ML20209B476
| ML20209B476 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/14/1987 |
| From: | Bjorgen J, Burdick T, Clark F, Mark Daniels, Keeton J, Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20209B464 | List: |
| References | |
| 50-263-OL-87-01, 50-263-OL-87-1, NUDOCS 8704280450 | |
| Download: ML20209B476 (65) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION III
-Report No. 50-263/0L-87-01 Docket lio. 50-263 License No. DPR-22 Licensee:
Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name:
Monticello-Examination Administered At:
Monticello, Minnesota Examination Conducted:
March 9-12, 1987 4//3/97 Examiner.
nsuuis ief Examiner Dat#
M I
ti 13/g-)
Date
~
j
$fdff7 g
Date D. Clark 7
Date A
Yt f/A7/87 Date w
Approved By:
. M.
-d, Cni af 4 f tLE'1 rating Licensing Section Date' Examination Summary Examination administered on March 9-12,1987 (Report No.50-263/0L-87-01)
Written, oral and simulator examinations were administered to seven Senior Reactor Operator (SRO) candidates.
Results:
Six SR0 candidates passed the written examination, six passed the oral examination and three passed the simulator examination.
A total of three candidates passed overall.
8704280450 870423 PDR ADOCK 05000263 V
REPORT DETAILS 1.
Examiners
- R. D. Lanksbury, Region III J. Bjorgen, Region III F. Clark, Region III M. Daniels, Sonalysts J. Keeton, Regian III
- Chief Examiner 2.
Examination Review Meeting Copies of the written examination and answer key were given to the facility aersonnel for review at the conclusion of the written examination.
- acility personnel provided their comments to the examiners on March 12, 1987.
Their comments as well as the resolutions are enclosed as to the report.
3.
Exit Meeting On March 13, 1987, an Exit Meeting was held.
The following personnel were present at this meeting:
Northern States Power J. Gonyeau, Manager Production Training D. Antony, General Superintendent, Operations D. Nevinski, General Superintendent, Engineer and Radiological Protection W. Shamla, Plant Manager R. McGillic, Operations Training Supervisor E. Earney, Training Seperintendent L. Eliason, General Manager, Nuclear Plants G. Neils, General Manager, Headquarters Nuclear Group NRC R. D. Lanksbury, Chief Examiner J. Bjorgen, Examiner M. Daniels, Examiner J. Keeton, Examiner (In Training)
The following topics were discussed in the Exit Meeting:
a.
The examiners identified two generic deficiencies in the candidates' knowledge and abilities. The first deficiency was that the candidates that do not interface with the day-to-day operation of the facility were unfamiliar with routine administrative functions such as work 2
I
g---.
requestprocessing, bypasses (jumpers),themethodsutilizedto-track surveillances and. Technical Specifications Limiting Conditions for Operation (LCO's). This included knowledge of the operating
-shift s participation.in refueling activities.
The second deficiency was the candidates were weak in their ability to locate and recognize proper keys needed for emergency operations.
This included obtaining access to the key storage cabinet.
b.
Two additional items of concern were identified.
The facilities Emergency Operating Procedure (E0P) format is not easy for the candidates to utilize during transient events.
The E0Ps require the candidate to use multiple procedures concurrently.
Since all procedures are contained in a single 8 1/2" x 11" binder, it is considered physically impossible to effectively utilize several procedures at once.
The examiners also noted that the facility did not have a procedure to respond to an inadvertent or spurious Emergency Core Cooling System initiation.
c.
A general discussion was held concerning the examination process, the future direction of the operator licensing program and the impact of the proposed change to 10 CFR 55 on the operator licensing process.
The examiners explained that the answers to many questions will be provided in a future meeting near.the Region III office, currently. scheduled for April 16, 1987.
t d
3
ATTACHMENT 1 WRITTEN EXAMINATION C0hrtENTS AND RESPONSES SENIOR REACTOR OPERATOR EXAMINATION ADMINISTERED MARCH 9, 1987 Docket No. 50-263 Sheet 1 of 16 COMENT: General ES-201 page 5 of NUREG-1021 states, "Although the written examinations must be appropriately thorough and comprehensive, they should not be so long that a knowledgeable candidate cannot complete the examination in the time allotted.
The exam administered on March 9, 1987, was of such length that no individuals finished early and most individuals did not have time to review their answers; which means they were not finished with the exam.
I believe they were knowledgeable individuals.
RESPONSE
The Region considers that " complete with the exam" means that the individual has completed answering the questions and does not necessarily include a review period.
All candidates were able to answer every question and in no case were any left blank.
The examiners observed that most of the candidates appeared to be done before the end of the allotted time and in a review made when the exams were called in.
However, in light of this comment the exam was reviewed using the guidance of ES-202, Paragraph E.9.
ES-202 states that as a rule of thumb multiple choice, true/ false, matching, and completion items generally require no longer than 2 minutes to answer and short answer questions generally require 3-4 minutes to answer.
The analysis that is believed to represent a very reasonable time for the exam to be completed resulted in a time of 336 minutes, or a total of 24 mi
'es for any review work.
This corresponds with the examiner reco11e
'on of the amount of time it took most the candidates to complete answering the questions.
COMMENT:
5.01 The core flow increase transient is the limiting transient when reactor power is between 45% and 100%.
When less than 45%, other transients may be more limiting.
See T.S. Page 217, bottom of page.
RESPONSE
Comment accepted.
If the candidate specifies that power is less than 45%, will accept "to protect the fuel (core) from a localized event such as a rod withdrawal error such that the safety limit MCPR requirement is not exceeded,"
Answer key changed.
COMMENT:
5.02 c.
APLHGR may or may not be MAPRAT MAPLHGR Examinee must assume MAPLHGR Limit before answer in answer key is correct.
Reference - M8104L-012 Revision 0, Page 14 of 24.
Other references - GE Thermodynamics book Page 9-74 and Reed Roberts Burns, Nuclear Power Plant Thermal Sciences Page 12-5.
RESPONSE
Comment noted.
Comment not addressed due to all candidates answering the question correctly.
COMMENT:
5.04 The most important reason that peff is 0.0070 instead of 0.0065 at BOL is that U-238 makes up 9% of the fission events in our core.
This results in an average Delayed Neutron Fraction (p)ision 2 Page 32 of 50).of 0.0072 at BOL, 10 (Ref. M8102L-007 Rev That is an increase of 0.0005.
It is true that delayed neutrons are born at a lower energy, but for the same time in core life this means that since they can induce fever fast fissions than prompt neutrons (Ref. Page 36 of 50) they are not quite as important as (I<1). Therefore, p is reduced by the importance factor of 0.972 resulting in a drop of 0.0002.
This drop in p due to the fact that delayed neutrons are born at a lower energy is less than the increase resulting from the effect of U-238 in our core.
Also, the use of the terms Beta and Beta Effective can be very misleading.
The term Beta Effective may lead one to believe the question is concerned with a core fueled only with U-235 which is not the case for our core.
When discussing a core with more than one fissionable isotope, one must discuss the average delayed neutron fraction (Beta Bar) and then proceed to the thermal energies (Beta Bar Effective more commonly referred to as Effective Delayed Neutron Fractions).
RESPONSE
Comment not accepted.
The question is theoretical and only requests a reason be provided for the difference in beta and beta effective for U-235.
COMMENT:
5.05 Answer Part 1:
It is true that the void coefficient, a
increases as core void fraction increases (Reference M5102L-016, Revision 0, Page 33 of 53).
However, the fuel temperature coefficient aD also becomes more negative with an increase in voids.
(ReferencePage27of53andPage16 of17 Attachment 1).
Therefore, the license candidate may select either a, c, or both and be correct.
2
Answer Part 2: Tcainee may answer c since'an increase of core inlet subcooling would cause a substantial decrease in core void concentration resulting in a large Jositive reactivity addition (Ref. M8102L-016. Revision 0 ) age 37 of53).
However, answer b is correct if the candidate assumes only the effects on a voidless core or node (same Ref. Page 10 of 53).
Answer Part 4:
The candidate may answer either a or c depending on how he interprets the word " advantage." The fuel temperature coefficient, aD, is taken advantage of to limit the rate of protection action. power increase prior to automatic plant The void coefficient, a, is taken advantageofwhentheATWScircuitrytripsEherecirc.
field breakers.
RESPONSE
Part 1:
Comment not accepted.
During the examination, the candidates were instructed that the question was asking for the most dominant coefficient and voids are much greater than doppler.
Part 2: Comment accepted. Will accept b, c, or both.
Answer key changed.
Part 4:
Comment not accepted.
The question clearly states "during ATWS." The purpose of tripping the recirculation pumps and maintaining a reduced reactor water level is to increase the core void concentration.
COMMENT:
5.06 The Operations Manual C.2 also gives reasons for rod patterns and exchanges.
RESPONSE
Comment accepted.
Reasons for rod pattern control in Operations Manual C.2 II.A>1-6 will be acceptable answers. Answer key changed.
COMMENT:
5.10 QuestionParta:
There is no fast start at Monticello, although start-ups with high Xenon are not uncommon.
Answer Part a:
Candidate may answer that rods will need to be inserted to stay within the power to flow map.
The candidate may also answer the PCIOMR and thermal considerations are of an increased concern during a rapid Xenon burnout (Ref. C.2-0026 and 0027).
QuestionPartb:
Thisquestionisdoublejeopardy.
First, the examinee must know that the shutdown margin test is performed Xenon free.
Second, he must know the approximate time for the core to become Xenon free, which are you testing for?
3
NUREG-1021 ES-202 Page 4 of 6 No. 13 states,
" Double jeopardy questions should not be used."
RESPONSE
Part a:
Comment accepted.
Will accept, "due to the burn out of Xenon, rods may"need to be inserted to stay within the power to flow map.
Answer key changed.
Part b:
Comment not accepted.
The candidates should know that shutdown margin tests are performed Xenon free and the approximate length of time, following a shutdown, for Xenon free conditions. Question is not double jeopardy.
COMMENT:
5.12 The rod need not be fully withdrawn.
Reference M8102L-016 Revision 0 Page 46 of 53, D.1.
The answer in the answer key does not answer the question.
It is unclear.
RESPONSE
Comment partially accepted.
Correct responses indicating, relative control rod worth change, change in zone of control, number of neutrons from adjacent cells being absorbed by other rods, will be accepted.
Answer key changed.
COMMENT:
5.14 What is the basis for asking"this question? The K/A catalog for 293003 Kl.22 is Explain the usefulness of the steam tables to the Control Room Operator." The NSP referenced lesson plan, M8104L-017 Revision 3, never uses gross enthalphy, H, for anything more than introduction (Page 5 of 17).
Gross enthalpy is not in the steam tables nor is it used by the operator.
Thisquestionisalsodoublejeopardy.
What are you testing for, the candidate's knowledge of the difference between gross and specific or the candidate's knowledge of the thermodynamic cycle across a moisture separator, or both?
RESPONSE
Comment not accepted.
Page 2 of 17, NSP Lesson Plan M-8104-L-017, Revision 3, Enabling Objectives, requires candidates to define enthalpy and determine specific enthalpy.
Questionisnotdoublejeopardy.
COMMENT:
5.15 Question Part c:
The wording of the question may be interpreted by the candidate as how rod worth at B0C compares to EOC instead of over core life.
If the question is interpreted in this manner, the candidate would answer increase (e.g., What would be the effect on the rate of melting of an ice cube at 40 F vs.100F? Answer:
It would melt slower.).
4
m
RESPONSE
-Comment partially accepted. Will accept, " increase,"
if the candidate. states the assumption that two times in core life were compared vice the core aging process.
Answer key unchanged.
COMMENT:
6.01 Upon loss of offsite power, the drywell fans stop running because they have no electrical power, the candidate will need to assume the diesels re-energizing the bus.
RESPONSE
Comment not accepted. The examination question did not specify a problem with the diesel generators.
The automatic progression is that the busses will be re-energized by the diesel generators on a loss of offsite power.
The question was intended to determine if the candidates knew whether or not the fans automatically restarted.
COMMENT:
6.02 a.
The answer presented in the answer key gives a reference of B.4.2.
The answer cannot be found in B.4.2.
The function of the system as stated in B.4.2 is, "To prevent ground level escape of airborne radioactivity from the Reactor Building during an isolated condition, and to remove radioactive particulate and halogens from exhaust gases when a high radiation condition exists."
B.4.2 does not specifically address the high drywell pressure initiation of SBGT.
Technical Specifications, Page 18, states the answer in the answer key, but it refers to SBGT design not specifically the high drywell pressure initiation.
We could not find information which directly supports the question.
b.
Cannot find any information to support the answer in the answer key under the references given.
RESPONSE
a.
Comment not accepted.
The answer can be empirically derived from information in Reference B.4.2, as described in Section 5.3 of the Updated Safety Analysis Report.
b.
Comment partially accepted.
In Reference B.4.2, the Charcoal Adsorber Bed is incorrectly and repeatedly referenced to as the Charcoal Absorber.
Reference B.4.2 discusses the operation of the train heater unit but does not address the adverse effect increased humidity will have on the charcoal adsorber bed.
The training material is considered to be deficient in this area, Reference U.S. NRC Regulatory Guide 1.52.
The reference for Part a was revised to include Section 5.3 of the UFSAR.
The reference for Part b was revised to include regulatory Guide 1.52.
5
-_.=. -
+
~.
' Device and panel number should not be required.
COMENT:
6.03-a j
Should also accept:. (1) SRV open annunciator; the device which turns on tailpipe pressure amber lig/ feed flowht inputs into the.annuciator circuit, and (2) Steam flow mismatch.
Reference:
C.4.B.3.3.A
~
RESPONSE
Comment accepted. Will_ accept name vice device number and general control board location.- The SRV open annunciator and steam flow / feed flow mismatch will be added to' answer key.
Answer key changed.
COMENT:
6.04 a.
-This would be true if inerting was in progress; after initial inerting the heater is no longer used. Make-up during plant operation is from the top of the nitrogen tank.
7 b.
The only answer which was found in B.4.1 Page 98 was Answer 2.
Would like to have reference for the other three answers in the answer key.
The IE Bulletin addresses cracking which makes Answers 2 and 3 about the same.
RESPONSE
a.
Comment not accepted. The question' addresses the use of steam heating and the candidate needs to know this is for initial inerting.
b.
Comment accepted.
Check additional reference; General l
Electric Company's Service Information Letter (SIL) 402 for remainder of the information.
COMENT:
6.05 a.
What is approximately? In the past, NRC exams have allowed setpoints to be i about 10%.
Technical i
Specifications allow the ATWS trip to be less than or equal to 1150 psig.
That'is very close to 1153 as stated.in the exam.
This could be confusing to 1
the candidate.
The exam question did not state whether to use the actual setpoint or the Tech.
j Spec. setting requirements.
Reference:
T.S. Page 60 i
d.
Part d does not indicate how the flow element fails.
If it fails high, only a rod block would occur.
If j.
It fails low, a rod block and half scram occurs (maybe).
t l
Reference:
B.5.1.2 Page 12
RESPONSE
a.
Comment accepted.
Setpoint should be 1135 psig. All candidates answered the question correctly.
Question l
and answer key unchanged.
6 l
d.
Comment not accepted.
Supporting documentation was not provided.
COMMENT:
6.07 Part a - This IE Notice was taken from Westinghouse instrumentation.
However, a similar incident occurred here at Monticello.
Since that time, Engineering evaluations 86-42 and 86-46 have been performed and a design change has been made to the IRMs.
IF 24 volt DC power is lost, an IRM inop trip occurs.
Part b - At Monticello, the APRMs cause a scram at
.58W + 62 in the S/U and run modes and would not be much help during a rod drop accident.
(Ref. B.5.1).
Therefore, at ' 0" recirculation flow reactor power would need to be 62 percent before the scram occurred.
The Monticello Technical Specifications address two other instruments which protect the plant during or after a rod drop accident.
The RWM (Rod Worth Minimizer).
T.S. Page 66 and the mainsteam line radiation monitors T.S. Page 38.
RESPONSE
a.
Commentgartiallyaccepted.
Answer key changed to accept, provides an IRM IN0P Trip, IRM IN0P alarm and a Rod Block."
b.
Comment accepted. Will accept:
1.
APRMS 2.
IRM IN0P Trip 3.
Main Steam Line Radiation Monitors 4.
Rod Worth Minimizers Answer key changed.
COMMENT:
6.09 This question does not lend its self to gaining the answer in the answer key.
1.
That are no automatic pressure relief system low-low 1Elel actuation switches at Monticello.
2.
How are the switches wired incorrectly?
3.
How is a candidate to answer this question the way it is asked?
4.
The references do not contain these switches.
RESPONSE
Comment partially accepted.
1.
System description B.3.3 Pages 8 and 9 identify the pertinent switches.
7
2.
During the examination, the question was clarified that the switches would not perform their required action.
3.
No response.
4.
System Descriptions B.3.3 Page 8 and 9, B.3.4 Page 31, B.3.1 Page 10 (with a typographical error that identifies the switches as 2-3-72A through D),
B.2.3 Page 2, and B.1.1 Page 22, identify these switches.
COMMENT:
6.11 Part b.
If the reactor is in hot standby then it is critical by T.S. definition of hot standby.
Therefore, over 12 rods have been pulled and the RWM may be bypassed with a second qualified individual verifying the rod pattern.
