IR 05000220/1987005
| ML20245C186 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/22/1987 |
| From: | Collins S, Crescenzo F, Keller R, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20245C168 | List: |
| References | |
| 50-220-87-05OL, 50-220-87-5OL, NUDOCS 8707020035 | |
| Download: ML20245C186 (48) | |
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U.S. NUCLEAR REGULATORY COMMISSION REG 10N 1
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OPERATOR LICENSING EXAMINATION REPORT'
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EXAMINATION REPORT NO.
87 - 05 -(OL)
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^ FACILITY DOCKET NO.
50 -220 q
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FACILITY LICENSE NO.
OPR-63
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LICENSEE:
Niagara Mohawk Pcwer Corporation
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.301 Plainfield Road
- % racuse, NY 13212 l
FACILTY:
Nine Mile Point Unit 1 i
y y
j EXAMINATION DATES:
}pril 6 - 9,1987 p
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CHIEF EXAMINER:
k ;d,
e (e /6 C
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a F. Crescenzo, Reactor Engineer (Elaniner)
Date.
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REVIEWED BY:
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j D. Lange, Lead R64cter EngtWeer (Examiner)
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REVIEWED BY:
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R. Keller, Chief, Pro;iect Section 1C Date
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APPROVED BY:
4 'l M
hM,ff I
f.. Collins,DeputyDirector,Divisionof Date i
Reactor Projects
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t SUMMARY:
Operat,or licensinh examinations were administered to '.four Senior
Reactor Operator candidates during the week of April 6,1987. One' candidate passed,the written and cperating examinati ns.,)0ne candidate passed?the writ-
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ten examination ' ut. failed the operating examination.
Two candidates' failed
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o the written 'and operating exarrinations.
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An inspectfun'of the Licensed Operator Waining Program was conducted concur-i
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rently during the examin& tion period.
Results of;this inspection are documen-ted in the resident inspector's monthly report (50-220/87-03).
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Y REPORT DETAILS TYPE OF EXAMINATIONS:
Replacement SRO
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EXAMINATION RESULTS:
[ Pass / Fail
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_ Written
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2/2
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[ Oral
[
2/2
[
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~~ Simulator
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_
2/2
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_ OVERALL
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1.
CHIEF EXAMINER AT SITE:
F. J. Crescenzo, Reactor Engineer (Examiner)
2.
OTHER EXAMINERS:
D. Lange, Lead Reactor Engineer (Examiner)
A. Howe, Reactor Engineer (Examiner)
T. Lumb, Reactor Engineer (Examiner)
J. Weschselberger, Resident Inspector, Oyster Creek W. Schmidt, Resident Inspector, Nine Mile 2 3.
SUMMARY OF GENERIC STRENGTHS OR OEFICIENCIES NOTED ON OPERATING EXAMINATIONS:
a.
The candidates demonstrated performance deficiencies when prioritiz-ing the actions and procedures required during severe transients.
This was evidenced by the candidates ' going back and forth from the Emergency Operating Procedures (EOPs) to the Emergency Action Proced-ures (EAPs) to the abnornial operating procedures without fully com-pleting the actions required by any of these procedures.
b.
The candidates' ability to properly execute the E0Ps was poor. More specifically, the following deficiencies were noted:
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1.
The candidates' demonstrated deficiencies in integrating the use of multiple procedures during severe transients.
2.
The candidates did not reenter the appropriate E0P following the subsequent receipt of a reentry condition.
I 3.
The candidates did not attempt to use the main condenser as a heat sink when this was directed by NI-EOP-3, " Failure to Scram"
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or NI-EOP-8, " Emergency Depressurization".
c.
When responding to component and instrument failures, the candidates referred to the abnormal or special procedures only some of the time.
There was no apparent correlation between the severity of the failure and procedure use.
4.
SUMMARY OF GENERIC STRENGTHS OR DEFICIENCIES NOTED DURING GRADING OF WRITTEN EXAMINATIONS:
The candidates performed poorly on the following questions.
Question No.
Topic Class Average 5.04 Effects of deep and shallow 62.5%
rods on axial and radial flux 5.11 Effects of various parameter 56.2%
changes on the void coefficient 6.01 Fill in the blanks regarding NMS 62.5%
7.04 Precautions regreding recirculation 67.5%
pump discharge and suction valves 7.07 Procedural requirements regarding 62.5%
the RWM 7.10 Basis for automatic initiation of E7.5%
emergency cooling 8.01-10 CFR 50 requirements to deviate 25.0%
from procedures or license conditions 8.05 Technical Specification definitions 55.0%
8.10 10 CFR 50 reporting requirements 68.7%
8.12 Technical Specifications problem 52,5%
regarding inoperable containment spray concurrent with inoperative diesel
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4.
EXIT INTERVIEW OETAILS Personnel Present at Exit Interview:
NRC Personnel F. J. Crescenzo, Reactor Engineer (Examiner)
A. G. Howe, Reactor Engineer (Examiner)
T. Lumb, Reactor Engineer.(Examiner)
W. Schmidt, Resident Inspector Facility Personnel K. F. Zollitsch, NMPC Training Superintendent R. Seifreid, NMPC Assistant Training Superintendent D. J. Straka, NMPC Training Supervisor, Unit 1 5.
SUMMARY OF NRC COMMENTS MADE.AT EXIT INTERVIEW a.
The items noted in paragraph 3 above were discussed, b.
The facility was informed that results of the examinations would not be available until after the grading of the written exam and after regional management review.
Every effort would be made to have the results within 30 days, c.
The facility was informed that the exam review was efficient and well organized.
Comments to the exams were pertinent and well documented.
d.
The results of the requalification program inspection results were discussed with the facility personnel.
Details of this inspection can be found in the resident inspector's monthlyL report (Inspection Report No. 50-220/87-03)
5.
SUMMARY OF FACILITY COMMENTS MADE AT EXIT INTERVIEW:
The licensee stated that the exams were well prepared ~and thanked the examiners for their efforts.
6.
RESOLUTION TO FACILITY COMMENTS ON THE WRITTEN EXAMINATION:
The " master answer key" is marked to ~ indicate resolution ~ of comments raised during the two hour exam review, conducted on April 7,1987 andLis attachment one to this report.
The following represent NRC resolutions of the facility comments which '
were documented in a letter from T. E. Lempges to :R. M. Keller dated April 15,1987 which is Attachment 2 to this repor.
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l NRC Resolutions-
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Resolution 5.04a
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Comment not accepted.
Full credit answer must-I include discussion in the answer key since the i
void reactivity is a significant contributor to i
the differences in the net reactivity added for the two conditions given.
Resolution 5.04b Comment accepted.
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Resolution 5.08d
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Comment accepted, tolerance band widened to
j psia.
Resolution 5.10a
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Comment will not be accepted as caryunder is not a significant contributor to NPSH. Additionally, no reference material was provided to support
this comment, j
Resolution 6.01
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Comment accepted. Parts b, c, d, of this question
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will be deleted from the exam.
Resolution 6.02b.1 -
Comment not accepted. The question does not ask what specific alarms and trips are caused 'by the Hi/Lo Lo-Lo Rosemount instruments but rather what will occur as level varies over the range of the instrument.
Normal range alarms will be
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required.
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I 6.02b.2 -
Comment noted.
Credit is only deducted for incorrect responses not additional correct responses.
Resolution 6.06
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Comment accepted, CST added to the answer key.
Resolution 6.07b
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Comment accepted.
Resolution 6.09 Comment accepted regarding pressure ranges,
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answer key changed to reflect correct ranges.
Resolution 6.10 Comment not accepted. Question asked why the
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valves are interlocked with. reactor PRESSURE.
Resolution 6.11a Comment not accepted. This answer can be corre-
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lated to NMP1 Learning Objective.
