ML20215J875
| ML20215J875 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 04/30/1987 |
| From: | Coe D, Collins S, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215J834 | List: |
| References | |
| 50-289-87-07OL, 50-289-87-7OL, NUDOCS 8705080305 | |
| Download: ML20215J875 (120) | |
See also: IR 05000289/1987007
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Evaluation Report No.: 87-07 (OL)
Facility Docket No.: -50-289
'
Facility License No.: OPR-50
License: GPU Nuclear Corporation
P.O. Box 480
Middletown, Pennsylvania 17057
Facility Name: Three Mile Island Unit 1
. Evaluation Dates: February 20, 1987 to March 12, 1987
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Chief Examiner: ,
- 2- @
D. Coe, Lead Reac Engineer IDateF
Reviewed By:
_R. Ke'ller, Chief, PFojects Section 1C
~
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Date
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Approved By: d N1/# MN7
S. Collins, Deputy Jirector Date
Division of Reactor Projects
Summary:
The administration of the facility's annual requalification examinations was
audited by the NRC. The effectiveness of the training department in conducting
the required evaluations was evaluated by the NRC as satisfactory overall.
However, as a result of this audit, the NRC identified two concerns which
require action by the facility licensee. These are related to the documenta-
tion of individual weaknesses identified in the simulator examination, and-the
necessity to evaluate senior operators who are routinely assigned only reactor
operator duties to an SRO level of knowledge and ability.
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DETAILS
1. Examination Results
R0 SRO Total
Pass / Fail Pass / Fail Pass / Fail Evaluation
Written 1/2 3/2 4/4 Marginal
Simulator 7/0 6/0 13/0 Satisfactory
Oral 1/0 3/0 4/0 Satisfactory
Evaluation of Facility Written Examination Grading: Satisfactory
Overall Program Evaluation: Satisfactory
2. Scope
The facility prepared R0 and SRO written examinations were reviewed by the
NRC prior to their administration on the three consecutive days beginning
with March 4,1987. Sections one (1) and five (5), Reactor Theory and
Heat Transfer, of the R0 and SR0 examinations, respectively, had previously
been administered by the licensee in December 1986, and were not reviewed
by the NRC. On March 5,1987, the NRC replaced 26% of the facility R0
examination and 21% of the facility SRO examination with NRC written ques-
tions as an independent check on the validity of the examination results.
The NRC then audited the facility grading of eight (8) randomly selected
written examinations administered on March 5,1987.
On February 20 and 27,1987, the NRC observed facility conducted simulator
examinations of three operating crews and on March 11 and 12,1987, ob-
served facility conducted oral examinations of three SR0's and one R0.
3. Review and Audit of Written Examination
The annual written requalification examination prepared by the facility
training department is identical in format to an NRC replacement written
examination. The NRC review of Sections 2, 3, 4, 6, 7 and 8 noted the
following:
a. Strengths
(1) Questions used on a given exam day were not used on any other
exam day. Some questions were shared by both the R0 and SR0
exams on any given day. This is considered desirable provided
generic training weaknesses can still be identified (see item
2).
3
(2) Topical coverage was similar on all eums, even though each
question was not used on more than one day, thus generic weak-
nesses in a topical area could be identified.
(3) Examination authors and supervisors were scheduled such that
they took an examination to which they did not contribute or
review. In this manner, all licensed operators took a written
examination; there were no exceptions.
b. Weakness
A detailed review of the R0 and SR0 exams to be administered on
March 5, revealed a large number of recall questions (65% for the R0
and 75% for the SRO) as opposed to questions which require analysis,
synthesis and evaluation. Recall questions typically ask for lists,
definitions, or require true/ false and multiple choice answers. Al-
though no regulatory standard exists which strictly defines this type
of question or places limits on its use, the NRC feels that 75% re- l
call oriented questions on a SRO examination is excessive. The other
two SRO exams reviewed contained approximately 64% and 38% recall
oriented questions. In addition, the licensee's requalification
program description 6211-PGD-2611.01 paragraph 7.5.1.B(3) requires
that " questions requiring analysis and or explanation should pre-
dominate."
The NRC rev:ewed the facility grading of eight randomly selected examina-
tions given on March 5, 1987. The guidelines of NUREG-1021, Quality
Assurance Checkoff Sheet ES-108-1, were followed and no inconsistencies or
objectionable grading practices were found. The NRC agreed with the pass /
fail results for all examinations reviewed.
An analysis was made of operator performance on NRC written questions as
compared to performance or facility written questions in terms of the per-
centage of available points attained for each. These results are shown on
Attachment 1. Senior Operator performance on the NRC written questions
was on the average 10.1's below that for facility written questions. Nor-
mally, a difference of 10*4 or greater is considered significant, but in
this case a significant contribution to this difference was directly at-
tributable to one part of one question for which 4 out of 5 operators
missed full credit. This question, which requires the operator to class-
ify a General Emergency in accordance with the facility Emergency Plan
given indications of significant fission barrier degradation, identified a
generic weakness and is noted as such.
This analysis shows that the NRC written questions were perhaps slightly
more challenging but substantially of the same level of difficulty as the
facility written questions. Furthermore, since operator performance on
NRC written questions was not significantly different than performance on
facility written questions, the validity of the examination results were
independently confirmed.
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The results of the written examination are also shown for all TMI-1 oper-
ators on Attachment 1. All written exam failures (three R0's and three
SR0's) occurred within t7e group examined on March 5,1987, and for which
NRC question substituticn took place. A review of marginally passing
examinations given-on the two alternate exam dates showed no significant
difference in grading practice between any of the three sets of examina-
tions. Therefore, because the examinations on all three days were of
approximately the same level of difficulty, the grading of all exams was
substantially the same, and NRC written question influence was not signif-
icant in general, the reason for the high failure rate on March 5,1987,
must be only due to kno41 edge weakness on the part of those operators who
failed.
For each of the six operators who failed the written requalification exam-
ination, a review was conducted of past written requalification examina-
tion performance. One of these operators terminated employment with the
licensee shortly after this examination. Two of the remaining five oper-
ators had passed an NRC licensing examination the previous year, and of
the other three operators, only one had a failure over the previous two
years. Thus, there does not appear to be a pattern of failure for any of
these individuals.
4. Audit of Simulator Examinations
Three operating crews were audited during facility administered simulator
requalification examinations. Two of these were shift crews normally
assigned to an operating shift and one was a group of off-shift licensed
staff personnel,
a. Strengths
(1) Each scenario prepared by the facility contained a detailed
expected sequence of events and key points for evaluation in-
ciuding references to applicable procedures.
(2) Ten scenarios were prepared from which two, three, or four were
chosen to evaluate each group of operators. The ten scenarios
covered a wide range of normal, abnormal, and emergency proced-
,
ures. In addition, single event drills (turbine trip) and on-
l the-spot modifications were sometimes used for variety. These
were adequate to sufficiently evaluate the operators.
(3) Strong operations department participation was evidenced by one
evaluator being from operations department management (Plant
Operation Director or higher) for each group evaluated.
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(4) Post-Trip Emergency procedures and Energency Plan actions were
heavily evaluated throughout all scenarios.
(5) Facility evaluators were highly observant of all actions taking
place during the simulator evaluation and' appeared to detect all
deficiencies which occurred.
b. Weaknesses
(1) Although some scenario descriptions contained events which re-
quired actions to be taken or decisions to be made based on the
guidance contained in Technical Specifications, they were not
explicit in their description of the key points to be evaluated
in this area. Technical Specifications were' not as heavily
evaluated in general as were the ATP's and other procedures.
(2) The evaluation forms used by the facility evaluators provide !
seven general areas for evaluation. Supporting documents which
define these seven areas strongly indicate that .the individual
is to be evaluated. This follows the requirements of the facil-
ity requalification procedure 6211-PGD-2611.01 Section 7.3.5,
.
Skills Evaluation System, which states, in part, "Each Itcensed
individual's performance shall be evaluated... annually during
Nuclear Plant simulator exercises." In practice, however, only
team performance is assessed by facility evaluators for each of
the seven categories. Therefore, no single document exists
which provides an evaluation of any one individual's performance.
(3) When facility evaluators noted team performance deficiencies,
there was little follow-up questioning to identify the individ-
'
ual weaknesses which contributed toward the overall deficiency.
Occasions were observed in which operator actions or behavior
- should have prompted follow-up questioning to determine the full
extent of individual knowledge weakness. #
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(4) The deficiencies noted by facility evaluators were generally
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attributed to the crew as a whole, whereas strengths and proper
actions were of ten attributed to individuals. Thus individual
weaknesses were not always documented.
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- ' (5) Although sufficient facility evaluators were present to provide
thorough coverage of all actions occurring during the simulator
i scenarios, evaluators were not assigned to evaluate a single
licensed individual but rather were given a " functional area"
! or physical portion of the control room to monitor.
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The program weaknesses identified in paragraphs 4.b.(2) through (5)
stem from the licensee's stated intent to conduct simulator evalua-
tions primarily from a team perspective. However, the requirements
of the licensee's requalification program and the use of these evalu-
ations to certify to the NRC that an individual has been found to
have the requisite knowledge and ability to renew their license at
the end of its term, require that the individual's performance be
fully assessed and documented during these evaluations. The facility
licensee has responded to this concern by explaining what actions
have been taken or are planned to improve this area of weakness.
5. Audit of Oral Examinations
The facility's oral requalification examinations of four operators were
audited. The four individuals consisted of three SR0's and one R0. One
of the SR0's was a shift foreman, another was assigned only to CR0 duties
(an R0 level of responsibility), and the third was off-shif t. The facil-
ity evaluators included training and operations personnel at the shift
supervisor / shift foreman level,
a. Strengths
(1) The evaluation form is thorough and requires detailed coverage
of theory, I and C, systems, procedures, and radiation control.
It also required coverage of recently modified or installed
safety systems.
(2) Three out of the four examinations audited were at an appropri-
ate level of knowledge and were adequate to identify individual
knowledge weakness,
b. Weakness
One examination was observed in which the examinee, who holds an SR0
license but is assigned only R0 duties, was examined at an R0 level.