The T.S. states that 12 rods in the startup must be removed not the first 12 rods in each power increase.
Is the candidate suppose to know how many rods are in each group? It appears so by the answer key.
Question has misleading information.
RESPONSE
Comment not accepted.
The question was developed from Page 25 of Procedure B.5.2 which requires the RWM to be operable for the first 12 rods to be withdrawn regardless of the position in the rod sequence, if the RWM is required to be operable.
COMMENT:
6.13 Part a.
Hydrogen should be a sufficient answer.
The question already states lost ventilation to batteries.
RESPONSE
Comment accepted. Answer key unchanged since the response intended is hydrogen.
COMMENT:
6.15 Knowledge and ability reference should be K6.01.
RESPONSE
Comment accepted.
K6.01(3.0) added to the reference.
COMMENT:
7.01 Part a.
See attached Procedure c.4.c.
j i
The procedure has been changed.
The candidate most likely will try to give five answers because the question asks for five; however, there is only one immediate action.
RESPONSE
Based on a discussion between R. McGillic, NSP, and T. Burdick, Region III on March 20, 1987, Part a was deleted.
For future exams, however, actions required i
by operators to be performed prior to abandoning the control room will be considered immediate actions.
i 8
l l
t
This response reflects the uncertainty associated with the lengt.h of time available to the operators, the habitability in the control room (availability of lighting, etc.), and the guidance of Section 5.3.9 of ANSI 18.7 - 1976 which describes immediate operator actions in actual or potential emergencies.
COMMENT:
7.02
.This question is very misleading. The indications given could be a loss of several power sources.
Loss of oower sources with a scram. What does the APRMs are drifting down scale mean? This does not lead you to answer the question as indicated by the answer key.
If the candidate assumed the wrong event by misleading information he could meter a manual procedure, abnormal procedure, E0P or any combination of the three.
RESPONSE
Comment partially accepted.
The question requires the candidate to apply his knowledge of the power supplies to the various indications and take a course of action based on the information supplied.
Rod )osition is not known.
Reactor power can be assumed to )e above 3% since no mention is made of the scram lights on the full core display.
There are two indications of a level problem (meters downscale and alarm).
Two worst case situations are possible; an ATWS or a level problem.
The entry condition for the level E0P is specifically identified (meters downscale).
Credit was the E0Ps for reactivity (power) given for entry into control or level control (C-5-1100 series).
COMMENT:
7.04 Part a.
Should not need service water for full credit.
Part c.
If the main condenser is available, having circulating water available is already assumed.
RESPONSE
Part a:
Comment accepted.
Answer key changed.
RBCCW as the heat sink assumed service water available.
Part c:
Comment accepted.
Answer key changed.
COMMENT:
7.05 Part b.
There is no procedure which specifically addresses a HPCI start at power. Therefore, there are no immediate actions.
The candidates will try to give three actions because the question asks for three responses.
i There are cautions which exist that instruct the operator to verify that a valid signal is not present before securing a ECCS system.
But these are in the E0Ps and given the questica, an E0P entry condition has I
not occurred.
The answer is not found in B.3.2 as indicated by the answer key.
9
. _. - -. ~... _ _
Also', the Simulator Cause and Effects book is referenced.
Are questions for NRC exams now going to be taken from this book?
It is not used as a training manual, it is used for simulator documentations and as an instructor aid.
' RESPONSE:
Comment not accepted.
The examiners noted during exam development that Monticello did not have a specific procedure to address an inadvertent HPCI initiation at power. Given a HPCI start, operators must assume a valid start signal until proven otherwise. Accordingly, the entry conditions for C.5.1100 and C.5.1200 can be considered to be present until the operators take the actions of the E0Ps. Answer key changed to accept " scram the reactor" as an immediate action.
C-5-1100 series and C-5-1200 series procedures added to references.
Credit was given for all reasonable answers.
The plant simulator i
manual was used merely to document the plant response for Part a.
It was noted that the initial plant reaction also is a power decrease due to increased voiding caused by HPCI steam fiow.
Answer key changed to accept decrease if supported.
COMMENT:
7.06 Part a.
The C.5-1103 procedure lists five other ways to insert rods:
individual rod scram, vent the scram air header, individually drive rods, ASDS panel air dump valve, CRD high point vent.
Reference:
C.5-1103 Page 11, 13, 19, 21 Part b.
What is this question asking for, the systems used in 1103, 2006, or 2007 procedure? This question is open-ended and is not the way an individual would be directed to the correct answer.
Part b.
In response to the answer keys.
Answer b.1.
If the actions taken in Part a were completed, the recirculation pumps would be tripped.
Answer b.2.
BoroninjectioncouldbebySBLC,RWCU, or CRD.
Answer b.3.
ADS is inhibited sometimes, not always.
Answer b.4.
Other systems for level control are acceptable.
Answer b.5.
Using the main condenser as a heat sink is acceptable.
The low-low group one isolation can be bypassed.
See C.5-1103, 2006, and 2007.
10
RESPONSE
Part a:
Comment accepted.
Credit was given for any method outlined in C-5-1103.
Answer key changed to add individual rod scram vent scram air header; individually drive rods with RMCS; Vent HCU's locally.
Part b:
Comment partially accepted.
The question is intended to ascertain the candidate's knowledge of the important considerations in dealing with an ATWS, i.e.,
(1) minimize cold injection; (2) lower level to maximize boron concentrations and utilize voids a::d steam cooling to limit power; (3) inject boron; (4) Use HPCI and RCIC to help draw off steam and smooth power spikes, but at the same time control level just above top of the fuel, and (5) initiate containment cooling. Answer key changed to add "will accept all reasonable answers that reflect the guidance of C-5 operations manuals." Since the reason for the Group I isolation was not given in the question, answers that reopened the MSIV's were required to be appropriately supported.
Answers that initiated ADS were also required to be supported.
Part b also states that if the actions of Part a were unsuccessful... so there i T no assurance that the ATWS switches tripped the recirculation pumps.
COMMENT:
7.07 Part a.
Should state number of responses required.
Part b.
What are core parameters?
RESPONSE
Part a.
Comment noted.
Since there were only three, stating the number was not considered necessary.
Part b.
Comment partially accepted.
This comment was discussed during the exam review meeting.
Sufficient clarification was provided to candidates during the examination.
COMMENT:
7.09 Part a.
Could not find answer in references given.
T.S. Page 18 states, "The low level scram assures the basis for the safety limit is maintained."
Part b.
There are three separate recirculation pump low level trips at -48 inches.
The question does not ask for a specific bases.
RESPONSE
Part a:
Corsment not accepted.
Procedure B.5.6 Page 4 is the answer location.
The bases for the safety limit scram at +9" y the answer given.
is essentiall A statement that the is to prevent exceeding a safety limit is insufficient.
11
Part b:
Comment not accepted.
The answer key was obtained from Technical Specifications Page 68.
Su) porting documentation was not provided for the otler two trips identified in the comment.
COMMENT: 8.01 Part a.
A correct answer would be after the valve was repaired.
Tech. Specs require only one valve in each line to be closed.
Part b.
None is correct answer.
Even if the plant was started up with one MSIV closed, neither Group I logic or the RPS logic would require modification.
Why? No Tech. Specs. change is required.
Part c.
_h
Reference:
T.S. Page 3 and 171.
Part a:
Comment partially acce Answer key changed
RESPONSE
to accept "after valve repair."pted.
Part b:
Comment partially accepted.
For other than this, if candidate states valve is to be repaired, the question asks about functions that would likely require review for the need of a temporary change.
This includes more than logic, e.g., the Group I isolation flow settings above 140%
steam flow.
It was understood that following review, the decision may be that nothing needed to be changed.
Part c:
Comment Noted.
Part C deleted.
COMMENT: 8.02 Why all the misleading information? This is not in accordance with 1021 guidelines.
The operator would have no way of knowing what the operating MCPR, the APLHGR were during the transient.
RESPONSE:'
Comment not accepted.
The question is testing for the candidates knowledge of Technical Specification bases that LC0 limits exceeded during transients are not reportable.
COMMENT: 8.04 Operation above 1670 Mwth does not constitute a violation of Safety Limit (Ref. 2.1.A) as indicated in the answer key.
The most serious violation in Tech. Specs. for this condition is that the APRMs, which are calibrated to the heat balance, were set unconservatively.
This may have resulted in operation without meeting the Limiting Safety System Setting (LSSS) for APRMs (Ref. T.S. 2.3.A.1 Page 6).
According to 10 CFR 50.36(c) Subparagraph (ii)(A), the corrective action for operating with the automatic safety system not functioning as required, the licensee shall take appropriate action, which my include shutting 12
O down the reactor.
For this event, no event has occurred requiring the APRMs to perform their protective action; therefore, no safety limit could have been violated.
The correct action would be to declare the APRMs inop.
per Tech. Soec. Reactor Protection System LCO 3.1.A (T.S.
Page 26).
To meet the requirement of Table 3.1.1 (T.S.
Page 28) we must either:
(1) Recalibrate the APRMs to the LSSS 2.3.A.1; or (2) Have all operable rods inserted with eight hours; or (3) Have power in the IRM range or below with the mode switch not in RUN, or (4) Reca11brate the feed flow instruments.
The most likely action would be to recalibrate the APRMs, or feed flow transmitters,-notify the Commission per 10 CFR 50.36(c), and review the matter and record the results of the review.
However, the candidate may commence a plant shutdown which is conservative.
Due to'the wording of Parts a and b of the question, the candidate may tie Parts a and b together, realize that this fits the number of actions required for a violation of a safety limits, and answer the question as if a safety limit has been violated.
Since the question is misleading, this answer should not be counted wrong either (the question does not elicit the required response per ES-202 Part E.17 Page 5 of 6).
Since Part a is incorrect, Part b will be incorrect.
Comment partially accepted.
Part a. changed to acce
RESPONSE
" Reduce power to <1670 MWT and Recalibrate APRM's." pt Part B deleted. -
COMMENT: 8.05 Part b.
Should except "those individuals required for shutdown outside the control room."
Reference:
T.S. Page 232
RESPONSE
Comment not accepted.
Technical Specifications Page 232 Section 6.1.C.6 specified minimum numbers.
COMMENT:
8.09 1021 ES-402 Page 4 of 4 for Category 8 states, "The candidate is not expected to memorize the exact details, numbers, and surveillance requirements contained therein."
This question expects the individual to know all the surveillance number requirements, the number of days in each month and whether it's a leap, year.
It should be enough to know that Technical Specifications allow a 25%
tolerance for surveillances.
This is not a fair question.
RESPONSE
Comment not accepted.
Liberal grading was given based on the candidate's assumptions.
The question was testing the candidate's ability to apply the 25% tolerance and the knowledge that the tolerance is not cumulative.
13
COMMENT:
8.10 The reference does not fit the question.
-RESPONSE:
Comment not accepted.
The question is testing the candidate's knowledge of operability.
COMMENT:
8.11 This question has little to do with operating the plant.
If 10 CFR 50.72 was supplied with the test, why was 10 CFR 55 not supplied? Should the 10 CFRs be committed to memory.
RESPONSE
Comment not accepted.
KnowleJqe of whether or not an operator's license is expired is considered an appropriate question consistent with the Examiner Standards.
COMMENT:
8.12 Should also accept "not fit for duty" as specified in administrative controls.
Same comment as 8.11.
RESPONSE
Comment partially accepted.
A licensed operator should know when he is not in compliance with the requirements of 10 CFR 55 and when he or the utility is required to notify the NRC.
This is considered to be consistent with the Examiner Standards. Answer key changed to accept alternate wording.
COMMENT:
3.14 Part b.
The way the question is worded, the mode switch in shutdown or refuel would be an acceptable answer.
Reference:
Tech. Specs.
RESPONSE
Comment noted.
Alternate wording to match the answer key was accepted during grading.
COMMENT:
8.15 Should accept Site Superintendent or Shift Supervisor.
Reference:
ACD 4.7 Paragraph 6.15.
RESPONSE
Comment accepted. Answer key changed.
COMMENT:
8.16 Should also accept:
NRC Operations Office via Silver Springs Central Office Commercial telephone system to NRC Office via Bethesda Central Office.
Reference:
4 AWI 3.9 Paragraph 6.2.
RESPONSE
Comment accepted.
Answer key changed.
Liberal grading was applied to accept any reasonable method that was considered viable and sufficiently timely.
Answer key so noted.
14 l
i COMENT:
CONCLUSIONS:
NUREG 1021 - ES-202 General Guidance for written exams gives guidelines for the written exams.
No. 10 "All examination questions should be " objective" in the following regard,"
"There should be only one answer"
- for some of the. questions in the exam, there was more than one correct answer-or there was no correct answer.
1 No. 10 "Short answer questions should be as precise and
(~
specific as possible to ensure that the candidate clearly knows what constitutes a fully correct j
response."
Some questions were not clear, precise, or specific, and were confusing to the candidate.
No. 13 There were several " double jeopardy" questions.
No. 16 Vague "open-ended" questions should be avoided.
There were several "open-ended" questions.
RESPONSE
Comments noted. The examiners always attempt to conform to the regulations and the Examiner Standards.
Resolution was provided for each specific problem where considered appropriate.
15
r I. u. >t.
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_UDHIICELLg______________
S n fA?/~ AT 4afr7 REACTOR TYPEt
_9HB:GE3.________________
FWaw:
6.' fdf7 DATE ADMINISTERED *_EZl92292________________
_EJDBGEH1_di_____________
j I%t pA TE, i J
\\
k IfATE INSIBUGIiOHE_ID_G6HDIQ8IEi Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70%.in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF,
CANDIDATE'S CATEGORY
__MeLUE_ _IDIok'
___EGORE___
_MobME__ ______________GGIEE9BI_____________
25.it fjf'_______
________ 5.
THEORY OF NUCLEAR POWER PLANT
.2f1E9__ _Elaif OPERATION, FLUIDS, AND THERMODYNAMICS g.,,. _/,, 24 1
_gp2pp__ _E;;1^
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
'Z1 tS 9&r$$318
_22:2E__ _MT9f
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL ts,zs W W3 i
_ffIff._ _SE 13
________ 8.
ADMINISTRATIVE PROCEDURES, i
CONDITIONS, AND LIMITATIONS 7~., # 95 7
_II ;;__
Totals i
i Final Grade f
l All work done on this exasiination is my own.
I have neither given
,{
not received aid.
1 Candidate's Signature
/\\ / j M,A V
- n. h
} Q \\/
I L
w.L\\.
L g
t.
n 1
i
\\
l 4
~~
s t
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS 4
- During' the administration of this examination the following rules apply!
- 1. ' Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil 901y to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7, Print your name in the upper right-hand corner of the first page of gagh section of the answer sheet.
P, Consecutively number each answer sheet, write "End of Category __" as 9ppropriate, start each category on a Ogw page, write 9012 9D 90211dg of the paper, and write 'Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 14, 6.3.
Skip at least ih gg lines between each answer.
10.
t sheets face
- 11. Separate answee'l sheets from pad and place finished answer down on your desh or table.
- 12. Use abbreviations only if they are commonly used in facility litgratung.
The point value for each question is indicated in parentheses after the
- 13. question and can be used as a guide for the depth of answer required.
assumptions used to obtain an answer
- 14. Show all calculations, methods, or to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
t
- 16. If parts of the examination are not clear as to intent, ask questions of the gxamingt only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
e, t-184 Wasn ycu cecploto ycur oxcoinctien, ycu challt a.
Assemble y'our examination as follows!
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
Turn in all scrap paper and the balance of the paper that you did c.
not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
rakt e
)
5
_IHEDBI_9E_HUCLE68 E9 WEB EL6HI 9EEB6119Ht_ELWIQft.eHE 4
IHEBH991HeHIGH QUESTION 5.01 (1 00) must Explain why the operating limit Minimum Critical Power Ratio (MCPR) be modified when the plant operates at less than 100% rated core flow.
QUESTION 5.02 (2 00)
Hatch each of the four lettered items with one of the numbered items below!
______ Thermal limit by which plastic strain and deformation are-a.
limited to less than 1%.
______ Safety limit that is not analyzed at less than 10% Power or b.
less than 800 psia.
An administrative limit to maintain less than 25% Reactor Power is used instead.
______ APLHGR divided by MAPLHGR.
c.
d.
______ LHGR divided by LHGR limit.
1.
HAPRAT 5.
MCPR 2.
APLHGR 6.
SEXL 3.
CPR 7.
Total PF 4.
FLPD 8.
LHGR QUESTION 5.03 (1.50)
HOW does xenon concentration affect peripheral rod worth following a scram from high power and WHY does this occur?
QUESTION 5.04 (1.50)
Beta for U-235 is 0.0065.
Why is beta effective approximately a.
0.007.
(1.0) b.
How does this affect reactor period?
(0.5)
(xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
Elit_ egg r nor.
Eg-.I IMEBUDRIH851GE QUESTION 5.05 (2.00)
Match the following statements uith the correct coefficient!
(May be used more than once) 1
______ Becomes more effective as steam flow increases.
1.
Will cause reactor power to change when Hi Pressure 2.
feedwater heater isolates.
______ Least responsive to fuel time constant.
3.
______ Taken advantage of during Anticipated Transient Without Scram 4.
(ATHS) conditions.
