IR 05000333/1987016
| ML20215D771 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/10/1987 |
| From: | Collins S, Keller R, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215D736 | List: |
| References | |
| 50-333-87-16OL, NUDOCS 8706190140 | |
| Download: ML20215D771 (79) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING REQUALIFICATION' EXAMINATION REPORT EXAMINATION REPORT NO.
50-333/87-16 FACILITY. DOCKET NO.
50-333
. FACILITY LICENSE NO.
LICENSEE:
Power Authority of the State of New York P. O. Box 41 Lycoming, New York 13093 r
FACILITY:
James A. FitzPatrick Nuclear Power Plant
~ EXAMINATION DATES:
May 6-8, 1987 CHIEF EXAMINER:
b lC&(AncAn,k M 6[9 / T T Dav$d J. Lange, d BWR Examiner Date REVIEWED BY:
M 6/7 /1/7 Robert M. Keller,' Chief, DRP Section 1C Date l
bi1[/$ki/)N j Q 97 l
APPROVED BY:
Samuel J. Collins,~ Deputy Director, DRP Date Inspection Summary:
The NRC administered written requalification examinations to seven (7) Senior Reactor Operators (SRO) and three (3) Reactor Operators (RO). One (1) SRO and i
two (2) R0s failed their respective written examinations.
The NRC also administered operating examinations to six (6) SR0s and three (3)
R0s.
One RO failed the operating examination. All remaining license holders passed their respective operating examinations.
The licensed operators participating in the evaluation were chosen so that an adequate distribution of on-shift and staff engineers was represented.
Based on the 60% passing rate for the evaluation, the FitzPatrick Licensed Operator Requalification Program was determined to be marginal.
Thi s -i s in accordance with ES-601, " Administration of Requalification Program Evaluation".
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8706190140 870610 t
PDR ADOCK 05000333 V
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REQUALIFICATION PROGRAM EVALU'ATION-Facility:
James Ai FitzPatrick Nuclear Power Plant-Chief. Examiner:
David J. Lange-Dates of Evaluation: May 6-8, 1987-Areas Evaluated:
X Written X Oral-NA Simulator Examination Results:
-R0 SRO Total-Pass / Fail Pass / Fail Pass / Fail Evaluation-Written Examination 1/2 6/1 7/3 Marginal Operating Examination 2/1
'6/0 8/1 Satisfactory Overall Program Evaluation: MARGINAL Six.(6) out of the ten (10) licensed operators examined'during t'his evaluation passed all portions of their NRC-administered examinations.
In ' accordance' with 'the guidelines provided in ES-601, " Administration of. NRC Requalification Program Evaluation", the 60% passing rate for this examination indicates a requalification program in marginal standing.
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Submitted:
Forwarded:
Approved:
h M alufti k DJL ChAef Examiner Section' Chief
" Deputy Divition Director
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DETAILS 1.
' Scope of Evaluation
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i On May 7, '1987, the NRC administered written examinations (60% ~ in length in comparison to the standard NRC licensing examination) to seven (7) SR0s
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and three.(3) R0s licensed at the James. A.
FitzPatrick Nuclear Power
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. Plant.
On May 6 and May 8,19' 7, the NRC administered operating examinations to
three (3) R0s licensed at JAFNPP. Due to time ' constraints, the NRC admin-istered operating examinations to six (6) of the seven (7) SR0s who had i
taken the NRC written examination.
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Exit Interview-The Chief Examiner conducted an exit interview on'May 8, 1987. The fol-lowing persons were present.
NRC Personnel
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David Lange, Lead BWR Examiner
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Lynn Kolonauski, Reactor Engineer (Examiner)
l Tracy Lumb, Reactor Engineer (Examiner)
Brian Hajek, NRC Region I Consultant Examiner Alan Luptak, Senior Resident Inspector Facility Personnel
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l Radford Converse, Resident Manager.
l William Fernandez, Superintendent of Power Douglas Lindsey,' Operations Superintendent Donald Simpson,-Training Superintendent
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Paul Walker, Training Specialist
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Fredrick Catella, Training Specialist
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Summary of NRC Comments The Chief Examiner stated that the requalification program evaluation would be based on the final results of the NRC-administered written and operating examinations.
He further emphasized (as stated in the Entrance L
Meeting conducted 5/5/87) that this was an evaluation of the requalifica-tion program and not an evaluation of the individual operators.
The examiners found the plant housekeeping to be satisfactory.
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I Summmary of NRC Comments (continued)
J The ' examiners identified. several generic strengths among the Senior Reactor Operators.while administering the _ operating examinations. These included a high level of familiarity with the facility's administrative
requirements and proficiency in using the-JAFNPP Emergency Operating Pro-i
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cedures and Technical Specifications.
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The examiners noted no generic weaknesses among the SR0s while adminis-tering the operating examinations.
i The examiners ident.ified a generic weakness among the Reactor Operators
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while administering the operating examinations.
The weakness involved
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verifying (from the Main Control Room) that the remote 02-ADS-71 panel was energized,- as required ' by F-AOP-36, " Stuck Open Relief Valve".
When asked, several on-shift personnel were also unaware of the control room annunciator used to verify the status of the remote ADS panel.
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One (1) Reactor Operator and one (1) Senior Reactor Operator were identi-fied as potential failures on the operating examination; their individual performance: deficiencies were discussed. -The Chief Examiner requested the
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licensee to. take immediate actions to ' verify the watch standing capabil-ities of these two -license holders in accordance with ITP-5 which de-scribes the'JAFNPP NRC-approved Requalification Program.
The Chief. Examiner stated that the written results would be reported to i
the Resident Manager by telephone once the regional review of the examina-
. tion grading was completed.
j No preliminary.results concerning the Requalification Program evaluation were given. The. Chief' Examiner stated that the final ~ program evaluation would be documented in the examination report.
The examiners identified several deficiencies in the training material provided for the preparation of the exams.
These included:
Instructor Outlines' referenced Student Guides which were not provided
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to the examiners.
i The ADS Operating Procedure referenced OP-1 for instructions to man-
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ually blowdown the RPV; however, OP-1 contains no such instruction.
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Several JAFNPP System Lesson Plans have conflicting isolation and actuation setpoints for RPV water level.
Several JAFNPP System Lesson Plans were presented in outline form;
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that is, they' contained headings without supporting tex ;
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Summary of Facility Comments and Commitments The JAFNPP Operator Training Department stated that ongoing efforts to improve the ' trainings material would continue, and requested a complete listing of all ~ deficiencies identified by,the NRC.
The -licensee : felt that 'the written examinations were fair and challenging and that the examiners displayed professionalism throughout the evaluation period.
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The licensee stated that in accordance with JAFNPP ITP-5, the JAFNPP Oper-ations ' Superintendent would evaluate prior to their return to licensed-duties.in the Main Control Room,' the watch standing capabilities. of the license ' holders identified as potential failures of the.NRC-administered operating exam.
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3.
NRC-Follow-up During' a telephone conference conducted on May 13, 1987, - the licensee-reported that both license holders identified as potential failures of the NRC-administered operating examinations had passed subsequently the exam-ination : administered by the Operations Superintendent. ' In addition, the NRC verified that.all identified performance deficiencies had ' been ade-quately evaluated by the licensee.
In a telephone conversation to Mr. Converse conducted on May 15, 1987,
Mr. Lange reported that an SR0 license holder had failed.the NRC written i
examination. This failure posed no immediate safety concerns because.the license' holder is not an on-shift operator at JAFNPP.'
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' Upon completion of the NRC Regional review of the written examinations, the results were reported to the Resident Manager in a telephone conversa-tion conducted on May 20, 1987.
The graded examinations were mailed.to the Resident Manager on May 21, 1987.
The NRC requested him to review the exam results and take. action in accordance with ITP-5 for all license holders who failed the written examination.
Mr. Lange informed Mr. Converse that no results of the JAFNPP Requalifica-I tion Program evaluation could be given at that time, but that the final report was expected to be issued in June, 1987.
4.
Conclusion Of the seven (7) Senior Reactor Operators evaluated by the NRC during the Requalification Program evaluation, only one (1) failed the written exam-ination.
(Due to time constraints, the NRC did not administer an opera-ting examination to this particular license holder.) All remaining SR0s took both the NRC written and operating examinations and passed all por-tions of these examination I
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The SRO failure on the written examination indicates an individual weak-ness, For the remaining SR0s, the NRC identified no generic weaknesses in either the written or operating examinations. The overall average for the SR0 written examination was 87.3%
All of the three (3) Reactor Operators evaluated by the NRC failed the NRC requalification evaluation.
Two (2) failed the written examination; the remaining R0 failed the operating examination.
q The group performed poorly in certain areas of the written examination; the average.for the R0 written examination was 78.2%. However, the NRC
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identified no generic weaknesses in administering the operating examina-q tions to the R0 license holders.
The following table lists specific weak i
areas identified from the R0 written examination results.
i Question No.
Question Topic Area Class Average
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1.02 Critical Power Ratio'
55.5%
2.02 RHR - Shutdown Cooling Mode 66.6%
2.04 Operation of EHC 63.3%
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3,01 RPV Level Instrumentation 66.6%
3.05 Rod Sequence Control System 62.5%
3.06 Feed Water Control System 56.6%
4.03 F-A0P-46 " Loss of 'B'
DC Power System" 52.7%
4.05 F-0P-65 "Startup and Shutdown Procedure" 66.6%
F-0P-20 " Condensate System" The NRC staff reviewed the quality of the training material submitted by the facility licensee and found many areas of lacking or incorrect infor-mation. A detailed list of discrepancies is contained in Attachment 3.
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The licensee should evaluate the examination results and establish a cor-
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rective action plan and schedule.
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Attachments:
1.
Written Examination and Answer Key (SRO)
2.
Written Examination and Answer Key (RO)
3.
JAFNPP System Lesson Plans Reviewed by NRC
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liASTE
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S.-' NUCLEAR REGULATORY = COMMISSION
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. SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY:
_F__IT_ZP_A_T_R__IC_K______________
_
REACTOR-TYPE:
_@WR-GE4_________________
DATE ADMINISTERED: _@Zf9@f9Z________________
-EXAMINER:
_K9L9NAU@Kiz_Lg__________
CANDIDATE:
_________________________
IN@lBU9119N@_19_C$Npip@IE1
.
Road the attached instruction page carefully.
This examination replaces the current cycle facility administered requalification examination.
Retraining requirements for failure of 'this examination are the same-as
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for failure of a' requalification examination prepared and administered by your training-staff.
Points for each question are indicated-in parentheses after the question.
The passing grade requires at least 70%
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in each category and a final grade of at least 80%.
Ex ami nati on papers I
.will be picked up four (4) hours after the examination starts.
-% OF CATEGORY
'% OF CANDIDATE'S CATEGORY
__yeLUE_ _IQ1@L
___@C96E___
_y@LUE__ ______________C@ LEG 96Y_____________
_19199__ _29199
__ _____ 5.
THEORY OF NUCLEAR POWER PLANT
___________
OPERATION, FLUIDS, AND THERMODYNAMICS
_19199__ _@9199. ___________
________ 6.
PLANT SYSTEMS DESIGN, CONTROL,
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AND INSTRUMENTATION l
_19199__ _29:99
________ 7.
PROCEDURES - NORMAL, ABNORMAL,
.
___________
EMERGENCY AND RADIOLOGICAL l
CONTROL-j
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25.00
________ B.
ADMINISTRATIVE PROCEDURES,
15.00 CONDITIONS, AND LIMITATIONS l
69t99__
________%
Totals
___________
Final Grade i
All work done on this examination is my own.
I have neither given i
nor received aid.
___________________________________
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
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- During the administration of this examination the following rules apply:
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Cheating on the examination means an automatic denial of your application
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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 gely 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 f or answers.
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 ggw page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least thtgg linet 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 litetatute.
13. The point value for each question is indicated in parentheses after the
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question and can be used as a guide for the depth of answer required.
Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examingr 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 complete. -,
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18.. Wben you complete your ex aminati on, you shall:
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a..
Assemble your examination as follows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3)
Answer pages including figures which are part of the answer.
)
b.
. Turn in' your copy.of the examination and all pages used to answer.
the examination questions.
c.
Turn in all scrap paper'and the balance of the paper that you did not use-for answering the questions.
L d.
-Leave the examination area, as defined by the examiner.
If after l eavi ng, you 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 (2.00)
JAFNPP Unit 1 is in hot shutdown. The Recirculation pumps are secured for 4160 volt bus maintenance. RHR Shutdown Cooling is in service.
f A review of Control Room instrumentation indicates the f ollowing conditions:
i Bottom Head Temperature (RWCU) =
153 deg F Reactor Pressure =
55 psig j
Vessel Level =
200" i
s.
In view'of these RPV parameters, what problem is indicated?
(0.75)
b.
What corrective action is required?
(0.75)
c.
If RPV pressure continues to riste, what automatic action is expected?
(0. 5)
QUESTION 5.02 (3.00)
a. Given the f ollowing two conditions and using attached. Figure 1,
determine which condition is operating MORE closely to its operating MCPR limit. Show all work and state all assumptions.
(2.0)
Condition 1 Condition 2 Rx pressure = 920 psig Rx pressure = 980 psig Core flow = 50 %
Core flow = 75 %
Rx power = 1220 MWt Rx power = 1830 MWt P-1 MCPR = 1.60 P-1 MCPR = 1.55 NOTE:
Assume that the operating MCPR limit as given by TS 3.1.B.1 is 1.27 for both conditions.
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b.
What is the primary basis f or the steady state MCPR operating
limit?
