IR 05000309/1985037
| ML20151U009 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/30/1986 |
| From: | Dudley N, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151T664 | List: |
| References | |
| 50-309-85-37, NUDOCS 8602100434 | |
| Download: ML20151U009 (43) | |
Text
{{#Wiki_filter:T' . .. . U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-309/85-37 (OL) FACILITY DOCKET NO. 50-309 FACILITY LICENSE NO. OPR-36 LICENSEE: Maine Yankee Atomic Power Company 83 Edison Drive Augusta, Maine 04336 FACILITY: Maine Yankee EXAMINATION DATES: December 10-12,_1985 CHIEF EXAMINERi / W/f /- D ~ 8 h Noel Dudley, Lead Re ' Engineer Date REVIEWED BY: /[3d d ' Robert. Keller, CfPff, Projects Section 1C - Date _ APPROVED BY: k I Ifo Harry B. KisfJr, Chief,( Date ' Projects Branch No. 1 SUMMARY: Written Simulator, and Oral Examinations were administered to five Senior Reactor Operator (SRO) upgrade candidates. All candidates passed and five SR0 licenses were issued.
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- . . REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS: l SR0 l l Pass / Fail l l l l l l l Written Examl 5/0 l l l l l l l l Oral Exam .I 5/0 I l l-
I I I l Simulator Examl 5/0 l I l l I I I Overall l 5/0 l i I I 1.
CHIEF EXAMINER AT SITE: N. Dudley, NRC 2.
OTH'ER EXAMINER: G. Streier, EG&G 3.
Summary of generic strengths or deficiences noted on simulator examinations: Curves posted on the simulator control panel are not controlled. During the first scenario the shift supervisor had to request the replacement of one set of curves.
On two occasions the candidates operated the simulator outside of written procedures. On a major steam leak inside containment, the shift super-visor ordered the Power Operated Relief Valve (PORV) to be used to reduce primary pressure. There was no approved procedure for this recovery technique. Also, during the restoration of off-site power, following a loss of off-site power, the candidates relied on their training since they were. unable to find a procedure to perform this recovery evolutio _ _ _ _ _ _ _ _ __ . .
4.
Summary of generic strengths or deficiencies noted during facility review of~the examination: During the examination review the facility noted that allowance for critical operations up to 10P. power with no reactor coolant pumps is contained in the Technical Specifications, but is not an approved mode of operation. Also, the facility noted that the Technical Specification exception is irrelevant and never expected to be used again.
The facility noted that the candidates had been informed of a letter from Byron Jackson indicating that extended operations of a backstop and lift pump is not an operational problem. However, a precaution in Procedure 1-10-7, warns against operating the backstop and lift pump more than 5 minutes to avoid foaming.
These items indicated a discontinuity between training and written procedural guidance.
This issue was addressed by the licensee during the exit interview.
5.
Personnel present at Exit Interview: NRC Personnel N. Dudley, lead Reactor Engineer (Examiner) NRC Contractor Personnel G. Streier, EG&G Facility Personnel J. Frothingham, Manager, Operations Department R. Bickford, Operations Training Section Head R. Nelson, Nuclear Safety Section Head M.' Evringham, Supervisor, Operator Training Group M. Swartz, Supervisor, Simulator Group ' J. Sanoski, Senior Operations Instructor ' 6.
Summary of Comments made at exit interview: The NRC summarized the number and types of examinations administered.
The NRC stated that no serious problems had occurred during the simulator examination, however, the simulator instructor was concerned that a pre-viously experienced simulator abnormality might interrupt the examination.
The NRC noted that it is the responsibility of the' licensee to maintain the plant specific simulator as close as possible to the Main Control Room and to assure the fidelity of simulator response.
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The NRC questioned the licensee's position on maintaining procedural and Technical Specification limitations, which are no longer required, and the training conducted on these limitations. The two items noted during the facility examination review were used as examples.
l The licensee responded that it is their policy to keep their procedures and Technical Specifications free from extraneous or inaccurate informa-tion. There is a mechanism for identifying minor discrepancies with the Technical Specifications which would result in the correction of the dis-crepancies in the next Technical Specification Revision Request. Speci-fically, the licensee committed to review the applicability and need for the exception to Technical Specification 3.3.A.2.
Also, the clarification was made that even though a report from Byron Jackson was received which indicated a reduced concern for foaming in the backstop and lift pumps, the limitation will be maintained in the procedure and appropriate train-ing conducted to ensure compliance with the limitation.
The NRC noted that during the simulator examination the candidates were operating outside procedures and relying on their training.
The NRC stated that it was understood that the Emergency Procedures were under-going changes and that the operators were going through a transitional period, however, the licensee should be sensitive to operators developing a feeling that it is acceptable to routinely operate outside of emergency procedures.
The licensee stated that operators were not being trained to operate outside of procedures.
Specifically, the procedure for reducing primary pressure with the PORV had been written and reviewed, and was waiting final approval.
7.
Changes made to written exam during examination review: All facility comments contained in Attachment 2 were considered during grading of the examination, however, not all comments resulted in a change to the examination or answer key.
Question No.
Change Reason 5.01 Change "PCM" to " PPM".
Corrects units for boron concentration.
5.12 Delete " Unit I".
Makes question plant specific.
6.0lb.
Delete.
Candidates are not responsible for memorizing specific plant setpoints which are not safety relate _.
.
. Question No.
Change Reason 7.03a.
Delete.
Question is ambiguous since it is not clear that a definition of Technical Overexposure.
is required.
5.01 Change to "3 (remains According to Maine unchanged)".
Yankee Technical Specifications definition, shutdown margin would not change.
5.02a.
Change to " Lower Securing a reactor cool-(Due to lost pump ant pump would cause heat)".
reactor coolant temper-ature to decrease adding positive reactivity.
5.02c Add "(Lower if Allows credit for alter-MTC is positive)".
nate answer if assumption is made that MTC is positive.