Resolution 7.01 Comment partially accepted.
Partial credit will
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be awarded for answer reflected in the commen i
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4 Resolution 7.04c Comment accepted although all three answers will
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be required for full credit.
Resolution 7.06 Comments accepted.
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j Resolution 7.09a
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This question will be deleted from the examina-i tion.
Resolution 8.01 Comment will be partially accepted.
Full credit
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answer must include discussion of the criteria in answer key.
Criteria in AP-4 may be accepted as partial credit, however, question asks for cri '
teria necessary to exceed license conditions, AP-4 refers to criteria for operating outside station procedures. Vague or broad interpreta-tions of either set of criteria will not be i
accepted.
Resolution 8.07b-Comment accepted.
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Resolution 8.08a Comment accepted.
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Resolution 8.10 Comment not accepted. Senior operators are ex-
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pected to recognize events which require one hour.
notifications without reference to 10 CFR 50.72.
Resolution 8.11
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Comment accepted.
Question deleted from the examination.
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Attachments:
1.
SR0 Examination and Answer Key
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2.
Facility Comments - Letter from T. E. Lempges to R. M. Keller, dated April 15, 1987
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NUCLEAR REGULATORY COMMISSION i
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SENIOR REACTOR OPERATOR LICENSE EXAMINATION l
l FACILITY:
_N__IN_E_M__IL_E__P_O__IN__T_________
_
REACTOR TYPE:
_@WB-GEg_________________
DATE ADMINISTERED: _@Zf93fgZ________________
i EXAMINER:
_HgWE _@ ________________
z CANDIDATE:
___
N k _________
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IN@l8UCllgN@_IQ_C@NDID@lE; j
Use separate paper for the. answers.
Write answers on one side only.
Staple question sheet on top od the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing l
grade requires at least 70% in each category and a final grade of at
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least 80%.
Examination papers will be picked up six (6)
hours after i
the examination starts.
% OF j
CATEGORY
% OF CANDIDATE'S CATEGORY
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__28LUE_ _Igl@L
___@CgBE___
_y@LUE__ ______________C@lEGQ8Y_____________
a
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_22:99__ _26:db 5.
THEORY OF NUCLEAR POWER PLANT
___________
________
OPERATION, FLUIDS, AND THERMODYNAMICS
_2}t99__ _29t@Z
___________
________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
AND INSTRUMENTATION
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_22199__ _23t}3
___________
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
1 l
_22199__ _2Sz23
________ B.
ADMINISTRATIVE PROCEDURES,
)
___________
CONDITIONS, AND LIMITATIONS J
_23199__
________%
Totals
___________
Final Grade All work done on this examination. is my own.
I have neither given nor received aid.
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___________________________________
Candidate's Signature j
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l NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
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During' the admi ni strati on 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.
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Restroom trips are to be limited and only one candidate at a time may
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leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
q I-3.
Use black-ink or dark pencil gnly to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the j
examination.
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5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only 'the' paper provided f or answers.
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7.
Print your name in the upper right-hand corner of the first page of gach section of the answer sheet.
B.-
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new paga, write gnly gn gne sidg of the paper, and write "Last Page" on the last answer sheet.
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Number each answer as to category am number, for example, 1.4, 6.3.
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10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. The point value for each question is indicated in-parentheses after the question and can be used as a guide for the depth of answer required.
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Show all calculations, methods, or assumptions used to obtain an answer
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i to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DD NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be'done after the examination.has
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18. When you complete your examination, you shal'Is
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a.
Assemble your ex ami nati on as follows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3) ' Answer-pages including figures which are part of the answer.
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b._
Turn.in your copy of=the examination and.all pages used.to answer the examination questions.
c.
Turn.in all scrap paper and the balance of the paper that you did not'use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving,Lycu.are found in this area while the examination is still in progress, your license may be denied or revoked.
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QUESTION 5.01 (1.00)
Assume the reector is subcritical and Keff equals O.9 and the count rate doubles on a rod pull.
i Calculate the new Keff. (Show All Work)
(1.00)
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QUESTION 5.02 (2.00)
The SRM reading is observed to change from 1.5 * 10^4 to 3.0 * 10^4 over a two minute time span. You also observe that the period meter is giving you a 60 second period.
Based on these observations, determine if the instrumentation is providing correct indication. Show All Work.
(2.00)
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QUESTION 5.03 (2.00)
The delayed neutron f raction is designated as bar beta.
c) How does the value of bar beta change (increase, decrease) from the beginning of core life (BOL) to the end of core life (EOL)?
(0.50)
b) What is the major cause for the change in bar beta from BOL to EOL?
(0.50)
c) How does bar beta affect reactor operations such that control reactor power is possible?
(1.00)
QUESTION 5.04 (2.00)
a) Why is the net reactsvity increase due to a rapid withdrawal of a deep rod much greater when the reactor is just at initial critical than it would be when the reactor is at full power?
(1.00)
b) What is the effect on the axial and radial flux shape of an operating reactor at full power:
1) when a deep rod is moved? Why?
(0.50)
2) when a shallow rod is moved? Why?
(0.50)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
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QUESTION 5.05 (1.50)
i Assume the reactor is at 100% power and the xenon has built up to 3% dk/k:
a) If the reactor scrammed, how long would it take for the
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xenon to peak.
(0.75)
I b) How long would it take in hours for the xenon to decay to approximately 0% assuming the reactor remains shutdown.
(0.75)
QUESTION 5.06 (2.50)
The reactor is operating at 75% power when the EPR system power is lost. How would the following parameters INITIALLY change and WHY?
i A.
Reactor pressure (1.00)
B.
Core flow (0.75)
C.
Reactor power (0.75)
QUESTION 5.07 (2.00)
You notice that MAPRAT is equal to 1.002 on the computer printout of P-1 (periodic core performance edit) that has just been printed.
, Answer the f ollowing questions wi th respect to the above:
a) What is the definition of MAPRAT?
(1.00)
b) Have any thermal limits been exceeded?
(0.25)
c) What is the basis of the limit MAPRAT monitors?
(0.75)
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CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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THERMODYNAMICS i
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e QUESTION 5.08 (2.00)
During your Shift, an SRV inadvertently opens from 100*/. power and 1000 psi a.
Use a Mollier Diagram or the Steam Tables to answer the following (ASSUME a saturated system and instantaneous heat transfer):
l a.
WHAT is the SRV tailpipe temperature, assuming atmospheric pressure in the Suppression Pool and No Reactor Depressurization?
(0.50)
b.
If the Suppression Pool Pressure were to increase, would the tailpipe temperature INCREASE, DECREASE, or REMAIN THE SAME?
(0.50)
c.
If the reactor is depressurized when the SRV is opened, will the Tailpipe Temperature INITIALLY INCREASE, DECREASE, or REMAIN THE SAME?
(0.50)
d.
At WHAT Reactor Pressure will the Tailpipe Temperature be at its MAXIMUM value (curing the depressurization)?
(0.50)
QUESTION 5.09 (3.00)
Describe HOW and WHY a centrifugal pump's discharge head is affected for each of the following.
(Consider each condition separately and assume NPSH is maintained in all cases.)
a.
Suction pressure increases.
b.
The discharge valve is throttled closed.
c.
The temperature of the fluid, being pumped, increases.
(3 & l.0 ea)
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QUESTION 5.10 (3.00)
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For each of the f ollowing events, or changes in the plant status, i
state whether the change will bring the retirc pumps CLOSER TD,
FARTHER FROM, OR HAVE NO EFFECT on the point where the recire
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pumps will cavitate. EXPLAIN EACH.
a.
Vessel water level increase.
(1.00)
I b.
Loss of a feedwater heater.
