ML20212P655
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| ML20212P655 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/09/1987 |
| From: | Collins S, Dudley N, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20212P619 | List: |
| References | |
| 50-029-86-20, 50-29-86-20, NUDOCS 8703160183 | |
| Download: ML20212P655 (68) | |
See also: IR 05000029/1986020
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U. S. NUCLEAR REGULATORY COMMISSION REGION I
REQUALIFICATION PROGRAM EVALUATION REPORT
EVALUATION REPORT NO. 86-20 (OL)
FACILITY DOCKET NO. 50-29
FACILITY LICENSE NO. DPR-3
LICENSEE: Yankee Atomic Electric Company
1671 Worcester Road
Framingham, Massachusetts 01701 (}
FACILITY: Yankee Nuclear Power Station
EXAMINATION DATES: November 5-7, 1986
CHIEF EXAMINER: 8&dd[ w&
Noel F. Dudley, Lead R6 actor Engineer
2-7-57
Date
REVIEWED BY: II[)
Robert M. Keller, Chief, Projects
2/7/I/
Date
Section 1C
APPROVED BY: MYml11[jf]]//1/yyj 1 9/pi7
1ariiuel J. C6111ns, Deputy Director, Dat'e
Division of Reactor Projects
SUMMARY: The administration of the facility's annual requalification examina-
tions was audited by the NRC. The effectiveness of the training department in
implementing the requirements of the requalification training program was
evaluated as marginal. As a part of our evaluation the NRC reviewed and
modified the facility prepared written examinations. The modified written
examinations and the facility administered oral examinations were utilized to
measure the effectiveness of the requalification program in identifying
individual and generic operator weaknesses. One of the five operators failed
the written examination. Problems were also identified in the timely training
of operators on plant modifications and procedural changes, and in addressing
individual weaknesses identified by facility evaluations.
8703160183
PDR 870303
ADOCK 05000029
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DETAIL
1. EXAMINATION RESULTS:
R0 SR0 TOTAL EVALUATION
Pass / Fail Pass / Fail Pass / Fail
Written Examination 1/0 3/1 4/1 Marginal
Oral Examination 1/0 4/0 5/0 Satisfactory
Evaluation of Facility Written Examination Grading: Marginal
Overall Evaluation of Program Implementation Effectiveness: Marginal
2. Scope
The facility prepared R0 and SR0 written examinations were reviewed by
the NRC and approximately 13% of each examination was replaced with NRC
prepared questions. The administration of the written and the oral
examinations was audited, and the results of past requalification
examinations were reviewed. Parallel grading of the written examinations
by the facility was conducted by the NRC and the results are presented in
Attachment 1.
3. Review of the Written Examinations
The annual requalification written examinations are utilized to evaluate
the knowledge level of operators and identify areas where retraining is
needed. These exams are prepared by a contractor and are reviewed and
approved by the facility training department prior to being administered.
A review of the facility approved written examinations was conducted by
the NRC using the general guidance for the scope of written examinations
contained in NUREG-1021, Chapter ES-202, Section E, the guidance concern-
ing depth of knowledge contained in NUREG-1021, Chapter ES-203, Section F
and examiner judgement. The following findings were considered to be
deficiencies in the examination.
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Section 1 was too heavily weighted towards thermodynamics and fluid
flow. Ffve of the twelve questions examined in this area and
comprised 47% of the point value of the section.
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Section 2 did not provide sufficient coverage of primary systems and
components. Only one question in this area was asked.
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Section 2 did not examine for a sufficient depth of knowledge. Six
of the ten questions required only recall of facts and comprised 74%
of the point value of the section.
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Section 3 questions 3.1(a) and 3.8(b) were duplicate questions.
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Section 3 did r.ot provide a broad coverage of the facility
instrumentation and control.. systems. Fifty percent of the questions
examined on the Reactor Protection System and about 20% of the
questions examined on Pressurizer Level indication.
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Section 3 did not examine for a sufficient depth of knowledge. Five
of the nine questions required only recall of facts and comprised
over 60% of the point value of the section.
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Section 4 did not provide sufficient coverage of normal and abnormal
procedures for primary and secondary systems. Only two questions
were asked in these areas. One concerned an automatic electrical
bus transfer and the other concerned a reactor start up.
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Section 4 did not examine for a sufficient depth of knowledge.
Eight of the ten questions required only recall of facts and
comprised about 75% of the point value of the section.
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The R0 examination did not provide a broad coverage of systems. The
following major systems were not examined: Pressurizer Level and
Pressure Control, Reactor Coolant Pumps, Shutdown Cooling System,
Control Rod Drive System, Control Rod Indication System, Condensate
System, Main Feedwater System, Main Steam Systems, Auxiliary Feed-
water System, Emergency Diesel Generators, Fire Protection Systems,
Service Water System, Refueling Systems, and Primary Containment
System.
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Section 6 did not provide sufficient coverage of the instrumentation
and control functions of the facility. Only two questions were
asked in this area. One concerned the pressurizer level detector
and the other concerned the listing of scram signals.
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Section 6 did not examine for a sufficient depth of knowledge. Six-
of the ten questions required only recall of facts and comprised
over 60% of the point value of the section.
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Section 7 included questions which should have been assigned to
l Section 8. Three questions were asked concerning administrative I
procedures which comprised over 20% of the point value of the
I
section. As a result, the coverage of operating procedures was
reduced.
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Section 7 did not examine for a sufficient depth of knowledge.
Eight of the ten questions required only recall of facts and
comprised about 75% of the point value of the section.
! -
Section 7 did not provide sufficient coverage of normal and abnormal
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procedures for primary and secondary systems. Only two questions
l were asked in these areas. One concerned an automatic electrical bus
transfer and the other concerned a reactor start up.
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In Section 8, question 8.9 appears to provide the answer to question
8.2.
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Section 8 was too heavily weighted towards the Emergency Plan.
Three of the eleven questions, examined in this area and comprised
about 28% of the point value of the section.
Sections 6 and 7 were duplications of sections 2, 3 and 4 of the R0
examination. Duplication of examinations is acceptable for requalifica-
tion. examinations if information is being examined at the SR0 level. Due
to the percentage of recall type questions in Section 6 and 7, these
sections were found to be inadequate for the training department to
properly evaluate the knowledge level of the senior operators and to
identify areas of weakness.
Based on the independent review and evaius; ion the NRC replaced 13% of the
questions in the written examinations. Tae resulting examinations were
considered adequate to identify iniividual and generic v.>aknesses. These
modified examinations are presented ,': Attachments 3 and 4.
4. Audit of the Oral Examinations
Oral examinations, which were administered by a contract examiner,
covered equipment changes during the last outage, material presented
during the requalification training cycle, and plant transic.ts.
The examinations were determined to be adequate to identify individual and
generic weaknesses. A wide variety of areas were covered during the
examinations, however there was sufficient overlap of subject areas
between candidates to be able to draw conclusions about generic strengths
and weaknesses.
Through observations and discussion with the contract examiner, it was
determined that one operator was weak in thermo-dynamics and that most
operators were weak in the area of system modifications and changes made
during the last outage.
5. Audit of Requalification Records
A review was made of previous 1986 annual requalification written and
oral examination results. The written examination grades ranged from 80%
to 90%. The oral examination evaluation sheets identified individual
weaknesses which included system knowledge, reactor theory, thermo-
dynamics, and the Emergency Plan. Several oral evaluation sheets, which
were completed in March 1986, identified a "need to get operators to read
the Pre-Startup Training Manual."
The developement of the 1986 requalification schedule did not include a
systematic review of the results of the 1985 requalification examinations.
There were.no facility records to support the information selected for
inclusion in the 1986 requalification schedules. However, the Training
Manager stated he could easily reconstruct the basis for selection of the
information included in the schedule.
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No action has been taken by the training department to provide feedback
or remedial training for the ir.dividual weaknesses identified during the
oral examinations. No action was taken by the training department on the
generic weakness concerning the information contained in the Pre-Startup
Training Manual which was identified in March. The Training Manager felt
that the information was adequately covered during the requalification
training. However, from the oral examinations that were audited in
November 1986, continued weaknesses in this area were identified after
the operators had completed the requalification training.
The requalification training program requires an operator to attend five
consecutive weeks of training once a year and to take the annual
requalification examination at the end of the training. As a result of
this schedule an operator must wait for over a year to receive training
in an area of identified operator weakness or to receive training on
recent plant and procedural changes. As an example, some recent operating
problems such as inadvertent actuation of the CARD 0X system and problems
associated with the operations of a new air compressor may not be
presented to operators for over a year.
The requirement in 10 CFR 55 Appendix A that requires requalification
training to be continuous is met by the licensee's present program. How-
ever, t5e requalification training program has no systematic method for a
systematic review of the results of operator evaluations for identifica-
tion of weaknesses'. As a result, the program does not provide for prompt
feedback to operators nor does it provide a mechanism for disseminating
necessary operational information in a timely manner to all potentially
effected operators.
6. Exit Interview:
NRC Attendees:
N. Dudley, Lead Reactor Engineer
H. Eichenholz, Senior Resident Inspector
Facility Attendees:
N. N. St. Laurent, Plant Superintendent
B. Drawbridge, Assistant Plant Superintendent
E. Chatfield, Training Manager
K. Jurentkuff, Jr. , Plant Assistant Operations Manager
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O. Nelson, PQS Corporation
Comments
The NRC examiner presented the following observations based on the onsite
evaluation: Although the written and oral requalification examinations
were adequate for identifying areas of individual and generic weaknesses,
no formal or systematic review of examination results was conducted prior
to establishing the 1986 requalification schedule. There has been no
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follow up to either individual or generic weaknesses identified by the
annual examinations. The facility provided at least ten 'cerrections to
the answer key after administration of the examination.
The facility stated that they had no plans to change the requalification
program schedule and that some other utilities were considering adoption
of a single training session per year.
7. Parallel Grading
Following completion of the onsite review parallel grading of the written
examination was conducted by the examiner. The results of the parallel
grading of the examinations are presented in Attachment 1. One senior
operator failed the written examination as graded by both the licensee and
the NRC. He has been taken off shift by the licensee and placed in an
accelerated retraining program which includes review of the examination
with an instructor and directed self study.
A second senior operator passed the facility graded written examination
and was evaluated as a weak pass on his oral examination. The contract
examiner who administered the oral examination recommended additional
upgrading prior to allowing this operator to assume licensed duties. The
NRC evaluated this operator's written examination as a failure. Since
this operator is not assigned to licensed duties, and would be required to
submit a certification to the NRC prior to resuming licensed duties, in
accordance with 10 CFR 55.31(e), no action is being taken to resolve the
pass / fail discrepancy between the results of the facility and NRC grading
of the written examination. No actions have been taken or are planned by
the facility to upgrade this operator's knowledge level.
All other operators passed all portions of the annual examinations. All
facility final written examination grades were within 4% of the grades
determined by the NRC. However, numerous changes were made to the
examination answer key as a result of the grading process. Corrections to
the answer key, provided by the facility, two days after the administra-
tion of the written examination, are presented in Attachment 2. The
details of all the changes made to the SR0 examination answer key are
presented below to demonstrate the number and types of corrections which
were required to be made to the facility prepared answer key.
Question Change Reason for Change
Number
5.05(b) Add " counts double Provides partial credit
means excess reactivity if no calculations are
has been reduced by h". performed.