5.02e Add "(Higher if MTC Allows credit for alter-is positive)".
nate answer if assump-tion is made that MTC is positive.
5.11b Change to "Subcooling Specific value for established".
amount of subcooling is not required for full credit.
5.11e Change to "PZR level Specific value for trending with charging / pressurizer level is letdown".
not required for full credit.
6.01 Add "(4 hour per Either battery amp-bus accepted)".
hour ratings or FSAR design criteria are acceptabl _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ .
. Question No.
Change Reason 6.03b Change " seal water Clarifies the loca-outlet temperature to tion of the temper- " seal' water HX outlet ature reading, temperature".
6.03c Change " Aux PZR spray" Modifies answer to to " fill header".
. reflect discharge paths specified in IAW E0P-2-70-5 " Emergency Bora-tion".
6.09 Add "CIS 2/3" and change Corrects logic of CSAS logic to "2/3".
Engineering Safety Feature logic.
7.02 Add "4. From ASP, shut Expands answer key-seal return MOV's; 5.
to include other RCS Drain OR-A-6".
objectives of an operator during a fire in the Protetted Cable Vault.
7.06 Add "(also accept Expands answer key to EFCV's)".
include another valve which should be checked if steam generator-pressure is low.
7.07b Add " 10'F".
Expands answer key 8.10a Add "(no, with correct Accepts other than the reference to RCP limits answer contained in the ' in OP-1-1)".
Technical Specifications 8.10b Add "two PORV's required Expands answer key to for LTOP considerations", accept answers for PORV operations under LTOP conditions.
8.13 Change to " Notify Plant Allows wider range of Engineering Department".
acceptable answers.
Attachments: 1.
Written Examination and Answer Key (SRO) 2.
Facility Comments on Written Examinations .
lYrrRGbrncn Y / ' ,'
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5. NUCLEAR REGULATORY COMMISSION SEtIIOR REACTOR OPE R ATOP. L ICENS E EXAMINATION FACILITY _d&lut_163dEE____________ PEACTOR TYPE _ EAR-CE _________________ DATE A DM IN I S T E R E D t _22412 /10__ ______________ EXAM!ilEP.
_SIREltEA_G4.____________ l APPLICANT! ()h.$_~[_k_I3___________ __ IUSIEUCIIQU1_IQ_aEELICAS11 Uso separate paper for the answers.
Wr i te 'an swers on one side only.
Staple question sheet on top of the answer sheets.
Points for each quOstion are Indicated in perentheses after the question. The passing grade requires at least 707 In sach category and a final grade of at least 807.. Examination paper s will be picked up six (6) hours after the examinati.on starts.
OF
, CATEGORY ?. OF APPLICANT'S CATEGOP.Y !__MALUE. _IDIAL ___SCOEE___ _ Ya L U E _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ C alL G uil _ _ _ _ _ _ _ __ _ _ _ 25.69 _Zis00__ _&trQ& ________ 5.
THEORY Uf flUCL E AR POWER PLANT ___________ OPER A TION, FLUIDS, A t40 Tile P MOD Y N A N I C S 24.oo 24 6 R _t2I29._ _ftr29 ________ 6.
PL APli S YS T E *15 DE S IGil, CONTROL, ___________ AllD IN S TRUMENT ATI0rl at. so 29.10 _&kr2ft__ _f2TUU ________ 7.
PROCEDUNES - NORMAL, A B il 3 R N A L, ___________ E N E P.G E N C Y AND RADIOLOGICAL C0flTROL 2 s. L 4 _22cQ2__ _&2r22 8.
ADMIN IS TR A T IVE PROCEDURES, ___________ ________ CONDITIONS, AND LIMITATIONS 91.E , '192x2C__ 100aQQ T3TALS ___________ ________ FIllAL GRADE _________________7 All cork done on this examination is ny own. I have neither given nor received aid.
___________________________________ APPLICANT'S SIGNATURE l l l
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IUEEdQQldad1C1 QUESTIUN 5.01 (1.00)
uith the plant operating at 855 power and a i systems in a nor7al configurations the operator bor a tes 100 Shutdown mar gin will
(1 0) ... 1.
Increase.
2. Decrease.
3. Remains unchanged.
QUESTION 5.02 (2.50) Coapare the C ALCULATED Estimatad Critical Position (ECP) for a startup to be performed 4 hours after a trip from 1000 ponar, to the ACTUAL centrol rod posi t ion (ACP)s if the following event s/condi tions occurred.
Considor each independently. Limit your answer to ACP is HIGHE4 than, LOWER thans or SAME as the ECP.
c. Une reactor coolant purp is stcpped two minutes prior to criticality.(Assume no reactor trip) (0.5) b.
The startup is del ayed until 8 hours after tne trip.
(0.5) c. The steam Jump pressure setoolnt is increased to a value just below the Steam Generator Saf e ty setpoint.
(0.5) d.
Condenser vacuum is reduced by 4 incnes of Merc ur y.
(0.5) e. All Stean Generator levels are being raised by 51 as the ACP is reached.
(0.5) QUESTION 5.03 (2.00) , a. Why is the indication of neutrons important durinJ Col d Shutdown? (1.0) . b.
State four posslote sources of neutr ons in the core durino Cold Shutdown.
(1.0) l . l QUESTION 5.04 (1.50) State the three basis for maintaining Rod Insertion Limitse, (1 5) l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****) l-
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IBEE50Q1B&BICE QUESTION 5.05 (1.00) Wnon the flow rate through a centrifugal pump is increased by Cp3ning the discharge valves the required NPSH _______, and the evallable NPSH ________. (INCPE ASE or DECREASE) (1.0) - w QUESTION 5.06 (1.50) TRUE or FALSE? a. The f aster a centr i f ugal pu ap rotatess the greater the NPSH required to preven t cavitation.
