IR 05000286/1986005
| ML20197J245 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/09/1986 |
| From: | Keller R, Kister H, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20197J224 | List: |
| References | |
| 50-286-86-05, 50-286-86-5, NUDOCS 8605190401 | |
| Download: ML20197J245 (44) | |
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EXAMINATION REPORT Examination Report N (OL)
Facility Docket No: 50-286 Licensee: Power Authority of the State of New York P.O. Box 215 Buchanan, New York 10511 Facility: Indian Point Unit 3 Examination Dates: March 10-14, 1986 Chief Examiner: /d/b Btrry S(,Norris, ctor Engineer dM N 0' ate (Examiner Reviewed by: ~
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Robert M. Keller,' Chief N8'? Da%
Projec s Sec io 1C Approved b _
_ .4 ) fp HafryB.K1s(ep, Chief ' [ Tate /
Projects Branch No. 1 Summary: Seven Senior Reactor Operator (SRO) candidates were examined during this period, six SRO-upgrades and one SRO-Instant (retake); five SR0 candidates received licenses. One SRO failed the written examination, and one SRO failed the simulator examination.
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-D"y 8605190401 860509 PDR ADOCK 05000286 gj PDR
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REPORT DETAILS TYPE OF EXAMS: Replacement EXAM RESULTS:
l SRO I l Pass / Fail l I I I I l l Written Exam l 6/1 I I I I I I I l Oral Exam l 6/0 l l l l 1 l l l Simulator Examl 5/1 l l l l l l l l Overall I 5/2 I I I I Chief Examiner at Site: B. S. Norris, USNRC Other Examiners:
D. M. Silk, USNRC P. T. Isaksen, EG&G l
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3 Summary of generic deficiencies 'noted on simulator / oral examinations: Candidates were unfamiliar with subcritical multiplication as shown by their inability to perform simple calculations of reutron popula-
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I tion in a subcritical reacto Candidates were unable to determine if a required surveillance per '
Technical Specifications had been accomplishe . Summary of generic deficiencies noted from grading of written examination:
Candidates demonstrated a general lack of knowledge with respect to radiological controls, radiation monitoring, and radwaste procedure (Questions 7.04, 7.05, and 7.06) Personnel Present at Exit Interview:
NRC Personnel .
B. S. Norris, Reactor Engineer (Examiner)
D. M. Silk, Reactor Engineer (Examiner)
N. F. Dudley, Lead Reactor E ;ineer (Examiner)
P. S. Koltay, Senior Resident Inspector Facility Personnel R. Tansky, Training Superintendent W. A. Josiger, Resident Manager E. Diamond, Operations Training Supervisor R. E. Roberstein, Operations Training Instructor (ETS)
8. J. Ray, Training Coordinator 4. Summary of Comments made at exit interview:
The examiner summarized the examinations which had been administered and presented the weaknesses noted on the operational examinations. The examiner stated that the Indian Point 2 simulator is not considered to be a plant specific simulator for the Indian Point 3 control room due to the major differences in the control board arrangement, facility equipment, and operating procedures. The examiner explained that the NRC no longer administers simulator examinations on non plant specific simulators due to the difficulty of evaluating a candidate's ability to operate his facility through observations of his responses in a dissimilar simulato The examiner emphasized that the Indian Point 2 simulator should be used as an integral part of the Indian Point 3 training program. However, future licensing examinations st Indian Point 3 would not include use of the Indian Point 2 simulato . _ _ _
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The facility stated that further evaluation should be conducted before disqualifying the Indian Point 2 simulator for examination purposes. The examiner conceded that some evolutions and board manipulations on the
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Indian Point 2 simulator may be similar enough to Indian Point 3 to allow
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adequate evaluation of a candidate's ability to operate the Indian Point 3 facility. The facility will evaluate the use of Indian Point 2 simulator for evaluation purposes and may submit a proposal to the NRC for consider-atio The facility commented that the simulator scenarios were not' credible, involved more than a single fault, and placed the simulator outside the designed basis of the reactor plant. The facility noted that the scenarios examined candidates on plant conditions beyond the license of the facility. The t7aminer responded that scenarios were developed to
'; evaluate a candidate's ability to effectively use the Emergency Operating Procedures and at times would place the facility outside license conditions. However,-the examiner acknowledged that the complexity of simulator scenarios was increasing and would need to be reviewed to I
ensure an appropriate level of difficulty was maintaine *
A discussion of the requalification program evaluation scheduled for June was conducte The examiner expressed appreciation for the support of the facility staff during the conduct of the simulator examinations. The facility
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commented that the written examination quality was very good overall and expressed appreciation for the depth of examination review allowed by the
- examiner . Examination Review
The facility comments on the written examination were discussed at the examination review held March 14, 1986. (See Attachment 2)
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Attachments: Written Examination and Answer Key (SRO)
- _ Facility comments on Written Examination and NRC Resolution i
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. . NUCLEAR PEGULaTOR' COnHIOSION SENIOP REACT 0F OPERATOR LICENSE EXAMINATION FACILITY: INDIAN POINT 3
_________________________
REACTOR TYPE: PWR-WEC4
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DATE ADMINISTERED: 86/03/11
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EXAMINER: SILKr _________________________
' INSTRUCTIONS TO APPLICANT:
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Use separate paper for the answer Write answers on one side oni Staple question sheet on top of the answer sheet Points'for each question are indicated in parentheses after the question. The passing grade requires at least 70% ir. each category and a final grade of at least 80%. Enamination papers vill be pickeo up sin (6) hours after the enamination start % OF CATEGORY % OF APPLICANT'S CATEGORT VALUE TOTAL SCORE VALUE CATEGORY
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_I'5.00
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____ ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND TNERM0 DYNAMICS 25.00 25.00
________ ______ __._________ ________ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUhEt4TATION
_'5.00l______ _l'5.00
____ ___________ ________ 7 PROCEDURES - NORMAL, ABNORNALe EMERGENCY AND RADIOLOGICAL CON 1ROL 2'"4.^'0 0 i 25*00
_;;_;;__ ___ __ ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS-
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100.00 TOTALS
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l FINAL GRADE _________________% j All work done on this examination is my own. I have neither l given not received ai ~~~~~~~~~~~~~~
5PPL5CAUTI5~55GU5TURE
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. THEORY OF NUCLEAR F0WER PLANT OPERATION, FLUIDS, AND PAGE 2
THERn0DrNAMICS
OUESTION 5.01 (1.50) How anc why will Deta-bar-Effective change over core life? (1.0) What affect does a larger Beta-bar-Effective have on reactor period? (0.5)
GUESTION 5.02 (2.00) The reactor is suberitical by 2.5% delta-k/k. The count rate is lia CP After a positive reactivity insertion, the count rate increases to 34 How much reactivity was added to the core? (1.5) Why does it tske longer, after each reactivity additso , for the neutron population to reach equilibrium as keff approaches 1.0? (0.5)
