IR 05000219/1986022
| ML20210R858 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/26/1986 |
| From: | Keller R, Kister H, Kolonauski L, Kolonavski L, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210R821 | List: |
| References | |
| 50-219-86-22, 50-219-86-22OL, NUDOCS 8610070471 | |
| Download: ML20210R858 (39) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-219/86-22(0L)
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FACILITY DOCKET NO. 50-219 FACILITY LICENSE NO. DPR-16 LICENSEE: GPU Nuclear Corporation P. O. Box 388 Forked River, New Jersey 08731 FACILITY: Oyster Creek Nuclear Generating Station EXAMINATION DATES: August 11-15, 1986 CHIEF EXAMINER: h bhawak L. folonauski, Reactor Engineer (Examiner) Date 9/l F / f(,
REVIEWED BY: o .
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D. Lange, LeVd B R ExaAiner Da(e REVIEWED BY: 7/ Z3/ TI6 R. Keller, Chief, Project Section 1C Date APPROVED BY: dL d Hdrry B. Ki' ster, Chief, Projects Branch 1 Date '
SUMMARY: Operator licensing examinations were administered to five Senior Reactor Operator candidates and two Instructor Certification candidates during the week of August 11, 1986. All candidates passed the written examinatio One candidate failed the oral examinatio A Requalification Training Program Inspection was also conducte No viola-tions or deficiencies were identifie atA188uBt8'olh'
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REPORT DETAILS TYPE OF EXAMS: Initial Replacement X Requalification EXAM RESULTS:
I SR0 I Inst. Cert I I Pass / Fail i Pass / Fail l I I I I I I I IWritten Exam I 5/0 1 2/0 1 I I I I I I I I 10ral Exam I 4/1 1 2/0 I I I I I I I I I 10verall l 4/1 1 2/0 1 I I I I I I I I
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1. CHIEF EXAMINER AT SITE: Lynn Kolonauski, NRC 2. OTHER EXAMINERS: Robert Turner, NRC Gordon Robinson, NRC Consultant Summary of generic strengths or deficiencies noted on oral exams:
No generic weaknesses were noted during the oral exam *
Most candidates displayed a responsible attitude toward the SRO positio *
The candidates were well trained in the use of the Site Emergency Plan Implementing procedure . Personnel Present at Exit Interview:
NRC Personnel Lynn Kolonauski, Reactor Engineer (Examiner)
David Lange, Region I Lead BWR Examiner Robert Turner, Reactor Engineer (Examiner)
Jaccb Wechselberger, Resident Inspector
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Facility Personnel John J. Barton, Deputy Director R. D. Fenten, Training Manager Rod Davidson, Operations Training Manager John Rogers, Oyster Creek Licensing Mike Heller, Dyster Creek Licensing Summary of NRC Comments made at exit interview:
The generic strengths given in Paragraph I were discusse *
The personnel of the Training and Operations Departments were cooperative throughout the examination perio *
The Oyster Creek Training Department failed to send the Systems Diagnostic Procedures (OPS 3024 Series) to the NRC examiners for use in examination preparation. In order to avoid this oversight for future licensing examinations, the examiners requested that a current listing of all Operations Department procedures be provided to the NRC as part of the examination preparation materia g5
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No violations or deficiencies were identified during the Requalifi-cation Training Program inspectio *
The staff expressed a concern about the licensee providing adequate time for personnel training on the modifications made to the plant during the current outag *
The NRC Region I Operator Licensing staff plans to send two examiners to monitor the Emergency Operating Procedure training to be conducted at Oyster Creek in Octobe . Summary of facility comments and commitments made at exit interview:
The facility staff stated that the Modifications Training Program was still under development and acknowledged the NRC comment on the amount of time to be devoted to this trainin *
The facility staff agreed to provide a copy of the Final Modifica-tions Summary to both the NRC Regional BWR Operator Licensing staff and to the Oyster Creek Resident Inspector's offic . Changes r.:ade to written exam during examination review:
All facility comments about the written examination were resolved during the two hour exam review conducted immediately after the completion of the written examination. No further comments were submitted by the facilit The following list represents the significant changes made to the examination.
