon failure (or suspected failure) of personal protective
equipment.
3.
Unexpected deterioration of radiological conditions.
4.
In the event tat the workers current accumulated dose
equivalent status becomes uncertain or his dose equivalent is
equal to the that authorized for the job.
5.
The "assembly" siren sounds.
6.
Completion of work assignment.
7.
Injury.
8.
Unexpected radiation monitor alarm and the area dose rate is
unknown.
(6 required, 0.25 ea.)
REFERENCE
LRP-1000-1 PG 7 AND 8.
294001XiO3
...(KA'S)
,
,
i
- -
- -
- -
.
1
~
a:,-
e
4
)
.u.. STER
l
,.
.
l .' . .
-
(
6
n
i
ATTACl1MENTS
..
4
9 M
M
a
9 6
s
.w
% -.
4
6
-
--- . _ _ _ _ _ _ . _ _ _ _ _ _
,
___ ______ ____________
'
g.w.
'
EQUATION SHEET
.
'
-
- . _
f = ma
y e s/t
Cycle efficiency = (Net work
out)/(Energy in)
I
2
w = mg
s = V,t + 1/2 at
2
,
E=E
,
2
g ,g ,-At
KE = 1/2 av
, , (yf , y )jg
g , 13
o
o
,
PE = ogn
Vf = V, + at
w = e/t
1 = En2/t
w 0.693/t
-1/2
1/2
1/2'ff * U t1/M *bU
2
t
-
w.y 3p
n0
A=
((t1/2)'ItI3
4
b
4 = 931 am
-
m = V,yAo
.m
I = 1,e
.
.
Q = mCpat
-ux
,
d = UA4 t'
g.ge
Pwe = W sh
!=I
10'*/IYL
f
TVL = 1.3/v
,
sur(t)
HYL = -0.693/u
P = P 10
P = P e*/I
o
i
SUR = 26.06/T
SCR = S/(1.- K,ff)
CR = S/(1 - K,ffx)
-
x
CR (1 - K ,ff)) = CR (I ~ eff2}
SUR = 26s/t* + (s - o)T
j
2
T = ( t*/o ) + ((s - o V Io ]
M = 1/(1 - K,ff) = CR)/CR,
T = 1/(o - 8)
M = (1 - Keffo)/(I - Keffl)
T = (s - o)/(Io)
SDM = ( -K,ff)/K,ff
a = (K ,ff-1)/K ,ff = AK,ff/K,ff
tw = 10
seconds
i
I = 0.1 seconds ~I
o = [(t*/(T K,ff)] + (a,ff (1 + IT))
/
Ijj=Id
d
2 =2 2
P = (IsV)/(3 x 1010)
Id
1d
jj
22
2
I = oN
R/hr = (0.5 CE)/c (eetes)
R/hr = 6 CE/d2 (feet)
,
Water Parameters
Miscellaneous Conversions
.
I gal. = 8.345 lem.
1 curie = 3.7 x 1010ep,
1 gal. = 3.78 liters
1kg=2.21lem
3 Sta/nr
1 ftd = 7.48 gal.
1 np = 2.54 x 10
Oensity = 62.4 lbT/ft3
1 m = 3.41 x 106 Stu/hr
Density = 1 gm/cW
lin = 2.54 cm
Heat of vaoorization = 9/0 5tu/lem
'F = 9/5'C + 32
'
He at of fusion = 144 Stu/lem
'C = 5/9 ('F-32)
1 Atm = 14.7 psi = 29.9 in. Hg.
i BTU = 778 ft-itf
I ft. H O = 0.4335 itf/in.
2
,
_.
.
-
-
,
L, . . '
,
'
e
FEED FLOW
.
4
!
)
l
,
4
I
L
I
-
'
05
.
50
60
'
.-
10
20
30
40
TIME SECOWS
+60
i
O
REACTOR WATER LEVEL
/
!
l
-60
-
_ . ,
..
!
10
20
30
40
50
60
100%%
'
APRM' S
.
4
,
.
,
i
.+
'
'
4
,
m
10
20
30
40
50
60
TIME SECONDS
l
1001 %
,
t
1
)
TOTAL STEAM FLOW
J
- - _ .
'
'
'
'
'
Of
--
'
'
10
20
30
40
50
60
TIML SEC0N05
l
.
1000
RLACTOR VESSEL PRESSURE
i
FIGURE 2
'
900
-
i
10
20
30
40
50
LU
TIME SECONDS
,
.
J
6
- .
,
.'.
MASTO
.
SELECTED TECilt1ICAL SPECIFICATI0t1S
SECTIO!1
NE
3/4.0
APPLICABILITY
3/4 0-1
3/4.3.3 ECCS ACTUATI0t1 I!1STRUME!1TATI0t1
3/4 3-23
3/4.4.1 RECIRCULATI0t1 SYSTEM
3/4 4 1
3/4.4.2 SAFETY / RELIEF VALVEG
3/4 4-6
3/4.5.1 ECCS OPERATING
3/4 5-1
.