(0.5) b. One of the pump laws f or centr if ujal pu ps s t at es that the m volumetr ic flow ra te is inversely propor tional to the speed of the pump.
(0.5) c. Pump runout is the term used to descrioc the condition of a centrifugal pump r unn ing wi th no volonetr ic flow rate.
(0.9) QUESTION 5.07 (.50) In order to maintain a 200 F subcooling margin in the RCS when reducing RCS pressure to 1600 psig, steam cener a tor pres sure mus t be r educed to approximately e.
245 psig b.. 445 psig c.
645 psig d.
845 psi g (0.9) (***** C ATEGORY 05 CONTINUED ON NEXT PAGE *****)
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IUE15021NA51El QUESTION 5.08 (4.00) - c. ish a t ido parameter s can the operator control to prevent pressurized therma l shock? (1.0) b. What is the sequence of events that could lead to pressurized thermal shock conditions? . (2.0) c. Why does the concern about brittle fracture of the r eactor pr essur e vessel Increase as the Maine Yankee plant ages? Include in your answer the specific matcci al property that is affected.
(1.0) ' . (***** CATEGDPY 05 CONTINUED ON NEXT PAGE *****) - - - _ .-
,' . . . , 23__IUE0dI_QE_UUCLEAE_EQWER_ELASI_ DEES &Il08t_ELU10St_AMd pAGE S IBEEddQIdadlCS QUES TI0tl 5 09 (1.50) Ch0ose the correct response to each of the following.
(1.b) 1. Moderator temperature coef f icient becomes more negative from BOL to EOL primarily because oft - a.
The larger change in resonant escape pr obabili ty per degree' change in moder ator tenperature.
b.
The l ar ger change in core Icakage per degree change in moderator temperature.
c. More thermal neutrons ar e available for absorption in the moderator.
d.
The smaller change in th?rmal utilization factor per degree change in moderator temperature.
2.
Doppler coe f fi cient (pcm/ degree F fuel) becomes more negative from t10L to EOL because of.
a. An increase in effective fuel temper ature.
o. Clad creep and fuel pellet swell.
c.
The production of plutonium-240.
d.
The overlapping of resonant peaks.
3. Control rod wortn is greatest.
a.
At higher boron concentr a tions b.
At higher modcr ator temperatures c. At low boron concentrations d.
At lower moderator temperatures QUES TI0l4 5.10 (2.00) a. List two causes of waterhammer.
(0 8) b. Give two examples of how waterhammer can be minimized.
(1.2) (***** C ATEGORY 05 CONTINUED ON NEXT PAGE *****)
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luEddQQ1UAtIC1 QUESTION $.11 (3.00) For each of the p ar am e te r s listed belows provide the desired indication or trending that would be expected for natural cir culati on cooli ng and what might result if the p ar ameter was not trending as expected.
a.
Th b.
Subcooling c.
Steam Generator Pressure d.
Steam Generator Level o. Pressurizer Level (3.0) .- ' QUESTION 5.12 (2.50) -Thc(U$ EIlbreactor is operating at 507 power, BOL, when a steam dump f all s open.
Assume rods are in manual, no operator action is takens and no reactor trip occurs.
Explain HOW and WHY reactor power ~and Tave will change.
(2 5) . QUESTION 5.13 (2.00) Why is the time to reach a stable count rate after each incremental withdrawal of the contiol rods not a constant? Assume reactor does not reach criticality.
(2.0) , (***** END OF CATEGORY 05 *****) a
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i. o o QUESTION 6.
(4,4 M -- I a..What is e design length of time the battery banks can supply power to all DC and 120 voit AC vital loads during a complete loss of AC~ power-with both diesel generators inoperable? (1.0) b- %2t ! s the--m-i-ft t ette--OC-v o4t-49 e-t h e t-t-he-i n y e r-t e r1-c s n-r-e 9 u 14 t-e-tc-110 vcM5 AC4 ( 1. 0 h QUESTION 6 02 (2.50) HOW and WHY will the 12 steam dump and turbine bypass sys tem val ves respond if tne system is attempting to main t ain no-load Tave in not shutdown and the tamporature is S35 deg F7 (2.5) .- .. QUESTION 6.03 (4.00) a.
If the Residual Heat Demoval ( F,H R ) System is lined up to the purification system and the temper atur e out of the RHR heat exchanger is above 140 dea F7 kill the demineralizer resin be ' damaged? Explain your an..seY (1.0) b.
Explain how RCP seal is maintained watersupply} temperature 145 deg (8, (1.5) between 140 deg F and c. What flow path other than through cnarging pu (P-14A, B or S) can be used to emergency b or a te the primary.
Include TWO (2) poss ib l e sources and TWO (2) oossible discharge points.
(1.5) QUESTION 6.04 (3.50) a. What will happen if a Backs top and Li f t punp is run for greater than 5 minutes? (.75) b.
What actionss if anys should be taken if the computer alarm " Thrust Runner Oil Fl ow Minimum" is recei ved after s ta r t ing an RCP7 Explain (.75) . c. What conditions must be met in order to operate with two of the lower three seal stages failed? ~(2.0) . (***** C ATEGORY 06 CONTINUED ON NEXT PAGE
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QUESTION-6.05 (2.50) ~ What two parameters are measur ed to provide a turoine trip signal to reactor plant protec tion sy stem (RPS)? Indi c at e the coincidence required and which paraneter is considered the backup signal.
(2.5) . QUES TION 6.06 (1.50) Explain how the design of the spent fuel stor age r acks will provent criticality even if th e boron concentr a tion is reduced to zero.
(1.5) QUESTION 6.07 (2.00) .- If the #1 steam generator (S/G) level detectors which supplies an input sign'ai to the level c omp ar ato r portion of'the feedwater con tr o l systers fails LOWS and the plant is at 707 power, steady s ta te : a.
expl ain the immediate ef f ec t on #1 S/G level.
