IR 05000322/1986022
| ML20213G515 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/07/1987 |
| From: | Collins S, Crescenzo F, Keller R, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20213G465 | List: |
| References | |
| 50-322-86-22OL, NUDOCS 8705180325 | |
| Download: ML20213G515 (93) | |
Text
{{#Wiki_filter:-sr . . U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 86-22 (0L) FACILITY DOCKET NO. 50-322 FACILITY LICENSE NO. NPF-36 LICENSEE: Long Island Lighting Company P. O. Box 618 Wading River, New York 11792 EXAMINATION DATES: December-19, 1986 CHIEF EXAMINER:
Mt/f/7 . Crescenzo, Re tor Engineer (Examiner) Date h 6/4/f7 REVIEWED BY: . Lange, Lead Reac r Engineer (Examiner) Date REVIEWED BY: Mt/ff7 R. Keller, Chief, Project Section 1C Date APPROVED BY: MM dNU 5'. J. ColTins, Deputy Director, DRP Dite SUMMARY: Operator licensing examinations were administered to three Senior Reactor Operator candidates, two Senior upgrade candidates, and five Reactor Operator candidates during the week of December 15, 1986. All candidates passed the written and oral examinations. Parallel grading of requalification examinations administered to the licensed staff was also conducted.
l 8705180325 870500 PDR ADOCK 050
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' REPORT DETAILS TYPE OF EXAMS: Upgrade (2), R0 (5), SR0 (3) { SR0 R0 EXAM RESULTS: Pass / fail Pass / Fail
Written Exam 5/0 5/0 ' Oral Exam 5/0 5/0 ____________________________________________ OVERALL 5/0 5/0 1.
. CHIEF EXAMINER AT SITE: F. J. Crescenzo, Reactor Engineer (Examiner) 2.
-0THER EXAMINERS: R. L. Turner, Reactor Engineer (Examiner) W. C. Cliff, (PNL) 3.
SUMMARY OF GENERIC. STRENGTHS OR DEFICIENCIES NOTED ON ORAL EXAMS: a.
No significant generic weaknesses were noted during the walkthroughs.
Generally, the candidates performed quite well and seemed very well prepared, b.
The candidates were proficient in use of the PID's and electrical logic diagrams.
c.
The Senior candidates were well trained on recognition of Technical Specification action statements and interpretation.
3.
SUMMARY OF GENERIC STRENGTHS OR DEFICIENCIES NOTED DURING GRADING OF WRITTEN EXAMS: ! The candidates performed well on all sections of the written exams. The candidates performed poorly on the following two questions in section 4 of the Reactor Operator exam.
Question # Topic Class Average 4.07 Operator cautions of 50% procedure SP 22.005.01 4.12 Operator cautions of 45% procedure SP 22.001.0 - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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EXIT INTERVIEW DETAILS Personnel Present at Exit Interview: NRC Personnel F. J. Crescenzo Facility Personnel J. L. Smith, Director, Office of Training J. A. Scalice, Operations Manager / Assistant Plant Manager L. J. Calone, Manager, Operator Training Division H. T. Carter, Operations Training Supervisor R. McNair, Licensed Operator Training J. Alexander, Operations Engineer 5.
Summary of NRC Comments made at Exit Interview a. The items noted in paragraph 3 above were discussed , b. The facility was informed that results of the examinations would not be available until after the grading of the written exam and after the in house QA efforts.
Every effort would be made to have the results within 30 days, c. The facility was informed-that that the exam review was efficient and well organized. Comments to the exams were pertinent and well docu-mented.
5.
' Summary of Facility Comments at Exit Interview: a. The facility stated that the exams were well prepared and thanked the examiners for their efforts, b. The facility invited the NRC to inspect the simulator once it is instal-led (planned installation in August).
6.
Resolutions to facility comments on written exams.
The " master answer copies" are marked and initialed to indicate resolution-of questions raised during the two hour exam review on December 15, and is one of the attachments to this report.
-The following represent only responses to those facility comments that were documented as an attachment to SNRC-1302 which was received December 18 and which did not include issues resolved during the exam review.
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. All of.the comments were considered during the grading of the exam. Those comments that required significant changes are listed below and addressed fully.
Facility Comments Question 2.04a - also accept absence of low dp annunciation for core spray piping integrity . Resolution - comment not accepted as non presence of dp annuciator does not provide indication that' the core spray system is operating Question 2.05 - also accept heatup of cond. pump motor bearings Resolution - comment not accepted as the suggested response is not a direct effect as requested in the question Question 2.06a - also accept for "yes" a rod out block with a third insert error Resolution - comment not accepted, no effect on grading Question-3.04a - also accept 603 four rod LPRM display, LPRM indicating lights or switch status Resolution - comment accepted except checklist alone not adequate response Question 3.07 - also accept LPRM prerequisite checklist (23.603.01) Resolution - comment accepted Question 3.10 - Also Accept: 1) Control rod reaches OUT limit 2) Control rod reaches IN limit 3) Wrong rod selected 4) Loss of RPIS 5) RSCS INOP Resolution - comment accepted to the extent that the suggested alter-nate nomenclature is equivalent to answer key statements.
Question 4.07 - states " Hot Shutdown" and answer key refers to " Hot Standby".
Resolution - comment accepted, will use for answer key a) Do not exceed the 100 degree F/hr croldown rate b) o cycle SRV's to close c) o start RHR in suppression pool cooling mode Question 4.09a,c-accept match a with 4,6 and c with-1,2 Resolution - comment accepted ., - - . . ._ - _ -.
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- Question 4.12b procedure does not caution against hydrogen explosion Resolution
- comment not accepted, all BWR's have potential for hydrogen explosion in mechanical vacuum pump immediately after shutdown from power.
Question 5.01c - Actual value in steam tables show approximately 3 F-4 F decrease.
Please also accept decrease.
1020 psia = 1192 BTU /lbm 100 psia = 1187.2 BTU /lbm Resolution Comment accepted graded accordingly Question 6.0lb - Change INCREASE to NO CHANGE due to no flow to charging (1)b(1) header during steady state operation.
Resolution - Comment accepted if steady state operation is declared.
Question 6.03 - MOV 35 (F006) close due to HOV-51 going closed on Hi level.
Should only require HOV-51.
Resolution - Comment accepted - only H0V-51 is required.
Note: SH 202 figure 2 MOV-33 appears to be mislabeled (should be MOV-35?) Question 6.07 - Do not take credit off if candidate states EDG will auto start due to an undervoltage condition, once stopped will auto start.
Resolution - Comment accepted.
Eliminate reference to RESET switch and change point values to +0.75 for last two.
Question 6.08 - (a) Also accept reduction in beta bar effective (E0C) and CR icw worth at full ou; position.
(b) Also accept reactor power greater than 30% as sensed by first stage shell pressure of 109 psig. Change high pressure EOC trip to 1120 psig.
Resolution - Both comments accepted. Add delta flow high 44 gpm and manual trips as acceptable answers.
Reference: Technical Specifications.
Question 7.01 - 1) Also accept: a) to get rid of a major heat load b) protect the RWCU pump seals 2) Also accept: a) to get rid of heat load b) to minimize vessel transiert when performing the manual scram Resolution - Comments accepted per given references.
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. Question 7.02 - May get more detailed information as per procedure.
Resolution - More detail, as long as it's correct, will never result q in point deductions.
Question 7.05 - Also accept RCIC auto initiation and isolations will be bypassed.
Resolution - It appears that this comment is intended to apply to question'7.04 and as such it is accepted (original answer allowed for such responses). Could not.varify this comment due to the lack of all odd numbered pages in reference material SP 23.133.01.
Question 7.06 - a) Also accept to insure MSIV closure will not occur at main steam line pressure of 825 psig when mode switch is in RUN.
b&c) Also accept to insure APRM downscale rod block or a scram on a APRM downscale with a companion IRM HI HI upscale.
Resolution - a) Not accepted. The question asks for the basis for the limitation which is the GEXL correlation limits (established during Atlas loop testing) not the safety action as a result of being outside the limit ~ envelope.
b&c) Accepted.
Question 8.01 - The reason for N0 is various, also accept 8.1.1.c.
Any of the Technical Specification action statements will.
address to the fact that you can not operate for 6 weeks.
Did not request the most limiting action statement 8.1.1.c.
Delete "all loads fed from DIV II emergency bus are considered INOP".
- Resolution - Comment accepted. Credit given for TS 8.1.1.c.