9
y e - c. A A u
/M9ks.4, ,
.7 ,n c r. .
.
......':...
'
.
.
l
-
-
,
I
t
3
-
,.
.
,
l
3/4.0 APPLICA8!LITY
.
,
LINITING CONDITION FOR OPERATION
1
3. 0.1 Compliance with the Limiting Conditions for Operation contained in the
!
!
succeeding.5pecifications is required during the OPERATIONAL CON 0!TIONS or other
J
!
'
conditions specified therein; except that upon failure to meet the Limiting
Conditions.for Operation, the associated ACTION mquirements shall be met.
l
.
3.0.2 Noncompliance with*4 Specification shall exist when the requirements of
j
the Limiting Condition for Operation and associated ACTION requirements are
!
not met within the specified time intervals.
If the Limiting Condition for
'
]
Operation is restored prior to expiration of the specified time intervals,
completion of the ACTION requirements is not required,
j
'
!
3.0.3 When a Limiting Condition .for Operation is not met, except as prow?ded
,
in the associated ACTION requiremenf t, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initteted
I
to place the unit in an CPEAATIONAL CONDITION in which the Specificatiun does
4
J
not apply by placing *it, as applice le, in
!
1.
At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
.
2.
At least NOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
)
,j -
3.
At least COLD SHUTDOWN witnin the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
!
When cornetive esasums are completed that pemit operation under the ACTION
j
j
[
requirements, the ACTION may be taken in accordance with the specified time
J
l
A
limits as measured free the time of failun to meet the Limiting Ceadition for *
,
l
Operation. Exceptions to these requirements are stated in the individual
i
Specifications.
j
This specification is not applicable in OPERATIONAL CONDITION 4 or 5.
'
3.0.4 Entr/ into an CPERATIONAL CONOITION or other specified condition shall
not be made unless the conditions for the Limiting Condition for Operation are
'
met without reliance on provisions contained in the ACTION requirements. This
provision shall not prevent passage through CPERATIONAL CONDITIONS as nquired
to coolly with ACTION requineents.
Exceptions to these mquirements are
stated in the individual Specifications.
,
i
.
1
3.0.5 When a system, subsystas, train, component or device is detemined to
j
be inoperable solely because its emergency power source is inopeable, or
solely because its normal power source is inoperable, it may be considered
- )
OPERA 8LE for the purpose of satisfying the requirements of its applicable
i
Lietting Condition for Operation provided:
(1) its corresponding nomat or
i
emergency power source is OPERA 8LE; and (2) all of its recunoant system (s),
'
subsystem (s), train (s), component (s) and device (s) are OPERA 8LE, or likewise
satisfy the requirements of this specification. Unless both conditions (1)
and (2) are satisfied, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> action shall be initiated to place the
unit in an CPERATIONAL CON 0! TION in which the applicable Limiting Condition
j
for Operation does not apply by placing it, as applicable, in:
)
1.
At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
2.
At least HOT SHUTCOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
a
j (
At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
3.
j
This specification is not applicable in CPERATIONAL CON 0!TICN 4 or 5.
LA SALLE - UNIT 2
3/4 0-1
1
j
- _ _ . . _ . . .
. . . . . . .
.- .
_ _ _ ._._.__ _ ,
- _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ . . _ . _ _ _ ,
_,
.___.
._ __
-
,
_ _ _ _ _ _ _ _ _ _ _ _ _ _
.
- -
!
'
'
'
...
. -
. . . . . .
-
.
.
.
<
INSTRUMENTATION
.V4. 3. 3 EMEj ". 20RE COOLING SYSTEM ACTUATION INSTRUMENTATION
.
'
LIMITING
1R OPERATION
3.3.3 The
.cy core cooling system (ECCS) actuation instnmentation
.
chanr.als snow.. in Table 3.3.3-1 shall be OPERABLE with their trip setpoints
set consistant with the values shown in the Trip Setpoint column of Table 3.3.3-2
and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.
APPLICABILITY: As shown in Table 3.3.3-1.
CTION:
1.
With an ECCS actuation instrumentation channel trip setpoint less
conservative than the value shown in the Allowable Values column of
Table 3.3.3-2, declare the channel inoperable until the channel is
.
restored to OPERABLE status with its trip .setpoint adjustad consistent
'with the Trip Setpoint value.
'
-
b.
With one or more ECCS actuation instrumentation channels inoperable,
take the ACTION required by Table 3.3.3-1.
.
(
c.
With either ADS trip system "A" or "B" inoperable, restort the
<
inoperable trip systen to OPERABLE status within:
-
1.
7 days, provided that the HPCS and RCIC sy' stems are OPERABLE.
2.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and
reduce reactor steam does pressure to less than or equal to 122 psig
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILURCE REQUIREMt'NTS
,,
.