Weight the statement
0.25 points.
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5.06 Weight MTC and FTC 0.4 Clarifies point
points each. Weight void distribution.
coefficient 0.2 points.
5.07(a) Curves for Tave and Steam This graph should be the
Pressure are reversed, answer to question 1.12.
Power change should be Properly labels diagram.
from 80% to 85%.
5.08(a)+(b) Reverse labels on flux Due to the fuel loading
diagrams. Change part (b) on the present cycle a
to " The flux in the core non-standard radial flux
becomes more evenly has been created.
distributed because of
burnout and shifts to the
core edges where new fuel
is located".
5.1. a Add " hydraulic horse- Provides definition of
power = horsepower out hydraulic horsepower.
(energy added by pump)".
6.01 Add "EDG and BT's open Adds additional abnormal
only if low voltage condition not included
conditions exist". in answer key.
6.04 Add "or any other Training department has
reasonable answer". not delineated what
other answers would be
acceptable nor how
partial credit should
be assigned.
6.05 a Add "or using HX bypass Provides other method of
valve". controlling temperature
even though method is not
explicitly addressed in
the procedure.
6.06(b) Weight EDG as 0.75 points. Clarifies point
Weight 125 VDC station distribution.
batteries as 0.25 points.
6.07 (c) Change to " decrease". Corrects incorrect
answer.
6.09 Expands answer key with Corrects incomplete and
additional paragraph. misleading answer.
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7.04 a Weight " Dose rate at Clarifies point
barrier = 100 mR/hr" as distribution. Corrects
0.5 points. Change calculation from line
calculation to "R,(d,)2 = source to point source
R2 (d
2 )2". method.
7.06(c) Change to "If system Clarified NRC provided l
pressure exceeded seal answer,
tank pressure".
7.9(a)+(b) Change answer to " prevent Corrects NRC provided
uplift from moving fuel answer.
internals".
8.3 (a) Accept answer of "50%". Training department
determined that 50% is
close enough to the
correct answer of 53% to
64% to be acceptable for
full credit.
The number and types of corrections made to the answer key by the
facility during grading of the examination is irdicative of a poor
technical and administrative review prior to administration of the
examination.
As a result of grading the examinations, some knowledge weaknesses were
identified in the understanding of recent procedural changes. Examples
of the procedures changes include: the Main Coolant Pump restrictions
added to the operating precautions and limitations as a result of
recommendations made in a letter dated November, 1985; revision to the
Steam Generator Tube Rupture procedure which was made in December 1985;
and the revision to the shutdown cooling procedure which was revised in
May, 1986.
8. Conclusion
The effectiveness of the training department in implementing the
requirements of the requalification training program is marginal. Due to
the past operating history of the facility and small sample size utilized
during this evaluation no immediate corrective actions are required to
upgrade or evaluate the knowledge level of all licensed operators.
The facility prepared written examination was not adequately reviewed by
the training department and resulted in an examination which did not
provide broad based coverage and did not test for a sufficient depth of
knowledge. The technical review of the answer key, by the facility prior
to the administration of the examination, was inadequate.
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Facility written and oral evaluations adequately identified individu'al
weaknesses. However, the -training department did not take aggt'essive
steps to correct identified weaknesses. Ay a result, some licensed
operators are not knowledgeable about some plaqd and procedural changes
that occurred over a year ago. '
Attachment 1: Requalification Results Summary Sheet x. ic'
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Attachment 2: Facility Additional Comments ,
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Attachment 3: R0 Requalification Examination and; Answer Key
Attachment 4: SRO Requalification Examination and Answer Key , -/'
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NRC'ADMif4lSTERED REGU ALIFICAT!ON EXAMINATION RESULTS SUMMARY ' ATTACHMENT 1
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/ CYCL,E VII LICENSE REQUALIFICATION
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Additional Comu}nts ,p s
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}J[J 1.5 The Answer Xey is in error. . Procedure,OP-2103, Step 8 f, page 5,
6 has the operator estimating CRP using the "two most recent plotted
points." This would give a rod position closer,to 85"..
(See Attachment'l),
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. Jg 2.8 Answer' Key is in jrror. Class II liquid'contains dissolved
oxygenated air not dissolved hydrogenated, air.
,
(Ref. STM, p. 24-4) t
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5 3.3 Answer Key is not. complete. l Low steam generator level should be
'? included * (Ref. STM, p. 33-48)
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3.7;and 6.9 The Answer Key is not complete and-is misleading. ,See attached
is S revised answer. O
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( (Ref. STM, p. 2-17)
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4.3 and 7.3 l Answer Key to part b is incomplete. See Procedure OP-3107, p. 3.
(See Attachment 3)
3.9 The Answer Key is incomplete. There are push buttens on the'b'reaker
y and turbine / generator trips.
(Ref. STM, p. 33-48)
5.6 The point distribution on the Answer Key is not clear. It should'shows
> - moderator temperature coefficient .4 4
- fuel temperature coefficient .4
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~ void coefficient .2
J i The Answer Key is incorrect.
5.8
? Part As
1 -1
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l Part B: The flux in the core becomes more evenly distributed
~! E because of burnout and shift to the core edges where -
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. new fuel is located. . ! l
'4 (Ref. Attachment 4, Cycle 17 flux maps
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- Fund, of Nuc. Reactor Physics,
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Westinghouse, p. 4-48
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- Reactor-Physics Manual,
p. 3-112 to 3-114-(Attachment 6))
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,. 5.10 Answer Key is incetaplete. Hydraulic horsepower is defined in
,d, Westinghouse Thermal Book as fluid horsepower. See appendixI A for "'i definition and section 10 for write-up. ,
(Ref. Westinghouse, Thermal Hydraulic
'
Principles)
6.7 Answer Key is incorrect. ' Answer C should be decrease.
f* (Ref. Core XVIII, Startup Manual,
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p. 109 (Attachment 5))
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PQS CORPORATION
RO AUDIT EXAMINATION
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Facility: YAbKEE ROUE
Reactor Type: LESTINGHOUSE - PtJR
Date Administered:
ANSlJER KEY
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P Q EEs CORPORATION
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RO A LID I T EXAMINATION
Facility ' YMJ<EE ROLE
Reactor Type: UESTINGHOUSE - PUR
Date Administered .
Examiner:
Candidate:
LNSTRUCT[0NS TO CNO!DATE:
.You may use calculators, Steam Tables and Data Book. Use separate paper
for the answers. Write answers on one side anh, leaving plenty of room
between answers. Separate answer sheets by Catesory and staple each
separately. Be sure your name is on each section's answers. Each question-is
worth one (1) point unless otherwise noted. The passing grade requires at
least 707. In each category and a final grade of at least 80%. Examination
papers will be picked up six (6) hours after the examination starts.
Category 7. o f Candidates's 7. o f
Value T_q t_a_I_ Score Cat. Value Category
19.0 1. Principles of Nuclear Power
Plant Operation, Thermo-
dynamics, Heat Transfer and
Fluid Flow
_192 0
2. Plant Design including
Safety and Emergency Systems
__19.0 3. Instruments and Controls
19.5 4. Procedures: Normal,
Abnormal Emergency, and
Radiological Control
75.5 TOTALS
Final Grade %
All work done on this examination is my own; I have neither siven nor received
aid.
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Candidate's Sienature
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1. PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS,
,HEATLTRANSFER AND FLUID FLOW (19.0)
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-1. The Dopp l e r .De f ec t . 'a t 1007. power is seen to decrease.
significantly over. core Iife. This i s . p r i ma r i l y'- due to
whichLone of the following causes? '(1.0)
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a '. The Fuel Tempera ture ' Coef f ici ent : decreases over core
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life.
b. The bui'idup of Pu240 over. core, life,
c. Mechanical changes in~ the . fuel pins- increase heat
transfer efficiency.
d. The buildup of lons Iived poisons.
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c
REFi E.O. 200912, REACTOR THEORY
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-2. Which one of the foilowing.most correctly deseribes the
change in differential control rod worth as its tip
approaches the bottom of the' core? (1.0)
a. Increases due to colder water and~ effects of MTC.
b. Increases due to increased leakage.
c. Decreases due to decreased flux density.
d. Decreases due to higher baron density.
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c
1 L REFi E.O. 200916 REACTOR THEORY ! l' 4 i !
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1a P_RI,N,Q,lP_l=ES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS,
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. HEAT,,_TR_ANSFER AND FLUID FLOU (CONTINUED)
3. The plant is operating at 757. power when a control valve
failure reduces steam flow to 607.. Which one of the
following correctly describes the response of Tave for the
four hours following the event? Assume that all plant
systems are in manual and no operator action is taken.
(1.0)
a. Initiaiiy inereases and then decreases to Iower than
its original value.
b. Initially increases and then decreases somewhat
stabilizing above its original value,
c. Initially decreases and then stabilizes at a value
below its original value.
d. Initially decreases and then returns to its orginal
value.
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a.
REFi E.O. 200914 REACTOR THEORY
4. Briefly explain how the buildup of Pu239 affects the
response of the reactor to reactivity changes. (1.0)
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You get a faster Start up rate for a given reactivity
insertion.
REFi E.O. 200904 REACTOR THEORY
5. During an appoach to criticality the following data is
taken:
Bank C Position CPS
(inches withdrawn)
0 400
20 500
40 890
65 1910
Using Figure 1-1, estimate critical position. (2.0)
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1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMl_GE
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IKAT. _T R A_N_SF_ER _ANQ _F_L U 1 Q _F_L QU __(,G,QNT_ l_liUEQ )
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See Fisure 1-1 9CDRAW L MC U S NO L AST Two C4rA PowrS
A a rreT SS #,
REFi E.O. 200922 REACTOR THEORY
6. 'FoiIowins a lenethy shutdown, near middle of Iife -the
reactor is brought critical at 1000 ppm baron. Considerins
just xenon-and power defect estimate boron concentration
when full power, equilibrium xenon is obtained. Use Fisures
1-3 and 1-4. (2.0 )
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-- From Fisure 1-3, 4 8 is -4 . 337. .
From Figure 1-4, baron worth is about 125 ppm %
AC 3 = -4.33% x 125 ap_m
= 541 ppm
C at fuli power a 459 ppm
D
REF; E.O. 200919 REACTOR THEORY
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7, At 100% power E.O.L. there is a top and bottom peak of axial flux.
--- ANSWER --- (DUDOOO1558) ---
Bottom peak due to lower moderator temperature (higher density) [O.53
Top peak due to fuel depletion at bottom of core with power
shift to the top.CO.53
--- REFERENCE --- (DUDOOO1558) ---
RO Requalification Examination 86-1, Question 1.08
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8. A certain centrifugal pump has a constant suction pressure
of 100 psis. When the pump runs at 1000 rpm, its discharse
pressure is 200 psis. If you wish to increase the
discharge pressure to 300 psis how fast must the pump run?
(Show your work) (2.0)
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1414 rpm
REF; E.O. 200701 MECHANICAL THEORY
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.
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,
1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYANMICS,
H.,li A_I_lR_A_N_S,F_ER _A_NQ _F_L_U I_Q_F_L_QU __(C_QN_T I_NUEQ)_
9. One indication of proper natural circulation cooling is a
core AT that is roughly the same as normal fulI power AT.