(1.00)
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Increase in recirc pump speed.
(1.00)
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QUESTION 5.11 (2.00)
State how the magnitude _of the VOID COEFFICIENT will vary (increase or
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decrease) as the f ollowing parameters vary with core age.
a. Burnable poison-concentration (0.5)
b.
Plutonium 240 concentration (0.5)
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Uranium 235 concentration (0,5)
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Control rod density (0.5)
j QUESTION 5.12 (2.00)
State how CRITICAL POWER varies (i e, increases, decreases, or is-not affected) by each of the following:
e.
If coolant mass flow rate increases (0.5)
i b.
If reactor pressure increases (0.5)
c.
If l oc al power increases (0.5)
d.
If inlet subcooling increases (0.5)
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(***** END OF CATEGORY'05
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6 __P6@NI_gygIEMg_ DESIGN _CgNIBOL _@ND_lNSIBUMENJSIlgN PAGE
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QUESTION 6.01 (1.00)
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Provide the correct numbers to fill the blanks in the followings l
l There are (a) _______ LPRM detector (s) in the reactor core.
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.The LPRM detectors are separated into (b) del eted group (s).
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Each group supplies (c) deleted si gnal (s) to (d) deleted APRM j
channel (s) and there are (e) ______ APRM channel (s) in the APRM system.
(1.00)
QUESTION 6.02 (1.50)
The Hi/Lo Lo-Lo Rosemount level instrumentation has a range of 100 inches.
a) How many inches above the top of the active fuel is the
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zero level?
(0.50)
b) Assume you are starting from normal vessel level. Within the indicated range of this instrument:
1) what are the trips and alarms which will occur if indicated level increase to 100 inches? Set points required.
(0.40)
2) what are the trips and alarms which will occur if indicated level decreases to 0 inches? Set points required.
(0.60)
i QUESTION 6.03 (1.50)
During startup, assume that three of the SRM's are not fully inserted and the count rate is 50 cps:
a) How would the above two conditions have to be changed so with all other permissives satisfied the rod block circuit would allow control rod withdrawal?
(1.00)
b) How could you enable the SRM system to be able to activate a trip of the Reactor Protection System.?
(0.50)
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(***** CATEGORY 06 CONTINUED ON NEXT PAGE
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QUESTION 6.04 (3.00)
Concerning the recirculation system flow units that input to the APRM system, name three flow unit trips and the condition or trip.
value that cause ROD BLOCK actuation.
(3.00)
QUESTION 6.05 (1.50)
Name the six automatic signals that will initiate closure of the
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Main Steam Line Isolation Valves.
(1.50)
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QUESTION 6.06 (1.50)
a) What are the three sources of water for the condensor shells of the Emergency Cooling System?
(0.75)
b) What valves are actuated on EC system actuation to put the i
condensers into service?'
(0.25)
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c) Why is there a ten second time delay af ter a signal-is received before the EC system is actuated.
(0.50)
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QUESTION 6.07 (3.00)
j Concerning the Automatic'Depressurization System (ADS):
a) If bus power is not available to the core spray pumps, would a i
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coincident Low-Low-Low Reactor water level and high drywell pressure trips actuate an automatic depressurization of'the reactor system? Why or why not?
( 1. 00)'
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b) If a blowdown is in progress, how can the operator stop-the blowdown?
(1.00)
c) What are.the normal pressure actuation setpoints.f or the ADS valves and how many valves operate at each setpoint?
(1.00)
(*****. CATEGORY.06 CONTINUED ON NEXT PAGE *****)
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. QUESTION 16.08 (1.50)~
During nor mal. operation what three conditions would initiate
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-the High Pressure Coolant Injection. System?
(1.50)
QUESTION 6.09 (3.00).
State the three different operating pressures WITHIN the CRD.
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hydraulic system and identif y the purpose of. each.
(3.00)
QUESTION. 6.10 (2.00)
With respect to'_the Core Spray. System:-
a) The inboard and outboard Core Spray -isolation valves are interlocked so'that they can not be opened'at the j
same time under certain reactor pressure conditions.
WHY7 (1.00)
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.b) Suppose there is a " Core Spray Sparger High Delta Pressure" alarm annunciated. What is this alarm ~ indicating?
(1.00)
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(***** CATEGORY 06' CONTINUED ON NEXT-PAGE
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. I6 PLANT' SYSTEMS DESIGN _CONTROLg_AND INSTRUMENTATION PAGE
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QUESTION 6.11 (2.00)
While-performing a' reactor startup per the attached rod group
eequence excerpt, the f ollowing conditions exist:
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- The currently selected rod is 30-23.
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- The currently latched RWM group is group 4
- One insert error exists on group 1 rod'34-43 As the operator is withdrawing the selected rod, control rod 18-11 of greap i begins to drift and eventually settles at position 44.
a.
EXPLAIN the response of the RWM system following the. rod drift.
Include in your answer WHY further withdrawal of the selected rod CAN or CANNOT. continue without restoring
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the drifted rod to its original position.
(1.50)
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b.
How would your answer differ if TWO insert errora existed prior
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to the rod drifting?
(0.50)
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DUESTION 6.12 (2.00)
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The plant is operating at 100% power.
A complete loss of instrument air occurs.
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With NO Operator action, answer the below questions?
i A.
The level in the hotwell would (increase, decrease, remain the same).
(0.50)
B.
Service water flow to the TBCLC heat exchangers would (increase, decrease, remain the same).
(0.50)
C.
Scram Discharge Volume drain valves would (open, close, remain as is).
(0.50)-
D.
Main Feedwater Control Valves would (open, close, remain as-is).-
(0.50)
(***** END OF CATEGORY 06
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7.
PROCEDURES NORMAL _ABNQRMAL _EMERGENQY_AND.
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i i-QUESTION 7.01 (2.00)
During reactor operati on a loss of Reactor Building Closed Cooling Water'has occurred. Per "N1-SOP-8,RBCLC Failure" a) Is it necessary to trip the recirc. pumps.?
WHY OR WHY NOT ?
(1.00)
b) Is a reactor scram required?
(0.50)
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c) How can the heat input ~to the RBCLC system be reduced?
(0.50)
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QUESTION 7.02 (1.50)
What are the whole body emergency dose limits that are authorized
]
by Nine Mile Point Nuclear Station procedures fort a)
non lif e threatening emergency whole body dose limit?
(0.75)
b)
life threatening emergency whole body dose limit?
(0.75)
..
QUESTION 7.03 (1.50)
At what reactor temperature and pressure is the shutdown cooling
,
.
!
mystem able to be placed in service.
(1.50)
i
,
t QUESTION 7.04 (3.00)
)
a.
Why are you cautioned that two recirculation loop suction and discharge valves must be open at all times?
(1.00)
b.
Under what special conditions may all the recirculation pump discharge valves be closed?
(1.00)
c.
What is the basis' f or the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> maximum time limit a ' recirculation pump can idle in hot standby or reactor power operation?
.(1.00)
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QUESTION 7.05 (3.00)
J Step 4.4 N1-EOP-4, " Primary Containment Control" states that if Torus temperature and RPV pressure control cannot be restored cnd maintained below the HCTL and less than 3 ERV's are open the operator must emergency depressurize the RPV.
a.
Why is RPV depressurization required?
(1.00)
b.
State whether or not the following situations would require emergency depressurization. Figure N1-EOP-4.1 is attached, j
1.
No ERV's open Torus water level 7.5 feet Torus water temperature 112 degrees F RPV pressure 500 psig (1.00)
2.
No ERV's open Torus water level 7.5 feet Torus water temperature 93 degrees F RPV pressure 750 psig (1.00)
QUESTION 7.06 (3.00)
In the " Control Room Evacuation" procedure there are a number of q
cctions that are required before the step " evacuate the control
'
room" is reached. List six actions and BRIEFLY explain the basis for each.