4.3.3.1
Each ECCS actuation instrumencation channel shall be demonstrated
OPERABLE by the pqrformance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and
CHANNEL. CALIBRATION uperations for the OPERATIONAL CONDITIONS and at the
frequencies shown in Table 4.3.3.1-1.
4.3.3.2
LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of
all channels e bli be performed at least once per 18 months.
4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3
shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all
channels are tested at letst once every N times 18 months where N is the total
number of redundant channels in a specific ECCS trip system.
.
U. SALLE - UNIT 2
3/4 3-23
.
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TABLE 3.3.3-1
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[HERGENCY CORE COOLING SYSTEM ACIUATION INSTRUM[NIA110N
f-
'a
MINIMUM OPERABLE
APPLICABLE
CilANNELS PER 1 RIP
OPERATIONAL
.
IRIP FUNCTION
FUNCIION(a)
CONDITIONS
ACTION
"
A.
DIVISION I TRIP SYSTEM
I
1.
RilR-A (LPCI MODE) & LPCS SYSTEM
2 'I
1, 2, 3, 4", 5"
30
II
a.
Reactor Vessel Water Level - Low Low Low, L vel 1
II
2 'I
1, 2, 3
30
f
l
b.
Drywell Pressure - iligh
.'
)
c.
LPCS Pump Discharge Flow-Low (Bypass)
~
l
1, 2, 3, 4 * , S*
31
l
!
.
d.
LPCS and LPCI A Injection Valve Injection Line
1/ Valve
1, 2, 3
32
R
Pressure-Low (Permissive)
4^ , 5*
33
!
$
,
e.
LPCS and LPCI A Injection Valve Reactor
2
-
l
2, 3
38
l
Pressure-Low (Permissive)
4g, 5*
33
-
,,
s.
f.
LPCI Pump A Start Time Delay Relay
1
1, 2, 3, 4 * , 5*
32
g.
LPCI Pump A Discharge flow-Low (Bypass)
-
1
1,2,3,4*,5*
31
h.
Manual Initiation
1/ division
1, 2, 3, 4 ^ , 5"
34
i
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
1
2 'I
1,2,3
30
II
a.
Reactor Vessel Water Level - Low Low Low, Level 1
i
coincident with
I
b.
Drywell Pressure - liigh
2
1, 2, 3
30
c.
Initiation Timer
1
1,2,3
32
l
1
1-
2
d.
Reactor Vessel Water Level - Low, l,evel 3 (Permissive)
1
1, 2, 3
32
h
e.
LPCS Pump Discharge Pressure-liigh (Permissive)
2
1,2,3
32
I
k
f.
LPCI Pump A Discharge Pressure-liigh (Permissive)
2
1, 2. 3
'32
i
y
g.
Manual Initiation
1/ division
1, 2, 3
34
,
j
h.
Drywell Pressure Bypass Timer
1
1, 2, 3
-32
m
i.
Manual Inhibit
1/ division
1, 2, 3
34
.
.
. - -
.
. .
.
. .
- -
-
m.
1
--
--
--
-
.
.
4
..
0
.D
D
..
_
g
1ABLE 3.3.3-1 (Continued)
O
[MERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
G
"*
MINIMUM GPERABLE
APPLICABLE
CilANNELS PER TRIP
OPERAll0NAL
IRIP FUNCIION
.FUNCTIOH(a)
CON 01110NS
ACTION
"
B.
DIVISION 2 IRIP SYST[H
1.
RilR 8 & C (LPCI MODE)
2 'I
1, 2, 3, 4*, 5*
30
II
a.
Reactor Vessel Water Level - Low, Low Low, Level 1
I
b.
Drywell Pressure - liigh
2
1, 2, 3
30
-
LPCI B and C Injection Valve 4.jection Line' Pressure-Low
1/ valve
1, 2, 3
32
c.
.
(Permissive)
4*, 5*
33
R
d.
LPCI Pump B Start Time Delay Relay
1
1, 2, 3, 4*, 5*
32
{
$
e.
LPCI Pump lischarge flow - Low (Bypass)
1/ pump
1, 2, 3, 4*, 5*
31
h
f.
Manual Initiation
1/ division
1, 2, 3, 4*, 5*
34
g.
LPCI 8 and C Injection Valve Reactor
.
2
1, 2, 3,
38
Pressure-Low (Permissive)
4*, 5*
33
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
.
a.
Reactor Vessel Water Level - Low Low Low, Level 1
2(b)
1, 2, 3
30
coincident with
II
2 'I
1, 2, 3
30
.
b.
Drywell Pressure - High
c.
Initiation Timer
1
1, 2, 3
32
l
d.
Reactor Vessel Water Level - Low, Level 3 (Permissive)
1
1, 2, 3
32
,
'
R
e.