In addition, the NRC noted that this individual was not evaluated in
an SR0 position during the simulator examination. Thus, only the
written examination tested this operator at a level commensurate with
his license. The licensee's requalification program description
6211-PGD-2611.01 paragraph 7.5.2.A(1) requires that "the oral exam-
ination should contain questions covering ... licensed duties and
responsibilities of the operating position corresponding to the
individual's license level."
The facility licensee is requested to respond to this concern by
explaining what actions have been taken or are planned to improve
this area of weakness.
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6. Exit Interview
An exit ' interview was held on March 12, 1987, following the completion of
all audits. The following persons were present: *
NRC Personnel
I
D. H. Coe, Chief Examiner
B. S. Norris, Examiner
R.-J. Conte, Senior Resident Inspector
F. I. Young, Resident Inspector
Facility Personnel .
t
H. D. Hukill, Director, TMI-1
4- M. J. Ross, Operations Director, TMI-1
j. O. J. Shalikashvili, Manager, Plant Training
{ W. W. Thompson, Manager, Operator Training
+
R. H. Maag, Supervisor, Licensed Operator Training
. .
Summary of NRC Comments
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- A discussion was held concerning the objectives and methods of this re-
qualification program audit and it was emphasized that the licensee's
,
requalification evaluation process was being evaluated. The NRC pre-
sented the observations of program weaknesses described 'in section 3, '4
, and 5 of this report with the exception of item 4.b.(1). The NRC stated
1 that the criteria for evaluating the requalification program using the
!
method of auditing the facility's evaluation process was not well defined.. "
i
No preliminary results were given.
p Summary of Facility Comments or Commitments
1_
The . facility committed to providing the NRC with a letter documenting the
actions they would take to address the concern raised regarding the con-
j. duct of simulator evaluations.
!
7. Conclusion '
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The Three Mile Island Unit I requalification evaluation program is eval- -
uated as satisfactory, but two areas of weakness were identified which
should be improved. These are as follows:
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'(1) Incomplete documentation of indiv'idual weaknesses identified from the
simulator examination.
(2) Lack 'of provisions to ensure SR0 license holders are evalu'ated at an
SR0 level 'during simulator and oral examinations, even though they
may be routinely assigned to R0 duties.
The licensee was requested to respond to these concerns by explaining what
actions have been taken or are planned to imprcve these areas of weakness.
The iicensee responded to item (1) by letter dated April 1,1987 from
H. D. Hukill to T. E. Murley. This letter committed the licensee to docu-
ment individual performance during annual simulator requalification exam-
inations.- This satisfies the basic NRC concern represented in item (1).
This commitment will be verified af ter the next annual simulator requalif-
ication examination (0 pen Item 87-07-01).
Attachments:
1. Written Examination Results
2. March 5,1987, Written Requalification Exam (RO)
3. March 5, 1987, Written Requalification Exam (SRO)
4. Facility Comments on NRC Questions
5. NRC Resolutions to Facility Comments
ATTACHMENT 1
Results of Operator Performance on Facility versus NRC Written Requalification
Exam Questions for March 5, 1987 Exam.
R0 Exams SR0 Exams
(3 sampled) (5 sampled)
Points available from facility questions: 80.5 84.2
Average points awarded for facility questions: 64.9 72.1
Percentage of points awarded: 80.6% 85.6%
Points available frcm NRC questions: 19.5 15.7
Average points awarded for NRC questions: 15.2 11.9
Percentage of points awarded: 78.1% 75.5%
Difference between NRC and facility question 2.5% 10.1% ;
performance
Overall Results
R0 R0 Average R0 Pass SR0 SRO Average SRO Pass
Exam Datq Exams Grade Rate Exams Grade Rate
3/4/87 4 88 100% 12 88 100%
3/5/87* 5 79 40% 10 84 70%
3/6/87 5 89 100% 9 90 100%
_.
Total 14 85 78% 31 87 90%
- NRC replaced 26% of facility written R0 questions and 21% of facility written
SRO questions.
Note: Fatsing criteria was > 70% on each section and > 80% overall.
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e N% MCdment 2
Attachment 1
.j WRITTEN EXAMINATION CERTIFICATION COVER SHEET
NAML (PLEASE PRINT) EXAM
(FIRST, MID. INITIAL, LAST): DATE: 5 March 87
EMPLOYER (CGIPANY): EMPLOYEE NO: 50C. SEC. NUMBER:
EXM EXAM LOCATION:
TITLE: Annual Requal Exam R0 CATEGORY: 1 118/119
I EXAM NO:
GENERAL INSTRUCTIONS AND GUIDELINES i R0-1
- PLEASE READ THE FOLLOWING INSTRUCTIONS CAREFULLY:
1. Remain seated and quiet during the examination.
2. Please raise your hand when: you have any questions on the examination
you have finished the examination
3. You are required to do your own work and you are not to help anyone else.
4 Use only the reference material authorized below.
5. If you must leave the room before you finish, the examination must be
returned to the proctor. Note that instructions #3 and #4 above still
apply while you are out of the room.
6. Misconduct or cheating on examinations will result in disciplinary action
on the part of the Company, and possibly additional civil and/or criminal
sanctions.
7. At the conclusion of this examination, you are to sign the following
certi fication. ~
CERTIFICATION
I certify that all answers contained in this examination are my own, that
I have neither received nor given unauthorized assistance, and that I have
not used any unauthorized references.
SIGNATURE: DATE:
- 00 NOT BEGIN THE EXAMINATION BEFORE THE PROCTOR REVIEWS THE REMAINDER OF THIS
PAGE WITH YOU.
- e*****************************************..************************
AUTHORIZED REFERENCE MATERIALS: 1 TIME START STOP
Attachments l LIMIT TIME TIME I I OPEN BOOK
l 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 1 xxl CLOSED 800K *
- PAGE CHECK THE EXAMINATION TO ENSURE YOUR COPY 15 C(NPLETE.
- SPECIAL INSTRUCTIONS:
1. Use only black ink or pencil (#2 or softer).
2. Answer on the exam pages.
SECTION l POINTS SCORE I % SECTION l POINTS I SCORE I % l
Previous Exam I IV
l j l 25
l l l
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1 I I I I I
III l 25 l l l TOTAL I l l l
MINIMUM ACCEPTABLE GRADES: EACH SECTION: 70.0 % OVERALL 80.0%
GRADED BY
(EXAMINER'S SIGNATURE): DATE:
Developed /Submi tted: 67 Date aVAB7
Reviewed: vn b Date /JO/7 l
Approved:
0412K
24 6 // Date & 7 ~
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Plant Design, Safety and Emergency Systems
(1.5) 2.1 What action have to be taken to allow the control room operator ,
to provide river water to the suction of the Emergency Feedwater
Pumps? ~ State any interlocks that may be imposed.
!Ans. 1. Spectacle flange between RF-V-4 & 5 has to be reversed.
!(0.5 ea.) 2. Must unlock and close breakers for EF-V-4 & 5.
! 3. Reactor River Pumps (s) have to be running before EF-V-4 & 5
can be opened.
!Ref. EFW change mod R0 B-2
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(1.5) 2.2 a. How long would it take EF-P-2A to come up to full speed
following an ESAS actuation. Explain how you arrived at
your time. (1.0)
b. What is the fail position of EF-V-30's on total loss of air
to the valves.
!Ans. a. 30 see to full speed (0.25)*
1 15 see to receive start signal (0.25)
! block 4 permissive than 5 sec delay (0.25)
! and 10 see to come up to speed (0.25)
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! b. closed (0.5)
'! RO A-3
! SRO A-3
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(1.5) 2.3 TRUE/ FALSE and EXPLAIN
Pressing the Manual ESAS 30 psig actuation pushbuttons will
start the Building Spray Pumps.
!Ans. (0.5) False
!(1.0) 8. S. pumps must be started at their extension controls
'
! or
'
! The 30# nunual actuation PB only causes the JHf valves to
reposition. /c/h/.5
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.
!
! 8.S. Pumps are auto started by 2/3 30# R. B. pressure Switches.
!Ref. LP-11.2.01.127 B.S. R0 A-2
.
!
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(2.75) 2.4 Answer the following questions concerning the Core Flood System:
a. What will cause the following valves to automatically
close: (1.25)
(1) CF-V-19A (Fill Line)
(2) CF-V-3A (Vent Valve)
(3) CF-V-20A (Drain Line)
b. What are the Tech Specs concerning CF-V-1A/B. (1.5)
!Ans. a. (1) CF-V-19A closes on RTI (0.25) and 4# ES (0.25)
! (2) CF-V-3A none (0.25)
! (3) CF-V-20A closes on RTI (0.25) and 4# ES (0.25)
! b. g
TheCFfV,,1/swillbeassuredcpenbyadministrativeEnd
contro position ind
The breakers shall be open and conspicuously marked.
A-1
fo ^n fo?zr)FkrEF
!Ref. Core Flood LP R0
! 11.2.01.014 SRO A-1
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(3.5) 2.5 One of the functions of the Decay Heat Removal System is to
provide suction to the sake-up pumps in a " Piggy-Back" alignment.
a. What conditions during a LOCA would make this alignment
necessary? (1.0)
b. Why are MU-V-14A/B lef t open in this alignment? (0.75)
c. Draw the Piggy-Back alignment showing water source, major
pumps, coolers and valves up to the point of injection into
the RCS. Include both trains. (1.75)
!Ans. a. If BWST reaches 10-10 level alarm (36") (0.5) before LPI
flow is established (0.5).
! b. To protect the make-up pumps from a loss of suction
pressure if the LPI pumps were to trip (0.75).