- a. Doppler Coefficient
- b. Temperature Coefficient
- c. Void Coefficient 00ESTION 5.06 (2.00)
Give TWO reasons for rod pattern exchanges throughout core life.
QUESTION 5.07 (1 00)
TRUE or FALSE damage produced by water hammer is proportional to The potential average fluid velocity and fluid density and inversely proportional to the length of the pipe.
9 GUESTION 5.08 (2.50)
Define Net Positive Suction Head (NPSH).
(1.0) a.
l b.
What effect would HP heater isolation have on Recire Pump NPSH at 85% power?
Explain your answer.
(1.5)
OUESTION 5.09 (2.00) is subcritical with a Keff of 0.95 and an SRM countrate of 200 l
The reactor Control rods are withdrawn and the new countrate is 400 eps.
Calculate how much reactivity was added?
SHOW YOUR WORK.
(1.5) cps.
l b.
Is the reactor now suberitical, critical, or supercritical? (0.5)
(
a.
l (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxx*)
rws v
Iz__IHE98Y_9E_HUGLEeB E9HE IHEBB001He51GE l
l GUESTION 5.10 (2.50)
Due to a personnel error during a surveillance, the plant scrammed after 60 days of high power operation.
preparations for restart are completed in four hours.
The Reactor (Nuclear) Engineer on call recommends doing a fast restart and pulling rods to the 100% rod line.
What is the problem with this recommendation?
(Include the effect a.
on the power to flow curve.)
(1.5)
The Nuclear Engineer calls back and wants to delay the restart long b.
enough to do a shutdown margin test for information.
How long would (1 0) you expect this delay to take to obtain meaningful data?
(1)
Two hours (2)
Four hours (3)
Ten hours (4)
Twenty-four hours (5)
Seventy-two hours GUESTION 5.11 (2.00)
Regarding the moderator temperature coefficient (MTC):
Provide Explain when this temperature coefficient can be positive.
a.
the core age and approximate moderator temperature.
(1.0)
Provide TWO reasons WHY the positive temperatere coefficient is not b.
a concern.
(1.0)
GUESTION 5 12 (1.00)
Explain what is meant by the term " rod shadowing'..
QUESTION 5.13 (1.50)
Heat transfer is primarily done by three methods:
Name them.
(1.0) a.
Heat transfer from the fuel cladding to the Reactor Ccolant is b.
primarily done by (1)
(.25)
, whereas, the heat transfer
.25)_________.
(
from the fuel to the cladding is done by (2) l (xxxxx CATEGORY 05 CONTINUED ON NEXT PAGE xxxxx)
raut o
52 IHEDBY_DE_HUGLE68 E9HEB_EkeHI_9EEBeIIDHt_ELW10Et_enu r_IHEBBQQ1H8HIGE QUESTION 5.14 (1 00)
From the H. P. Turbine exhaust to the L. P. Turbine inlete the gross anthalpy (1)_____________ (increases, decreasest remains the same) and the cpecific enthalpy (2)
(increases, decreases, remains the came).
(Choose either increases, decreases, or remains the same).
GUESTION 5.15 (1 00)
Concerning rod worth, indicate if the following will cause rod worth to increase, decreasee or remain the same.
o.
Core void fraction increases b.
Fuel temperature increases c.
Core agei BOL versus EOL d.
Rod height; a rod inserted from its fully withdrawn position, with no rod shadowing l
l
(***** END OF CATEGORY 05
- )
PECE 6
64..tL6HI.EIIIEHE_DESIGHz GQUIB9Lt eH9_INEIBU5ENI6IIDH OUESTION 6.01 (2.0Cl Describe how the drywell ventilation recirculation fans respond to a o.
Assume no Loss of Coolant Accident (LOCA) signal is present.
(1 0)
Explain the interlock between the drywell recirculation fans 1 and 3 b.
and the RHR pumps A and C.
Offsite power is not available and the LOCA signal has cleared.
(1.0)
GUESTION 6.02 (3.00)
Provide TWO reasons why the Standby Gas Treatment (SBGT) system is
)
a.
designed to start when primary containment pressure reaches approximately 2 psis during an accident.
(1.5)
State the purpose of the Standby Gas Treatment (SBGT) System b.
activated carbon filter bed and HOW and WHY it would be affected if the train heater were lost following an accident.
(1.5)
QUESTION 6.03 (2.00)
Upon receipt of alarm 3-A-9 ' Auto Blowdown Relief Valve Leaking,"
a.
list FIVE plant parameters (for example, vessel pressure) or indications (for example, pressure recorder on C05 panel) to be used to verify that a relief valve is leaking.
(1.25) b.
If the leak were confirmed to be severe, Procedure C.4.B.3.3 ' Relief Valve Failure' requires a manual scram and THREE system flows reduced or stopped to lisiit vessel cooldown rate?
Identify them.
(0.75)
GUESTION 6.04 (1 50) l The containment inerting systesi uses steam to heat the nitrogen.
a.
Identify THREE adverse effects that loss of this steam. could have on system / plant operation.
l I
l t
(xxxxx CATECORY 06 CONTINUED ON NEXT PAGE rrrxx)
PACE 7
62_ EL6HI_f1EIEUf_QESIEUt_GQUIBQLt 6HQ_IHEIBUHEHIBIIQH QUESTION 4.05 (2 00)
Indicate whether the following ere TRUE or FALSE.
The recirculation pumps are tripped at approximately 1153 psis in a.
~~
order to rapidly reduce reactor pressure.
For a given total jet pump flow, driv?'Y1ow typically exceeds driven b.
flow.
The recirculation pump speed is controlled by varying the voltage c.
applied to the pump motor via the Recirculation System Motor Generator set.
d.
Failure of the Recirculation System Loap (drive) flow instrument (FE-2-153A) at 100% power provides a rod block and 1/2 scram.
QUESTION 6.06 (1.00)
During evacuation of the control room it is desirable to leave the Reactor Mode Switch in RUN.
WHY?
QUESTION 6.07 (2 50)
How would the Intermediate Range Monitor (IRM) respond to a failure a.
of the negative 24-volt power supply fuse (s)?
(Two items required)
(1.5) b.
During plant startup, if a rod drop accident were to occur that caused a power spike, what instrument (s) provide plant protection besides the IRM's?
(1.0) 00ESTION 6.08 (1.00)
During a shift turnover following a feedwater transient, you notice that level recorder indicates level reached approximately + 3 inches.
the Discussion with the offgoing shift finds that a 1/2 scram was received in the
'B' channel and subsequently reset. Which one of the following is the The transient lasted less than 2 minutes total time.
appropriate action?
Request maintenance to check the calibration of the level recorders.
a.
level switches b.
Request maintenance to check the calibration of the that feed the Reactor Protection System.
Since the plant is stable, contact the Operations Superintendent as I
c.
time permits.
Make a los entry of what you noticed.
d.
Place RPS Channel
'A' in the tripped condition and initiate an orderly shutdown.
Contact the maintenance department to install a special level e.
recorder and slowly lower level to check recorder accuracy.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
' PACE 8
di._theHI_f15IEHE_QESIGHt_GQUIBQLt6HQ_INSIBudEHIGIIDH QUESTION 6 09 (2.00)
The Instrument Maintenance Department reports that all the Automatic Pressure Relief System Low-Low level actuation switches (LIS-2-3-672 A, Be have been found to be wired incorrectly.
The Operations Managere C
and D) directs the plant to be shutdown. What automatic system (s) when notifiede operating feature (s) have been lost?
GUESTION 6.10 (1 00) level The Main Steam Isolation valves are closed at -47' decreasing water (CHOOSE ONE) to:
Prevent having to close the valves later with no steam in the lines.
a.
To ensure that a reactor scram has occurred.
b.
To assure that steam is available for RCIC and HPCI to start.
c.
To assure that reactor water loss is minimized.
d.
QUESTION 6.11 (1.50)
What parameter (s) does(do) the Rod Horth Minimizer utilize to monitor a.
(.75) reactor power?
With the plant in hot standby with all the Group I rods fully withdraw b.
can the Rod Worth Minimizer be bypassed when directed to commence a power increase?
(.75)
GUESTION 6.12 (2.00) 125-volt DC battery Identify TWO adverse effects that loss of the til and its vital would have on the associated emergency diesel generator bus (415).
QUESTION 6.13 (1.00)
What hazardous substance would be of primary concern if the ventilation system to the station batteries was lost? (.50) a.
What type of problem would exist? (.50) b.
GUESTION 6.14 (1.50)
Describe the TWO major functions of the refueling interlocks that prevent inadvertent criticality during refueling operations.
CATEGORY 06 CONTINUED ON NEXT PAGE *****)
(*****
PACE 9
62.. EL6HI_IIIIEnt _QERIGu t _CQUIBQL t _eHQ _IN EIBUHEMIeIIQH QUESTION 6.15 (1.00)
At what position (OPEN or CLOSED) does the low flow feedwater regulating valve fail whenever the following occurst A. Loss of electrical control signal. (0.5)
B. Loss of control air. (75 psis)
(0.5)
(xxxxx END OF CATEGORY 06 **xxx)
PACE 10
' Zn_EBQGEQUBER_:_UQBuekt_etHQBuekt_EMEBGEUGI_eUE BeQIQLQEIGek_GQUIBQL
(
^
S (( M QUESTION 7.01
,, :n. ;, W mer & dese.nean n,
o IdET.ti y T!": i--- i=+= arrr:t... c t i
- 7. - r:,rir-d a*ine to evacuating the -- ^
' ---- --- "--- '"-- P 8 r-(2.0)
F1.0) b.
Describe the preferred evacuation path 4 JCCer7 yyt acWm i mfy Ib The $erni=e Stu?;pmJ 8%eL.
QUESTION 7 02 (1.00) o i
/.
While reviewing the operator's los on your shift, you notice all lights on the full core display so out.
The APRM's'are drifting slowly downscale.
A quick glance at the narrow range eva(1 meters shows them to be downscale.
The low level alarsi i up; Numerous other alarms start to cound.
o.
Based on the information given, what procedure should be referred to first?
Why?
Assume the plant was initially at 90% power.
QUESTION 7.03 (2.00)
For the following symptoms, seatch the appropriate initiating event:
a.
Rapid decrease in reactor power to less than 6% within seconds in conjuriction with an initial level decrease then increase.
b.
Reactor level increase, rapid reduction in core flow, loop flow-and reactor power; rapid decrease in main generator output and decrease in reactor pressure.
c.
Increase in drywell tesiperature and pressure, excessive drywell floor drain sus,p pump operation.
d.
Reactor Water Cleanup (RWCU) system isolates on high non-regen. heat exchanger outlet temp, RWCU posip bearing temperatures hight recire pusips low coolirig water flow.
Localized torus water heatup accompanied by a sudden small loss in f
e.
generator megawatts.
t l
Initiating Events.........
1.
Feed Pump Runout 6.
HPCI Injection 2.
Feed Posip Trip 7.
Loss of RBCCW 3.
Reactor Scram 8.
Steam Leak in Drywell 4.
Seiall Break LOCA 9.
Recirc Punip Runout 5.
Stuck Open SRV 10.
Recirc Pusips Trip OUESTION 7.04 (2.25)
Upon a loss of shutdown cooling, describe THREE alternate means of establishing a shutdown cooling flow path.
Include the pumps used and the heat sink.
(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE ***xx)
I
PCCE 11 N ZikEBDGEDUBE8_:_HOBuekt_etHQBuekt_EUEBGENGI eHQ BGRIOLOGIGeL_GDHIBOL l
i GUESTION 7.05 (2 50)
With the plant operating at 80% power, the High Pressure Coolant Injection system initiates.
(HPCI)Describe the INITIAL reactor response in terms of POWER and LEVEL.-
o.
(1 0) b.
Describe THREE immediate actions.
(Assume the plant has not scrammed) (1.5)
GUESTION 7.06 (3.20)
Following a Group I isolation, you notice that the control rods apparently Using the guidance of Operations, Manual C.5:
did not scram.
Identify THREE methods which should be used to attempt to scram e.
the reactor, per Procedure C-5-1100 series, " Emergency Operating Procedures." (1.2)
If all attempts to scram were unsuccessful, describe which systems b.
and actions should be used to mitigate the consequences of this event.
(Include the system (s) to be used for level control and the desired level range). (2.0)
GUESTION 7.07 (3.50) a.
List the immediate actions required by Procedure C.4.B.1.4.A,
' Trip of One Recirculation Pump,' when one recirculation pump trips f rom 100% power.
(1.5)
Identify FIVE CORE parameters that must be adjusted for single loop b.
(1.5) operation.
Procedures desire that an idle recirculation pump be restarted as soon c.possible. WHY? (0.5)
'i QUESTION 7 08 (2.50) i 8l18 While observing the start of tig ;;;; r,-ey mp operability test, you notice the minimum flow recircut....... JTTe is closed.
You contact the in plant operator to observe the pump and the local a.
indications for abnormal behavior.
List four things he should look for (problems with the pump only). (1 0)
After several minutes the control room operator recommends opening b.
the suppression pool test valve.
Should this be authorized?
Why?
Assume the local operator reports smoke coming from the pump casing.
(1.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. ~
PAGE 12 Z2...tBOGEQUBES_:_HGBuel1 6BNDBU6Lt_EDEBEEMEI 6HQ BeQIDLOGIG6L_G0HIB9L QUESTION 7.09 (1.50)
Regarding reactor water level:
The plant is scrammed at a decreasing level of approximatley +9'.
This is far above the top of the active fuel.
Why scram at this a.
level?
(0.75)
The recirculation pumps are tripped at -48' decreasing level.
Why?
b.
(0.75)
GUESTION 7 10 (2.00)
The A condition arises that requires entry into a HIGH RADIATION AREA.
You have operator will receive an estimated whole body dose of 80 mrea.
the following data available:
Candidates 1
2 3
4 Sex Male Male Female Male Age 27 38 24 20 Nk/ exposure 15 mrem 280 mrem 0
30 mrem GTR/ exposure 1210 mrem 1950 mrem 450 mrem 800 mrem 54730 mrem 5200 mrem 9950 mrem Life exposure 3 months Remarks History Unavailable Pregnant Each candidate is technically competent and physically capable of task.
Emergency limits do not apply and time constraints performing the do not permit obtaining authorization for an administrative exposure limit increase.
An NRC Form 4 is on file unless otherwise indicated.
Give your reasons for accepting or rejecting each candidate for the job.
l l
l I
OUESTION 7 11 (1.50) p#
he perations Manual C.4-Ar prohibits starting the mechanical vacuum pump f the following conditions exist:
He# n A ~0'ondenser vacuum is greater than 22 inches Hg.
B. High Main Steam Line radiation.
C. Indications of fuel failure.
State the reason for each of the conditions listed.
(marra END OF CATEGDRY 07 a****)
PccE as 94.. 6951HIIIB613YE_tB9GEDUBEtt_G9991139531.659.k1911011988 ouESTION 8 01 (3.00)
It was decided to Local Leak Rate Test (LLRT) the Main Steam Isolation Valves (MSIV's) during a short maintenance outage four months prior to the The outboard valve in the
'A' line was found to have ocheduled refueling.
gross leakage.
(1 0)
Under what circumstances could the plant resume operation?
c.
Identify TWO safety functions that would likely require review for b.
the need of a temporary system change.
(1 0) c.
- C ;F revel r:Uld -v=i likely be is vir;d Fri;r t0 r:: tert.
Z77 (1.0) 00ESTION 8.02 (1 00)
The plant has scrammed from a spurious Group I isolation.
During the subsequent pressure transient the Technical Specifications Limiting Condition for Operation (LCO) limits for Minimum Critical Power Ratio (MAPLHGR) and Maximum Average Planar Linear Heat Generation Rate (MCPR) 10 CFR 50.72 requires a red phone call for this event.
were exceeded.
WHY?
10 CFR 50.72 attached.
GUESTION 8.03 (2.00)
Due to vacations and required training, you have been asked to work the following schedule next week.
Review the schedule and identify the Technical Specification overtime guidelines that you would be violating.
Sunday 0700-1900 Monday 0700-1800 Tuesday 0700-2400 Wednesday 0700-1600 Thursday 0700-2000 Friday 0700-2000 Saturday 0700-1800 DUESTION P.04 (1.50)
Thr Reactor (Nuclear) Engineer reports that, due to minor problems with in calibration of the feed flow the procen computer and an error reactor power has actually been at approximately 1720 Mwt for instr uments,
most of the previous month.
What TWO immediate actions are required to be taken by Technical a.
Specifications?
(.75) b.
4DwFL e J ill C0huition must be satisfied prius is centinuins r
cmt i e..?
?.7M UPdehe.,
(sr**a CATEGORY 08 CONTINUED ON NEXT PAGE sumum) 1
-,--,,.n---,
,,--,,,_,_n-n
-.,.g.,,,.----..-_.-_.n
,,,,.--,,n.-------
PCCE 14
'. BAL._6DUINIEIB611ME_EBQGEDUBElt GQHDIIIDHit 6HQ.kIMII6IIDHE QUESTION 8.05 (1.00)
(Fill in the blanks)
A fire brigade of (1)__(0.5)________ members are required to be onsite, per 0.5)________
(
Technical Specifications, at all times.
This excludes (2) cembers of the operating shift organization. (1.0) 00ESTION 8 06
(.75)
What are the Shift Supervisor's THREE responsibilities for Work Request Authorizations (WRA's) as outlined in 4ACD-15.4, "WRA Processing.'