LPCI Pump B and C Discharge Pressure - liigh
'
g
(Permissive)
2/ pump
1, 2, 3
32
k
f.
Manual Initiation
1/ division
1, 2, 3
34
l
g.
Drwall Pressure Bypass Timer
1
1, 2, 3
32
.
P
h,
thy =: Inhibit
1/ division
1, 2, 3
34
4
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.
.
.
-
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.
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D
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.
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TABLE 3.3.3-1 (Continued)
%
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
!
MINIMUM OPERABLE
APPLICABLE
'
CHANNELS PER TRIP
OPERATIONAL
FUNCTION (*)
CONDITIONS
ACTION
TRIP FUNCTION
"
C.
DIVISION 3 TRIP SYSTEM
1.
hPCS SYSTEM
a.
Reactor Vessel Wder Level - Low, Low, Level 2
4
1, 2, 3, 4*, 5*
35
b.
Drywell Pressure - High
4(c)
1, 2, 3
35
c.
Reactor Vessel Water Level-High, Level 8
2(d)
1, 2, 3, 4*, 5*
32
d.
Condensate Storage Tank Level-Low
2(d)
1, 2, 3, 4 * , 5*
36
e.
Suppression Pool Water Level-High
2
1,2,3,4*,5*
36
y
f.
Pump Discharge Pressure-High (Bypass)
1
1,2,3,4*,5*
31
B
g.
IIPCS System Flow Rate-Low (Permissive)
1
1,2,3,4*,Sa
31
'
w
h.
Manual Initiation
1/ division
1, 2, 3, 4*, 5*
34
E
l
D.
LOSS OF POWER
MINIMUM
TOTAL NO.
INSTRU-
APPLICABLE
'
OF INSTRU- MENTS TO INSTRU-
OPERATIONAL
,
MENTS
.
TRIP
_ ,MENTS(a)
CONDITIONS
ACTION
1.
4.16 kV Emergency Bus Undervoltage
2/ bus
2/ bus
2/ bus
1, 2, 3, 4**, 5**
37
(Loss of Voltage)
i
2.
4.16 kV Emergency Bus Undervoltage
2/ bus
2/ bus
2/ bus
1, 2, 3, 4**, 5**
37
l
-(Degraded Voltage)
l
TABLE NOTATION
>
,
2
(a) A channel / instrument may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during periods of required
i
E
surveillance without placing the trip system / channel / instrument in the tripped condition provided at least
2
one other OPERABLE channel / instrument in the same trip system is monitoring that parameter.
To
(b) Also actuates the associated division diesel generator.
.
(c) Provides signal to close HPCS pump discharge valve only on 2-out-of-2 logic.
- e
P
(d) Provides signal to HPCS pump suction valves only.
Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.
^
w
Required when ESF equipment is required to be OPERABLE.
"
Not required to be OPERABLE when reactor steam done pressure is i 122 psig.
i
i
.
-
-
-
- -
-
.
.
_ _ _ _ _ - _ _ _ _ - _ _
.
..
'
..
.
(
TABLE 3.3.3-1 (Continued)
EMERCENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
'
ACTION
i
ACTION 30 -
With the number of OPERABLE channels less than required by ae
Minimum OPERABLE Channels per Trip Function requirement:
a.
With one channel inoperable, place the inoperable channel
in the tripped condition within one hour * or declare the
'
associated system inoperable.
b.
With more than one channel inoperable, declare the
associated system inoperable.
l
ACTION 31 -
With the number of OPERABLE channels less than required by the
Minimum OPERABLE channals per Trip Function, place the inoperable
channel in the tripped condition within one hour; restore the
inoperable channel to OPERABLE status within 7 days or declare
the associated system inoperable.
-
ACTION 32 -
With the number of OPERABLE channels less than required by
the Minimum OPERABLE Channels per Trip Function requirement,
i
declare the associated ADS trip system or ECCS inoperable.
ACTION 33 -
With the number of OPERABLE c'han'nels less than the Minimum
'
OPERABLE Channel.s per Trip Function rcacirement, place the
inoperable chanr.el in the tripped condition within one hcur.
ACTICH 34 -
With the number of OPERABLE channels less than required by the
Minimum OPERABLE Channels per Trip Function requiremeht, restore
the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or
declare the associated A05 trip system or ECCS inoperable.
l
ACTION 35 -
With the number of OPERABLE channels less than required by the
Minimum OPERABLE Channels per Trip Function requirement
For one trip system, place that trip system 'in the tripped
a.
condition within one hour * or declare the HPCS system
)
j
b.
For both trip systems, declare the HPCS system inoperable.
ACTICH 35 -
With the number of CPERABLE channels less than required by the
Minimum OPERABLE Channels per Trip Function requirement, place
at least one inoperable channel in the tripped condition within
one hour * or declare the HPCS system inoperable.