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Question N Change Reason 5.02 Samarium is an acceptable Tech Spec Bases alternate answe .04, 5.05 Point distribution on key Credit given as long as not strictly followe major concepts state .07 Key changed to accept five SGTS considered a part responses instead of si of secondary containmen .01 c The 86G relay trip causes More specific answe the generator tri .02 a Original answer not Answer changed to reflect accurat corrected lesson pla .04 a The EPR, MPR, and the These are also turbine Bypass Opening Jack are components that regulate acceptable answers, steam flo .05 b Manual control of FWR Acceptable answe valve also acceptabl .06 Additional answers include: Additional correct differences in enrichment, answers accepte discrimination, and U308 coating thicknes .07 Additional answers include Additional correct closing of Recirc sample answers accepte lines, isolation of IC vents, and V-6-39 .02 b It was announced to the Additional information candidates that startup require had progressed past the withdrawal of the twelfth control ro .05 a The tube side pressure Additional correct indication on IF/2F is answer accepte also acceptabl .10 Actions of the FWCS Sufficiently answers Failure procedure should the question, also be accepte I
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Question N Change Reason 5.02 Samarium is an acceptable Tech Spec Bases alternate answe .04, 5.05 Point distribution on key Credit given as long as not strictly followe major concepts state .07 Key changed to accept five SGTS considered a part responses instead of si of secondary containmen .01 c The 86G relay trip causes More specific answe the generator tri .02 a Original answer not Answer changed to reflect accurat corrected lesson pla .04 a The EPR, MPR, and the These are also turbine Bypass Opening Jack are components that regulate acceptable answer steam flo .05 b Manual control of FWR Acceptable answe valve also acceptabl .06 Additional answers include: Additional correct differences in enrichment, answers accepte discrimination, and U308 coating thicknes .07 Additional answers include Additional correct closing of Recirc sample answers accepte lines, isolation of IC vents, and V-f-39 .02 b It was announced to the Additional information candidates that startup require had progressed past the withdrawal of the twelfth control ro .05 a The tube side pressure Additional correct indication on IF/2F is answer accepte also acceptabl .10 Actions of the FWCS Sufficiently answers Failure procedure should the questio also be accepte W
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Question N Change Reason 8.12 The information contained Table not provided as in Table 3.1.1(E) given to part of the Tech Spec candidates during exa handou Attachments:
1. Written Examination and' Answer Key (SRO)
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'~*me, U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: OYSTER
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p REACTOR TYPE: BWR-GE2
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DATE ADMINISTERED: 86/08/11 -------------------------
EXAMINER: _ TURNER 1 _R.______________
CANDIDATE: _________________________
INSTRUCTIONS TO CANDIDATE:
Une separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the questio The passing grade requires at least 70% in each category and a final grade of at Isast 80%. Examination papers will be picked up six (6) hours after the examination start % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE
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TOTAL SCORE
VALUE
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CATEGORY-29199-- 29 99 ___________ ________ THEORY OF NUCLEAR POWER PLANT
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OPERATION, FLUIDS, AND
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THERMODYNAMICS 25.00
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25.00 ----_------ -------- PLANT SYSTEMS DESIGN, CONTROL,
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AND INSTRUMENTATION 25.00
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25.00 ----------- -------- PROCEDURES - NORMAL, ABNORMAL,
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EMERGENCY AND RADIOLOGICAL
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CONTROL
- 29199__ _29199 ___________ ________ ADMINISTRATIVE PROCEDURES,
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CONDITIONS, AND LIMITATIONS i
! 100
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Totals j
t Final Grade
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All work done on this examination is my ow I have neither given nor received aid.
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Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application cnd could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil gnly to facilitate legible reproduction Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gn,e side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no IS. Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the ex,amtner_
{ onl . 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 examinatio This must be done after the examination has been complete pc , * ' *L
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18. When you complete your examination, you shall: Assemble your examination as follows:
(1) Exam questions on to (2)* -E-ram aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination question Turn in all scrap paper and the balance of the paper that you did not use for answering the question Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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PAGE 2 IME509DINQUICS
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QUESTION 5. 01 **- (2.00) . The reactor is operating at 335.3 psig reactor pressur What maximum pressure may the reactor pressure be one hour later that assures the reactor vessel heat up rate limit (according to Technical Specifications)is NOT exceeded? Show all wor (2.00)
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QUESTION 5.02 (3.50)
a) The Standby Liquid Control System is designed to shutdown the reactor in the event that the control rod drive system fails to insert the control rods when activated by a scram signal or manuall Name the contributors to positive reactivity that must be counteracted by addition of the baron poison if the reactor was at full power when the ATWS occurre (five required) (2.50)
b) Why are minimum and maximum SLC injection times specified? (1.00)
DUESTION 5.03 (2.00)
For each of the following events, state the INITIAL reactor power response. BRIEFLY explain. Include in your answer the reactivity coefficient (s) responsible for the chang a) Generator load reject at 100% powe (1.00)
b) Inadvertant Isolation Condenser initiation at 100% powe (1.00)
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QUESTION 5.04 (2.00)
The reactor has been operating at 95% power for several days. An operator RAPIDLY reduces reactor power to 60% by reducing the speed of the recirculation pumps. During the next 2-3 MINUTES the operator notices that the reactor power slowly increases by approximately 3% (with no operator action). EXPLAIN the cause of this effec (2.00)
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. THEORY OF NUCLEAR -
POWER PLANT OPERATION, FLUIDS, AND
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QUESTION 5.05 (2.00)
In the main condenser, circulating water flow rate is approximately 20 times that of the steam flow rate. Why are these flow rates different?