How and why will indicated 6T change if natural circulation
starts to fail? (2.0)
--
AT will increase. The core will always put out the heat and
that heat will rise. However, if there is no flow the T c
indtruments will read the temperature of the stagnant water
in the cold ~les. This will slowly decrease due to losses to
ambient- The water and/or steam heated by the core will
rise into the hot less. Since there is no significant heat
sink as yet the temperature will increase d r i v i n s AT u p .
REF; E.O. 200513 THERMO
10. During cooldown,'with pressurizer pressure at 900 psis,
pressurizer relief valve starts slowly leakins. Pressurizer
relief tank pressure is 5 psis.
a. Usding steam tables, or a Mollier diagram, estimate
relief line temperature. (1.0)
b. Is the steam entering the pressurizer relief tank
saturated, subcooled, or superheated? (1.0)
--
a. About 3100
b. Superheated
REF; E.O. 200506 THERMO
11. a. Define net positive suction head. (1.0)
b. Give two (2) ways minimum NPSH is provided for the
condensate pumps. (1.0)
.
--
a. Actual pressure at pump suction minus saturation
pressure of the fluid at pump suction plus velocity
head of the fluid at pump suction.
or
NPSH is a measure of the pressure above saturation
pressure at the pump suction required to prevent
cavitation.
I
..
.
.
L P_R_l_liC_1 P_l,E_S _QF__NQC_l,E_A_R _P_QQER_ _P_l, ANT _qP_ER AT I ON . THERMODYNAMICS,
H_E_A T _1R A_NS F_E_R. _A_NQ _F_l,U 1 Q _F_L_QQ _ _ (C_QN T_ I_ N U E_Q )
--
b. 1. Condensate depression
2. Elevation difference between pump and condensate
hotwe1l
REF; E.O's 200702 AND 200703 MECHANICAL THEORY
12. With the plant at 507. power, a 107. step load increase
occurs.
a. Assuming beginning of life, sketch reactor power, Tave
and steam pressure vs. time. (1.0)
6. In dotted lines, sketch the same parameters for this
event at end of life. (1.0)
--
a. 5EE li l/5WER To 5. 7
TW CdHCAh TG RffACfAt A TC C it r1 g - G S r o r it t F T6 u d 6 ,
b. See above figure
REF; E.O.'s 200911 AND 200912 REACTOR THEORY
,
.
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2. PLANT DESIGN INCLUDING SAFETY AND EMERGEN_C_Y _SYSTEfiq__(1_hD_1
1. The plant has received an SIAS; the operator notes the
following: "All HPSI and LPSI pump operating BTs are open
and diesels are operating, all three charging pumps are
operating". Is this situation normal for safety injection
actuation? Explain your answer . (1.5)
--
This situation is not normal. The charging pumps shouldn't
be operating. One of two things could have occurred. A
circuit problem didn't trip the charging pumps or only WL-1
tripped. If this occurred, the charging pumps are off but
the indicating lights remain lit.
REFi E.O. 100419 CVCS
2. How will the following valves fail on a loss of air? Choose
from SHUT, OPEN, or AS IS. (2.0)
a. LCV - 222
b. -T4 - 405
OcV
c. Low pressure gland steam valve
d. LCV - 405 and 406
--
a. Shut
b. Shut
c. Open
d. As is
REFi E.O. 100419 CVCS
3. Sketch a one line diagram of the CVCS showing all major
equipment and valves. (3.0)
-- Figure 15-17, attached.
REFi E.O. 100471 B, CVCS
"
.
..-
*
a
'2a PLANT ~ D E S I G N _J_N_C_l,U DJ_N_G__ S A_F_E_T_Y _ _A_N_D _ E_M_E_I1G E N_C_Y,_ S Y S T EM S
(CONTINUED)
How-is pramary cooldown rate.and/or primary temperature
O maintained during Shutdown Cooling System operations with:
a. Main Coolant temperature above 140 DEG F? (1.0)
b. Main Coolant temperature below 140 DEG F? (1.0)
--- ANSWER --- (DUDOOO1545) --- -
a. Flow is regulated by throttling the SD cooling pump discharge
valve.
.
b. CC-TCV-2OO control signal is shifted from the LPST temperature
to Shutdown Cooling Pump suction temperature EO.53 and close
the component cooling temperature control bypass valve
..-
CC-MDV-631 CO.53
'
---
REFERENCE --- (DUDOOO1545) ---
STM 20-15, 21-12
DP 2162, ATT. A, p 4
5. Would the hot les injection system be more important in
mitigating the consequences of a cold les MCS break or a' hot
+
les MCS break? Explain your answer. (2.0)
-- Cold les break, because for this case most of the injected
water bypasses the core and goes out the break. The only
water added to the core makes up for steaming, which
concentrates the boric acid in the core. When the baric '
acid becomes too concentrated > precipitation, plate out
and channel blockase is possible.
REF; E.O. 100601 SIAS
6. Procedures call for injecting 77,000 sailons of water before
switchins to V.C. recirculation.- A plant modification was ~
made to the cavity drain line that requies this line to be
open to the brass drain box durins operation. This
procedure requirement and the plant modification are both
concerned with the same problem Explain. (1.5)
..
o
e
2. P QN_T_ _Q E S_I_G N_ _ _!_N_C_L_y 9 1_N_Q _ S A_F_E_T_'( _ _A_N_Q _ E_M_E_R_G E1J_C_Y_ _ S Y_S_T_O1S
(CONT INUED )
-- The 77,000 gallon requirement is to insure NPSH for V.C.
recirculation in the event of a large LOCA. The plant
modification is also a net positive suction head concern
associated with a LOCA in the RX head regions resulting in
MC water remaining in the cavity resulting in insufficient
NPSH for VC recirculation.
REFi E.O. 200659 SIAS
7. The Comunent Cooling pumps are normally started and stopped
with the discharge valve closed. Explain why this is done.
(1.5)
-- Prevent possibility of a pump accelerating in the reverse
direction due to back flow if the discharge check valve
should fail to close when pump is stopped.
A CC ct'T A N Y QTitC R RCn 5CfM OL E /) psiv ER fet r uu. cuor r,
REFi E.O. 100817 COMPONENT COOLING
8. The processed waste liquids are divided into two
classifications. Griefly list what liquids comprise the two
classifications. (1.0)
--
Class 1
- liquids containing dissolved hydrogen and
fission product gases.
Class II - Ilquids containing dissolved CM: -ecen lWA ff0air
REFi E.O. 1, RAD WASTE DISPOSAL
9. Provide the location and source of power of each motor
control center. (2.0)
.
-- There are six emergency motor control conter. EMCC No. 1 is
in the switchgear room and is fed from manual throwaver No.
1. EMCC No. 2 is in the Si building and is fed from manual
throwaver No. 2. EMCCs No. 3 and No. 4 are in the PICS
building and are fed from emergency busses 1 and 3
respectively. EMCC No. 5 is in the switchgear room and is
fed from No. 3 EMCC. EMCC No. 6 is in the switchgear room
and is fed from EMCC No. 4.
REFi E.O. 106263 AC ELECTRICAL DISTRIBUTION
- _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
0
*
0
L. P_L A N_T_ _Q g S_1_qN,_J_N_Q,l.U QJ_N_Q_ S A_F_E,,T y ,_A_N.Q_ E_M_E_R,Q g N_C_Y_ _ SJ_S_T1f15
LC.QtlT.1NtKQ}
.
10. a. Describe the function and purpose of the Electr.ical
System Station Service System. (1.5)
b. List the backups to the off-site electrical power.
(1.0)
--
e. The Electrical System also includes a Station Service
System which provides the power necessary to drive
auxiliary equipment within the plant. The Station
Service System is designed to ensure that sufficient
main coolant flow is maintained to keep the thermal
rise in the reactor core within safe limits in the
event electrical disturbances causs a partial loss of
station electrical power. The Station Service System
is comprised of three systems, each normally supplied
from a different source. Two are normally supplied by
each of the 115 kv transmission liness and the third is
supplied from the station generator,
b. As a backup to off-site electrical powere an Emergency
Power Systems consisting of three emergency diesel
sencratorse and three 125 v OC station batteries is
provided to maintain service to vital equipment.
REF: E.O. 106201 AC ELECTRICAL DISTRIBUTION
.
.
CHARGING AND VOLUME CONTROL -
-
.
.
Q \$ ~$7 ~
i 16"31
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'
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TO SPRAY LINE+ &V V . INSIDEVC .OUTSIDE VC
1 - -
REFERENCE: 9699 FM 8A 13
g
- 4 3 *- 2 1 * M
_
"
i M TD -4 l'
CH MOV 528
q FROM
l
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TT: 1i: T
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-
COOLER
D FEED & BLEED
l -TO LO.OP 4 a CH MOV 526
-
LOW PRESSURE'
HOT LEG HEAT EXCHANGERS SURGE TANK
,
LCV 222
I M +D .
M r,g
_
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l FROM r (
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i) COLD LEG CH MOV-525 +D VARI CH MOV 527
TD y LPST
' u ORIFICE "
COOLING '
i
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.
SAMPLE A .
FROf4
s
CH MOV 521
g o
-
-
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C'5G. S.G. FED
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.
-L LNSTJIMWLS AND _CONLROI=S __(L8 Jll
For each of the following control system alignments, explain how,
hp if at all, a loss of feedwater transient from 100% power will vary
from the FSAR analysed transient. Include in your response which
parameter will be most effected. Consider each control system
separately. (2.5)
a. Atmospheric steam dumps operate properly in automatic.
b. Pressurizer pressure control operates properly in automatic.
c. Pressurizer level control opreates properly in automatic.
d. MCPS trip 30 seconds after the reactor trip.
e. Emergency feedwater system begins delivery of water with the
steam driven feedpump 10 minutes after the initiation of the
transient.
--- ANSWER --- (DUDOOO1554) ---
,
a. Reduce severity [0.33 reducing peak MCS pressure by lowing Tave E0.23
b. Reduce severity 00.33 reducing peak MCS pressure with sprays CO 23
c. No difference [0.53
Reduce severity [0.33 by reducing Tave [0.23
($A Increase severity E0.33 by increasing Tave since less heat is trans-
p' fered in Natural Circulation CO.23
--- REFERENCE --- (DUDOOO1554) ---
FSAR; 406:2
SRO/RO Retraining; Loss of Feed water Transient: Terminal Objective
2. Provide the seven blank values in the following CVCS
paragraphs. (2.0)
During normal operation the bleed flow is adjusted to about
_____
spm with the vari-orifice and one or three charging
pump is running with its station selector in automatic.
During steady state, the charging pump speed will be such
that the charging flow is equal to the bleed flow and
pressurizer level will be constant at about _____
inches. A
_____
uF chanse in Tave will cause a pressurizer level change
of about _____
to _____
inches. Tavs will go up if reactor
power is incrL25ed or if turbine load is decreased. Tave
will go down if reactor power is decreased or if steam flow
or feedwater flow is increased.
A decreasing Tavs or a Main Coolant System leak will cause
pressurizer level to decrease and the charging pump speed
will increase to pump more water into the Main Coolant
System. The charging pumps which have their control switch
set to automatic will start if the pressurizer level
decreases to _____
inches.