(1.0) b.
what will be the long term effects on plant operations if no operator action is taken? (1 0) . (***** C ATEGORY 06 CONTINUED ON NEXT P AGE *****)
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QUESTION 6.08 (3.00) True or False c. The Power Range safety channels use a fission cha.iber for detecting neutrons.
(0.5) b.
The core loading channel uses an uncompens ated ion chas! J 'or detecting neutrons.
(0,5) c.
The wide range indication i s made up of the Campbell Circuit signal and the output of th e countrate circuit at high power levels.
(0.5) d.
The Zero Moce 6ypass distable nper ates contacts below 157.
powers to allow oyoassing cer tain reactor trips.
(0.5) -- c. Uncompensateo Ion chanbers are b or on-10 lined, n i tr o ge n filled . detectorss operating in the Ionizatson region of the gas amplification curvet (0.5) f.
The power r ange con tr ol channel output gain potentiometers are located in the rear of the core loacing cabinet.
(0.51 QUES TION 6'09 (3.00) . If during reactor plant operations at 957 power a feedline rupture were to occur inside the containment, what are the THREE ^ Engineering Safety Features (E SFs ) that could possibly be actuated and wnat signals will cause th es e actuations? Include setpoints and logic.
(3.0) QUESTION 6.10 (1.00) TRUE or FALSE? a.
Each S af ety Injection Tank nas a flow restricting orifice in its di scharge line which is provided to extend the accumulator blowdown time which in turn reduces the peak fuel cladding temper ature in the event of a LOCA.
(0.5) b.
The design condensing capacity of the quench tank is based on accepting all postulated load rejections with no steam dump system availability.
(0.5) . (***** END OF CATEGORY 06 *****) -
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EaQ1DLDGICaL_COMIEDL QUESTION 7.01 (3.50) If a Xenon oscillation occurred as a resul t o f a r apid power reduction followed by a return to full powert e. how could the oscillation be dampened? (2.0) b.
what are the upper and lower limits associated wi h gr ou p 5A and B? (1,0) c.
what restriction exists when c on tr o l l i n g on group 5A and 587 (0.5)
OVES TION 7.02 (1.50) In accordance with EOP 2-90-1 (pl ant shutdown plan for fire) when a fire is detected in the Pr otected C able Vaults an operator is dis patched to per form THREE najor objectives. State those THREE major objectives.
(1.5) s. c o QUESTION 7.03 F2. 50 ) a,--Un4+r-what-T-W0-con d it-E ons c7 o a R 4 d ; a t-+-,eWerher-r-ec e 4-v e-a- -TEG+WI-gat-OV EREXPOS UR E-?- (-1.H A b'. What ac is required when a TECHNICAL OVEREXPOSURE has been receive (1,0) QUESTION 7.04 (1.50) What TWO peoples by position titles have to approve a loop cntry while oper a ting at power? (1.5) QUES TION 7.05 (2.00) What FOUR criteri a must be met to allow termination of an " unjustified SIAS"? (2.0) i I (***** C ATEGORY 07 CONTINUED ON NEXT P AGE *****) - -- .
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RAQ1DLDGICAL_CDUIEDL . QUESTION 7.06 (4.00) a.
Wh at F OUP. valves per EOP 2-70-4, must be checked closed if one or more steam generators are less than 400 pseo during a steam line break? (2.0) b.
TRUE or FALSE - a 1.
It is possible to damage tn'eJreactor vessel Dy allowing safety injection to continue woen it isn't needed? (1.0) 2.
An SIAS must be reset before the atmospheric dump valve can be used? (1.0) QUESTION 7.07 (3.00) ~ ,_ a.
In accordance with ADP 2-3 (HELB isolation system act lon), what FOUR syst~ ems are involved in the HELB i so l a t ion s che me, (2.08 What t emp er ature in th lower level PAB will caus e a HELS b. isolat ion actuation? (eInclude the logic require.d).
( 1 0) 00ES TI U;4 7.08 (1.00) During a loss of off-site power why is the auxiliary feedwater not realigned to the first point heaters until the steam generator levels are above 407 narrow range? (1.0) QUESTION 7.09 (4.00) a. What THREE required immediate actions are performed when 11 or core CEA's fall to insert upon receiving a reactor trip signal? (1.5) b.
State FIVE methods that may be use to increase subcooling if it is < 50 F f o l l ow i n g a reactor tri (1.5) c. Wnat TWO actions shall be performed following a Reactor trip if the core exit thermocouples indicate > 800 F? (1.0) . (***** CATEGDP.Y 07 CONTINUED ON NEXT PAGE *****)
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EAQ1DLOGICAL_CD3IEDL .QdESTION 7.10 (2.00) The following concern the Waste Gas Release procedure (OP 3-21-3): e. Who prepares the r elease permi t for the gas release? (1.0) b.
Who's permission is required to commence a waste gas release? (1.0) . (***** END OF CATEGORY 07
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QUESTION 8.01 (1.50) Tha inner door of the containment air lock leaks excessively.
The plant is at 100% power when a Maintenance Request (MR) to rcpair the gasket is brought to the PSS.
State the ac t ions PSS sh0Jld take and your reasoning.
(1.5) . -QUESTION-8.02 (1.50) For each of the f ollowing si tuati ons shoul d a WHITE tag, YELLOW tag, or NO TAG be used? (1.5) ~ a.
Reactor Protection System high pre s sur i ze r pressure channel A is byp assed.
b. ~Following maintenance on a service water heat exchanger a relief valve.ls blocked in accordance with a bydros tatic test procedure.
c.
A valve on the service wa te r s ys tem is to De repackea.
QUESTION 8.03 (2.00) a.
What is'the minimum shift crew composition for a cold shutdown plant according to the Tecnnical Specifications (1.25) b.
How long can the crew composition be below minimum before a Technical Specification violation occurs? (.75) QUESTION 8.04 (2.00) If during a reactor startup, with power at 10 E-10 power, the pressure in two Safety Injecti on Tanks drops to 200 psia, is it permissable to continue raising power? Justify your answer.