(3.00)
QUESTION 7.07 (2.00)
s) Will the Rod Worth Minimizer enforce rod blocks and stop rod motion if a control rod is being driven in using the
" Emergency Rod IN " mode of the RMCS? Reactor is in STARTUP MODE. (1.00)
b ')
What are the minimum monitoring requirements for. rod withdrawal if the RWM fails during a startup? (Consider for the first group withdrawn and for later in the rod withdrawl sequence)
(1.00)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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QUESTION 7.08 (2.00)
l What,per the site S-RP-1 " Access and Radiological Controls", are the radiation exposure limits used for personnel radiation exposure control? (Weekly, Quarterly and Yearly required)
(2.00)
!
QUESTION 7.09 (.50)
a) Deleted b) In the event of loss of air or DC power to the Emergency Condenser condensate return IV's causing them to fail open, how would you control reactor pressure?
(0.50)
QUESTION 7.10 (1.00)
What is the basis f or automatic ittitiation of the Emergency Cooling System rather than manual initiation per the basis stated in the Technical Specifications?
(1.00)
i QUESTION 7.11 (2.00)
a.
During a transfer of RPS bus #11 from the MG power to the maintenance power supply, the power to RPS bus #11 is lost momentarily and then restored. What two actions should be taken to reset the ATWS cabinet 1948 ?
(0.75)
b.
Who may authorize a parallel bus transfer of a' reactor trip MG set ? (two required)
(0.5)
c.
TRUE OR FALSE: All fuel must.be removed from the reactor core in order to completely shut down the RPS system.
(0.75)
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-QUESTION 7.12 ( 1. 50 ).
What~are the three entry conditions for Procedure N1-EOP--2, RPV Control ?
( 1. 50) -
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QUESTION B.01 (1.50)
Describe the conditions per 10 CFR 50.54(X) under which it'is j
permissible to deviate f rom license conditions or technical
)
epecifications.
(1.50)
]
l i
i QUESTION 8.02 (3.00)
Tech Spec's define Reactor Building integrity to mean that the reactor j
building is closed and THREE conditions are met.
'1 i
a.
List these THREE conditions.
(1.50)
i b.
When must the Reactor Building Integrity be in effect?
(1.50)
i l
QUESTION 8.03 (1.00)
a.
A fire' brigade of at least (a) members shall be maintained
______
on-site at all times.
b.
Unexpected absences of the fire brigade members may be tolerated
]
for up to _________(b)
hours, while cction is taken to fill the
.
vacant position (s).
j I
QUESTION 8.04 (1.50)
Technical Specifications state that temporary changes to-g procedures may be made provided three conditions are met.
'
What are these conditions?
(1.50 l
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I QUESTION B.05 (2,50)
'
i Based on NMPNP1 Technical Specification DEFINITIONS, answer the f ollowing either TRUE or FALSE:
a.
The TEST INTERVALS that are specified in Tech. Spec.s are only valid during periods of power operation and do not apply in the event of extended Station Shutdown.
(0.50)
b.
An OPERATING CYCLE is that portion of station operation between the end of one operating cycle and the end of the i
next operating cycle.
(0.50)
c.
CORE ALTERATION is the addition, removal, relocation, or other manual movement of fuel or core components in the
,
reactor core, including movement of control rods with the control rod drive hydraulic system.
(0.50)
j i
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d.
A FIRE WATCH PATROL is a patrol that requires an area with inoperable fire protection equipment to be inspected at least every four (4) hrs.
(0.50)
e.
A TRIP SYSTEM is an arrangement of sensors and auxilary equipment required to generate and transmit a plant parameter to an instrument channel for the purpose of i
satisfying a component response.
(0.50)
QUESTION B.06 (2.50)
- NOTE: USE THE ATTACHED SECTION OF THE TECHNICAL SPECIFICATIONS TO
- ANSWER THE FOLLOWING QUESTION.
FULLY REFERENCE ALL SECTIONS YOU USE. *
i The plant is at 100 */. steady state power, it is two a.m.
and one of your operators informs you that the closed position indication light for MS ERV # 6 is out. The problem is determined not to be a burned out light bulb but a maintenance problem that can't be resolved until day shift.
May the plant continue to be operated at power? Give a brief description and the bases for any actions you would take.
(2.50)
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QUESTION 8.07 (2.00)
- NOTE: USE THE ATTACHED SECTION OF THE TECHNICAL SPECIFICATIONS TO
- ANSWER THE FOLLOWING QUESTION.
FULLY REFERENCE ALL SECTIONS YOU'USE. *
-The plant is operating at 75% power. You are. informed that the Biweekly MSIV Closure Surveillance Test is seven days late and
'
should be done now. Halfway through the test, ONE Outboard MSIV fails to meet the specified closing time. Per the Tech Specs:
a) What action.is required due to the test being late?
(1.00)
b) What action is required due to failure to meet MSIV closure times?
(1.00)
)
QUESTION 8.08 (2.50)
What is the Technical Specifications basis for the f ollowing:
a)
Condenser Low Vacuum Scram (1.00)
]
{
b)
Turbine Control Valve Fast Closure Scram (1.50)
i i
QUESTION 8.09 (2.00)
As Shif t Supervisor you have declared an alert and-have assumed the j
position of Emergency Director. List four responsibilities of the-
)
Emergency Director that may NOT be delegated.
(2.00)
.
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QUESTION 8.10 (2.00)
l From the events listed below, identify those which require a one hour notification to the NRC as given by 10 CFR 50.72.
1.
The TS limit for the RPV cooldown rate is intentional 1.y violated while executing procedure N1-EOP-2.
2.
The Reactor Protection System automatically initiates on a valid high Drywell pressure signal.
3.
An orderly plant shutdown is initiated in order to comply with the NMPNS1 Technical Specifications.
4.
A fire breaks out in the Technical Support Center.
(3.00)
QUESTION 8.11 (.00)
l DELETED l
QUESTION 8.12 (2.50)
- NOTE: USE THE ATTACHED SECTION OF THE TECHNICAL SPECIFICATIONS TO
,
- ANSWER THE FOLLOWING QUESTION. FULLY REFERENCE ALL SECTIONS YOU USE
i i
During the beginning of your shift it is determined that one of the Con-
'
tainment Spray pumps in Loop
"A" is inoperable. Half way through your shift the Diesel Generator associated with Containment Spray Leop A" fails its
"
operability surv. test. Using the attached TECH SPECS what action is required for the containment spray system when:
1.
Only the one pump is inoperable.
(1.0)
2.
The one pump and the diesel generator are inoperable.
(1.5)
- THE REACTOR IS OPERATING AT RATED POWER ***
(***** END OF CATEGORY 08
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(************* END OF EXAMINATION
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
.
ANSWER 5.01 (1.00)
cr1/cr2 =(1-Keff2)/(1-Keff1)
.5=
(1-Keff2)/(1.9)
(.5 pts)
(1.00)
,
Keff2 =.95 (.5 pts)
REFERENCE GE Reactor Theory ; Ch.
3, Objective 3.1.3 292OO3,K1.02 KN of Subcritical Multiplication 2.4/2.4 292OO3K102
...(KA'S)
ANSWER 5.02 (2.00)
From Doubling Time (the time required for the reactor power to increase by a factor of 2) you can see that the reactor period i
should be 1.44 *120 sec.or 173 seconds (1.5) which is inconsistent l
with the period meter reading of 60 seconds.(.5)(The period can also be calculated from the power equation:
From the power equation - when P1
=2 * Po then in 2 = t/T or
. 693 = t/T or Period equal s 1. 44 times doubling time.