An increasing pressurizer level will cause the charging pump
speed to decrease pumping less water into the Main Coolant
System. All running charging pumps will trip if the
pressurizer invel increases to _____ inches.
_ - - _ _ _- . - _ - _
o
- .
3. INSTRUMENTS AND CONTROLS (CONTINUED)
--
20 120 10 5 to 6 112 129
REF; E101316: I&C PRIMARY
3. a. List all the reactor scram signals that are activated
at full power and deactivated at zero power. Include
setpoint. (1.0)
6. What scrams provide protection during startup and are
deactivated at full power? Include setpoint. (1.0)
--
a. e Low MC flow AP >80%
e Law MC flow Amp A + B > 240 < 960
e Turbine trip
e Generator trip
, L o w GG- WA TE R L e vet.
b. a Int. range SUR <5.2 amp
e Low power scram <35%
REF: E.O.'s 101213 AND 101224 I&C - RPS
4. a. Explain how gammas are eliminated from the source range
signal. (0.5)
6. List the signal and setpoint that automatically applies
source range high voltage. (0.5)
c. If the source range " white memoory light" is on, what
does it indicate? (0.5)
--
a. A discriminator circuit allows only pulses of a certain
value to pass. Gamma pulses produce lower signals than
neutrons interacting with baron, and can therefore be
removed.
b. Intermediate Range (10-9 amps.
c. Scram cutout switch is in the " normal" position and a
scram signai (5 OPM) has been initiated.
REF: NIS DETECTOR OPERATION
- _-- __-__ - ___-_____
o , .
.
'&,
^
..,
'O 4
.L (NST.,R.,1),MXN_T_S _AND _C QNTRAS___(CONT I NUED ) '
.
5. Answer the followins TRUE or. FALSE about the Geiger _Mueller
-Tube (G.M.) Radiation Detector. .
.
a. Uses a baron trifluoride coatins inside'the detector
with an arson quench gas. (0.5)
b. Ha s .-a' shor t period of time when no radiation'may b'e
counted. (0.5)
c. Will completel.y saturate when a samma . enters . but .only
slightly saturate when a neutron enters. (0.5) '
d. Uses a photomultiplier-tube to amplify the radiation it
detects. (0.5) .
--
a'. False
6. True
c. False
d. False ,
REF: E.O. 200406 RAD. PROTECTION
Lh,' Explain what - the probable cause of the automatic stoppage of control
rod motion is in each of.the following situations. Explain what must
-
ube done to' establish outward rod motion. Assume-the manual rod control
switch was positioned to rods out at.the time of rod motion stoppage.
a. The' reactor plant is critical at 10 E-7 amps and power is being
raised to 5%.
b. The reactor plant is at 75% power and power is being raised to
100%.
---
ANSWER --- (DUDOOO1555) ---
a. SUR 1.5 DPM rod block [0.53
,
Depress push button on MC Board when SUR condition has cleared CO.5
b. 130-MWe single rod step EO.53 ,
- Release rod control switch to off posi t'i on CO.53
---
REFERENCE --- (DUDOOO1555) ---
l STM; 34-13, 16 i 4 P I ,
L
- _ - _ _ _ _ - - -
_ , _ ._ . . _ _ ._ ___
f f
-
:.- ,
'
3. [NSTilyMEN_T_S _kND__QQNLR_Ql=,S __( QQML LMQE_Q1
_
.
7. Describe t h e . p u r p o s e. of the NRVs and the sequence'of
operation'once they are actuated. (2.0).
-- The non return valve ~ automatic contro l '- s ys tem serves the
plant's abi~lity to handle a main steam - l ine ' break 'on -the
turbine side of the non return valves'(NRVs). The NRVs!are
provided with automatic quick closure-to ensure that 2 or 3
s t e'a m generators will be available,for heat removal;and
m i n i m i z e ,t,he etfeet of transients i mposed. on the' Main
Coolant Sykstem. In addition to the. automat i c closure of
.the closure of the NRVs, a signal is sent 'to the. protective
system to trip the L eeactor to-reduce the severity of the
cooldown and ensuing transient effects resulting from a main
steam line break. A Main Coolant System high pressure trip
.is provided in the protective s y s t e m -. t o protect against
'I overpressurization caused.by a loss of load incident. A
signal is also sent to.the condensate pump trip circuit,. in.
conjunction with a high containment pressureisignals to stop
t h ee. p um p when there is a steam'line break inside the. vapor
~
containerr
REF: E.O. 101233 I&C - RPS
~-
R E Pl.n cG v/TiH F A CIGT T V P d O VI C E C f A (r [ TTTLGQ
, R TTnciln wT 'A .
-
.
4
ATTACEMurT 2
.
'
' b. P_L A,N_T _ _S_YJiT g[LS_i_ __ AE_S 19_N_ _C p N_T_R_Qk, AN D __1_N_S IByg,N I A T I O N
/ KQNT,(NijdD1
Y' Desceibe the purpose of the NRVs and the sequence of
operati.on once they are actuated. (2.0)
--
The Non-Return Valve Automatic Control System serves the
plant's ability to handle a main steam line break on the
turbine side of the non return valves (NRVs). The NRVs are
provided with automatic quick closure to ensure that 2 or 3
steam generators will be available for heat removal and
minimze the effect of transients imposed on the Main Coolant
System. In addition to the automatic closure of the NRVs, a
signal is sent to the protective system to trip the reactor
to reduce the severity of the cooldown and ensuring
transient effects resulting from a main steam line A-
ga,- cc,e nn, em- ge_ t: t - - - - _ - ..im :- - _ - . AAbreak t- k.L.
; Otcct: .c r"e+a- te protect a ua...=6 uvu. r. ;;;mri: tica
c = u r : J 'v 7 o J ne e nf load incidan+ ^
0 . u . . o's is also sen6 6u-
tLo r a r, d c 7 ,; ;; -.-n +-in r i .- r . . i + . :- : .gr,c;;un witn a ~
'
,s' : C r. t - . r. t i. . ; ; O u r O ;.u..oi , su =6up Lne pump u h u n-
inere is a s t e o ... l...c'uruoi inWde the vapor contaiu %
REF: E.O. 101233 ,
18.C - RPS
The closure of a valve is accomplished by energizing the main
solenoid operated valve operating coil installed in either
train. The main solenoid operated valves are the mechanisms
through which hydraulic fluid is vented from under the piston
to the hydraulic fluid reservoir. Each operator is also
equipped with an exercise (test) solenoid operated valve
. located, one in each train. The test solenoids operated valves
actuate a restriction orifice installed in series with each
main solenoid operated valve. The operation of the test
solenoid operated valve together with the main solenoid
operated valve will closed the valve at a reduced rate.
'
The hydraulic fltiid pressure is developed by an air-driven
pump. The air supplied to the pump passes thru two solenoid
operated valves. Energizing either of the solenoid operating
valve operating coils will secure the air supply to the pump.
.
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--
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8. The plant is at' steady. state, full power.
a. The reactor operator ~ turns on pressurizer backup
heaters to raise pressure and increase spray-flow, in
so doings the pressurizer water temperature increases
by about 1500 F and actual pressurizer level increases.
Does indicated pressurizer level increase, decreases or
remain the same? Explain. (1.0)
6. A slow leak develops in the presssurizer level
reference leg. Would you expect indicated pressurizer
level to increase, decrease, or remain the same?
Explain. (1.0)
--
a. It remains the same because the mass of water in the
pressurizer does not change.-
b. Increase, because the differential pressure between the
reference les and the pressurizer level decreases.
REF: E.O. 101314 1&C - PRIMARY
'9. List five (5) ways (manual and auto) the scram breakers can
be tripped. (2.0)
--
e Any of 3 scram buttons on MCB
e Control switch at scram breaker cubicle
e Signal from scram ampilfier relays
e Signal from main coolant channels of either high or low
pressure
e Non return valve trip
REF: E.O. 101217 - 1&C - RPS
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1. A turbine trip occurs from full power without reactor scram.
The rods cannot be moved by depressing the Manual Scram
Button. List three immediate actions the reactor operator
should take. (1.5)
--
e Emergency borate
e Manually open scram breakers in switchgear room.
o Open Circuit 12. Control Rod Drive Power Supply on
battery Switchgear No.2.
REFI E.O. -
ATUS
2. With the plant at powere an unisolatable pressurizer steam
space break occurs. List five (5 ) indications typical of
this event. (2.0)
-- Any five (5) of the followinsi
e Low or erratic pressurizer pressure
e High pressurizer level
>
e High containment pressure
e High VC drain tank and containment level
e High containment humidity
e High containment temperature
REFI E.O. nl - EMERGENCY PROCEDURES (EP'S)
3. With the plant at powere a steam generator tube rupture
occurs. Main coolant pressure rapidly decreases to 1400
psig and then begins slowly increasing due to safety
injectlon.
a. Lint two (2) unique symptoms at a tube rupture. (1.0)
b. What actions are required to isolate the faulty steam
generator? (1.0)
_- - _ _ _ - _ _ _ - _ _ _ _ . _ _ _ _ _ - _ _
.
.
..
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G.QNT.R A__(G,QN[LWiiQ1
.
--
a. e Hish air ejector radiation level .
e High pressure and level on one steam generator
b. e Stop feedwater flow
e Verify closure of all safety valves on the faulty
SG
e Close the non return' valve on the faulted loop
a Trip MC pump associated with faulted SG
e l' t.0 S e q s ove Tc HCV Stop yist yg
REF E.O.'s 1, 7 and 8 - EP'S
4. During refueling a survey of the vapor container shows a
valve reading SR/hr on contact (valve is approximately 2" in
radius),
a. How far from the valve should a "High Radiation Area"
barricade be constructed? (1.0)
b. Based on maximum alIowabIe quarter 1y extremity dose
rate only, how long can a worker have his hands in
contact with the valve for repair work (assume no-
previous quarterly exposure)? (0.5)
-
--
a. Oose Rate at barrier = 100 mc/hr
EEih1_ " Olstange_ttqm_galgg
-
0.1R/hr 1/e ivus
50 x 1/i - 0:; ten;; -
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THEllT 116 fait.iT so uncc reverso or rirset coa w .
b. LQu2%_8 = 3 u2%_h,tn 8, lde l * : #2 [da)
5 R/hr
REFI E.O. 200401 - RADIATION PROTECTION
5. List all the automatic actions that shou'Id take place on an
emergency shutdown from power where there is automatic
isolation and re-energiration of the No. 1-2400 volt station
service bus. (2.5)
_ _ _ . _ _ _ _ _ _ _ _ _ _ - - . _ _ _ _ _ _ _ _ _ . _ _ ,
y
..
.
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P_R_QC E011FR_E_S__- _tlOR_M_A_L ,_ _ ABNgRtjAb_ftlERGEta_CJ_
b. -
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...
-- e Reactor scram .
e Turbine trip
e Z-126 and/or Y-177 OCBs open auto
e- No. 124 448 ACBs and the' exciter field breaker open
auto
e 1224 or 1324 ACBs close auto
e Emersency generators associated with de-enersized buses
auto start.
e Boiler feed pumps auto trip
e Condensate recirculation valve opens auto
e Automatic operation of the steam dump valve
REFi E.O. #3 - EP8S
( ,, What actions, if any, should be taken if it is determined that the
facility has been operating with a rod in group C, Justify 12 steps below
the average group position for the past 12 days. your
answer. (1.5)
--- ANSWER --- (DUD 0001557) ---
Attempt to restore proper rod alignment. [0.51
Monitor core flux during restoration. [0.53
High peaking factors and fuel damage can occur if recovery is made
too rapidly. [0.53
--- REFERENCE --- (DUD 0001557) ---
License Event and Plant information Report.> lesson plan
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7. Yankee Rower Tech. Spec +,. require you to maintal.o a e
-specitled SHUTDOWN MePGIN c. t a 1 i ~~ t i m e s . AJhat action 's i $.