(1.0)
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QUESTION 5.03 (2.50)
JAFNPP is brought critical at 50% on range 2 with a stable positive
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period of 45 seconds. The. point of adding heat has been determined to be'50% on range 8 of'the IRMs.
c. What' is the expected doubling time if period remains constant?
(1.0)
b.
How long will it take for power to reach the point of adding (1.5)
heat if period remains constant?
QUESTION 5.04 (2.25)
Reactor power has been at 50% for the past three days. Power is now raised to 100%.
Five (5) hours after reaching full power, HOW will each of the following parameters change (i e., INCREASE, DECREASE, or NO CHANGE)
compared to their initial values UPON reaching 100% power?
i Assume no operator action. EXPLAIN the reason for each change briefly.
. Values are not required.
a. Reactor power (0.75)
6.' Megawatt output (electrical-MWe)
(0.75)
c.
Core flow (0.75)
QUESTION 5.05 (1.50)
i Even.if the Main Condenser and associated systems were absolutely j
eir tight, the SJAEs would still be required to remove other non-condensibles from the main condenser.
List two (2) additional sources of noncondensibles to the main condenner other than air in-leakage.
(1.5)
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QUESTION 5.06'
(2.50)
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s c. If. a BWR core is inadequately cooled, hydrogen will be produced.'
j List THREE (3) sources of hydrogen from a degraded BWR core.
(1.0)
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. b. Li st T)#MI (.Z-I 7) methods used to' reduce the hydrogen concentration
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in Primary Containment;&t JAFNPP.
(1.5)
I QUESTION 5.07 (1.25)
, Attached Figure 2 shows the'resulting peak fuel enthalpy f or a rod
. rod drop accident.
Explain why aJrod drop occuring at 3% of rated power (Curve 1)
,
-rosults inLa' HIGHER peak fuel enthalpy than one occuring at-l
30% of rated power (Curve 2),
(1.25)
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-QUESTION 6.01 (2 00)
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D3 scribe the interrelationship between the Rod Worth Minimizer and the following systems:
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inputs to the RWM and outputs from the RWM)
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a.
Feedwater Level Control System (0.67)
.b.
Reactor Manual Control System (0.67)
c.
Rod Position Indication System (0.67)
QUESTION 6.02 (2.50)
Concerning the. recirculation pump speed mismatch limitations (refer to cttached Figure 3):
a.. WHAT is-the' PURPOSE of-the mismatch operating limitations?
~ (1.0)
b.
FILL-IN THE BLANK When operating one pump, the idle pump must not be started until running pump speed is ________________ percent.
(0.5)
c.
Is operation allowed in the " PROHIBITED" regions (crosshatched regions) at any time?
Briefl y EXPLAIN.
(1.0)
QUESTION 6.03 (2.00)
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a. While performing a RCIC full flow test, a Reactor Operator (RO)
controls RCIC turbine speed with the RCIC Turbine Test Potentiometer.
I During the test, RCIC automatically initiates on reactor vessel low-low level.
Assume reactor vessel level has been restored. In order to avoid a RCIC trip on high reactor vessel level, the RO attempts to y
control RCIC turbine speed using the test potentiometer.
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Will he be able to control RCIC turbine speed using this method?
Explain your answer.
(1.0)
b.
Assume that the
"A" 125V DC bus is lost. Explain RCIC system response if an automatic RCIC initiation signal is then received.
(1.0)
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QUESTION 6.04 (2.25)
j Each of the conditions listed below initiate an automatic response in I
the Reactor Water Cleanup System.
For each of the conditions give the SETPOINT of the initiation signal and the PURPOSE of the RWCU system response.
c. Low Pressure Upstream of the Discharge Flow Control Valve (0.75)
b.
High Pressure Downstream of the Discharge Flow Control Valve (0.75)
c.
Filter /Demineralizer Effluent Low Flow (0.75)
I QUESTION 6.05 (2.25)
Concerning the Standby Gas Treatment System:
c. What conditions give an auto initiation of the system?
Include setpoints.
(1.50)
b.
How does the system respond to an initiation signal?
Include expected system flowrate and reactor building differential pressure.
(0.75)
QUESTION 6.06 (1.00)
The HPCI system auto initiated on low reactor water level.
The cystem operated normally f or two minutes then the turbine tripped.
State whether the f ollowing statements concerning the HPCI turbine trip are TRUE or FALSE.
a.
If the turbine trip was due to turbine overspeed the turbine WILL NOT auto restart when the trip condition clears even if an auto initiation signal is present.
(0.5)
b.
If the turbine trip was due to an auto isolation of the HPCI
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system the turbine WILL auto restart when the system isolation I
is reset if an auto ini ti ation signal is present.
(0.5)
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b QUESTION 6.07 (3,ooy (
A loss'of all site AC power _ including UPS (i. e. Station Blackout)
has occurred.
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c.
What reactor water level indication (s) are available in t'he
. control room f ollowing this event?
(1.0)
b.
What reactor water level indication (s) are available outside the control room f ollowing this event?
' ( 1. 0)
c.
Why does F-EOP-4 require RPV Flooding if drywe11' temperature reaches the RPV saturation temperature corresponding to RPV pressure?
( 1. 0 )
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699196991G96_QQN189(
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QUESTION 7.01 (3.00)
c.
For each of the following Radiation Work Permits (RWPs), give
its duration limit.
(1.0)
1.
Regular RWP 2.
Special RWP
3.
Continuing RWP 4.
Extended RWP b.
List three (3) conditions that could require a Special Radiation Work Permit.
(1.5)
c.
TRUE or FALSE?
A Continuous RWP can be issued for conditions requiring a Special RWP with the approval of the Health Physics General Supervisor.
(0.5)
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QUESTION 7.02 (1.50)
F-AOP-12, " Loss of Instrument Air", cautions that various conditions may occur due to the decrease in air pressure.
Explain how the following systems would respond if the air supply was completely lost to system components.
Include all system components that would respond (i.e.,
valves, controllers, etc) unless otherwise noted.
c.
Main Steam (0.50)
b.
Control Rod Drive Hydraulics (excluding scram valves)
(0.50)
c.
Off-Gas (0.50)
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QUESTION 7.03 (2.00)
c.
What requirements are placed on reactor pressure during a hot.
startup per F-OP-65, "Startup.and Shutdown Procedure", so that j
reactor vessel level can be controlled?
(1.0)
b.
F-OP-20, " Condensate System", cautions that condensate pump 33-P-BC and' condensate booster pump-33-P-9C should be the
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1ast pumps turned on and the'first pumps turned off in their
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respective group.
What is the reason for this caution?
(0.5)
c.
F-OP-20 also cautions that no more than one condensate pump
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i can be operated when the downstream system is " deadheaded".
What is the reason for this caution?
(0.5)
QUESTION 7.04 (1.00)
Under what conditions may the Core Spray pumps be stopped or the Core Spray. inboard injection valves be throttled af ter an automatic initiation?
(1.0)
QUESTION 7.05 (1.00)
State whether the following statements concerning Rx Analyst Procedure 7.1.3, Refueling Procedure are TRUE or FALSE.
If the statement is FALSE briefly explain WHY it is false.
c.
When fuel is being moved to, from,.or within the Rx vessel a licensed RO shall be stationed in the Control Room in direct communication with the Refueling Floor.
He shall not have'any concurrent duties other than refueling duties.
(0.5)
b.
A fuel array of 3 fuel bundles may be outside of a normal storage area or their normal shipping containers as long as an edge to
,
edge spacing of 12" or more is maintained between them and all other fuel.
(0.5)
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QUESTION 7.06 (2.50)
SRY The reactor is at 100% power and it is suspected that an JMMf7 i s st uc k open.
Concerning F-ADP-36, " Stuck Open Relief Valve";
SRV-c.
If no position indicating lights f or the SOMY on panel 09-4 are
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lit, what are-FOUR (4) other control room instrument indications, including back panels that you could use to verify that the associated Relief Valve is stuck open?
(1.0)
b.
Having unsuccessfully attempted to shut'the stuck open Relief Valve by cycling the valve control switch on panel 09-4, what
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other methods can be used to shut the valve ?
Give THREE (3).
(1.5)
l i
QUESTION 7.07 (1.00)
)
F-EOP-4, " Primary Containment Control", requires operation of
Standby Gas Treatment to control primary containment pressure but prohibits SBGT operation when temperature in the space being
)
ovacuated is at or above 212^F.
What is the purpose of this restriction?
(1.0)
s
4 i
(*****
CATEGORY 07 CONTINUED ON NEXT PAGE
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4,
7.'
. PROCEDURES NORMAL _ABNQRMAL _ EMERGENCY _AND PAGE
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.Be91969GIC86_QQNI6QL
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QUESTION-7.08 (3.00)
c. Given Table F-EOP-2.1 (Figure 4 attached), state whether Emergency RPV Depressurization is REQUIRED or NOT-REQUIRED under each of the f ollowing situations.
Consider each situation separately.
(2.0)
1.
RPV water level is 0" and decreasing, RPV pressure is 350 psig and decreasing, HPCI and RCIC are available, LPCI-A and CS-A are lined up for injection with pumps running.
2.
RPV water level is 0" and increasing, RPV pressure is 350 psig and. increasing, HPCI and RCIC are unavailable, LPCI-A and CS-A are lined up for injection with pumps running.
~
3.
RPV water 1evel11s 0" and decreasing, RPV pressure is 50 psig and increasing, HPCI and RCIC are available,.LPCI-A and CS-A are lined up for injection with pumps running.
4.
RPV water level is 0" and increasing, RPV pressure is 50 psig and-increasing ~ HPCI and RCIC are unavailable, LPCI-A and CS-A.
,
are lined up for injection wi th pumps running, b.
What are two (2) purposes of Emergency RPV Depressurization?
(1.0)
(***** END OF CATEGORY 07
- )
y e _3epdlNISIBSIlyE_E8pCEDUBE@t_CQNpl11pNS _9NQ_(Id11gIlpNg PAGE 1'2 z
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...
.
-QUESTION 8.01
.
.
(1.50)
According to JAFNPP Plant Standing Order #53, " Post Trip Evaluation":
c. WHO is responsible f or implementing this procedure after any unplanned plant' shutdown?
(Give position title.)
(0.5)
b.' What plant personnel are required to attend the post trip
-
critique?
(0.5).
c.
WHO is responsible f or ensuring that the personnel listed for part
"b" attend the post trip critique?
(Give position title.)
(0.5)
-
QUESTION 8.02 (1.00)
Fill in the following blanks with the appropriate position title as given in JAFNPP Operations Department Standing Order #3,
" Procedure for Temporary Operating Procedures",
" __________________________ shall be responsible f or the Temporary Operating Procedures."
(0.5)
"Saf ety related temporary operating procedures shall be reviewed by l
PORC and approved by ___________________ prior to implementation. "
(0,5)
.
I (*****
CATEGORY 08 CONTINUED ON NEXT PAGE
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_
i ac__09 MINI @I6811yE_PBQGEQUSES _QQNQlligNS _@NQ_(lMlI@IlgN@
PAGE
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QUESTION 8.03 (2.00)
l l
Answer the following questions in accordance with JAFNPP Operations Department Standing Order #1, " Operating Staff Responsibilities and Authorities".
l Answer each as TRUE or FALSE. If False, explain WHY.
(2.0)
l l
c.
In operating the JAFNPP Unit, the Shift Supervisor is the only person who may assume " command reponsibility and authority".
b.
Assume that the Shift Supervisor leaves the control room during routine operations and designates the Assistant Shift Supervisor as having control room command.
The Assistant Shift Supervisor may not alter plant condi tions without verbal consent from the l
Shift Supervisor.
'
c.
The Assistant Shift Supervisor has the authority to shutdown the reactor if he determines that the reactor is in Leopardy.
l
QUESTION B.04 (2.00)
It is discovered today, May 7, 1987, at 0800, that a monthly eurveilllance item which was due Friday, May 1,
during the
]
midnight shift (0000-0800), was NOT performed.
I I
In addition, you learn that the last monthly surveillance j
for this item was performed six (6) days late.
)
It was performed on time f or all months previous to last month.
I Hcs the extended time interval for this surveillance as given
{
by the JAFNPP Technical Specifications been exceeded?
i Explain your answer.
(2.0)
j l
l I
<
{
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. - _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ -
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'PAGE -14 t
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QUESTION 8.05 (3.00)
i c. According to the JAFNPP Technical Specifications, is it PERMISSIBLE to go f rom STARTUP to RUN if IRMs
"A",
"B",
and
"C" have been declared inoperable?
Explain your answer.
(1.0)
i b.
If the same IRMs were found inoperable while in RUN, would it be a violation of Tech Specs if yous 1.
stayed in the RUN mode?
Explain.
(1.0)
2.
placed the mode switch in STARTUP?
Explain.
(1.0)
USE THE ATTACHED TECHNICAL SPECIFICATIONS TO DEVELOP YOUR ANSWER.
BE SURE TO REFERENCE EACH APPLICABLE TECH SPEC LCO AND ACTION STATEMENT BY NUMBER.
QUESTION 8.06 (2.00)
The plant is at 75% power and is proceeding to f ull power in cccordance with F-OP-65.
The. Shift Supervisor is then notified that the motor operator for the HPCI Outboard Steam Isolation valve (23-MOV-16) has been discovered as incapable of moving the valve to the f ully closed position.
The decision is made to open the breaker for the valve and manually close the valve.
Con the power ascension continue to f ull power? Give ALL applicable Tcch Spec LCOs and action statements used to develop your answer.