GUEf. TION 5.03 (2.00)
Pipe elbows are used for flow measurements of the RC Describe how pressure olfferences occur at the elbows and state how different1s1 pressure is related to flo QUE STI0rJ 5.04 (2.5 Give three reasons whv the Doppler Temperature Coefficient becomes
?ess negative as fuel ten. pet stur e increases? (1.5i How ano whv does the Doppler Temrerature Coefficient change over core 12fe' (1 0, GUESTION 5.05 (2.50>
The p l a re t is at 5 0 *. Powe What affect will an increase in condenser cir-culating water flow have on pressurl:er l e v e l '- Explain your answe Assur.ie All control svstems in manua (****z CATEGOR) 05 CONTINUED ON NEXT PAGE *****)
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REACTOR C00LANT FLOW ,, > 323600 GPM
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RATED POWER (PERCINT)
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.. THEOR) 0F NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3
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QUESTION -5.06 (2.00) What is the difference between minimum NPSH and available JPSH? (1.0)
l ?lhen does cavitation occur? (1.0)
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GUESTION 5.07 (2.00) -
' If a reactor core safety limit was exceeded, what physically could
Jhappen to dama3e the fuel? (1.0) If the reactor is operated beneath the curves in Figure 1,.what two adverse core cond2tions are prevented? (1.0)
00ESTION 5.08 (2.50)
' Why does the noderator Temperature Coefficient (NTC) become positive-during high boron concentrations and high RCS. temperatures? (1.5)
- Explain how and why hTC changes with core age? (1.0)
l- DUESTION 5.09 (2.00)
A reactor has the following characteristics at 100% power: Teve is 573.5 F-1 and Tstn. is 513.8 The plant is shutdown and 5% of the steam generator ,
j tubes are plugge The plant returns to.100%. powe Given that Tave is
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agais. 573.5 F. c e t e r nii r.c the pressure of the steam leaving the steam gen-erato State. assumptions and shoo all wor QUESTION 5.10 (3.00 Describe a Xenon oscillation by explaining how it starts and whv it
- oscillstes? (3 0)
l GUESTIO4 5.11 (3.00)
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1 Define F0(Z)? (1.0)
' List the four conditions that' ensure that FQ(Z) is maintained within
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limits. (2.0)
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. PLANT SYSTEh5 DESIGN. CONTROL, AND INSTRUhENTATION PAGE 4
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GUESTION 6.01 (1.50)
What two oesign features of the component cooling water system minimize the effects of a rupture of the thermal barrier?
QUESTION o.02 (1.00)
What is the design basis for having all three pressurl:er code safety valves operable woen tne reactor is above cold shutdown condition? (1.0)
GUESTION 6.03 f.1.50) Hou is leakage into the Incore Flux Detection System detected during norn.a3 oper ations? (0.5) What two conoitions would alert the operator of a reactor coolant lea *
In an incor e thimble) (1.0)
GUESTION 6.04 ti.00's What is the purpose of R-16 ano R-23, Containment Fan Cooling Water honitor si (1.0>
,C QUESTION 6 05 s2.50)
Assume all s v s t e n.s in automatic and plant Power is 100 What Automatic Actions will occur for each of the f ollo'aing f ailures? A cold leg temrerature channel fails hign? List five. (1.50) st stage pressuie channel A (Channel 1) falls low? List thre (1.0)
Consider ecch indepenoentl (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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. FLANT S) STEMS DEEIGt4, CONTROL, AND INSTRunENTATION PAGE 5
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DUESTION o.06 (2.00i What conditions must exist for an automatic initiation of the carbon dioxide fire protection system in the Cable Spreading Room? (1.0) How can a carbon dioxioe discharge be prevented once the system is activated? (1.0)
QUESTION 6.07 (2.40)
State how the following components respond when instrument air pressure drops below 30 psis with the plant at 100% powe Seal return valves (261 A-D-C-0) Charging pump VCT makeur valves (FCV-1108) Containment. pressure relief valves (PCV-1190) nainstesm isolation valves (MS-l'si Auxiliary feedwater regulators (FCV-405's and 406's)
GUESTI0tv 6 08 (3.00)
How oc the 480 volt buses, and related systems, respond when a blackout condition occurs and a safety injection signal develops?
GUESTION 6 09 (3.10> The MSI"'s are closed cn High Oteam Line flow and High-High Containment Pressure ESF Signals. Explain the reason for each and include the type of sccident for unich each is designe (1.5) High-Higo Containment Pressure causes a Phase B Containment Isolatio What two actions must ce tak en before another Phase C can occur? (0.81 Safety injection 25 sctuated. After the two minute interlock has timed out, the safetv injection signal is rese An automatic init-lation signal cannot reactivate safety injection until what two actions have been taken? (0.8)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE xxxxx)
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. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUhENTATION PAGE 6
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GUESTION o.10 (3.00) How could a reactor trip occur if both Intermediate Range Detectors are overcompenssted? Explai (1.5) The plant is at 100% powe NI-41 fails hig Why should the oper-ator put the Delta-T Defeat Switch to the detest Loop-1 position?
(1.5)
GUESTION 6.11 (4.00) Indicate whether the OT Delta-T and the OP Delta-T setpoints will in-crease, decrease. or not change if the following changes occur. Con-sider each change independentl (2.0) The N-41 lower oetector fails low Overdilution of the RCS, which causes rods to insert slowly to maintain constent 1cao ano Tave Steam generator safetv relief valve opens with rods in manual Justifv your answers for part a. sbov (2.0)
(***** END OF CATEGORY 06 *****)
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. PROCEDURE 5 - NORh4L. A E:NORnAL, ENERGENCY AND PAGE 7
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DUESTION 7.01 (1.50) What is the Immediate Operator Action following an Indication of a failure of a heat tracing circuit on original plant esulpment? t.75> Can power operation continue if one channel of heat tracing associateo with the boron injection tank is out of service? Justifv vour answe t.75/
OUESTION 7.02 (1.00)
In accorcance with 50p-RCS-1, Reactor Coolant Pump Operation, explain uhv t n e before star ting a RCF. with no other RCP's running, ther e must be a gas bubble in the RCS or a complete temperature equallustion between water in the reattot ves.sel and water in the steam genersto QUESTION 7.03 (2.00) What Indication would alert an operator of a I iC A outside of contain-ment? (0.5> If an NPO is dispatched to search for the leak, where would be the two most likely places to start to search in accordance with ECA-1.2, LOCA Outsice Containment? (1.0) In a attempt to identifv and isolete the break, various valves are sequenciallv closeo and opene What Indications would the control rcom operctor nave that ihe break had been isolated? t0.5)
GUESTION ' .04 (2.50) What three approval signatures are required on a Joe Specific REA prior to starting work? (1.0) What four factors go into the determination of the ha::imum Permiss-ible Concentration? (1.0) How does the operator distinguish between a ' crud burst' and a fuel failure? (0.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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. 7 PROCEDUREE - NORnAL, ABNORnAL. EhERGENCi AND ,
'FAGE 8
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OUESTION 7.05 (2.00) / When stripping fission gases from the RCS while at Power, a precaution is to closelv observe VCl pressure and leve Whyet (1.0)