GUESTION 8 07 (2 00)
If the On-Duty Shift Supervisor becomes disabled as discussed in 4ACD-4.7, "Genere1 Plant Operating Procedures,*:
o.
Identify the position (s) and order of succession for immediate assumption of the Shift Supervisor duties and responsibilities. (1.0) b.
At least one, and possibly two, persons must be called and informed of the situation.
What are their titles?
(1.0)
QUESTION 8.08 (2.00)
Indicate whether the following statements are TRUE or FALSE regarding
' Equipment Control Procedures' as described in 4ACD-4.5:
a.
The control room operator is responsible for authorizing the repositioning of locked equipment.
b.
The Job Supervisor for whom a tag is installed is responsible for i
authorizing the installation of safety tags on plant operating l
equipment.
The Unsafe Card shall be used when it is essential that the I
c.
specified position be maintained to safeguard equipment.
d.
The Shift Supervisor is required to authorize a work control document to utilize previously installed safety tags.
Hold Cards need not be physically removed from equipment during e.
temporary removal f or testing.
l QUESTION 8.09 (1.25)
A quarterly surveillance becomes due on January 15th.
Due to an ongoing Limiting Condition for Operation, the surveillance is not completed until January 29th.
Due to similar problems, the next surveillance is completed on May 10th.
Is this acceptable?
Show your work.
(*zuzz CATEGORY 08 CONTINUED ON NEXT PAGE **xxx)
Er__eDHINIEIB8IIEE_EBDGEDUBEEz_GDHDIIIDHIz_6HD LIMII6IIDHf PACE 15 QUESTION 8.10 (1 50)
During repairs to the safety-related scram discharge volume vent valve, it eas necessary to install a spring retainer machined in the shop.
The eriginal cast retainer was found broken, causing the valve to fail open.
A Quality Assurance " Hold' was placed on the work package pending ongineering approval of the part substitution.
The valve was reassembled and functions perfectly.
Is the valve ' operable? ' WHY or WHY NOT?
OUESTION 8.11 (1 00)
A control room operator was issued an NRC Operating License on January 30, 1985.
His application for renewal was submitted on December 10, 1986.
As of February 20, 1987, he still had not received his renewal.
Is this individual violating the provisions of 10 CFR 55 by continuing to operate the plant with an expired license? (1.0) 00ESTION 8.12 (1.50)
While playing softball on Sunday afternoone you severely sprain your ankle.
Crutches are needed to report to work on Monday as the Shift Supervisor.
a.
Can you legally assume the shift? (0.5)
Why or Why Not?
(0.5) b.
Must a report be sent to the NRC?
(0 5)
DUESTION 8.13 (1.00)
With the plant operating at 100% power and the normal shift manning in the Control Room, the plant experiences a feedwater transient and scram.
Three NRC license examiners and the license candidates are in the Control Room in the middle of exams.
The examiners start directing questions to the candidates about the event.
What should the candidates do to comply with 4ACD-4.7 ' General Plant Operating Activities"?
(Choose the appropriate answer) a.
Continue to answer questions because the NRC has unrestricted access to the Control Room.
b.
Continue to answer questions but try to stay out of the operator's way.
Inform the NRC examiners that they must leave the Control Room.
c.
d.
Ask the NRC examiners if they would like to leave the Control Room.
(****x CATEGORY 08 CONTINUED ON NEXT PAGE
- )
PC.GE 16 921.69BINIBIB6IIVE tB9GEDUBEft_GOURIIIQuit_6HQ_LIBII6IIQUE QUESTION 8.14 (2.00)
The
'A' Main Steam Line Radiation Monitor has failed upscale.
A jumper is According to required to be installed to eliminate the half scram.
' Bypass Contro1*:
Procedure 4ACD-4.8, What minimum reviews must be performed and by whom prior to o.
installation of the jumper?
(1.0)
When is Assume the planned installation has been properly reviewed.
b.
(1.0) a second verification of the installation Not Required?
DUESTION 8.15 (2.50)
AccordinS to Procedure 4ACD-4.7, " General Plant Operating Activities,'
with proper authorization, licensed operators can knowingly deviate from License Conditions or Technical Specifications Identify the THREE conditions that must be satisfied before the o.
deviation is allowed?
(1.5) b.
What TWO individuals can give " proper authorization'?
(1.0)
DUESTION 8.16 (1.00)
Following a significant event, you are ready to make a red phone call via (ENS).
The red phone is dead when you the Emergency Notification Systemlong distance lines to the NRC Operations pick up the receiver.
The Center in Bethesda, Maryland, are busy and stay busy.
What alternate method may be used to make the required notification.? (1.0)
I (zzzra END OF CATEGORY 08 mmzzz)
END OF EXAMINATION ***************)
(*************
I,quAlaun ami.e l*,
Dele effittency o Ctmort Ya se o o s/t eWL)/(EnsfSFle) s = bt + 5 att w. es, E. act A. All A = %p-At e = (V - %)/L EE. 5 avt t
PE och a = ant /tg. o.493/tg
= = e/t vg g + st tgeff. [(tg) (t )3 b
W-f I(t ) + (t )3 t
t
.g, af = 931 as i=sCpat 4
I = lot h = UAal I = 1,10-x/TVL Pvr = N ah TYL = 1.3h f
W L = -0.693A P = Po 0" U 1
P = P et/T o
SCR = 5/(1 - K,ff)
SUR = 26.06/T CRx = 5/(1 - Keffx) l CA:(1 - Keffi,) = CR (1 - keff2) 2 SUR = 26p/s* + ( s-p)T M = 1/(1 - Keff) = CRt/CRo T = (**/,) + [(a - p)/ip)
M = (1 - % fro)/(1 - Keff t) l T = a/(p - s) sw. = (1 - r,ff)/Keff
. T = (s - e)/(ae)
A* = 10-5 seconds a = (Keff-1)/Keff = AKeff/Keff i = 0.1 seconds *1 e = [(a /(T Reff)) + [s ff;/(1 + IT))
e
'It
= 12 2 d
d 2
2=3d22 l
lidt P = (reV)/(3 x 1010) 2 R/hr = (0.5 CE)/d (meters)
I =.N R/hr = 6 CE/d2 (feet)
Miscellaneous Conversions l
Water Parameters 10 ps d
1 curie = 3.7 x 10 1 gal. = 8.345 lbs.
1 kg = 2.21 lhe 3 1 gal. = 3.78 liters 1 hp = 2.54 x 10 Stu/hr 6
1 ft3 = 7.48 gali.'
1 er = 3.41 x 10 Stu/hr Density = 62.4 ths/ft 1 in = 2.54 ce Density = 1 gn/cm3
- F = 9/5'C + 32 l
Heat of vaporization = 970 Stu/lbe
- C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lbe 1 STU = 778 ft-1bf 1 Ata = 14.7 psi = 29.9 in. Ng.
1 ft H O = 0.433 lbf/in!
2
' t,
- "Y
,=, _=.
=
=.w g,
WI M F U) r.
Al & lL I 5diI 5
~
~
//h ThBLn JRii
~
IMLW i%!ftl%
~
lh WiM lRWi 4L 1
~
llMMUunt U ~;
~~
IM/2V8 Ni40
~
//hQf14% 'ir%'JJ
~
i JhDW. 'idi t /r _i
~
M
'73 I
h"%W$W/D f~
W 0
/ 4 B
~
//BHM
/
lW7W14 /%WD
~
- 1M%'7MW//7
~
WWhWAVL D'
~
MMMHY
~
MXM7x'HY
~
D'):MWM
"'t W M Of( M
.. z,
-=-
._A.
ma
-.r--
_a.
9 9
34
.o e
3 A< DN@o0MC ;,
NNEN MGWC 2w? sM3<
W G4 < 5NM&d; NNx%>MQM&;
Scc 9<%NM%M; e
Nd<GM% MEA ;
~
NC><
Q(SRWW;1
^
C>GEMRh
~
> <D
=-=can:gNg9,i w
m
- zgss==s x 3,
~ ~
m q_~
~
k YC 7
-=====4 l
322 2
==ume
=====saissiisiiiiig l
7====r'EEE!np N-
-~~--
D I
_@E5===== =ssis p
=
=
4 l
i 1,
~,..%.,_
I{
l m
e4 j
etantoo l_
I_ I I
I I
I I
I I
g i
1
~~--.d-.-
g
- T*e
H gema/ftenestemmesersense
~'
Shenhe e Austna sleem WMus and 941 hous been eskuess tem 7hemmenomk and 7mupet senswees d semen gospywsm.1987, ty the 8sertson essesty er sammenises egneers4 TeMs1 Fmporties of saturated steam and estersted water (temperature) l Volume,ft'/m Enthalpy.Stuft Entropy.Stu/h a F i
asser
- Eve, Steam Water Ever Steam Water Euer Steam fT 5
m em859 eAla02 3305 3305
-o.ca 10n.5 10n3 0.0000 2.18n 2.18n m
l 1
l N
029991 Salt 02 2948 2948 3.00 1073A 10762 0.0061 2.1706 2.1767 5
40 0.12163 Ome02 2446 2446 em 1071D 1079.0 0.0162 2.1432 2.1594 40 48 0.14744 OA1402 2037.7 2037A 13.04 106d.1 1001.2 0.0262 2.1164 2.1426 48 to 0.17796 OA1602 1704A 1704A 18 05 1065.3 1083.4 04361 2.0901 2.1262 to 80 0.2561 SA1403 1207.6 1207.6 38 06 1059.7 1087.7 0A555 2.0391 2.0946 80 j
35 03629 OA1605 068 3 068 4 30.05 1054.0 1092.1 0.0745 1.9900 2.0645 70 80 0.5068 021407 633.3 633J 48.04 1048.4 1096 4 0.0932 1A426 2.0359 to 90 0 6961 0A1610 468.1 468.1 58 02 1042.7 1100.8 0.1115 1A970 2.0086 to 100 03492 OA1613 350.4 350.4 48.00 1037.1 1105.1 0.1295 12530 1.9825 100 110 1.2750 0A1617 265.4 265.4 77.98 1031.4 1109.3 0.1472 12105 1.9577 lie i
i l
120 1.6927 0A1620 203.25 203.26 87.97 1025.6 1113.6 0.1646 1.7993 1.9339 120 l
130 2.2230 0A1625 157J2 157J3 97.96 1019.8 1117A 0.1817 1.7295 1.9112 130 1
140 23892 0A1629 122.96 123.00 107.95 1014.0 1122.0 0.1985 1.W10 1A095 140 150 3.718 021634 97.05 97.07 117.95 2006.2 1126.1 0.2150 1A636 1.9686 100 l
140 4.741 0A1640 77.27 77.29 127.96 1002.2 1130.2 0.2313 1A174 1.8487 100 I
170 5.991 021645 62 04 62.06 137.97 996.2 1134.2 0.2473 1.5822 13295 170 100 7.511 021651 50.21 50.22 148 00 990.2 1138.2 0J631 1.5400 1A111 180 190 9.340 0 01657 40.94 40.96 158.04 984.1 1142.1 0.2787 13148 1.7934 190 300 11.526 OA1664 33.62 33.64 168 09 977.9 1146.0 0.2940 1.4824 1.7764 300 210 14.123 CA1671 27.80 27A2 178.15 971.6 1149.7 0.3091 1.4509 1.7600 210 l
212 14.696 0A1672 26.78 26to 180.17 9703 1150.5 0.3121 1.4447 1.7568 212 l
320 17.186 0A1678 23.13 23.15 188.23 966.2 1153.4 0.3241 1.4201 1.7442 230 j
i 330 20.779 OA1685 19.364 19.381 196.33 958.7 1157.1 0.3388 1J902 1.7290 330 i
340 24.968 0A1693 16.304 16321 20845 952.1 1160.6 03533 1.3609 1.7142 340 Me 29225 0A1701 13302 13A19 218.59 945.4 1164.0 0.3677 1.3323 1.7000 ISO
]
i 300 35.427 021709 11.745 11.762 228.76 938.6 1167.4 0.3819 1J043 1.6862 200 I
270 41A56 0A1718 10.042 10.060 238 95 931.7 1170 6 03960 1.2769 1.6729 270 l
MD 49.200 0 41726 8.627 8.644 249.17 924.6 1173.8 0.4098 1.2501 1.6599 380 I
200 57.550 0A1736 7.443 7.460 259.4 917.4 1176.8 0.4236 1.2238 1.6473 290 1
300 47.005 0A1745 6.446 6.466 269.7 910.0 1179.7 0.4372 1.1979 1.6351 300 j
]
S10 77.67 021755 5.609 5.626 230.0 902.5 1182.5 0.4506 1.1726 1.6232 310 1
830 99 64 OA1766 4A96 4.914 2904 894.8 1185.2 0 4640 1.1477 1.6116 320 t
340 117.99 OA1787 3.770 3.738 311J 878.8 1190.1 0.4902 1A990 1.5892 340 1
350 153.01 SA1811 2.939 2.957 332.3 362.1 1194.4 0.5161 1A517 1.5678 380 NO 195.73 0A1636 2J17 2.335 353.6 844.5 1198.0 0.5416 12057 1.5473 300 400 247J6 90 W 13444 1A630 375.1 325.9 1201.0 0.5667 01607 1.5274 400 480 308 78 L ag94 1.4808 1.4997 396.9 306.2 1203.1 03915 03165 15080 430 440 351.54 021926 1.1976 1.2169 419.0 785.4 1204.4 0.6161 02729 1.4890 440 400 466.9 0A196 0.9746 0.9942 441.5 763.2 1204.8 0.6405 0 3299 1.4704 440 400 9662 0m00 0.7972 OA172 464.5 739.6 1204.1 0.6648 0.7871 1.4518 400 000 sec.9 0204 0.6545 0.6749 487.9 714.3 1202.2 OA890 0.7443 1.4333 900 S30 812.5 0509 0.5386 0.5596 Sith 487.0 1199.0 0.7133 0.7013 1.4146 830 S40 9624 0A215 0.4437 0.4651 536.8 457.5 11943 0.H78 0.6577 13954 540 900 11334 0m21 0.3651 0.3871 962.4 625.3 1187.7 0.7625 0.6132 1.3757 SGC 900 13262 SA228 0.2994 0.3222 589.1 589.9 1179.0 0.7876 03673 IJ550 S00 000 15432 SA236 0.2438 0.2675 417.1 550.6 1167.7 OA134 0 5195 1J330 000 480 1786 9 0A247 0.1962 0.2208 646.9 906.3 1153.2 03403 04689 1.3092 430 440 2059.9 0A260 0.1543 0.1002 679.1 454.5 1133.7 02686 04134 1.2821 640 000 23651 0m77 0.1146 0.1443 714.9 392.1 1107.0 0A995 03502 1.2498 860 30 2708 6 SA304 0.0008 0.1112 758.5 310.1 1068.5 0.9365 0J720 1.2006 800 300 30943 Sand 6 02386 0D752 322.4 172.7 995.2 0.9901 0.1490 1.1390 700 33082 48508 0
0.0508 906.0 0
906.0 1.0612 0
1.0612 206.