ACTION 37 -
With the number of OPERABLE instruments less than the Minimum
OPERABLE INSTRUMENTS, place the inoperable instrument (s) in the
i
tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> * or declare the associated
emergency diesel generator inoperable and take the ACTION
required by Speci ication 3.8.1.1 or 3.8.1.2 as appropriate.
"Ine provisions of Specification 3.0.4 are not applicable.
LA SALLE - UNIT 2
3/4 3-27
Amendment No.27
-,w--
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,
.
TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
.
ACTION
ACTION 38
With the number of OPERA 8LE channels less than required by
the Minimum OPERA 8LE Channels per trip function requirements:
a.
With one channel inoperable, remove the inoperable channel
i
within one hour; restore the inoperable channel to
OPERABLE status within 7 days or declare the associated
ECCS systems inoperable.
b.
With both channels inoperable, restore at least one
channel to OPERA 8LE status within one hour or declare the
associated ECCS system inoperable.
l
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LA SALLE - UNIT 2
3/4 3-27(a)
.
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3/4.4 REACTOR COOLANT SYSTEM
3/4.4.1 RECIRCULATION SYSTEM
.
PECIRCULATION LOOP _S
LIMITING CONDITION FOR OPERATION
3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.
)
APPLICABILITY: "0PERATIONAL CONDITIONS in and 2".
1
ACTION:
a.
With one reactor coolant system recirculation loop not in operation:
,
1.
Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
a)
Place the recirculation flow control system in the Master
Manual mode, and
b)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety
i
Limit by 0.01 to 1.08 per Specification 2.1.2, and,
'
c)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting
-
Condition for Operation by 0.01 per Specification 3.2.3, and,
d)
Reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE
I
(
(MAPLHGR) limit to a value of 0.85 times the two recirculation
'
loop operation limit per Spscification 3.2.1, and,
e)
Reduce the Average Power Range Monitor (APRM) Scram and
i
Rod Block and Rod Block Monitor Trip Setpoints and Allowable
Values to those applicable for single loop recirculation
loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.
2.
When operating within the surveillance region specified in
Figure 3.4.1.1-1:
a)
With core flow less than 39% of rated core flow,
initiate action within 15 minutes to either:
1)
Leave the surveillance region within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or
2)
Increase core flow to greater than or equal to 39% of
rated flow within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b)
With the APRM and LPRM# neutron flux noise level greater
than three (3) times their established baseline noite
levels:
"See Special Test Exception 3.10.4.
- Detector levels A anc C of one LPRM string per core octant plus detector levels
A and C of one LPRM string in the center region of the core should be monitored.
LA SALLE - UNIT 2
3/4 4-1
Amendment No. 32
.
~
.
__
_
_
.
.
.
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , _.._ _ _____ _ .
-
.
.-
',-
'
.
.
LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued)
1)
Initiate corrective action within 15 minutes to restore
the noise levels to within the required limit within
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, otherwise
2)
Leave the surveillance region specified in
Figure 3.4.1.1-1 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3.
The provisions of Specification 3.0.4 are not applicable.
4.
Otherwise, be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
b.
With no reactor coolant system recirculation loops in operation,
immediately initiate measures to place the unit in at least HOT
SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.1
Each reactor coolant system recirculation loop flow control valve
shall be demonstrated OPER\\BLE at least once per 18 months by:
~
a.
Verifying that the control valve fails "as is" on loss of hydraulic
pressure at the hydraulic power unit, and
(
b.
Verifying that the average rate of control valve movement is:
1.
Less than or equal to 13% of stroke per second opening, and
2.
Less than or equal to 11% of stroke per second closing.
,
4.4.1.2
With one reactor coolant system recirculation loop not in operation:
a.
Establish baseline APRM and LPRM# neutron flux no'ise level values
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> upon entering the surveillance region of Figure 3.4.1.1-1
provided that the baseline values have not been established since
last refueling.
b.
When operating in the surveillance region of Figure 3.4.1.1-1, verify
that the APRM and LPRM# neutron flux noise levels are less than or
equal to three (3) times the baseline values:
1.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and
2.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after completion of a THERMAL POWER increase of at
least 5% of RATED THERMAL POWER, initiating the surveillance
within 15 minutes of completion of the increase.
c.
When operating in the surveillance region of Figure'3.4.1.1-1, verify
that core flow is greater than or equal to 39% of rated core flow at
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- Detector levels A and C of one LPRM string per core octant plus detector
(
1evels A and C of one LPRM string in the center region of the core should be
(
monitored.
LA SALLE - UNIT 2
3/4 4-2
Amendment No.32
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LA SALLE - U111T 2
3/4
4-2a
Amendment No. 32
. _
_ - _ _
'
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3/4.4.2 SAFETY / RELIEF VALVES
LIMITING CONDITION FOR OPERATION
'J . 4. 2 The safety valve function of 18 reactor coolant systes safety / relief
valves shall be OPERABLE with the specified code safety valve function lift
settings." #
a.