Question 8.02b - Also accept 3.6.3 Table 3.6.3-1 MSIV's (821*A0V81A, B, C & D and 821*A0V 82A, B, C & D).
Resolution - Section 3.6.3.a is also applicable and will be accepted along with its LCO. Also change 4.0.3 to 4.0.4 in first paragraph of b.
Question 8.08b - Also accept if the candidate breaks up the two pumps.
SBLC True CRD False Resolution - Comment accepte '. l . Question 8.09 - Delete last sentence, "Section 3.5.1 requires both CSS subsystems to be operable prior to entry into mode 1."
Question has already stated that CSS is inop with a 7 day LCO. Candidates would not need to repeat statement.
Note grading change.
Resolution - Full credit to be given if the candidate does not repeat this statement.
Question 8.10 - Also accept Technical Specifications 3.6.1.1 and.3.5.1.c.
Resolution - Will not deduct points if mentioned, but answer remains as is, and is required for full credit.
Question 8.11 - Also accept: New York State Suffolk County LERO Communications - Hot Line (red phone), commercial telephone Resolution - Comment accepted.
7.
RESULTS OF PARALLEL GRADING OF REQUALIFICATION EXAMINATIONS: Parallel grading of facility administered requalification written examinations was conducted by NRC examiners. Three Reactor Operator and four Senior Reactor operator examinations were graded by both.the facility and the NRC.
One Reactor Operator and one Senior Reactor Operator failed the examinations per facility grading.
These individuals were removed from licensed duties, must complete additional training and successfully pass another requalification examination prior to resuming licensed activities.
NRC results of the parallel grading were, in most cases consistent with those obtained by the facility. All discrepancies were investigated on a question by question basis. Most discrepancies were due to harsher grading by facility graders.
In some instances, the NRC examiner graded harder than the facility but these discrepancies were due to differing interpretations of both the candidate answers and the answer keys. None of the discrepancies were noted to be significant.
In addition to parallel grading, the examinations were reviewed for content adequacy. The examinations were found to be adequate from both an admin-istrative and technical viewpoint.
It was noted that many of the questions were objective in nature or required " lists" for answers. This was par-ticularly evident in those examination sections which dealt with plant systems and instrumentation (sections 2, 3, and 6).
These types of ques-tions are normally used on NRC administered examinations however, the frequency of such questions would be significantly less on an NRC examina-tion.
It is suggested that the facility consider reducing the percentage of such questions in preparation of future examinations.
In summary, the facility's.requalification written examination preparation and grading techniques appear to be reasonably consistent with current NRC procedures and as such are adequat.
. Attachments: 1.
Written Examination and Answer Key (RO) 2.
Written Examination and Answer Key (SRO) l-
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NUCLEAR REGULATORY COMMISSION .e REACTOR OPERATOR LICENSE EXAMINATION y.
, . . FACILITY: _SHgREHAM________________ ' ' REACTOR TYPE: _BWR-gg4_________________ DATE ADMINISTERED: _@6/12/1g________________ EXAMINER: _ TURNER g._______ _ ___ i CANDIDATE: __ M 5I______________ ___ INSIBUCIIQNS_IQ_C8NDID81El Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%. . Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY __Y869E_ _IQI@b .___SQQBE___ _y860g__ __.,___________CBIEQQBY_____________ _2Es99__ _2EsgQ ___________ ________ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW . _25z99__ _2Ez99 ___________ ________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS _2Ez99__ _2Ez99 _______ -__ ________ 3.
INSTRUMENTS AND CONTROLS _2Ez99__ _25299 __ -_____ _ _ _ __4.
PROCEDURES.- NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 199z99__ __ _____ Total s Final Grade All work done on this examination is my own.
I have neither given nor received aid.
_ __ . - _ - -_ . - - - _ _ _ - Candidate's Signature J
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS , Daring th2 administration of this examination the following rules apply 1.. Ch' eating on the examination means an automatic denial of your application
and could result in more severe penalties.
- 2.
Rest room trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil gnly to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of gach section of the answer sheet.
B.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write gnly gn gng sidg of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least thrgg lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility litgtatutg.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of l the gxamingt only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in-completing the examination.
This must be done after the examination has been completed.
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, 'S 10. When you complete your. examination, you'shall - a. ' Assembleryour. examination as follows: , . (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b.- Turn in your copy of;the examination and all pages used to answer the examination questions.
c.. Turn in all scrap paper and the balance of the paper that'you did-not use for answering the questions.
d.
Leave.the examination area, as defined by the examiner.
If after l eavi ng, you are found in this area while the examination is still in' progress, your license may be-denied or revoke __- 'ig__PRINGIELES OF NUCLEAR POWER PLANT OPERATION PAGE
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. - - . DUESTION 1.01 (2.00) Assume that your reactor is-being made critical by periodically withdrawing equal reactivity value control rod increments, how does: ^ a) the time to reach steady state change as keff approaches unity? (1.00) b) the neutron level change for each rod increment as keff approaches unity? (1.00) DUESTTON 1.02 (1.50) What are the three major functions of the control rods as they apply to each operational mode? (1.50) I DUESTION 1.03 (2.00) Name the three primary modes of heat transfer and give an example of each common to a nuclear power plant.
(1.50) QUESTION 1.04 (2.00) With respect to a fuel element in an operating reactor, define what i s meant by " onset of transition boiling".
(2.00) DUESTION 1.05 (3.00) What are the three " thermal limits" observed for the operation of the SNPS Reactor? What is the basis for each? (3.00) (***** CATEGORY 01 CONTINUED ON NEXT PAGE
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IHE80QQYN6dlCSz_HE61_IB6NSEg8_@ND_E6UID_E6QW . ,, . QUESTION .1.06 (2.50) Due to a computer software malfunction, a heat balance has to be performed manually on your shift. Given.the following data, arrange the terms into an equation by which the reactor thermal power could be determined. (CALCULATION NOT REQUIRED but do (2.50) write the terms (i. e (MFW x HFW)) so that the calculation could be performed.)
6,400,000 lbm/hr Given: MFW (Feedwater flow) = MCU (cleanup demineralizer flow) 110,000 lbm/hr = MG (steam flow) = 6,423,000 lbm/hr MCRD (control rod drive flow to reactor) = 23,000 lbm/hr HFW (Enthalpy of feedwater) = 345 BTU /lbm HCU,in (enthalpy of cleanup flow in ) 506 BTU /lbm = HCU, out (enthalpy of cleanup flow out) = 419 BTU /lbm 1194 BTU /lbm MG (enthalpy of steam ) = Hcrd (enthalpy of CRD flow to reactor) = 68 BTU /lbm Op (input energy from the recirc. pumps) = 26,500,000 BTU /hr Ofl (heat losses f rom the nuclear boiler) = 2,040,000 BTU /hr DUESTION 1.07 (2.00) In a BWR, reactor power nust be changed prior to changing turbine power. Assuming no reacter act i on, explain the initial effect on reactor power and pressure if: a) turbine power was reduced by closing the turbine control valves.
(1.00) b) turbine power is increased by opening the turbine control valves.
(1.00)
QUESTION 1.08 (2.00) During startup you have established a stable 65 second period and have declared the reactor critical. By definition, is the . reactor critical, sub-critical or super-critical? Explain the difference between the three states.
(2.00) (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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, - ,. QUESTION 1.09 (1.50) a) Explain what function the delayed neutron fraction, Beta, has in reactor control.
(1.00) b) _1f a reactor is at prompt critical, what is the difference betweer the value of Beta and the excess reactivity added? (0.50)
- REFERENCE SNPS Reactor Physics Module - Lesson 15 292003 kl.07, 3.3
- QUESTION 1.10 (1.00) The suction pressure of a centrifugal' pump is 50 psig and the entering.
water temperature is 225 degrees Fahrenheit: a) What is the Net-Positive Suction Head? (0.50) b) Is this a cavitating condition? (0.50) OUESTION 1.11 (2.00) , The reactor has been operating at 100% power for one month when a scram occurs in which several control rods FAIL TO FULLY INSERT.
Enough rods DO insert to bring the reactor subtritical at the time of scram.
If reactor moderator temperature is maintained CONSTANT, j and control rods are NOT moved, about HOW LONG will the operator [ have to wait before he can be reasonably sure that the reactor will remain subcritical? EXPLAIN.