i In accordance with SOP-WDS-7, Gaseous Waste D1'scharge, when must a gas release be discontinued? (1.0) '
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OUESTION 7.06 (3.00) What four Automatic Actions should be verified if the Fuel Storage Building (FSDe Area Radiation honitor. R-5. s l a r nis due to an a r r adiat-eo fuel assemblv being damaged curing handling? (2.0) Unoer what ec.nditions may tne FSB Emergencv Ventalation System be placeo in sn inoperable conottion when spent fuel is in the FSE? (1.0)
GUESTION 7.07 (3.50) Describe hou a noncondensaole bubble coulc form l'r i the reactor vessel headi (0.5) What potential problem covid arise as a result of a noncondensable bubble in the reactor head? (1.0)
, With RCP's availabloe what three methods can be used to remove the nonrnnnenubia bubble from the reactor vessel head? Enplain how these methods result in reducirg the bubbl (2.0)
GUESTIOr1 7 . 0 E, (3.00, During an accident recovery proceoure, the operator who has been monitor-Ing tne Critcal Safetv Functions r eports the follouang inf or matio Primary integritv -
orange Reactor coolant inventory -
yellow Core cooling -
yellow Containment -
green Suberiticality -
orange Heat sink -
red Rank the CSF's according to their priority in numerical order from 1 to 6 with 1 having the highest priorit (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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. PROCEDURES - r1 0 R h A L , ABNDRnAL. EhERGENCi AND PAGE 9
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GUESTION 7.09 (3.00)
According to ONOP-RCS-2 (Malfunction of Pressurizer Control System), what five actions / verifications are directed to be followed if pressurizer Pressure is decreasing below the control / alarm setpoint and a reactor trip has not yet occurred? Contingency actions (reponse not obtained) are not required for these five actions.-
GUESTION 7.10 (3.50)
The RCS is in a solid condition with temperature and pressure at 100 F and 300 ps19 The RHR s y s t e n. is operating 'alth one pump and is letting down to the CVC One charging Pump is operatin The ' Valve 730 or 731 tJ o t Fullv Open* znnunciator is teceive What immeaiste operator actions are required? (3.0) Approntmatelv how long after the alarn. receipt uill the valves be fully closeo? (0.5)
(***** END OF CATEGORY 07 *****)
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. ADMINISTRATIVE FROCEDURES, CONDITIONS, AND LIhTTATIONS PAGE 10 t ----------------------------------------------------------
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I QUESTION 6.01- (1 00) ) g [g fg a[
On No em .r 30, 1985 Indian Point 3 tripped during a startup from high a
, steam ierator water leve LER-011-00 was issued and committed to l chang i- a procedure to prevent r e oc c u r e n'c e . What was the cause of this t
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QUESTION 8.02 (2.50)
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' Who makes the initial determination if a Significa'nt Occurrence Report
- (SOR) should be filled out and when must a SOR-be filled out? (1.0)
' What are four responsibilities of the Shift Supervisor following a Significant. Occurrence? (1.0)
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00ESTION 8.03 (2.00) i i It equipment is removeo from service that results in a load reduction, who should be notified? List two. (1.0) -
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! Who must the Shift Supervisor notify after any reactor. trip? List j three. (1.0)
QUESTION 8.04 (2.00) In accordance with Technical Specifications, who nas administrative l control over hevs to doors wnich allow entry in to a radiation field greater than one tem per hour? (0.8)
( List f our responsibilities of the on-coming Shif t Supervisor DURING ,
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sh2ft turnover? (1.2)
i OUESTION 8.05 (2.50)
' What are the purposes of the 'Do Not Operate' tags? (1.0)
' Who must authorire a jumper if it is to be installed to defeat a trip or nuclear safety circuit in a non-emergency situation and i s not in an approved procedure? (1 0)
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' Who must authorize a jumper if it is specified in an approved procedure? (0.5)
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. ADnINISTRATIVE Ph0CEDURES, CONDITIONS, AND LINITATIONS FACE 11
00E511DN 5.0c (2.00) If the specific activity of the RCS is greater than the Tech Spec l nists, whv is the reactor brought to hot standby with Tave less than 500 F? (1.0) If failed fuel is indicated and RCS activity is below Tech Spec Limits, why are large rapid load changes to be avoided? (1.0)
GUESTION 8.0, s2.00)
The plant is at 100% power and Control Bank D is at 220 step A mal-function occurs in the rod control system which will not allow rods to move in or out in either n,a n u a l or aut Is the control rod system inoperable? Explai (1.0i Per proceoure OrJ0P-RC-2, Control Rod Svstem nalfunction, what two T e c h r. l e a l Specifications l i n,1 t c m u s t be checked during this tin.e?
L1.0)
00ESTIOrJ 5.05 (2.80) In accordance with SOP-CB-2, Containment Entrv and Egress, what are the maximum ano n.i ni mum size of groups allowed in containment while the reactor is at power) (0.6) For Technical Specifications, when does containment integrity exist?
(2.2)
QUESTION S.07 (2.20)
Discuss the relationship t't.4een Limiting Conditions for Operations, Liniiting Safetv Svsten. Settings, and Safety Limits in terms of preventing release of radioactivitv to the environmen (***** CATEGORY OB CONTINUED ON NEXT PAGE *****)
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. ADnINISTRATIVE PROCEDURES. CONDITIONS, AND LInITATIONS PAGE 12
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GUESTION 8.10 13.001 In acecordance with AP-3, ' Procedure Preparation, Review and Approval',
temporary changes mav oe made to plant procedures governed bv Techn2cel Specifications if four criteria are me Provide these four criteri QUESTION 6.11 (3.00)
Assume that it is 0300 on 2-19-66 and the reactor is presently at 45%
powe Considering the Delta-I penalty history listed below, when may power be increased above 50% power? Justifv your answe Date Time (ou T i n.e ( i rd Power (%)
2-18-96 0300 0318 85 2-16-Es 1557 1633 65 2-10-86 0135 0300 45 (***** END OF CATEGORY 08 *****'s (************* END OF EXAMINATION ***************)
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EQUATION SHEET
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Where mi = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2
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KE = mv2 PE = mgh PEi +KEi +P1Vi = PE +KE 2 +P 7 Y22 where V = specific li volume
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i P = Pressure
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Q = mcp (Tout-Tin) Q = UA (T ave -Tstm) Q = m(hi -h2 I
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P = Pg10(SUR)(t) p . p oe t/T SUR = 26.06 T = (B-p)t T p
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delta K = (Keff-1) CR1 II~Keff1) = CR 2 (l'Xeff2) CR = S/(1-Keff)
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M = (1-Keffi) SDM = (1-Keff) x 100%
I1'Aeff2) K eff
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decay constant = In (2) = 0.693 A 1 = An e-(decay constant)x(t)
t t 1/2 1/2
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Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 10 3 Btu /hr
Density =62.4lbg/ft 1 MW = 3.41 x 106 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 1 Atm = 11.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec2
....... _......______...... ___........ __.......__..__ _.. __ ....______
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< THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, AND PAGE 13
_______------_-----__----------_----..----------
THERh0 DYNAMICS
-____---_-----
.
ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, '
ANSWER 5.01 (1.50) Beta-bar-Effective will decrease over core life (0.5) because of the increased amount of plutonium in the core (0.5). Reactor period will increase-(0.5)
-REFERENCE Rx Th ch. 4, pgs. 14-19, 24, 25
_-__-_---------_--___--_------__-------------------------------------------
3.1 001 000 K 5 47 .
ANSWER 5.02 (2.00)
6 rno = 2.5% delta-l./L Keff = 1/(1-rho) = 1/(1-( .025)) = 0.9756 (0.5)
CR1/CR2 = (1-Keff2)/(1-Keff1)
j 115/345 = 1/3 = (1-Keff2)/(1 .9756) (0.5)
j Keff2 = 0.992 Reactivity-added = (Keff2-Keff1)/(Keff1)(Keff2) = 0.0169 delta-K/K (0.5) hore neutron generations will be involved with each.new neutron leve (0.5)
REFERENCE Reactor Theorv (Ry Th) c , pgs. 61-63, 57
'
_ --------_---------
3.1 004 000 K 5.08 ANSWER 5.03 (2.00)
As the fluid flows around the bend in the elbow, its velocity increases (0.5), the centrifugal force on the outside of the radius will increase (0.5) and result in a pressure increase on the outside~ radius and.a de-crease in pressure on the inside radius (0.5). -The flow rate is then pro-portional to the square root of the pressure difference (0.5).
REFERENCE ~ .
Thermodynamics (Thermo) ch. 5, pas. '13-15
_ _ - _ _ _ - _ _ _ _ _ - - - - _ - - _ - - - - - _ _ _ _ _ _ - _ _ _ _ - - _ - _ _ - _ _ - - - _ - _ - - - - - - _ _ _ _ _ - - - - _ - - _ - - - - _
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OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND FAGE 14
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ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWEh 5.04 (2.50> Reduced broadening cf - ::cr r.c c pcchc per unit temperature change at higher temperatures (0.5)
S ccdcr :i i c a; 3 tcnd to overlap at energy betmeen 7 est e s v.5m a1 224e.fSelf shielding r eduefwn:-r" tier e x F c ;:.: r - to f- cl ,0.5l- du M * > db $ 5 W ',
Lewsr e< .m a st. a baory b.s e rvsa -Jes k.m of hp len- nos &J fG 5) Doppler > Temperature Loefficient der e- in ma nitude over core life (0.5). CI:d c :- r e ; derren:c thE Yr N e_;cl d 3:p 012: t hr: i r.c r e c ;;;;g
+s e ; g. t s_ ;; 7 .g . . g : i t ., 9 , g y , 4, ,, h c, 4 4,,, fg ,f 42%
REFERENCE * "'d h IC' M Rx Th c , pga. 45-53, 58
- - - _ - - - _ - - - - - - - - - - - - - _ - - _ _ _ _ - - - _ _ _ _ _ - - - - - - _ - - _ - _ - _ _ _ - - _ _ _ - - - _ - - - _ _ - - _ - - - - - _
3.1 001 000 l, 5.49 ANSWER 5.05 (2.50)
As concenser cirevlating water flow increases, condenser vacuum will in-crease and draw more steam causing Pstm and Tstm to decrease (1.0).
As Tstm decreases, the delta-T across the steam generators increases which causes more heat. to be transferred from the primary system (0.5).
As more heat is transferred, toe specific volume of the RCS cecreases (0.5) which will cause pressurz er level to decrease (0.5).
REFERENCE Thermo ch. 9 pgs. 39-42, 46
---_--_--------_--__-_-_---___-------_--___________________--___-______-___
3.2 002 020 h 5.0B ANSWER 5.06 (2.00) ninimum NPSH is that required to prevent cavitation (0.5)
Available NPSH is that which is actually present (0.5) When minimum NPSH is greater than available NPSH, cavitation occurs (1.0).
REFERENCE Thermo c , pgs. 36, 39
-__________-__-____________________________________________________________
App pg. A-9 .
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. THEORY OF NUCLEAR POWEF PLANT OPERATION, FLUIDS, AND PAGE 15
____ _ _ ____ ______________________________________
______________
ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWER 5.07 (2.00) Excessive cladding temperatures could result because of the onset of de-Parture from nucleate boiling and the resultant reduction in the heat transfer coefficient. (1.0) To ensure DNBR is no less than the designed DNBR value (0,5)
To ensure the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid (0.5)
REFERENCE Technical Specifications (TSr pgs 2.1-1, 2.1-2
_-___________________________________________________._____________________
3.9 012 000 K 5.01 ANSWER 5.08 (2.50) When the c.odorator temperature increases it becomes less dense and moderates fever neutrons but when the moderator expands Doron is re-moved from the core and thus causes nTC to be positive. (1.5) MTC becomes more negative at EOL (0.5) because with reduced boron con-centration the expansion of the moderator is the dominate negative reactivitv effect (0,5).
REFERENCE Rx Th ch. 5, pgs. 24-27 31-33
____-_____ ---_____________________________________________________________
3.1 001 000 h 5 . 2 .6 ANSWER 5.09 (2.00)
O == U1 A1 (Tavs - Tstml) = U2 A2 (Tavs - Tstm2) (0.5)
U1 U2 A2 = .95 A1 (Tavs - Tstml) = .95 (Tavg - Tstm2) (0.5)
(573.5 - 513.8) = .95 (573.5 - Tstm2)
573.5 - 59.7/.95 = Tstm = 510.6 F (0.5)
From steam tables Pstm = 748.5 psia (0.5)
REFERENCE Thermo ch. 8, pg. 42
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. THEORT OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE le
---_ - -_ --- - _--------------__--__----_--_----_-_-_
_----------_--
ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, .5 041 020 K 5.02 ANSWER 5.10 (3.00)
Assume Xe equilibrium existed prior to an increase in neutron flux at the top of the cor The increased f.lun will be removing Xe in the top of the core but producing I-135 (Xe's parent), while in the bottom of the core, Xe is building in and I-135 production is less due to the flux shift (1.5).
But after a while, Xe builds in to the top of the core as it decays at the bottom of the core, thus causing flux to shift towards the bottom of the core and then the process repeats itself (1.5).
REFERENCE Ru Th c , pgs. 16, 17
___-_---_--------__-_-_-_-_------------_----------_-------_----_-_---------
3.1 001 000 L 5.38 ANSWER 5.11 (3.00) I
. It is the ratio of the mculmum local flux on the surface of a fuel rod at core elevstion Z to the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods (1.0). Cor.tv ol rods in a single bank move together with no individual rod insertion ciffering bv more than 15 inches from the bank demand posit-son (0.5>
Control rod banks are sequenced and overlapped (0.5)
Control bank 2nsertion l i nil t s are not violated (0.51 Antal power distribution is maintained within limits (0.5)
REFERENCE TS pgs. 3.10-8a,10,11
_----_------_--_____----______-_--_---_--_____-_--___--__-_-_____--___--_-_
3.9 015 020 K 5.05 l y + - +
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. PLANT SYSTEh5 DESIGN, C 0 rJ T R O L , AND INSTRUr1ErJT ATION PAGE 17
_----------------------------------
ANSWERS -- INDIAN P OIrJ T 3 -86/03/11-SILK, ANSWER 6.01 (1.50)
Al p the outlet of check valve 774A to the inlet of isolation valv/pingfrom 789 is high pressure pipe (.75) and isolation valve FCU 625 will
/clo pe n a high cooling water return flow (.75).