4 safety / relief valves e 1205 psig + N , - N
b.
4 safety / relief valves # 1195 psig + 3. -5
c.
4 safety / relief valves 01185 psig + 3, -5
,
d.
4 safety / relief valves 9 1175 psig + M , - 5
e.
2 safety / relief valves 8 1150 psig + N , - 5
-
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With the safety valve function of one or more of the above required
a.
safety / relief valves inoperable, be in at least HOT SHUTDOWN within
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
-
With one or more safety /nlief valves stuck open, provided that
b.
suppression pool average water temperature is less than 110*F, close
the stuck open relief valve (s); if unable to close the open valve (s)
(
within 2 minutes or if suppression pool average water temperature is
110*F or greater, place the reactor mode switch in the Shutdown
-
position.
c.
With one or more safety /reifef valve stas position indicators
inoperable, restore the inoperable stes position indicators to
CPERABLE status within 7 days or be in at least HOT SHUTDOWN within
the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.2.1 The safety / relief valve stes position indicators of each safety / relief
valve shall be demonstrated CPERA8LE by performance of a:
a.
CHANNEL CHECK at least once per 31 days, and a
b.
CNANNEL CALIBRATION at least once per 18 months.**
4.4.2.2
The low low set function shall be demonstrated not to interfere with
the OPERABILITY of the safety / relief valves or the ADS by performance of a
CHANNEL CALIBRATICH at least ones per 13 scnths.
.
"The lift setting pressure shall correspond to ambient conditions of the
valves at nominal operating temperatures and pressures.
- Up to two inoperable valves may be replaced with spare OPERABLE valves with
lower setpoints until the next refueling outage.
- The provisions of Specification 4.0.4 are not applicable provided the surveil-
'
lance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate
to perform the test.
LA SALLE - UNIT 2
3/4 4-6
Amendment No.15
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3/4.5 EMERGENCY CORE COOLING SYSTEMS
3/4.5.1 ECCS - OPERATING
LIMITING CONDITION FOR OPERATION
3.5.1
ECCS divisions 1, 2 and 3 shall be OPERABLE with:
a.
ECCS division I consisting of:
1.
The OPERABLE low pressure core spray (LPCS) system with a flow
path capable of taking suction from the suppression chamber and
)
transferring the water through the spray sparger to the reactor
'
vessel.
2.
The OPERABLE low pressure coolant injection (LPCI) subsystem "A"
of the RHR system with a flow path capable of taking suction from
the suppression chamber and transferring the water to the reactor
'
vessel.
-
3. .
At. least 6 OP.ERAB.L.E"" ADS .v.a.lves.
l
)
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.
.
(
b.
ECCS division 2 consisting of:
1.
TheOPERABLElowpressurecoolantinjection(LPCI) subsystems
"B" and "C" of,the RHR system, each with a flow path capable of
taking suction from the suppression chamber and transferring the
water to the reactor vessel.
)
2.
At least 6 OPERABLE"* A05 valves.
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l
.
ECCS division 3 consisting of the OPERABLE high pressure core spray
c.
(HPCS) system with a flow path capable of taking suction from the
suppression chamber and transferring the water through the spray
sparger to the reactor vessel.
APPLICABILITY: OPERATIONAL CONDITION 1, 2"# and 3*.
"The ADS is not required to be OPERABLE when reactor steam dome pressure is
less than or equal to 122 psig.
- "See Specification 3.3.3 for trip system operability.
- See Special Test Exception 3.10.0.
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LA SALLE - UNIT 2
3/4 5-1
Amendment No.27
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EMERGENCY CORE COOLING SYSTEMS
LIMITING CONDITION F0,R OPERATION (Continued)
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ACTION:
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a.
For ECCS division 1; provided that ECCS divisions-tand 3 are OPERA 8LE:
. 1,
With the LPCS system inoperable, restore the inoperable LPCS
systes to OPERA 8LE status within 7 days.
2.
With LPCI subsysten "A" inoperable, restore the inoperable LPCI
subsystes "A" to OPERABLE status within 7 days.
3.
With the LPCS system inoperable and LPCI subsystem "A" inoperable,
restore at least the inoperable LPCI subsystem "A"
or the
'
inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.
Othenvise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
For ECCS division 2, provided that ECCS divisions 1 and 3 are OPERABLE:
.
1.
With either LPCI subsystem "B" or "C" inoperable, restore the
(
inoperable LPCI subsystem "B" or "C" to OPERA 8LE status within
7 days.
,
2.
With both LPCI subsystems "B" and "C" inoperable, restore at least
the inoperable LPCI subsystes "B" or "C" to OPERABLE status
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
,
3.
Otherwise, be in at least HOT SHUTDOWN with n the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />".
c.
For ECCS division 3, provided that ECCS divisions 1 and 2 and the
i
RCIC systas are OPERABLE:
,
.
1.