. required ii SHUTOOWN MARGEN is' denrmined to be !Iess/ th'&n
required? (2.0)
'
y- ;, - ,,i,, ,
-- With the SHUTDOWN MARGIN less ihan reqU4rede immediately
initiate and continue boration .st 126 spm of 2200 ppm b2.ran
concentration or equivalent until the r e q b i r e d - SHUTCMUN r
MARGIN is restored.
.jJith the SHUTDOWN MARGIN (with all control rods inserted) <
5. 07. AK / K , immediately continue boration at 126 som of 12200
ppm baron concentration or equivalent and es tabl ish and
maintain CONTAINMENT INTEGR1TY untiI the requirei! SHUT 00WN <
MARG!N is restored.
.
REFi,,E,0. #6 TECH. SPECS.
-
(0.5)
( a. Why can on!y two Main Coolant Pumps be opreratedibelow 250 F7
Why should the two Main Coolant Pumps b+p in opposite loops? '( 0. 5 )
6. 1
/ -
'
i
' --- ANSWER --- (DUDOOO1560) ,
---
'
I
a. " : cent tr!tt!r ' :ture rf the r2ctr- - eew4 --46,4 2 { C^' >
Cr c "" nr Liiting fuel.
b. Ennurr r;cr ther-21 -! M ! m; L' f the -eir ree! at_ C0_"! { C")
lnrt W6. Fa t t. ;
--- REFERENCE --- (DUDOOO1560) ---
Memo YRP 115/05 JKT to LHH "MCP Operating Rostrictions", Nov. 13; 1905 l
9. Describe the three (3) authorities granted to e, control room '
'l
operator, according to AP-2001. (2.0) ,
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-, , .-- ' e To shutdown the reactor . i s ' he determines that . the
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, safety of the raector is in Jeopardy. or if operatins.
47 /' parameters exceed : reactor . protect ion. setpoi nts and'
"
,. automatie shutdown does not occur,
e To recommend to the Shift Supervisor and/or the:
Supervisory Control Raos Operator chanses in plant
4 - #" * operations or equipment status.which would improve
overalI plant efficiench and safety.
,
~
U' To respond immediately in the absence-of hisher
e
author'i ty ~ t'a abnorma l or emergency. conditions.
p ,
E.O. #4 - ADMINISTRATION
'
REFI
,
!
.i ;
10. An approach to critica.lity is in prosress.
a. What is the minimum number of licensed personnel
required in the control room? (l'.0)
a. What is the minimum control rod position for
; criticality? (1.0)
'
c. Thu reactor operator decides to go critical by
s er . dilbtlons uith rods at the minimum control rod position
'
M I' for e r i t l 'c a l i t y . Esplain why this approach would
' '
'
,
probabiy risult in either an excessive startup rate or
e violction of the minimum critical rod position.
'
'
,
--
a. Iwa (2) RO's and. i Sro
1
b. 0" stops on Group C.
' 2
-
c. If the o p e r a t o i- terminates the dilution when the
'
, reactor is criti'al, c the LPST will continue to supply
.
the charging pumps with water that is less barated than
the core. Startup rate will increase and may exceed
9 the administrative limit of 1 DPM, unless rods are
- '
1
insertede which would violate the zero power insertion
l Iimit.
PEFI E.O.'s 6 and 9 , ADMINISTRATION ,.
.
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REACTOR OPERATOR
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< N SENIOR REACTOR OPERATOR ,
'RLQl!ALIFICATION EXAMINATIONS-
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ANSWER GAUGES
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POS ' CORPORATION
SRO AUD I-T EXAMINATION ~
.
.
. -Facilitys. 'fAM(EE ROLE
Reactor Types LJESTINGHOUSE - PLR
Date Administered ~
ANSWER KEY
.
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ATTACHMENT 4
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POS CORPORATION
.
SRO A t_J O I T EXAMINATION
Facility YAhKEE ROWE
Reactor Typer WESTINGHOUSE - PUR
Date Administered:
Examiner:-
Candidate
INSTRUCTIONS TO CAbOIDATE:
You may use calculators, Steam Tables and Data Book. Use separate paper for
the answers. Write answers on one side only, leaving plenty of. room between
answers. Separate answer sheets by Category and staple each separately. Be sure
your name is on each section's answers. Each question is worth one (1) Point
unless otherwise noted. The passins grade requires at least 70% in each catesary
and a final grade of at least 80%. Examination papers will be picked up six (6)
hours after the examination starts.
Category % of Candidates's % of
_Value . Total Score Cat. Vafue -Category
19.0 5. Theory of Nuclear Power Plant
Operation, Fluids and
Thermodynamics
_18.0 6. Plant Systems Design, Control
and Instrumentation
19.5 7. Procedures - Normal, Abnormal,
Emersency and Radiolo9ical
Control
21,0 8. Administrative Procedures >
Conditions and Limitations
77.5 TOTALS
Final Grade %
All work done on this examination is my ouni I have neither siven nor received
aid.
__
Candidate's Signature.
.
.- _ _ _ _ , - _, _ ,- _ _-_ x - - . , -
.
-
.
5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND
THMMODlN Adi CS._ _U_tuO_1
_
1. Which one of the following best describes the influence of
Samarium (by itself) upon shutdown margin for one week
following a trip from full power? (1.0)
a. No effects the Samarium concentration does not change.
b. Shutdown margin increases for the first three days but
then decreases after that.
c. Shutdown margin increases for several days then remains
constant.
d. Shutdown margin decreases over the course of several
days and then remains constant.
--
c
REF: E.O. 200914 REACTOR THEORY
2. Which one of the following is true concerning the reactivity
effects of fuel temperature? (1.0)
a. The Fuel Temperature Coefficient decreases over core
life but the full power Fuel Temperature Defect
increases over core life.
b. The Fuel Temperature Coefficient increases over core
life but the full power Fuel Temperature Defect
decreases over core life.
c. Both Fuel Temperatu.e Coefficient and full power Fuel
Temperature Defect increase over core life.
d. Both Fuel Temperature Coefficient and full power Fuel
Temperature Defect decrease over core life.
--
b
REF; E.O. 200912 REACTOR THEORY
_.
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.
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- 5. : T H g O R Y - O F ' NU C L E A R _ P_0_WliR __P_L A,N T__QP_E R A_T_I_0_h__F_l t_J
m I O S__A_N_D -
TJdE_Rl10D_Y,NA.tl 2(C.S __(G.ONT I NUED )
~
-
' 3. -WithJ the -plant nearins-EOC, an.ECP has been calculated for a
r e a c t o r -. s t a r t u p -that is'to be performed 15Lhoursc after a-
' trip foilowins a lons fulI power run. For EACH o'n e . o f the
followins actions, assuming that they occur after the ECP.is'
performed, indicate if it-will result in a HIGHER'than.
-
predicted critical position -a' LOWER-than-predicted position ,
or have N0' INFLUENCE on critical position. (1.5)
.
(i)~ Feeding.the-Steam senerators to increase
' level _by 157. causing a drop in Tave
(ii)' Delayin's the :startup by four hours
~
. longer than anticipated-
-
.(ii1) Increasing -the Steam Dump Setpoint by
100 psi
--
(i) Lower
(ii) Lower
~
-
(iii) Hisher
-REFi E.O. ~200922 ^ REACTOR THEORY
4. Near'end of life, the reactor is Just criticals with Bank C
. putied to'60" withdrawn.
..
a. Using Figure 5-2, estimate startup rate if rods are
,
pulled to 70" withdrawn. (1.0)
I
b. Give two-(2) reasons why the same rod withdrawal at
beginning of cycle would result in a lower startup
'
rate. (1.0)
!
-
( --
a. T =
O_e_f3'_ P .0052 .002
'
AA (.1)(.002)
=
u0032 = 16 sec
: 1 2x 10-4
I SUR = 26 = 26 = 1.63 DPM
T 16
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51 THEORY'OF NyCLEAR P O W E R _P_L_A_N T__QP_li R,A_T_I_QL _ FAQJ,DS__A_N_D
T,H_QUQD Y_fiA_d (C,!i _ _ (C,QN_T_ (NUE.,Q1
--
b. e Less reactivity is associated with rod withdrawal
in this region at BOL.
e- Beff'is greater.at"BOL.
REF: E.O.'s 200904-and 200916 REACTOR THEORY
I
5. Figure 5-2 sh'ows baron worth vs. burnup
a. Explain the change in baron-worth as Tavs increases.
(0.5)
6. During an approach to criticality, the reactor operator
. notes that count rate doub'les as baron concentration is
reduced from 500 ppm to 400 ppm. Estimate' Keff after
the dilution. (1.0)
~
--
a. B a r o n ' w o'r t h decreases because the baron molecules s e t'
further apart as Tavs increases.
b. * From Figure 5-2, dilution added about 813 pcm. An
additional 813 pcm would. bring the reactor cr i t i ca'I .
Therefore> = -813 pcm
t.
Kett = 1 = 1-
1-P 1 + .00813
= 0.992 [0,78]
a SHurown ng urg 1s wtwen wer ccvars Opu&C Co.25]
REF: E.O. 200919 REACTOR THEORY
6. List the reactivity coefficients which makeup the overall
power coefficient. (1.0)
--
e Moderator Temperature Coeffielent (0.4]
e Fuei Temperature Coefficient gg,9y
a Void Coefficient [O.D
1
REF: E.O. 200913 REACTOR THEORY
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.
-
.
5 T_tj E.O_R_Y__ O F_ _N U C1._E_A_R _ P_0_U_E R PLANT OPERATION, FLUIDS AND
ItiE R_R1001N M1 LG.S _ _(C O_N I LN_U EiD_1
7. With the reactor at 807., a 57. step load increase occurs
without rod motion.
a. Sketch reactor power, Tave and steam senerator
pressure. (1.5)
b. Write a heat transfer equation relating Tavs, reactor
power, and steam pressure. (0.5)
. . . .
--
a. gg7
807, (b> Power -40%-
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Sdout.d L CA O
Steam IAV
Pressure (b)
--- - -- - _._ _.,
b. Power = UA (Tavs - Tsat)
REF: E.O. 200513 THERMO
8. a. Sketch radial flux profile for the core at beginning of
cycle and end of cycle. (1.0)
6. Why does it change? ((1.0)
--
a.
Boc ECC
-EOG4. OGEN.
b. u: ;t 4 i..,- - tug . ente. ,, 4
+Le c ,, - e ca.2ges i+ +n
L..-... 4,-.,,_ +L,- +Lm
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Ike ~[aNo h~AA'6t* As t.cn D' 50 AT~ ~ [[R tpH s*RY s3 MT FLUV
REFx 200511 THERMO WILL PEAN TCa nRD T ec. Ed6th un'irt
Bur:Caf L Ec'EL S r et E FL ur fdOfftj*
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-S. ' THEORY OF NUC L.E AR ~ P O W El3__PAAJ T__QP_E R A T I O N . - FLU 10S AND
. IHQUQllYNAM[C_S __(C.ONILN_L!E_D1
.