(2.00)
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. 9tSTION 1.12 (2.OO)e TRUE OR FALSE: For a CONSTANT reactor period, it takes the SAME AMOUNT OF TIME to change reactor power from 1% to 5% as it does to change it from '10% to 50%. EXPLAIN YOUR ANSWER FULLY.
(2.00)
OUESTION 1.13 (1.50) -ANSWER THE FOLLOWING THREE QUESTIONS REGARDING CONTROL ROD EFFECTS ON CORE POWER WITH TRUE OR FALSE.
n a.
Withdrawal of a " deep control rod" generally has an appreciable-effect on total core power output because the power increase is spread throughout the core by the relatively high void content in the area of withdrawal.
(0.50) b.
Withdrawal of a " shallow centrol rod" generally changes the power shape, while affecting total core power - very little, . because the relatively large local power increase is off-set tnr an increase in void centent.
(0.50) c.
The " reverse power effect" or " reverse reactivity effect" occasionally observed when a shallow control rod is withdrawn.
one or two notches is due to a relatively large local power increase being off-set by a moderator temperature related power decrease.
(0.50)
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. DUESTION 2.01 (3.00) a.
What are the normal values for CRD HYDRAULIC SYSTEM FLOW and DRIVE WATER DIFFERENTIAL PRESSURE? (1.00) b.
Approximately what percentage of the flow in "a" above is supplied to the cooling water header? (1.00) Explain HOW/WHY requesting single rod insertion causes cooling at.
heeder flow to vary (include by how much the flow varies). (1.00) DUESTION 2.02 (3.00) For each of the following statements regarding the High Pressure Coolant Injection System (HPCI), indicate whether the statement is.TRUE or FALSE, and EXPLAIN your answer.
a.
In the event Low Pump Suction Pressure i s sensed during HPCI system operation, the turbine will trip, and the signal must be manually reset before the turbine will restart, if initiation.
signals are still present.
(1.00) b.
Upon a HPCI system isol ation, due to Low Steam Pressure, the system cannot restart until the pressure rises above the isolation setpoint and the isolation signal is reset.
(1.00) c.
If the HPCI turbine trips due to an overspeed condition, it will restart when the speed coasts down to between 3000 and 4000 RPM.
(1.00) QUESTION 2.03 (3.00) a) The RCIC system is in standby when the minimum flow valve (MOV-036) inadvertently opens. What would be the consequences j of this situation if it went unnoticed by the operators? (1.00) b) Why should operation of the RCIC turbine at speeds of less than 2200 rpm. be avoided ? (two reasons required) (2.00) , (***** CATEGORY O2 CONTINUED ON NEXT PAGE
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. OUESTION 2.04- '(2.00) In the event'of a large break in the reactor primary' system piping: a) what control room indication-is available to provide. assurance that the core spray system is able to provide adequate cooling .to the reactor core? (1.00) ~ b) What is considered adequate cooling with the core spray. system? ( 1'. 00 ) DUESTION 2.05 (1.50) What effect would a loss of TBCLCW have on the Condensate System? (1.50) (consider direct effects only) DUESTION 2.06 (3.00)
- During startup,the Rod Worth Mi nim 2 :er operator panel shows that control rods 22-37 and 26-41 are in " insert error" status, a) Does a Rod Black exist or not?
(1.00) b) Why or Why not? (1.00) c): What panel indication should tell you if a rod block is in force.
(1.00) QUESTION 2.07 (3.00) The steam flow signal to the FWLCS fails to zero during reactor operation. What is the immediate effect (increase, decrease, no change) on: a) the reactor level? (0.75) b) the reactor pressure? (0.75) c) the reactor power? (0.75) d) feedwater flow (0.75) . (***** CATEGORY O2 CONTINUED ON NEXT PAGE *****) . . __. - -, _ _, _ - _. - - - - -.. -. - ,.... - -.. - - - -.. - - -,,,,, _ - -
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. QUESTION .2.03-(3.00) What are the three conditions.that must be -present to permi t the recirculation flow control 30 % speed limiter to be bypassed AND what is the reason for each? (3.00)- QUESTION 2.09 (1.50) ~ Assume a Diesel Generator has started as a result of a high drywell pressure. What locations and what MANUAL controls or switches are available to stop the Diesel Generator.
(1.50) -
DUESTION 2.10 (2.00) What four conditions have to exist for the automatic ? depressurization system (ADS) valves to open.
(2.00).
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OUESTION 3.01 '(2.50) The following plant conditions exists ' - Reactor Vessel Level is -140" (confirmed, both channels) and decreasing.
- The RHR System is operating with a pump discharge pressure of >125 psig in system A and B.
- Reactor Vessel Pressure is 1000 psig and steady.
- All ADS INIT INHIBIT switches in NORMAL position.
- ADS Timers are at 110 seconds (both A & B).
- On the Automatic; Blowdown Relay Panels ( 1H11*PNL-628 and 1H11*PNL-631), the white indicating lights above each of the SRVs show - ONE light DUT and ONE light LIT BRIGHTLY, for all valves.
(1.5) a.
What.has happened ? EXPLAIN.
. (1.0) , b.
What are your actions in this situation? DUESTION 3.02 (2. 50) ' For the f ollowing combination of Radiation Monitoring System (RM5) control room annunciators, which have just been' activated; -RMS RAD MONITOR ALERT-REFUEL LEVEL VENT EXH RAD ALERT-REFUEL LEVEL DIV I VENT RAD HI-RMS RAD MONITOR RAD HI-FUEL STOR POOL RAD HI-REFUEL LVL DIV II VENT RAD HI State whether or not each of the following events should have occurred.
a.
Reactor Scram (0.5) b.
NSSSS Isolation (0.5) c.
RBSUS Initiation (0.5) d.
Liquid radwaste discharge isolation valve closure (1G11-ADV-158) (0.5) e.-CRAC. isolation in the emergency mode (0.5) (***** CATEGORY 03 CONTINUED ON NEXT PAGE
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. QUESTION-3.03 (3.00) ' Consider the RECIRCULATION FLOW CONTROL system. 'The following conditions exist:. 1.
The MASTER flow contro11er'has failed full scale HIGH.
2.
Both recire pumps being controlled on their MANUAL / AUTO transfer stations, in-MANUAL at MINIMUM SPEED.
3. Reactor LEVEL is within the normal operating band and may be considered to stay in the normal band for purposes of answering the following questions.
Assume an operator inadvertently transfers the "A" Recirc Pump' M/ A transfer station to AUTO. WHAT will be the final stable Pump A speed with: a) to*al feed flow < 20% and WHY? (1.50) b). total feed flow > 20% and WHY? (1.50) QUESTION 3.04 (2.00) a) There is an administrative limit that requires each APRM to receive inputs from each of the four LPRM axial detector locations (A,B,C & D). How does one determine the availability of LPRM axial detector inputs? (1.00) b) How does the APRM Inop Trip Circuit assure that required LPRM detectors inputs are present? (1.00) DUESTION 3.05 (2.50) What are three basis for setting the -38" reactor level setpoint for HPCI and RCIC initiation? (2.50) (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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, , ' QUESTION 3.06 (2.00) With respect to the control rod drive system scram dump volume: a). Why is the scram dump volume level monitored? (1.00) b) Why are both float type level switches and level transmitters used to monitor the scram dump volume level? (1.00) i QUESTION 3.07 (3.00) l ' Name six of the-eight parameters that are monitored to provide a closure signal'to the Main Steam Isolation Valves during reactor operation. Briefly explain why each parameter is monitored? (3.00) (Mode switch in "run") , ' QUESTION 3.08 (3.00) a) Name the signals that should cause an automatic initiation of the Law. Pressure Coolant Injection (LPCI) system (1.00) < -b) State how to manually initiate the LPCI system as.per procedure SP 23.204.01 if failure to automatically operate is identifi ed. (di scuss in standby AND SDC modes) (2.00) 4 -
1 OUESTION 3.09 (1.50). l The.Ch. A IRM is reading 100 on range 8.
What will the reading
be when switched to range 9? Why? (1.50) ! ' QUESTION .3.10 (2.00) What four-(4) conditions will cause the RSCS to apply a rod block.
(2.00) , , , (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****) . -.. . -. -...- - .... - -. - -. - .. -. . -..-. , -.. -.
Iz__INSIBUMENIQ_6NQ_QQNIBQ6S PAGE
l ' .