(Consider thermodynamic principles in your answer) (2.00)
QUESTION 5.06 (2.00)
a) What is the basis for the MAPLHGR, LHGR and MCPR limits? (1.00)
b) How does the Critical Power vary (i . e. increase, decrease or not affected) if: Reactor pressure increase
. Inlet subcooling increases Mass flow rate through core increases (1.00)
QUESTION 5.07 (3.00)
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in studying " Accident Safety Analysis" an accident is defined as
"any single event not reasonably expected to happen, but postulated for analysis purposes , that causes or threatens to rupture any of the barriers which prevent radioactive release". List six of these barrier (3.00)
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21-_ldE901_9E_USEhE00_C9 DES Ch0U1 9CE00119UL Eh91SSL AND PAGE 4 IUEBdODYN@dICS
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_ OUESTION 5.08 (3.00) -
Assume that an automatic initiation of the core spray system has occurred and the initial injection pressure is 350 psig.The reactor pressure then decays to 100 psig. For each of the core spray paramete-s listed below, determine any change (i . e. increase, decrease, or remain the same).BRIEFLY EXPLAIN the reason for your answe a) Pump flow (1.00)
b) Pump discharge head (assume constant NPSH) (1.00)
c) Pump power requirements (1.00)
DUESTION 5.09 (2.00)
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, There are two types of control rod worth of interest in BWR technology, differential rod worth and integrated rod wort Show these two concepts graphically and state units for eac (2.00)
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QUESTION 5.10 (1.50)
The reactor is operating at 90 % power and a central control rod at position 14 is withdrawn two notches and the power decreased. What is this phenomenon called and what is happening in the core? (1.50)
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4 OUESTION li .1 1 (1.00)
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a) If the reactor is on a 60 second period, how long will it take for the reactor power to increase from 10 KW to 100 KW? (0.50)
! b) Assuming the period stays constant at 60 seconds, will it take a longer time, a shorter time or the same time for the reactor power to go from 10 MW to 100 MW than from 10 KW to 100KW? (0.50)
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
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OUESTION 5.12 (1.00)
When operating in the sub-critical region and Keff is made to rpproach unity by the same incremental changes in Keff:
a) Will the time to reach each new equilibrium neutron level be less, more or the same as Keff gets closer to one? (0.50)
b) Will there be less, more, or the same increase in neutrons ~
at each equilibrium level as Keff gets closer to one? (0.50)
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DUESTION 6.01 (2.50)
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If a turbine trip occurs during full power operp.t i on :
a) What turbine control equipment would be activated and in what (1.00)
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b) Assuming the resulting pressure transient causes the reactor pressure to increase beyond the turbine control equipment capabilities what equipment would be called on to turn the w pressure transient?. (1.00)
c) What protective device causes the generator trip? (O.50)
DUESTION 6.02 (2.50)
a) What is the basis for automatic bypass of the RWM when reactor power is greater than the the Low Power Setpoint of 10% 7 (1.00)
b) What process signal is sent to the RWM to indicate reactor power level? (0.50)
c) Explain what the " transition region" of the RWM i s, and how RWM operation in this region differs from RWM operation when reactor power is below 5%. (1.00)
DUESTION 6.03 (3.00)
The CRD pump is normally operated at a 1500 psig discharge pressure supplying 65 gpm to the CRD hydraulic system. Name the three pressure stages and identify the purpose of eac (3.00)
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The purpose of the turbine control system is to regulate the flow of steam to the turbine so that the reactor pressure is maintained at a preset value, monitor critical values so steam flow can be interrupted as necessary to protect the turbine and condenser, and send a signal to the reactor protection syst?m when the flow of steam is interrupted in anticipation of the resultant reactor pressure spike, a) Name three turbine system components that regulate steam flo (1.00)
b) Name three critical values (system parameters) monitored for turbine protectio (1.00)
w, c) What turbine component sends a signal to the RPS when steam flow is interrupte (0.50)
QUESTION 6.05 (2.00)
The reactor is operating at 100 */. power when the vessel level input to the feedwater control circuit fails to zero. Using the attached feedwater control diagram:
a) Describe what effect the vessel level signal failure will have on the feedwater flo (1.00)
b) What operator action would be required to minimize a significan.t reactor level transient? (1.00)
j QUESTION 6.06 (1.50)
What are Three major differences between the SRM and the IRM i fission chambers? ,
(1.50)
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QUESTION 6.07 (2.50)
r For the Main Steam Line Radiation Monitoring Subsystem:
a) Where are the detectors specifically located? (O.50)
b) ON a trip of this system, what four other automatic actions (as a direct result of the MSL Hi Hi Rad)
occur besides the closure of the MSIV's ? (2.00)
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QUESTION 6.08 (1.50)
Identify and explain the purpose of each of the three filter sections in a Standby Gas Treatment System trai (1.50)
DUESTION 6.09 (3.00)
Assume the plant is operating and a LOCA occurs which causes the diesel generators to be activated and idling. If off-site power is then lost , what would happen to running equipment and DG's? (3.00)
DUESTION 6.10 (2.00)
If the plant was being operated at 100% and 4160 V bus 1A power was lost :
a) How and why would reactor power change? (1.00)
b) Why is there a note in the manual corrective action that states the suction and discharge valves in two retir loops must be open? (1.00)
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QUESTION 6.11 (2.00)
The emergency 4160 buses C & D are designed for supplying equipment vital for safe shutdown of the plant. List the pumps supplied by 4160 Volt Switchgear 1C (numbers not required). (2.00)
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QUESTION 7.01 (3.00)
In the " Control Room Evacuation" procedure (Qape are a number of actions that are required before the step." evacuate the control room" is reached. List six actions and BRIEFLY explain the basis ~~-
for eac (3.00)
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QUESTION 7.02 (2.00)
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With respect to the Rod Worth Minimizer; a) Will the Rod Worth Minimizer (RWM) enforce rod blocks if the " EMERGENCY ROD IN" mode of the RMCS is being used? BRIEFLY explai (1.00)
b) What are the minimum monitoring requirements for rod withdrawal when the RWM fails during startup? (0.50)
c) What is the purpose to limit startups without the RWM to one per year? (0.50)
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QUESTION 7.03 (1.00)
The Power Control portion of the RPV Control Emergency Procedure (EMG-Ol) defines 273 pounds of baron as the the cold shutdown boron weight. On what plant conditions is this weight based? (1.00)
QUESTION 7.04 (2.00)
During reactor operation a loss of Reactor. Building Closed Cooling Water has occurred. Per "2OOO-ABN-32OO.19 ,RBCCW Failure Response": ,
a) Why is it necessary to trip the cleanup and the recir. pumps if RBCCW flow can not be reestablished in one minute? (1.00)
b) is a reactor scram required? (0.50)
c) How can the heat input to the RBCCW system be reduced? (0.50)
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DUESTION 7.05 (2.50)
Station Procedure 307, Isolation Condenser System, provides instructions on maintaining the isolation condenser in standby readines a) What four parameters would you use to assure standby readiness? (2.00)
b) What should be the status of the condensate return valves? (0.50)
QUESTION 7.06 (2.00)
Station Procedure 205 is the controlling document for normal reactor refueling operations. What are the four responsibilities of the Group Shift Supervisor according to this document? (2.00)
OUESTION 7.07 (2.00)
On loss of 125 V DC, there is a caution in ABN-32OO.13 that states
" REACTOR LEVEL SHALL BE CONTROLLED BY ABN-32OO.17,FEEDWATER SYSTEM FLOW CONTROL FAILURE".