With ECCS division 3 inoperable, restore the inoperable division
to OPERA 8LE status within 14 days.
2.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
i
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
'
d.
For ECCS divisions 1 and 2, provided that ECCS division 3 is OPERABLE:
1.
With LPCI subsystem "A" and either LPCI subsystem "B" or "C."
inoperable, restore at least the inoperable LPCI subsystem "A"
or inoperable LPCI subsystem "B" or "C" to OPERABLE status within
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
"Whenever two or more RHR subsystems are inoperable, if unable to attain COLD
(
SHUTDOWN as required by this ACTION, maintain r63ctor coolant temperature as
low as practical by use of alternate heat removal methods.
LA SALLE - UNIT 2
3/4 5-2
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- -EMERGENCY;CORfC00tlim 5Ynu45
LINITING CONDITION FOR OPERATION (Continued)
._
ACTION: (Continued)
2.
With the LPCS system inoperable and eithe'r1PCI subsystems "B" or
"C" inoperable, restore at least the inoperable LPCS system or
inoperable LPCI subsystem "B" or "C" to OPERABLE status within
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.,
-
3.
Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />".
~
e.
For ECCS divisions 1 and 2, provided that ECCS division 3 is
OPERA 8LE and divisions 1 and 2 are othemise OPERABLE:
1.
With one of the above required A05 valves inoperable, restore the
inoperable ADS valve to OPERABLE status within 14 days or be in
at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor
steam dome pressure to 1122 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
With two or more of the above required ADS valves inoperable,
be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor
steam dome pressure to i 122 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
_.
f.
With an ECCS discharge line "keep filled" pressure alarm instrumenta-
(
tion channel incperable, perform Surveillance Requirement 4.5.1.a.1
at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
.
g.
With an ECCS header delta P instrumentation channel inoperable,
restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
or determine ECCS header delta P locally at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />;
othemise, declare the associats4 ECCS inoperable.
h.
With Surveillance Requirement 4.5.1.d.2 not~ performed at the required
interval due to low reactor steam pressure, the provisions of Specifi-
cation 4.0.4 are not applicable provided the surveillance is performed
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform
the test.
i.
In the event an ECCS system is actuated and injects water into the
Reactor Coolant System, a Special Report shall be prepared and
submitted to the Comission pursuant to Specification 6.6.C within
90 days describing the circumstances of the actuation and the total
accumulated actuation cycles to date. The current value of the
usage factor for aach affected safety injection nozzle shall be
provided in this Special Report whenever its value exceeds 0.70.
,
j.
With one or more ECCS corner room watertight doors inoperable, restore
all the inoperable ECCS corner room watertight doors to OPERABLE
status within 14 days, othemise, be in at least HOT SHUTDOWN within
the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
"Whenever two or more RHR subsystems are inoperable, if unable to attain COLD
(
SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as
low as practical by use of alternate heat removal methods.
LA SALLE - UNIT 2
3/4 S-3
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EMERGENCY CORE COOLING SYSTEMS
. SURVEILLANCE REQUIRENENTS
4.5.1 ECCS divisions 1, 2, and 3 shall be demonstrated OPERA 8LE by:
a. .
At least once per 31 days for the LPCS, LPCI, and HPCS systems:
1.
Verifying by venting at the high point vents that the system
piping from the pump discharge valve to the system isolation
valve is filled with water.
2.
Performance of a CHANNEL FUNCTIONAL TEST of the:
a)
Discharge line "keep filled" pressure alarm instrumentation,
'
.
and
b)
Header delta P instrumentation.
3.
Verifying that each valve (manual, power-operated, or automatic,)
fn the flow path that is not locked, sealed, or otherwise
-~
secured in position, is in its correct position.
4.
Verifying that each ECCS corner room watertight door is closed, .
except during entry to and exit from the room.
b.
Verifying that, when tested pursuant to Specification 4.0.5, each:
1.
LPCS pump develops a flow of at least 6350 gpm against a
test line prassure greater than or equal to 290 psig.
2.
LPCI pump develops a flow of at least 7200 gpa against a test
line pressure greater than or equal to 130 psig.
3.
HPCS pump develops a flow of at least 6200 gpm against a test
line pressure greater than or equal to 330 psig.
c.
For the LPCS, LPCI and HPCS systems, at least once per 18 months:
1.
Performing a system functional test which includes simulated
automatic actuation of the system throughout its emergency
operating sequence and verifying that eacn automatic valve in
the flow path actuates to its correct position. Actual injection
of coolant into the reactor vessel may be excluded from this test.
.
LA SALLE - UNIT 2
3/4 5-4
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bkbRGENCYbbREbOOLINGSSTEMS
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SURVEILLANCE REQUIREMENTS (Continued)
2.