Why -is a :cooldown- more restrictive than a-heatup' for a
~
:: 9.
pressurized thermal shock standpoint? (1.0)
--
Pressur'e stress and thermal stress are additive.
:REF: E.O. 20110'4 MATERIALS (PTS)-
10. a. Define brake horsepower and h'ydraulic horsepower for a
centrifugal pump. (1.0)
6. -Would you. expect pump amps to increase or dec'rease if a-
centrifugal pump (e.g., reactor coolant-pump) loses net
positive suction pressure and begins pumping a'
saturated mixture? Explain. (1.0)
Brake horse power - horsepower in' from pu
'
--- a. motor)
llYMllML1l, ll015f PCWEA - H0dSrPOWfW par ($fd6Y A $l'd 8V f*WW) -
b. Decrease, because fluid. density and therefore pump work
' decreases.
REF: E.O. 200701 MECHANICAL THEORY
11. Using basic heat transfer- equations, explain why RCS flow
rate is about 10 times feedwater flow rate. (2.0)
-- Primary Power =
I"ACS Cp AT
Secondary power = m fw Ah
Since Primary Power = secondary power,
!
InRCS Cp AT = In f wA h
'
r
.5np
= 8 ^h~
G h~T
Since Ah involves a phase change, steam tables show that Ah
'
-is about 10 times Cp A T.
REF: E.O. 200513 HEAT TRANSFER .
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' 5; T HE O_RX _ O F__N1J.C1._ EAR _ P O W E R PLANT- OPERATION, FLUIDS AND
~
IHER_MQQ1NAR1[C_S __(C.Ori[[NLED_1
- .
Deseribe the mode-of-heat transf er .f rom:. f uel' rod to
~
1 - 12. a'. -
coolant a t' s . (1.0)
e- Core' entrance
e- Core midplane.
b. W i t h . t h e :- r e a c t o r - at power > a s ma l 1 ' m e t a 1.1 i c ob.j ec t
becomes' loose.in the reactor coolant: system.and lodges
-
in a. channel entrance-between f uel: rods's partially-
blockins flow to the channel. How would the following
- change in the affected channel? (1.0)-
i 'e Linear heat generation rate
e Departure from Nucieate Boiiins Ratio
-~ sa. e -Core entrance - subcooled convection
Core midp' lance
'
e - nucleate boilins, in which_ steam
bubbles . form on the-surface of'the fuel rod'and
are swept'into the coolant.
b. Both-decrease.
REF: E.O. 200509 THERMO
..
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6_t P_L_AN_T _SlST,EM,S i__QES I_GN _G.QNTR_QL _ANDjNST_RUMENT_A_T I_qN __(1_8_. Q)
_
1. The plant has received an SIAS, the operator notes the
followins: "All HPSI and LPSI pump operating BTs are open
and diesels are operating, all three chargins pumps are
operating". Is this situation normal for safety injection
actuation? Explain your answer. (1.S)
--
This situation is not normal. The chargins pumps shouldn't
be operatins. One of two things could have occurred. A
circuit problem didn't trip the chargins pumps or only UL-1
~
tripped. If this occurred, the chargins pumps are off but
the indicating lights remain lit.
CDC. + BTM CNLY CF Vri. T AG f [$ LCu Civ p16(ts
REF: E.O. 100419 CVCS
2. a. Describe the response of the accumulator- to a Safety
Injection Signal (include approximate values for
pressure, and time frame to reach full pressurization).
(1.0)
b. If the contents of the accumulator are injected to the
MCS during a large break, what automatic valve
manipulations take place on low level? (1.0)
--
a. Two timers activate to open three nitrogen regulator
trip valves, which pressurize the accumulator to 473
t10 psis.
The time frame is approximately 12 seconds.
b. e Nitrosen supply valves trip closed.
m Accumulator outlet closes
e Accumulator relief valves open.
REF: E.O. 100640 SIAS
3. Would the hot les injection system be more important in
mitigating the consequences of a cold les MCS break of a hat
les MCS break? Explain. (2.0)
I
I
. .
.. ;
*
' ,
6. P {, A N T SYSTEglS_;_ _ _ _D_(S_ I G N__G 011_T_R_ql, _ _A_N Q _ _I_ tis T Ril_M_(N T A T_I_O_N_
tc_QNT,! NLED1
-- Cold.les break, because for this case most of the i n.iec ted
wa ter - bypasses the core and - soes out the break. The only
water added to thy core ma kes . .up for steamins, which
concentrates the boric acid in the core. When the boric
acid becomes too concentrateds precipitation, plate out, and
channel blockage is.possible.
-
-REF: E.O. 100601 SIAS
- .
.
.
,
^
4. The Conibo ne n t Coolins pumps are normally started and stopped
with the discharse valve closed. Explain why this is done.
(1.5)
. ..
~
-- P r e v en t passibility.of a pump acceleratins in the'roverse-
direction due to back flow if the -di scharse check valve
should fall to close when pump is stopped. s.
CR A NY 0T fire s tNGL E FEA Scda OLC A r/SwCA Fod Fuu. cdcor7
REF: E.O. -100817 COMPONENT COOLING
How is primary cooldown rate and/or primary temperature
*. maintained during Shutdown Cooling System operations gith: . en ---
4
Main Coolant . temperature above 140 DEG F? (1.0)
a.
b. Main Coolant temperature below 140 DEG F? (1.0)
---
ANSWER --- (DUDOOO1559) ---
Flow is regulated by throttling the SD cooling pump discharge .
!
a.
valve.
'
CA 4SC HC47 ExoMN6cn Gy-M% vacgsg
b. CC-TCV-2OO control signal is shifted from the LPST temperature
to Shutdown Cooling Pump suction temperature.[O.53 and close
[ ' the component cooling temperature control bypass valve l
CC-MOV-631 CO.5]
---
---
REFERENCE - - (DUDOOO1559)
! STM 20-15, 21-12 l OP 2162, ATT. A, p4 , !
'
l '
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c6. P ( A N T = S Y S T_E_M.S__ _D E_S_LG}i__G Obl_T_R QL _'_A_N_Q__1_NRTg_M E N T A T I O N .
(CONTINUED)
6' , .a. Describ'e the ' function and purpose - of the Electrical
System Station Service System. (1.5)
~b. : List the backups to:.the'off-site. electrical power.
(1.0)
~
--
.a. . The ' El ectr i ca l System also includes ia - Station Service
SystemLwhich provides; the power necessary taidrive
-
auxiliary equipment within.the- plant. The .Sta tion
Service System . i s des igned . to ensure that sufficient
m a i n . c o o l a n t f l o w' - i s' ma i nta i ned to' keep the the~rma l
-
rise in 't h e : reactor-core within safe limits in the
eve nt _- e l ectr i ca l disturbances cause a partial loss of
station e l ec tr i ca l ' . power'. The Station Service System
is comprised of three sys tems , ~ each norma l l y 1 supp l l ed
from a different source. Two are normally supplied by
each of the 115 kV transmission Iines, and the third'is
. supplied from the station senerator.
'
b. As a backup to off-site eIectricai power, an- Emersency
Power
senerators,Systg,addg.,cthree o ns i~s125t iVnsDC
of station
three emersency diesel
batteries is
provided to maintain service to vital equ i p m e n t . CO. 2 5]
'
REF:- E.O. 106201 AC ELECTRICAL. DISTRIBUTION
.
7. Describe the response (increase,- decrease or no chanse) of
the narrow ranse pressurizer level instrument to the
following conditions:
a. Leak in the water column which is at ambient
temperature. (0.5)
b. PRZR water temperature increases durins plant heatup
with pressorizer futi. (0.5)
c. Loss of detector voltase. (0.5)
'
f
--
a. Increase
b. Decrease (or it could remain pessed hish)
,.
c. 1 .= c 21 e e
O rtWrA SE
i REF: E101314 I&C - PRIMARY ,
~
-4 .
*
.
: e- ,
,
b P_L A_N_T___S_Y_S_T E11S_;__ __0, E S 1 G N__ C O N_T_R_QL __AfiQ__I_N_S_T R11_MXN T A_T_I_QN
tcONTUNE.D1
Indicate whether or not the following coincident instrument failures
8 *. will result in a reactor scram. Consider each' set of failures
separately-and justify your answer. (3.0)
a. One intermediate range power N1 and one power range NI both fail
high when both coincidence switches for the scram logic are set.to
the coincidence position.
b. Three pressure switches to the non-return valve trip system
fail low: The pressure switch in loop 1 channel 1 (MS-PS-11)
The pressure switch in loop 2 channel 2 (MS-PS-12)
The pressure switch in lo'op 3 channel 3 (MS-PS-13)
c. Main coolant loop 1 pressure channel fails high and main-coolant
loop 2 pressure channel fails low,
d. The over current relay on phase A for the main coolant pump
in11oop 1 fails high and the under current. relay on phase B
for the main coolant pump'in loop 2 fails low.
- --- - ANSWER --- (DUDOOO1541) ---
V
a. Scram CO.253 Any two of the 6 High Power bistables will cause
a trip. CO.53
b. No Scram EO.25] Requires 2 of 3 press switches in a single loop.CO.23
.
c. No scram CO.25] High and low . pressure trips provide seperate signals
, therefore 2/3 coincidence is not met. [O.53 i d. Scram E0.25] Each failure will indicate MCP failure and meet the
2/4 logic for scram. [0.51
3 --- REFERENCE --- (DUDOOO1541) --- 1- l STM: Chap 31; Excore Instruments, p 31-39 .
Chap 33; Reactor Protection, p 19, 23, 34, 36
K+A: 3.1-44:EA2.04 4.4
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6
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.
.
.
0 :. N_ _S E [LS.A. _ _ p_E_S 1 G N_ _C Q N T R_Q( _ _A_N,Q _ _I_N_S I R1(MJEI AJ_I_O N
Y' Describe the purpose of the NRVs and the sequence of
operati.on onc'e-they are actuated. (2.0)
-- The Non-Return Valve Automatic Control System serves the
plant's ability to handle a main steam line break on the
turbine side of the non return valves (NRVs). The NRVs are
provided With automatic quick closure to ensure that 2 or 3
steam generators will be available for heat removal and
minimze the effect of transients imposed on the Main Coolant
System. In addition to the automatic closure of the NRVs> a
signal is sent to the protective system to trip the reactor
to reduce the severity of the cooldown and ensuring
transient effects resulting from a main steam line break. A
Main Coolant System high pressure trip is provided in the
protective system to protect against overpressurization
caused by a loss of load incident. A signal is also sent to
the condensate pump trip circuits in conjunction with a
high containment pressure signals to stop the pump when
there is a steam line break inside the vapor container.