a) What is the reason for this caution? (1.00)
b) How can the feedwater flow be controlled? (1.00)
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OUESTION 7.08 (3.00)
5 If all five recirculation pumps trip during reactor power operation, what are three actions that should be taken? (3.00)
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-OUESTION 7.09 (2.50)
While performing the Heating and Pressurization portion of the Plant Heatup to Hot Standby Procedure 201.2, the following equipment is put into service: (not necessarily in the order listed) Reset turbine trips and commence warming the steam chest Start SJAE Establish turbine shaft sealing Start one feedwater pump Put cleanup pump in service Match up the following pressure ranges with the equipment to be put into service, above: (2.50)
1. SO psig 2. 60 - 120 psig 3. 130 -150 psig 4. 250-300 psig 5. 400-600 psig OUESTION 7.10 (2.00)
The reactor power was being increased and you noted that the reactor vessel level was decreasing. What would be two immediate responses of the GSS to that situation? (2.00)
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DUESTION 7.11 (3.00)
What are the six entry conditions for Procedure EMG-32OO.01, RPV Control ? (3.00)
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E1__O99191510011YE_E09EES90EE1 E9UE1119dE1_OUS_h}gg}}gNS PAGE 12 QUESTION 8.01 (1.50)
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After an automatic initiation of t'he core spray system, what two conditions must be satisfied before the core spray system (as as ECCS) can be secured or placed in manual? (1.50) OUESTION 8.02 (1.00) Per Technical Specifications, LIST two of the conditions which ~- must be satisfied prior to BYPASSING the refueling interlock to permit removal of a Control Ro (1.00)
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QUESTION 8.03 (1.00) Caution No 1. on page 32 of procedure 302.1, CRDH System states
" Monitor the rod position indicators during all periods of rod motion for indications of abnormal rod motion". Define abnormal rod motio (1.00)
OUESTION 8.04 (3.00) What i s the minimum Station Operating Complement as given in the Procedure " Conduct of Operations" N ? (3.00)
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_____ _ ___AN _________D _8_. ___ADM I N I STR A T I VE PRO _CE_ DURE _S ,_________________ __ L_ __CONDI_TIONS, IMITATIONS _________ PAGE 13
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DUESTION 8.05 (2.00) Procedure 108, Equipment Control, indicates that the tags used for control of switching and tagging are color coded. Match the color with the correct descriptio a) RED d) RED / WHITE b) DLUE e) WHITE c) YELLOW f) ORANGE ~
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_r s e e> % w u.; Descriptions 1) To provide electrical safety clearance 2) To provide electrical safety / mechanical equipment 3) Electrical caution tag 4) Mechanical tag / equipment shall not be operated (2.00) OUESTION 8.06 (2.00) What are the two main requirements for making a temporary change to a procedure which cannot be delayed for normal review and approval? (2.00) QUESTION 8.07 (2.00) The plant is operating at 80 */. pcwer and with only four recirc. loops in service. The limiting MAPLHGR is for type V fuel at 2 foot core height and 20 GWD/MTM. Based on section 3.10 of the TS attached: a) What is the limiting MAPLHGR ?(Value) (1.00) b) What is the basis for the MAPLHOR limit? (1.00) OUESTION 8.08 (2.50) Give the definition for primary containment integrity as specified in your Technical Specifications. (exact wording not required) (2.50)
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91_ 09U19151091125 509E5S_Uggg3_gggg}}}ggg3_ggg_b}g}TATIONS _ _ PAGE 14
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QUESTION 8.09 (3.00) What are the TWO safety limits _specified in the Oy. ster Creek Technical Specifications? Provide a brief description of the basis for eac (3.00)
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DUESTION 8.10 (2.00) -~ During refueling #1 SBGTS fan trips on overload during its 10 hour test run and needs extensive repairs. #2 SBGTS fails to ar;t.durino.the subsecuent-operabilitv test. What two . requirements are placed on the plant for this condition? (2.00) DUESTION 8.11 (2.00) Referring to the attached sections of the Technical Specifications, STATE the required action for the f ollowing si tuation whil e at powe Reference all TS's that you use to develop your answer (2.00) V-14-30 has been out of service for four days in the open position when the valve operability test on the other isolation condenser fails due to an inoperable condensate make-up valv QUESTION 8.12 (3.00) Following a three month refueling / maintenance outage, a startup is scheduled fcr your shift. All required surveillance and operational prerequisites have been complete I&C reports a surveillance logic functional test of the High Drywell pressure trip function for the Containment Spray Logic indicated that one half of the logic was inoperative. (The troubleshooting and repair will take about 16 hours.)