_----------------------------------------
3.3 000 009 EK 3.15 ANSWER 6.02 (1 00)
The combineo capacity of the three pressurl:er safety valves is greater than the m e::i m u m surge rate resulting froni a complete loss of load without a direct reactor tri (1.0)
REFERENCE TS 3.1-2, 3
_---_______--_____-_-__-_--------_---_--_-_-------__-_--___--
3.3 010 000 L 4.03 ANSWER 6.03 (1.50) Leak Detection Svsten, a l e r (0.5) Abnormal ractation levels within plant containment (0.5)
Dif icult detector 2nsertion (0.5)
Les de fens.,a .n a. u k, /
R L F E R ErJ C M7 N '
IP3 SD Incore Instrumentation, pg. 22
3.2 002 000 K 1.12 .2 002 000 K 4.05 ANSWER 6.04 (1.00)
These channels monitor the containment fan cooling water for radiation which is indie of a leak from the gontainment atmosphere into the coolingwate(gtive Ubring a LOCA '1.0',or 51wf det<< k s 'nJs ile Em Mfr,w /- (o,5),
'
4.0/
s) P. pang fr u s y pfy ek ck +o 10 e> Jury no/ 4 ,- h ij 4 p s u ~ <. .
M $.pply cbetk ua(vo pe ued ba.c4 Nina s) Fev-4Ar w s't / e fo.t e o>r /r<y'fr a kr n /4v n) PreHurt cefoi/ inslaHe/ e ,1 hyli pee.u u n p }ilyr fuy ko a t e.y ea c C
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. PLANT SYSTEns DESIGN, CONTROL, AND INSTRUhENTATION PAGE 16
-__--__------__-----_-_-_---_-----_---____----------_-
ANSWER 5 -- INDIAN POIN1 3 -86/03/11-SILK, D.
't REFERENCE IP3 SD Radiation Monitoring and Protection System, pgs.-27, 7, 8
, _----________--------_-------__------_---------_---------_--__--_----------
3.10-008 000 K 1.03 ANSWER 6.05 (2.50)
' Reduction in trip setpoints'for OTdT and OPdT for affected loor Actuation of auto withdrawal and insertion rod stop Charging pump speed will increase
, Steam dump circuit will have an initiating signal-
Lower insertion limit will be. calculated (0.3 each) Tripping of the 'A' channels high steam line flow SI bistable (0.4)
j Rod insertion will be initiated (0.3)
i Steam dump circustrv will be sent an actuation signal (0.3)
"
REFERkhE h"9 "7 b#
ONOP-RPC-1 pgs 6 and .27, Fgare AE-s s
___-------__-_----_---_-___ -----___----_ ---_________-_____-_---
3.9 016 000 L 4.05 j ANSWER 6.06 (2.001 ^ct:' tie ^ :# : c'e'r detecte* fill c 2:e c r- E r l '. " r :i ra L i f t te
! ill :nct' '^ -' Two Heat Actuating Devices have to sen e-c high temperature in the area and an alarm will sound-locallv - * * A two minute predischarge time must be completed before actuation of the carbon dioxide % (e.5)
' Holding tne deadman switch (0.5)
Abort / automatic toggle switch (0.5)
i REFERENCE IP3 SD Fire Protection, p g . 44- N o -41
--_------_----___-_____--___----__-__--_____-_--_---___________--__---___-_
3.11 086 000 K 4.06 3.3 I
-
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-v.v-m. , , . , , ,.e-, , , ,e,e- ,,e,- -y,
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. PLANT SYSTEMS DESIGN, CONTFOL, AND INSTRUnENTATION PAGE 19
______________________________________________________
ANSWERS -- INDIAN POINT 3 -86/03/11-SILL, ANSWER 6.07 (2.40) Fail open Goes to full spee6 F a i l c lo se d pc v4ts A Q, p c v- u ca r_g, , ps y- n o 8/no dy Fail closed Remain open (due to the accumulator for each valve actuator) Remain functional (will be serviced by their standby nitrogen supply)
(open) (0.4 each)
REFERENCE ONOP-IA-1 p , 3
___________________________________________________________________________
3.8 078 000 L 3.02 3 . ANSWER 6.08 (3.00)
480 volt buses isolated from 6.9KV buses bv tripping the breakers Bus supp1v breaker is trippeo whc- undervoltage condition l' ..toued 480 volt tie breakers are tripped Equipment on buses trip off due to undervoltage All auto-starting non-safegaurds equipment is locked out Diesels start then breakers close when buses are isolateo (0.5 each)
REFERENCE IP3 SD 10.0 pg. 31
___________________________________________________________________________
3.7 Oo2 000 k 3.01 K 3.02 h 4.02 J
- . _ - _ _ _ _ . _ _ _ _ _ _ _ . _ - _ _ _ - - . . . _ _ _ . _ _ _ _ _ _ _ _ _
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- PLANT SYSTEh5 DESIGN, CONTROL, AND INSTRUhENTATION PAGE 20
ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWER 6.09 (3.10) High Steam Line Flow - prevents excess positive reactivitv insertion (.25) caused by s_ steam line break downstream of the MSIV (0.5)
Hi-Hi Containment Pressure - Limits containment pressure by isolating containment from the steam header (.25)
in the event of a steam line break inside containment (0.5) The spray signal must be reset (0.4)
Phase 8 must be reset (0.4) Reset the reactor trip breakers (0.4)
Depress the SIS test buttons (on the logic panels) (0.4) ,
REFERENCE IP3 ED Engineered Safeguards, pgs. 11, 13, 26, 27
.
3.5 039 000 K 4.05 .6 003 000 K 4.06 .1 001 000 K 4.10 l ANSWER 6.10- (3.00s Overcompensation results in a lower indicated flux which could cause j the Source Range Detectors to be reinstated too early during a shut-f down (1.0i. thus causing a trip from source range high flux trip (0 5). This action will prevent an inadvertant turbine runbach vis the OPdT and the OTdT circuits when the OPdT and OTdT bistables are tripped for '
the affected channel (1.5).
REFERENCE IP3 SD NI, pg. 37 ONOP-NI-1, pg. 8
3.9 015 000 K 6.01 .9 015 000 K 1.05 f . lo . eg ?- b 14 rueb a.e hnc ( 8erh r, ep /s.e Y' $r'af f civr b$
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21
ANSWERS -- INDIAN POINT 3 -86/03/11-SILKr ANSWER 6.11 (4.00) OTdT OPdT Decrease Deeresse Decrease Decrease Increase No change (.33 each) . Indications of unoesireable flux cistribution will cause the set-points to be reduced (.66) An increase in Tave increases the heat capacity of the RCS and the rods moving in will push the flux lower in the core hence the re-duced setpoints (.66) OTdT - As Tavg cecreases the heat capacity of the RCS decreases thus allowing the setpoint to increase (.33)
OPdi - Setpoint cannot increase (.33)
REFERENCE IP3 SD 28.0 pg. 6 -11
3.9 01" 000 A 1.01 .