Period = 1.44*120 = 173 seconds)
REFERENCE GE - Reactor Theory - Objective 3.3.6 292OO3,K1.09 KN of Doubling Time 2.5/2.6 292OO3K109
...(KA'S)
ANSWER 5.03 (2.00)
c) Decreases (0.50)
b) The contribution to the delayed neutron population by U 235 decreases as the U 235 is burned out and the contribution f rom plutonium increases, decreasing bar beta.
(0.50)
c) Delayed neutrons make control of a reactor power i
possible since the delayed neutrons make up part of the succeeding generation neutron population and are added to the population at a rate such that the reactor period can be controlled by the relative slow rod movement.
(1.00)
._ ---
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
-
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REFERENCE GE - Reactor Theory - Objective 3.4.6 292OO3,K1.04 KN'of delayed neutrons 2.4/2.4 292OO3K103
...(KA'S)
ANSWER 5.04 (2.00)
c) On withdrawal of a deep rod at power, the reactivity worth of the rod is immediately compensated f or by the increase in voids making the net reactivity increase quite small, (.5) At initial critical the reactor sees the full rod worth on withdrawal which results in a period rate increase until the reactivity added is reduced by moderator temperature increase or insertion of rods (.5)
(1.00)
(Consider for partial credit a discussion strictly on rod worths)
b) 1) Change in deep rod position effects the radial flux shape over a number of fuel cells due to the coupling effect caused by the change in voids (.25) and the axial effect is minor (.25).
(0.50)
2) Change in shallow rod position ef f ect the axial flux shape primarily (.25) as the change in flux is localized due to rod shadowing and the lack of voids lower in the fuel bundles (.25).
(0.50)
(Discussion of reverse power effect will be accepted for full credit)
REFERENCE GE - Reactor Theory - Objective 5.2.5 292OO5,K1.09,K1.12, KN of Parameter Effects On CR's 2.5/2.6 292OO5K109 292OO5K112
...(KA'S)
ANSWER 5.05 (1.50)
c) Seven to eleven hours (0.75)
b) Seventy hours +/- ten (0.75)
REFERENCE GE - Reactor Theory - Objective 6.2.5.4 292OO6,K1.07 KN Xenon after Scram 3.2/3.2'
292OO6K107
...(KA'S)
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
ANSWER 5.06 (2.50)
l A.
Increases (0.25)due to the control valves going shut (in response to the MPR becoming the controlling signal.) (0.5)
B.
Increases (0.25)due to the reduction in the void content (of the two phase mixture in the core). (0.5)
C.
Increases (0.25)due to the collapse of voids from the higher pressure (which adds positive reacti vity).
(0.5)
REFERENCE G.E.
Reactor Theory, ch.
4, pg. 4-24 G.E.
Heat Transfer and Fluid Flow, ch.
8, pg. 8-41. Objective 6 KA 292OOO,4K1.11 KN of Void Coeff.
2.5/2.6 KA 241000, K1.01-3.9,1.02-4.1,3.01-4.1, 3.02-4.3 292OOOK111 241000K302 241000K301 241000K102 241000K101
...(KA'S)
ANSWER 5.07 (2.00)
a) MAPRAT = Actual APLHGR/APLHGR Limit b) Yes(.25),
(0.25)
c) The average planar linear heat generation rate is the limit established to keep fuel rod surface temperatures from exceeding 2200 degrees Fahrenheit during a design basis LOCA (.75).
(0.75)
REFERENCE GE Heat Transfer and Fluid Flow, Objective 9.4 (check on 241)
293OO9K110
...(KA'S)
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
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ANSWER 5.08 (2.00)
i i
a.
295 degrees F (+/- 5 percent C15 degrees FJ)
b.
Increase c.
Increase d.
450 psia (+/- 50 psia)
REFERENCE
'
Steam Tables /Mollier Diagram GE Heat Transf er & Fluid Flow Ch. 3 Objective 3.1.2 293003, K1.23 Kn Steam Tables 2.8/3.1 293OO3K123
...(KA'S)
.
ANSWER 5.09 (3.00)
.
a.
Head increases (0.5) the pump is still putting the same amount of work into the fluid, therefore the same delta pressure increase across the pump, so as suction pressure increases so will the discharge head (0.5).
b.
Head increases (0.5) as system resistance to flow increases, pump head increases (0.5).
c.
Head decreases (0.5) as temperature increases system resistance to flow decreases (lower vi scosi ty) ; therefore head decreases (0.5).
(3.00)
l l
REFERENCE GE Heat Transfer & Fluid Flow - Objective 6.10.10 291004, K1.13 KN of Cent. Pumps 2.6/2.7 291004K113
...(KA'S)
5 __IHEQBy_gE_NQCLE@B_EQWEB_EL@@l_ GEE 6@llgN _ELQlgS _@NQ PAGE
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ANSWERS -- NINE MILE P' INT-87/04/07-HOWE, A.
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ANSWER 5.10 (3.00)
l a.
Farther from cavitation (0.5). As the water level increases, the static head of water component in NPSH determination is also j
increasing adding NPSH (0.5).
(1.00)
j b.
Farther f rom cavitation (0.5). When feedwater heating is lost inlet subcooling increases (inlet temp. decreasing) which brings the water farther from saturation.(0.5)
(1.00)
c.
Closer to cavitation (0,5).
As pump speed increases the pressure in the eye of the impeller decreases, which will cause cavitation earlier for the same NPSH. (required NPSH increases) (0.5)
(Alternate answer : Increased flow increases power and feedwater flow thus subcooling increases and this provides more NPSH)
(1.00)
REFERENCE G.E.
HEAT TRANSFER AND FLUID FLOW P.
6-76 to 6-81, SLD-6.10.5, 6.10.6 KA 293006 K1.10-2.8, KA 291004 K1.06-3.3, KA 202001 K1.03-3.3 291004K106 293106K110 202OO1K103
...(KA'S)
ANSWER 5.11 (2.00)
a.
increase b.
increase c.
increase d.
decrease (0.5 each)
REFERENCE General Electric Reactor Theory pgs. 4-23,24 Learning Objective 4.3 292004 K1.13 2.1/ __TSEQBy_gE_NgCLE@B_EgWE6_EL@NI_ GEE 6@llgN _ELglgS _@NQ PAGE
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
)
ANSWER 5.12 (2.00)
a.
Increases (0.5)
b.
Decreases (0.5)
c.
Decreases (0.5)
d.
Increases
'(0. 5 )
.-
I REFERENCE l
HC HT&T No. 11, Learning Objective 2, pages B and 9.
FJC 294
)
Pilgrim THT&FF pgs. 9-26 thru 9-30 Nine Mile 2 THT&FF pgs. 9-26 thru 9-30.
Nine Mile 1 THT&FF pgs. 9-26 thru 9-30
293009 K1.22 thru K1.25 3.2
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
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i ANSWER 6.01 (1.00)
(a) 120 (b) deleted (c) del eted (d) del eted (e) 8 (.50 each)
(1.00)
{
)
REFERENCE Nine Mile Point 1,
Operations Technology Chapter 9 c.
Objective 3, LPRM Systems q
215005,K1.04 KN OF LPRM &APRM INTERCONNECTION 3.6/3.6 I
215005K104
...(KA'S)
i ANSWER 6.02 (1.50)
c) 86 inches (0.50)
b) (Normal range i s 65-83 inches), alarms hi or low at.the extremes of this range.
(.4)
Turbine trip at 95 inches (.2)
Lo Level Scram at 53 inches (.2)
Lo Lo Isolation at 5 inches.