34 Table 3 l
Properties of estursted steem and estursted water (pressure) sseelpy.em snee9y.s e ae sm.'ar.ewt g
i tuwn n9&
teme mesm wenn s, m
wenn ant seem ein, mm l
y
y' weew a,
a.
s 5
s 2,
!l Egens 32438 SS1602 3302A 3302.4 6 00 10753 1975.5 8
3.1872 2.1872 e
le1J tm Ett 35.023 0A1602 2945.5 2945.5 3.03 1073A 10764 00061 2.1706 2.1796 BAB 1022J E30 Ell 45 453 DD1602 2004.7 2004.7 1830 1067.9 1 3 1.4 0.0271 2.1140 2.1411 13.50 1025.7 AJS ff I
E30 53 140 0A1403 16MJ 35NJ 2122 1063.5 154.7 0.0422 2.0730 2.1140 2122 10283 tag
& 30 44 484 0.01604 1039.7 1039.7 32.54 1057.1 1 2 9.7 0.0641 2A148 2.509 32.54 1052.0 4 30 See 72A00 041806 792A 792.1 40 92 1052.4 10933 00799 1.9762 2A562 40.92 1054.
te al 79.586 001607 M1.5 441.5 47A2 1048 6 1096 3 00825 1.9446 2.0370 47A2 1036.9
&S E6 85218 041609 540 4 M O.1 5325 1045.5 1098.7 0.1028 13186 2 2 15 E324 1038.7 Es S.7 to 08 0.01610 446 93 466 94 54 10 1042.7 11008 OJ 1A966 2.0083 SS 10 1040.3
&7 OA 94.N 001611 411.67 411.69 62.39 1040.3 1102.6 0 1817 1A7M 1.9970 62J9 1041.7 GA ES 9524 001612 364.41 368 43 66 24 1038.1 1104.3 0.1264 1A606 1.9570 4624 1042.9 ES 1.0 101.74 0 01614 333 59 333 60 69 73 1036.1 1105.8 0.1326 1A455 13781 69 73 1044.1 la 3A 126 07 0 01623 173.74 173.76 94 03 1022.1 1116 2 0.1750 1.7450 13200 94 03 1051A 3.0 3.0 141A7 0.01630 118 71 118.73 109 42 1013.2 1122 6 02009 1.6854 1AS64 109 41 1056.7 3.0 4A 152.96 001636 90A3 90 64 120 92 1006 4 1127.3 02199 1.6428 13626 120 90 1060.2 48 i
10 16224 0 41641 73.515 7333 13020 1000 9 1131.1 02349 1.6094 13443 130 18 1063.1 80 10 170.05 0.01645 61367 61.98 19803 996.2 1134.2 02474 1.5420 1 A294 1N 01 1065.4 40 I
l t
7A 176 84 0 01649 53 634 53 45 144 83 992.1 1136.9 02581 1.5587 13168 14431 1067.4 7A I'
E0 18246 0 01653 47.328 47J5 150A7 988 5 11393 02676 1.5384 13060 15034 1069.2 SA 9.0 15 27 0 016M 42385 42.40 156.30 985 1 1141 4 02740 1.5204 1.7964 IM 28 10704 9A 10 19321 0 01659 as 404 38 42 161.26 982.1 1143J 02836 1.5043 1.7879 16123 1072J N
14.006 212.00 0.01672 26 782 26A0 180 17 970.3 1150.5 0J121 1A447 1.7MS 180 12 1077.6 14.9 5 15 21303 0 01673 26.274 26.29 101.21 969 7 1150.9 0.3137 1.4415 1.M52 181.16 1077.9 18
/
30 227.96 0 01643 20 070 20 087 196 27 960 1 11 % 3 0.3358 1J962 1.7320 196 21 1082.0 30 I
30 250 34 0 01701 13 7266 13.744 218 9 945 2 1164.1 03642 1J313 1.6995 218 8 1087.9 30 40 N7.25 0.01715 10 4794 10 497 2M1 933 6 11698 03921 1.2844 1A765 2M0 1092.1 40
[
30 mim 0 01727 84967 8.514 250.2 923.9 1174.1 0 4112 12474 1.6546 250.1 1095.3 to 40 292.71 0 01738 7.1%2 7.174 M2.2 915.4 1177.4 0 4273 1J167 1A440 262.0 1098.0 80 70 3D233 0.01748 6.1875 6205 272.7 907A 1180.6 0 4411 1.1905 14316 2723 1100.2 70 to 312.04 0.0lM7 5.45M 5 471 282.1 900 9 1183.1 0.4534 1.1675 1A208 N1.9 1102.1 80 to 320J8 0 01766 4AIT7 4395 290.7 894 6 1185.3 0 4643 1.1470 1A113 290 4 1103.7 to 100 st?A2 0 01774 4.4133 4A31 298 5 808 6 11&P.2 0 4743 1.1294 1A027 298 2 1105.2 100 130 341.27 0.01789 3.7097 3.728 312 6 877A 1190.4 0 4919 1.0960 1.5879 312.2 1107.6 130 140 353 04 0.01803 32010 3219 325.0 868 0 1193 0 0.5071 1.0681 15752 324 5 1109.6 140 140 363 55 0 41815 2A155 2A34 3M 1 859.0 1195.1 0.5206 1.0435 13641 335 5 1111.2 300 130 373.0B 0 01427 23129 2 531 346.2 850.7 1196.9 0 5328 14215 13543 345 6 1112.5 180
+
300 3B1A0 0 01439 2.2689 2287 355.5 842A 1198.3 0 H38 1.0016 15454 354A 1113.7 300 350 400 97 0.01865 1A245 13432 3761 825 0 1201.1 0 5679 0 9585 3.5264 375J 1115A 350 300 417.35 0 01889 15238 1.5427 394.0 808.9 1202.9 03882 0.9223 13105 392.9 1117.2 300 350 431.73 001913 IJ064 1.3255 4098 794.2 1204.0 0 8059 0 4909 1A968 4086 1118.1 350 400 444 40 0 0193 1.14162 1.1610 424.2 780 4 1204.6 06217 0400 1AS47 422.7 1114.7 400 400 456JS 0.0195 1Al224 3A318 437.3 M7.5 1204A 06360 03378 1.4738 435.7 1118.9 480 t
900 467Al 00198 090787 0.9276 449 5 755.1 1204.7 0 6490 0 3144 1A439 447.7litt S 900 960 4M to 00199 CA2183 03418 460.9 743.3 1204.3 06611 0.7936 1.4547 458 9 1118 000 000 dB6JO 00201 074962 0.7698 471.7 732 0 1203.7 0 6723 0.7738 1A461 4695 11182 GOS 700 503 OB 00205 063505 0 65 % 491.6 710.2 1701A 0 6928 0.7377 1A304 488 9 1114.9 780 300 518 21 00209 034809 0.5690 509A G89.6 1199 4 0.7111 0.7051 1A163 506 7 1115 800 900 531.95 0 0212 047968 0.5009 526.7 669.7 !!96 4 0.7279 0.6753 1A032 523 2 11130 900 1000 544.58 0 0216 042436 0 4460 542 6 650 4 1192 9 0.74M 0 6476 1 A910 SSB 6 1110.4 1000 1100 566 3 0 0220 SJ7863 04006 557.5 431.5 1189.1 0.M78 0 6216
- 1J794 5531 1107.5 110C 3300 567.19 0 0223 SJ4013 0.3625 571.9 613 0 1884 8 07714 0.5969 1A683 566 9 1104.3 IS 1300 577At 0.0227 OJ0722 0.3299 585.6 5946 11802 0 7843 0 5733 1 3577 580.1 11009 130C.
1400 587.07 00231 027871 0.3018 598 8 576.5 llM 3 0.7966 0 5507 13474 592 9 1097.1 140f l 1500 59620 0.0235 025372 02772 611.7 558 4 1170 1 0 4085 0 5288 1J373 405 2 1993 1 190s :
3000 63530 0 0257 016266 0 1883 672.1 466 2 11383 0 4625 0 42 % 1.2001 862 6 106 3500 668 11 0.0286 010209 0 1307 731.7 361.6 1093.3 09139 0J206 12345 Fit 5 1032 4
3000 405J3 0.0343 OA5073 0.0850 301A 218 4 1020.3 0 9728 Elett 1.1619 782 3 973.1 300 330 & 3 706 47 0.0508 0
0 060s 906.0 0
9060 1A612 0
1.0612 - 875 9 SM9 330 C -m
34,.
gg d resowve pospween at suportested senem and esmereened moto
.'J~
m Weempp ass ano ano eso eso ano too suo too soon slao asco asco taso asas 4RIA tilS 671A St.1 000.7 l
1384 4 00 11902 31M.7 3M1A 13 ABS 1336.1235 S.1162 2.1722 161.M 173A6 38575 197.70 3262 221 61 J M2.5 v 801 3
4 5 1639A 18933 37480 18033 901.74) s 0.1296 12615 13B 05 15001 1506 7 3 84 RJ1M 2Jt09 2 2 11 2A101 9024 80224 11421 1335.9 1884J 14336 353J S&MJ l
1IMA 8241J 1388233369 13943 2De40 2AB32 2.13
, OA181 35.14 StM S7.04 S AB 9920 74.98 '80 94 86 9 5
6 501 184&A 1
1639 5 1893J 1747.9 1803.4
'~~
06224) e 0.1295 1A744 1384 0 1433 4 1483 5 15346 19866744 IJoe4 2A337 12406 1287A 1835.51.9692 24146 E.0603 2.1011 e 0A141 MA4 44SB 33 a 302 114L6 IIt3 7 41305 Matt Walt 73A13 77A07 3 3593 13173 15345 15065 1639.4 IW32 17473 1 29299 33963 37.985 41386 45978 49.
0521) e 0.129511.75BB l
13352 1383 4 14332 1443 4 e 00163 SS146 t 15 4 See 15A9 1192.5 1239.9 1297J 1A720 13242 3435 46420 49 405 52.3BB 96370 6 1A134 1639J 1983.1 1747A 13033 013 03) s 03295 0 3 4022J% 25428 28457 31466 34 465 37A5a 40447 4 1534.3 1546 32.0991 2.1336 2.14 so 4 5 05 l e 11 1191 4 1239.2 1286 9 1334 9 13835 1432 9 1483.2 l
e 0A161 SA146 01295 0 2940 1.7805 1A397 1A921 13397 1.9836 2 4244 2A628 24 489 26143 27.676 2918E l
14 165 15 685 17.195 18 699 20 199 21 H7 23 194 1585.8 1638 8 1992.7 1747.5 18 Q27.96) e 14321 1482 5 1533 7 2A%9 2.0599 2.1224 2.1516 l
e 00161 CAIM 11.036 12424 1285.0 1333 6 1982.51.9065 1.M76 13860 2.0224 s
12M 4 4810 19 15 11M 6 1.Mos 1A143 13624 16 450 17A48 la 445 19 441 40 4 12 446 13 450 14.452 15 452 g67.25) s 0.1395 0 3 40 1A992 1585J 16384 1492 4 1747.1 1802 l
00161 RAIM 7257 8354 9 400 to 425 11 438 2A120 2.0450 24765 2.1068 233 515 Im20 1181 6 1233.5 1283.2 1832.3 1381.5 143 1.7134 1.M81 tales IA612 1.9024 1.M10 13U4 3
12331 13001 13 A29 145U e
80 6 0.1295 4 2139 i1 4492 9 319 10.075 10 329 11381 1638 0 1892.0 1746 2 18 9 92.71) s 7A18 7.794 8.560 15326 1944912500 2A131 2Ae46 2 G421 34524 M9 74 1230.5 1281J 1330 9 1380 5 1430 00161 ED196 00175!
6218
.6790 1.7349 1.7S42 IA289 12702 1.9009 13454 9
9A60 10400 11.060 11 A5 e
S 1295 0 3 39 0 4371 1 30 6 4135 550E 6 216 6 833 7A43 8 050 8 655 9258 18022 IG74 34916 17465 0 12.04) s 1279.3 1329 6 1379 5 1429 7 1480 4 1132.0 1584 13e83 24199 2.0502 2A79 4
IA036 13451 1A439 12205 1.9552 60161 E0144 00175 1227.4 l
100 t. S 26 184 29 M9 77 9 713:
e 14516 1.70ns 13546 6 1928 6 7006 7.2060 7.7096 8 211 027A2) s 0 1295 0 2 939 04371 I
4 0786 4 6341 5.1637 $6831 2
13640 1.9996 2.0300 2 05t 120 6 EJ1 364J3 26981 1224 1 1277.4 1328 1 13784 14288 1479 8 1531 4
@4127) e 4.1295 8 2939 04371 1 6286 14872 1J376 1.7829 1A246 1 A635 1300 08161 6 0166 0 0175 33 e
6 1709.6 4036 7 2 49 7 4652 7 3 1220 8 1275J 13268 1377.4 1428 0 395M 44119 43545 SJ995 57364 1.9525 2 4129 2 A421 3.4661 12071 13461 1526 13176 1350B 60161 6 4146 0.0175 18085 1A446 1.7196 13652 54132 53945 5.n41 4.1522 45293 11 SJ7 35 38 26925 e
140 6 1490 5 1745 6 18014 E1295 8339 04370 053.04) e 3 0060 3 4413 3 8480 4 2420 4 6295 S 42 344 42 M939 1217.4 1273J 1325 4 1376 4 1427.2 1475 4 73 13906 1A522 1J039 1.7499 1.7919 1 A310 1A6 4D161 50166 00175 e
6 1363 64704 4 4505 47907 51299 5 4657 5A014 4
300 1477.7 1529 7 1542.4 14359 30 a 812M S2938 04370 3.0433 3 4093 3 7421 4.1 5 4 (363 55) 1.9649 2 A142 1 Al?6 1 A545 1 ABM 13227 1.9545 8 47 19 47 M9 92 1213 3 12712 1324 0 1375J 1426 3 2 6474 80141 R O1 M 0 0174 14743 1 A376 14900 1J362 1.U34 53209 5A219 e
40008 430U 4A125 4$165 82191 6
nao e A12M S 3 38 04370
& W 52 368 51 269 96 1210.1 IM9 0 2.3598 21247 3 05E3 3 3783 3 4915 1 A776 13109 1.M27 13732 0 73.00) e 40161 041M 00174 14242 14776 11239 IJG63 1A057 13426 131 4A546,
e 412M 4338 04M9 l13593 2 4662 2.4872 2.M10 31909 3 4382 3 300 1423 4 14753 15274 1540 4 IO4 08130) 137M :
1.4976 11405 1J801 1A173 1A524 1 e 40181 041M OA174 0 0186 ! 2.1504IM3 5 1319.0 13714 15E3 270A5 3M.10 1.5951 1 4502 36746 BA W 350 4 5 46 A2937 04368 0.M47 2.2263 24407 26509 23545 30643 SJ6A 15794 1633J 168E0 1743 (400 97) s E1294 1A317 1A652 32972 13278 6 48 79 18874 27014 375.15 1257J 1315.2 1964 9 1421J 1473 6 15M2 1Je45 2.0044 e 00161 0Clu 00174 00186 1.M93 13964 1AMS 1J192 0
13970 2 0R32 22652 2 4445 2421 13703 16274
- 5) e 0.12M S2937 04307 05445 300 1578 2 16323 1487.1 1742 (417J 13105 1340 e s t2 188 85 27024 375.21 1251 5 1311.4 1966 2 1419 2 14713 1524 7 00186 1A913 1J028 1.5483 1.0077 1A571 13009 1.7411 1J7 e 00161 64146 00174 23515 2.902 1DM9 2.1339 22901 2A450 23947
- 3) e 01253 029M 04367 0.5664 330 1741.9 1790 1523.3 15749 16312 14A62 1 A955 IJB41 1.4763 1 4499 13151 (431.7 1245.1 1307.4 1863 4 1417 4 14701 M897 27033 37527 1.5252 1.S901 1A406 1.4450 1J255 136 o 0A161 &0144 04174 0A162 132M 13507 20744 21977 a 3 05 401s 41293 02935 04306 0M63 ago 001M 09919 1.15M 13037 1 4397 13708 1 6992 1629 3 1484 4 17403 1798, t a 3.32 309 19 270 51 SM.38 12312 1299.1 1.n30 1A059 1Apl3 13702 1A9 (444 o 00161 001M OA174 1.5595 1A123 14678 1 A990 1J371 D1)s ;&latt SJtM 043M 0.9040 l1A021 ge (447 b
i m+
N tema/florestrasades etsamen
~
TaWe3 properties of superbested steem and r,.;:M water tiemperature and pressure)
T*ugershme F j
Abs poem.