USE THE ATTACHED TECHNICAL SPECIFICATIONS TO DEVELOP YOUR ANSWER.
BE SURE TO REFERENCE EACH APPLICABLE TECH SPEC LCO AND ACTION STATEMENT BY NUMBER.
!
(***** CATEGORY 08 CONTINUED ON NEXT PAGE
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QUESTION B.07 (1.50)
The JAFNPP unit is operating at 100% power when all MSIVs close.
A review of the alarm printer shows that the reactor scrammed on APRM high flux.
Have any Tech Specs been violated because of this event?
Explain.
(1.5)
USE THE ATTACHED TECHNICAL SPECIFICATIONS TO DEVELOP YOUR ANSWER.
BE SURE TO REFERENCE EACH APPLICABLE-TECH SPEC '.CO AND ACTION STATEMENT BY NUMBER.
QUESTION 8.08 (2.00)
i
~JAFNPP Unit 1 is operating at 100% power. Core Spray pump
"A" was determined to be inoperable six (6). hours ago.
c. What actions would be required per the Tech Specs if the
"B" Emergency Diesel Generator was then found to be inoperable in addition to Core Spray pump
"A"?
(1.0)
b.
What actions would be required per the Tech Specs if INSTEAD of the
"B" E D/G, Core Spray pump
"B" was then found to be inoperable?
(1.0)
!
!
j I
USE THE ATTACHED TECHNICAL SPECIFICATIONS TO DEVELOP YOUR ANSWER.
BE SURE TO REFERENCE EACH APPLICABLE TECH SPEC LCO AND ACTION STATEMENT BY NUMBER.
t
!
(*****
END OF CATEGORY 08
- )
(************* END OF EXAMINATION ***************)
'
r'
,' $i;; THEORY'OF" NUCLEAR POWER PLANT OPERATION _ FLUID @z_ANDL PAGE
x TH_ERM_O_D_YNA_M_ICS
_ __ _
_ _ __
. ANSWERS'-- FITZPATRICK-
-87/05/07-KOLONAUSKI, L.-
im gc.,
' ANSWER 5.01 (2.00)
'l i
I E
'a.
Thermal stratification - delta T = 302.93 - 153 = 150.deg F-(0.75)
l l
b.
Maximize natural' circulation by raising water ' level to 234.5"
{
as indicated on the Refueling GEMAC.
(0.75).
j j
c.
SDC isolates when L RPV pressure reaches 75 psig increasing.
_.
[10-MOV-17, 18,.32, 33 close and.RHR. pumps trip 3 ( 0. 5 )..
,
.
.
!
REFERENCE JAF. Transient & Accident Analysis TAA 101.2 Enab Obj 1.'7 Core Coolant Temperature increase j
RHR LP SDLP-10 F-OP-13..pages 12-15 l
295021 Loss-of SDC i
K3.01 knowledge of reason f or. raising Rx water ~1evel 3.3/3.4
.i
'K1.02 Operational implication of. thermal stratifiction 3.3/3.4
'
-
295021K102" 295021K301
...(KA'S)
I J
ANSWER 5.02 (3.00)
a. For. condition 1:
For' condition 2:
Kf = 1.11 Kf'=
1.0-MCPR =-(1.27)(1.11)
=-1.41 MCPR ='(1.27)(1.0)
1.27
=
dMCPR.= 1.60 - 1.41 = 0.19 dMCPR = 1.55 - 1.27 = 0.28 (1.5)'
j So Condition 1 - i s c1'oser.
(0.5)
b.
The operating MCPR' limit provides adequate margin to OTB (onset of-transition boiling)1such that during any analyzed transient, the resulting delta MCPR would not result in OTB.
(1.0)
!
REFERENCE JAFNPP-HT LP, Thermal Limits pages 31-35
- Enab Obj 228.9.1.23, 228.9.1.21 JAFNPP=TS page 31, 47a 293009 Core Thermal Limits K1'.19 Explain basis of LCO for MCPR 2.8/3.6 K1.27 Explain purpose of Kf for MCPR 2.7/3.3 293OO9K119 293OO9K127
...(KA'S)
...
($1_,ThEORYOFNUCLEARPOWER' PLANT _QPERATigN_FLUIpS_AND PAGE
.,.
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- ANSWERS-- :FITZPATRICK-87/05/07-KOLONAUSKI, L.
j
,,
>
cANSWER 5.03 (2.50)
c. P = Po>e**(t/T)
.for doubling time, P = 2 Po-T = 45 see t'= T'In(P/Po)
T in 2,
=
= 45 (0.69)
31.2 sec (1.0)
=
b.
50% on range 2 = 0.05% on range B Po = 0.05 Pt = 50 T = 45 sec.
Pt = Po e**(t/T)
T In (P/Po)
t =
= 45 sec in(50/O.05)
310.8 sec = 5.2 min (1.5)
=
REFERENCE JAFNPP Rx Heatup and Power Range Ops NET 237.8, page 5 Enab-Obj-237.8.1.1.c 292003 Rx Ki net i'c s K1.08 Solve for changes in period and power with period eqn 2.7/2.8 292OO3K108
...(KA'S)
l ANSWER 5.04 (2.25)
(gg p. Xe dXhh 4-6 h(5. )
]
l a.
INCREASES (0.25), the increased Xe burnout would add posi tive i
reactivity which would cause a power increase (0.5).
I b.
INCREASES (0.25), MWe increases due to the power increase (0.5).
c.
DECREASES (0.25), as power increases, two phase flow increases which decreases core flow due to increased flow resistance (0.5).
REFERENCE JAFNPP Instructor Lesson Plan-Fission Product' Poisons NET 237.5 Enab Obj 237.5.1.1, 237.5.1.2 pages 6-10 292006 Fission Product Poisons K1.04 Describe the removal of Xe-135 2.9/2.9 K1.06 Describe maneuvering Xe effect on Rx Ops 2.7/2.7 292OO6K104 292OO6K106
...(KA'S)
5:_ THEORY'dF' NUCLEAR' POWER PLANT _QPERATiQN _FLQIQS _ANQ-PAGE
~
-
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,
.IHERdQQyN@ digs Y-.
ANSWERS.-
FITZPATRICK-87/05/07-KOLONAUSKI, L.
.
- =
ANSWER 5.05 (1.50)
' Activation gases from'the irradiation of reactor coolant.
~
Fission gases produced in the fuel will leak to the coolant.through'
' cladding' defects.
Ent' rained water' vapor.
(Any two, 0.75 each)_
REFERENCE-
.JAFNPP Offgas Lesson Plan SDLP'01A page 12'
Enab Obj
' LOR-01A-1.01, l'.02 291006 HX.and Condensers K1.18 LReasons<for noncondensible. gas removal 2.8/2.9-291006K118
...(KA'S)
ANSWER'
5.06 ( 2. 50 )'
a. Zirconium-Water (Any three, 0.33 each)
-Steel-Water Zinc-Water Concrete-Water Radiolytic decomposition of water pgm 0.75'
.b. ; Add any other gas ( % th g, Of each)
(N2, CO2, etc.)
Thermal recombiners H2 burners ggggy Containment igniters VVent / Purge. operations
, REFERENCE JAFNPP Mitigation of Core Damage, 301.2 Gas Generation pages 4-14
'Enab;Obj 301.2.1.1, 1.7
223001 Primary Containment K4.04 Design features for avoiding H2 explosive mixture 3.5/3.8 223OO1K404
...(KA'S)
i I
'
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ETHEORY-OF' NUCLEAR' POWER PLANT OPERATION _ FLU 1DS _AND PAGE
t x
'ISEBdQDYN@DIQS ANSWERS --LFITZPATRICK-87/05/07-KOLONAUSKI, L.
..
!
!
. ANSWER 5.07 (1.25)
!
!
Increased power causes increased core voiding (which flattens the
!
.pcak to average. flux ratio surrounding a ' control rod.) The reacti vi ty
)
change will be less pronounced and would result in lower local l
host generation.
(1.25)
l
REFERENCE JAFNPP Transient and Accident TAA-101.3 Section~1.1, Enab Obj 1.1.
'
RWM LP SDLP-03D Section II.f.2),
Enab Obj 1.02 201006 RWM K5.02 Operational implications of LPSP 2.9/3.0
.201006K502
...(KA'S)
I i'i i
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..
.
.......
.
.
.
.
.
...
.
.
.
.
L.
l6:_gPL@NI,@lSIEd@,pE@l@N _CQN169L _@NQ_1N@l69dENI@llgN PAGE
'20 '
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ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
.
.
ANSWER 6.01.
(2.00)
c. Total steam flow input (0.67)
b.
Rod drift input, withdraw and insert block output 5(ltchd (Od InfWk (0.67)
c. ~ Rod position [and selected rod inpu{vaghC S (rod drift input via RMCS)) (0.67)
,
REFERENCE JAFNPP Instructor Lesson Plan SDLP-03D, " Rod Worth Minimizer", Rev 1, section VI Learning Objective - SDLP-03D 1.13 K/As - 201006-K1.01 (3.4), 201006 K1.02 (3.4), 201006 K1.03 (3.2)
i 201006K101 201006K102 201006K103
...(KA'S)
bO(h nam S bLY' -039 RM cs - Kur l, pay 9
_ ANSWER 6.02 (2.50)
i o.
To avoid excessive vibration of the jet pump risers (1.0)
b..Less than 50 %
(0.5)
c.
YES. (0.25)
Operation in the prohibited regions is only i
allowed during coastdown when a recirc pump trip occurs during two pump operation. (0.75)
.
L7 gM ORhf St } \\0 y of REFERENCE JAFNPP Instructor Lesson Plan, SDLP-402I, "Recirculati on Flow Control",
Learnin Obj c v
- SDLP-4021 1.07 K/A - 202001 K4.11 (3.5)
202OO1K411
...(KA'S)
h_ P($NI_@f@lEdg_QEg1GN _CQNI6QL _@NQ_lN@IBUdENI@IlgN PAGE
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ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
.
.
ANSWER'
6.03 (2.00)
'c.
No (0.25). when-an automatic initiation signal is received, the RCIC turbine test potentiometer is removed from the turbine
,
control circuit (0.75).
4D WC4 C Cey0M.(gh uAi.( k N0 CCLC KffXm.(0.6)
0.6)
.
b.
125V DC
"A" supplies the DCIC ircer.cr.
'5_
i r c e r_t s r supplies-the-j f'C I C est a f l aw -i nd i c e+ i nn ran+raller '13 IIC-71). (G.3)
The-RC-It systee wi11
=&=rt-h'+
will F e-i n an av=r= peed. ' O. 5 )-
REFERENCE-i JAFNPP RCIC SDLP-13, pages 23-38/ Enab Obj 13-1.10, 11, 12 I
F-OP-19 p g,49 9 d
"A" I7C f0Wt r fyb ", ybr 0 ye) 6 217000 RCIC K2.03 KN of elec power supplies to RCIC FC 2.7/2.8 K6.01 KN of loss of-elec power effect on RCIC ops 3.4/3.5 A1.05 AB to predict impact of power loss on RCIC ops 3.7/3.7
!
217000A105 217000K203 217000K601
...(KA'S)
ANSWER 6.04 (2.25)
gmh (gen-gg( d h l) d.L h k& b s. 5 pri;;
'O.25),
prevents condenser vacuum from draining the RWCU system (0,Af{g yg )
b. 1-40 poig (A<"25'), protects the low pressure piping to radwaste and the main condenser (OMM
,
(03%
( 0 16)
c.
60 ;pr
'O.25),
retains the resin on the filter elements ( 0,AI).
REFERENCE JAFNPP Instructor Lesson Plan SDLP-12, " Reactor Water Cleanup System",
Rev 1,
pg 15, Table II Loarning Objectives - SDLP-12 #23 & #15 K/A's - 204000 K4.02 (2.9), 204000 K4.07 (2.9)
204000K402 204000K407
...(KA'S)
l
. - _.
...
...
-
.
- _ _ _ _ _
6 _,PL@N1_SYSIEdg_gESIGN _CgNIBgLi_8Np_1NSIBgdENISIlgN PAGE
z
,
ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L..
.
.
. ANSWER 6.05 (2.25)
c. - High-drywell pressure - 2.7 psig
- Low RPV water level 177 inches
-
- High rad levels'in Rx b1dg exhaust plenum-
- 10,000 cpm or detector INOP l,
- High. rad levels in L
the refuel floor L
exhaust plenum 10,000 cpm or detector INOP
-
- Auto. initiation of HPCI
- (126.5 inches RPV level or
'.
l 2.7 psig drwell pressure)
(0.3 each)
b.
Both fans start and all. appropriate valves in each filter train open, motor-operated suction valve (s) open depending on which signal caused the. initiation (0.50).
Rx building dp more negative than ~0.25,with a g
corresponding flow of < 6000 scfm (0.25).
A REFERENCE JAFNPP F-EOP-4, " Primary Containment Control", Rev 1,
pg 17 JAFNPP F-OP-20, " Standby Gas Treatment System", Rev 12, pg 4
-
JAFNPP Instructor Lesson Plan SDLP-01B, Standby Gas Treatment, Rev 1,
pg 18 l
Loarning Objectives - SDLP-01B 1.05 & 1.09 l
K/A's - 261000 K1.02 (3.4), 261000 K4.01 (3.8)
l 261000K102 261000K401
... (KA' S)
ANSWER 6.06 (1.00)
a. FALSE (0.5)
b..TRUE (0.5)
REFERENCE l
JAFNPP Instructor Lesson Plan SDLP-23, "High Pressure Coolant' Injection System", Rev 1, pgs 14 &.15 Learning Objective - SDLP-23 23.04 K/A - 206000 K4.03 (4.1)
206000K403
...(KA'S)
J
_ _ - - _ _ _ - - _ - _ - _ -. - _ _ _ _ _ _ _ _ _ - _ _ _ _ _
,
6r,...fLONI_Sy@ led @_QESIGN _CgNI69!=t_6NQ_lN@l6UdENI@IlgN PAGE
t
,.