(2.0) l l l . (***** C ATEGORY 08 CONTINUED ON NEXT PAGE *****)
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QUESTION 8.05 (1.50) a. How many consecutive hours is an operator permitted to work while meeting the crew staffing requ i remen ts ? (.75) b. How many hours.may that person work in a 4 8-hour per io d? (.75) QUESTION 8.06 (2.00) What are TWO (2) required actions if a Safety Limit is violated according to the Technical Specifications? (2.0) O UE S T10N 8.07 12.50) ... In accordance with Maine Yankee's OP's Memo 9-E-3, prior to removing a Diesel Generator from service, you must perform a review of five (5) i n f o r.g a t. i on a l items to ensure compliance with Tech. Specs. What are these five i n f or m a t i on a l i t'e m s ? (2.5) . .. QUESTION-8.08 (2.00) ~ For each of the following leak locations, give the maximum leak rate allowable according to the Technical Specifications.
a.
Unknown location.
b.
Through a pr essur izer code safety valve to the Pressurizer Relief Tank.
c.
Through the wall of the line between the pr essur izer relief valves and the pressurizer.
d.
TOTAL Steam Generator' tube leakage.
(2.0) > (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****). . - - . 34 _ aDuld11IE AIIEE_EdDCEDUEESz_CudQ1Il0GSt_adt_L1111 All0NS PAGE
QUESTION 8.09 (2.50) Wh2never the RCS temperature is less than MPT and the RCS is not vcnted, the RCP's may not be s tar ted unless WHAT TWO condi tions oxist? (2.5) ' \\ - _ % r [! /A/ QUESTION 8.10 (2.50)
3 '- c vg a. Do the Maine Yankee Tech. Specs. ever allow operation of the qP,P z (J plant in a critical condi ti on with no reactor coolant pumps ssR c operating? Explain.
(1.5) g N b.
Explain when the Power Goer at ed Pel i e f Valve (s) are required by Tech. Specs. to oe operable.
(1.0) _.
_, QUESTION 8.11 (1.50) Explain what each wor d below means when it is used in a Maine Yankee procedure.
1. May 2.
Should 3. Shall (1.5) QUESTION-8.12 (2.00) What are FOUR of the FIVE required conditions necessary to consider the Steam Generators oper le for decay heat removal according to the Maine Yankee Tech. Specs.
(2.0) . (*****' CATEGORY 08 CONTINUED ON NEXT PAGE *****). - . . ... , , 2 __aDdlNISIHAIIME_EBUCEDUEESt_CONDIIl0diz_ABQ_L1511A110NS PAGE
QUES TION 8.13 (1.50) During a b ac ks h i f qqro r k is being done to replace and reposition supports for the discharge piping of an HPSI pump.
Tne workers cannot find the supports specified in the work package and have determined that it is impossible to ~ install the supports at the required locations.
Tho workers have located some pipe hangers in the shop and want per m i ss i on' t o.. i n s ta l l them as close as possible to the specified locations.
What actions, if any, should the shift supervisor take? Suppor t your answer.
- _m . (***** END OF CATEGORY 08
- )
(************* END @F @XQQRGoFROM ***************D
' \\ . . . , . h__ INEQ R1_DE_U u L L Ea3_EDh EE_ E La 3I_ DEEE AIIUu t_E L u1DS t_ AU Q PAGE
-IUE150DidadICS ANSWERS -- MAINE YANKEE-8 5 /12/10-S T R E I E R, G.
ANSWER 5.01 (1.00) -4-44no+-etse4-3 ( r e.m *'. n s ecbme.O (1.0) - REFERENCE C -E Mot-o r-Theo r-y_ m,n.3anwec.
tech.spcc. del.w'.W n ANSWER 5 02 (2.50) a. -G4#E-Lowec (du b ha pump WoA) b.
HIGHER ~ . . '5 f * 'n'* M c.
HIGHER (Lue<\\4 MTC
d.
SAME -' '? , -. ' ".. ' t (.. : - > < ? - - ,. u'.A.
n ' ' - ' - . each) (2.5) e.
LOWER L--~~ '" [35 R E F E R E t4C2 - C-E Reactor Theor y . ANSWER 5 03 (2.00) a. (Without some neutrons ther e is no instrument indication.
Without instrument indications you do not know core reactivity status or whether the status is changing.) Neutrons are needed to assure thst the reactivity status of the reactor is known and the instrumentation indication is available to guide operator action.
(1.0) b.
1.
Spontaneous fission of fuel 2. Reactor start-up sources 3.
Sustainer sources 4. Pho to neutr on s our ces 5.
Alpha-Boron reaction 6.' Alpha-D-16 reaction [Any 4 9 0.25 each] (1.0) lie F ERE NCE Gsncr al Physics Vol II chap. 5
- . . . . . 12__IUEDE1_DE_BUCLEAE_EQWER_ELa3I_DEEEaIl0ut_ELU1QSt_ABQ PAGE
IBEEdDuldadlCS ANSWERS -- MAINE YANKEE-95/12/10-STREIER, G.
ANSWER 5.04 (1.50) 1.
Ensure that' acceptable power distribution l imi ts ar e maintained.
2.
Ensure that the minimum SDM is naintained.
3.
Limit the potenti al effects of rod m i s al i gnmen t on tne as s oc i a te d accident analyses. (rod ejection) [0.5 each] (1.5) REFERENCE T. S. 3.10 basis ANSWER 5.05 (1.00) Increasess decreases (1.0) ^" REFERENCE General Physics Heat Transfer and Fluid Flow ANSWER 5 06 (1.50) a.
True b.' Fatse c. False [0.5 each) (1.5) REFERENCE Gsneral Physics Heat Transfer and Fluid Flow
' ANSWER 5 07 (.50) a.. (0.5) REFERENCE ! stoam tables I
. _,, a
_.