. . _ _ _ _ _ _ _ ._ .. -_ . _
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1 . PROCEDURES - NORnAL, ABNORMAL, EMERGENCY AND PAGE 22
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~~~~RE5i5E551CEL C5NTR5L
_----_---______-__-- ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, D.
4 .
t
.
ANSWER 7.01 (1.50)
!
) Notify NPO to proceed out to the EHT Panel and report the abnormal i condition (s) to the Con t r+5fp) ,2 5 Ao de r** ff g s' e to
$ w M 4 sitten.k P/s to. H)t r o l R o om , Yes ( . 4% ) Allowec by Tech Specs if daily checks of operability of redundant channel are made (0.5)
REFERENCE ONOP-EL-1, pgs. 2, 3
- 3.2 006.050 Sys Gen 1 4.0 i- Svs Gen 5
-
4.3
!.
'l ANSWER 7.02 (1.00)
These conditions prevent RCS pressure spikes which exceed the RCS t pressure - temperature relationship requirements of Tech Spec (1.0)
REFERENCE l SOF-RCS-1 pg. 1
- - - - - - - - - - - _ - - - - - - - - - - - - - - - _ _ - - - _ - - - _ _ _ - - - _ _ - _ - - - - - - _ - - _ _ - _ _ _ _ - _ _ - _ - _ _ _ - - _ _
i 3.4 003 000 0 6.14 :
-ANSWER 7.03 (2.00) Abnormal radiation levels in the av::iliary building (0.5)
/ * **g t.u M le jpgtog t,)
b . ,(,51 and RHR pump relief valves) ( 0,5 ) and the mechan 2cag penettation
-
area (0.5)
, RCS pressure increase (cr other indication of break isolation) (0.5)
REFERENCE ECA-1.2, pg , 3, 4
3.3 000 009 EK 3.21 .3 000 011 EK 3.12 i
!
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-, v- e y v%,- e e 4 e ,, e,ey--, - , er>- ,ee -em,y,eg<-y ,,r*-- y ..s-wt --g- = c -e -e-,,-ye,w---w ee,-, a ,p y-- -+tt--+W- r
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.
. PROCEDUREG - NORhAL, ABNORnAL, EhERGENCY AND PAGE 23
~~~~ ---------------------~~~
R Si5t55ICEL C5 SIR 5t
_-______-______-____
ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWER 7 04 (2.50) Job Supervisor Shift Supervisor Health Phvsics (.33 each.) , Affected organ Radioactive half-life Biological half-life Isotope (.25) Both R636 and R63E. (Gross Faileo Fuel Detectors) have increased on a fuel element failure (0.5)
REFERENCE Radiological Controls, ch. 7, pgs. 36, 37 ch. or pgs. 6-12 ONOP-RCS-4 pg. 2
___________________________________________________________________________
Plant-Wide Gen 15 ANSWER 7.05 (2.00) To ensure seal backpressure (0.5) and charging pun.p NPSH do not fall below minimum (0.5), If gross radiogas monitor R-14 becomes inoperable (.33), the I-131 monitor becomes inoperarie (.33), or plant vent activity exceeds set-point i.33).
REFERENCE SOP-CVCS-7. pg. 1 SOP-WDS-7, pgs. 1, 2
___________________________________________________________________________
3.1 004 000 K 1.21 .11 071 000 K 1.06 .
._ _ . _ _ _ _ . . . . . .
.
. PROCEDURES - NORnAL, ABNORMAL- EMERGENCY AND PAGE 24
~ ~~~~~~~~~~~~~~~~~~~~~~~~
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____________________
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ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, D.
.
I, '
ANSWER 7.06 (3.00s . Air Tempering Units are shutdown and their outlet dampers close
! Inlet and outlet dampers of the charcoal filter will open and bypass dampers will clo . FSB slidins door will close
- Station air will inflate the seals on all FSB perimeter doors ,
(0.5'each)
i Irradiated fuel is not being handled (0.5)
Neither the spent fuel cask nor the cask crane are moved over the spent fuel pit (0.5)
- REFERENCE
- SOF-V-2, pg. 31 TS 3.8-2
___________________________________________________________________________
i 3.11 034 000 K 6.02 33 5.11 034 000 Sys Gen 5 ; ANSWER 7.07 -(3.50) RCS pressure drops belou saturation conditions (.25) initiating local-1:ed boiling and water disassociation (.25) ,e ,ffe r . ,t%gy,gje ug io wJ
,
i . When the RCS is depressurl:ec, the noncondensable bubble.will expand and possibly reduce core cooling capacity (1.0)
{
l Depressurl e the RCS by pressurl:er sprays or burping' the pressurizer
small bubbles are carried away in the coolant 44,44. These bubbles i
are removed by the spray line to the pressuri=er vapor wm19 or through the CVCS letdown system to the UCT-++vi space *I'8 7)
Open the reactor head vents ++v44 removes noncondensables to the RC REFERENCE drain tank 4 ,44. Pgf ,&
.h h % e mW '6 kT w vM4Q 1 ONOP-ES-2, pg , 3
___________________________________________________________________________
3.4 000 074 EK 3.11 4.4-
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, PROCEDURES - NORnAL, ABNORMAL. EhCRGENCY AND PAGE 25
- - - - ~ ~ - - - - - - ~ ~ - - - ~ ~
~~~5d555L555CUL 55 sir 5L
____ __________-__ ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWER 7.08- (3.00) f e a 4 C b d (0.5 each)
REFERENCE Function Restoration Procedures, CSFST pg. 2,3
____________________ ____________________________-_____---_____-_-_-___----
System-wide and Plant-wide Generic K&A 11 ANSWER 7.09 (3.00)
Ensure both PORV's are shut (0.6)
Ensure both spray velves are shot. (0.6)
Ensure backup heaters are energized (if pressure < 2185 psis) (0.6) ;
Check av::ili ar y spray valve (212) shut (0.6)
Check either normal or alternate charging isolation valves (204 A/B) are open (0.6) sj ' / 5 o (c. A (Lh a r3,'nj ( le(d e REFERENCE hcc ep f- fr y ,% r f p ropp r- n sJ w )0 'M A A ' "
ONOP-RCS-2 pg. 2-3
___________________________________________________________________________
<
3.3 000 027 El; 3.03 4.1 a
ANSWER 7.10 (3.50) M; 2:+ Le r. . g s g g . g comtegy tre_igt3 +g mgim+3im Sn peig Secur e char ging pump (f.o)
Secure RHR pumps (s.sh n,___ ev, m :_ ___ .__
Er bhe ORV s prbeehky operate (l 8)
0 : .- '20/721 Oc ccen se thef clere '^ " ="$- /2 minutes (3-4-accepted) (0.5)
REFERENCE
- ~ b I ' r' 'O PDP-3.4 pg. 3
___________________________________________________________________________
3.3 010 000 K 4.03 4.1
!