t/sq km, gest,9say) 300 300 300 400 900 000 700 MO 900 1000 1100 120 1MID 1400 tam r
, Galet SAIM 0 4174 0 0186 03944 OM% 1A726 1.1992 1300B 1A083 1A180 1Atll 1.7252 1A294 1A30s age a es3s 100 42 270 70 3M 49 1215.9 1290.3 1351 2 140s.3 3463 4 1517.4 15713 1627A 3GR2 6 173BA 17964 4406J0) e 41292 02913 0 4%2 03667 1AHO 15329 1.5444 14351 1A799 1.71 % 1J517 1.7859 1A144 134M 13732
, &D161 E01M 00174 0 4106 0 0204 01928 0 9072 12102 1.1078 IJ023 12948 13B58 1 A757 1.5647 1530 7e0 e 43A4 18B M 27029 375 61 487.93 1281A 1545 6 14037 1459 4 1514 4 1569 4 1424 2 19I07 1737.2 1794J (503.0s) e E1291 02932 0 4%0 0 %55 0.6a89 1.5090 1.%73 1 A154 1A600 1A970 1J335 1J679 13006 12318 1A617 ED161 0.0lM 0.0174 0 0186 0 0204 06U4 0 7828 OEM9 0.9631 18470 1.1339 12093 IJEB5 1.3669 1 A446 e
see a 70.11 14928 273.07 375 73 487AB 1271.1 1139.2 1999.1 14553 1511 4 1%69 16223 1678 9 1735.0 17922 (51821) e 0 1290 02930 0 4358 0 % 52 0 6845 1A469 1.5444 15900 1 A413 1A007 1.71M 11522 13B51 1A164 1A464 e
CD161 0 0166 0 0174 0.0186 0.0204 0 5069 0.6858 0 7713 0 2504 0 9262 0 9998 1A720 1.1430 1.2131 1JB25 900 t 7037 170 10 27126 37524 487 53 1260 6 1332.7 1394 4 14522 15083 1564 4 1620 6 16U.1 1734.1 17914 (531.95) e 01290 OJ929 0 4357 0 5649 0 6881 1A659 15311 1.5422 1A263 1.6M2 13033 1.7382 IJ713 13028 13329 e 0 4161 0 0166 0 0174 0.0186 0 0204 0 5137 0 6060 0 6875 0.7603 0 8295 02966 0.9622 1.0766 1.0901 1.1529 1000 4 70 63 17033 211 44 3M 96 487J9 1249 3 1325 9 1389 6 1448 5 1504 4 1%1.9 16184 1675J 1732 5 179CJ (54458) e 0 1289 0 2925 0 4355 0 % 47 06876 1 4457 1.5149 1.%7) 1.6126 1 6530 1.6905 132 % IJ5B9 13905 12207 e
00161 0.01M C0174 0 0185 0 0203 04531 0 5440 0 6188 0 6865 0 7505 02121 0 3723 0.9313 0.9894 IDe68 1100 4 70 90 170 56 271.63 376 08 487.75 12373 1318.8 1384 7 1444.7 1502 4 1559 4 1616.3 1673 5 1731.0 1789D (5%28) e 0 1289 0 2927 0 4353 0.% 44 06872 1.4259 1A996 15542 14000 14410 1.6787 1.7141 1.7475 1.U93 13097 e
00161 0 01M C0174 0 01B5 0 0203 0 4016 0.4905 0 5615 0 6250 0 6M5 0 7418 0 1974 02519 09055 0994 1200 4 71 16 17015 77132 37620 447.72 1224.2 1311.5 1379 7 1440 9 1449 4 15569 1614.2 16714 1729 4 1757A (567.19) e 0 1298 02926 0 4351 0 % 42 0 6668 1 4061 1.4851 15415 13883 14296 14679 1.7035 17371 13691 1.7996 e
00161 0 01H 00174 0 0185 0 0203 03176 0 4059 0 4712 0 5282 0.5809 0 6311 0.6798 01272 0 7737 0A195 1400 6 71 68 171.24 272 19 376 44 487.65 11941 1296 1 1369.3 14332 14932 15513 1609 9 1668 0 17263 1781D (587J07) e 0 1267 0 2923 04348 0506 0 6459 13652 1 4575 13182 1.%70 1 6096 1.64M 1.6H5 IJ185 1.7508 11515 i
e 00161 0 01M 0.0173 0 0185 0 0202 0 0236 0 3415 0 4032 0 4555 0 5031 0 5482 0 5915 0 0 36 0 6748 0 7153 1
1600 4 7221 171 69 272.57 376 69 487.60 616 77 1279 4 13585 1425 2 1484 9 1544 6 1605 6 1%4.3 17232 1712 3 (60437) e 0 1286 02921 0 4344 0 % 31 0 6851 03129 1.4312 1.4968 1.5478 15916 1.6312 1.M 78 1.7022 1.7344 1.M57 e
0 0160 0 0165 0 0173 0 0185 0 0202 0 0235 0 2906 0.3500 0 3988 04426 0406 05229 0 %09 0 5960 0 6343 1800 6 72 73 172.15 272.95 376 93 487.% 615 58 1261.1 IM7.2 1417.1 14806 1541.1 16012 IMO7 17201 1D97 (62122) e 01284 CJ918 0 4341 0.% 26 0 6643 0A109 1.4054 1 4768 15302 1.5753 1.61 % 14528 1.8876 1.7204 1.7516 o
CD160 00165 00173 CDlp 0.0201 0 0233 02488 03072 03534 03942 04320 04680 E5027 0 $M5 0.M95 3000 e 7326 172 60 27332 377.19 487.53 614 48 1240 9 1353 4 1408 7 1447.1 1536 2 1596 9 1657.0 1717.0 1777.1 f
.(E3520)e R1213 02916 0 4337 05621 0 6434 02091 1.3794 1.4578 1.5138 13603 14014 11391 14743 13075 113B9 e
CD160 00165 0 0173 0 0184 0 0200 0 0230 0 1681 0 2293 02712 03068 0.1390 0 3692 0 3980 0 4259 0 4529 I
I 2500 4 74 57 173 74 274.27 37722 487.50 612 08 1176 7 1303 4 1386 7 1457.5 1522.9 15859 16473 17092 Ino4 (M8.11) e 0 1200 0 2910 0 4329 05409 0 6815 0 2048 1.3076 1.4129 1.47M 15269 1.5703 1A094 14456 1 6796 12116 e
CD160 0 0165 0 0172 0 0183 0 0200 0 0228 0 0982 01M9 02161 024M 02770 03033 GJ282 0.3522 03753 3000 4 75 28 174AE 27522 37847 487.52 610 08 1060 5 1267.0 190 2 14402 1509 4 15742 1638 5 1701 4 17613 ft9533) e 0 1277 0 2904 0 4320 03597 0 6796 0 3009 1.1966 1.3692 1 4429 1 4976 15434 1.5M1 1A214 16561 1ABBS e
CD160 00165 0 0172 0 0183 0 0199 0 0227 0.0335 0 1586 0 1987 0.2301 02576 0.2827 03065 0 3291 03510 Baco 4 764 1753 2756 378 7 447.5 609 4 800 5 1250.9 1353 4 14331 15032 15703 16342 1698.3 1761.2 (705 00) e C1276 02902 0 4317 03592 0.6738 01994 0.9708 1.3515 1.4300 14e66 15335 1.5749 1A126 14477 1AB06 e
0D160 CDIM 0 0172 CD183 0 0199 9 0225 0 0307 0 1364 0 1764 02066 02326 02 %3 02784 02995 CJ198 3000 t U2 176 0 2762 3791 487.6 606 4 779 4 1224 6 1338 2 14222 1495 5 1 %33 1629 2 1693 6 1757.2 e 01274 02899 0 4312 03585 067U 01973 0 9508 13242 1A112 1.4709 13194 13618 14002 14358 1A691 e
0 0159 0 0164 0.0172 0 0182 0 0198 0.0223 0 0287 01052 01M3 01752 01994 02210 02411 02601 02783 l
4000 e 73 5 in.2 277.1 379 A 487.7 606 9 763 0 1174J 1311.6 1403 6 14813 1552 2 16193 1685 7 1M06 e
0.1271 023B3 0 4304 05573 0 6760 0 7940 0.9M3 12 44 13807 1.4461 1A976 1.5417 1.5812 14177 1 A816 e
CD150 0.0164 0 0171 E01El 0 0196 0 0219 0 0268 0 0591 0 1038 01312 01529 01718 A1990 02050 0.2203 0000 6 31.1 1793 2791 3Bl.2 448 1 604 6 7MO 1042 9 1252 9 1364 6 1452.1 1529.1 1600 9 1670 0 1737.4 e
0 1265 CJEBI 04287 03550 0 6726 01880 0.9153 1.1593 1.3207 1A001 1.4542 13061 1.5481 1.5463 1 4216 CalH 00163 00170 E0lto 00195 00216 002% 00397 00757 01020 01221 01M1 01544 016M 01817 e
0000 6 33 7 181J 281.0 382 7 488 6 602 9 7M1 945 1 1.48 8 1323 6 1422J 15059 1542 0 1654.2 17242 e 41258 CJB70 04271 05528 OM93 0 7826 0 9026 1D176 1.2615 13574 1 4229 1.4748 1.5194 1.5593 1596 2 e 0 2150 0 0163 0 0170 00100 00193 0 0213 0 0244 003M 0 0573 0 0sl6 01004 01160 al298 01424 01%2 7900 6 06J 134 4 283 0 384 2 401 3 Scl.7 729.3 901A 1124 9 12517 1992J 1442 6 1%31 1638 6 1781.1 e 61351 CJR54 0 42% 05507 OM63 0 7777 02926 10350 1J055 IJi?! 13904 1A446 1A938 1A355 13735
50.72(b)
Sh1(g)
PART O o DOMESTIC UCENSING OF PRODUCTIOJ AND tmutAfl0N FACILmES te) Reewes which are neutme er
[ (l)Ds declaratieselany of te the segulations in this part. by lleense k haestsacy Classes spedSedin the eenauen. er by technical specifies-s uen shall be maintained for the (3) The submittal ahan include (l) a hcensee's approvedBasegencyFles er perled specified by the appropriate sortification by a duly autherteed offi-regulation, lleense esedition, or tech-eer of the lleensee that either the in-alcal specifientian If a retention formation escurately presenta chanses
" (16)Of those non4nergencyevents saade since the previous subenittal.
o ecifiedinparagraph[blef this section.
period is not otherwise specifled, such records shau be maintained untu the necessary to reflect informauen and <. p(2)lf tbetnergensyNetiScades th-waa authorises their dispost-analyses subenitted to the th=i=laa System is inoperettwo.6e licensee shaR tion.
or prepared pursuant to th= Man 9utred sel$ cations ela make Ibe ",g mkphoneservice.other requirement, or that no such changes tdM1) Records which saust be asain-
,,,,,gg tained pursuant to this part may be were made; and (ii) an identificauon of E, dedicated tehphone system, er any the original er a..r.1M espy or changes made under the provisions of,I other method whichwE enom &at a l t salcroform if such reproduced copy or B 50.50 but not,..m_5 subenitted to '
report is made as essaas practicalto
'i microform is duly authenticated by the th=i=daa I tuthorised personnel and the micro-(3MI) A revision of the original
,,tbe NRC Operationstater.s 2 form is capable of producing a clear FSAR containing those ortsinal pages and legible copy after storage for the that are stiu appHcable plus new re.
(3)The licensee skalmotify the NRC period specified by Ca==i= ton regu*
placement pages ahan be filed within Isamediately after motiBoation of the lations.
24 months of either July 22,1980, or appropriate State erlocal agencies and (2) If there is a conflict between the the date of lasuance of the operating not later than one bour aher the time the Commission's regulations in this part.
license, whichever is later, and shall licenm declares one af the Emerpacy license condition, or technical specifi-bring the FSAR up to date as of a Ol******
cadon, or other written Commission vnartmum of 6 sponths prior to the a report under (4) When maldag&is section.the approval or authorization pertaining date of fDing the revisioen.
Sraph (s)(3) of to the retention period for the same (D) Not les t 15 days herm
{cene shau identify:
type of record, the retention period 8 SO.*llte) becomes effective, the Direc.
(illbe Emergency Osse declared; or apecuted in the regulations in this s ter of the Office of Nuclear Reactor part for such records shah apply : Regulation shau notify by letter the (ii} Either paragraph (b)(1). *'One-Hour f
rt." e th par p of t a chon in i ect to t NR s
m as of [ tequiring notificatian af the Non-e e se p n e
re tion requirements specified in the res I
- no comply pro Emugency hent.
.ulations in this part.
this section while the program is being (b)Non-Emergencyfrents. (1) One-(e)Each person licensed to operate a conducted at their plant.The Director. NourReports.lf not reported as a
~
of the Office of Nuclear Reactor Reg.
nuclear power reactor pursuant to the provisions of I 50.21 or i 50.22 of this ulation will notify by letter the beens.
declarstionof anEmergencyClan part shall update periodically. as pro-ee of each nuclear power plant being under paragraph (elef this section. the vided in paragraphs (eM3) and (4) of evaluated when the systematic evalua-licensee shall notify the NRC as soon as this section. the final safety analysis tion program has been completed.
practical and in allcases within one report (FSAR) originally submitted as Within 24 months after receipt of this hour of the occurrence of any of the part of the application for the operat*
notification, the Heensee shall fue a g*y,N ing license. to assure that the informa-complete FSAR which is up to date as (i)(A)The initiation of any nuclear l
tion included in the PSAR contains of a maximum of 6 months prior to plant shutdown required by the plant's the latest material developed. This the date of illing the revision.
submittal shall contain all the changes (4) Subsequent revisions shall be TechnicalSpeci5catient.
necessary to reflect information and filed no less frequently than annually (B) Any deviation from the plant's analyses submitted to the Comminston and shall reflect all changes up to a Technical Specificaticas authorized by the licensee or prepared by the 11 maximum of 8 months prior to the
= pursuant to i 50.54(a)of this part.
censee pursuant to Commission re.
date of filing.
(5) Each replacement page shall in r-ll Any event etcondition during quirement since the submission of the clude both a change indicator for the la op(er)ation eat muhe in se condibon o
- original FSAR or, as appropriate. the area changed. es., a bold line vertical j the nuclear pown plant including its
= last updated FSAR. The updated ly drawn in the margin adjacent to the principal safety barriers, being seriously i
a FSAR shall be redsed to include the portion actually changed, and a page chanse 44utificidon (date of change { degraded;or resultsinthe nuclear
' etfects of; all changes inade in the fa.
3 eility or procedures as described in the FSA'.t: au safety evaluations per or change number or both).
, powerplant being (A)In an unanalysed condition that formed by the licensee either in sup-
,,significantly compromises plant safety; port of requested Mcense amendments or in support of conclusions that changes did not involve an unreviewed 7 (B)In a condition that is outside the safet.= quesdon; gad all anstyses of 3 designbasis of theplar.t;or new anfety issues performed by or on 2
behalf of the Heensee at Cammtmaion AIn a condidw con @ 6e
- P ant *e opeaung and emergency t l request. The updated information shall be appre;riately located within 4.prwadas.
,,,,the FSAR.
I 50.y2 knmeeste nennedten W a.
for opersens nuchr power ico.,,
menis en samed=w enacemen of es pate by luensed spesees seeseer po-or
[oectors.(a) Genera 1 Requirements.8 (1)Each
'g**'*" *", Q',*h"$*g'*,E "
l)The licensee shad eubmit revisions j@containing updated information to the r
l g Commission, as specified in I stL4. on a 3 nuclear power reactor licensee lic.ansed enees smersee ese are addressed m replscament.page bests that is a under $ 50.21(b) or $ 50.22 of this part ayyendu a et en m a
i accompanied by a tiet which identines
% shall notify the NRC Operations Center
- cemmerdet ielec see==mber et se rate Ormaans C**ier as ratimi-assa
. the current pages of 6e FSAR following [,vis the Emergency Notification System o.f:
{page replecement November 30,1984 5044 tenetseesse se. nee)
PART W e DOWlESDC LICENSINO OF PRODUCT ON AND UTILIZATION FACILITIES
[ tui) Anymetwelphenomenoneresber applicable soccentrations of the halls
'ggsys Lamusessewentyeysessa, esternaleen& tion that poses an actual spectned in.*,--h-3. Table B of part holde threat to the safety of the nuclear 3D of this abapter in enrestricted areas.
,g(
gas.tt gy g,
e power plant or significantly hampers when everaged over a thee period of wer plant (boensee)abaB subunit a elle personnelin the performance of one bow.
nue Event y Min any as I du6n necessary fw ee safe operaton (B) Any hquid effluent please that event of the type a - ^ J in able exceeda 2 times the limiting combined paragraph within 3D days aner the af the event.Unless otherwise liv) Any event that results or should Maxin um permissible Concentration 0 38 beve resulted in Emergency Core (MpC)(see Note 1 of Appendix B to part p Q
Cooling System (ECCS) discharge into 3D of this chapter) et the point of entry pint mee w ne w l. u d
- ; the reactor coolant system as a result of into the reortving water (La.
regardlus d signincance d he g a valid signal.
unrestricted ares) for all radionuclides
,g,,,,,,,,pg,,,,,,,7,
,4 &a t (v) y event that resulta in a major except tritium and dissolved noble hitlehd Ibe event.
g loss o emergency assumment gases, when everaged over a time period (2) h tir='iaee shaB reput-
, capability,orfalle response capabDity.or of one hour. (Immediate actincetions M N @ud
'de
= communications capability (eg, made under this paragraph also satisfy plant etdown mquired by the plant $
significant portion of control room the requimments of paragraphs [a)(2)
TechnicalSpdficaHm w indicabon. Emergency Notificabon and (b)(2) of I 20.403 of part 30 of this i
System. or offsite noti $ cation system).
chePter-)
prohibited by the plant e Technical (v) Any mnt mquiring the transpat SpeemceSonam
{ (vi) Any event that poses an actual
= threat to the safety of the nuclear of a rs&uctively contaminandpmon (C) Any deviation from the plant's I power plant or significantly hampers alte to an offsite me& cal facility for Technical Specifications authorized personnelin the performance of dutie's treatment.
pursuant to i 50.54(a) of this part.
E necessary for the safe operation of the (vi) An) event or situation.related to
[A $'e ' d1uon udear d
. nudear power plant snduding fires. toxic the health and safety of the public or power plant. induding its principal j
{ gas releases. or ra&oactive releases.
onsite personnel, or protection of the Wety barriers being seriously environment. for which a sewe release degreded. or that neulted in the nuclear (2)rour'NourReports.1f not reported is planned or notification to other power plant being-under paragraphs (a) or (b)(1) of this government agencies has been or will be (A)in an enanalyzed condition that
{ NRC as soon as practical and in all section. the hcensee shall notify the made.Such an event may indude an significantly compromised plant safety; onsite fatabty or inadvertent release of (B)In a condition that was outside the a
= cases.within four hours of the A radioactively contaminated materials.
3 design basis of the plant.or Loccurrence of any of the follow g (c)FollowupNotifecofion.With g *(C)In a con & tion not cov'end by the a f, "
,raung and emergency (i) Any event found while the reactor a respect to the telephone notificat ons
~
is shut down. that, had it been found
. made under paragraphs (a) and (b) of
' ex(ternal con & tion that posed an actua
=
. this section.in ad& tion to making the while the reactor was in operation.
would have resulted in the nudear aged initial notincation.each threat to the safety of the nudear power i
power plant,induding its principaj licensee,shall during the course of the plant of significantly hampered site event:
personnelin the performance of duties I
safety barriers. being seriously degraded s
i or b ing in an unanalped con & tion that (1) hrunedorely reporte (i) any further necenary for the safe operation of the
' signihcantly compromises plant safety.
degradation in the level of safety of the avdear power plant.