.
~' ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
' '
..
gg1(b i f I M C d. O N K M.f ht/> ep m scre prwh>.
(CObridJr Mn
$M uj m p % ]
ANSWER 6.07:
(3.00)
c.
"B"
&
"C" Narrow Range GEMAC on panel 09-5,
'r'-1 7ane CM)
(1.0)
!
!
b.
4. e c. '
"O" a
"C" Norr;.
9= ;; CE."^C,-All Local Yarways,
!
"B" Narrow Range Barton, '" r 1 Zen:
C.^"'-
(1.0)
l
.
!
c.
The reactor vessel level reference legs.will begin to flash when the drywell temperature approaches saturation temperature for the corresponding RPV pressure, (0.5) causing erroneous high reactor l
1evel indication. (0.25)
RPV Flooding is required to assure
_l adequate core cooling. (0.25)
l REFERENCE j
JAFNPP Instructor Lesson Plan SDLP-402B, " Reactor Vessel Level j
Instrumentation", Rev 2, pgs 17-24
JAFNPP F-EOP-4, " Primary Containment Control", Rev 1, pg 11 Learning Objective
.SDLP-402B 1.08 K/A's - 216000 K6.01 (3.3), 216000 K5.07 (3.8)
CAF - see JAFNPP F-AOP-21, " Loss of UPS", Rev 2, pg 3 216000K507 216000K601
...(KA'S)
l
<
,
j f
i
{
i L
- _ - - _ _ _.
-
_ -
.Zz__EB99EgyBES_;_NgBM96t_$@Ng5M861_EME5GENgy_6Np PAGE
.
5891969 GIG 86_GQNIBg6 I
~ ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
l l
ANSWER 7.01 (3.00)
c.
1.
24 hrs (Three 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts max)
2.
One shift (no longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)
3.
One month ( job Arnijm, M mt > l mcwth )
4.
One year (Calendar year)
(0.25 each)
b.
- High Radiation Areas with special controls
- Airborne radioactivity concentrations requiring the use of respiratory protection or MPC-hour tracking
- Surface contamination levels that could result in airborne based on the proposed work
- Unknown conditions
- Neutron radiation requiring the use of neutron dosimetry or exposure tracking (3 required.at 0.5 each)
c.
' FALSE (0.5)
REFERENCE JAFNPP Radiation Protection Manual, Chapter 8,
" Radiation Work Permit" K/A - 294001 K1.03 (3.8)
294001K103
...(KA'S)
ANSWER 7.02 (1.50)
a. Outboard MSIV's drift closed (0.25)
Inboard MSIV's drift closed and air is lost to the SRV's if drywell is on instrument air (0.25)
b.
CRD flow control valves fail closed (0.25)
SDV vent and drain valves fail closed (not listed in F-ADP-12) (0.25)
c.'Offgas discharge valve (01-107-ADV-100) fails closed (0.5)
( M0k-OHy > be ctdMttmd A0Vs)
JAFNPP F-ADP-12, " Loss of Instrument Air", Rev 4, pg 3 LGarning' Objective - SDLP LORT 1.10 K/A - 295019-System Generic #8 (4.1)
295019GOOB
...(KA'S)
- - __- _ __-_ _ __-___-_--
- - - - _ _ - _ - _ - _ _ _ _ _ _ _ _ - _
_
_
-
2 _iPRQgEggBE@_;_Ng6d@bt_@@hg6d@bt_EDE6GENgy_@NQ PAGE
.
809196901Geb_GQNI6QL ANSWERS - ;FITZPATRICK-87/05/07-KOLONAUSKI, L.
.
ANSWER 7.03 (2.00)
a.
The pressure in:the reactor vessel will.have to be controlled below that'which can be overcome by condensate booster pump pressure, [0.53 since no steam will be available for the feedwater pump turbine (until approximately 700 psig). CO.53 b.
The pumps are powered by 4KV bus (10700) which is powered by the main-generator.
No power will be available unless the main generator is on line and. house service transfer has taken place. (0.5)
c.
Minimum flow conditions are saf e f or only one pump.
(0.5)
REFERENCE JAFNPP F-OP-65, Startup and Shutdown Procedure, Rev 36, pg 16 F-OP-20, " Condensate System", Rev 12, pg 5 K/A's - 256000 K1.13 (3.5), 256000 K2.01 (2.8), 256000 K4.03 (2.8)
256000K113 256000K201 256000K403
...(KA'S)
i ANSWER-7.04 (1.00)
.The requirements of F-EOP-2 are met. (1.0)
,
'Varify by two (2) independent indications (0.2) that
- auto initiation was/is not required (0.4)
- adequate core cooling is assured (0.4)
REFERENCE JAFNPP F-OP-14, " Core Spray System", Rev 11, pg 6 F-EOP-1, "EOP Cautions, Rev 2, pg 21 K/A - 209001 System Generic #10 (3.6)
209001G010
...(KA'S)
i
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_ _ _ _ - - - - - _ - _ - _.
_ _ _ _ - _ _ -.
.. - - - _ - _ -..
I.
..
.
..
.
.
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,
7. PROCEDURES - NORMAL ABNORMAL ;EMERGENQy_ANQ PAGE. 26.
t t
Be9196991G86_GQN16Q6
'ANSWERSL-- FITZPATRICK-87/05/07-KOLONAUSKI, L.
,
l1 i
ANSWER 7.05 ( 1. 00 )'
a. TRUE
.b.
TRUE (0.5 each)
{
l i
)
. REFERENCE JAFNPP.Rx1 Analyst Procedure 7.1.3, Refueling Procedure, Rev 10, pg 3,9 K/A - '294001 A1.03 (3.7)
294001A103
...(KA'S)
.
ANSWER 7.06 (2.50)
c. - Acoustic monitor (alarm)
'
Tailpipe temperature (alarm)
- Generator electrical output (decrease)
Torus temperature (increase)
- RPV pressure and water level (oscillations)
(any 4 at 0.25.each)
b. - Pull the. control power fuses for the S/RV(at panel 09-45)(relay room)
'i
- Cycle the control switch for the SpRV at remote relief panel (02-ADS-71)
- Open the power supply circuit breaker inside the remote relief' panel (0.50 each)
'
UM M prti Mtt.
~
I
@ yf((j (
REFERENCE
- uhtd)
3rt JAFNPP F-AOP-36, " Stuck Open Relief Valve", Rev 1, pgs 2,3 Lsarning Objective - SDLP - 4029 1.04 K/A - 239002 A2.03 239002A203
...(KA'S)
I i
,
Zz__EBgggDUBEg_ _NQBd@6t_6BNQBd@kt_Edg6GENQy_@NQ PAGE
,
'8891969@lGOL_G99I696
'
-
' ANSWERS -- FITZPATRICK-
-87/05/07-KOLONAUSKI, L.
.
)
ANSWER 7.07 (1.00)
-j l
l SBGT operation. evacuates non-condensible gases potentially leaving j
o saturated steam environment (0.5).
Subsequent rapid cooling of
the space may result in exceeding design negative pressure leading
.)
.to containment failure. (0.5)
'
- REFERENCE JAFNPP F-EOP-4, " Primary Containment Control", Rev 1, pg 17
'JAFNPP F-OP-20, " Standby Gas Treatment System", Rev 12, pg 4 j
~JAFNPP Instructor Lesson Plan SDLP-01B, Standby Gas Treatment, Rev 1, pg 18 j
Learning Objectives - SDLP-019 1.05 & 1.09 K/A's - 261000 K1.02,(3.4), 261000 K4.01 (3.8)
261000K102 261000K401
....(KA'S)
,
ANSWER 7.08 (3.00)
a.
1.
REQUIRED 2.
NOT REQUIRED 3.
REQUIRED 4.
REQUIRED (0.5 each)
b.
1._ Maintain adequate core cooling (allow low pressure injection)
(0.5)
2. Preserve primary containment integrity ( fj @ jr, p wi t Hm(t (0.5)
% OMtisyt - Wleet, )
' REFERENCE JAFNPP F-EOP-2, "RPV Control (Boron Injection Not Required)", Rev 1, pg 13 K/A - 295031 EK3.05 (4.3)
295031K305
...(KA'S)
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__
_ _ _ _ _ _. _.
_
_
[s_,epMINI@lB@IlyE_PBQCEQQ6E@t_CQNQ111QU@t_@NQ_LlMil@IlgNg-PAGE 5H3 s
~-ANSWERS --JFITZPATRICK-
-87/05/07-KOLONAUSKI, L.
.-
.
ANSWER O.01
'(1.50)
,
(
'6. Shif t ' Supervisor.
(0.5)
6. '. Operations Shif t crew and other plant personnel involved in maintenance'and testing during the event.
(0.5)
c.
Shift Supervisor
. ( 0. 5) -
REFERENCE ~
'JAFNPP PSD #53, pages 2-3 s
294001. PW Generics A1'.031-Use'of procedures:related'to shift activities 2.7/3.7 i
~294001A103-
...-(KA'S)
,
' ANSWER-8.02 (1.00)
.
-
I The Operations Superintendent (0.5)
The' Resident Manager'
(0.5)
REFERENCE l
FAFNPP Ops Dept PSD #3, page 2 294001~ PW' Generics A1.03.Use of procedures related to shif t ac ti vi ti es 2.7/3.7
'
294001A103
...(KA'S)
j i
l
- - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_;
- _ - _ _. _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
8:._,eQd1Ni@l8@IlyE_P8QCEQUBE@g_CQNQlIlgNS _@NQ_(;1dlI@IlgNQ PAGEl 29 t
.
ANSWERS -- FITZPATRICK.
-87/05/07-KOLONAUSKI, L.
<
,
ANSWER-8.03-(2.00)
c.-FALSE (0.25). The Operations' Superintendent may assume command-responsibility and authority from the Shift Supervisor-(0.75).
b.
TRUE'
. (0. 5 )
c.,TRUE (0.5)
. REFERENCE-JAFNPP Ops Dept SO #1, pages 4-7 (5.4.6, 5.4.8, 5.5.3)
294001 PW Generics 2.7/3.7 A 1. 03. AB to.use procedures and station directives for shift activities 294001A103
....(KA'S)
ANSWER B.04 (2.00)
No.
The extended surveillance time may not exceed 25'/. of the surveillance interval, or-30/4 = 7.5 days. At this time, the surveillance is only
six days late.
(1.0)
Also, the total time for the last three consecutive surveillances may not exceed 3.5' times the surveillance interval. For the last ' three intervals, time = 30 + 36 + 36 = 102. And 3.5 * 30 =.105 days.
(1.0)
pgk - d-y q $ pc g [ g1 p ( b ct g g h REFERENCE JAFNPP TS:4.O.B EhdAdick a.k itT dyM M do1A, Ni)T 5(X dag.
294001 PW Generics kw( Ad htg K1.02 KN of tagging and clearance procedures 3.9/4.5 a
294001AK10
...(KA'S)
l
. _ - - _ _ _ - - _ - _ _ _ _ _ _ - _. - _ _ _ _ _ _ _ _ - _ _ - _ - - - - - - - _ _ - -
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
-
._
_
Eh_,9 MIN 1EI6811yE_P6QCEQUBES _CQNQ111QNSg_@NQ_ LIM 11811QNg
.PAGE..30
!
t
.
' ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
,
e :
'4
<
,.,
.\\~.
I L
ANSWER B.05 (3.00)
!
c. Yes (0.25). TS 3.3.1.a (by Tabl e 3.1-1 Note 1). requires'you to put:
the RPS system " A" in the tripped condition while in startup.
-
~
!
Once in RUN,-there are no IRM RPS requirements (0.75).!
'
l'
l b.
1.
No (0.25), by TS Tables 3.1-1 (RPS) and 3.2-3.(Rod Block), the IRMs are not required in the RUN mode (0.75).
hC 1 2.
Yes (0.25), unless you trip RPS " A" to comply with TS 3.1. A -(0.'75).
.
REFERENCE JAFNPP TS 3.1, 3.2 215003 IRMs SG #5 KN of TS LCOs 3.2/4.2 215003 GOO 5
...(KA'S)-
'
-
ANSWER 8.06 (2.00)
J
,
Bscause HPCI is now inoperable, power operations may continue for noven (7) days provided that ADS, CS, LPCI, and RCIC prove operable
_
,
par TS.3.5.C.1.a (1.0)
l In addressing the Primary Containment Isolation Valve, TS 3.7.D.1-hg etates that all isolation valves listed in TS Table 3.7-1 be
' -49 operable, except as given in TS 3.7.D.2-which is met by closing (Jj the' outboard isolation valve.
(1.0)'
REFERENCE
j
~"
!
k-206000 HPCI j
SG #5 KN of Tech Spec LCOs 3.6/4.3
,
206000 GOOS
...(KA'S)
>
>
'N
i
-_____
-
- _ _ _ -
_
-_ _ i
_ - _ _
- - - - - _ _ _ - _
- - - - - _ _
- -
-
_
,
' "Mb. ' add 1NINI66I1VE_P6QCEQLJ6Eh_CQNQlIlgNh_@ND_LidlI@IlgNS PAGE. 31 e
fM;:!!J ANSWERS -- FITZPATRICK-87/05/07-KOLONAUSKI, L.
y
-
%
,
.