. DUESTION 3.11 (1.00) What.is (are) the source (s) of operating gas to the Inboard and Outboard Main Steam Line Isolation Valves? Indicate the normal supply and why system is used.
(1.00) (***** END OF CATEGORY 03
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s '4 __ESQGERUBES_ _NQBd@6t_6BNQBM86i_EMEEGENQY_@NQ PAGE 13- ' t E89196091G86_GQNIBQL .
,
. QUESTION 4.01 (2.00) Explain whether either of the following situations describes a case where 10CFR2O Limits f or individuals working in restricted areas have been violated. (Both men have updated NRC Form 4s on file) a.
A skin dose of.7000 mrem is received by a 44 year old. male over a full quarter.
His lifetime dose is 55 rem.
(1.003 , ~ b.
A whole body dose of 1500 mrem is received by a 44 year old male over a three (3) month period.
His lifetime dose ' is 22 rem.
(1.00) QUESTION 4.02 (2.00) According to Procedure SP 20.004.01, " Emergency Use of SLC"; i a.
WHO is responsible for determining if Standby Liquid Control-initiation is necessary? (0.50) b.
Name five (5) items which should be checked in the control-room to verify that-the SLC tank contents are being injected into the vessel.
(1.00) c.
What is the purpose of the heat tracing that is provided on the SLC system piping? (0.50)
' , OUESTION 4.03 (2.00) i ' According to SP 23.116.01, " Main and Auxiliary Steam": a.
What should be your immediate actions if you have an s inadvertent MSIV closure during startup with reactor < power at 40% and reactor pressure at 1000 psig? (1.50) , T +7.
, , b.
According to the procedure, what is the limit on differential pressure across the MSIV's prior to _ opening them? (0.50) e (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
. JW' . ... 'n.
- ~ V- ~ ~ ~ ~ ' ' j A__EBDGEDUBES - - NgRMALc ABNORMA'.j2_GMERGENCY[AND LPAGE-14 , q, Begig60GICAL'= CONTROL: f, f . ,
hi c d) $.. ',
- '
' ' ' . q ~ _r_.N , m
-QUESTION' '4. O'4 I ,r' (3.00) , + , 'f e ,- . < > . . Priop to starting an idle recirculation loop, the fpilowing three - conditions must be met.by procedure: ~, s tj i, , steam space.and. bottom head r , ., ' I *[ A. Delta T between the vessel i.
il.
drain must'Ybe less than or equal to 145 degrees F.
' -
,sb, f . + j 27 Delta T between the-idle and operating loops muAt be; . less than or squal to 50 degrees F.
' . .., . - s '3.
Operating. loop flow.must be less than or equal to . ' 50% of rated.' ' ' T .(3.00) ,p , . , . the above can be verified.
L Briefly explain one way that each of
Vn . , ' QUESTIOt! 4.05 & (2.00) - H if - According to EP 23.604.01, "APRM System": i ' f. t.'a ) When;is'an'APRM channel defined;as inoperable as LPRM inputs (1.00)- 4' are lost? ', .1 - \\ lj.
_c , 4b')-During a normal shutdown, when should the operatorq witch s the IRM/APRM recorder switches to the IRM position 7 (1.00) t - , i.
. a, f '. / ( . s -
', , 'f ' k(2.50) , -QUESTION 4.06
- l4
y 1: % , N. - N!?Iat.are the six pararreters that can be monitored from ' i .!, fl'the :Nemote shutdown p,dnel accord'yng to SP 29.022.01, , ,J - Sbutdo.wn From Outside Control Room - Emergency Procedure?- (2.50) .
.e s.s ' !;i.
, a pr( - .,. S
- J
'~ ' i g , < . , - .,. ,, , i j, f , \\.
,'
i e !' _ ll " \\ l' ' ' !, Zs y- ' , i 4,. ,t., tr y . .e (~ ' (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) ., '\\
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s > '> b .---~,-.,n-,.. ,c., ,c.
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PAGE
L4___EB9GEDUBEsi-NORMAL 3_ARNQRMAL2_EMERQgNQX_AND BODI96991G96_GQNI8g6 , . . . QUESTION 4.07 (2.00) When using procedure, SHUTDOWN - FROM 20% POWER, SP 22.005.01, and.you have a achieved a " hot shutdown" condition, what cautions apply with respect to: a)' operator control of reactor pressure? (1.00) b) operator action on inadvertent safety relief valve actuation?. (1.00) . ~GbESTION 4.08-(2.00) L%a " MAIN TURB EXH HOOD TEMP HI" alarm annunciates during startup e sc. 4s confirmed by a high exhaust hood temperature: C What control room indications are'available to confirm
- his high temperature condition?
(1.00: . u) What is your immediate action? (1.00)' . QUESTION 4.09 (1.50) Match the notification signal (a - c) with its' tone description.
(1.50) a.
Evacuation 1.
distinct steady tone b.
Fire 2.
one steady tone c.
Security 3.
ringing gong 4.
pulse tone 5.
siren 6.
warble . QUESTION '4.10 (2.00) With the reactor caerating at power, a loss of both CRD pumps occurs. Briefly explain two (2) undesirable effects of this loss.
(2.00) i.
!. i f (***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) i ., . -
_ =St__EB99EDWBES - NORMAL _ABNQRMAL _EMg89ENgy_AND PAGE
i i BSD196991G66_G901896- - .. .. , QUESTION 4.11 (2.00) What are the " Administrative Radiation Dose Guides" per SP 61.012.01 " Personnel Dose Limits" for: a) whole body per week during operation? b) whole body per week during outages? c) whole body per quarter? d) whole body per year? (.5 each) (2.00) OUESTION 4.12-(2.00) In accordance with SP 22.001.02 " Reactor Criticality Procedure" complete the following: a) During.startup, the minimum administrative positive period is _ _ Reactor period may-be calculated by . multiplying by 10.5 the time in _________ required to increase reactor power by ______.
(fill in the blanks) (1.00) b) During plant heatup, why are-you are cautioned against operation of-the mechanical vacuum pump when the reactor is above 5% thermal - power? (1.00).
(***** END OF CATEGORY 04
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(************* END OF EXAMINATION ***************)
- - _ _ - _ _. _ _ - _ _ _ 957(ilbiEB V% - 1x__EBIgCIE6ES_gE_Nyg6E68_EgWEB_E68NI_QEEB6IlgN PAGE
' i IHEBMggYNSMJCgi_HEBI_IB6NSEEB_8Ng_ELQIg_ELQW - . ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
. ANSWER 1.01 (2.00) a) The time to reach steady state becomes progressively longer.
(1.00 b) The neutron level change for successive rod increments becomes (1.00) greater.
REFERENCE SNPS Reactor Physics Module - Lesson 16 292002 kl.07, 3.5
ANSWER 1.02 (1.50) The three major functions of the control rods are: _ a) to change the degree of criticality for the purpose of (0.50) startup, and shutdown.
b) to compensate for changes in reactor reactivity during (0.50) power operation.
c) to provide a rapid shutdown capability (0.50) REFERENCE SNPS Reactor Physics Module - Lesson 13 292005 kl.01, 3.2 ANSWER 1.03 (2.00) 'f} Conduction - heat transferred through the tube wall of an RHR Heat exchanger due to difference in temperature between the inside and outside of the tube wall.
(0. 50) f, Convection - heat transferred to the molecules of a fluid and 'b transferred by the movement of the fluid as in the cleanup regenerative heat exchanger.
(0.50) Radiation - heat transferred through space by wave energy or .f1 photons as in the heat transferred from the vessel wall to the vessel concrete shielding.
(0.50) _
. -. " I2__EBINCIELgg_gF_NyyLEAR_POWEB_ELANT_gEgRAIlgN 2 _ PAGE
, _,. IHEBMQDyNOMICg2_HEBI_IB6NSEEB_6ND_ELUID_ELQW - ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
., > , REFERENCE SNPS Heat Transfer and Thermodynamic Module - Lesson 5 292OOB kl.01, 3.2 . ANSWER 1.04 (2.00) ONSET OF TRANSITION BOILING is the point where there is a sufficient heat flun to cause the bubbles on the clad surface to effectively group together and blanket the clad in ' localized areas.
(2.00) REFERENCE SNPS Heat Transfer and Thermodynamics Module - Lesson 7 _293008 kl.10, 2.9 ANSWER 1.05 (3.00) a) LHGR - Linear Heat Generation Rate (.5) designed to limit the pin power at any node in the reactor to a value that limits the fuel clad strain to less than one percent plastic st-ain (.5).