(.2)
(1.00)
REFERENCE NMPNS1 Operations Technology - Reactor Vessel Instrumentation -
)
Objective 3 216000,K1.16 KN ves. inst. & main turbine 3.0/3.1 216000,K5.01 KN ves. inst. & level 3.1/3.2 2160000K50 216000K116
...(KA'S)
ANSWER 6.03 (1.50)
a) Either more than 100 cps (.5).or the SRM detectors must be inserted (.5).
(1.00)
b) The Non-coincident bypass switch must be placed in the non-coincident position.
(0.50)
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PLANT SYSTEMS DESIGN _CONTROLt_AND INSTRUMENTATION PAGE
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
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REFERENCE NMPNS1 SRM Objective 10 215004,K1.02 KN Connection Between SRM/RMC 3.4/3.4 215004K102
...(KA'S)
J ANSWER 6.04 (3.00)
l a)
1.
Flow Upscal e Trip (. 5)
b)
1.
105% (. :25 )
2.
Flow Unit Inoperative trip (.5)
2.
a) Mode switch not in operate (.25)
b) Any internal module unpluq:Md (.25)
c) Upscal e trip (.25)
3.
Flow Comparator Trip (.5)
3.
More than seven (7) %
,
difference in flow output signals from each flow unit (.5)
(3.00)
)
REFERENCE NMPNS1 - APRM Objective 6 215005, K1.10 KN cause effect APRM/RMCS 3.3/3.3 215005K110
...(KA*S)
ANSWER 6.05 (1.50)
Low-Low-Reactor Water Level Low-Low-Low Condenser Vacuum Low Reactor Pressure with Mode Switch in Run Main Steam Line High Radiation High Steam Flow High Temperature in Steam Tunnel (.25 each)
(1.50)
REFERENCE NMPNS1 Primary Containment System - Objective 5
-
223OO2,K1.01 KN cause effect PC/MS 3.8/3.9 223OO2KO1
...(KA'S)
6 __P(@NI_SYSIEMg_QESIGN _CQ@l896t_@NQ_1 Ngl 8QMENI@ll@N PAGE
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ANSWERS -
NINE MILE PDINT-97/04/07-HOWE, A.
-.
ANSWER 6.06 (1.50)
a)
1.
Makeup Tanks 2.
Condensate Transfer System (CST's acceptable)
3.
Fire system (.25 each)
(0.75)
b) Condensate return isolation valves, one for each loop (0.25)
c) To allow f or short term pressure spikes caused by a turbine trip to be ignored.
(0.50)
REFERENCE NMPNS1 - Emergency Cooling System - Objective 9 207000,K1.01,K1.04,&K1.06 KN EC & Rx,Cond. Sys & Fire System 3.8/4.0 207000K106 207000K104 207000K101
...(KA'S)
ANSWER 6.07 (3.00)
a) No (.5),
The same power bus that supplies the respective core spray pump supplies the ADS actuation circuit (.5).
(1.00)
b) By depressing both reset buttons
---OR--- by placing both ADS inhibit switches in bypass (1.00)
c) 1090 psig (.25)
1095 psig (.25)
1100 psig (.25)
two at each set point (.25)
(1.00)
REFERENCE NMPNSI ADS - Objectives 6 & 7.
-
21BOOO,K4.01,K4.03 KN of ADS Logic Control 3.7/3.9 3.8/4.0 210000K403 21BOOOK401
...(KA'S)
6 __ELONI_SygIEMS_ peg 1GN _CQNIBg61_@NQ_lNSIBgMENI@IlgN PAGE
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
i
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I ANSWER 6.08 (1.50)
l a) Turbine Trip, or b) Low Reactor Water Level of (53")- or c) Anytime f eedwater flow exceeds (1.9*10^6 lbs/hr) at the discharge of either motor driven feed pump.
(.5 each)
(1.50)
REFERENCE NMPNS1 - HPCI - Objective 9 206000,K4.07 KN of initiation 4.3/4.3 20600K407
...(KA'S)
,
i ANSWER 6.09 (3.00)
"CRD Charging Water pressure" (.5) supplies high pressure water, q
normal l y 1390-1510 psig to the accumulator charging header (.5).
l l
"CRD Drive Water Pressure" (.5) supplies water at a pressure
'
approximately 260 psi above the reactor vessel pressure for
'
motive force to move control rod drives (.5).
"CRD Cooling Water Pressure" (.5) supplies water at approximately 17-40 psi above reactor vessel pressure to cool the graphitar seals (.5)
(3.00)
REFERENCE NMPNS1 - CRD Hyd. - Objective 14 201001,K1.10 KN relation between CHD Hyd. & CRD 2.8/2.8
...(KA'S)
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
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ANSWER 6.10 (2.00)
a) Protects the sections of the core spray piping designed for low pressure from high reactor pressure (during flow tests.while operating the reactor).
(1.00)
b) This alarm is indicative of a core spray sparger break inside the reactor vessel.
(1.00)
REFERENCE NMPNS1 - CORE SPRAY - Objective 5 & 11 209001,K4.01,K4.OB KN to prevent overpressure & initiation 3.2/3.4 3.8/4.
209001K408 209001K401
...(KA'S)
ANSWER 6.11 (2.00)
'
a.
The sequence is automatically unlatched, (.25) the core is_ scanned (.25)
and all blocks are applied (.25). Since only two insert errors will exist rod withdrawl c are continue following scan. (.75)
b.
Same as above except rod withdraw 1/ insert blocks will remain following core scan due presence of three insert errors.
(0.5)
REFERENCE
,
NMF1 ops tec chap. 6 pg. 19 L.O.
- 10 201006 A2.03 3.0/3.2 K4.01 3.4/3.5
"
ANSWER 6.12 (2.00)
A.
Decrease B.
Increase C.
Close D.
Remain as is (4 9 0.5 ea = 2.0)
REFERENCE NMPNS1
- Instrument Air, Objective 3 System K/A's K6.01 KN of the effect of Loss of Plant Air 201001K603 256000K601 259001K601
...(KA'S)
,
,.
.....
.
..
.
.
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..
.
- - - -. - - - - - - - -. - -
-,
2 __PBQCEQUBES_ _NQBM861_8BNQBM8L _EMEBGENCY_@NQ PAGE
l x
889196991C9L_CgNIBQL j
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
ANSWER 7.01 (2.00)
i'
a) Yes (.5) pump coolers and pump motor coolers are lost (.5).
(1.00)
b) Yes(.5)
(0.50)
3.
Isolate the Cl ean-up System (.5).
(0.50)
REFERENCE Nine Mile Point 1; N1-SOP-8,RBCLC Failure l
Objective 9, RBCLC Course Objectives.
295018K1.01 KN OF EFFECTS OF SYS. LOSS ON COMP.
3.5/3.6 295018K101
...(KA'S)
l l
ANSWER 7.02 (1.50)
a) 25 rem (0.75)
b) 75 rem (0.75)
j REFERENCE
,
NMPNS1 procedure S-RP-1 (6.4)
HP Objectives 2 294001,K1.03 KN FAC RAD. CONTROL REO.
3.3/3.8 294001K103
...(KA'S)
ANSWER 7.03 (1.50)
Temperature equal to approximately 350 degrees F. (.75) and Pressure equal to 120 psi g (. 75).
(1.50)
REFERENCE NMPNS1 - Shutdown Cooling - Objectives 7&B 205000,K1.01 KN of SDC service & Rx Pr.
3.6/3.6 205000K101
...(KA'S)
.
__
___
Zg__P6QgEQWBES_ _Ng6086t_@@Ng60@6t_EdE6GENgy_8dp PAGE
B891969G1986_GQNI696
.
ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
ANSWER 7.04 (3.00)
To maintain communication between the core area and the annulus (0.5) this ensures accurate level indication (0.25) and promotes natural circulation (0.25)
b.