<
!m
'
!
? (,t, <
,.s.
' ',
.
. ANSWER:
B. 07.
(1.50)
-
'q '.l { '-
~
y
,
"
Ycs-(O.3)..A safety limit shall be assumed to be exceeded-when a
,s'
tecram is accomplishpd.by means'other'than the expected scram signal (0.6).
'A scram should have' occurred due to the MSIV valve closure (0.6).
i k'
. h
'
l REFERENCE.
+
i.
JFNPP Tech' Specs 1.1 Fuel Cladding Integrity Saf ety Limits lC
.
.
!
l 212000 RPS?
,
.
SG #5f KN.of'LCOsiand safety limits' 3.8/4.5 212000G005
.(KA'S)
y a
..
l
'
' ANSWER B.OB (2.00)
'
'
!
'a. Per[TS 3.O.E.
with the
"B" E D/G out of service and a component'in
'
' the oWs(Ttrai n' i noperabl e - (the " A" CS pump), JAFNPP must.be in E
.
j h6f shutrJt3wn within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
h., O )
'
.
COLO d
@
b} Peri kS 3. 0.C,] thi s condi tion representd; a cir'cumstance in excess of
'
those id~ dressed by TS. The. unit shall bo' placed in cold shutdown.
L
- ~
-
'
%.
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a g
.
<
'( 1. 0 ) '
'
'
( ' ' l t 6 v,_
'
Cold SID' ip 24h.
35A
-
ch dio s+M LK
.
'
REFERENCE A'
)
'
s
JAFNPP T,S
' '
s
->
.
'
209001 LPCS,
.,
,
,.
<
,
1SG'#5. pl.of. applicable TS LCOs 3.3/4s.2s (
,:L
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...(KA'S)
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TEST'CRDSS REFERENCE PAGE
-
,
-QUESTION-VALUE REFERENCE.
.,_____a__.. ______.
__________
05.01
'2.00 BANOOOO687
'05.02'
3.00 BANOOOO688 05.03 2.50 BANOOOO689 05.04 2.25 BANOOOO690 05.05 1.50 BANOOOO691 05.06 2.50 BANOOOO692 05.07 1.25 BANOOOO693
______
15.00 06.01 2.00 BANOOOO701 06.02 2.50 BANOOOO696 06.03.
2.00 BANOOOO704 06.04 2.25 BANOOOO703 06.05 H2.25 BANOOOO745 06.06 1.00 BANOOOO700 06.07 3.00 BANOOOO695
______
15.00 0 7-. 0 1 3.00 BANOOOO699-07.02 1.50 BANOOOO698-07.03 2.00 BANOOOO694 07.04 1.00 BANOOOO744 07.05 1.00 BANOOOO697 07.06:
2.50 BANOOOO702 07.07 1.00 BANOOOO706 07.08 3.00 BANOOOO707
______
15.00 08.01 1.50 BANOOOO733 08.02 1.00 BANOOOO734 l
08.03 2.00 BANOOOO735
!
08.04 2.00 BANOOOO736 08.05 3.00 BANOOOO737 08.06 2.00 BANOOOO738 08.07 1.50 BANOOOO739
,,
08.08 2.00 BANOOOO740
______
15.00
______
______
60.00
/
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,
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220 l
l l
i
.
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~
210
%
-
%
-
C 200
%-
-
_
190
-
-
180 m FINAL MAXIMUM ENTHALPY
-
170
-
g
-
\\
160
-
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N EXCUR$10N
\\
INCREMENT 150
-
-
g 140
\\
-
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U 130
-
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CURVE
' h 120
-
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No.
INITIAL POWER
.
-
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l 3*, OF RATED 5 110
-
g
30*,0F RATED
-
100
\\
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INITI AL MAXIMUM
\\
CENTERLINE ENTHALPY
\\
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-
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-
-
-
_
-
0 0.01 0.02 0.03 0.04
.s k CONTROL RCD JAMES A FITZPATRICK FS AR UPDAT; ROD DROP ACCIDENT (POWER RANGE)
PEAK FUEL ENTHALPY -INITIAL CORE j
AEV.0 JULY.1982 l FIGUAE NO.14.6-5
.
, - - -
_._
...
. _
_
FIGURE 3
"
.
Wintuou OPL R Al tNG 5P([0
' *
ZERO Otf f u5ta r(Ow e AN ALYT8 Call
.
N* f Ut L SPEED RassCl Pf k
~
ount%G ONL Pur Of'l Ra T ION
.
.
-
-
-
.
100
, --.
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STF ADY STATE
'_
M')vt N T A R Y Of SI A f l0M
_
j DLibite., eNMP FRCHISITED 1/>
START
,
'
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-
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-
10
30
50 dio
80
400 SPtED OF PtidP B(5)
TIGURE 0?!-10 PICIRCULATIGi PUMP SPEED HISMTCH OPERATING LIMITS I
J
FIGURE 4
.
..
..
NEW YORK POWER AUTHORITY JAMES A.
FITZPATRICK NUCLEAR POWER PLANT
.
'
EMERGENCY OPERATING PROCEDURE TITLE:
RPV CONTROL (BORON INJECTION NOT REQUIRED)*
NO. F-EOP-2 TABLE F-EOP-2.1 ( ~~ RPV i
RPV PRESSURE REGION I
l WATER i
i I
LEVEL i
100 psiq 335 psig i
!
TREND l
LOW l INTERMEDIATE l
HIGH I
j am ness e s. eses + mm e s s ema s s es s a nu m mes s e s s a s se ss amm as s e s s s e n es e s s s s s e m ss m e s sa s s s a + s s s s s s s s s ss s sa m m e m o mm se m an j
{
l o IF...HPCI or RCIC is available, I
l
i WHEN-RPV water level reaches 177 in.,
I l
l l
THEN-return to Step 3 (Page 5) of this procedure.
l l
I i
Return to l
l l
o IF...neither HPCI nor RCIC is available,
Step 3 (Page 5)
l l'
AND..RPV pressure is increasing, I
of this procedure.
l INCREASING l THEN, EMERGENCY RPV DEPRESSURIZATION IS REQUIRCD l
l l
(accomplished in Step 4) WHEN RPV pressure l
l I
l is decreasing, THEN return to Step 3 (Page 5).
l l
l l
l l
l I
o IF...neither HPCI nor RCIC is available,
l l
l AND..RPV pressure is not increasing, I
l l
l THEN. return to Step 3 (Page 5) of this procedure.
1 l............+...........................................+.............................................l l
l Start pumps in the alternate injection sub-l
I l systems lined up in Step 3.3.1 (Page 11).
I l'
I l
l I
'
I o IP...RPV pressure is increasing, 1 o IF...either.HPCI or RCIC is not l
l THEN-EMERGENCY RPV DEPRESSURIZATION IS l
operating, I
I I
REQUIPED (accomplished in Step 4).
I THEN. RESTART the non-operating system.
I i
l l
i l o IF...RPV water level drops to 0.0 in.,
I o IF...no injection subsystem is lined up i
l i
THEN. perform Steps 1 through 3 below I
for injection into the RPV with l
!
I (Core Cooling Witheat Level Restora-l at least one pump running, I
l l
tion):
THEN. START pumps in the alternate l
I i
I injection subsystems which were
.1 l
1.
OPEN all ADE l CAUTION 14 !
lined up in Step 3.3.1 (Page 11).
I l
l valves.
I RAPID l
l l
l IF... any valve I COOLDOWN l o IF...RPV water level drops to 0.0 in.,
I l
I cannot be IMAY BE REQD.I AND..any system, injection subsystem, l
i DECREASING l opened,
l or alternate injection subsystem is l
1 THEN. OPEN other SRVs until 7 1 lined up for injection with at least l l
l valves are open.
I one pump running, I
l l
l THEN. EMERGENCY RPV DEPRESSURIZATION IS l
l
2.
Inject water into the RPV from I required (accomplished in Step 4).
l I
l the suppression pool with l
l l
available CS subsystems.
I o IP...RPV water level drops to 0.0 in.,
I l
t
AND..no system, injection subsystem or l
I i
3.
WHEN.at least one CS subsystem i alternate injection subsystem is
!
I I
is injecting with suction I lined up for injection with at least i I
l from the suppression pool, I one pump running,
!
,1
AND..RPV pres nt d s A 30 psig_.I THENJ TEAM COO (INO.IS REQQI_REDj accom-l
1 above suppression pool i
plished in Step 4).
l l
l pressure, I
I l
l THEN. terminate injection into I
l l
I the RPV from sources ex-
l i
I ternal to the primary I
i-
)
i containment.
I
.
Rev. No.
Date 12/84 Page
_ of
g;
.
U.
S.
NUCLEAR REGULATORY CDMMISSION REACTOR OPERATOR REQUALIFICATION EXAMINATION
FACILITY:
_ FIT 2 PATRICK,____________
REACTOR TYPE:
_@WR-GE4_________________
DATE ADMINISTERED: _@Zf9gggZ;_______________
- HAJEK _@z_______________
EXAMINER:
t CANDIDATE:
_________________________
IN@l69CllgNS_Ig_C@N91981El Rzed the attached instruction page carefully.
This examination replaces tha current cycle facility administered requalification examination.
Rstraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training staff.
Points for each question are indicated in perentheses 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 four (4) hours after the examination starts.
% OF
,
'
CATEGORY
% OF CANDIDATE'S CATEGORY
__YelyE_ _Igl@L
___@Cg6E___
_y@LUE__ ______________C@IEGQBy_____________
.19c99__ _25:99 1.
PRINCIPLES OF NUCLEAR POWER
___________
________
PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_19199__ _29 99
________ 2.
PLANT DESIGN INCLUDING SAFETY
___________
AND EMERGENCY SYSTEMS
_19:99__ _2Dz99
________ 3.
INSTRUMENTS AND CONTROLS
___________
_19A99__ _29199
________ 4.
PROCEDURES - NORMAL, ABNORMAL,
___________
EMERGENCY AND RADIOLOGICAL CONTROL
_b9:99__
________%
Totals
___________
Final Grade I
I
'
All work done on this examination is my own.
I have neither given nor received aid.
___________________________________
Candidate's Signature
_-
!.
l
.
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
'
3During the administration.of ' this examination the f ollowing rules apply:
'
1. -
Cheating on the examination means an automatic denial of your application end could result in more revere penalties.
2.
Restroom trips are to be limited and only one candidate at a time may j
leave.
You must avoid all contacts with anyone outside the examination
!
room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil gnly to f acilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the ex ami nati on. -
5.
Fill in the date on the cover sheet of the examination (if necessary).
'
6.
Use only the paper provided f or answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
-8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, fdr example, 1.4, 6.3.
th ee lines between each answer.
10. Skip'at least t
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in f acility literature.
13. The point value f or each question is indicated in parentheses af ter the question and can be used as a guide for the depth of answer required.
l 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE i
i l
QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examinet only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been complete.:
...
'
v
.
18. When'you complete'your examination, you shalls
'
o.
Assemble your examination as'follows:
(1)
Exam questions on top.
(2)
ExamLaids - figures, tables, etc.
(3)
Answer pagen including figures which are part of the-answer.
b'.
' Turn in your copy of the examination and all pages used to answer the examination questions.
c.-
Turn in alliserap 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, you'are found in thi9 area while the examination is still-in progress, your license may be denied or revoked.
.!
c
.
n i
-.
Y
1 PAGE
PRJNCIPLEg_QE_ NUCLEAR _PgWER_P(ANT _9PER9TIgN 1 __THEggggyN@glCg1_HE@l_I6@N5EEB_9Np_E(Ulp_E69W
'
'
t
.
.
QUESTION 1.01 (2.50)
a.
What fuel cladding condition must be met to
,
determine that the core is adequately cooled?
(1.0)
l b.
1.
Under what core spray operating conditions j
will adequate core cooling be assurred?
(0.5)
i 2.
Assume that Core Spray is being used to maintain adequate core cooling.
Give two observations you can make to assure that
{
adequate core cooling is taking place.
(1.0)
'
QUESTION 1.02 (1.50)
State HOW each of the f ollowing parameters will change (INCREASE, DECREASE, or REMAIN THE SAME) if Recirculation Pump speed is increased, a.
Actual bundle power.
(0.5)
.
,
(0.5)
b.
Critical power.
c.
Critical power ratio.
(0.5)
QUESTION 1.03 (3.00)
With regard to Xenon-135 in the core, j
a.
Explain WHY the Xenon level initially increases after a power decrease, and initially decreases after a power increase.
(2.0)
6.
Expl ain HOW and WHY control rod notch worths are altered by the presence of Xenon in the core for a startup ten hours after a reactor scram from (1.0)
full power.
(*****
CATEGORY 01 CONTINUED ON NEXT PAGE
- )
_ __
_ _
PAGE'
H
. la_,PBlNCIELES_QE_NQCLE68_EQWE6_EL@NI_QEE6811gNt
..fd5,BdQ9XN651CSc_UE8I_18@NSEE8,6NQ_E6Ulp_E(QW
.
l
.
' QUESTION 1.04
.(2.00)
m Assume for a specific SRV-tailpiece, pressure is 15 psip, and the tailpiece temperature is 320 degrees F.
a..Is.there any.superheat in the vessel if the. vessel l
pressure is 500 psi y Show all calculations and state
{
any assumptions use in arriving at your answer.
(1.0)
b.
What if the vessel pressure were 300 psi ?