)
- _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _-_
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. . ~ ADhINISTRATIVE PROCEDURES, CONDITIONS, AND LIhITATIONS PAGE 26 4 --_------------_---_--__--__-------..-------------------_.
i ANSWERS -- INDIAN FOINT 3 -86/03/11-SILK, D.
- l i
- r
_ ANSWER 8.01 (1.00) 3 dE A 6 7 6 b The ips result d from starting a second condensate pump at an inappro-priate t - 'c sing high steam generator water levels (1.0).
!
. REFERENCE -
PDP- LER 85 11-00
-_--___--_---_----__-_---_-_-__---__----___----_--_-__---_--_-------_
SWP GKA 1 .
I ANSWER 8.02 (2.50)
i
' The Shift Supervisor makes the determination (0.5) and a SOR must be -
filled out for all events determined to be reportable (0.5) and any-event which places the plant in-an-LCO (0.5). Assure the Teen Spec requirements are being rnmeaied with
- Assure the the applicable Department-Head is notified habe the necessary notification to the PRC Operations Center- L Complete a SOR prior to leaving the station- (.25 each)-
[ REFERENCE
]
AP-8 pg. 1
j System-Wide and Plant-Wide Generic K & A's'(SWPWGHA) 3 3.8 i ANSWER S.03 (2.00)
a Production Control Center and Operations Superintendent-(0.5.each)
- Operations Superintencent Shift Technical Advisor NRC (.33 each)
!
! REFERENCE AP-21. pg. 2 AP-21.2 pg. 1
--_____-__--_-_---_---___----__-_--__----____---__-_-____--__----_-_-___---
SWPWGKA'l !
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. ADnINISTRATIVE PROCEDURES, CONDITIONS. AND LInITATIONS PAGE 27
__-----________--___---------_----_-__----_------_---_----
ANSWERS -- INDIAN FOINT 3 -86/03/11-SILh, ANSWER 5.04 (2.00> Shift Supervisor ***4or Raolological and Environmental Superintendent (0.8) . Revieu Shift Supervisor's log for preceeding shift Review Night Order Book en.try for the previous and current working day Discuss any areas of concern with off going Shift Supervisor Sign the Shift and Relief Turnover Checklist in the control room (0.3 each)
REFERENCE TS pg. 6-19 AP-21.3 pg. 2
_________-_____-_____________-_--_--_--_________-_---_____-_-______-_____-_
SWPWGh6 17 .9 ANSWER 6.05 (2.50) For the control of jumpers in conjuction with the Jumper Log (0.5)
For tagging equipment, valves, control switches, etc, which are operable but may require additional instruction before the equipment is opeisted or its status changed (0.5). Operations Superintencent Superantendent of Power PORC (.33 eachi Shift Supervisor (0.5)
REFERENCE AP-10.1 pg. 5 AP-13 pg. 1-3
__________________________________-________________________________________
SWPWGKA 14 . - , _ _ . _ _ _ . . . . . . -_. - . . . . . . .
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. ADhINISTRATIVE PROCEDURES, CONDITIONS, AND LIriIT ATIONS PAGE 28
_------------------_----------------------------------
ANSWERS -- INDIAN FOINT 3 -86/03/11-SILK, ANSWER 8.06 (2.00)
! Saturation pressure of the RCS is below the lift pressure of the ,
atmospheric steam reliefs ( 0 /7 ) if a steam generator tube rupture should occur ( 0 .) ) .
- Avoid possible iodine following changes in thermal power spikin.gp[henomenonwhichmayoccur (1.0). T/ree me/ # 7h-c.rJ -[e e p a r /11 / c ed de i REFERENCE TS pg. 3.1-16 TS pg. 3.1-15
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3.11 000 076 Sys Gen 5 ANSWER S.07 (2.00)
, The h nt'ol rods are operable provided they can be tripped (.33) and a no rod misalignments (.33) and the drop time requirements
[ there are me ). Power Distribution Limits (0.5)
Quadrant Power Tilt Limits (0.5)
REFERENCE 1 TS 3 10-7 ,
} ONOP-RC-2 pg. 3
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- 3.1 001 050 Svs cer, 5 I i
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T. 0 7. x '
jn km a ,n o ee ru bic (o s ) b ee a.1 e ;+ cun/
wtrol ,,J f v [*'m sYJ m Nendrel p ary Arc (b.7)
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. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIhITATIONS PAGE 29
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ANSWERS -- INDIAN POINT 3 -86/03/11-SILK, ANSWER 8.08 (2.80) and 2 (0.3 each)
{ All penetrations required to be closed during accident conditions are 1 either i Capable of being closed by.an operable containment auto-isolation valve system (0.5), or ,
Closed by manual valves, or blind flanges (0.5)
i Equipment door is properly closed (0.5)
Both doors in each personnel air lock are properly closed unless being used at which time at least one air lock door shall be closed (0.7)
REFERENCE SOP-CB-2 pg. 2 TS 1-4
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SWPWG KEA 5 ANSWER 8.09 (2.20)
t LCO's indicate lowest performance level of equipment required for safe operation of the f acility (0.7). If improper automatic action occurs
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prior to reaching Limiting Safety System Settings, then Safety Limits will not be exceeded (0.S>. If Safety Limits are not exceeded then. fuel and RCS integrity will be maintained (0.7).
REFERENCE 10 CFP 50.3o .__________-_________-___. ...______________________________________________
- SWPWG KEA 5 3.9 I ANSWER 8.10 (3.00)
The intent of the original procedure is not altered (.75).
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The change is approved by two members of the plant staff, at least one of whom holds a SRO license on Indian Point 3 (.75)
If the procedure is an AP, then OA approval is also required (.75)
The change is documer.ted, reviewed by PORC and approved by the Resident Manager within 14 days of implementation (.75)
REFERENCE AP-3, pgs. 6-7
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TS P3 6-13a
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s ADMINISTRATIVE PROCEDURE 5, CONDITIONS. AND LIMITATIONS PAGE 30
ANSWERS -- INDIAN POINT 3 -86/03/11-5 ILK. SWPWGKA 5 ANSWER 8.11 (3.00)
2/19 0300 - 0138 = 8 2 > 41 penalty minutes (0.5)
60 - 41 = 19 minutes available (0.5)
2/18 1633 - 19 = 1614 t'0.5)
With no further operation outside the Delta-I band (0.5), the penalty would be reduced to 60 minutes at 1614 on 2-19-86 and power could then be increased (1.0)
REFERENCE TS 3.10. TS 3.10. .1 001 050 Sys Gen 5 J
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ATTACHEMENT 2 FACILITY COMMENTS ON SRO WRITTEN EXAMINATION & NRC RESOLUTION
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Question: 5.03 Comment: Is it required to discuss each related aspect of the created Delta-P?