REF: E.O. 101233 1&C - RPS
/} /V Sa/ Cif o u hiGC
R E PLn C C Wir n FACJLTY PACVICCd
T ITt ED A TTnCHMGtVT 1,
,
/
'ATreseeruT 2
.
g P_L A_N_ T _ _S_Y_S T E tijU_ _ _ _0_E_S 19.N,_C Q N_T_R_Q(_ _A_N Q __I_N S IBy_M_lU31 AJ_I_O_N_
KQNT [N UgQ1
Y* Desceibe the purpose of the NRVs and the sequence of
o p e r a t i.o n once they are actuated. (2.D)
-- The Non-Return Valve Automatic Control System serves the
plant's ability to handle a main steam line break on the
turbine side of the non return valves (NRVs). The NRVs are
provided uith automatic quick closure to ensure that 2 or 3
steam generators vill be available for heat removal and
minimze the effect of transients imposed on the Main Coolant
System. In addition to the automatic closure of the NRVs, a
signal is sent to the protective system to trip the reactor
to reduce the severity of the cooldoun and ensuring
transient effects resulting from a main steam line break h A -
M e .' - Cea 'an+ C" rte- '- ! 5F 7 7 : ; ; u .- - t i- le --etided '. t%
pretc;t; : r"e+e- te r-etect oue . .m u uvm. r. ::; c ::tica
ce>::d by W nee nf Icad ineidon+ ^
-
tv is aiso sent ur
tLo r : 7, d ; 7 , ;; - mn + in rireni+. :- : .;u7,;;;un witn a ~
*
. L' C C r. t ; ' - . O r. p.c;curr ; . m .. o ; , tu stup t r.e pump u h u n'
-
tnere is a s t e a .o l...m break iiTTd e the vapor contaiu %
REF: E.O. 101233 ,18.C - RPS
The closure of a valve is accomplished by energizing the main
solenoid operated valve operating coil installed in either
train. The main solenoid operated valves are the mechanisms
'
through which hydraulic fluid is vented from under the piston
to the hydraulic fluid reservoir. Each operator is also
equipped with an exercise (test) solenoid operated valve
_ located, one in each train. The test solenoids operated valves
actuate a restriction orifice installed in series with each
main rolenoid operated valve. The operation of the test
solenoid operated valve together with the main solenoid
operated valve will closed the valve at a reduced rate.
The hydraulic fluid pressure is developed by an air-driven
pump. The air supplied to the pump passes thru two solenoid
operated valves. Energizing either of the solenoid operating )
I
valve operating coils will secure the air supply to the pump. i
1
!
.
,
, ,
. - .
,
,
7
-
,.~
4
..= :.
'7.. LPROCEDURESi' ' NORMAL, ABNORMAL', E
EMERGENC,y__A_N_D__R AQLQL_QG,LC Al
'~
.' G_QNTRQl,,,_,,Ll?,,_51.
.
- _
l'. A' turbine tr.ip occurs-from full power without reactor scram.
The' rods cannot: be moved by depr es s i ns , the .Manua l Scram-
Button. . List three immediate~ actions ' the reactor- operator.
- should.~take. (1.5)
---
; s' JEmergency bcrate
-
e- Manually open scram breakers in-switchgear-room.
e- -Open: Circuit 12, Conteal Rod Drive Power Supply on
battery Switchsear No.2.
REFi E.O. -> ATWS
2. With the p'l a n t at powere an unisolatable pressurizer steam:
. space break occurs. List-five-(5 ) indicatians typical of
~
-this event. -(2.0)-
-- Any five (5) of the followins:
e ' Low or erratic pressurizer pressure
e' -High pressurizer- level
e H'ish containment pressure
-e -High VC drain tank and containment level
e High containment humidity-
e High containment temperature
REFi E.O. #1 - EMERGENCY' PROCEDURES (EP'S)
3. With the plant at power, a steam generator tube rupture
occurs. Main coolant pressure rapidly decreases to 1400
psis and then begins slowly increasins due to safety
injection.
a. List two (2) unique symptoms of a tube rupture. (1.0)
b. What actions are required to isolate the faulty steam
generator? (1.0)
. .
7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND RADIOLOGICAL
CONTROL (CONTINUED)
--
a. e High air ejector radiation level
e High pressure and level on one steam senerator
b. e Stop feedwater flow
e Verify closure of all safety valves on the faulty
SG
e Close the non return valve on the faulted loop
e Trip MC pump associated with faulted SG
REFi E.O.'s 1, 7 and 8 - EP'S
4. During refueling a survey of the vapor container shows a
valve reading SR/hr on contact (valve is approximately 2" in
radius).
a. How far from the valve should a "High Radiation Area"
barricade be constructed? (1.0)
6. Based on maximum allowable quarterly extremity dose
rate only, how long can a worker have his hands in
contact with the valve for repair work (assume no
previous quarterly exposure)? (0.5)
--
a. Dose Rate at barrier = 100 m r / hr [0, .f]
3ft1l.. = Distance from valve
0.1R/hr leo ;uct _
50 1/6 - O!ctance -
S _4_SS+_ _^_ _ LSSESS
+rnm "a l v e-
Tff A T A .5 PCIU T So apu 10 STEAD of h A us .souM .
b. 18.75 R = 3.75 hrs. f, [ g, ) 2. , g (g,)2
5 R/hr
E.O. 200401 - RADIATION PROTECTION
'
REF;
5. List all the automatic actions that should take place on an
emergency shutdown from power where there is automatic
isolation and re-enersization of the No. 1-2400 volt station
service bus. (2.5)
l
r
..
*
. e:.
L R
P_R Q G EQU R_E_S_. - _ N O R_M_A_L_,_ _ ABr_4Q R r_i AL
2 _ fitf R GEE NCJ_ _A_N_0_ _R A Q I_,qL QG I_G,AL
CONTROL (CONTINUED)
t-
%, ,
--
a Reactor-scram
. Turbine trip
e Z-126 and/or Y-177'OCBs open auto
e No. 124 448 ACBs and the exciter- field breaker open
auto
a 1224 or 1324 ACBs close auto
e Emersency senerators associated uith de-enersized buses
auto start
e- Boiler feed pumps auto trip
a - Condensate recirculation valve opens auto
e Automatic operation of the steam dump valve
REF; E . O '. #3 - EP'S
h, Answer the following concerning the Primary Pumps Seal Water System:
a. How .is level normally maintained in the Seal Water Tank? (O.75)
b. How is level maintained during periods of excessive leakage? (0.75)
c. Under what conditions could water from the cooled pump leak
back to the Seal Water Tank? (1.0)
t
--- ANSWER --- (DUDOOO1549) ---
a. By remote operation of the Primary Pumps Seal Water Makeup Pumps from
t
the PWST to Seal Water Tank. [O.753
'
!
b. By operation of the LPST Makeup Pumps from the PWST to the Seal Water
Tank. [O.753
c. If the nitrogen pressure in the Seal Water Tank is allowed to drop to
less than 25 psi greater than the highest discharge pressure of any
operating primary pumps. [1.03
- 6YST C m f) AES$ > SE*4L rdWK fff55uf[
--- REFERENCE --- (DUDOOO1549) ---
STM 23-5, 23-7
i OP 2165, pp. 1,2 L
. _ .
-
.
*
, .
~
7. LPROCEDURES;~ NORMAL, ABNgRtIAl=2_ EtMf R GENC
R Y_ _A_N,_D__IlA,Q1_QlmqG I_C_A1
C, QN_IllQt,._ _(C_QNILNUE_Q1 ,
7. Yankee Rouer~ Tech. Specs. require ~ you to maintain a
specified SHUTDOWN MARGIN at all times. What action is
required-if SHUTDOUN MARGIN is ' determined to'be less than
required? (2.0)
-- With.the SHUTDOWN. MARGIN less than required >J immediately
initiate-and continue -boration at 126 spm'of.-2200 ppm baron
concentration or equivalent until the required SHUTDOWN.
'
MARGIN is restored.
Uith the SHUTDOUN MARGIN-(uith alI control rods inserted) <
5. 07.A K / K , immediately continue baration.at 126 spm of 12200
-ppm baron concentration or equivalent and establish and'
maintain CONTAINMENT INTEGRITY until the required SHUTDOWN
MARGIN is restored.
REFi E.O. #6 TECH. SPECS.
1
g According to OP-3010, Fire or Forced Evacuation of'the Control Room,
what can be used to substitute for, or regain control of, the'following
if they should be lost during the casualty. Be specific as to what
is used and where it is located.
(0.5)
. a. Pressurizer level (0,5)
b. Charging pumps
(0.5)
c. Core temperature
.
--- ANSWER --- (DUDOOO1552) -- -
a. Pzr. level may be approximated using the low pressure surge tank
level.[.53' on S /f f E S !f W DOWM SYSTEM
b.' Charging pumps can be operated locally by overriding their controllers
and jumping the CCP breaker control circuits.E.43 Instructions and
jumpers are inside the breaker cabinets.E.13 n S/f TE St/ uT&## $ 6 TEq
c. Core temperature by using a portable pot;entiameter [.33 and
monitoring the ICI T/C's at the #1 VC blister.E.23
M Sifit Sif47 b /d/ 5YSTEM
--- REFERENCE --- (DUDOOO1552) ---'
Rowe, OP-3010, p.4.
-
er
3-- g- ,. -- 3 - _ , , , c. .y, _ __ y---
:.
.
.
O
P_i1QG ED Ll_R_E_S_;___N_QR_M_A_L_,
C _A_N_D_ _R AQ l Ql=QG l_C,AL
_ _ ABr_4OR M Ab _ f Mg R Gf N CJ_
?,
CONTROL (CONTINUED) -
a. Why can only two Main Coolant Pumps be operated below 250 F? (0.5)
b. Why should the two Main Coolant Pumps be in opposite loops? (0.5)
---
ANSWER --- (DUDOOO1560) ---
'
a. --' A' "-'**' '-~--- ' ' - --- ' - ----- ' '^ "' ~~~-
Car: OP er Lifting fuel.
b. E ur : een the- 21 -ixin; ci t r.: reir e s:1 r.t . CO.5: (c^n
l. 1 7 T i M FUEL
--- REFERENCE --- (DUDOOO1560) ---
Memo YRP 115/85 JKT to LHH "MCP Operating Restrictions", Nov. 13, 1985
10. An approach to criticality is in prosress.
a. What is the minimum number of licensed personnel
required in the control room? (1.0)
6. What is the minimum can'rolt rod position for
criticality? (1.0)
y
c. The reactor operator decides to go c r i t i c a.l . b y
dilution, with rods at the minimum control rod position
for criticality. Explain why this approach would
probably result in either an excessive startup rate or
a violation of the minimum critical rod position.
--
a. Two (2) RO's and 1 Sco
6. O steps on Group C.
l l
c. If the operator terminates the dilution when the
' reactor is critical, the LPST will continue to supply i the cha rg i ns pumps- w i th wa te'r that is less borated than
the core. Startup rate will increase and may exceed
l l
the administrative limit of 1 OPM, unless rods are
i
inserted, which would violate the zero power. insertion.
Iimit.
i
REF; E.O.'s 6 and 9 - ADMINISTRATION
l t I l . i
r:? .. 3 ~ ;. --
-
2
. 8. ; A D M I N I S T R A T I V E = P R O C Ep_U_R_ES,t_ C OND I T I ONS AND LIMITATIONS
..Gi_ al
.
'
'l. ' List all. emergency centers that ~are activated during a SITE-
AREA Emergency ~. (2.0)
-- TSC,- OSC, EOF, ESC,' MEDIA ~ CENTER
~
RE'F: OBJECTIVE.,#4 -
EP
:2.