! c. See Attached
! 11.2.01.019 A-2 A-2
! OP 1210-6 A-3 A-3
! A-1 A-1
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(3.5) 2.6 An ESAS actuation has occurred and you have bypassed the ESAS
signals and are regaining control of the equipment when a loss
of off-site power occurs. Both diesel generators start.
a. What three conditions have to be met for the diesel
generator breakers to close onto the bus? (Other than bus
UV.) (1.5)
b. What happens to the ESAS equipment when the diesel
generators energize the buses. Why? (1.5)
c. True/ False
The 27/86 lockout relays for non-essential equipment will
not reset unless the amber disagreement lights have been
reset for the equipment. (0.5)
!Ans. a. - Up to voltage (0.5)
!(any three) - Up to frequency (0.5)
! - Bus normal supply breaker open (0.5)
! - EMF timer (0.5)
! b. Previously running Block 1 equipment energizes (0.5) block
loading does not occur (0.5) because the ESAS signal is
> bypassed. (0.5)
! c. False (0.5)
! 11.2.01.129 A-1 A-1
! A-3 A-3
! A-3 A-3
6.0 1434R
, . ., .. .- .- . . . --
-- . .. -- .
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.
(1.0) 2.7 There are numerous design features built in to the spent fuel
cooling system to prevent or reduce the loss of spent fuel
pool (s) inventory under normal operating conditions. Describe
or list four (4) of these features. (Like features at different.
locations may be treated individually if the location (s) are
identified.)
!Ans. 1. No Bottom Drains (on Pools)
!(any 4) 2. Syphon Break on A Pool suction
!(0.25 ea.) 3. Syphon Break on Cask Load Pit drain
! 4. Normal suction lines are located high in the Pools
! 5. Gate can be inserted between Pools (A&B)
!Ref. SF Cooling LP R0 A-2
! 11.2.01.146 SRO A-1
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(3.00) 2.8 Refer to Figure 1 to answer the following questions, consider each
case separately:
(0.80) a. How will the system lineup change if the second Letdown cooler is
to.be placed in service?
(0.80) b. Assume a spurious 30 psig Reactor Building ESAS signal is
received. What automatic actions will occur in the ICCW system?
(0.60) c. Assume the 30 psig Reactor Building ESAS signal is concurrent
with an undervoltage on the 1D 4160 volt bus. What automatic
actions will occur in the ICCW system?
.(0.80) d. What TWO automatic actions would happen directly as a result of
ICCW flow decreasing to 500 gpm followed immediately by a loss of
the running Makeup pump?
!(0.40 each) a. Start the second ICCW pump
1 and put the second ICCW cooler into service
./f*
1(0.10 each) b. IC-V2/3/4/6 shut
1(0.66). IC-V74 opens
.so
!(0.30 each) c. Pump 1A will trip and be locked out
! Pump 1B will start
1(0.40 each) d. 1. Standby pump will auto start (<550 gpm ICCW flow)
! 2. RCPs will auto trip (<550 gpm ICCW flow & <22 gpm seal flow)
! Ref: THI-1 OPM Vol 1, Chpt B-10, pgs 4-8, & 12
l TMI-1 OPM Vol 1, Chpt B-2, pg 25
I K&A 008000K102/IF 3.3
1 000026K302/IF 3.6
1 000015A210/IF 3.7
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(0.5) 2.9 One of the Limits and Precautions in the Instrument Air
Procedure (1104-25) state that Backup Instrument Air should not
be used to pressurize instrument air lines except during Bul A
actuation conditions. What is the purpose for the precaution?
!Ans. BUIA is not DRIED clean and oil free (1104-25 only mention dry
others are exceptable)
!Ref. 11.2.01.053 R0 B-1
! SRO A-1
.
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(1.25) 2.10 Describe how the operation of the Miscellaneous Waste Evaporator
can give false indication of an OTSG Tube Leak.
!Ans. (g f)
! if the MWE has a leak in the Steam Tube Bundle contamination can
enter the condensategeturn unit and be cumped to the Main
Condensen?PRM-A-iP46 bid detect this contamination and alert the
operator to a Dossible OTSG Tube Leak. (1.25)
!Ref. R0 A-3
! SRO A-3
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(1.0) 2.11 The Recovery procedure from an MU-V-3 closure on High Letdown
Temperature has the operator close MU-V-6 A/B and open MU-V-70.
Explain what operation of these valves accomplish and why this
is necessary.
!Ans.
!(1.0) These valves isolate (6's) and bypass (70) the demineralizer to
Drevent damaqe to the resin due to high temperature (Potential for
melt / damage to resin).
!Ref. MU & P LP R0 A-1
! 11.2.01.069 SRO A-1
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(2.0) 2.12 Leakage from the RB Emergency Cooling Coils is a concern during
normal and emergency conditions. How is leakage detected during
each of '4hese conditions? Describe how the leakage is detected.
!Ans. (0.5) Normal - Rotometer indicating
!(0.5) flow f rom the Nuclear Services Closed Coolina System to the
coils.
!(0.5) Emergency - RB Leak Detection System
!(0,5) River water inlet flow is compared to temperature compensated
outlet flow.
!Ref R0 A-1
! SR0 A-1
12.0 1434R
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(1.0) 2.14 What is the reason for having separate source of cooling water
to the A and C :nake up pumps? (i.e., A DCCW vs B DCCW)
!Ans. Separate cooling water source are used to meet the separation
and redundancy requirement of an ESAS system.
!Ref. 11.2.10.069 R0 A-4
! T.S. SRO A-4
14.0 1434R
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(1.0) 2.15 Indicate whether each of the following control rod breaker trips
WOULD or WOULD NOT result in a reactor trip.
a. 11, CC and E
b. 10 and 11
c. CC, 00, E and F
d. 10, CCand E
!Ans. (0.25 ea.)
! a. would
! b. would
! c. would
! d. would not
,
!Ref. 11.2.01.132 R0 A-3
! SRO A-3
.
A
4 End of Section II
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SECTION III
INSTRUMENTATION AND CONTROL
(1.75). 3.1 Refer to Figure 2. The reactor is at 100% power with all ICS
stations in automatic, Group 7 control rods are 70% withdrawn, and
NI-5 is selected for ICS input. Assuming no operator action, what
will be the effect on the ICS and the plant if NI-5 fails low?
l(0.25 each) 1.- Rods will withdraw at 30"/ min
1 2. Unit will go into track (neutron cross limits)
1 3. Feedwater flow will decrease (due to neutron cross limits)
1 4. Hegawatts will decrease (due to turbine header pressure
'
! decreasing)
1 5. Tavg will increase rapidly (due to My decreasing while rods are
! moving out)
! 6. Pressurizer level will increase (due to Tavg increasing)
! 7. Reactor trip on high pressure
! Ref: TMI-1 OPM Vol 3, Chpt F-3, pgs 173-175 ,
.I K&A 015000K304/IF 3.4
!
!
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.
- -
. (1.25) 3.2 TRUE/ FALSE
On a loss of ICS-NNI Auto Power feed pump speed fails to 50%
demand and blocks manual pump control. This condition provides
the potential for an severe plant transient. Explain your
answer.
!Ans. (0.5) False
!(0.75) Feed Pump control transfers to hand with no speed change and
gives the operator control. (Control scheme change removes auto'
powered module from blocking hand control.)
!Ref. TMI-l Loss of ICS-Auto Power B-3
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(2.0) 3.3 List, in sequence, the four (4) major automatic action that
occur in the Reactor Building Emergency Cooling System following
an ES signal without loss of off-site power. Assume the system
initially in a normal line up for power operation.
!Ans. (0.5) 1. The AH-E-1 fans operating will trip on a block one loading
signal.
!(0.5) 2. All three (3) fans will start automatically and operate at
their slow speed on block two loading signal.
!(0.5) 3. The Emergency Cooling System will go into its emergency
mode of operation, opening the necessary valves
automatically and starting the river water pumps (RR-Pl A/B)
to establish flow through the emergency cooling coils.
!(0.5) 4. The nornel cooling coils penetration isolation valves
(RB-V2 and RB-V7) will close.
!Ref. RB Emergency Cooling Lesson B-2
! 11.2.01.126
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3.4 During operation at power, will pressurizer level read higher or l
(2.25)
lower than actual (and explain why) if: l
a. Temperature compensation is lost. l
b. dp Cell connection to tank top ruptures.
RCS rapidly depressurized to 600 psig.
'
c.
!Ans. a. Reads low (0.25) - PZR water density is reduced so level is
higher than at lower temperature (0.5).
! b. Reads high (0.25) - less pressure from top indicates more
water weight than actual (0.5).
! c. Reads high (0.25) - reference leg boiling or outgassing
reduces pressure there, indicating more water weight than
actual (0.5).
,
!Ref. NNI A-3
,
! 11.2.01.080
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--n.e- -w -w-, ---- w e , - - ,,-.-- , , -ee--,wmanw,, ,----,,we .'n,-r w, , --,, e,m ----re- p-,-,a- y,- -
-. . . . . .. . _ . .
.
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(1.0) 3.5 Explain why the Mod Comp NAS software calculation for tilt may
not show any effect of a SPN0 failure.
!Ans.
!(l.0) The Mod Comp automatically substitutes another predetermined
SPN0 sieral for the failed detector, so the tilt calculation
4
source will appear norinal
i
!Ref. 11.2.01.296 A-1
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3.6 Answer the fo11 ewing questions about the Heat Sink Protection System
(HSPS):
(1.00) a. List the FOUR conditions that will cause EFW initiation, include
r setpoints where appropriate.
(0.50) b. List the TWO conditions that will cause MFW isolation, include
setpoints where appropriate.
(0.75) c. WHEN and WHY is TYPE II compensation used?
(0.75) d. What would be the effect on components and indications on a loss
of all power to Train A of the HSPS?