(1.00) b).APLHGR - Average Planner Linear Heat Generation Rate (0.5) designed to limit average pin power at any node to a value such that following a design basis accident the maximum fuel clad temperature will not exceed 2200 degrees F.C.5) (1.00)
I c) MCPR - Minimum Critical Power Ratio - (. 5) designed to limit the power of any fuel element to below the value that any point in the bundle will experience the onset of transition boiling (.5) (1.00) REFERENCE SNPS Thermal Based Limitations Module - Lesson 1 '293008 kl.10, 2.9 . , ,- -- r-y y-<-.----- -- - , ,,,,,s -,,--,v- - - -. - - - --n,ry- --- - ,-y ---+c-- - - - - - - -, -
'It__EBINGLELES_QE_NQGLE68_EQWEB_E6@NI_QEE66IlgNz - PAGE 19J IHg85QQYN6d1GSt___ HEAT _T86NSEEE_8NQ_E6Q1Q_ELQW . - - g.
_ ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
' e ' ANSWER 1.06 (2.50) f Reactor Power =P W P= (MG-x HG) + (MCU x (HCU,in - HCU,out)) + Ofl-- (MFW x HFW) - Op - (MCRD x HCRD) (2.50) REFERENCE SNPS Heat Transfer and Thermodynamics Module, Lesson 3, Objective 1 292008 kl.11, 3.7 ANSWER 1.07 (2.00) a) The resulting reactor pressure increases, decreasing voids, thus adding reactivity (until the system is tripped by either high flux or high prescure.)
(1.00) b) The resulting reactor pressure decrease will cause an increase in voids thus reducing reactivity and reactor power (and this effect will continue to " snowball" until the unit is shutdown.)
(1.00) REFERENCE SNPS SH # HL657, objective 1 241000, k4.01, 3.8 i- . _.. .. _... .. . .... _ _ _. _ _ _ _ _ _ _ _. __ _ _... _ _ .... _, - _ _. ..., _ _ _.. _ _.. _ _ _ _., _.
_ _, ~ _ -it;_PRINQlELES_OF NQQLEA8_EQWEB_ELONT OPERATigNs PAGE 20-IMEBM90YN@M1GS _Bg81_IB8NSEEB_@ND_E(QlD_E6QW t . . ~ ANSWERS -- SHOREHAM- .-86/12/15-TURNER,R.
. . ANSWER 1.08 '(2.00) u Super-critical (0.50) Definitions / example Critical - reactor power is constant, neutrons per generation is constant.
(0.50) ~Subcritical - reactor power is decreasing, neutrons per. generation are decreasing.
(0.50) Supercritical - reactor power is increasing, neutrons per generation are increasing.
(0.50) . REFERENCE SNPS Reactor Physi cs Module, Lesson 7, pg.7-95, objective 1; 292002, K1.07, 3.5 l ANSWER 1.09 (1.50) a) Delayed neutrcos allow safe reactor operation by making it possible for an operator to control reactor period.
(0.75) b) Beta is equal to a.s+e. sycesy Fe4c,b e7 (0.75) REFERENCE SNPS Reactor Physics Module - Lesson 15 292003'kl.07, 3.3 ANSWER 1.10 (1.00) a) (50+14.7-18.9)= 45.8 psi (NPSH) (0.50) b) No (0.50) REFERENCE SNPS Fluid Mechanics Module - Lesson #5, objective #9; 293003, kl.23, 2.8 l
. - - . . It__EBINglELES_OE_Nyg6 EAR __ POWER _ELANT DEER @llQN PAGE
m-IMEBMQDYN8dlCS _ME81_188NSEEB_QUD_ELylD_EbgW t . .. ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
. ANSWER 1.11 (2.00) 60 - 70 hours (1.00) It will take approximately.70 hours for the Xenon to peak and .then decay after the scram.
If the positive reactivity inserted by the decay of Xenon is less than the shutdown reactivity due to rods, then the reactor will remain subcritical.
(1.00)' . REFERENCE SNPS Reactor Physics Module - Lesson 11, pages 7-154 to 7-161 Student Objectives #1 & 2 292006, K1.07, 3.2
ANSWER 1.12 (2.00) TRUE (0.50) x In time / period and Using -the equation Power = Power (initial) solving.for time results-in the equation: time = Period x In Power /PowerCinitial) From this-it can be seen that since 5/1 yields the same value as 50/10, and since all other-factors in the equation are equal, the (1.50) time is equal.
- REFERENCE SNPS Reactor Physics Modul e - Lesson 15 Student Objective # 1
292003, K1.05, 3.7
Y ' ANSWER 1.13 (1.50) (0.50) a.
True (0.50) 6.
True (0.50) c.
False REFERENCE SNPS Reactor Physics Module - Lesson 13, pages 7-201 & 202 - Lesson 15 Objectives 292005, K1.12, 2.6 . , ., ,. . . -. - -. .. - ..-. . ~.._-. -.., -..
2m__ELeNI_QEgigN_lNQLyDINQ SeEETY ANQ_EMER@ENGY_@Y@TEMS' PAGE' 22 '
- ANSWERS -- SHOREHAM-86/12/15-TURNER,R.-
. ANSWER 2.01 (3.0 ) a.
30 - 46 gpm (C/ g[6 0.50)
( 260 psid, accept 250 - 270 psid (0.50) ht.L e p (1.00) b.
100% / _ stabilizing valves,/ rom the.valv - m 'c.
When a' rod is inserted, one set of to the CRD and away .close (0.5) to direct 4 gpm (0.5) cooling water header.
(1.00) REFERENCE SNPS Student Handout # HL106 Student Objectives F.a, F.e.
F.d, & F.c 201001 kl.10, 2.8 ANSWER 2.02 (3.00) a.- Fal se (0.5)- Once the low suction pressure signal is clear, the turbine will auto restart if the initiation signals are still present.(0.5) (1.00) b.
True (0.5) - The l ow steam pressure auto isolation signal seals in, and must be manually reset (psing the AUTO ISOLATION SIGNAL RESET pushbuttons on the *PNL-601 after the reason for the isolation has been determined and corrected)((0.5) ft.n (1.00) lA c.
True (0.5) - The oil pressure will be restored when the turbine coasts down, thereby causing the stop valve to open. (0.6) (1.00) REFERENCE High Pressure Coolant Injection System and, SNPS Procedure 23.202.01, Rev. 9 206000 k4.01, 3.8
-, -. . -,.,. - , .,, -, -. - - -.. r, ,-. ..,.....-.,-,e.. - - _.,-
r '3&__EleNI_ DESIGN _INCLUQ1NG,,S8EEIY_6ND_EUER@ENCY_SYSIEUS PAGE 23.
. ~ * ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
. ANSWER 2.03' (3.00) (a) A drain path would be established from the CST to the Suppression Pool.
(1.00) (b) 1.
Water. hammer may result in the exhaust line.
(1.00) 2.
At. lower speeds, adequate cooling and lubrication of 'the turbine may not be ensured.
(1.00).
REFERENCE a-Handout'119 Section 8.3.2.1 b-SP 23.119.01 page-4 217000 kl.03, 3.6; k4.01,2.8 ANSWER 2.04 (2.00) a) CR indications: core spay flow (and pump motor amperage) (1.00) Hb) If one core spray system is providing design flow it can be ,. assumed that the core is being adequately cooled.
(1.00). REFERENCE Core Spray System Student Handout No. 203.
209001 kl.14, 3.8 ANSWER 2.05 (1.50) Lube oil would heat up in the Condensate and Booster Pumps.
(1.50) (accept Condensate and Booster Pumps may have to be tripped) REFERENCE SNPS'SH # 126, Objective 7.4 295018 AA2.01, 3.3 ,
I' N - - -. -.. J
_ - _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ 2i__EL8MI_REE1EN_INGLUQ1NG_SAEETY ANQ_EMER@ENCY'SYSIEMS P A G E-.
~' 4 ANSWERS'!-- SHOREHAM' -86 /12/15-T URNER, R.
'
. , . ANSWER 2.06 (3.00) a).No-(.5) It takes three rods for the rod block (.5) or (1.00) Yes (.5) If'there is a third insert error (.5) b) Either of the above is correct if the assumption is stated as to whether there is a third error.
(1.00) c) If Insert Rod Block Window is lit (red) a insert rod block is in force.