(ii) An} event or con & tion that results plant or other worsening plant (iv) Any event or con & tion that e
neulkdin saanual or automatic
' in manual or automatic actuation of any con &tions. inclu&ng those that require actuation of any Engimnd Safety
" Engineered Safety Feature (ESF).
the dedarat on of any of the Emergency Future (ISF) tadudmg the Reactor indu&ng the Reactor protection System Classes.if such a declaration has not ch,ga*g,$p3,
). Howe ata ian been previously mede, or (ii) any change y
6"I from one Emergency Class to another, or that resulted from and was part of the a ed segue c (iii) e termination of the Emergency preplanned sequence during testing or
(,th p
- H-nacer opstion need not be reported.
not be reported-(2)lmmesotely nport-(i) the results (v) Any event or con & tion that alone of ensuing evaluations or assusments of could heve prevented the fulEllment of
' (iii) Any event or condition that alone plant con & bons. (ii) the effectiveness of the safety function of stmetums or could have prevented the fulfillment of nsponse or protective measuns taken, systems that are needed to; the safety function of structuns or and (iii)information nlated to plant (A) Shot down the reactor and systems that are needed to:
behavior that is not understood, maintain itin a safe shutdown
- "&b
I k ma(A) Shut down the ructor and (3) Maintain an open. contmuous Rem nsi intain it in a safe shutdown communication channelwith the NRC a&oactin
. constion.
Operations Center upon request by the
- y)gMit ate the consequences of an (B) Remove reeldual beat.
NRC.
6
(
(C) Control the reluse of re&oactive g,
E cove in ph
) t ste the consequences of an
],
e g
de one
)
)' Any airboree re&oactive or more procedural errors. 'Quipment failures and/or ese very of design.
nlease that exceed: 2 tunes the September 30,1983
PACE 17 52_'_INEQBI_DE _U U GLE6B _EQUEB _EL6HI_DEEB 6IIQ H t _ELUIQit _6 H Q IHES5DDIHeUIGE c
-87/03/09-BJORGEN, J.
ANSWERS -- MONTICELLO MASTER COPY ANSWER 5.01 (1.00)
To protect the fuel (core) from an inadvertent core flow increase (0.5) cuch that the safety limit MCPR requirement is not exceeded (0.5).
4 4LJf. sow, wict 4ccA.tr re /korte,r fuel
/f C M JsD4/4 Meren.s to REFERENCE fn'cor n e.otd u e.e9 e M sucs p des WsrWJweenL FMe%
F#ew ruer scrA Technical Specifications pg. 217; K/A 293009, K1.27 (3.3).sgrery e.snsr
/9 Nor o e e c.s s.P.
ANSWER 5.02 (2.00) a.
8 6/r6/?-
b.
5 />fcAC c.
1 ef4/y*.#
d.
4 g6 /?O (4 9 0.5 each)
REFERENCE Northern States Power Company Lesson 4M8104L - 012, Rev. O BWR Thernichydraulics and Thermal Limits, pages 14, 15, 19 K/A 293009, 1.21(3.6), 1 06(3.8), 1.09(3 7), 1 10(3.7)
ANSWER 5.03 (1.50)
Peripheral rod worth will increase.
(0.5)
High Xe concentration in the center of the core (highest power before scram) will depress the thermal neutron flux in that region causing the relative neutron flux in the peripheral bundles to be higher, thus increase the relative rod worth in peripheral rods.
(1.0)
REFERENCE NSPC Less 4M 8102L-016, Rev.
1, Fission Product Poisoning Effects, Page 45, K/A 292006, 1 07(3.2)
ANSWER 5.04 (1.50)
Delayed neutrons are born at a lower energy than prompt neutrons, a.
therefore, more reach thersial energy implying that their effective number is greater than the actual number produced.
(1 0) b.
The period is longer.
(0.5) l M
..r. g-,
q
- b. e
/\\ N v e{
J i
l y (,f-g j i e-N'
- ... s s
PCGE 18 li__IHEQBI_DE_HUCLE6B_tQHEB EL6HI_QEEB6110Ht_ELVIRit_6HQ
,IHEBEDDIHe5 IGE
-87/03/09-BJORGEN, J.
'CNSWERS -- MONTICELLO REFERENCE 32, 33, & 39.
NSPC Lesson GM 9102L - 007, Rev. 2, Neutron Kinetics, Pages M/A 292001, 1.02(3.1),293003, 1 06(3.70)
ANSWER 5.05 (2.00)
Vd / D#
1.
c b* l^ fC/7/' **- Vil * '
2.
b, C. n 3.
a perpcs(
4.
c V0//S (4 0 0.5 each)
REFERENCE BWR Inherent Reactivity Coefficients NSPC Lesson 4M8102L-016, Rev. Or H/A 292004, 1.14(3 3)
ANSWER 5.06 (2.00)
(1.0)
Even out fuel depletion by flattening the flux.
(This Limit Pu 239 buildup in areas adjacent to inserted rods.
a.
b.
prevents exceeding local peaking factors on subsequent rod 0.Q,fffCf$114t"nep withdrawals.) (1.0)
W/t t-itt.5 o n cceer Adenreds s4 REFERENCE BWR Inherent Reactivity Coefficients NSPC Lesson #M8102 L-016, Rev. O, (Pase 51 of 53)
K/ A 292005 M1.10 (3 3)
ANSWER 5.07 (1 00)
False.
REFERENCE I
NSPC Lesson #M8104L-015, Rev. 3, Fluid Flow (page 7 of 16)
K/A 293006 1.05 (3.3) l l
1*
Reviewed, Fuy 29,1986 II.
CONTROL ROD PATTERNS AND ROD SEQUENCE EXCHANGES r.
W A.
Control Rod Patterns t the beginning of a fuel cycle, rod patterns are set up for the entire cycle with the following factors in mind:
l i
1.
Easy withdrawal or insertion procedures for the operators, c
2.
Minimum rod worths at all times throughout the cycle.
(Less than 1.3*.'
a K, for an entire control, rod as stated in the bases"for Technical Specification 3.3.8.3.)
e.910" gh 3.
Uniform burnup of fuel and control rods.
l 1
tv fpW Minimizing fuel assembly peaking factors.
4.
5.
Maintaining the optimum power shape throughout the fuel cycle; and
/-
(
6.
Reaching the end of cycle with all rods fully withdrawn from the core witn the optimum power distribution, and with t
acceptable peaking factors.
e I f WP/jms C.2 - 0016.0.0
PACE 19 Ei__IHEQBI_DE_HUGLE6B_t0HEB EL6HI 0tEB6110Ht_ELVIDHz.6HD IMEBBQRIHoHIGH ANSWERS -- MONTICELLO
-87/03/09-BJORGEN, J.
C.NSWER 5.08 (2 50)
The difference between the suction pressure and the pressure at o.
which boiling occurs for the existing temp of the fluid entering the pump.
(1 0) b.
HPSH would increase.
(0 5)
Because the feedwater tamperature would decrease, decreasing the annulus temperature and the temperature at the suction of the pump.
(1.0)
REFERENCE NSPC Lesson 4M8104L-015, Rev. 3, Fluid Flow, Page 3 K/A 293004 1.12(3.1) and 293006 K1.10(2 8)
ANSWER 5.09 (2 00)
(0.5) a.
CR1(1-Keff 1)=CR2(1-Keff2) 200(1.95)=400(1-Keff2)
[200(1.95)/4003-1=-Keff (0.25)
.975=-Keff2 Delta P=[Keff2-1/Keff2)-(Keff1-1/Keff1)
(0.5)
Delta P=C(.975-1)/.9753-E(.95-1)/.953 Delta P=(.0256)-(.0526)
(0.25)
Delta P=.027 +/
.002 b.
Subcritical (Will accept answer that matches the solution of part a)
(0 5)
REFERENCE NSPC Lesson 48102L-008, Rev. 3, Subcritical Reactor Theory (page 11 of 19)
K/A 292002 1 07(3.5); 292003 1.01(3.0), 292008 1 05(4.3)
ANSWER 5.10 (2 50) a.
Due to the buildup of Xenon (.75), power may exceed the power limits of the power to flow curve.
(0.75) s/L JPuc fo S4Md b.
(5)
(1 0)
Of kCA) cab McAs /Yotr N/A 10 BA- /N.s tMN >
f*c syny w sfM/a) fe wtA rd
/CL4 W QuAW.
REFERENCE Fission Poisoning Effects, M8102-L14, pg. 201 K/A 292006, K1.07 (3 2)
A
=
PACE 20
. 52__IHE0BI_DE_HUChieB_t95EB_EL6HI 9tEBeIIGHz_ELUI951_eHg IHEBHQQIH6HIGH
- 'CNSWERS -- MONTICELLO
-87/03/09-BJORGEN, J.
ANSWER 5.11 (2 00) o.
Late in core life (0.5) at a moderator temperature of 200-220 degrees F.
(0.5) b.
The coefficient is sas11 (0.5) and becomes negative as temperature increases.
(0.5)
REFERENCE BWR Inherent Reactivity Coefficients M8102-L-016, pg. 12 K/A 292004, K1 02 (2.6)$ kl.01(3.2)
ANSWER 5.12 (1 00)
A condition where a center rod is out with all other surrounding rods fully inserted.
Thus, the flux that each bundle sees comes primarily from its adjacent bundles.
WrLL.
M cego*r g t.n p* w.se.s sM,asce rssst yearin CN sAs e *Asa co w fold Ls Mund *nt ap we u roteHJ
/kon #nta4 dos Wggren HD /JFcessr ce ss gems mangen gy emeA A*os.
REFERENCE BWR Inherent Reactivity Coefficients, M8102L-016, pg. 46 H/A 292005, M1.12 (2.9); K1.04(3.5)
ANSWER 5.13 (1.50) o.
Conduction Convection Radiation (Total of 1.0;.34 will be assigned to first correct answ b.
(1)
Convection (.25)
(2)
Conduction (.25)
REFERENCE Student Handout M8104-2-16, pg. 51 K/A 293007 K1.01 (3.2)
ANSWER 5.14 (1.00)
(1)
Decreases (0.5)
(2)
Increases (0 5)
REFERENCE Northern States Power Lesson Plan H-8104-L-017e Rev. 3, Thermodynamics, Pages 12 and 19 - K/A 293003, K1.22 (3.2)
.-.. ~.
P^CE 21 Mi..IHE9BI_QE_HUGLEeB_t0HEB_EkeUI_DEEBeIIGHt_ELUIDHt_6HD IMEB50018651GH iCNSWERS -- MONTICELLO
-87/03/09-BJORCEN, J.
ANSWER 5.15 (1 00) o.
RW - decreases (.25) b.
RW - remains the same (.25) c.
RW - decreases (.25) d.
RW - increases (.25)
REFERENCE NSP Lesson Plan M-8102-L-016e Rev. 0, BWR Inherent Reactivity Coefficiente pages 43 K/A 292005, K1.09 (2.6)3K1.04(3.5) i e
e
PACE 22
' 62_.tL6HI.IIIIEHE_QERIGHz G9HI69Lt.6HR.IHRIBUNENI6IIGH
-87/03/09-BJORCENe J.
pWSWERS -- MONTICELLO ANSWER 6 01 (2.00)
Fans that are ON will trip and automatically restart when the diesel o.
(1 0) provides power.
A Recirculation fan will not start if both RHR pumps are running.
b.
(1 0)
One pump must be shut down before a fan will start.
REFERENCE Operations Manual B.4.18 M/A 295003e AK 2.04 (3.5), AK 3.02 (3 1), AA 1.01 (3.8), AA 1 03 (4 4)
ANSWER 6.02 (3.00) 2 To maintain secondary containment at a negative pressure.
o.
1.
(0.75)
To process any leakage from primary containment (to e,inimi::e 2.
the chances of a radioactive release).
(0.75)
(0.75)
The bed The purpose is to adsorb radioactive iodine.
b.
efficiency would decrease because the additional moisture (humidity) in the flow decreases the carbon bed's ability to adsorb iodine.
(0 75)
$' M N M " # A secma.1:.1 W piorgt REFERENCE 4
System Description B.4.23 K/A's 290001, K1 04 (.9) and Generic Knowledge 47 (3.8)
ANSWER 6.03 (2.00) i (Any five of the following 9 0 25 each)
Tailpipe temperature recorder TR 2--lii n. P:.:1 C21 shows an increase i
a.
Relief valve indicating lights er *:r.r! m l
Tailpipe pressure trip units
. *:. 1 C253 sho ~' > Me fw//,el f/.~ NA i
Torus water temp inereasins g,gy 9,, uw6.lu
~
Mwe decrease Total steam flow decrease Control valves going closed Nj,P ' Xf
- N TW6 r
_Cfd. cooling flow is reduce d o minimum ACJ MM 22/~~
Nd#dem N b.
B,1LC1) is shut down
- dW2 3NI PUilRis are closed 4 Ash (0.25 each - 0.75 total)
REFERENCE (3.9), K4.06 (3.7) and Operations Manual C.4.B.3 38 K/A 239002, K1.01 Generic Knowledge No. 1 (4.0) and No. 10 (3.4)
[
l
PAGE 23 it..tL6HI 51RIEUR.9ERIGHz C9HIB9Lt.6HD.IMEIBUBENI6IIDH CN3NERS -- NONTICELLO
-07/03/09-BJ0RGENs J.
1 ANSWER 4 04 (1.50) a.
1.
Line could freeze and stop the flow of nitrogen. (Would be unable to maintain < 4% oxygen inside containment i.e.e Tech Spec on containment inerting).
2.
The systen piping could get so cold it would fail by brittle fracture.
3.
The containment might crack from the cold nitrogen.
(Loss of primary containment.)
4.
Liquid nitrogen could enter containment and damage equipment inside.
(Any 3 9 0.5 each) 4 GtMt. 6 teem /t. 3lL Y $03
- A REFERENCE Systes. Description B.4.11 IE Bulletin 84-01.
M/A 223001, K6.08 (3.4),
K3.05 (3.2)
ANSWER 6 05 (2 00)
(0.5 each) a.
False b.
False c.
False d.
True REFERENCE Systes. Description D.1.4, Siniulator Malfunction RR158 K/A's 202001 K1.14 (3.2), K1.02 (4.1), M1.06 (3.6), K1.12 (3.6)
ANSWER 6.06 (1.00)
To allow auton.atic closure of the MSIV's on Low Reactor Pressure of 840 psis.
REFERENCE Operations Manual C.4-C.
K/A 295016 AK3.03(3.7), Generic Knowledge K/A 47.
l
PAGE 24
- 61. EL6HI_IISIEHE_9ERIGHz_GOHIB9Lt eut_INIIBUHEMI61195
-87/03/09-BJORGEN, J.
,dNEWERS -- MONTICELLO
/
ANSWER 6.07 (2 5 g, g y p g y g g, g,
Provide--an-IRM-inop-slarn-(T75) and vould not-track properly-to-powh--
,,t w,a e s a n changes.-(Will accept ~t.ose-Trip-FunctrohS(.75) a.
The APRM's (1 0)Q4 fan Appo/* 7%84/f.54_y ge4J / 784-4 b.
6A #0)
WeMM #9sa4sst L tA..
REFERENCE IE Information Notice 86-1058 K/A 215005, K6.05 (3.1), K/A 215003, M6.02 (3.8)
ANSWER 6.08 (1.00) d.
REFERENCE Technical Specification 3 1, H/A 212000, A2.05 (3.7)
ANSWER 6.09 (2 00)
(approx.-48') initiation signals for HPCI, RCIC, ADS The Low-Low level core spray and LPCI and the diesel start have been lost.
(APR),
(any 5 0.4 ea)
REFERENCE H/A's 218000, H3.01 (4.4), H6.03 (3 9), H1.01, H1.02 (4.1)
OPS Manual B.3 and 0.2 3 ( S.I.l ANSWER 6 10 (1.00) 4 d.
REFERENCE (3.7)
Systes Description D.4.1 and 0.5 6 - H/A 223002, H3.01 ANSWER 6.11 (1 50)
(.75)
Main Stessi Flow
(.75) a.
(not until 12 rods have been withdrawn in Group 2) b.
No, m.,_.
,-,-,,.__-_,,,-._.,_,,--.._,.__....._._,..m.
,.y..
.w.,_,
P^.CE 25 Ai_ _tL 6HI_IIIIE HE _DE RIGH t _G9HIB9L t _6HQ.IHEIBUHEMI6IID W
-87/03/09-BJORCENe J.
,C.N3WERS -- MONTICELLO REFERENCE K4.04 (3.5)
System Description B.5.2 - N/A 201006, M1.04 (3 2),
ANSWER 6.12 (2.00)
(Any two of the following 9 1.0 each)
Loss of one source of power to the D/C start circuit panel C-91.
a.
(Start CMT 02) b.
Loss of D/C field flashing panel C-91.
415 4 KV Bus Breaker operation and bus protection logic.
c.
REFERENCE (3.8), K3.02 (3.8)
System Description B.9.10, 0.9.6 - K/A's 263000 K3.01
/
ANSWER 6.13 (1.00) g fydky%
The buildup of hydrogen _ gas in the battery room. (0.50) a.
b.
Increased danger of fire or explosion.
(0.50)
REFERENCE System Description 0.9 10, M/A's 263000, K6.02 (2.6)
ANSWER 6.14 (1.50)
To block control rod withdrawal while fuel handling operations are in a.
progress.
(.75) b To interrupt power to the hoists on refuel and service platforms when all rods are not inserted.