1(0.15)- a. Low OTSG }evel #
I(O. Mr)o.af w .4) < 10" { < /f
!(0.25) Loss of all RCPs
1(0.25) Loss of both MFW pumps
!(0.15) High R3 pres
!(0.10) > 4 psig
f(0.15) b. High GTSG level
!(0.10) > 94%
1(0.15) Low OTSG pressure
1(0.10) < 600 psig
I . ) c. FW as i e ,[F cf ,L
!(0.5p- More accurate indication of OT G level
!(0.15 each) d. One set of EFW valves per OTSG inop (EF-V-30A/C)
! Auto start of EF-P-2A W % _4 /
1 Auto isolation of one set of MFW valves,(F@ V-5A/92A/16B/17B)
! Lass of half of HSPS control room indication
! Ioss of ICS input to High & Low level limits
! Ref: TMI-l OPM Vol 3, Chpt F-10, pgs 2-3, 5, 9, & 11
1 K&A 059000A201/IF 3.4
1 059000K419/IF 3.2
1 061000A205/IF 3.1
1 061000K402/IF 4.5
21.0 1434R
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0
(2.0) 3.7 Some of the Reactor Core Safety Limits are based on two
parameters not directly observable. One to these is DNB. What
four (4) observable parameters are used by RPS to determine the
proximity to DN87
!Ans. Neutron Power
!(0.5 ea.) RCS Flow
! Temperature
! Pressure
!Ref. 11.2.01.132 B-3
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(0.5) 3.8 Define degree of redundancy. (Mathematical or word statement
acceptable)
.
.!Ans. (# Channels Operable) - (#C hannels Required for Trip' Signal)
! or
! Difference between the number of operable channels and the
number of channels which will cause an automatic trip actuation.
!Ref. 11.2.01.082 8-1
. ! 11.2.01.132
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.-
.
(1.75) 3.9 Answer the following questions concerning the Reactor Protection
system (RPS):
a. What four (4) reactor trips are bypassed when the RPS is
placed in " Shutdown Bypass?" (1.0)
b. During power operation the "C" RPS Channel is placed in
" Manual Bypass." What is the trip logic with RPS in the
above configuration? Explain. (0.75)
I
!Ans. (0.25) a. Power / Flow / Imbalance
!(0.25) Power / Pumps
!(0.25) Variable Press / Temp
!(0.25) Low Pressure
!(0.25) b. Trip Logic - 2 out of 3
!(0.5) Why - in manual bypass the "C" Channel will not trip so
any 2 of the remaining must trip to cause a reactor trip.
!Ref. 11.2.01.132 B-2
24.0 1434R
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(1.00) 3.10 Why does the reactor trip on a total loss of ICS/NNI HAND power when
the plant is at 100% power?
!(1.00) '-t icip-t ry trip a-
la-- af '^*h = 7" ,- -
I.
Y SU
Ref: TMI-l'1202-41, pgs 1-2
'
l 'INI-1 OPM Vol 3, Chpt F-2, pg 11
1 TMI-1 OPM Vol 3, Chpt F-3, pgs 101-102
1 .TMI-1 Training Handout 3210-86-0164 dated April 10, 1986
I K&A 016000K301/IF 3.4
,
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_ . . .
.
4
(1.5) 3.11 Define High Impedance Fault. Explain how a high impedance fault
might affect an electrical buss and the actions to be taken to
recover from this condition.
!Ans. (1.0) A kiah Impedance Fault is excess current (wire-to-wire) due to
fire damage to wire insulation. This current, while not enough
to trip the component breaker affected but can cause a bus over
current triD causing loss of needed equipment.
!(0.5) To recover f rom a bus trip all breakers should be ODened the bus
reenergized then close breakers on essential loads only.
!ReF. 11.2.01.262 B-3
.
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.(3.00) 3.12 Refer to Figures 2 - 4. The reactor.is at 100% power with all ICS
stations in automatic with the exception of. Loop A Feedwater demand.
Assuming no operator action, how will the below subsystems respond to
a reactor trip and what will be the final value of the indicated
parameter?
Subsystem Parameter
a. Integrated Master Header pressure
b. Loop A Feedwater Demand A OTSG level
c. Loop B Feedwater Demand B OTSG level'
1(0.30) a. On the reactor trip, the setpoint will shift to the 125 psig bias
1(0.30) Atmospheric dump valves will relieve initial pressure surge
!(0.40) Header pressure - 1010 psig
B7K $ Ygjag f*sJ,,,ygn4C,
1(0.30) b. High lezel li=it vill crere feeductor i clation
f(0.30) Low level limit will ecure E" te etert cg Wy pv A'
1(0.40) A OTSG level - 30" on Start-Up range
1(0.50) c. Low level limit will control Start-Up FW valve
1(0.50) B OTSG level - 30" on Start-Up range
1 Ref: TMI-1 OPM Vol 3, Chpt F-3, pgs 150-153, 208-209
! THI-1 OPM Vol 3, Chpt F-10, pg 3
1 K&A 059000K107/IF 3.2
1 061000A101/IF 3.9
I 059000A307/IF 3.4
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(2.0) 3.14 Give the controlling level setpoints for the EF-V-30's (EFW) and
under what condition each is used. Include the level setpoint
when the system is in standby.
- !Ans. (8 parts)
!(0.25 ea.) S/U 0" - EFW not required
! S/U 30" - Loss of MFW
! - 4# RB Pressure (RCP's on) .
i
~
,
! - OTSG's <10" (S/U Range)
! OP 50% - All RCP's off
)
!Ref. 611.2.01.311 , A-1
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(2.0) 3.15 a. What is the reaction utilized by the SPND (Self-Powered
Neutron Detector)? (1.0)
b. Why must we wait 10 minutes af ter a power change before
using SPN0 output for calibration? (1.0)
Ans. (1.0) a. Rh + gn 4 Rh *$ Pd
!(1.0) b. The half life of Rh104 is such that it takes
approximately 10 minutes to reach equilibrium.
!Ref. 11.2.01.050 A-3
End of Section III
30.0 1434R
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.
Procedures; Normal, Emergency and Radiological
(1.0) 4.1 The plant is operating at 55% reactor power with three (3) RCP's
running. The Diamond and Rx Demand stations are in manual. The
following indications are received:
1. In-limit indication on rod 2-2 and Group 2.
2. Reactor power dips to 49% on NI-6
-
What manual actions (s) is/are required based on the above
information?
!Aqs. Run the reactor back in hand (0.5) to 60 percent of the reactor
power (45%) allowed for the RCP combination (0.5).
!Ref. EP 1202-8 B-3
.
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(3.00) .4.2 In accordance with EP-1202-37 "Cooldown from outside the Control
Room" where would you go to transfer control'of the below components
to the " EMERGENCY" mode of operation? Match the letters to the
numbers:
a. Communications System 1. RSTSP-A
b. Diesel Generator 1B 2. RSTSP-B
c. EF-P-2B (Emergency Feedwater Pump) 3. RSTSP-C
d. MS-V-4A (OTSG Atmospheric Dump Valve) 4. Control Room
e.. MU-P-1B (Makeup Pump) 5. Locally at the
f. MU-V-3 (Letdown Line Isolation Valve) Breaker Panel
g. NR-P-1C (Nuclear Service River Water Pump) 6. None of the
h. Pressurizer Heaters Group 9 above
1. RC-V-2 (PORV Block Valve)
j. RR-V-1B (RB Emergency Cooling River Water
Pump Discharge Valve)
'l(0.30 each) a. '2 f. 1 ggry a [ /, 2, f ""
l . S .
a CC f )l-*- l
"J QMGf
I d. 1 1. 2
I e. 3 j. 2
1 Ref: THI-1 OPM Vol 3, Chpt F-9, pg 13
i THI-1 1202-37, pgs 3-4
! THI-1 Lesson Plan 11.2.01.262 Remote S/D Panel,
! Learning Objectives O & P
l K&A 000068A121/IF 3.9
3 000068K201/IF 3.9
.
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(2.0) 4.3 The plant is operating at 70% power with four RCP's running.
The following alarms and indications are then received: ;
1. Hi motor stand vibration alarm on the computer.
2. RCP 1A Bently Nevada System - Alarm lights
3. Total RCS flow at approximately 118 x 106 lbm/hr.
4. Loose Parts Monitoring system alarm.
5. Low motor current indicated on RCP 1 A.
,
a. Identify the event taking place.
b. What immediate manual actions are required for this event.
!Ans. (0.5) a. Pump Motor Separation - Dropped Impeller
!(0.5 ea.) b. Manual Actions
! Reduce Power
! Trip the Reactor
! Secure affected RCP
!Ref. 1203-16 8-3
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4.4 In accordance with TMI-1 Technical Specifications:
(0.75) a. What is the Minimum Temperature for Criticality?
(1.50) b. What action must be taken during power operations if the actual
temperature is less than the Minimum Temperature for Criticality?
110.75) a. 525 F (NO tolerance allowed)
<
l I.75) b. The reactor shall be cuberitic:1 S/u f b
f (&rP9} by en e-aunt equel te er greater +'-- the c21cu12ted reretifity
1 incertier d:: t: depressuri ticr..
I Ref: THI-1 TS pg 3-6
! K&A- 002020SG08/IF 3.5
,
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(3.5) 4.5 State the Immediate Actions required by ATP 1210-3 for Excessive Cooling.
(Including If/Then statements)
'n.
!Ans. (0.5) 1. IE OTSG level is greater than 94%; THEN verify HSPS MFW isc
!
!(0.5) 2. IE OTSG level is greater than 97%; THEN trip the MFW pumps.
3. IE OTSG pressure is less than THEN verify HSPS MFW isolation
!(0.5) has actuated.
! 600 psig;
!(1.0) 4. Isolate the af fected OTSG (both if affected generator cannot be
identified).
,
!
! FW-V-16 A FW-V-168
FW-V-17A FW-V-17 B NOTE
!
! FW-V-5A FW-V-5B ( .05 each valve missed)
! FW-V-92A FW-V-928
! MS-V-30 MS-V-3A
! MS-V-3E MS-V-3B
! MS-V-3F MS-V-3C
! MS=V-4A MS-V-4B
! MS-V-1A MS-V-1C
! MS-V-18 MS-V-10
5. IE OTSG level and pressure did THEN close the fcilowing valves
!(1.0) not stabilize; on the OTSG with the lower
!
pressure: (if no pressure
difference exists, than
isolate both if the leak is
in the intermediate bldg.)
!
EF-V-30A EF-V-308
!
EF-V-300 EF-V-30C
! MS-V-2B
MS-V-2A
!