Might shift point distribution to;
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1 pt create Delta-P 1 pt Delta-P proportional to flow Resolution: Not accepted, question specifically asked for a description of how-the differential pressure was develope Question: 5.0 Comment: Answer should be: Reduced broadening per unit temperature change at higher temperatur . Rate of self-shielding reduction decreases at higher temperature . Low resonance absorbtion cross-section at higher temperatur Resolution: Accepte Question: 5.0 Comment: Answer Incorrect! FTC More negative at lower temperatur Answer should be: Doppler Temperature Coefficient (DTC _or FTC)
becomes more negative over core life. Three (3) effects:
U-238 depletion (minor) Less neg Pu-240 buildup (major) More neg Lower fuel temperature (minor) More neg Overriding effect, Pu-240 buildup makes DTC more negative.
, Units on FTC (DTC) PCM/ F Resolution: Answer corrected.
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Question: 5.05 i
- Comment
- Should also accept an increase in condensate depression causing j a decrease in overall plant efficiency. Reactor must put out '
more power to maintain turbine load, more heat transfer.to S/G therefore RCS specific volume will decrease and pressurizer i level will decrease.
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Resolution: Will be considered in grading.
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Question: 5.06.b
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Comment: Consider an additional answer to the key:
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Cavitation occurs at Tsat Resolution: Will be considered in grading.
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- 5.0 Comment: Very misleading. Could have asked why or how MTC could be positive at high boron concentration Resolution: Subjective Question: 6.01 Comment: There are four (4) specific design features that protect the CCW system;
- Piping from supply check to 789 is designed for high
, pressure.
4 Supply check valves prevent back flow of high pressure to low pressure pipin . FCV-625 will close on high return flow.' Pressure relief installed on high pressure pipin Since question asked for two (2), should accept any two (2) of the abov Resolution: Answer changed.
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Question: 6.0 Comment: Consider an additional answer to the key: Leak Detection indicator R-7 Seal Table Radiation Monitor I R-2 80 Foot Airlock Radiation Monitor Resolution: -Added number 1 to the key, numbers 2 and 3 are considered the same as abnormal radioactive levels.
Question: 6.04 Comment: Is it required to reference a LOCA specifically?
] Resolution: Answer changed to allow for "LOCA or Steam break".
! Question: 6.05.a i
Comment: Consider an addition to the key:
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Pressurized level-increase and an increase in the setpoin Consider a reduction-in the OP AT and 0 TAT as separate-item Resolution: Pressurizer level increase is synonymous with changing pump speed increase +
The changes in the ' protective setpoints are considered one ite Question: 6.0 Comment: Delete an answer-from the key:
Smoke Detector is a warning and not~an actuation devic Resolution: Deleted-3 Question: 6.07.c
Comment: Consider an addition to the Key:
The V.C.T. has several make-up valves:
,. FCV-111A PW to Blender
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-tec-- p +--
-*wf -9 .--g epqp
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4 FCV-110A B.A. to Blender FCV-110B & FCV-111B VCT Make-up valves The question does not name the valve but lists the numbe It also is in the plural for . FCV-111A fails close . FCV-110A fails ope . FCV-110B & FCV-1118 fail close Resolution: Considered in gradin Question: 6.0 Comment: Delete explanation of accumulator as no explanation was asked for in the questio The key should state simply "open".
Resolution: Information contained within parenthesis on the answer key is not required for full credi Question: 6.08 Comment: Consider additions to the key: MCC 36a, b, c breakers remain close If question is interpreted as a Blackout condition and subsequent S.I. initiation: stripping of the buses except MCC 36a, b, c and reloading of S.I. equipment except Component Cooling Water Pum Resolution: Considered in grading, based on assumptions stated by candidat Question: 6.0 Comment: Revise the answer / key:
One action only Reactor Trip Breakers or SIS Test Buttons Resolution: Not accepted, question specifically asked for two action *
Question: 6.1 Comment: Consider an addition to the key:
The key addresses a plant shutdown condition. Consideration should be given to a start-up condition, i.e. P-6 is not achieved; therefore, the plant trips on a high flux-source rang Resolution: Answer added to ke Question: 7.0 Comment: Accept SR0 directing NPO to swap to alternate power suppl Resolution: Modified key to make this part of the required answe Question: 7.0 Comment: Accept Yes! T.S. allows this condition for 7 day Student indicates reference to T.S. which will detail required action such as daily check Answer to b. appears in T.S. and 0 WOP but as subsequent actio Resolution: Considered in gradin Question: 7.03.b j Comment: " Mechanical penetration area" should be changed to " piping penetration area".
Resolution: Key change Question: 7.0 Comment: While the answer is right out of the procedure, the candidates know that the relief valves for RHR and SIP's are located inside Resolution: Key modified; training department has notified the operations department'to initiate a correction to procedur . . . _. .. .. .. ,-. -
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Question: 7.0 Comment: There are more ways to get a non-condensible bubble. Should also accept; H2 production from Zr-H 2O reaction
, N2 injection from accumulators Gases coming out of solution Consider accepting these als Resolution: Considered in gradin Question: 7.0 Comment: Consider accepting; Per steam space sample
, Start RCP to flush gases up to pressurizer; then use PORV's, or sample lin As per emergency procedures, objective would be to flush steam bubble out of head with RCP's to establish core coolin Resolution: Answer modifie '
Question: 7.09
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Comment: PORV's Shut Sprays Shut HTR's On 4 Auxiliary Spray Isolated Normal / Alt Chg Open
, Isolate charging and letdow Check for instrument failure .
, Trip Rx if below trip setpoint I Resolution: Allowed for 5 or 6'above; allowed instrument failure and/or reactor trip if proper assumptions made. .
Question: 7.10.a
- Comment
- Student should not be required to memtrize all actions
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specified in POP's. Should give credit for discussion of concerns and general' action Resolution: . Madified answer for major action .
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Question: 7.10.b
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Comment: Should not be required to know time. They know it's not a quick closure. This should be sufficien Resolution: Considered in-gradin Question: 8.01
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- Comment: We do not require students to memorize LER's by number and/or dat j
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Resolution: Question deleted from examinatio >
l- Question: 8.0 i Comment: The contact at the PCC, the System Operator (or S.O.) should be i an acceptable answe ~
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Accepted.
] Question: 8.0 l Comment: Either answer acceptable?
i Resolution: Yes, key modified.
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Question: 8.04.b i
Comment: In practice this is done some time after turnove Resolution: Not accepte ..
Question: 8.0 .,
Comment: Ops. Supt, authorizes. (Supt. of Power concurs ar.d PORC reviews and approves safety evaluation)-
Resolution: Not accepted, all signatures _ are required prior to wor '
Question: 8.0 Comment: Iodine spiking results from thermalistressing of cla t Therefore, thermal stressing of clad should be an acceptable
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answe Resolution: Considered during gradin s
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Question: 8.0 Comment: The control rod system is inoperabl Resolution: Key modified.
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Question: 8.09 Comment: Our Technical Specifications do not define them in terms of preventing release. Obscure question.
Resolution: Not accepted. The question did not ask for a Technical i Specification definicion.
- Question
- 8.10
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Comment: The following should also be considered acceptable answers:
.- Procedure cannot be performed or completed as require Not to be used for convenience or to correct typo's.
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TPCN completed.
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Resolution: Not accepte Question: 8.11 Comment: Consider 1615 an acceptable -answer to ensure <60 min, out of i ban '
Resolution: Accepte ,
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