- . .. , 2A__IUEDEE_DE_UUCLEAE_E0 HEE _ELASI_DEEEAllDut_ELul0Sz_ABQ PAGE
IUERdQQ1UA51CS ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
ANSWER 5.08 (4.00) c. Coolant temper atur e [0.5] Coolant pressure [0 5] (1.0) ' b.
Rapid cooldown and depressurization [1.0] followed by rapid repressurization [1.0]. (2.0) c. Neutron exposure (integrated) [0 5 3 makes the material more brittle (raises NDTT) [0.53.
(1.0) REFERENCE C-E Lecture Concept of the Fracture Analysis Diagram --g - ANSWEF.
5 09 (1.50) 1.- d C.50] . 2.
c [.50] 3.
b [.50] 'e (1.5) REFERENCE C-E Reactor Theo r y ANSWER 5.10 (2.00) c.
Valve op er a t i o n, opening or closin; Pump starting or stopping (TLo Regu. ired ) Osci ll at ion of auto control valves (0.8) b.
Slowly opening of valves between voided and full sy s tems Proper venting of components Adequate level on tanks in systems where the tanks provide supply or surge function Proper use of s te a m traps and vents Pr op er sequencing of valves in pressurized systems (lido EetuVEd (1.2) . REFERENCE Nuc lear Ene rgy Tr ai ni n g, Thermodynamics, NET 4-2, pg. 2.1-4,5
' . . . . . , 2s__luEQal_QE_uuCLEa8_EDEE8_ELadI_DEEEaI10dz_ELu125t_aMQ PAGE
IUEEdQDIULd1CS ANSWERS -- MAINE YA NK E E-85/12/10-STREIER, G.
+ ANSWER 5.11 (.3. 0 0 ) s U P I* W bgw "T ~ ~a.
Th stable or decreasing [0.3) -5 Loss of Watural c i r c u l a t i on..f l o w.
[0.3J 5k. d subcooling b. -1G F i n co r e b[0. 3 3 *, At.t: Voiding r-nut leg whi ch would interrupt. flow. [0.3] c.
SG pressure tr acking Tave saturation pressure. [0.31 SG not removing heat. [0.31 d.
SG level in Narrow Range [0.31 SG no longer available as heat sink. [0.3] c. PZR Ieve1 5 C '. [0.3] Ped % w h caus % lt,+ b e Voiding of hot leg which would interrupt ficw. [0.3J ,d.-TCT iner ess es-[0.-3-1-becattse-th e cfTO heat-trmns f er-cao ac i ty. i s-r-educed- ( i ncr ease t prp e r a tu r e d u e t o l o we r-- c ond u c t i.v. i.t y_o f. c r.u d ). ~ -EO.43 - (3.0) REFERENCE General physics; Heat Transfer Thermodynamics and Fluid Flow fundamentals, p.
356-351.
ANSWER 5.12 (2.50)- T ave decr eases s ince more energy is being removed. [0.7) Rx Power increases due to the. positive reactivity added through MTC and doppler. [0.8) Power stabilizes at a higher.value. [0.5J Tave stabilizes at a lower.value. [0.53 (2.5) R G E R E ^) c-T. * c.E h <_ b v- %eg roonuaA ANSWER 5.13 (2.00) With each incremental rod withdrawal, Keff is incr eased towar d a value o f 1.00.
The larger the value of Keff the l on g er it will take to reach a new higher neutron. level since more generations will have s i gn i f i canc e.
[2.03 REFERENCE Introduction to Nuclear Engineerings Lamarch, C hap 8, Sec 8.2, p 313 Chao 7, S ec 7.5, p 295-298 Chap 4, Sec 4.1, p 102
' . . . 6A-ELAHI IXSIEUS_DESIGut_CDMIEDLa_AUQ_luSIEudENIAIIQU PAGE
ANSWERS -- MAINE YANKEE-85/12/10-STREIER. G.
3.00 ANSW ER 6.01 42.00) a.
8 n o u rfmle ma(9 bucs e,s lou s utopW.1) (1.0) t w lO-5-vo4t= AC 4h0 )- REFERENCE Ch. 33 pgs. 48-49 ANSWER 6.02 (2.50) Dump valves shut.
Valves do not open until 5 deg F above setpoint (552 deg F ). C. I. G 2 nypass valves open 3 bypass va! ves r amped open Two groups of bypass v al ves open sequentially on difference in pressure above 900 psia.
Tave of 535 deo F pr oduces 915 ps i a in SG which will fully open one set of valves and ramp open second set of valves.[i 5] (2.5) REFEkENCE Vol V Cnp. 25 pgs. 8 C 12 and Fig. PSG 14-8 ANSWER 6.03 (4.00) a. Yes. ( Res in is damaged by temperatures above 140F) RHR in terf aces wi th CVCS downstream of neat exchanger.
Bypass valve senses temperature downstream of HX and will not bypass O demineralizer.
( 1./') 9x b.
The T on condensate discharge line is controlled oy seal w a t e r" 2u t t e_t_ t empe r a t u r e and regulates HX condensate level.
Condensate level e f f ects HX heat tr ansf er surface.
(1.5) c.
Tnrough aux i l i ar y charging pump RWST, BAST, BAMT (any two) f' charging to Loop 2 or 3 r -Au ; "ZP r p r ay, HPSI header (any two) ( L.,6) c;o A e de v REFERENCE Val I Chp. 4 pg. 39 C 67 and Fig. NS-4-3 _
' , ' . .. , hi__ELaGI_11SIEd1_DESIGBt_CDUIEDLt AdQ_IUSIdudEMIallOS PAGE
ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
ANSWER 6.04 (3.50) lost.)
d ev e l o p (a nd a.
Foaming will oil pressure will be (.75) Co G b. No action.4When RCD comes up to speed the back stop pump auto issensea.Eo.35] (.75) shuts down and a low oil pr e s s ure @ lee 2.
c. Lower seal cavity temperature (below 200 deg F.