!Ref. ATP 1210-3 8-1
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4.6 Answer the following questions concerning a fire in the cable !
spreading room:
(0.5) TRUE/ FALSE ,
a. Alarm from the smoke detectors in the cable spreading room
on Panel PRF 5-1 " Relay Room Fire" means the relay room
CO2 system has also actuated.
-(1.25) b. What immediate manual actions are required for a serious
fire in the cable spreading room per EP 1202-31?
.
!Ans. (0.5) a. False
1
!(0.2'; ea.) b. - Manually trip the unit
! - Manually actuate the CO2 system if not done
automatically
1 - Page announcement " Fire in the Relay Room (type) Fire
Brigade report to Control Building 3rd floor, Evacuate the
Relay Room"
! - Actuate the station fire alarm
! - Complete the follow-up action in the main body of this
procedure.
!Ref. EP 1202-31 8-1
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'4.7 Answer the following in accordance with 1203-16 " Reactor Coolant Pump
and Motor Malfunction":
(l'.00) a. What are the TWO indications of a failure of Number i seal?
L(1.00) b. What are the TWO indications of a failure of Number 2. seal?
(0.50) c. If Number 1 seal has failed, what Immediate Action must be taken
within five minutes?
1(0.50 each) a. High No. I seal leakoff flow (> 5 gpm)
! 'High'No. 1 seal outlet temperature (> 200F)
'
I b. High standpipe level (> 60")
l High vibration (> 0.002")-
I c. Close the No. 1 seal leakoff valve.
.
l' Refs THI-1 1203-16, pgs 2-3
-l K&A 003000A301/IF 3.3
. 1 003000K103/IF 3.3
4
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4.8 With the plant stable at Hot Shutdown Conditions, answer the
following ouestions concerning a " Total or Partial Loss of
ICS/NNI, HEX, HEY or Aux. Power" (EP 1202-43):
(1.0) a. What could cause the MS-V-3's to close if in AUTO on a loss
of HEX / HEY power?
(1.0) b. What are the Immediate Manual Actions reouf red for a loss
of HEX / HEY power in accordance with EP 1202-43 to deal with
the event identified in part a?
4
! Ans. (1.0) a. If affected OTSG pressure transmitter is selected for
control, indication will fail to O psig causing MS-V-3's to
go closed.
!(1.0) b. If turbine bypass valves have closed and are needed to
control turbine header pressure, take HAND control and open
to maintain set point.
,
!Ref. EP 1202-43 B-3
B-1
.
.
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4.9 A loss of ICS AUTO power has occurred followed by a reactor trip.
.
(1.5) a. What immediate action do you take to control feedwater flow?
(1.0) b. Does this loss of ICS AUTO power af fect the EF-V-30's? l
I
Explain.
!Ans. (0.5) a. -Immediately establish control of feedwater by closing the
main and startup feedwater control values.
!(0.5) -Reopen the startup control valves to maintain 30" level in
the OTSG.
!(0.5) -If control of feedwater cannot be established trip the
I main FW pumps and verify EFW actuates.
!(0.25) No
!(0.75) EF-V-30's receive power via the HSPS which is not powered
by ICS.
!Ref. EP 1202-42 B-2
! HSPS LP B-3
O
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l
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-- . -.-
e
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o
(2.5) 4.10 State the HPl throttling criteria including notes and cautions.
!Ans. High Pressure Iniection (HPI) Throttling Criteria
! Throttle HPI only if one or more of the following criteria are met:
!(0.5) 1. HPI must be throttled to prevent pump runout (550 gpm/ pump). ,
- -----
! -==- - - - - - - NOTE
!(0.25) l
Do not throttle to less than 500 gpm/ pump unless one of the l
! l following three criteria is met. __
l
_- - ____ _ _ _ - - - -
-
!(0.5) 2. HPI must be throttled to prevent violation of the
applicable brittle f racture/ Thermal shock curve limitations.
!(0.5) 3. HPI may be throttled if LPI flow is greater than 1000 gpm
in each line and stable for 20 minutes.
!(0.5) 4. HPI may be throttled if the required 25'F subcooling margin
exists.
!
CAUTION---- -- ==
!(0.25) l Open MU-V-36 and MU-V-37 when HPI is manually throttled below l
400 gpm/ pump. l
! l -- -
____________ _ ____________
g _____
!Ref. ATP 1210-10 8-1
40.0 1434R
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TilIS PAGE WAS INTENTIONALLY LEFT BLANK
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(1.0) '4.12 What are the company exposure / dose limits for .< hole body, skin,
and extremities?
i
-!Ans. Whole Body
!(0.3) 2.7 Rem /Qtr p 4JTw[e[/r y
!(0.3) 5 Rem /yr yggp
! Skin
- - !(0.2) 5 Rem /Qtr
! Extremities
!(0.2) 15 Rem /Qtr
!Ref. GET 202 A-1
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r
ob
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(1.0) 4.13 The below are two hazardous materials used at TMI-1. State two
(2) locations where each is stored.
a. Liquid Chlorine
b. Sulfuric Acid
!Ans. (0.25) a. RW Chlorinator Bldg.
!(0.25) CW Chlorinator 81dg.
!(0.25) b. CW Acid 81dg.
!(0.25) IWT Acid First Floor Turbine Bldg.
!Ref. PPC Plan A-1
,
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End of Section IV
End of Exam
43.0 1434R
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P 7.
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Attachment 1
WRITTEN EXAMINATION CERTIFICATION COVER SHEET
Ibsgr. (PLEA 5E PRINT)
~
-
EXAM
- (FIRST MID. - INITIAL LAST): DATE: 5 March 87
EMPLOTER (GwrANY): EMPLOYEE NO: 500. SEC. NUMBER:
EXAM EXAM LOCATION:
TITLE: Annual Requal Exam SRO
CATEGORY: ' I 118/119
I EXAM NO:
GENERAL INSTRUCTIONS AND GUIDELINES l SRO-t
- PLEASE READ THE FOLLOWINCs IN5TRUCTION5 CAREFULLT:
1. Remain seated and quiet during the examination.
2.- Please raise your hand when: you have any questions on the examination
you have finished the examination
3. - You are required to do your own work and you are not to help anyone else.
4. Use only the reference material authorized below.
5. If you must leave the room before you finish, the examination must be
returned to the proctor. Note that instructions #3 and #4 above still
apply while you are out of the room.
6. Misconduct or cheating on examinations will result in disciplinary action
on the part of the Company, and possibly' additional civil and/or criminal
sanctions.
7. At the conclusion of this examination, you' are to sign the following
certif1 cation. ~
CERTIFICATION
I certify that all answers contained in this examination are my own, that '
I have neither received nor given unauthorized assistance, and that I have
not used any unauthorized references.
'
SIGNATURE: DATE:
- D0 NOT BEGIN THE EXAMINATION BEFORE THE PROCTOR REVIDS THE REMAllWER OF THIS
PAGE WITH YOU.
AUTHORIZED REFERENCE MATERIALS: l TIME START STOP
Attachments l LIMIT TIME TIME I I OPEN BOOK
l 4.5 hrs xxI CLOSED BOOK
- PAfaE CHEGK THE EXAMINATION TO ENSURE YOUR COPY 15 C(MPLE FE.
- SPECIAL INSTRUCTIONS:
1. Use only black ink or pencil (#2 or softer).
! 2. Answer on exam pages.
? SECTION I POINTS SCORE I % SECTION l POINTS l SCORE I % i
!, Previous Exam i I I I I i
- V l VIII I 25 l
l l l l
l VI 25 i
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l l
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!
I ,
I i I
VII
,
?
l 25 l l l TOTAL I l l l ,
i MINIMUM AGGEPTABLE faRADES: EACH SECTION: 70.0 % OVERALL 80.5 ;
GRADED BY
- (EXAMINER'S SIGNATURE): DATE:
! Developed / Submitted: [ Date it.Me7
j Reviewed: m b Date/S # M /
Approved: DateM//
l 0412K y ~
i -
!
T i
!
--. _ . - - . . . _ . _ _ _ . . _ . _ . _ . . _ _ _ _ _ . . . , _ . _ . _ . . . _ _ . _ _ , _ . . _ _ _ . - . . . _ _ . __.-._,_,,_._,._,,..-_w
_ _ _ _ - . - -- ..
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SECTION VI: PLANT SYSTEMS: DESIGN, CONTROL
AND INSTRUMENTATION
'(1.50) 6.1 a. Why does the reactor trip on a total loss of ICS/NNI HAND power
when the plant is at 100% power?
b. Why does the reactor trip on a total loss of ICS/NNI HAND power
when the plant is at 5% power?
!(0.75) a. Mticiprtery trip en ' err ef 5:th ==' 7 ;;. D
A4sb Ras pressure. b p
f(0.75) b. RCS low pressure on enve-cooling (E" ;- ,
over
tert).
i Ref: TMI-1 OPM Vol 3, Chpt F-2, pg 11
! TMI-1 OPM Vol 3, Chpt F-3, pgs 101-102
1 THI-1 1202-41, pgs 1-2 ,
-
1 TMI-1 Training Handout 3210-86-0164 date April 10, 1p86 (
l K&A 016000K301/IF 3.6
.
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(3.00) 6.2 Refer to Figure 1 to answer the following questions, consider each
case separately:
(0.80) a. How will the system lineup change if the second Letdown Coole'.
is to be placed in service?
(0.80) b. Assume a spurious 30 psig Reactor Building ESAS signal is
received. What automatic action will occur in the ICCW system?
(0.60) c. Assume the 30 psig Reactor Building ESAS signal is concurrent
with an undervoltage on the ID 4160 volt bus. What automatic
actions will occur in the ICCW system?