The' reactor is flooded to above the main steam nozzles (0.5)
or the steam dryer and the moisture separator are removed (0,5)
c.
1.
To prevent thermal shock which would occur when the pump-is restarted. (0.33)
2.
To prevent pump seal damage (0.33)
3.
For cold water reactivity considerations (0.33)
REFERENCE N1-OP-1 section c, pg. 11 202002 SG 10 3.3/3.3 LO #10 ANSWER 7.05 (3.00)
a.
The RPV must be manually depressurized while the Heat capacity of'the suppression pool is still sufficient to accomodate the blowdown.
(1.0)
b.1 YES (1.0)
c.2 NO (1.0)
REFERENCE N1-EOP-4 295026 Hi Supp Pool water temperature knowledge of reasons for emergency depressurization EK3.01 3.8/4.1 i
l
.
J
Z___P69CEQUBES_ _NQBM@6t_@BNQBM86t_EME6GENCy_6ND PAGE
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5091969G1C@L_CQNI6Q6 t
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
ANSWER 7.06 (3.00)
)
a) Scram the reactor (.25) to reduce heat input (.25)
(-OR-place the reactor in a safe condition)
(0.50)
b) Verify that the main turbine and generator have tripped (.25)
To verify that automatic actions have occurred as expected (.25)
(0.50)
c) Initiate both ECs (.25) to provide heat sinks (.25)
(0.50)
d) Turn both VESSEL ISOLATION switches on Panel E to isolate (.25)
to reduce cooldown rate (.25)
(-OR-to conserve inventory)
(0.50)
{
e) Take radios (.25) for communication (.25).
(0.50)
f) Verify that HPCI has initiated (.25) for level control (.25)
(0.50)
)
g) Activate the E-PLAN (.25) to alert appropriate personnel (.25)
(0.50)
h) Sound the fire alarm (.25) to notif y station personnel (.25)
(0.50)
i) Obtain keys (VA-1 and GE-75) from the CSD desk or the SSS office (.25)
To open the shutdown panel gates and to operate S/D panel keyl oc k switches (.25)
(0.50)
(any 6 of 9 f or full credit)
REFERENCE i
Nine Mile Point 1 - Control Room Evacuation, N1-SOP-9 EDP Objectives 2.2, Use N1-SOP-9 I
295016GKA10 PERFORM WITHOUT REF. TO PROC. THE IMM. OPER.
3.8/3.6
295016GK10
...(KA'S)
I i
,
!
i
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I j
,...
.
.
.
.
...
.
.
.. _
.
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.
.......
.....
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..;
-
l Z __669CEDQBE@_;_NQBd@L t_@@NgBd@L _EDEB@ENgY_@NQ PAGE
t
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50DigLQ@lC@L_CQNIBQL
.
ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
j
]
-
.
!
ANSWER 7.07 (2.00)
a)
Yes (1.00) (The RWM will not be bypassed if reactor power is below the LPSP).
(1.00)
b)
If more than 12 crds have been withdrawn a Second licensed operator verifies that the first operator is f ollowing the rod pattern.(.5) If less than 12 ced's have been withdrawn then the startup is halted until the RWM i s repaired (.5).
(1.00)
REFERENCE Nine Mile Point 1,
Procedure N1-OP-37 " Rod Worth Minimizer (RWM)"
NMP1 - Technical Specification 3.1.1; 201006,K1.01, KN OF CONN. BETWEEN RWM & RMCS 3.4/3.4 201006K101
...(KA'S)
C ANSWER 7.08 (2.00)
100 mrem /Wk 1000 mrem / Calendar Qtr 4000 mrem / Calendar Yr (.67 each)
(2.00)
REFERENCE NMP1, S-RP-1, V Objective 10 Si te Admini strati ve Control 294001 PWG K1.03 KN OF FAC RAD CONTROL REQ.
294001K103
...(KA'S)
ANSWER 7.09 (.50)
a) Del eted l
b) Close the steam supply valves to the Emergency Condenser.
(0.50)
REFERENCE Nine Mile Point 1,N1-OP-13 Emergency Cooling Objective 11;
,
Zi__C89EE993EE_2_U9806ht_6EU9006ht_EDE6@ENgy_@NQ PAGE
i
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689196991C86_CQNI@g6
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
j
-
.
207000,K1.01 KN OF CONN. BETWEEN EMERG. COND. & RX VESSEL 3.8/4.0 207000K101
...(KA'S)
i l
ANSWER 7.10 (1.00)
To conserve reactor vessel water inventory.
(1.00)
REFERENCE
Emergency Cooling, Objective 6, NMP1 T/S bases 3.1.3
!
207000,K1.03 KN of EC/Rx Level Relation 3.7/3.8 294001K103
...(KA'S)
ANSWER 7.11 (2.00)
)
a.
1.
verify no ATWS lights on F panel [0.53 2.
push the appropriate reset button EO.253 i
b.
1.
SSS [0.253 2.
Ops. Super vi sor 00.253 i
c.
TRUE EO.753 REFERENCE Nine Mile Point 1 - Procedure N1-OP-40, Reactor Protection System p.
10, 1 RPS - Objective 13 212000, K 2, 3.5/4.3; K 10,3.8/4.2; A4.14 3.8/3.8.
i
ANSWER 7.12 (1.50)
a) RPV Level below +53 inches (0.50)
b) Drywell Pressure above 3.5 psig (0.50)
c) RPV pressure above 1080 psig (0.50)
i REFERENCE Nine Mile Point 1 - RPV Control - Procedure No. N1-EOP-2, Objectives 1.1, Emergency Operating Procedures, 295031,EK1.01 KN OF Rx LEVEL & AD. CORE' COOLING 4.6/4.7 295024,EK1.01 KN OF DRYWELL PR. & CONT. INTEGR.
4.1/4.2 295025,EK1.02 KN OF PRESS EFFECT ON Rx INTEGR.
4.1/4.2 295024K101 295025K102 295031K101
...(KA'S)
9.__8901NIEI6811VE_P69CEgyBE@t_CQNDlll@@@t_@NQ_(ldll@llQN@
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
ANSWER B.01 (1.50)
In an emergency when this action is immediately needed to protect the
,
public health and safety and no action consistent with license l
conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent.
(1.5)
(Exact wording not required. Criteria from AP-4 will be accepted for par ti al credit. Vague or broad criteria will not be accepted.)
REFERENCE 10 CFR 50.54(x)
ANSWER 8.02 (3.00)
l a.
1.
At least one door in each access opening is closed.
(0.50)
'
2.
Standby gas treatment system is operable.
(0.50)
3.
All reactor building automatic ventilation system i sol ati on valves are operable or are secured in the isolation posi ti on.
(0.50)
b.
1.
During the refuel operating condition.
(0.50)
'
2.
During the power operating condition.
(0.50)
3.
Irradiated fuel or irradiated fuel casks are being handled in the reactor building.
(0.50)
REFERENCE NMPNS Tech Spec 1.0 X Definitions & 3.4.0 objective 1.2.H 290001,K1.01 Kn of phy. relation 3.3/3.5 29001K101
...(KA'S)
ANSWER B.03 (1.00)
a)
five (0.5)
b)
two (0.5)
REFERENCE NMPNS Tech Spec's 6.2.2.g objective 1.2.k Fire Watch
,
294001,PWG,K1.19 Kn of Fac Prot. Req.
3.5/.
-
__ _ _,
.8 __@QulNISIB@IlyE_P8QCEQQBEQt_QQNQlllQNQt_6NQ_LidlI@IlgNQ PAGE
t
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
t
)
-
.
294001K119
...(KA'S)
ANSWER 8.04 (1.50)
a)
The intent of the original procedure is not altered.