(1.00) ' REFERENCE-SNPS SH # 607/609,' objective I C 201006 k4.01 3.4 , ANSWER 2.07 (3.00) a) decreases b) decreases c) decreases d) decreases (.75 each) (3.00) REFERENCE SNPS SH # 656, 259002, k6.03. _ . 12c__EL8HI_ DESIGN _lNCLUDING_S@EETY ANQ_gMERGENGY_SYSIEMS PAGE
- ANSWERS -._SHOREHAM.
-86/12/15-TURNER,R.
. . , ANSWER 2.08 (3.00) a)-Recirc.. pump discharge valve must be greater than 90% open (.5) to reduce axial thrust on the pump bearings during high' speed, low flow conditions- (.5). (1.00) 'b) Reactor water level must be higher than 12.5 inches (.5) and if tripped must be-manually reset before the 30% speed limiter is bypassed (.5). (g g gg payer),g hpg#) (1.00) W c) Feedwater flow must be greater than 20% or greater of , rated for at least 15 seconds (. 5). to prevent pump cavitation (.5).
(1.00) REFERENCE SNPS SH # HL658, objective 1 202001, k4.16, 3.3 - ANSWER 2.09 (1.50) Control Room - manual pushbutton (.5) mode selector switch in LOCKOUT (.5) Local - mode selector in LOCAL and local STOP - push button depressed (.5) (1.50) REFERENCE ' SNPS SH 4 307, 7.6 295024, ek2.06, 3.9 t ANSWER 2.10 (2.00) a) Triple low vessel level, and b) confirmation low vessel level, and c) 105 second timer timed out, and d) a low pressure ECCS pump running (.5 each) (2.00)
. 3a _ PLANT _DEglgN_JNGLUDING_gAggIY_AND_gMERQgNCY SYSIEMg PAGE
,. ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
. REFERENCE SNPS SH # HL-201, Objective 5 218000, k4.03, 3.8 i i l i I l i I ' st__INSIBUMENIS_6ND_CQNIBQLS PAGE
~
ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
. ' ANSWER 3.01 (2.50) a.
All parameters are normal for an ADS initiation, but the ADS system has not initiated.
The status of the white indicating lights on the Automatic Blowdown Panels indicates that only one of the 2 channels in each of the 2 logic systems energized.
Since the logic systems are 2 out of 2 once, the valves did not open.
(1.5) b. Attempt to manually initiate the system by(verifying that at least two RHR or one CS pump is running, and try to open the ADS valves using the individual control swi ches.
(1.0) Y
REFERENCE 1.
Automatic Depressurization/Saf ety Relief Valve System Student Handout
- HL-201 Student Objectives # 3& 10 2.
SP 23.201.01.
21S000, k4.02, 3.8 ANSWEF 3.02 (2.50) ": ;; i :"' w a m is s u o. w wa m w 's" Y37N6 M.
c.
YES EO.53 d.
NO [0.53 e. ~N9-CO.53 $ws3 BCG f,f M Y/L REFERENCE SNPS SH HL-631 Radiation Monitoring System, objective 3 272000, kl.06, 3.2 ANSWER 3.03 (3.00) a) "A" pump speed will stay at minimum speed (0.75) because the 30% limiter is limiting speed with feed flow < 20% (0.75).
(1.50) b) "A" pump speed will go to high speed limit of the master controller dual limiter (0.75) because with feed flow > 20% and the discharge valve open the 30% limiter is bypassed (0.75).
(1.50) _
L__INSIBLJt!ENIS_OND_GQ$IBOLg - PAGE
'
ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
e , ' ~ REFERENCE SNPS SH 658 Recirculation Flow Control System, Objective #4 202OO1, k4.16, 3.3 ANSWER 3.04 (2.00) a) The availability of axial detector inputs can be determined by by examining the detector status indicating lights and the ' OPERATE / BYPASS / CALIBRATE switches on H11-P608 in the control (1.00) room.
b) APRM Inop Trip Circuit accepts a trip setpoint signal corres-ponding to 11 LPRM detector inputs from the Trip Reference.
unit. It also accepts a signal proportional to the number of LPRM inputs. It the latter is less than the-former an Inop ' signal is generated.
(1.00) REFERENCE SNPS SH 4 603,604,6D2, objective 3 215005, Lt.0), 4.0 . ANSWER 3.05 (2.50) 1) Approximate level expected after void collapse on scram from (0.50) full power.
2) Avoid false starts of RCIC & HPCI after a reactor scram due to low level (12.5"). (1.00) 3) High enough so that HPCI & RCIC can prevent level falling to level 1 (-132.5"). (1. OO) H REFERENCE SNPS'SH HL-621,622, & 624, Objective 7.1 216000, kl.1*., 3.6 k gi r a gig n - e e
~ . 3u__IN@l8QUENI@_AND_CQNI69L@ PAGE
^ - e.
ANSWERS --'SHOREHAM- -86/12/15-TURNER,R.
' . ANSWER 3.06 (2.00) a) The SDV should normally be drained during operation (.5), if excessive leakage or inadvertent closure of vent and drain valves occur, the SDV could fill and potentially hinder scram of the control rod drives (.5).
(1.00) b) Float type level switches and level transmitters are connected to the instrument volume to monitor for abnormal water' level to assure reliability (.5) of the monitoring system due to the importance of assuring space for the exhaust water following a scram (.5) (1.00) REFERENCE SNPS SH HL-106, Objective 1 a 201001, k4.11, 4.0 ANSWER 3.07 (3.00) 1) Main Steam Line Low Pressure (.2S) - protects against open control or bypass valves cr turbine pressure regulator failure (.25).
2) Main Steam Line High Area Temperature (.25) - protects against a breach in main steam line outside of primary containment (.25) 3) Reactor Vessel Low-Low Level (-38") (.25) - protects against a breach in the nuclear primary pressure boundary (.25) 4) Mein Steam Line High Radiation (.25) - protects against a gross release of fission products from the fuel (.25) protects against a break in 5) Main Steam Line High Flow (.25) - the main steam line(.25).
6) Low Condenser Vacuum (.25) - protects condenser and isolates reactor vessel (.25) (3.00) LM h _ :k 7. ML w. % Q C - s. w %..o h. q. - % M, m REFERENCE SNPS SH # HL-650, Objective 2 239001, k4.01, ~5. 8
- _ _ - - _ _ - _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ , 3,___1NSIEUMENIQ_AND_CQNIBQLS PAGE
$ . ANSWERS -- SHOREHAM ,
e . v ~ ANSWER 3.08 (3.00) a) Auto initiation should have occurred at (-132.5") decreasing water level (.5) and/or high drywell pressure of (1.69 psig) (.5). (1.00) 'b) 1.
If RHR is in the Shutdown Cooling Mode, the pump suction valves (MOV-31 A-D and MOV-32 A-D will have to be realigned by the operator and the (MOV-037A(B)) SDC' Isolation Logic must be reset.
(1.00) 2.
Arm and depress both Manual Initiation push button switches.
(1.00) ) REFERENCE SNPS SH # 204, Objective 5 SP 23.204.01, Pg. 7 203000, F4.01, 4.2 ANSWER 3.09 (1.50) Reading will equal 10 as the ten ranges cover only five decades and the value between ranges is dependent on the specific ranges that are adjacent.
(1.50) REFERENCE SNPS SH # 602, Objective 7.7 215003, L4.04, 2.9 ANSWER 3.10 (2.00) a) When a control rod reaches one of its group bank position.
(0.50) b) When control rod movements do not meet the Rod Pattern Control rod pattern criteria.
(0.50) c) When a control rod passed by an even reed switch that is' failed.
(0.50) d) When the RPC has failed, or cannot find a correct core (0.50) rod patter =. - _..- - . ... A__INSIBU!'gNIS_8Np_ggNIBQ65 PAGE
- [-ANSWERS -.SHOREHAM
.-86/12/15-TURNER,R.-
, ., , REFERENCE-SNPS'SH #609,' Objective D 201004, k4.02,k4.03,3.1,3.3 g . ' ' ANSWER 3.11 (1.00) . ' Inboard MSIV's'- operated with nitrogen. gas.(.25) so as not to dilute.the inert drywell atmosphere (.25) (0.50) ,. Outboard MSIV's - operated with instrument air (.25) 4.o take ' advantage of the backup to the insr*rument air system for reliability (.25) -(0.50) REFERENCE-SNPS SH # 116,. Objective CG 239001, k5.06, 2.8 . e P P P ,w.