According to Technical Specifications, can you take the Mode Switch to Start Up and start pulling control -rods? Justify your answer by fully referencing ALL applicable . (3.00) i
! < (***** END OF CATEGORY 08 *****) ; (************* END OF EXAMINATION ***************)
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91__19590__g{_ggggggg_{g_{g_gggy]_ggggg}}g_N3_{{g},Dgg_g_ND Y _ _T_HE_R_M_O_D_Y_N_AM_I_C_S _ _ ANSWERS -- OYSTER CREEK -86/08/11-TURNER, he ANSWER 5.01 (2.00) a7"335.3 psig + atmospheric pressure 14.7 = 350 psia Tsat = 431.73 degrees F 431.73 +100 (hourly heat rate limit) = 531.73 degrees F (1.00) b) Psat = (approximately) 900 psia (1.00) m . .- . - . -
- __
REFERENCE Steam Tables and Oyster Creek Technical Speci f i cati ons 3. 3. A. (ii ) ANSWER 5.02 (3.50) a) Rated Void Collapse Fuel Doppler Effect Decrease , Xenon Decay Temperature Decrease to 125 degrees F Shutdown Margin 4% delta K (.5 each) (2.50) b) minimum (60 minutes) to prevent power chugging due to uneven mixing (.5) and maximum (120 minutes) to compensate for cooldown following xenon peak (.5). (1.00) ' REFERENCE Oyster Creek Technical Specifications, 3.2 Reactivity Control - Basis LORM #53 KR's 8 and 12
.
ANSWER 5.03 (2.00) a) Reator power increases (.25) due to the pressure increase causing void collapse and the resultant positive reactivity insertion due to the void coefficient (.75). (1.00) b) Reactor power increases (.25) because cold water in the IC return leg is discharged to the suction of the "A" or "E" recirculation loop which causes a positive reactivity insertion due to the cooling of the moderator (moderator temperature coefficient). (1.00) _ _ _ _ _ . - - _, . . - - . _ - -. - - - _ _ _ _ _ _ , _ _ _ , __ . - _ _ _ _ - _ _ _ _ __
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 16
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______________ ANSWERS -- OYSTER CREEK -86/08/ll-TURNER, REFERENCE Oyster Creek LP2Ol, BWR Operating Characteristics, Pg's 23-24,33 ANSWER 5.04 (2.00) The reactor is now producing less steam to go to the turbine. There will be less extraction steam and reheater drain steam going to the feedwater heater ( 1. 0) . Therefore, less feedwater heating will occur resulting in colder feedwater entering the vessel (.5) which will cause reactor power to increase about 3 /. trom the pousLivereactivily -- cddition (alpha m) (.5) . (2.0) REFERENCE Oyster Creek - GE Thermodynamics - Heat Transfer - L Fluid Flow Ch. 2 ANSWER 5.05 (2.00) Circulating water is maintained subcooled while the steam undergoes a change in phase. ( 1. 0) The heat removal required to condense the steam (i . latent heat of condensation) accounts for-the large difference in flow rates.(1.0) (2.00) REFERENCE Oyster Creek GE Thermodynamics - Heat Transfer & Fluid Flow - Ch. 8 ANSWER 5.06 (2.00) a) The Thermal Limits are established to protect the barriers to fission product releas (1.00)
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b) Critical Power: 1. decreases increases increases (.33 each) (1.00) REFERENCE GE Thermodynamics / Heat Transf er/ & Fluid Flow Ch. 9 Pg 9-19,9-68
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. i PAGE 17 5.-... THEORY OF NUCLEAR
-- - ----- POWER PLANT OPERATION, FLUIDS, AND
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_ . . , fNSWERS -- OYSTER CREEK -86/08/ll-TURNER, ANSWER 5.07 (3.00) a) Fuel pellets (designed to contain approximately 707. of the fission
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b) Fuel clad (integri ty maintained by saf ety limits and limiting conditions for operation) c) Reactor Vessel (designed for 1250 psig and protected by RPS, EMRV, , q,
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d) Primary containment (designed and treated like a pressure vessel and serves as backup to the reactor vessel) o) Secondary containment f) SGTS (to control and release of radioactivity)
(.50 for each correct answer) (3.00)
REFERENCE Oyster Creek - LORM Vo , Lesson Plan 201, Pg. 8 , ANSWER 5.08 (3.00)
- a) Increase (.25), as the pressure of the system decreases,the flow increases due to the pump head / flow characteristic (.75). (1.00)
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b) Decrease (.25), as the pressure of the system decreases,the operating point on the pump characteristic curve is shifted to a lower pump discharge pressure (.75). (1.00) I c) Increase (.25), from the pump characteristic curve, as the flow capacity increases the power requirements also increase (.75). (1.00) REFERENCE Oyster Creek GE Thermo HT &FF Chapter 7 l i l i
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ANSWER 5.09 (2.00) a) 1 Differential Rod Worth T 1 . . R 1
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b) Reactivity / unit distance of travel - Differential Rod Worth (0.50) Reactivity / entire distance of travel (0.50) REFERENCE Oyster Creek - LORM - Lesson Plan #300.08 Pg.53
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ANSWER 5.10 (1.50) a) Reverse power effect (0.50) b) The rod is being withdrawn and the power is going down because the positive reactivity due to CRD withdrawal is less than the negative reactivity due to local generation of void (1.00) REFERENCE Oyster Creek - LORM - Lesson Plan #201 Pg.12/13
ANSWER 5.11 (1.00) . t/T t/SO a) P = Poe 100 = 10 e In 10 = t/60 t = 2.3*60 = 138 sec. (0.50) b) The same (O.50) REFERENCE ' Oyster Creek - LORM - Lesson Plan #300.11 P . - . _ _ . .. .. . , . . .- - _ _ - - - .