(.75)
REFERENCE System Description 0 5.5 - K/A 234000, K1.04 (3.6), K4.01 and 4.02 (4.1)
ANSWER 6 15 (1.00)
A. Closed D. Open REFERENCE Procedure B.5.7 ' Reactor Level Control', page SlH/A 259001, K6.03( 3.1),K6.07 (3.8), /f (,,6/ (3, c)
PAGE 26 Zl..tB9CEDUBEE.:.W9886Lt.etH9886kt.EBEBGENGI.699 B$919LOGIG66.G9dIBGL
-87/03/09-BJORGENe J.
CNSNERS -- MONTICELLO ANSNER 7.01 (3 00)
(Any five of the following 9 4 each)
- s. s 1.
Manually scram the reactor
'2.
Verify scram
- 3. ' Notify operators outside control room 4.
Obtain Key 26 and 41 5.
Notify Security 6.
If time permitse any immediate actions of C.4.A. Reactor Scram (Except Mode Switch should.go to shutdown)
Any ren'aining ites s/ listed in.0PS Manual C.4.C.
2nd Floor through TSC to 7Control Room Main entrance to Admin. Blds.e b.
3rd Floor of EFT (Panel C292).
(1 0)
REFERENCE (4.2)
Operations Manual C.4-Cf H/A 295016, M/A 3.01 ANSWER 7.02 (1.00)
C.5 1100 series, RPV control (0.5) because the low level entry condition is satisfied.
(0.5) 04 AN A7W J/MC 4.
TM FMdf a.
.wov,4LJ MAN 1cteenth, REFERENCE (4.4), EH1.01 (4.7)
Operations Manual C.5 - N/A 2950318 EH2 11 ANSWER 7 03 (2.00) a.
3 b.
10 c.
O or 4 d.
7 e.
5 (0 4 each)
REFERENCE 15 (4.0)
Operations Manual C.48 H/A's Generic M/A's No.
l L
PACE 27 Z 4__tB9GE9UBEE.:_H9BB6L t.68H9866L t.EDERGENGL699
. R$919L9EIC6L_G95IB9L
-87/03/09-BJ0RGENe J.
~
ANSWERS -- MONTICELLO l
A' g,d C.NSWER 7.04 (2.25) as the Use RWCU pumps on maximum flow with RBCCW and err"i-- "
8--
o.
heat sink (RWCU and RBCCW heat exchangers).
(0.75) use condensate If the main condenser is available as a heat sinke b.
and feed pumps and RWCU and steam drains as a flow path.
- Y;;eleting M*=r - rfidre r:nd:n;;;..eli.g. (0.75) 4.nse/7'sJ Raise level and utilize the SRV's to the torus and RHR pumps in torus l
cooling.
The RHR heat exchanger (s) are the heat sink.
(.75) c.
REFERENCE Operations Manual C.4.E.3 4.Al M/A 2050003 K5.03 (3.1)
ANSWER 7 05 (2.50)
/5' 3 "A'W** {N " *
'g#',"
Wstk ncteff' P9 Wed JF4C A*^**'
(1.0fu W/p/vs p,ower and level will both initially increase.
i b.
1.
Verify that the initiation signal is or is not valid by checking i
s.
drywell pressure (0.25) and at least two independent level instruments.(.25)
Attempt to secure HPCI or at least place it in the test mode.
2.
(0 5)
(0.5)
Turn the power increase by reducing recirculation flow.
3.
(gpv 3 g.s) l
+,
eennn we n enern REFERCNCC Simulator Malfunction HP-018 Operations Manual D.3 28 H/A 206000 A1 01 (4.4)e H3.01 (4.0), dior'.s 0,.5".//do y/> /ge d sc4 pe.e.
i ANSWCR 7.06 (3.20) a.
1.
Place the mode switch in shutdown.
[
2.
Depress the manual scran, pushbuttons.
3.
Depress the ATWS trip pushbuttons on panel C05.
(
(3 9.4 each) if not automatically trippge Manually trip the recirculation pumpse i
b.
1.
Initiate Boron Injection using Standby Liquid Control (SBLC).
2.
l Inhibit ADS.
i 3.
Use CRD and SDLC for level control at greater than -126 inches 4.
with power less than 3% and SRV's remaining closed.
5.
Use SRV's to control pressure.
(5 0 4 each) (WoLL 4estst 4 0, L A4AJe A)pg L.n.
AAlp poses ftfM' R tf t te,T*
Q
,$,gdgso typAju p )
1 i
we--, - - ---.-m.ww-~w-.,--.mw-.m.w-e
..-*.,-w-y-wow,
.%w mg, y
-,y%,--yww
-9y-__
gy wwy mp
%ww -
ywewym,,,
-v
Pact ze
'ZI__tB9GE99 bel _:_W9856Lt_etH98Hekt_EBERGEUGI_659
. BeQI9k9EIG86_C9HIB9L
-87/03/09-BJORcENe J.
' ANSNERS -- MONTICELLO REFERENCE System Description B.5.4, Procedures C-5-1103, C-5-20078 M/A's 295037e EK 2 01 (4.3)e EK 2 03 (4.2), and EK 2 10 (4.1)
CNSWER 7.07 (3.50)
Verify / assist required automatic actions have occurred (such as o.
1.
reactor scram or feed flow response to the transient).
Enter Monitor plant for emergency procedure entry conditions.
2.
if required.
3.
Notify Shift Supervisor. (Three f 0 5) b.
MCPR Limit MAPLHGR APRM scras, and rod block setpoints Core flow
)
or Core Plate DP noise and Core Thersial Power
)
APRM/LPRM noise levels (Any 5 9 3 each)To minisii e thers al stratification in the vessel and recirculatio c.
loops.
REFERENCE AM 2 06 Procedure C.4.0 1 4. A e Tech Spec 3.5.Hf H/A 295001, HA 1.03 (4.1),
(3 8), AM 2 03 (3.7)
ANSWER 7 08 (2.50) a.
Heat (ssiote), noise, vibratione discharge pressure fluctuatinge seal leakage increase.
(Any 4 9 25 each) (Will accept any other reasonable answers).
b.
No. (0.5)
(Shutdown the pusip.)
Opening the valve siay cause additional punip or system daniage (0.5) due to flashing or water hamtier (0.5).
REFEREUCE H/A 291004, H1.04 (3 1), H1.01 (3.2)
ANSWER 7.09 (1.50)
To reduce the fission heat generation in the core because the core is s.
in danger of being inadequately cooled.
(0.75)
This is an ATHS trip to insert negative reactivity if for sos,e b.
reason the rods did not scras..
(0 75) l
PACE 29
- ZI.".tBOGEQUBEl_ _HORU6Lt etH98dekt.EUERGENGI.eHR
.BeQ19LOGIG6L GQUIB9L
' d.NSWERS -- MONTICELLO
-87/03/09-BJORCENe J.
REFERENCE Tcch Spec Bases 3 2e System Descriptions B.1 4 and B.5 4 M/A 295031e Generic M.04 (4 3), EK 2 13 (4.2)
ANSWER 7 10 (2.00)
Ccndidate 41 - has exceeded the 1000 mtem quarterly limit.
Reject Ccndidate 42 - would exceed the 2000 stem quarterly limit.
Reject Ccndidate 43 - would exceed the 500 mrem pregnancy limit.
Reject Ccndidate 44 - would exceed the 5(N-18) rem exposure limit. Reject (4 9.5 each)
REFERENCE AWI 11 1.141 H/A 294001e K1.03 (3.8)
ANSWER 7.11 (1.50)
A. Starting the mechanical vacuum pump at high vacuum could cause a loss of vacuum in the condenser. (0.5)
D.and C. Both actions could cause a release to the environsient because the cechanical vaevum pump bypasses the off as filters and goes directly to the 3
stock.(1 0)
REFERENCE Monticello Oper ations Manual Procedure C.4-Ae Step 0,6 and Dases Section KA System 271000 K/A K1 01 (3.1), M1.02 (3.3), K1 10 (3.3)
PAGE 30 i
th_CDIWIIIB6IHE tt9CEQWBElt.E9H91139 Bit.6HQ.LIBII6110NE 87/03/09-BJORGEOe J.
,AN55ERS -- MONTICELLO p 4 vMnM '
g,VO
' bp,M OM a
g fp,
I
.00)
W n L. Mo-<.4 y 'y tsor f** *pe4 S.01.
CNSWER
' Isolate (close both valvedo the
'A' noin steam lin (. (1 0)
- b. W.
Reactor Scram From MSIV Closure Logic.
(0.5) a-L f M tW dv4.ht,'.,
o.
2.
Group I isolation logic.:nd : 1798 = -
(0 5)
MbNrd c4ti.3.
Techni;:1 he i'icati:n then;;; sestd West-tuely-c.
See:=: :
he re-" ire 6
' cur unillance requires,enLLa-Peesibly th: 07 eeble nn) rstiveente pei henne1e end p;sstBIy sostin5854 01294 REFERENCE System Description B.5.6e Technical Specifications 3.28 K/A 223002, K4.04 (3 4)e K6.08 (3.7), A1 02 (3 7) i ANSWER 8.02 (1.00)
The actuation of the Reactor Protection Systes. (RPS).
REFERENCE Technical Specification Bases, Section 3 118 to CFR 50.72, paragraph (0)2ite Procedure 4AWI 3.9.5e '30 CFR 50.72 and 50.73 Reporting.'
ANSWER 8.03 (2.00) 1.
No s> ore than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
(Tuesday 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />) 2.
No siore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> period.
(Honday - Tuesday, Tuesday - Wednesday, Thursday - Friday).
3.
At least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rest between work periods.
(Tuesday - Wednesday) 4.
No more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in any 7 day period.
(86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> Sunday -
Saturday) 5.
No s. ore than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight.
(Tuesday)
(5 9 4 each)
REFERENCE Technical Specificationse page 233, ACD 3 18 H/A 294001, A1 03 (3 7) 04 e,4l.
A P* 6 04Pw L 4. wpo mWt a A v e < uhi+ MA % (.M) 8 (1 50) i ANSWER t
Shutdowfe-4he-t lent-end-immediately twithin one-hour) notif y-the I a.
NRC-end-corpor a t e-he a d qu a r t e r sv-6 75 )-
l b. y @RC appr. eve 1-4 e-eegos eed-f er -teetarh (.75 )
Dc%,
t
.,,--_,,,,,,_.gr,.,,,,,_,n
_-,__.n.
' 32..eR6INIIIBeIIVE_tBOCERUBEft_CONDIIIDWiz.6HO LIBII6'IIDWI PACE 31
,,ANSNERS -- MONTICELLO
~57/03/09-BJORCEC, J.
REFERENCE Technical specification 6.4.
K/A 290002 K5 07 (4.4)e Generic N/A 011 (4.2)
ANSWER 0.05 (1 00)
(1) 5 (0.5)
(2) 4 (0.5)
REFERENCE Technical Specifications, page 232e Section 6 1.D., K/A 294001, K1.16 (3.8)
CNSWER B.06
(.75) 1.
Approve work to cominience.
(.25) 2.
Signing off cosipleted WRA's.
(.25) 3.
Updating WRA status.
(.25)
REFERENCE 4ACD-15 4, paragraph 5.9.,
H/A 294001, H1.06 (3.4), H1.07 (3.6)
ANSWER 0.07 (2.00) a.
1.
The Senior Off-Duty Shift Supervisor onsite.
(.5) 2.
Senior Lead Plant Equipsient and Reactor Operator on duty in the Control Roosi.
(.5) b.
1.
General Superintendent Operations.
(.5) 2.
Flant Manager.
(.5)
REFERENCE 4ACD-4.7, paragraph 6.1.1.f., M/A 294001 CNCWER 8.00 (2.00) a.
False b.
False c.
False d.
True e.
False (5 9.4 each)
PAGE 32 i
St. 6951HIIIB61IVE.tBOCERUBElt.G9991IZ9 bit.6HR.LIBII6IIONE
-87/03/09-BJORCENe Je
,,AN6hERS -- MONTICELLO REFERENCE 4ACD-4.5, M/A 294001, K1 02 (4.5)
ANSWER S.09 (1 25)
No. (0 50)
The acceptable limit is 12 weeks + 25% or 15 weeks between due dates.
See from January 15th to April 15the is the quarterly schedule.
This may be extended 3 weeks until May 4th.
The 25X period was exceeded by approximately 4 days.
(.75)
REFERENCE Technical Specification 4.0, M/A 294001, A1.02 (4.2)
ANSWER B.10 (1.50)
No. (0.5)
The substitute part has not been approved.
There is no assurance, therefore, that the part is capable of performing its intended function through all design conditions. (1 0)
REFERCNCE Technical Specification 1.08 H/A 201001, Generic H/A's 42 (3.5), 46 (3 5)
ANSWER 0 11 (1 00)
No. (1 0)
REFERENCE 10 CFR 55.33 ANSWER B.12 (1 50) a.
No.
(0 5)
The basic medical qualifications of 10 CFR 55 are no longer met.
(0e5)
WoLL Mcener MLf4AA)ttrt W6ts/AJG b.
Yes.
(0.5)
StJOH ns
yppp peg n u yy "
REYCRENCE 10 CFR 5511e 55.41, 4ACD-4 7e paragraph 5 6e H/A 294001, H1 16 (3 0)
PAGE as
? EI _6953HIIIB61HE tB9GEDUBElt_C9991119 Hit _699 LIBII611985
-87/03/09-BJORCENe Je o, AN04ERS -- NONTICELLO CNSWER 8.13 (1.00)
O.
REFERENCE Procedure 4ACD-4.7e paragraph 6.1 1.kl K/A 294001, K1 05 (3.7)
ANSWER 8.14 (2 00)
The Shift Supervisor (0.5) and another licensed SRO (0.5) must review a.
the installation.
b.
1.
When operability of the equipment is not required.
(0.5) 2.
The functioning of any interlock or logic affected is not required by plant conditions.
(0.5) 4Gren. unit. Wetw W 6 #9 ces/ren.
Procedure 4ACD-4.88 K/A 294001, K1 02 (4.5), H1.16 (3 8)
ANSWCR 0.15 (2.50) a.
1.
The deviation is necessary to protect the health and safety of the public.
2.
Adequate or equivalent protection consistent with (i.e.,
within) the license is not apparent.
3.
There is insufficient tise for an asiendsient to the license to be approved by the NRC.
(3 90.5 pts each) b.
1.
The General Superintendent Operations 2.
The Shift Supervisor eA p/rt Jru /gA/ pro /pgg r (0.5 pts each)
RCrCRCNCC Procedure 4ACD-4.78 H/A 294001, H1.16 (3 0)
ANFWCR 0.16 (1 00)
(1 0) fly calling the ReDion III of fice in Glen E11yneI111no s.
rak:n At45enAntA AN.ekt4/Ls W o L s.
Jt MCcet*r@, xl. 6 //enust.
EfthCNCE f*M YJF / C 9 im % d'ft)
Procedure 4 AWI-3 9 3,
'Cuidance for Identification of Security Eventse' Paragrsph 6 2 4.e H/A 294001e A1 04 (3 2) i
TEST CROSS REFERENCE PACE 1
..o puts' TION VALUE REFERENCE g.......
05.01 1 00 C8J000 05.02 2 00 CBJ000 05.03 1 50 CSJ0001070 05 04 1 50 CBJ0001071 05 05 2 00 CBJ0001072 05 06 2 00 CBJ0001073 05 07 1 00 C9J0001074 05 00 2 50 C8J0001075 05.09 2 00 CBJ0001074 05.10 2 50 CBJ0001077 05 11 2 00 CBJ0001078 05 12 1 00 CBJ0001079 T 05 13 1 50 CBJ0001080 05.14 1 00 CBJ0001081 05 15 1 00 CBJ0001082 24 50 06.01 2 00 CBJ0001083 06.02 3.00 CDJ0001084 06 03 2.00 CDJ0001085 06 04 1 50 C0J0001086 06 05 2 00 C0J0001087 06 06 1 00 CCJ0001088 06 07 2.50 CDJ0001089 06 08 1.00 CDJ0001090 06 09 2 00 CDJ0001091 06 10 1 0C CDJ0001092 06 11 1 50 CDJ0001093 06.12 2 00 CDJ0001094 06 13 1.00 C0J0001095 06 14 1 50 C0J0001096 06.15 1.00 C0J0001124 25 00 07.01 3 00 CDJ0001097 g7h 9
3 m/)/
07.02 1.00 C0J0001090 07 03 2 00 CDJ0001099 07.04 2.25 CDJ0001100 Ip, L-lg oM 07.05 2 50 C0J0001101 07.06 3.20 C0J0001102 07.07 3.50 C0J0001103 07 00 2.50 CDJ0001104 07 09 1 50 C0J0001105 07.10 2 00 CDJ0001106 07.11 1 50 C0J0001125 24 95 00.01 3 00 C0J0001107
F TEST CROSS REFERENCE PAGE 2
'GUTITION VALUE REFERENCE 68.02 1.00 CSJ0001108 08.03 2 00 CSJ0001109 08 04 1 50 CSJ0001110 08.05 1 00 CBJ0001111 09.06
.75 CSJ0001113 08.07 2.00 CBJ0001114 08 05 2 00 C0J0001115 08.09 1 25 CBJ0001116 00.10 1 50 C9J0001117 08.11 1 00 C8J0001115 09.12 1 50 C8J0001119 08.13 1 00 CBJ0001120 08.14 2 00 CBJ0001121 08.15 2 50 CBJ0001122 08.16 1 00 CBJ0001123 25 00 99.45 s,s*
O