.a ---. ,v.m..w,--,-c.m.m-.,.r, ...~w.._.,w.37...,-.m,,_,.,e -,-..m.w..,,-.m-,,,,.m,,.._,,. ,. -.,,. ~,,. .. _... -
4 __88QG'EQUBES_:_NQBd@bs_8@NQ8d@bi_EMEBQENQY_6NQ EPAGE 13 t . ,1 88Q1060Q1G86_GQNIBQL ,.- ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
- * - j.
ANSWER-4.01-(2.00) 44 = N -5(N-18) = 130 rem.
NeitherEcase has exceeded the lifetime dose under 10CFR2O-(1.00) .a.17000 mrem to the skin does not exceed the 7500 mrem limit /qtr.- (0. 50).
b.
The 1250 mrem /qtr whole body limit has not been exceeded-because he has-an updated NRC Form 4.
(0.50) ! REFERENCE 10 CFR 20, Section 20.101 294001,.kl.03,H3.3 ANSWER 4.02 (2.00) J (0.50) a.
The Watch Engineer 6. - Squib valve loss of continuity alarm - SLC pump discharge pressure >= reactor pressure - Both squib valve lights out . Selected pump running light on - SLC Tank level decreasing - Reactor power decreasing-( 5 of 6 9 0.2 pts each) (1.00) c._to prevent precipitation and crystallization of the solution in the pump suction lines (0.50) REFERENCE SP 29.004.01 and Student Handout " Standby Liquid Control" # 123 '211000, A1.01-1.0,, 3.6-. '2x__E89GEQUEES_:_NQBd86t_8EN98586t_EMEB@gNCY '6NQ PAGE
88Q196991986_GQUIB96
.; , ANSWERS -- SHOREHAM-es/12<1e-11mnEs;s.
. ANSWER-4.03 (2.00) a.
Monitor and ~ maintain reactor pressure using control-rods to adjust power as required.
(0.7G) Start RCIC to assist in controlling reactor pressure and level.
-(0.75) b.
200 psid $6 - (0.50) REFERENCE SP 23.116.01, page 10 SH #116, Objective 1 223002, k3.O',-3.7 ANSWER' 4.04 (3.00) 1. Verify Delta T between steam dome and drain by Process Computer printout, Vessel Pressure. Convert to temperature with steam tables and compare with RWCU drain temp.
(1.00) 2.
Verify by comparing Process Computer points and parameters on PNL 602 (1.00) 3.
Verify loop flow by recorders on PNL 602, Process Computer or Jet Pump Flows.
(1.00) REFERENCE SNPS SH # 120, Objective #4 202001, k4.12,3.2 . ANSWER 4.05 (2.00) a. When thers are less than 2 LPRM inputs per level or less than 11 LPRM inputs total to the APRM.(will accept not working) (1.00) b. When the first APRM Downscale alarm light illuminates.
(1.00) ,
c- -- 6;-- , , di__PRQQEQURES NQRMAbt_ABNQRMAL _EMERQENCY ANQ PAGE
t .
Benig6001CeL_QQNIBQ6 ANSWERS -- SHOREHAM-86/12/15-TURNEP.P.
. REFERENCE SP 23.604.01, pages 2 and 4 SNPS SH #604, Objective 4 215005, A2.03, 3.6 ANSWER 4.06 (2.50) 1) Reactor Vessel Water Level (LI-OO4) 2) Reactor Vessel Pressure ( P I'-OO6 ) 3) Suppression Pool Level (LI-026) 4) Suppression Pool Temperature (TI-022A & B) 5) Drywell Pressure (PI-012) 6) Drywell Temperature (TI-021) (.5 each up to 2.50) (2.50) REFERENCE SPl Number 29.022.01 294001.11.01,3.7 [h ANSWER 4.07 (2.00) a) The operator should not attempt to control reactor pressure with contral rod movement.
g ioo g g 4 (i,00) n b) If inadvertent safety relief valve actuation should occur while in hot standby, manually scram the reactor.
6/tg 3 (1.00)
REFERENCE Wf4 SP Number 22.005.01, Pg. 13
m 201001,gl.a10,3."
ANSWER 4.08 (2.00) a) Spray header pressure (.5) and exhaust hood temperature (.5) (1.00) b) Open the Exhaust Hood Spray Bypass Valve (1N21-MOV-11)(.5) to obtain about a 7 puig in the spray header (.5) (1.00) REFERENCE SP Number 23.127.01, Pg. 29 245000,k4.05,.. "I di.__EBQGEQWBES_:_NQEd66t_6@NQBd66t_EdE8GENCY __ ANQ - PAGE ;35
- 8001QLQQ1G86_GQNI6QL
c .. ANSWERS - SHOREHAM-86/12/15-TURt!ER,R.
to' , ANSWER-4.09 (1.50) / a) 4,6 b.-
2 y1 _{- c.
. 0.5 each) ( REFERENCE SNPS General Employee Training-294001,kl.16,3.5 ANSWER 4.10 ,(2.00) The loss of the CRD pumps causes a loss of cooling water-f l ow (0.5).
~which shortens CRD seal life (0.5), and also allows the CRD hydraulic accumul ators to slowl y ceprescurl:e (0,5) which reduces the scram capability of L:m .. L : 7> ' <. t p e d i + : i - s' 'c.
prewd.
et.
.(Also accept loss of seal purge to RWCU and RR pumps which can shorten seal life and loss of ability to move control rods with RMCS as . hydraulic pressure iC lost.)
REFERENCE' SNPS SP' Number 23.106.01 Pg. 9 201001,L5.01, 2.4 ANSWER 4.11 (2.00) a)-200 mrem /wk b) 500 mrem /wk c) 1000 mrem / quarter d 4000 mrem / year (.5 each) (2.00)
'53__EBQGED9BES_:_N9BdeLz_ GEN 9Bd862_EMEB9ENQY_8ND ' PAGE
802196991986_GQNIBQL - - . ANSWERS -- SHOREHAM-86/12/15-TURNER,R.
.. J.WVVT' ** A --^^^ ^ ] A ANSWER.
4.12 (2.00) --- /IC#"~' (1.007 a) 60 seconds (0.5), seconds (0.25), 10% (0.25) b) Significant levels of hydrogen and oxygen would be present in the condenser,(0.33) and could result in a detonatable mixture of hydrogen and oxygen (at atmospheric pressure) (0.33). If combustion were to occur, the event would pose a personnel hazard since the pump and water separator are not designed to be detonation resistant.(0.34) (1.00) REFERENCE SP 22.OO1.01," Start Up - Cold Shutdown to 20%" 201001,GKA 10, 3.2 .
o =.
- TEST CROSS REFERENCE PAGE
. tuESTION: -VALUE REFERENCE -.--___-
4: ~ 01.01 2.00 TRLOOOO122 01.02 1.50 TRLOOOO123' 01.03 2.00-TRLOOOO124- -01.04 2.00 TRLOOOO125 ' 01.05 3.00-TRLOOOO126 01.06 2.50 TRLOOOO127 01.07' 2.00 TRLOOOO135 01.08 2.00 TRLOOOO115 01.09 1.50 TRLOOOO121 01.10 1.00 TRLOOOO117 01.11 2.00 TRLOOOO103-01.12 ~2.00 TRLOOOO104 01.13 1.50 TRLOOOO105 ______ 25.00 02.01 3.00 TRLOOOO106 02.02 3.00 TRLOOOO107 02.03 3.00 TRLOOOO108 '02.04 2.00 TRLOOOO128 02.05 1.50 TRLOOOO130 02.06 3.00 TRLOOOO131
O-' % 00 1F>! 0000) 72 02.00 3.00 TRLOOOO133 02.09 1.50 TRLOOOO136 02.10 2.00 TRLOOOO137 . -_--__ 25.00 03.01 2.50 TRLOOOO109 03.02 2.50 TRLOOOO110 03.03 3.00 TRLOOOO111 03.04 2.00-TRLOOOO138 03.05 2.50 THLOOOO139 03.06 2.00 TRLOOOO140 03.07 3.00 TRLOOOO141 03.00 3.00 TRLOOOO142 03.09 1.50 TRLOOOO145 03.10 2.00 TRLOOOO148 03.11 1.00 TRLOOOO149 ______ 25.00 04.01 2.00 TRLOOOO112 04.02 2.00 TRLOOOO113 04.03 2.00 TRLOOOO114 04.04 3.00 TRLOOOO146 04.05 2.00 TRLOOOO147 04.06 2.50 TRLOOOO150 04.07 2.00 TRLOOOO151 - - - - . - - - - - - . - -
- -
. TEST CROSS REFERENCE PAGE
,
- 1-NUESTION VALUE REFERENCE
,04.08.