Seal water return flow temperature (below 200 deg f.) Low 6DIncreased leakage can be handled by CVCS.
6) Seal water supply flow to RCP seals creater (than 5 gpm)above seal water return flow.
P' # S* 4 **WC (2.01 REFERENCE Proc. # 1-10-7 pgs. 4E8 .- ANSWER 6.05 (2.50) afg.
Os Main turbine auto s':op oil p r e s s u r e [ 0. 51(}14.,[ 0. 5 ]. Tur b i ne generator'stop valvas shut [ 0. 5 ] 4/4 (in par ~allel) [0.5]. Valve position i s the backup [0.5]. (2.5) - REFERENCE Chap. 11, pg.22 and Fig. NS-12-9 } ANSWER 6 06 (1 50) , Storage facility' designed to maintain 20-inch center to center spacing between each assembly and shleided wi th Boral plates.
(1.5) R E FE etnc t'. ANSWER 6.07 (2.00) a.
FWRV control system will "s e e" a low a/G water level and attempt to recover level [0.5]. #1 S/G water level will raise ~[0.5]. -(1 0) b.
Turbine trip on nigh S/G level (at 917.)
(1.0) _ _.
_
- . . . . , ka__ELANI_11SIEd1_DE11ENt_CDMIEQLt_AH2_lBSIEudEblallDU PAGE
ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
-REFERENCE System description s30 and #25 pg. 22 . ANSWER 6.'08 (3.00) ' a. False (0.5) b.
False (0.5) c. True (0.5) d.
False (0.5) e. True (0.5) f.
True (0.5) REFERENCE Excore Instrument s ys t em des cr i p t i on -.. ,. ANSWER 6.09 (3.00) SIAS [0.5] and CIS [0.5) - High c ontai nment p r essur e [0.3], ,, 5.0 psia [0.~33, 2/4 [0.2) C16 'v2 CSAS [0.5) - High containment pressure C0.3] 20 osig [0.3) 2//3[0.1] (3.0) REFERENCE _ ~~ ~ Volume II sect 1, 3 E 4 ANSWER 6.10 (1.00) a. False (.50) b.
False (.50) REFERENCE Chapter 6 pas. 26-28 _
- . . - . 22__EEQCEDURES_:_dDEUALt ABUDEL4L4_EdEEGEUC1_a3Q PAGE
EAQ10LDGItaL_CDBIEDL ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
ANSWER 7.01 (3.50) a.
If the symmetr ic of f set (S/0) is above the upper control line insert rods [0.53 un til the S/D is at the upper control line [0 5]. As S/0 decreases below the upper control line CO.5] withdraw rods [0.53.
-(2.0) b.
90 steps [0.25] or above PDIL [0.25); ARO [0.51.
(1.0) c. Ma in ti an subgroups SA and 50 within 3 s teps of each other.
(0. 5 ). REFERENCE Procedure 1-8-1 (3hyd<cs) ANSWER 7.02 (1.50) 1.
Remove. power from the p0RV's 2. Man the Alternate S hu tdown P a n e l (ftS P') 3.
Start Di ese l G ener a tor #2 (1.5)
- i. From A SP~ shi seaA reb.% maid 5. R C S b m h M 'l'
-PEFEKENCE E0P 2-90-1 1. 0 ANSUER 7.03 (-3,-54 a l._W herua n-i ndhr Fdua t r'ec el'res-ex t'er na l-r-etf Fa ti-o n-ex po sure-t r-e xces s-o f-r e7utTro rrt inri tr-i n any-cel-ender-quar _ter [ 0. 7 5 jn
Wnen-en-indbritua-i is exposed Io a i r o o r n e-rs dFonet14e-mat er-i-e I-4n-emces s o-f i e y JtTr0 r y IImIts (B77t h 6-1 s+} l a.. h(' Th e individual shall be r emoved from further exposure during the remainder of the applicable period.
(1.0) REFERENCE Radiation Protection manual pg. 2-6 l l' l . .
, - - - - .
. .. . , Ic__EaQCEDURE3_=_UDEdaLA_aBUQEdakt_EdEEGEUC1_aUD PAGE
BADIDLOGICAL_CDUIRQL ANSWERS -- NAINE YANKEE-8 5 /12/10- S TR EI ER s G.
- ANSWER 7.04 (1.50) 1. Plant Manager 2. Radi ol ogica l Contro ls secti on head.
(1.5) REFERENCE Radiation Protection manual pg. 2-9 ANSdER 7.05 (2.00) 1. RCS subcooling >50 F 2.
Pzr. level >SO 7 3. One S/G at least lgp" WR <- - 4. SIAS no longer required to maintain Pzr level o r p r e s s'U r e. (2.01 RE F E R E N.C E A.
E0P 2-70-3 pg. 2 ANSWER 7.06 (4.00) ,, a.
1.
Feed reg valve 2.
Bypass feed reg valve 3.
Aux feed reg valve '4.
Aux feed isolation valve [0.5 each) (2.0) Cans acca p + a ve v'13 b.
1.
True (1.0) 2.
True (1.0) REFERENCE EDP 2-70-4 pg. 1-2 . +
- e
"' ; ~ _ p.
. - -- -
..-. - -. -- - -. - . - -. -- -
- . - . . . , Zs__ERQCEQuRES_=_BDEd&Lt_ABUQ2dalt_EdElGENC1_aBQ PAGE-26 SADIDLDGICAL_CDUIEDL ANSWERS ---MAINE YANKEE-8 5 /12 /10-S TR E I ER, G.
ANSWER 7.07 (3.00) a.
1.
S/G blowdown 2.
Letdown system 3.
Aux steam to PAB 4.
Aux steam to spr ay building (2.0) .. b.