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_-_--__------- ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ANSWER 5.12 (1.00) d . %.
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a) greater time (.5) (0.50)
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b) greater increase in neutrons (.5) (0.50) REFERENCE
' Oyster Creek, 842.0.95 Reactor Theory Review P _ - - .
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50)- - c) The generator will trip (due to the closure of the turbine steam admission valves, generator anti-motoring). (0.50) ' REFERENCE Oyster Creek - LORM - Lesson Plan #201, Pg 21/2 ANSWER 6.02 (2.50) a) A rod drop reactivity addition is less severe when reactor power is in a region that results in voids in the core because the presence of voids provides a rapid feedback mechanism that dampens the reactivity addition resulting in slower power changes than would occur if there were no voids , in the core regio (1.00) b) Total steam flow signal (0.50) c) The " transition region" is reached when 10% steam flow is exceeded for 60 seconds. The region ends when 15% power is exceeded for 60 seconds (.5). (i . e. between the low power set point and the low power alarm) In the " transition region, the RWM no longer applies rod blocks, but updates the RWM display every five seconds (.5). (1.00) REFERENCE Oyster Creek LP #49, RWM, KR 5, Pg 2-4,7,8
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, s. -
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9e ANSWER 6.03 (3.00)
"CRD Charging Water Pressure" (.5) supplies high pressure water, --
normal l y 1390-1510 psig to the accumulator charging header (.5).
"CRD Drive Water Pressure" (.5) sup' plies water at a pressure approximate +y 250 psi above the reactor vessel pressure for motive force to move control rod drives (.5). , . .
"CRD Cooling Water Pressure" (.5) supplies water at approximately r rrj rgsprmyg7pgi tg ggggig ggec; g .ty gggy--(;,g g , o p;,1 ty , - 9 ,, ,1 g ; , 3 ; ,-( 3 ,c o ;- -----
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REFERENC Oyster Creek - LORM - Lesson Plan #9, Pg's 12/15 ANSWER 6.04 (2.50) a) 1. turbine control valves i 2. turbine stop valves 3. turbine intercept valves 4. turbine bypass valves (1.00) i b) 1. condenser vacuum (Will accept any of the turbine 2. steam header pressure trips listed on pg 26 of the reference) 3. exhaust hood temperature (1.00) c) Turbine stop valve 10 '/. closure switche (0.50) I ! REFERENCE Oyster Creek - LORM - Lesson Plan # 79, Pg's 23/26 -
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. e Pt. ANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22
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ANSWERS --- OYSTER CREEK -86/08/11-TURNER, ANSWER 6.05 (2.00) a) Loss of signal will cause the level somparator to produce a large positive error signal that will call for full open FW regulating valves (1.00).
b) The operator will switch to the other vessel level signal and/or manually scram the reactor before it is scrammed auton.a ticall y by
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high level.(1.OO) (2.00)
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_ _ . . - _ . ___ , - - - __ _ _ _ , _ _ _ REFERENCE Oyster Creek - LORM - Lesson Plan #44, Feedwater Cc.ntrol Figure
i ANSWER 6.06 (1.50) a) SRM fission chambers operate at a higher argon pressure than IRM fission chamber (0.50) . b) SRM fission chambers high voltage is 350 VDC while the IRM fission chambers are operated at ISO VD (0.50) c) SRM detector output is current pulses while IRM detector output is piled up current pulse (0.50) ] REFERENCE
Oyster Creek - LORM - Lesson Plan #37, Pg 6-11 ANSWER 6.07 (2.50) s a) Detectors mounted in steam pipe tunnel, next to main steam piping, upstream of outer MSIV's (0.50)
,
. b) 1. Reactor Scram Off-gas bypass valve close AOG isolated Mech. vacuum pump tripped (.5 each) (2.00) . REFERENCE Oyster Creek - LORM - Lesson Plan #69, P /13 ! i
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of the' absolute fi1ters. (.5) Absolute Filter - removes fine particles of dust and contaminants from the air stream. (.5)
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Charcoal' Filter - removes radio-nuclide's ~'(i od i nesi from the exhaust air.(.5) " _.,
(1.50) (Second Absolute filter - collects break-through from the CF.)