Perfoming a CHANNEL CALIBRATION of the:
a)
Discharge line "keep filled pressure alarm instrumentation
d
and verifying the:
1)
High pressure setpoint and the low pressure setpoint
of the:
(a) LPCS system to be 5 500 psig and > 55 psig,
respectively.
(b) LPCI subsystems to be 1 400 psig and 1 55 psig,
respec*.i vely.
2)
Low pressure setpoint of the HPCS system to be 3
63 psig.
b)
Header delta P instrumentation and verifying the setpoint
_.
of the:
(
1)
LPCS system and LPCI subsystems to be t 1 psid.
,
,
2)
HPCS system to be 5 1 2.0 psid greater than the
normal indicated AP.
3.
Verifying that the suction for the HPCS system is automatically
transferred from the condensata storage tank to the suppression
,
chamber on a condensate storage tank low water level signal and
on a suppression chamber high water level signal.
4.
Visually inspecting the ECCS corner room watartight door seals
and room penetration seals and verifying no abnomal degradation,
damage, or obstructions.
d.
For the A05 by:
1.
At laut once per 31 days, performing a CHANNEL FUNCTIONAL TEST
of the accumulator backup compressed gas system low pressure
alam system.
2.
At least once per 18 months:
a)
Performing a system functional test which includes simulated
automatic actuation of the system throughout its emergency
operating sequence, but excluding actual valve actuation.
b)
Hanually opening each ADS valve and observing the expected
,
change in the indicated valve position,
c)
Perform 1ng a CHANNEL CALIBRATION of the accumulator backup
compressed gas system low pressure alam system and verifying
an alarm setpoint of 500 + 40, - O psig on decreasing pressure.
LA SALLE - UNIT 2
3/4 5-5
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EMERGENCY CORE COOLING SYSTEMS
3/4.5.2 ECCS - SHUTDOWN
.
LIMITING CONDITION FOR OPERATION
3.5.2 At least two of the following shall be OPERABLE:
a.
The low pressure core spray (LPCS) system with a flow path capable
of taking suction from the suppression chamber and transferring the
water through the spray sparger to the reactor vessel.
b.
Low pressure coolant injection (LPCI) subsystem "A" of the RHR system
with a flow path capable of taking suction from the suppression
'1
chamber upon being manually realigned and trauferring the wster to
the reactor vessel.
c.
Low pressure coolant injection ( OCI) subsystem "B" of the RHR system
with a flow path capable of taking suction from the suppression chamber
upon being anually realigned and transferring the water to the reactor
vessel.
d.
Low pressure coolant injection (LPCI) subsystem "C" of the RHR system
with a flow path capable of taking suction from the suppression
(
chamber upon being manually realigned and transferring the water to
the reactor vessel.
e.
The high pressure core spray (HPCS) system with a flow path capable
of taking suction from one of the following water sources and trans-
ferring the water through the spray sparger to the reactor vessel:
1.
From the suppression chamber, or
)
2.
When the suppression pool level is less than the limit or is
drained, from the condensate storage tank containing at least
135,000 available gallons of water, equivalent to a level of
14.5 feet.
APPLICABILIT/: OPERATIONAL CONDITION 4 or 5*.
ACTION:
a.
With one of the above required subsystems / systems inoperable, restore
at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or
suspend all operations that have a potential for draining the reactor
vessel.
,
b.
With both of the above M auired subsystems / systems inoperable,
suspend CORE ALTERATIONS aad aP 6perations that have a potential
for draining the reactor vessel. Restore at least one subsystem /
'
system to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish SECONDARY
CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
"The ECC5 is not required to be OPERABLE provided that the reactor vessel head
is removed, the cavity is flooded, the spent fur' por* qates are removed, and
water level is maintained within the limits of cee'
tions 3.9.8 and 3.9.9.
LA SALLE - UNIT 2
3/4 5-
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ENERGENCY CORE COOLING SYSTEMS
SURVEILLANCE REQUIREMENTS
..
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4.5.2.1 At least the above required ECCS shall be demonstrated OPERA 8LE per
Surveillance Requirement 4.5.1, except that the header delta P instrumentation
is not required to be OPERA 8LE. .
, ,
-
4.5.2.2 The HPCS system shall be determined OPERABLE at least once per
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensate storage tank required volume when the
condensate storage tank is required to be OPERABLE per Specification 3.5.2.e.
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LA SALLE - UNIT 2
3/4 5-7
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GELECTED GENERAL GTATIONS EMERGENCY PROCEDUREG
LZP-1200-1
CLASSIFICATION OF GSEP CONDITIONS
LZP-1200-2
CLASSIFICATI0t10F NOBLE GAS RELEASE
LZP-1200-3
CLASSIFICATION OF AN IODINE RELEASE
LZP-1200-5
GSEP GUIDELINES FOR RECOMMENDED OFFSITE PROTECTIVE
ACTIONG
LZP-1210 2
NUCLEAR ACCIDENT REPORTING GYSTEM (NARG)'FCRM
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