150 'Ds 75 ] s 2_ ou t of 2 logic [0.25] (1.0) " flO*F REFERENCE AOP 2-3 p2
.. enes ANSWER 7.08 (1.00) To prevent thermal stress [0.51 and water hammer to the feedring [0.5]. (1.0) REFERENCE E0P 2-70-9 pg. 1 .- .. ANSWER 7.09-(4.00) a.
1.
Open mg set output bre ak er s 2.
Emergency berate 3.
Reenergize CEDM's and dr i v e roos (1.5) b.
1. Increase chargi ng/r educe letdown 2.
Prevent unnecessary heatup 3.
Energi ze heater s when level > 28f.
4.
Reduce spray 5.
Isolate PORV .. 6.
Initiate SIAS [3ny five 9 0.3 eacn] (1.5) c.
1.
Ini ti ate SI AS 2.
Open PORV's and PORV's block valves (1.0) REFERENCE EOP 2-70-0 .
. - . .
' Is__ EEQC EQuRE1'_=_d Dad &LA_aShl0 R5 &L t_EHERGENC1_ A NQ PAGE
R&Q1DLOGICAL_CDMIEQL ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
ANSWER 7.10 (2.00) a. Chemistry department (1,o) -b..The PSS (1,o) i REFERENCE OP 1-21-3 pgs. 1-3 . fY , '*w f .
.- , . - . Ha__AQd1HISIEAIIVE_EEQCEQUEElt_CONDIIl0 Ult ANQ_LidlI&llDUS PAGE
ANSWERS -- MAltlE YANKEE-8 5 /12 /10-STR EI ER, G.
ANSWER 8.01 (1.50) Do not en te r a i r-lock for repairs.
Entry to air lock would violate Tech Spec cont ainment requirements.
(Walt until SD or enter through escape hatch) (1.5) REFERENCE T..S.
3-11-A ANSWER 8.02 (1.50) a. No tag b.
Yellow tag c. White tag (0.5 each) (1.5) ,f REFERENCE Proc 16-1, Cps meno 9-E-8 A NS tJ ER 8.03 (2.00) a.
1 SQL 1 OL 1 Non Licensed (1.25) b.
2 hr (.75) REFERENCE T. S. Table 5.2-1 ANSWER 8.04 (2.00) No., cannot change-modes with reliance on remedial a ct i on s. (Can commence power escalation af ter SIT pressure is restored) (2.0) REFERENCE T. S.
3-0 pg. 1 .
. .. a ..- . ~Rs__ AQuid1SIE AIIEE_ESQCEDUEES t_CDd u1110dS t_ AdQ_ L1311& IlQd3 PAGE
ANSWERS -- MAINE-YANKEE-85/12/10-STREIER, G.
ANSWER d.05 (1.50) a.
16 hours (.75) b.
24 hours (.75) R EF ER EllCE Proc. 1-201-3 pg. 1 ANSWER 8.06 (2.00) 1.
The facility shall be placed in at least a hot shutdown concition within one hour.
(1.0) Ga.6 3 2. The NRC shall.be notified as expeditiously as possibles4but within 24 hours.
Ops shall not be resumed antil au th o r i zed by the commission]. (1.0) (0.W] REFERENCE T.
S.
2.0 pg. 1 , wgy:* ANSWER 6.07-(2.50) 1.-White tag book 2.
Yellow tag book 3.
ECCS s ta tus bo ard 4.
Shift turnover 5.
Control room log b o ok [0.5 each] (2 5) REFERENCE Ops memo 9-E-3 ., . . . - - - _.
.-- ._. - _ _.
r - . - . ... , E4__AQ51d11IEAIIME_EEQCEDUEElz_CONDII10 Nit _AUQ_L1511AI1081 PAGE
ANSWERS -- MAINE YANKEE-85/12/10-STREIER, G.
ANSWER 8.08 (2.00) a.
1 ypm [0.5] b.
10 gpm [0.5) c.
O gpm [0.5] d.
1 gpm [0.5] (2.0) REFEREllCE Technical Specification 3.14 ANSWER 8.09 (2.50) -a.
Pressurizer level is <807 [1.25] D.
S/G Temperature is <100 F a bo-v e RC S temp. [1 25] (2.5) R EF EP. ENC E Maine Yankae T.S. 3.4 ANSWER 8.10 (2.50) a. Yes E0.75]- During initial te ting to permit power lev els not to exceed 107 of rated powers natural circulation is permitted.
[0.75].4 v;g 4ce,g no, w 4 gyee,y u4,.<cwee 4o Re? L td.h Wo9-8-h(l. 5 ) - D.
At least one PORV is requ ir ed operable whenever RCS temp >210F.
(1.0) [Two Ford's op;<< d for 'T * " ^I'6"'O'"5 3 " "' REFERENCE ' ' Naine Yankee T.S.
pg.
3.3-1 .
TL ' , . . ..- - - , as__&QB181EIEAIIEE_EEDCEDUEESt_CD3DIllDUEt_auQ_ Lid 11allDUS PAGE
ANSWERS -- M AIN E YANK E E-85/12/10-STREIER, G.
ANSWER 8 11 (1.50) 1.
Denotes permission 2. Denotes a recommendatio,n 3. Denotes a requirement (1.5) REFERENCE Proc. 0-06-1 pg. 1 ANSWER 8.12 (2.00) e.
~ 1.
RCS pr ess. - 100 ps i > saturation press.
~2.
Tc & Th stop vavles open 3.
S/G water level above the top of the tube bundle 4.
Inventory of > 100,000 gallons of primary grade feedwater 5.
Feed pump available C4 a 0.5 each] (2.0) - REFERENCE T.
S.
3.8-9 "- - - - .. _ ANSWER 8.13 (1 50) In +3a e a Repair n.de r-e nd-for-.44 r d t e the Pl ant Engineer ing c Department [0.6] Job cannot be handled as On-the-S pot C hanges. [0.9] (1.5) REFERENCE Proc. 0-07-4
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