REFERENCE Oyster Creek - LORM - #50, Pg' /12 ANSWER 6.09 (3.00) If normal off-site power should be lost during a LOCA, the follcwing happen: a) Running equipment will trip; (Core Spray, Containment Spray, ESW). Breakers 1A1P (181P) to USS 1A1 (181) trip, Breaker 1C (1D) trips ope (1.00) b) The DG's will accelerate from the idle condition and close in on the buse (1.00) c) The loads will come back on according to their sequence timer (1.00) REFERENCE Oyster Creek - LORM - Lesson Plan #65, Pg. 26
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ANSWER 6.10 (2.00) a) Drive motor breaker for recirculation pumps A, C and E would trip on undervoltage(.5) and reactor power would decrease due the increased voids due to the reduced core flow.(.5) (1.00) b) To assure that there is always a path for natural circulation and indication of core water leve (1.00) _ . . - - . . - . . ~ _- _ - - . _ . . _ _ ~ ~~ ' ~ ~ ~ REFERENCE
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, _ Oy.s.ter . Creek - LORM - Lesson Plan #48,- Pg . 5, and Alarm Response ,. Procedures, Vol 1. Section E, F-1-a
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ANSWER 6.11 (2.OQ) . Emergency Service Water Pump 1-1 (.5) Emergency Service Water Pump 1-2 (.5) Core Spray Pump NZO1A (.5) Core Spray Pump NZO1D (.5) (2.00) REFERENCE Oyster Creek - LORM - Lesson Plan #39, Attachment #1
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. PROCEDURES - NORMAL, ~~~~~~~~~~~~~~~~~~~~~~~~ ABNORMAL, EMERGENCY AND PAGE 25
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ~~
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ANSWER 7.01 (3.00) a) Scram the reactor (.25) to reduce heat input (.25)
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_ _ b) Trip recirc, condensate and feedwater pumps (.25) in preparation for reactor system isolation (.25) c) Close MSIV's (.25) to isolate reactor system (.25) s,. .
- - - - - - - . . . . . . . _ _ . _ . . _ . . .. ,_ . . ..
, e) Have both CRD pump's running (.25) to have source of makeup water (.25).
f) Take radios (.25) for communication (.25). (3.00) REFERENCE y Oyster Creek - Control Room Evacuation - 2OOO-ABN-32OO.30, Pg. 2/3 ANSWER 7.02 (2.00)
:
a) Yes (.25). The RWM will block Emergency Rod Insertion if reactor power is below the LPSP to prevent rod insertion 4 that is not compatible with the latched rod withdrawal sequence (.75). (1.00)
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b) Second licensed operator verifies that the first operator is following the rod pattern.(.5) (0.50) c) To assure high operability of the RWM in preference to procedural control s. ( .5) (0.50)
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REFERENCE Oyster Creek - Procedure " Approach to Criticality" No. 201.1 Pg. 7
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Procedure "CRD RMCS" No. 302.2 Pg.12 ANSWER 7.03 (1.00) , , Reactor is cooled down and depressurized (.OO),the reactor water j level is at the high trip point (.5) and the reactor is isolated with only the Shutdown Cooling System in operation (.5) (1.00)
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ABNORMAL, $MERGENCY AND PAGE 26 '
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_ _____ __ ANSWERS -- OYSTER CREEK -86/OB/ll-TURNER, ANSWER 7.04 (2.00) a) Seal cooling is lost to Recirc. Pumps & Clean-up pump (1.00) b) Yes(.5) (O.50)
- r. - _ c). Isolate the3 Clean-up System (.5). . _ _ . ( 0. 50.)
REFERENCE :_ ,, Oyster Creek 2OOO-ABN-32OO.19, RBCCW Failure Response SLO #12 Lesson Plan #41 ANSWER 7.05 (2.50) e) Shell and Steam inlet temperatures of isolation condensers Isolation Condenser area r'aom temperature Area and vent radiation monitors Isolation Condenser leve (.5 each) (2.00) b) Confirm that the condensate return valves are in the closed position with their control switches in Aut (O.50) REFERENCE Oyster Creek - Isolation Condenser System - Station Procedure 307, Pg 6/7 ANSWER 7.06 (2.00) 1) Has responsi bili ty and authority to cease refueling operation (0.50) 2) Ensures documentation of refueling activities (O.50) 3) Ensures that all Licensing Requirements for refueling are met (0.50) 4) Ensures that all personnel on his shift assigned to refueling are currently qualified in bridge control and refuel operations before any fuel handling is execute (O.50) REFERENCE Oyster Creek - Station Procedure 205.0, Pg 5/6
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. "P; PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27 ------------------------------------------------
SeDI96991Ceg_CgNJgg6 ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ' m s34, ANSWER 7.07 (2.00) a) On loss of DC signal, the feedwater control locks in as is valve positio (1.00) b) Fee'dwater valve can be controlled with the local _ . _ . _
.. h an d wh eel until dc. power is restore .
_
(1.00) ,
REFERLNCE Oyster Creek ABN-32OO.13, Response to lost of all 125 VDC Pg. 4 ANSWER 7.08 (3.00) 1) Confirm that all pump and suction valves are ope (1.00) 2) Maintain'the reactor water level between 155 and 165 inches above Top of Active Fue (1.00) 3) Scram the reacto (1.00) REFERENCE Oyster Creek - Recirculation Pump Trip - 2OOO-ABN-32OO.02 Pg. 6 j ANSWER 7.09 (2.50) a-3, 130- 150 psig b - 3, 130 -150 psig
c - 1, 50 psig d.- 4, 250-300 psig e.- 2, 60-100 psig C5 @ O.5*eal (2.50) REFERENCE Oyster Creek - Procedure 201.2 - Plant Heatup to Hot Standby Pg 3-11 Oyster Creek - RBCCW Failure Response - 2OOO-ABN-32OO.19 i I i ,
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PAGE 28 80919h991gg6_ggN_ Igg ( -W ,, ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ANSWER 7.10 (2.00) 1) Stop the power increas (1.00) 2) Enter Feedwater System Flow Control Failure Procedure (1.00) "~ _ - - . . . . _ . _ _ . .