Show all calculations and' state any' assumptions used in arriving
(1.0)
at your answer.
QUESTION 1.05 (2.00)
L I, ?3l IIg u
,
Indicate whether the following events will3 cause the-
-i M INCREASE or DECREASE in core reactivity during operation, AND state which reactivity coefficient will cause the change.
.
,
a.
Moderator temperature increases during a reactor (0.5)
,
startup.
!
b.
Fuel temperature increases.
(0.5)
c.
A feedwater heater is lost.
(0.5)
d.
RPV pressure suddenly decreases.
(0.5)
QUESTION 1.06 (1.00)
)
!
For the f ollowing statements concerning subcritical multiplication, choose the word that makes the statement Correct.
a.
As K-effective approaches unity (1.0), the fractional increase in neutrons per generation _________________
(INCREASES / DECREASES).
(0.5)
>
-
l b.
As K-eff approaches unity (1.0), a _________________
l (SHORTER / LONGER) period of time is required to reach the equilibrium neutron level for a given change in K-eff.
(0.5)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE
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,
21. - ~ PRINCIPLES'OF' NUCLEAR _PQWER_ PLANT _QPERAllgN PAGE ll t
,-
IdE60QQyN8dlC@t_dE@l_IB@N@[E8_@NQ,[(Q1Q_[(QW
.
,
'
..
-QUESTION: 1.07 (2.00)
Assume that the reactor is shutdown and the Recirculation pumps'are' tripped.
!
a.
Give two -(2) methods of enhancing natural circulation in the RPV.
(1.5)
.b.
State what additional' condition is required to assure that natural circulation' will continue after it-has been established.
(0.5)
QUESTION 1.08 (1.00)
State whether the reactivity' worth of a single control rod would INCREASE or DECREASE given the f ollowing changes:
a.
if the void content around-the rod increases.
(0.25)
% 4 '. f +
-b.
if_the moderator' temperature decreases..
(0.25)
'
% 1, L k, l. f f c.'is an adjacent control rod is withdrawn.
(0.25)
lin 4
.d. if Xe-135 concentration around the rod decreases.
(0.25)
4%4 :.f 4 u)
)
--
l
'
b/
.
s (***** END OF CATEGORY 01
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!
!
-)
r-L. "4 1;iELBNI_QEg'l@N_ LNG 69plNG_S9EgIy_80D_EDEBGEggy_gy@IgDS.'
PAGE 5-
o
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4
- QUEST 10N 2.01 (2.50)
L.
'a.-Once the' ADS valves-have been automatically opened-
.by - the. rel ay l ogi c, what conditions will cause or
- allow them to close? -
(1.5)'
)
- b. How can'the operator prevent the ADS valves from closing automatically?
(1.0)
QUESTION 2.02 (1.50)
Concerning operation of the RHR System in the. Shutdown Cooling Mode, j
.a.
Uncer what condition (s) can the high RPV pressure isolation signal be reset?
(0,5)
!
b..
Why is Head Spray used in conjunction with Shutdown Cooling?: Give.two reasons.
- ( 1. 0)
- % bl VALO,byd V45 WyA W W60.
,
.
QUESTION 2.03 (3.00)
With regard to operation of the Standby Liquid Control
-
System, state whether each of the - f oll owi ng statements is either'TRUE or FALSE.-
If the statement is FALSE,
~
state WHY it'is FALSE.
a.
Turning the-keyl ock swi tch to either the " Start System A" or the " Start System B" position will fire both squib valves and will start both pumps.
(1.0)-
.b.-
If.a squib continuity indicating light on the control panel should burn out, continuity can be checked on a back panel meter by observing that current is greater than three milliamps.
(1.0)
c.
During normal day-to-day operations, a heater in the storage tank is used to keep the sodium pentaborate in solution by maintaining a (1.0)
temperature of approximately 150 degrees F.
(***** CATEGORY O2 CONTINUED ON NEXT PAGE
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i-2:__E60NI_DEglGN_lNC6UplNg_g@ Eely _9Np_EDERgENCy_SySIEds PAGE
,
.
QUESTION 2.04 (1.00)
With the plant operating at 100% power,an operator inadvertently dscreases the EHC pressure setpoint by 5 psig.
What will be the i
final status of the f ollowing due to this action af ter the plant has stabilized?
Assume load limit set at 100% and max combined flow set at 105%.
(1.0)
c. TCV position.
b.
BPV position.
c.
Reactor power, d.
Reactor pressure.
QUESTION 2.05 (3.00)
For LPCI System operation, a.
What two signals will cause an automatic initiation?
Setpoints are required.
(1.0)
b.
1.
Under what condition (s) will the pump minimum flow valves automatically open?
(0.5)
2.
When will they automatically close?
(0.5)
c.
Assume LPCI has automatically initiated.
If vessel level has been restored to an acceptable level, and is continuing to increase to a higher than desired level, in accordance with F-OP-13,
" Residual Heat Removal System", HOW and WHEN can you manually control level?
(1.0)
QUESTION 2.06 (2.00)
State FOUR (4) conditions that may (depending on the sensed condition) prevent movement of the refueling equipment, OR withdrawal of control rods during refueling operations?
(2.0)
1 I
i l
l (*****
CATEGORY O2 CONTINUED ON NEXT PAGE
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e
- gt_;fLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
.PAGE-
- 7
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l
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,
'
..
. QUESTION 2.07 ( 2. 00)-
Concerninglthe RCIC System,
.a.
What two (2) conditions will cause the' suction to swap f rom the normal source to the alternate source?
(1.0)
b.
What valve actions will occur, and in what sequence will they occur, when suction swaps f rom. the normal-to'the alternate source?
(1.0)
\\
'
.
I (*****
END OF CATEGORY O2
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j
l
'2:__INSI69MENIS_AND_CQNIBOLS PAGE'
-1 i
'.
.
-
o-QUESTION 3.01 (2.50)
For each of. the f ollowing parameter changes and operational conditions, state whether the INDICATED LEVEL'will INCREASE, DECREASE, or REMAIN-THE SAME for
'the'specified level instrument if the ACTUAL LEVEL
' REMAINS THE SAME.
,
l a.
The Drywell temperature increases about 30 degrees i
due to a small steam leak.
How will the NARROW RANGE level instrumentation respond?
(0.5)
b.
The Drywell temperature increases about 200
)
degrees.
How will the NARROW RANGE level instrumentation respond?
(0.5)
c.
-The Drywell temperature increases about 45
- degrees.
How will the WIDE RANGE level instrumentation respond?
(0.5)
d.
A reactor startup is in progress.
The head vent has been cl osed..
Vessel temperature and pressure are increased from atmospheric a,d 220 degrees F to 800 psig and 518 degrees F.
How will the NARROW RANGE-level instrumentation respond?
(0.5)
e.
A reactor startup is in progress.
The head vent has.been closed.
Vessel temperature and pressure are increased from atmospheric and 220 degrees F to 800 psig and 518 degrees F.
How will the WIDE RANGE level instrumentation respond?
(0,5)
QUESTION 3.02 (1.50)
WHAT automatic actions occur ons a.
A Main Steam Line Radiation Monitor High-High radiation (>3 x normal) trip?
(1.0)
6.
An Off-Gas Radiation Monitor High-High radiation (>4 R/hr)
trip?
(0.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
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3:_ INgl@UMENIS_@ND_ CONI 6QLS PAGE
A.
..
J
.
QUESTION 3.03 (3.00)
a.
For each Rod Block Monitor (RBM) channel- (i e.,
RBM'A and RBM B), give its normal and alternate
!
reference APRM channels.
(1.0)
b.
An LPRM input to the selection matrix has failed high.
The LPRM is also an input to APRM D.
1.
Explain how bypassing this LPRM in the APRM cabinet will affect operation of the Rod Block Monitor.
(1.0)
i 2.
Explain how NOT bypassing this LPRM in the APRM j
cabinet will affect operation of the Rcd Block
"
Monitor.
(1.0)
-QUESTION 3.04 (2.50)
a.
Give all signals (with setpoints) that adtomatically l
initiate an MSIV closure (PCIS Group I Isol ati on).
(1.5)
b.
Under what condition (s) is(are) the MSIV closure scram bypassed?
(0.5)
QUESTION 3.05 (2.00)
,
,
Concerning the Rod Sequence Control System,
i a.
At what power levels is the RSCS active, and how is
!
reactor power determined (i e.
source of signal to j
RSCS) for this purpose?
(0.5)
'
b.
TRUE or FALE.? All RSCS Rod Blocks are bypassed in the " Transition Zone".
(0.5)
c.
Prior to reactor startup the Sequence Mode Select Switch is in the " WITHDRAW" position.
.
l WHEN is this switch placed in " NORMAL" and WHAT EFFECT does changing the' switch position have on the mode of RSCS control?
(1.0)
(*****
CATEGORY 03 CONTINUED ON NEXT PAGE *****)
,
4t__lUSI69DENIg_8ND_CgNIB9LS PAGE
t
-
/',
,
,
b QUESTION-3.06 (2.50)
With the reactor operating at 100 percent power under steady state conditions in three-element control, an-instrument technician mistakenly isolates and equalizes the pressure across one of the Main Steam Line flow
' transmitters which inputs to the Feedwater Control System.
Describe the response of the Feedwater Control System until steady state. conditions are again established.
Include in your. answer the reasons for this response.
(2.5)
Assume no operator' intervention.
i QUESTION 3.07 (1.00)
The HPCI system auto initiated on low reactor water level.
The system operated normally for two minutes then the turbine tripped.
State _ whether the f ollowing statements concerning the HPCI-turbine trip are TRUE or FALSE.
a.
If the turbine trip was due to turbine overspeed the turbine WILL NOT auto restart when the trip condition cl, ears even if an auto initiation signal is present.
(0,5)
'
b.
If the turbine trip was due to an auto isolation of the HPCI system the turbine WILL auto re' start when the system isolation is reset if:an auto initiation signal is present.
(0.5)
!
.
(***** END OF CATEGORY 03
- )
4:.__E69CEgUBES_ _Ng6M L _6BNDBt % _E_MEBGENCLA_ND PAGE.11.
+
-
-
6991969GIC96_cgN16Q6
.
QUESTION 4.01 (2.50)
According to F-ADP-31, Loss of Condenser Vacuum, a.
What are THREE (3) symptoms (other than automatic trips) that you could observe in the Main Control Room if Main Condenser vacuum was decreasing?
(1.5)
b.
How would you determine if the cause c4 vacuum loss was due to air in-leakage DR loss of cooling?
(1.0)
QUESTION 4.02 (2.50)
For.each of the 'f ollowing conditions, indicate whether or not EMERGENCY OPERATING PRDCEDURE entry is required.
If entry is required, state which procedure (s) to enter.
If entry i s not required, state "None."
Consider _each sub part as a separate item..
Assume no additional abnormal condi tions are present f or each individual item.
.
,
a.
RPV level is 1BO. inches.
(0.25)
b.
Reactor power is 12 percent, Startup mode.
(0.25)
c.
Reactor power is 93 percent seven minutes af ter a load reject.
(0.25)
d.
Power operations, Group I isolation occurs.
(0.25)
i f
e.
Suppression Pool l evel is -0.5 inches.
(0.25)
f.
Drywell pressure is 2.9 psig.
(0.25)
g.
East - CRD Accumulators Area Rad Monitor f ull sc al e.
(0.25)
I h.
Suppression Pool temperature is 9B degrees F.
(0.25)
1.
Reactor shutdown, RPV pressure is 1D90 psig.
(0.25)
_
j.
Drywell temperature is 160 degrees F.
(0.25)
_
i QUESTION 4.03 (3.00)
According to F-AOP-46, " Loss of
"B" DC Power System", how would the Reactor Feed Pumps respond AND how would reactor j
vessel water level be affected if the "B" DC Power System I
is lost while operating at 100% power:
a.
If the
"A" vessel level column is selected when the loss
<
occurs?
(1.5)
)
6.
If the
"B" vessel l evel column is selected when the loss occurs?
(1.5)
l
(***** CATEGORY 04 CONTINUED DN NEXT PAGE
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'
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_ - _ _ _ _ _ _ -
i4.
PROCEDURES - NORMAL _AgNQRMAL _EMER@ENCY_AND-PAGE 12.
t
' dBD196D91G86_CQN16QL
.
iQUEST2DN 4.04 (1.50)
According to F-AOP-12, Loss of Instrument Air, how do each of'the following components respond (FAIL OPEN, FAIL CLOSED, or FAIL AS IS) to low air header pressure?
a.
Dutboard MSIVs (0.5)
'b.
CRD Flow Control Val ves (0,5)
Building ventilation 3 ampers (0.5)
d c.
,
l holdin QUESTION 4.05 (2.00)
c.
What requirements are placed on reactor pressure during a hot startup per F-OP-65, "Startup and Shutdown Procedure", so that reactor vessel level can be controlled?
(1.0)
l
.
F-OP-20, " Condensate System", cautions that condensate pump
.
.
b.
33-P-BC and condensate booster pump 33-P-9C should be the last pumps turned on and the first pumps turned off in their respective group.
What is the reason for thi,s caution?
(0.5)
-c.
F-DP-20 also cautions that no more than one condensate pump
.can be operated when the downstream system is " deadheaded".
What is the. reason for this caution?
(0.5)
'DUESTIDN 4.06 (2.00)
Fill in the blanks in the f ollowing statements concerning DDSD-1, " Operating Staff Responsibilities and Authorities",
a.
The on-shift CD shall be the cire Brigade Leader.