(0.50-b)
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
(0.50 l
c)
The change is documented reviewed and approved by the j
,
General Superintendent - Nuclear Generation or his i
designee within fourteen days of implementation.
(0.50 REFERENCE Technical Specifications 6.8.3 Admin. Procedures, Objective EO-1
,
294001,A1.03 Ability to use procedures 2.7/3.7 l
ANSWER 8.05 (2.50)
a.
TRUE j
b.
FALSE c.
FALSE d.
FALSE i
e.
FALSE (0.50 for each correct ans. )
(2.50)
REFERENCE NMPNPSI-Tech. Spec. definitions.
Objective S __89MINISI6611VE_P6gCEQUBES _CgNQlligNS _@NQ_LIMlI@IlgN@
PAGE
t t
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
..
ANSWER 8.06 (2.50)
No (0.50)
LCO 3.1.5, requires six valves to be operable (0.25) to permit rapid depressurization to allow for full flow core spray operation in the event of a small line break. (0.25)
(0.50)
,
!
'
LCO 3.2.9 (0.5) (requires that only 5 of the 6 valves need to be operable. As five valves would limit reactor over pressure below the lowest safety valve set point in the event of rapid reactor i sol ati on )
(0.50)
The LCO for section 3.1.5, where all 6 valves shall be
,
operable, is the most limiting and should be adhered to. (0.50)
(0.50)
Be at 110 psig or less and at saturation temperature or less (0.50)
j within ten (10) hours (0.50)
(1.00)
l REFERENCE
'
NMPNS1 Tech. Spec.'s 3.1.5 & 3.2.9 Objective 1.1.b 218000,K1.02 KN OF RELATION BETWEEN ADS & CORE SPRAY 4.0/4.1 ANSWER 8.07 (2.00)
a) Reportable violation of Technical Specifications (per TS 6.6)
(per 50.72 & 50.73 of 10 CFR 50)
(1.00)
b.
Action
'b'
of 3.2.7 states that with the one MSIV INOP due to j
exceeding the allowable closing time, the affected steam line j
shall be isolated. (If the problem is not corrected initiate J
,
an orderly shutdown within one hour and have the reactor in cold shutdown within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.)
(1.00)
]
REFERENCE j
NMPNSI-TB-SEC. 3.2.7 Reactor Coolant System Isolation Valves.
j Objective 2.1 b l
223OO2,K1.01 KN OF CAUSE/EFFECT BETWEEN MSIV & MAIN STEAM 3.8/3.9 l
223OO2K101
...(KA'S)
l l
i l
)
!
i
--.
.. - _
_ _ - _ _ _ _ _ _
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PAGE
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
.
.
i
ANSWER 8.08 (2.50)
a) (To prevent the cladding safety limit from being exceeded l
due to the sequence of events following a loss of vacuum, j
'
a reactor scram is initiated.) The condenser low vacuum.
scram is anticipatory to the stop valve closure scram and causes a scram before the stop valves are closed (0.5) and thus the pressure transient is less severe (0,5).
(1.00)
b.
(The turbine control valve fast closure scram is provided to scram.) the reactor in anticipation of the rapid increase in pressure and neutron flux (0.75) that would result from fast closure of the turbine control valves and subsequent failure of the bypass valves to open (0.75).
(1.50)
REFERENCE NMPNS1 Technical Specifications Bases 2.1.1 Fuel Cladding Objective II B 1.4 241000,K1.27 KN of Rx/ Turbine Pr REG. SYS. \\7 Cond. Vac.
3.1/3.1 241000,K1.07 KN of
& Turb. Stop Valves 3.4/3.6
"
"
241000K107
...(KA'S)
ANSWER 8.09 (2,00)
a.
The decision to notify Off-site Emergency Management Agencies.
b.
Any Protection Action Recommendations made to Off-site Emergency Management Agencies.
c.
Classification of the event.
d.
The decision to require on si te evacuation.
e.
The decision to exceed normal radiation exposure limits.
(4 & O.5 ea.)
(2.00)
<
REFERENCE NMPNS1 EAP-1 Objective II EO 3 SYSTEM GENERIC KAIS
a
_
.... _
... -
_
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ANSWERS -- NINE MILE POINT-87/04/07-HOWE, A.
]
.
.
ANSWER 8.10 (2.00)
i
.
1-2-3-4 (.5 for each correct answer)
(2.00)
REFERENCE 10 CFR 50.72 System Generic K/A's 3 ANSWER 8.11 (.00)
<
DELETED j
REFERENCE NMPNS1 TS Table 3.6.14-1 note c Objective TS II A.2 294001,K1.03 Kn of Fac Rad Control Procedures 294001K103
... '. K A ' S )
ANSWER 8.12 (2.50)
l 1.
One pump inop. -- operation may continue for 15 days provided (0.5)
the other pump is demonstrated operable immediately and daily
'
thereafter. T/S 3.3.7.b.;
4.3.7.d.
(0,5)
2.
With the Diesel inoperable, the
"A" Containment Spray Loop becomes inoperable.
(0.75).
T.S.
3.3.7.d requires that plant operation can only continue f or 7 days and that C/S loop. "B" be shown operable immed.
and daily thereafter T/S 4.3.7.d.(0.75)
(T/S 3.0 requirements to declare the system operable are not satisfied.)
REFERENCE NMP-1 T.S.
Sec.
3.3.7.
209001 GOO 6
...(KA'S)
!
.)
TEST CROSS REFERENCE PAGE
.
hUESTION VALUE REFERENCE
________
______
__________
.
05.01 1.00 AXAOOOO425 05.02 2.00 AXAOOOO426 05.03 2.00 AXAOOOO427 05.04 2.00 AXAOOOO428 05.05 1.50 AXAOOOO429 05.06 2.50 AXAOOOO430 05.07 2.00 AXAOOOO431 05.08 2.00 AXAOOOO432 05.09 3.00 AXAOOOO433 05.10 3.00 AXAOOOO434 05.11 2.00 AXAOOOO450 05.12 2.00 AXAOOOO451
______
25.00 j
i 06.01 1.00 AXAOOOO407 06.02 1.50 AXAOOOO435 06.03 1.50 AXAOOOO436 06.04 3.00 AXAOOOO437 06.05 1.50 AXAOOOO439 06.06 1.50 AXAOOOO440 06.07 3.00 AXAOOOO441 j
06.08 1.50 AXAOOOO442 06.09 3.00 AXAOOOO444 06.10 2.00 AXAOOOO445 06.11 2.00 AXAOOOO452 06.12 2.00 AXAOOOO446
______
23.50 07.01 2.00 AXAOOOO401 07.02 1.50 AXAOOOO418 07.03 1.50 AXAOOOO443 07.04 3.00 AXAOOOO447 07.05 3.00 AXAOOOO448 07.06 3.00 AXAOOOO400 07.07 2.00 AXAOOOO406 07.08 2.00 AXAOOOO409 07.09
.50 AXAOOOO403 07.10 1.00 AXAOOOO408 07.11 2.00 AXAOOOO399 07.12 1.50 AXAOOOO404
______
23.00 08.01 1. h'v AXAOOOO410 08.02 3.00 AXAOOOO412 08.03 1.00 AXAOOOO413 08.04 1.50 AXAOOOO414 08.05 2.50 AXAOOOO415
,
TEST CROSS REFERENCE PAGE
,
.
QUESTION VALUE REFERENCE
________
______
__________
- 08.06 2.50 AXAOOOO416 08.07 2.00 AXAOOOO417 08.08 2.50 AXAOOOO419 08.09 2.00 AXAOOOO420 08.10 2.00 AXAOOOO421 08.11
.00 AXAOOOO422 08.12 2.50 AXAOOOO449
______
23.00
______
______
94.50 i
.
)
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