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ANSWER 7.11 (3.00) 1) RPV Level below +138 TAF 2) Drywell Pressure above 2.0 psig 3) A condition that requires a reactor scram and power above 2 %. 4) Main Steam Isolation Valve Closure 5) RPV pressure above 1050 psig . 6) A condition which requires a scram in the judgement of the operator to either conserve RFV inventory or reduce the release of radioactivity to the environment.
! (.5 each ) (3.00) REFERENCE , Oyster Creek - RPV Control - Procedure No. EMG-32OO.01 Pg. 3 t I i
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h _ _ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 29
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* ,4 ANSWERS -- OYSTER CREEK
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-86/08/11-TURNER, *
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ANSWER 8.01 (1.50)
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The below must be confirmed by at least two independent indication (0.5) Misoperation in automatic mode (0.5) g. ,, ____.._,;_.;. ,_
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r 4 % 'i ) - - REFERENCE Oyster Creek, EMG-32OO.01 Pg. 6 ANSWER 8.02 (1.00) The Mode Switch must be locked in the REFUEL position (0.50) At least two (2) source range monitors (SRM) shall be operable and inserted to the normal operation leve (0.50) REFERENCE Oyster Creek Technical Specifications 3. ANSWER 8.03 (1.00) Abnormal rod motion is defined as motion in the wrong direction, motion when no motion is intended, or motion of a rod not selecte (1.00) REFERENCE Oyster Creek - Procedure 302.1 - Control Rod Drive Hydraulic System Page 3 "
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ANSWER - 8.04 (3.00) a) One Group Shift Superviser.yo
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b) One Group Operating Superv,isor c) Two Control Room Operators d)
._ - . - . . -Three Equipment
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e) One Shift Technical Advisor f) One Radiation Technician (.5 each for a thru e) (3.00) REFERENCE Oyster Creek - Procedure 106 - Conduct of Operations - Pgs 34/35
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ANSWER 8.05 (2.00) a) 1,2 b) 1 c) 3 d) 4 (.5 each) (2.00) REFERENCE Oyster Creek - Procedure 108 - Equipment Control, Pg's 15/23 ANSWER 8.06 (2.00)
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' a) The temporary change does not change the intent of the original procedur (1.00) , b) Two members of the GPUN management staff who are qualified Responsible Technical Reviewers approve the chang (1.00) REFERENCE Oyster Creek - Procedure 107 - Procedure Control Pg 26 i i l
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PAGE 31 ANSWERS -- OYSTER CREEK -86/08/11-TURNER, _
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ANSWER 8.07 (2.00) D a) From figure 3.10-2 and 3.10-3 the value is (1.00) b) Assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 degree Fahrenheit (specified in in 10 CFR 50.46). (1.00)
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I Oyster Creek Technical Specifications 3.10 Core Limits-. *
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ANSWER 8.08 (2.50) Primary containment integrity means that the drywell and adsorption chamber are closed and all of the following conditions are satisfied (.5): All non-automatic primary containment isolation valves which are not
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required to be open for plant operation are closed.(.5) B. At least one door in the airlock is closed and sealed (.5) C. All automatic containment isolation valves specified in (Table 3.5.2) ar operable or are secured in the closed position (.5).
D. All blind flanges and manways are closed (.5). (2.50) REFERENCE Oyster Creek Technical Specifications 1.13 ,1985 LRPLP 850.O.95 LO4 ANSWER 8.09 (3.00) a) Fuel Cladding Integrity (.5) - safety ~ limit is set such that no fuel camage is calculated to occur if the limit is not violated (.5) and is defined as the critical power ratio in , the limiting fuel assembly for which more than 9 9 . 9 */. o f the l fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties (.5). (1.50) b) Reactor Coolant System Pressure (.5) - The pressure safety limit was derived from the design pressures of the reactor pressure vessel (.33), coolant piping (.33) and isolation condenser (.33). (1.50)
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une- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32
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ANSWERS -- OYSTER CREEK -86/08/11-TURNER, ~ ~ ~ .
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ANSWER 8.10 (2.00) Cease all fuel handling (1.0) and cease all operations that could reduce the shutdown margin (1.0). (2.00) REFERENCE I -- z pg fer}ssN ' ~ l tiesii Cud o"' 8-63 ~ ~~~M--
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Oyster Creek Technical Specifications 3.5.8.3. . . . ANSWER 8.11 (2.00) Cold Shutdown wi thin 30 hours (1. 0) , TS 3. O ( . 5) , TS 3.8(.5) (2.00)
(2.00)
REFERENCE Oyster Creek Technical Specifications 3.0 & Oyster Creek Item Code 8-36 ANSWER 8.12 (3.00) YES, (.75).
The containment spray function is only required to be operable when primary containment is required to be operable. Table 3.1.1 (E), and note-(u) on pg. 3.1.13.(.75) Primary containment is not required,(.75) < 212 deg., RX. not critical., T.S-3.5.A. pg 3.5.1 (a).
A startup can continue until P/C is required, or if repair is complete prior to criticality.(.75) CAF for required administrative control (3.00)
.
REFERENCE Oyster Creek requi:ements for P.C, cnd l l }}