(0.80) d. What TWO automatic actions would happen directly as a result of
ICCW flow decreasing to 500 gpm followed immediately by a loss
of the running Makeup pump?
f(0.40 each) a. Start the second ICCW pump
I and put the second ICCW cooler into service
.6
!(0.16*each) b. IC-V2/3/4/6 shut
f(0;jg) IC-V74 opens
f(0.30 each) c. Pump 1A will trip and be locked out
! Pump 1B will start
f(0.40 each) d. 1. Standby pump will auto start (<550 gpm ICCW flow)
I 2. RCPs will auto trip (<550 gpm ICCW flow & <22 gpm seal flow)
i Ref TMI-1 OPM Vol 1, Chpt B-10, pgs 4-8, & 12
1 TMI-1 OPM Vol 1, Chpt B-2, pg 25
l K&A 008000K102/IF 3.7
1 000026K302/IF 3.9
1 000015A210/IF 3.4
,
2.0 1306V
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a
(1.0) 6.3 Explain wny the mod comp NAS software calculation for tilt may not
snow any effect of a SPND failure.
l(1,0) Tne mod comp automatically suostitutes anotner predetemined SPND
signal for the failed detector so the tilt calculation source will
appear nomal.
! Ref: 11.2.01.296 B -2
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. - . _ . _ _ .
.
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(2.25) 6.4 During operation at power, will pressurizer level read nigner or
lower than actual (and explain wny) If:
a. Temperature compensation is lost.
b. do Cell connection to tant t3p ruptures,
c. RCS rapidly depressurizes to 600 psig.
l
! a. Reads low (.25) - PZR water density is reduced so level is
nigner than at lower temperat are (.5). ;
! o. Reads nign (.25) - less press:are from top indicates more
water weignt than actual (.5).
! c. Reads nigh (.25) - reference leg boiling or out gassing
reduces pressure there, indicar.ing more water weight tnan
actual (.5).
! Ref: 11.2.01.080 g B -3
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w = - +9-y- - -- ,- y,vr.- - - ,.,.-,- v--9.,, vr -
g ----, -,+-e.- --
r.-- - - -y--ww-.m------ew,-r m..g- wwv--ag y ---oy--e - - -*
. . . . . - .
.
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(1.0) 6.5 Indicate wnether eacn of tne following control rod oreaker trips
WOULD or WOULD NOT result in a reactor trip,
a. II, CC and E
o. 10 and iI
c. CC, 00, E and F
d. 10, Ccand E
l(.25 ea) a. would
! o. would
l C. would
! d. would not
l Ref: 11.2.01.132 SR0 A-3
t
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(2.0) 6.6 Give the conditions wnicn will initiate an auto start of tne
emergency feedwater pumps. (Include setpoints)
!(.5 ea) Loss of main feed water pumps - less tnan 50 psig across pumps
! Loss of all RC pumps
! Low leveT UTSG - 15" SU Ind.
! RB pressure - 4#
! Ref: 11.2.01.311, 11.2.01.028 A-1
)
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6.0 1306V
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(1.25) 6.7 True/ False and Explain
Pressing tne manual ESAS 30 psig actuation pusnouttons will start
tne Building Spray Pumps.
l(.5) False
!(.75) B.S. pumps must be started at their extension controls.
! or
! Tne 30# manual actuation PB only causes tne B.S. valves to
reposition.
! or
! B.S. pumps auto start on 2/3 R.B. pressure greater tnan 30#.
! Ref: LP 11.2.01.127 B.S.
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(1.0) 6.8 Describe now tne operation of the Miscellaneous Waste Evaporator
can give false indication of an OTSG tube leak.
! If the MWE nas a leak in the steam tube bundle contamination can
l
enter tne condensate return unit and be pumped to the main
f condenser. RM-A-5 would detect tnis contamination and alert the
operator to a possiole OT5G tube leak. (l.0)
! Ref: 11.2.01.164 SRO A-3
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8.0 1306V
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(1.0) 6.9 Tne new startup range level instrument used type 2 level
compensation. Describe tne corrections made to ootain an accurate
type 2 compensated OTSG level. (Neglect reference leg temperature
compensation).
l(.5) Develops Tsat based on OTSG pressure for water density (in OTSG).
!(.5) Corrects for steam volume (space) effects in tne OTSG (delta P
correction).
! Ref: 11.2.01.311 C-1
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(1.5) 6.10 Define nign impedence fault. Explain now a nign impedence fault
mignt affect an electrical Duss and tne actions to be taKen to
recover from this condition,
i
l(1,0) A nigh impedence fault is excess current (wire-to-wire) due to
fire damage to wire insulation. Tnis current, while not enougn tc
trip the component Dreater arrected but can cause a bus over
current trip causing loss of needed eouipment.
!(.5) To recover f rom a bus trip all breakers should be opened the bus
re-energized tnen close breakers on essential loads only. .
! Ref: 11.2.01.262 C-3
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(1.5) 6.11 Explain the function of eacn position on tne tnree-position "69"
I switches located on some of tne ES areaker cabinets. Include in
your explanation any safety functions wnicn are operational or
removed.
, !(6 parts, .25 each)
! Normal - Control in control Room
! -
All interlocks functional
! Bypass - Control locally and in Control Room
! - All interlocks functional
! Emergency - Control of eouipment at breaker only
! - Start interlocks removed (breaker interlocks remain
functional)
! Ref: 11.2.01.262 C -2
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(2.0) 6.12 Answer tne following ouestions concerning tne Reactor Protection
System (RPS):
(1.0) a. Wnat four (4) reactor trips are bypassed when the RPS is
placed in "Snutdown Bypass?"
(1.0) o. During power operation tne "C" RPS Channel is placed in
" Manual Bypass." Wnat is the trip logic witn RPS in tne
above configuration? Explain.
l(.25) a. Power / Flow /Imoalance
!(.25) Power / Pumps
!( 25) Variable Press / Temp
!(.25) Low Pressure
!(.5) o. Trip Logic - 2 out of 3
!(.5) Why - in manual bypass tne "C" Channel will not trip so any 2
of the remaining must trip to cause a reactor trip.
! Ref: 11.2.01.132 C-1, C-2
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(2.0) 6.13 List, in seouence, tne four (4) ma.ior automatic acticas that occur
tn the Reactor Building Emergency Cooling System following a 1600#
ES signal without loss of offsite power. Assure system initially
in a normal line up for power operations.
!(.5) 1. Tne AH-E-1 fans operating will trip on a clock one loading
signal.
l(.5) 2. All tnree (3) fans will start automatically and operate at
their slow speed on clock two loading signal.
- (.5) 3 Tne Emergency Cooling System will go into its emergency mode
of operation, opening tne necessary valves automatically and
starting the river water pumps (RR-PI A/B) to estaolisn flow
tnrough the_ emergency cooling coils.
l(.5) 4 Tne normal cooling coil penetration isolation valves (RB-V2
, and RB-V7) will close.
! Ref: ESAS C-2
! RB Dnergency Cooling
! 11.2.01.126, 11.2.01.029
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(/.7f)
M 6.14 One function of the Decay Heat Removal System is to provide suction
to the makeup pumps in a " Piggy-Back" alignment.
(1.00) a. What conditions during a LOCA woul make this alignment
necessary?
(0.75) b. Why are MU-V-14A/B left open in this alignment?
1(0.50 each) a. If BWST reaches lo-lo level alarm (36")
I before LPI flow is established
!(0.75) b. To protect the makeup pumps from a loss of suction pressure if
1 the LPI pumps were to trip.
i
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14.0 130tV
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(.75) 6.15 Wnat effect does transferring an ES component control to the
remote snutdown panel nave on tnat component's response to en ESAS
actuation?
! That component will not respond to the ESAS signal (ES signal is
blocked).
! Ref: 11.2.01.029 C-2
.
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(1.5) 6.16 a. How long would it take EF-P-2A to come up to full speed
following an ESAS actuation. Explain now you arrived at your
time. (1.0)
o. What is tne fai t position of EF-Y-30's on total loss of air
to tne valves. (.5)
!(.25) a. 30 see to full speed
!(.25) 15 see to receive start signal
!(.25) oloct 4 permissive then 5 sec delay
l(.25) and 10 see to come up to speed
!(.5) o. closed
! SRO A-3
!
End of Section VI
16.0 1306V
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SECTION VII: PROCEDURES; NORMAL, ABNORMAL, EMERGENCY
AND RADIOLOGICAL
7.1 A loss of ICS auto power nas occurred followed by a reactor trip.
(1.0) a. Wnat immediate action do you take to control feedwater flow?
(1.0) D. Does tnis loss of ICS auto power affect the EF-V-30's?
Explain.
! a. Immediately establisn control of feedwater oy closing the
main and startup feedwater control valves (.25). Re-open tne
startup control valves to maintain 30" level in the OTSG
(.25).
! - If control of feadwater cannot be establisned trip tne main
FW pumps and verify EFW actuates. (.5)
! o. No (.25) EF-V-30's receive power via tne HSPS wnicn is not
powered by ICS (.75).
! Ref: EP 1202-42 B -2
! HSPS LP B -3
.
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17.0 1306V
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, . - - - - - . ~ ,,.,4 e --- e ,. -,
_ _ _ _ _ _ _ _ . _ _
_ _ _ - _ _ - _ _ _ _ _
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(2.0) 7.2 The plant is operating at 70% power with four RCP's running. Tne
following alarms and indications are then received:
1. Hi motor stand vibration alarm on the computer.
2. RCP I A Bently Nevada Sys. - alarm lignts
3. Total RCS flow at approximately 118 x 106 lom/nr
4. Loose parts monitoring system alarm
Low motor current indicated on RCP I A
'
5.
a. Identify the event taking place.
D. What imediate manual actions are reouired for this event.
!(.5) a. Pump motor separation - dropped impeller
! D. Manual Actions
!(.5) Reduce power
l(.5) Trip tne reactor
!(.5) Secure affected RCP
! Ref: 1203-16 B-3
B -1
l
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18.0 1306V
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7.3 Answer the following ouestions dealing witn a reactor trip per ATP
1210-1:
(1. 5) a. Wnat steps are taken to deal witn ATWS event.
(1.0) n. At wnat point following a reactor trip would tripping tne
main FW pumps be reouired.
(1.25) c. Wnat steps are reoutred if suocooled margin drops cetow 25*F
following tne reactor trip.