Describe the three .(3) authorities granted to a control room-
operator, according to AP-2001. (2.0)
e To shutdown the reactor is he dete'emines that . the
--
safety of the reactor is in Jeopardy or if operating
parameters exceed reactor protection setpoints and
occur.
automatic shutdown does not
e To recommend to the Shift Supervisor ~and/or the
Supervisory Control Room Operator changes in plant
operations or equipment status which would improve
overall plant efficiency and safety.
e To respond immediately in the absence of higher
authority to abnormal or emergency conditions.
.
REF; E.O. #4 - ADMINISTRATION .
3.
During power escalation, a Group C control rod sticks at the
4D" withdrawn position and cannot be moved.
a. Using Figure 8-2, determine the maximum-power level
allowable for prolonged operation in this condition.
(1.0)
If the same Group. C cod hadres'teiction stuck'at about i?"
b. on power
withdrawn, would there be any
operation? (1.0)
Accept either 5 3 7. (for 40") or 64% (for 48").
-- a.
Important point is that the answer is based on aligning
all Bank C rods within 8" and that insertion limit will
prevent full power operation.
ACCCPT 50 % rot f act Cffdir
b. Yes,~ power would stilI be Iimited to f,7S%
REFT E.O. #9 TECH. SPECS.
.
- _ _ _ - _ _ ._ . _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _
-,
. .
8. ADMINISTRATIVE PROCEDURES, COND1TIONS AND L iM I T_A_T_1_O_NJ
LCO_NT LtlUEIll
4. Match the proper action statement and fill in time or word
for the conditions. (2.5)
a. Main Coolant System >2735 a. Within _____ _____
psi in Mode 1 establish fire watch
b. SOV 90 out of commission b. Fix in _____ _____ or
be in hot standby in
c. MOV 1 half close in Mode 1 12 hours
d. PRZ code safety valve will c. _____ suspend all core
not life until it sets to alterations or posi-
2490 tive reaetivity
changes
e. Moving fuel and all d. Be in hot standby in
charging pumps are lost _____ _____
f. Shutdown margin at 5.4% e. Fix in __________ or
AK/K in Mode 1 close and de-enersize
block valve
9 Reducins baron in Mode 6
and shutdown cooling flow f. _____open or be in hot
drops to 840 spm standby in _____
hour
h. All fire detectors found 9 Be in hot standby in
to be out in cable tray _____
house
h. _____ suspend all
i. At 100% power the Z-126 dilution activities
high line sees dead
i. _____ start baration at
126spm of 2200 ppm
barated water
--
a. d or s a. One hour
b. e b. 15 minutes
c. f c. Immediately
d. 6 d. One hour
e. c e. One hour
f. i f. Immediately, one hour
9 h 9 One hour
h. a h. immediately
i. d or 9 i. Immediately
REF: E.O. #6 TECH. SPECS.
c: ': .
_....
*
a. a
T _
,
. 8. N
'. A D M I N I S T R A T I Vli__P_R QC @_U_R_E_S_t_ C.ON_D_I_'[.!_QN,S _ A_N_D__l, I (1 IJ_A_T_I_QfiS -
LG.QNTLNtKQ1
.
. .
5 .' ~ Fo i I ow i ns ex tended ' .f u l l' . . power operation, 'the reactor. trips.
A Main Coolant Sample shows-a spike'on dose equiva' lent:1-131
-
.to lasci/sm.
a. Wh a t .i s t h e' : m o s t I-i k e l y caus'e of' the- iodine spike?.
'
(1.0)
,
b. - F i s u r e ' 8--4' shows the allowable short term 11imits .f o r.
iodine spikes. Assumins all other' admin'istrative-
requirements are satisfied >.can startup proceed?. (1.0)
--
a '. Pin hole fuei. leak (s) may_ exist. For those fuei rods:
with leaks, the iodine ~. sap activity tends to wash into
the coolant following a trip, because the steam blanket
inside Ieakins fueI rods - at fulI power collapses when
fuel- rod temperature' decreases _ to zero power values.
The - iodine spike is n.o_t ' n e c e s s a r i l y - indicative of_an_
increase in she number of fuel leakers,
b. No. -- A made_ change requires the l'imitins co'ndition for:
operation to be met without reliance on provi'sions of
the action statements.
REF: E.O. #9 TECH SPECS.
'
6. The Reactor Core Safety Limit establishes combinations of-
'
~ Main CoolantiSystem Tavs, pressure, and reactor power for
which op e r a't i o n is acceptable.
a. What is.the purpose of this limit? (0.5)
t 6. What is the basis for construction of the curves (i.e.,
what two thermodynamic parameters, or limits,-would be
'
potentially violated by operation above the curves)?
(1.0)
c. What action is required if the Reactor Core Safety
Limit is exceeded? (0.5)
i 1 c 4- W
[
l P
.
- - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - _ - _ _ . -- . - - - - -
. . .. . .
.. .
.,
. .
i .
-P.' s
8. ADM I NI STR AT IVE~ PROCEDURES, CONDITIONS AND LIMITATIONS-
(CONT INUED )
--
a. Prevent overheating of the fuel cladding,
b. e DNBR < 1.3~
e . Average enthalpy at core exit greater than
saturation enthalpy
c. Be in HOT. STANDBY'within one hour
REF: ~E.O. #2 TECH. SPECS.
7. TECH. SPEC. 3.5.1 (Accumulator LCO) lists seven (7) . items
which must be satisfied in order for-the low pressure safety
injection accumu.lator to be operable. . List five (5) of.
these items, including required values of each item. (2.5)
--
Any five of the followings
e Isolation valves SI-MOV-1 and SI-TV-608 open,
e A minimum useable contained barated water volume of 700
cubic feet of borated water, equivalent to an indicated
level of 261" in the accumulator.
e A' minimum baron concentration of 2200 ppom.
e An accumulator nitrogen cover pressure of less than 15
psis,
e The nitrosen supply system with three supply pressure
regulating valves set at 473 t 10 psis and at least
-
Sixteen 48 cubic foot nitrogen bottles 1 1390 psis
or
-
Seventeen 48 cubic foot nitrosen bottles 1 1340
psis or
-
Eighteen 48 cubic foot nitrogen bottles 11294 psis
a Two OPERABLE low level venting system and
e Timers set to operate between 11.85 t 0.23 seconds.
REF: E.O. #5 TECH. SPECS.
.. .
.
. _ _
_ _ _ _ _ _ _ _ _ _ _ _
_
, ,
.
y *
-...
82 A_D M_} R(q l 83_T_(Mg _ _PRQG E D_URE[q r, _ _GpKD_!J,j_QN S,,f_N_D_ _( 10,j Jf_D.O_N_5 -
(C0_NI(WED1
-
_
. ; .
.
,q
,
.'
8.
There are three types of Main Coolent System leakage defined in your Tech. *
- 2
Specs. Foras:
classified each occurrence below, qqte the type of leakage it would be
~.
a.
Leakage
tank.
through a valve packing that is routed to a coolant drain '
(0.25)'
b.
Slight
line. seepage through an elbow s,0cket weld on a pressure sensing (0.25)
.
.I
c. Steam generator tube leakage. '
-4
(0.25)
d. Valve packing leaks of unknown origin.'
(0.25)
's -
--
a. Identified leakage
b. Pressure boundary leakage (0.25)
c. Identified leakage (0.25)
d. Unidentified leakage (0.25)-
,
.
(0.25)
.
$
t
,
.9. Per OP-3106 (Loss of Main Coolant) answer the following:
j
a. When is S.I. recirculation flow initiated? (0.5) y
j
b. When is hot leg injection initiated? Why- is this needed? (0,5) (
,
--
a.
Af ter operation with unrestricted flow with 77,'C00 gallons injected (0.5)
(11 feet SI tank). t
'-
b.
20 to 24 hours to prevent boron precipitation in the core.
(0.5)
10.
.
In an Emergency Condition at Yankee, the lead control room -
'
.
operator assumes the responsibility of the. plant Emergency
Director. The Reactor Operator must know his responsibility in
order to communicate the proper information for him to carry
out his duties. List four of his duties that require team work 1
s
and clear concise information exchange. (2.0)
.
e
b
. !
..._ t .
--
r.
-
~_ ,
-- . ,
,,; a;;
.
- "
+;u. --
:y .
u
_
_
M ,
a
t n !bS ,
.g ' 8. A 0 M 1 N 1 S T R A_T_!_V_E_ _P_R O G E.D_U R_ELS_t_ C OfJ_D I T I O N S AND LIMITAT1ONS'
e. GMLLNtKDJ
gf tjj.h
y . . , -
W;
-
y
pci:
'
.
Ifo 2-
..
--
An'y .iohr (4-) of the falIowinst
-
.. . .
s' Mu s t . clear i y input infor_mation that ailows him to
determine proper, emergency. action.. levels.
't .: ,
[ e Must h' ave the proper information to knnw.he is in'the
>
'Froperloperating/ emergency procedure,
&. isg
, X ,
'.h!Jz ,
'
,
t
u Must have information to properly classify emersency.
o g.-
, ~
.
,
o ;Information transfer must be indepth to allow an-
a c c u i; a t e , efficient and timely turnover to the
' technical support center coordinator, or recovery
f.
I
, manassr.r
a Certain of the notifications are a responsibility that
c l'
'#
'
can be delegated. If this response.is used the grader
f / 'l ' nust be ccnvi~nced the RO has dem'onstrated in his answer
* that iihe _dg l egat i on .is of'a controlled nature and not
.
_'
# because he does not possess the information to carry-
'
' N' q
f -
out the re6pansibility.
REF: E. . O . #1 EMERGENCY DIRECTOR
,i
11. A major 't.O C A _ i s in progress. You are the primary plant
operator- and oper:ating the plant per OP-3106 (loss of
coolant) procedure. Explain when you would use the
i fail'oLine prncedures:
c
.i
' ~
NOTE If these procedures would be used as a stand alone
procedure or in conjunction with OP-3106.
1
*
a. Fuel c'ladding faifure (OP-3100) (0.5)
A
t
la . Loss 'o f coolant pressure and/or safety injection
initiation (OP-3051). (0.5)
c. C lass i fi ca t i on of emergencies (OP-3300). (0.5)
d. Inadequate core coolins (OP-3053). (0.5)
l, v.
,
-?
* ~
>
4
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.
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E
70 90
i
*
CONTROL ROD GROUP r 4 fl' a (INCHESWITHDRA '
-
.
*combination Allowable' in operation. THERMAL. Powern pump Eased on the main c
O
.
.
r
,
FICURE 3.1-1
*
.
l .
YANXEE-ROWE '
3/4 Ir29
,
Amendment No. g, sg .
.u s)
I I
;
,
e
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,
0:
c .
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Figure 8-4
.
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.
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. 20 30- 40 50 60 70
, so so too
, I t
PERCENT OF RATED THERMAL. POWER
t l 1
DOSE EQUIVALENT I-131 Primary Coolant v y Limit
Specific Acti it
versus Percent of RATED THERMAL POWER with the Primary Cool
Specific Activity > 1.0 pCi/ gram DOSE EQUIVALENT
.
-
.
ant I 131 -
.
FIGURE 3.4-1 '
!
.
1
-
l
YANKEE-ROWE i
3f4 4-18
/ O
.
4 .
.....p I
/
w_
I
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