I f n o _ __ _ _L_QL _ _ __ _ _ _ _ _ _ _ i s o n s h i f t, the
________CJCL________ shall be the Fire Brigade Leader.
(1.0)
b.
The Fire Brigade consists of ___til_______
(#) members of which at least ___jJLl______
(#) have operational exper-ience.
(1.0)
i
!
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
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.o 19z_,E8QgEQWBEg_:_UgBueLi_eg8g8 bel 1_EDEB@gNgy_@ND PAGE
,
bed 196991G66_gQNIBQ6
'-
'
.
. QUESTION'
4.'07 (1.50)
The Reactor Water Cleanup System will automatically isolate on high.non-regenerative heat exchanger outlet temperature.
a.
What are two (2) possible' causes of this high temperature?
(1.0)
b.
Prior to unisolating the RWCU system after an auto isolation the filter /demineralizers must be bypassed and the system flushed to'radwaste.
What is the purpose of these actions?
(0.5)
,
.
l (***** END OF CATEGORY 04
- )
(************* END OF EXAMINATION ***************)
Y[SI[Q-
" ' " '
11i__EBINQlELgg_QE_NQQ(E@B_EQWEB_E(@NI_QE[6@llgNi PAGE: 14
-
- 10EBD99XN8D1QE2 dE@l.168N@EEB_@ND_ELylD_E(gy-
-
l ANSWERS"-- FITZPATRICK.
-87/05/07-HAJEK, B.
+ =,,,
ANSWER
~ 1. 01'
(2.50)
a.
.The core is adequately cooled as.long as the cladding temperature anywhere'in the core
-
'does not exceed 22OO. degrees F.
( 1. 0) '
b.-
1.
. Adequate core cool'ing will be assurred as long as the core spray system is' operating at or above its design condition.
(0.5)
2.
CS system flow, spray sparger differential
)
pressuref(or others).
(two atEO.5'each)-
j REFERENCE MIT-301.4-(Tab G),-ppg. 16
'17.
j EO-1.04.
'295031K101-
...(KA'S)
l
'
i
-
ANSWER 1.02 (1.50)
a.
Increases b.
Increases-
' c '. '-
Decreases f
REFERENCE H-228.9, Thermal Limits, ppg. 27 - 28.
J ED 15, 16,J19.
293OO9K117
- 293OO9K118 293OO9K123'
...(KA'S)
-
i
<
_
_
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _
_
_ - _ _
PAGE
1.e._ _ESI NCI E6 E S_QE _ NQC6E 98_ EQWE 8_869NI_QE E B8Il gNt IdEBD99XN9dlCSz_bE81_I68NSEEB_6NQ_E691p_E69W
'
~ ANSWERS;-- FITZPATRICK'
-87/05/07-HAJEK,.B.
,
' ANSWER 1.03 (3.00)
f a.
. Xenon is, produced primarily f rom the. decay of L.
1-135. CO.53 It is removed primarily by burnup.
[0.53 On a power decrease, it is initially still being produced at the higher power _ level' rate, while the burnup rate has decreased. [0.53 The.
opposite is true for a power increase. CO.53 b.
Xenon will have built up or' peaked in regions of
' the core that previously had high flux (i. e.
the core center),.and will be much lower in regions that previously had low flux (i. e.
the periphery).
[0.53 This will result in a shift in flux during
,
the startup condition to the previously_ low flux regions, and a corresponding increase in the.
reactivity of the control rods in those regions.
CO.53 REFERENCE
.
'
NET 237.5,-pg. 10. -
EO-1.1, 1.2.
292OO6K103 292OO6K104 292OO6K106 292OO6K107 292OO6K108
'292OO6K110
...(KA'S)
yA
-
e s 9s.c jno2.o y e 4 4 12 iL L
,.'.128
ANSWER 1.04 (2.00)
G \\toz 13 ' *!
M4g j
(1.0)
a.
500 psig nov.7 No gy s b.
300 psig stos.9 h -Leson Phn ( 1. O)
N8 is 4 n REFERENCE MIT-301.8 (Tab K), pg. 22.
EO-7.b, 7.c, 7.d, 8.
239002K504 293OO3K123 295031A204
...(KA'S)
i
_ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _... _. _ _ _ _ _ _ _
-.. _ _ _ _ _ - - -
,
1:__ERIUC*ELES_QE_NQQLE@B_EQWEB_E(@NI_QEEB@IlgN PAGE
1 ISE5M991N@MICg3_dE@l_lB@N@EgB_@NQ_E691p_E699
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
ANSWER 1.05 (2.00)
a.
Decreases CO.25] due to moderator temperature coefficient. CO.253 b.
Decreases CO.25] due to Doppler coefficient. CO.253 c.
Increases CO.25] due to moderator temperature coefficient. CO.253 d.
Decreases CO.25] due to void coefficient. CO.253 REFERENCE NET 237.4
)
EO-1.1, 2.2, 3.2, 4.3.
,
292OO4K102 292OO4K105 292OO4K110
...(KA*S)
!
ANSWER 1.06 (1.00)
.
.
a.
DECREASES (0.5)
b.
LONGER (0.5)
REFERENCE Lesson Plan 237.7, Subcritical Multiplication, pgs 6,7 EO - 237.7.1.3 K/A - 292003 K't.03 Concept of Subcritical Mul ti pl ati on (2.9/3.0)
292OO3K101
...(KA'S)
ANSWER 1.07 (2.00)
a.
1.
Couple the in-shroud and downcomer areas - 1.
e.,
increase reactor water level. CO.753 2.
Establish a heat sink (steam flow). [0.753-lAst,emQ.d k'
b.
Both methods require feed. CO.53 REFERENCE MIT-301.10 (Tab M), pg.
7.
EO-1.1.d.
293OO8K137
...(KA'S)
. _ _ - -.
__
_
PAGE ~17 1c_.EBINC1ELES_QE_NQCLE@6_EQWEB_E(@NI_QEE66IlgN,.
-
-
" ' ISEBdQQyN@DlC@,._dE@l_I6@NSEg6_8NQ_E6Q1Q_E6QW ANSWERS -- FITZPATRICK
.-87/05/07-HAJEK, B.
,
ANSWER 1.08-(1.00)
a.
Decrease (0.25)
M Car Y (O.25)
b. Decrease-L ky g*d c.. Increase (0.25)
d.
Increase (0.25)
REFERENCE
>
FITZPATRICK NET 237.4 Rev. O
'-
E.O.
237.4.5.3 i
201003
.K5.06 2.7/2.9 292005 K1.09 2.5/2.6 201003K506 292OO5K109
...(KA'S)
.
l
-
_ _ _ _ _ --- _ - - _ - - - -
- -
-
PAGE
. 2 __g69NI_DEgl@N_1NCLyDlN@_g8EEIy_@ND_EDE6@ENCy_SYSIEdg ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
t ANSWER 2.01 (2.50)
a.
The valves will remain open until
'
- the initiating conditions clear and the logic circuits are manually reset. [ahWkS[0,3 3)
!
'
- no RHR or Core Spray pump is running-GepD4[0,7h UNLESS the l ogi c kv$s r& 4s v&W pElois & [OJ32 #
(wn code &~d
- ^
'
At W4 < % 9ss erride switch is in override. [O.53 rwk th4rb ro
. < *' l b b.
The operator can take the ADS switches (on panel 09-4)
"' 4 to the OPEN position. [1.03 REFERENCE SDLP-02J (Vol 1 Tab K). F-OP-68, pg.
3, 7.
EO-LOR-1.29, 1.30, 1.31.
218000K403 218000K501
...(KA'S)
'
ANSWER 2.02 (1.50)
a.
With RPV pressure below 75 psig (both relays energized). (0.5)
l
b.
1.
To assist in uniform cooling 2.
To prevent pressure buildup during flooding OR To allow the vessel level to rise (to limit thermal stress during cooldown)
3.
To maintain saturated conditions
,
(0.5 each for any two)
)
i REFERENCE SDLP-10 (Vol 2 Tab B). F-OP-13, ppg.
5, 12, 13.
EO-LOR-1.01, 1.07A, 1.15.
205000K302 205000K303 205000K402
...(KA'S)
I l
l I
--___----
- _
gz__ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
.
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
ANSWER 2.03 (3.00)
a.
False. [0.53 Only one pump starts. [0.53 b.
True [1.03 c.
False. [0.53 150 degrees is used only f or initial mixing.
Normal temperature is maintained between 76 and 85 degrees. [0.53 REFERENCE SDLP-11 (VOL 2 TAB C).
F-OP-17, PPG.
4, 5.
EO-LOR-1.05, 1.14.
211000K403 211000K404 211000K407 211000K408
...(KA'S)
i ANSWER 2.04 (1.00)
TCW at-- ! OO%
smi L i ca-- M Ade io @d F'i"^"'
f
"'
(.25)
a.
b.
BPV shut (.25)
c. power sl i ght y l ower-bd we,S t.t e M.t. 4 o u t. W.
(.25)
'
d.
pressure slightly lower (.25)
l J
REFERENCE FITZPATRICK SDLP-94C Rev. 1 E.
O.
1.03 and 1.05 241000 A1.01 3.9/3.8 A1.02 4.1/3.9 A1.07 3.8/3.7 A1.08 3.3/3.2 2410004101 241000A102 241000A107 241000A108
...(KA'S)
PAGE
?g__E68NI_DE@l@N,1NC(yplN@_@@ Eely _$ND_EDEB@ENCy_Sy@IEd@
.
ANSWERS --:FITZPATRICK-87/05/07-HAJEK, B.
.
ANSWER 2.05 (3.00)
4.
1.
I ow-l ow-l ow l evel -459.5"
-
[0,53 IT" IS * I*'h 5 yac. rh * -
2.
High drywell pressure - 2.7 psig
[0.53 541t%Itd MS"dd *
@ ct tg Ming b.
1.
If flow is less than 400 gpm f or greater than 10 seconds, the min flow valve will open.
CO.53 2.
It will close automatically when flow exceeds
'
400 gpm.
CO.53 c.
After LPCI has been running for at least five minutes E0.53, the LPCI throttle valve [MOV-27.A and B3 can be positioned by the operator to control flow. [0.53 REFERENCE SDLP-10 (Vol 2 Tab B). F-OP-13, ppg.
3, 9,
10.
EO-LOR-1.01, 1.04c, 1.04e, 1.16.
203OOOA409 203OOOK301 203OOOK401 203OOOK403
...(KA'S)
'
.
ANSWER 2.06 (2.00)
All rods not inserted
-
Refueling platform positioned near or over the core
-
Ref ueling platf orm hoists f uel-loaded [ grapple,
-
f rame mounted or trolley mounted hoists]
Fuel grapple full up
-
Service platform hoist fuel-loaded
-
(0.5 each for any four)
REFERENCE
_SDLP-OBB (VOL 2 TAB YZ), PG.
5.
EO-LOR-3.
234000K502
...(KA'S)
- _ _ _ _______ _ _ __ - _ _.
__.
2A__ELONI_DEgl@N_ LNG (UDIN@_S@EEIL@ND_gDE6GENCy_SYSIEDE PAGE '21
.
ANSWERS --.FITZPATRICK-87/05/-07-HAJEK, B.
..
,
' ANSWER-2.07 (2.00)
4.
Suction will swap on low CST level [15,600 gals 3 DR
manual isolation valves not f ull open (0.5 each)
j b.
Torus suction valves [MOV-39 and 41J will open first CO.53, and then the CST suction valve'[MOV-
'
183 will close. [0.53 REFERENCE
.
l 12.
SDLP-13 (Vol 2,
Tab F), ppg. 11
-
EO-LOR-2.04b.
217000K101 217000K103 217000K407 217000K604
...(KA'S)
[
-
.
<
I l
I l
>
. - _ _ _ _
_ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _._j
13 __J.NSI6UMEN]@_6Np_CQNI69hE'
PAGE
2,
.
ANSWERS:-- FITZPATRICK.
-87/05/07-HAJEK, B.
.,
ANSWER-3,01 (2.50)
a.
Remains the same.
b.
Increase c.
Increase d.
Remains the same e.
-Decrease REFERENCE SDLP-402B (Vol I Tab D), ppg.12 - 16.
LOR-1.05, NLO-1.03.
-
216000K501 216000K507, 216000K510
...(KA'S)
ANSWER 3.02 (1.50)
l a.
- Reactor SCRAM (0.33)
- Containment Isol ati on - Grw "E bidio (0.33)
(.- MSIV' Closure MS Line Drain isolation valves close
-
Recirc System Sample isolation valvds close):2.McP Md40
-
- Mechanical vacuum pump trips (if running)
(0.33)
b.
After the 15 minute timer, then the Off Gas isolation val ve (01-107-ADV-100) closes.
(0.5)
REFERENCE
'
-JAFNPP Instructor Lesson Plan SDLP-17, " Process and Area Radiation Monitoring Systems", Rev 1, pgs 12, 17.
Learning Objective - SDLP-17 #1.04 K/A - 272000 K4.02 (3.7)
272OOOK402
...(KA'S)
.
!
-
- _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ -
-
-_
i aA__.1.NEIBUMENI@_86D_QQUI60L@
PAGE 23-
.
l
ANSWERS--- FITZ#ATRICK-B7/05/07-HAJEK, B.
,
ANSWER 3.03-(3.00)
i a.
Normal Alternate RBM
"A" C
E (0.5)
"B" D
F (0.5)
b.
1.
If the LPRM is bypassed, its output would j
drop to zero, and would trip the downscale trip unit. [0.53 This would remove it from the count circuit, and result in a RBM input based on the readings of the properly operating LPRMs. [0.53 l
2.