!(.5) a. - Initiate HPI, maximize letdown
!(.5) - Trip 1G and IL buses ,
!(.5) - Maintain primary to secondary neat transfer
!(1.0) o. If MFW flow is still excessive after attempt to control witn
MFW regulating valves.
!(.25) c. Trip all RCP's
!(.25) Initiate HPI
!(.25) Initiate EFW
!(.25) , Raise OTSG level 90-95%
!(.25) Go to ATP 1210$ g' absent nigner priority symptoms
i
! Ref: ATP 1210-1 B -3, B -1, B -1
,
19.0 1306V
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7.4 Answer the following ouestions concerning an OTSG Tube
Leak / Rupture:
(.75) a. An OTSG tube leak has been identified wnile operating at
power. What actions are reouired for tnis per ATP 1210-5.
(Note: ATP step contains actions and reasons both are
reouired)
(.5) o. Under wnat conditions can the fuel pin in compression curves
ne violated for a shutdown with OTSG tube degregation.
(.75) c. Witn a VALIDATED OTSG tune leak of 2 gpm, is any E-Plan
declaration reouired? If so, wnat EAL.
(.25) a. - Reduce load at rate specified by SS
!(.25) - To minimize the risk of lifting MS safeties
!(.25) -
Close MU-V-3 as needed to maintain PZR 1evel
!(.5) o. OTSG tube rupture greater than or eaual to 50 gpm.
! c. Yes (.25) Unusual Event (.5)
Ref: ATP 1210-5, EPIP's B -2, B -1, B -2
.
20.0 1306Y
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(2.25) 7.5 Tne immediate manual actions of AP 1203-10 (Unanticipated
Criticality) nas actions to deal witn three (3) possible causes of
an unanticipated criticality.
a. What are tnose tnree possible causes,
b. Identify the reouired manual actions for each cause
identified in part a.
!(.25) a. Rod withdrawal
!(.25) RCS dilution
(.25) Cooldown (Overcoo11ng)
!(.5) n. Start insertion of control rods
(.5) Insure no flow into RCS or stop dilution in progress
!(.5) Onect TH and Tc and staoilize temperature if possible
! Ref: AP 1203-10 B-3
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21.0 1306V
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(3l00) 17 . 6 In accordance with EP-1202-37 "Cooldown from outside the Control
Room" where would you go to transfer control of the below components
to the " EMERGENCY" mode of operation?. Match the letters to the
numbecs:
a. . Communications System- 1. RSTSP-A
b. Diesel Generator 1B 2. RSTSP-B
c. EF-P-2B (Emergency Feedwater Pump) 3. RSTSP-C
d. MS-V-4A (CTSG Atmospheric Dump Valve) 4. Control Room
e. MU-P-1B (Makeup Pump) 5. Locally at the
f.
MU-V-3 (Letdown Line Isolation Valve) Breaker Panel
g. NR-P-1C (Nuclear Service River Water Pump) 6. None of the
-h. Pressurizer Heaters Group 9 above
i. RC-V-2 (PORV Block Valve)
j. RR-V-1B (RB Emergency Cooling River Water Pump Discharge Valve)
!(0.30 each) a. 2 f. 1
gg j gJ,
I
i
c.
d.
5
1
h. 5
i. 2
M
1 e. 3- j. 2
l Ref: TMI-1 OPM Vol 3, Chpt F-9, pg 13
I TMI-1 1202-37, pgs 3-4
1 TMI-1 Lesson Plan 11.2.01.262 Remote S/D Panel,
l. Learning Objectives O & P
I K&A 000068A121/IF 4.1
1 000068K201/IP 4.0
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7.7 Witn the plant stable at Hot Snutdown Conditions, answer tne
following ouestions concerning a " Total or Partial Loss of ICS/NNI
Hex, Hey or Aux Power" (EP 1202-43):
(.75) a. What could cause the MS-Y-3's to close if in AUTO on a loss
of Hex / Hey power?
(1.0) n. Wnat are tne insediate manual actions reouired for a loss of
Hex / Hey power in accordance witn EP 1202-43 to deal with tne
event identified in part a?
l(.75) a. If affected OTSG pressure transmitter is selected for control
indication will fail to O psig causing MS-V-3's to go closed.
l(1.0) o. If turbine bypass valves have closed and are needed to
control turbine neader pressure, take HAND control and open
4 to maintain setpoint.
! Ref: EP 1202-43 B -3
! B -1
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(.75) 7.8 On a " Loss of "A" DC Distribution System" wny does EG-Y-I A start?
!(.75) Tne solenoid valves in tne diesel air start s.Ystem de-energize on
loss of DC power.
! Ref: EP 1202-9A B -3 4
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7.9 Answer the following ouestions concerning a fire in the caole
spreading room:
True/ False
(.5) a. Alarm from the smote detectors in tne cable spreading room on
Panel PRF 5-1 " Relay Room Fire" means the relay room CO2
system has also actuated.
(1.25) o. What immediate manual actions are reouired for a serious fire
in the cable spreading per EP 1202-31.
l(.5) a. Fal se
!(.25 ea) o. - Manually trip the unit
! - Manually actuate the CO2 system if not done
automatically
! - Page announcement " Fire in the relay room (type) Fire
Brigade report to Control Building 3rd floor, evacuate
tne Relay Room"
! -
Actuate the station fire alarm
! - Complete the followup actions in the main oody of this
procedure
! Ref: 1202-31 B -1
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(0.75) 7.10 TRUE/ FALSE
In accordance with 9100-IMP-4200.03 " Rad Con / Chemistry Actions
Required when RMS Malfunctions": If the vacuum pump on the RMS
_
Monitor (example: Fuel Handling Building monitor, RM-A4) becomes
inoperable, the iodine, gas, and particulate channels shall'also be. ,
considered inoperable.
l(0.75) TRUE.
I Ref: TMI-1 9100-IMP-4200.03, pg 2
i K&A 072000SG08/IF 4.0
26.0 1306V
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(1.75) 7.11 State tne EFW tnrottling criteria.
!(.5) - To prevent RCS overcooling due to excessive feed rates,
manually control EFW flow as necessary to maintain OTSG
pressure to within 100 psig or desired pressure.
!(.25) - Monitor RCS Cold Leg Temperatures to insure tnat EFW flow is
not causing a significant RCS temperature decrease. ,
!(.5) - To insure adeouate EFW flow, verify decreasing incore T/C
temperatures.
!(.5) - If incore T/C temperatures are not decreasing, increase EFW
flow to at least 450 gpm (225 gpm per SG) until OTSG 1evel
setpoint is reached. If incore T/C's are decreasing, tne
i overcooling criteria takes priority.
! Ref: ATP 1210-10 B -1
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'(2.50) 7.12 Answer _the below questions in accordance with 1507-3 " Fuel Handling
Bridge Operating Instructions":
-(1.00) a. What is the reason fot' the administrative precaution "The bridges
shall not be de-energized when the fuel and control rod masts are
centered over the transfer tube axes."
(0.50) b. Why must you push the button on the end of the bridge control
handle prior to moving the bridge?
(1.00) c. The bridge operator is removing an Axial Power Shaping Rod (APSR)
from the core, but he has not placed the Function Selector Switch
to the Control Rod position. What is the consequence of this
action?
1 :l(1.00) a. Damage may occur due to movement of the bridge (the system
I interlocks are not operative when the bridge is de-energized).
!(0.50) .b. To alert personnel that the bridge is moving (warning bell)
l
1(1.00) c. It is possible to move the bridge with the APSR not fully
I withdrawn into the mast.
i i Ref: TMI-1 1507-3, pgs 6,-11, & 25
.I K&A 034000K601/IF 3.4
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! 034000K401/IF 3.0
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(.5) 7.I'5 a. Wnat is the limit placed on station disenarge maximum delta
temperature? (Detween inlet and disenarge)
(.25) O. Is tnis an administrative limit or a limit reouired by tne
NPDES permit?
l(.5) a. Max Delta-T 10*F
l(.25) o. Admin. limit
! Ref: NPDES Pemit A-1
! OP 1104-37
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31. 0 1306Y
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SECTION VIII: ADMINISTRATIVE PROCEDURES, CONDITIONS
AND LIMITATIONS
(1.0) 8.1 Differentiate between a channel test and a channel check as
defined by tne TMI-I Technical Specifications.
' A channel check only involves verifying acceptable performance (Dy
comparison witn similar devices, for example) (.5) a cnannel test
involves putting an artificial signal into the channel to verify
alarms, trips, etc. (.5).
! Ref: D4I-I TS definitions
! Code: VI II JB -3 -1. 0
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(2.00) 8.2- The plant is in Hot Standby when the only heat sensor in the 1A
Diesel Generator room is found hanging by its wires. In accordance
with the TMI-1 Technical Specifications, what actions, if any, should
be taken by the Shift Supervisor?
' *(1.00) 0 :lere the fire ryrter ineperrble
'
l(er50)(r.co) Establish a fire watch
!(0.50 kt.ov) within one hour
i Ref: TMI-1 TS, pgs 3-86 & 3-87
I LLER 86-012-00 dated October 6, 1986 -
! K&A 086000SG25/IF 4.0
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(1.5) 8.3 The ' A' Diesel Generator fails a surveillance. DC-P-1B is out of
service for an oil cnange. Tne plant is at 97.5% power. Is there
any ouestion aoout tne operanility of notn LPI trains? Wny or wny
not?
!(.5) Yes
! The emergency power supply for tne ' A' train is 00S (.5) and tne
cooling water supply for the 'B' train is 00S (.5).
! Evaluate alternate responses.
! Ref: THI-I TS 1.3, 3. 7. 2.c, 3.3. l .4.c
' Code: VIII-B -3-1. 5
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(3.00) 8.4 Using the Emergency Plan Classification procedure provided, classify
each of the following sequence of events. ,EEnsider each sequence
separately. Include in your JUSTIFICATION the specific paragraph-
cited.