If the LPRM is not bypassed, the high reading will l
be averaged with the other LPRM signals. [1.03
- C F***************************
REFERENCE SDLP-7C (Vol 2, Tab V), ppg. 14, 25.
^;
.
EO-12, 14.
215002A203 215002K101 215002K102 215002K401
...(KA'S)
i
!
ANSWER 3.04 (2.50)
a.
RPV l ow-l ow-l ow l evel 59.5" (o,tS) (1. 5)
MSL High Radiation 3x NFPB (o,ty)
MSL High Flow 140%
(3,te)
Low Turbine Inlet Pressure B25 psig in RUN
[0,25)
MSL Tunnel High Temperature Tamb + 40^F 03 10 Low Condenser Vacuum B" Hg (0.1T)
(Bypassed if <1005 psig, TSV closed, and not in RUN)
b.
At < 1005 # RPV pressure when not in RUN. [0.53 krm u ann e h M m s<s)
REFERENCE SDLP-05 (Vol 2,
Tab R), Figures 3 and 5, Tech Specs, ppg. 41a - 42.
EO-NLO-1.03, 1.05.
212OOOA211 212OOOK114 212OOOK402 212OOOK412
...(KA'S)
_ _ _ _ _ _ _ _ _ _ _ - _ - - _
_ _ __ - __ _ _ _ _ _ _ _ -
_ __
.
PAGE
Uz,_JNSI6UMENIS_$ND_CQNIBQLS
-
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
. ANSWER 3.05 (2.00)
a.
At less than 20% reactor power [0.253 as determined by first stage turbine pressure. [0.253 b.
TRUE
[0.53 c.
The SMSS is placed to NORMAL when 50 percent rod density is reached. CO.53 This changes the i
control mode from group control to group notch control.-[0.53 REFERENCE SDLP-03E (Vol 2,
Tab 0).
EO-1.05, 1.07.
1201004A402 201004K102 201004K404 201004K406 201004K503 201004K604
...(KA'S)
ANSWER 3.06 (2.50)
(i.e.,decrcin h C
,
.
l Total steam flow would indicate 75 percent 4while actual steam flow would remain at 100 percent. [O.53 The FWLCS would reduce FW according to the change in indicated steam flow, [0.53 Level will begin'to decrease, and a level error signal will be generated. [0.53 Feed flow will increase to bring l evel back to near normal.-LO.53 Final level will be established slightly lower than the original level. CO.53 Clevel will be such that the level error is equal to but opposite the flow error.3 REFERENCE SDLP-06 (VOL 2 TAB S).
259002K102 259002K103 259002K104 259002K301 259002K603
...(KA'S)
l l
l s
.-_
..
_ _ _ _ _ _ _ _ _ - _ _ _ _.
_ - _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _
PAGE
g___1pSIBUdENIS_@ND_QQNI60LS
.
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
ANSWER 3.07 (1.00)
a.
FALSE (0.5)
b.
TRUE (0.5)
\\
REFERENCE
~
JAFNPP Instructor Lesson Plan SDLP-23, "High Pressure Coolant Injection System", Rev 1, pqs 14 & 15 Learning Objective - SDLP-23 23.04 K/A - 206000 K4.03 (4.1)
t 206000K403
...(KA'S)
-
.
_... _ _ _. _ _
I
~4i__EBQGEQQ6ES_;_NQBD86t_@@696D$(t_EDEBGEDQy,@NQ PAGE
e JBee1969G1996_ggNI6QL.
'
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.,.
i ANSWER 4.01 (2.50)
cese( M N @ @ )b.y a.
Decreased generator output [0.53 ud.#di te-pv S*4kN M
Low' condenser vacuum alarm [0.53
'l Increased Off-Gas flow [0.53 b.
Air in-leakage would be indicated by excessive Off-Ges flow (normal flow is less than 100 SCFM). [0.53 Loss of ceoling would be indicated by Hotwell level increase and/ Cire Water outlet temperature increase. [0.53 REFERENCE-F-AOP-31 295002 GOO 5 295002K207 295002K208
...(KA'S)
ANSWER 4.02 (2.50)
a.
None (0.25)
.
'
b.
None (0.25)
c.
EOP-2, RPV CONTROL kk4 %4Icipdt phrg 'do EOF-d (O.25)
d.
e.
None (0.25)
f.
g.
EDP-5, SECONDARY CONTAINMENT CONTROL (0.25)
l h.
EDP-4, PRIMARY CONTAINMENT CONTROL (0.25)
l 1.
EDP-2 (0.25)
J.
EOP-4 (0.25)
REFERENCE EDPs-1,
-2,
-3,
-4,
-5.
295024G011 295025G011 295026G011 295027G011 295028G011 295029G011
...(KA'S)
1
E8QgEQU6ES_ _NQ6066t_QBNQBb661_EDEB@ENgy_@NQ PAGE 27.
4:r_E00196991G06 GQNI6QL-
-
_
ANSWERS -- FITZPATRICK-
-87/05/07-HAJEK, B.
,
ANGWER 4.03 (3.00)
a.
A RFP speeds up in response to the FWLC system. (0.5)
B RFP continues to supply the vessel at a constant rate due to loss of control and tripping power. (0.5) Vessel level increases about 2 inches and stabilizes. (0.5)
b.
A RFP speeds up in response to the FWLC system until vessel level increases to 222.5 inches, then trips due i
to high vessel level. (0.5)
B RFP continues to supply the vessel at a constant rate due to loss of control and tripping power. (0.5)
Vessel level increases to 222.5",
then decreases when A RFP trips. (0.5)
Molt p/Ad Ets noY
,
REFERENCE x4.guire, oMn4 O Nd F-ADP-46.
s gg yf.umM a f8M 295004 GOO 9 295004K102
...(KA'S)
?
by ndJch ="tt g-
'
ANSWER 4.04 (1.50)
a.
FAIL CLOSED b.
FAIL CLOSED c.
FAIL CLOSED REFERENCE F-AOP-12 295019K201 295019K205 295019K207 295019K209
...(KA'S)
i
QRD PAGE
r 86!ggDURgS_ _NgRdAbt_AgNgRDAl _gDgRGgNQy_ANg
t
.
-
1969G1966.ggNI696 ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
.
ANSWER 4.05 (2.00)
o.
The pressure in the reactor vessel will have to be controlled
<below that which can be overcome by condensate booster pump l
pressure, [0.53 since no steam will be available for the
,
feedwater pump turbine (until approximately 700 psig). CO.53
b.
The pumps are powered by 4KV bus (10700) which is powered by the main generator.
No power will be available unless the main j
generator is on line and house service transfer has taken place. (0.5)
'
c, Minimum flow conditions are saf e f or only one pump.
(0.5)
.
REFERENCE JAFNPP F-OP-65, Startup and Shutdown Procedure, Rev 36, pg 16 F-OP-20, " Condensate System", Rev 12, pg 5 K/A's - 256000 K1.13 (3.5), 256000 K2.01 (2.8), 256000 K4.03 (2.8)
256000K113 256000K201 256000K403
...(KA'S)
'
.
.
ANSWER 4.06 (2.00)
a.
1.
Assistant Shift Supervisor (0.5)
2.
Senior Nuclear Operator (0.5)
b.
1.
five (5)
(0.5)
2.
two (2)
(0.5)
REFERENCE JAFNPP Standing Order, DDSD-1, " Operating Staff Responsibilities j
and Authorities", pg 14.
J
, 294001K116
...(KA'S)
{
l
<
ANSWER 4.07 (1.50)
.
c.
- Low RBCCW flow (0.5)
- excessive blowdown flow (0.5)
6.
To prevent the filter cake on the F/D's from being flushed to the RPV if the cake was lost on the isolation.
(0.5)
l
.
$
hi. ' -r.
j
..
t ABNORMAL-~~~~~~~t~ EMERGENCY AND PAGE
4.bEROCEDURES-NORMAL 6#15CQilC6CCQ9iBDL
,
~---~~~~~~~~~
"-
ANSWERS -- FITZPATRICK-87/05/07-HAJEK, B.
,.
.
~ REFERENCE
.JAFNPP Operating Procedure F-OP-28, " Reactor Water Cleanup System", pg 20.
204000K404
...(KA'S)
'
.
.
.-
U
_ _ _ _ - _ - - _ _ _ _ _ _ _ - _ _
-l TEST CROSS-REFERENCE PAGE
.,,,
,
,;,
DUESTION VALUE REFERENCE
______-_
______
__________
.
01.01'
2.50 BANOOOO746 01. 02 ~
1.50 BANOOOO709 01.03 3.00 BANOOOO710 t
'01.04 2.00 BANOOOO711 01.05 2.00 BANOOOO712 l
01.06 1.00 BANOOOO713 l
i 01.07 2.00 BANOOOO714 01.08 1.00 BANOOOO708
______
15.00 l
02.01 2.50 BANOOOO715
'
02.02 1.50 BANOOOO716 02.03 3.00
'BANOOOO717 02.04 1.00 BANOOOO718 02.05 3.00 BANOOOO719 02.06 2.00 BANOOOO720
'
02.07 2.00 BANOOOO721
______
15.00 03.01 2.50 BANOOOO722 03.02 1.50 BANOOOO723 03.03 3.00 BANOOOO724
.
03.04 2.50 BANOOOO725 03.05 2.00 BANOOOO726 03.06 2.50 BANOOOO727 03.07 1.00 BANOOOO700
______
15.00
- I a
.
l 04.01 2.50 BANOOOO729 j
'
~O4.02 2.50 BANOOOO730 04.03 3.00 BANOOOO731 04.04 1.50 BANOOOO732 l
04.05 2.00 BANOOOO694 l
04.06 2.00 BANOOOO741
!
'O4.07 1.50 BANOOOO742
!
______
15.00
______
,
______
60.00
_ - - - - - - - - - -
.
e Attachment 3 f
JAFNPP SYSTEM LESSON PLANS REVIEWED BY NRC Lesson Plan Revision Discrepancies / Comments Off Gas Rev 0 none Standby Gas Rev 1 some sections in outline Treatment form Rx Vessel Rev 1 outline form; just a list of Internals vessel components Rx Vessel Level Rev 2 sketchy information on power Instrumentation supplies; setpoints not clearly listed Rx Vessel Temperature Rev 0 some sections in outline Instrumentation form Rx Vessel Pressure Rev 2 outline form; contains tables Instrumentation with contradicting setpoints Rx Vessel Flow Rev 1 outline form Instrumentation RPV Head Flange Rev 1 outline form
Leakage Nuclear Fuel Rev 2 some sections in outline form
'
Rx Recirculation Rev 1 operational information sketchy; no explanation of trips /runbacks Recirc Flow Control Rev 1 some sections in outline form ADS Rev 2 outline form with brief l
descriptions only; no initi-
{
ation signals
)
CRD Mechanism Rev 1 some sections in outline form Control Rod Blade Rev 1 outline form CRH Hydraulics Rev 2 some sections in outline form i
Rod Worth Minimizer Rev 1 some sections in outline form; F-0P-64 contains more information than the lesson plan.
l
,
o D
'
JAPNPP SYSTEM LESSON PLANS REVIEWED BY NRC-Continued l
Lesson Plan Revision Discrepancies / Comments Rod Sequence Control Rev 1 handwritten lesson plan
System
{
Rx Manual Control Rev i some sections in outline form System j
Rod Position Rev 0 outline form Information System RPS Rev 2 some sections in outline form; includes design bases only.
I FW Level Control Rev 1 none System NI Detectors Rev 1 some sections in outline form SRM, IRM Rev 1 outline form; the lesson plan references tables that were not
[
included; has table with incon-sistent setpoints.
LPRM, APRM, RBM Rev 1 some sections in outline form; setpoint in table conflicts with Recire Flow Control lesson plan.
TIP Rev 1 some sections (component descrip-tions) in outline form Refueling Equipment Rev 0 none Refueling Interlocks Rev 0 some sections in outline form Process Computer Rev 0 outline form Residual Heat Removal Rev 1 some sections in outline form Standby Liquid Control Rev i some sections in outline form Reactor Water Cleanup Rev 1 outline form RCIC Rev 1 some sections in outline form
Core Spray Rev 2 some sections in outline form RBCLC Rev 1 some sections in outline form Primary and Secondary Rev 1 some sections in outline form Containment i
O e
JAFNPP SYSTEM LESSON PLANS REVIEWED BY NRC-Continued Lesson Plan Revision Discrepancies / Comments PC Auxiliaries Rev 1 some sections in outline form PCIS Rev 1 some sections in outline form Process and Area Rev 1 some sections in outline form Radiation Monitors Fuel Pool Cooling Rev 1 none l
Radwaste Overview Rev 0 outline form HPCI Rev 1 outline form Main Steam Rev 2 some sections in outline form; designates 8 ADS valves - ADS lesson plan designates only 7.
Extraction Steam and Rev 1 operational summary incomplete FW Heaters Feed and Condensate Rev 1 some sections in outline form; operational information sketchy Cire Water Vacuum Rev 1 some sections in outline form Priming I
TBCLC Rev 2 none Gland Seal / Gland Rev 1 some sections in outline form; Exhaust handwritten Instrument, Breathing, Rev 1 some sections in outline form; j
Service Air system interrelationships scarce
{
Service Water Rev 2 some sections in outline form; handwritten Emergency Service Water Rev 1 some sections in outline form Rx Bldg Ventilation Rev 1 outline form MG Room Ventilation Rev 0 outline form Turbine Building Rev 0 outline form Ventilation i
t i
i i