The plant is at 75%, EOC
The below equipment is out-of-service for each of the events:
2A Emergency Feedwater Pump
1B Nuclear Service Water Pump
1B Makeup & Purification Pump
.
a. Hot leg leak of 20 gpm - identified and confirmed
Loss of Bus ID - cannot be reenergized
Update on Hot leg break - NOW 250 gpm
Loss of offsite A/C - Reactor trip on power to flow, ES actuation
on Low RCS pressure
1B DHR Pump will not start
b. NI-7 (PR) fuses blow - replacement fuses also blow
Rapid load decrease to 50% ordered and completed due to problems
on the grid
B OTSG Main FW valve fails closed - Reactor trips on high
pressure
Report of explosions in the area of the Emergency Feedwater Pumps
c. lA Nu Q g r Service Water pump trips
RM-A-S,"Feading 2E5 cpm
RM-L-1 reading IE4 cpm
Maxeup tank level decreasing after a second MU pump started
Steam line rupture outside RB - ES actuation on low pressure,
automatic Reactor trip does not occur but manual trip functions
One control rod does not insert on the trip
Main Steam Isolation valve MS-V-1A does not close, fully
ANSWERS ON PAGE 35.0 A
35.0 1306V
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' ANSWERS TO QUESTION NO. 8.4 .
- !(0.50) a. Site Area Emergency
-l(0.50) 'RCS leakage > makeup capacity - only IC MU pump avail
! .
(para 3.5.1.A)
1 OR Failure,of.any ECCS to start and.run - Bus ID was de-energized
I and 1B DHR pump would not start (para 3.7.2)
.!(0.50)- b. General Emergency.
.l(0.50) Loss of physical security control of the vital. area - explosions
.I _ 9)" in the area of the EFW pumps (para 3.3.1)
{!(0.50) c. General Emergency
,1(0.50)~ Potential for release of large amount of radioactivity - loss of
all three barriers (para 3.4.1)
i .I Ref: TMI-1.6410-IMP-1300.1,-Enclosure 2
! K&A 194001A116/IF 4.4
$Ne Akn fmgency pc 6 %k. 3. c.2 s
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(1.0) 8,5 - State the RCS activity limits (2 limits).
!(.5) Less than or eoual .to 1.0 microcurie / gram dose eouiv.1-131
- (.5) And less than or eoual to IOO/E microcurie / gram
! Ref: TMI-1 T. S. 3.1.4.1
! Code: VIII-B-1-1.0
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(1.0) 8.6 State tne two OTSG leakage levels wnicn reouire plant shutdown per
Tecn. Specs.
! I gpm (.34)
0.1 gpm (.33) aoove baseline (.33)
t e t .o gy k
! Ref: TMI-1 T.S. 3.1.6. 3
! License p.6b Para 8.2
- Code: VIII -B -1 -1. 0
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(1. 5) 8.7 TS 3.3.1.3a reouires, in part, tnat two RB spray systems and two
RB emergency cooling fans and emergency cooling units De operable
prior to criticality. If, during power operation, it is
discovered that botn spray systems and one of the three cooling
fans are inoperaDie, does sufficient capacity remain to supply
emergency building cooling? Explain.
!(.5) No
!(i.0) RB emergency cooling reouires 3 coolers in the aosence of RB spray.
! Ref: TMI-1 TS 3.3.1.3a and bases
! Code: VI II-B -4 -0. 5
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(.5) - 8.8 Under wnat conditions do Tecn Specs permit snutting down all
active means of decay neat removal? (Specific numoers not
l
reouired) ,-
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l(.5) 1.ow decay heat generation
I Ref: TS 3.4.2.1!; . CW YN FW<
! Code: VIII-B-1-0.5
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(.5) 8.9 True/Faise
A spent fuel . snipping cask is not permitted in TNI-1.fuei nandling
building.
! Fal se
! Ref: T.S. 3.11
! Code: VIII-8-1-0.5
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(1.0) 8.10 With tne plant operating at power, one of the two licensed CR0's
must leave tne site immediately. There are no add 1tional licensed
personnel on site. How long can tnis situation exist without
violating Tech Specs? Does this rule apply to tne Snift
Supervisor positions?
l(.5) 2 nours
.
!(.5) No
! Ref: TMI-I TS 6.2.2
! Code: VIII -8 -1 -1.0
41.0 1306V
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(2.0) 8.11 A control md is inoperable if: (4 items)
q al o.f)
Id!(.A) Cannot be exercised
! M) Cannot be located
!M) Does not meet flight time
!W Assynetric or misaligned b
! .CWry erly spr?frmw/y greater than 9 incnes
! Ref: TMI-l TS 4.7 bases
! Code: VIII-8 -1 -2.0
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(1.25) 8.12 True/ False and Explain
-Per Aministrative Procedure 1067, Independent Verification
Program, it is permissible to do an independent verification of a
manual valve position visually.
!(.5) True
! Can use tnis option if second party can verify that operation by
first party (.5) was sufficient to verify proper position (.25).
! Ref: AP 1067, Rev. 0 4.3.b.2
! Code: VIII -A-4-0. 5
! 1-0.75
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(.5) 6.13 A surveillance test is performed early. Subseouent performance l
(late /early) dates will ne biased toward the seneduled date.
Pict one.
!(.5) Late
l Ref: AP 100lJ, Rev. 6 3.1.4a
l Code: VIII -A-3 -0. 5
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(1.0) 8.14 Under wnat conditions may a surveillance procedare data step oe
skipped? (2 items)
!(.5) If specifically excluded by individual procedure
!(.5) If not, must ne identified by a test exception
! Ref: AP 1001J, Rev. 6 3.2.6
!. Code: VIII-A-1 -1.0
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45.0 1306V
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.(1.0) 8.15 Identify two systems below wnicn are eliginie for " personal tags"
per AP 1002.
,
a. Auxiliary Building Aircraft Door
'
b. Welding Circuits
c. Weatner station
d. Domestic water
e. Sewage Treatment Plant-
l(any two, .5 eacn)-
! D, C, e
l Ref: AP 1002, Rev. 45 Encl. 4
! Code: VIII -A-3 -1.0
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8.16
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(.5) 8.17 How are temporary openings to primary systems tracked per AP 1030?
l(.5) Logged in SF logbook.
! Ref: AP 1030, Rev.10 Step 2.3.3~
! Code: VIII-A-I-0.5
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(1.0) 8.18 A licensed training crew CR0 not on duty is present in the Control
Room when tne turnine trips. Tne duty crew responds to tne trip,
out fails to notice one of the IF's in 1210-I. Tne training crew
CR0 does notice it, and takes appropriate action, and informs the
duty crew of his action. Is this permitted? Explain.
!(.5) Yes
!(.5) AP 1029 permits cualified operators to take action in emergencies.
! Ref: AP 1029, Rev. 24 Step 5.2.1
! Code: VIII-A-1 -0. 5
! 4 -0. 5
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(.5) 8.19 Wnat plant condition change reouires a change in snift manning?
. !(.5) RCS temperature at 200*F
! or difference between cold shutdown and heatup/ cool'down.
! Ref: AP 1029, Rev. 24 Step 5.7.a
! Code: VIII -A-1 -0. 5
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(1.0) 8.21 It is discovered that TMI-1 was operating in an unanalyzed
condi tion. Wnat is the difference in reporting tnis item to NRC
if the plant is operating or shutdown?
!(.5) If found while operating,1 nour reporting time.
!(.5) If found wnile shutdown, 4 nour reporting time.
! Ref: AP 1044, Rev. 16, Encl. I
! Code: VIII-A-I-1.0
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(1,0) 8.22 Define " Contaminated Area" for TMI-1. (2 items)
!(.5) 1000 dpm/100 m2 3-g ,
!(.5) 20 dpe/100 cm a-
! Ref: 9100-ADM-4110.01, Rev. 3 Step 3.5
! Code: VIII -A-1 -1. 0
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(1.25) 8.23 Wnat radiation level reouires the posting of an area as " Locked (
'
High Radiation Area" and wnat two places can the keys be found for
these areas?
! If someone could get a dose grrater than 1000 mrem in I hour
(.75) Keys in locked nign rad area key locker (.25) and in '
'
Control Room key locker (.25).
! Ref: 9100-ADM-4110.06, Rev. 5, Steps 4.1, 4.2, 4.6
! Code: VIII -A-1 -1. 25
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54.0 1306V
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(1.0) 8.24 During a declared emergency, tne ED/ESD (may/must) comply witn NRC
advice, and (may/must) comply witn NRC directives. Circle the
appropriate response.
!(.5) may comply, w/ advice
!(.5) must comply, w/ directives
' Ref: 6410-IMP-1300.02, Rev.1, Encl. 5
- Code: VIII-A-3-1.0
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End of Section VIII
End of Exam
55.0 1306V
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ATTACHMENT 5
NRC Resolutions to Facility Comments on NRC Questions on March 5,1987
Requalification Examination
Question NRC Resolution
2.8.b Point values redistributed.
l
3.1 Will consider during grading on case basis. '
3.6.a Both setpoints will be required for full
credit.
3.6.c Agree that a more specific answer is warran-
ted. Proposed answer added to NRC answer and
required for full credit.
3.6.d Proposed answer will not be accepted for full
credit. Question requires differentiation of
affected and non-affected components.
3.10.a Accepted. NRC answer was a best guess as to
which trip would occur first.
3.12 Accepted based on given reference.
4.2 Answer modified to allow credit for a response
which indicates the candidate knows which coni-
ponents are transferred at one of the RSTSP's
and which are transferred locally. Based on
4.4.b Delete the portion of the answer which specif-
ies the required minimum shutdown margin and
substitute "The reactor shall be shutdown"
which is defined as having a shutdown margin
greater than the above minimum.
6.1 Some resolution as for 3.10.a based on given
reference.
~
6.2 Some resolution as for 2.8.b.
7.6 Some resolution as for 4.2.
________________a
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Attachment 5
,
- Question NRC Resolution
8.2 Delete the words " declare the ' fire system
inoperable" since -.taking action to establish
a fire watch presumes the local fire- system
is inoperable.
- 8.4.b Accepted based on given reference.
8.4.c Not accepted. Emergency Plan gives sufficient -
basis for the stated conditions to be class-
ified'as a General Emergency.
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