2.00 TRLOOOO152
- -
04.09 1.50 TRLOOOO153 .04.10 2.00 TRLOOOOl'54 04.11 2.00 TRLOOOO155 .04.12 2.00 TRLOOOO156 ___ 25.00 -_____
100.00 -- _ _ - _ _.
_ _ _.. _ _ _ _ _
Nk - 't U.
S. NUCLEAR REGULATORY COMMISSION /[rrw hmen/ P- ? SENIOR REACTOR OPERATOR LICENSE EXAMINATION ' .., FACILITY: _@HQREHAM_____ _______
e REACTOR TYPE: _EWR-@E4 ______ DATE ADMINISTERED:_g6/12/1g________________ EXAMINER: _JARRELL _D3 _
CANDIDATE: ____ ___________ , INSIBUCIlgNS_IQ_Q8NDIp61El Use separate paper for the answers.
Write answers on one side only.
Stcple question sheet on top of the answer sheets.
Points for each qu stion are indicated in parentheses after the question.
The passing grcde requires at least 70% in each category and a final grade of at Icest 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY __Y8LUE_ _IQIBL ___SCQBE___ _y@6UE__ ______________Q81EGQBY_____________ 24.Sc 2 S. 3 _@@I99__ _EEz99 ________ 5.
THEORY OF NUCLEAR POWER PLANT ___________ OPERATION, FLUIDS, AND THERMODYNAMICS 25.7 25.00 05:00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION I f. E _2Ez99__ _3DrQ9 .________ 7.
PROCEDURES - NORMAL, ABNORMAL, ___________ EMERGENCY AND RADIOLOGICAL CONTROL .Sc 23.l0 sEEI99'_ _25z99 ___________ ________ 8.
ADMINISTRATIVE PROCEDURES, _ CONDITIONS, AND LIMITATIONS 97.00 ' _ _ _ _ _ _ _ __ Totals 199I21__ Final Grade All work done on this examination is my own.
I have neither given j nor received aid.
=_____ _-=__ Candidate's Signature l > I I ,e n.y-. m ,- , _ _ _ _ _ _ _ _ ,-__r -. _.- ..c ,,-___..__-____.-9._.. .. _., -
' ' NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS . During the administration of this examination the following rules apply: .l.
Eheating on the examinat2on means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil gnly to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Categorv __" as appropriate, start each category on a ngw page, write gnly gn gng side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example.
1.4 6.3.
10. Skip at least thtgg lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility litetatute.
13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examinet only.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been complete. .,
- 18. When you complete your. examination, you shalls o.
Assemble your examination as f ollows: ' . (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still
in progress, your license may be denied or revoked.
, a }
i
, .- ,_,..,._..,-,,-,_._.c.
c_my__,. _, _ _. _ _ _., _, _ - .....,_..,,,_.-.___.-.._-_,.r__,_,,,,mm..._.-,___.
, _...
~5 __IhEggy_gF NUGLgAR_PQWER_PbANT_gPERATIgN _FLUIgg2_ANg PAGE
2
- .IHERMggyN8 MICE
. QUESTION-5.01 (2.50) With the reactor operating at 100% power, an SRV suddenly fails , open.
Using the attached Mollier Diagram, STATE all assumptions and answer the following.
I a.
WHAT SRV Tailpipe Temperature would you expect immediately after the valve opened? (0.75) b.
IF in case "a." the drywell pressure were to increase to 45 psig, WOULD the Tailpipe Temperature INCREASE, DECREASE, or REMAIN THE SAME7 (0.5) c.
With normal drywell pressure and the reactor pressure at 85 psig, WHAT would you expect the Tailpipe Temperature to be relative to case "a."? (INCREASE, DECREASE, or REMAIN THE SAME) (0.5) d.
At WHAT reactor pressure will the Tailpipe Temperature be at its maximum value (with normal drywell pressure)? (0.75) t DUESTION 5.02 (1.00) On a reactor trip from 70% power and equilibrium xenon conditions.
peak xenon will be reached in approximately ______ hours.
(1.0) (a.) 4 to 5 (b.) 8 to 10 (c.) 12 to 14 (d.) 72 i ! QUESTION 5.03 (2.00) . With the reactor operating at 75% power, recirculation flow control fcils high, rapidly increasing flow.
, , c.
WHICH reactivity coefficient will act first to limit the ! power excursion? (0.5) i t b.
JUSTIFY your choice of coefficient.
EXPLAIN this choice ' relative to all other coefficients.
(1.5)
i (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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, _. -. -__r._ - _,...,. ...., _. _ _ _. -, _ _m.,_...m.,_-,,..__.y_ ,, _.,,, _.. ,. _,., _._,_, _,,,..._-. _..,,, y _.,, .,_,,,.r,,..
Dz TbE98Y_9E_NUGLE88_EgWEB_EL9NI_QEEB8IlgN _ELUIDS _8ND PAGE
1
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. IHEBdQQXN@dJg@
. QUESTION 5.04 (1.00) Shoreham is operating at 50% power.
WHICH of the following best dsscribes the process that contributes the most significant amount of energy (provides the most thermal power)? (1.0) (a.) The fission process yields large fission fragments with high kinetic energy, which is converted into heat.
(b.) The fission process yields neutrons with high kinetic energy, which is converted into heat.
(c.) The fission process yields gamma radiation with high energy, which is converted into heat.
(d.) The fission process yields beta particles with high energy, which are converted into heat.
QUESTION 5.05 (3.00) Shoreham is operating at 100% power and BOL when you lose partial feedwater heating.
a.
If the STA tells you that reactor coolant temperature decreased by 10 deg F and voids decreased by 2%, WHAT would be the corresponding temperature change of the fuel? (ASSUME no rod movement, no recirculation flow changes and the reactor reactivity returns to zero.)
(1.0) (1.) ancrease by 30 deg F (2.) decrease by 30 dog F (3.) increase by 300 deg F (4.) decrease by 300 deg F b.
If the same situation was to occur at 50% power and BOL, WOULD the fuel temperature change be LARGER THAN, SMALLER THAN, or THE SAME as that in part a7 WHY7 (Include the effect of each coefficient.)
(1.0) c.
If the same situation were to occur at 100% power and EOL, WOULD the fuel temperature difference be LARGER THAN, SMALLER THAN, or THE SAME as that in part a? WHY7 (Include the effect of each coef fici ent. ) (1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
Qb_ IM QBY QE_N WLEAR POWER PLAN 1' OPERATION,[ELUI_QQ3,,,ANQ PAGE' ~ ^ ^ ' ~
4 I!!EBt!O YN9t!1CS .. - .. e' , - OtJES'l 10N 5.06 (2.00) The figure given below shows plots of the coolant temperature and coolant enthalpy as a function of the height in the core or flow path length, a.
EXPLAIN WHV the coolant temperature remains constant over much i of the flow path while the enthalpy continues to rise.
(1.0) b.
EXPLAIN WHY the slope of the coolant enthalpy curve increases from Pt. A to Pt. B then decreases from Pt. B to Pt.
C.
(1.0) j.
Coat 10P ~
a , \\. k
r OfA
' \\ \\ \\ \\ COOLANtituPtRATUnt FUEL fl0D SURFACE \\
- ft wtMATVA,1 I
\\ . . t Putt it0D Cf NTERLNE \\, '("****'U** E -
aI [ t I - - l h OULK00 LNG C00LANT ENTHALPY I . i AT l / oft FUEL g l flu ROD g / l ' , i f NUCLEATE.
l . / - , sotNo
_ ./ ,U,. .U., s.
. COOLNo C00tN0l / g (Bluttim)l
g p) , / Wt ___ "f GORE 8011064 7tWPER ATURC (NTHALPY,SE AT FLUE I ', (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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... _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _.., _.. - - _ _ _ _._ _ ______,. _.. _.
. _ _ _ _ _,.
Dz__I'Hgg8Y_QE_NQGLE98_EQWEB_EL@NI_QEEB@IlgN _ELQJQgz_@NQ PAGE
z . ISEBdQQYN@dIgg .
. OUESTION 5.07 (2.5GT 2 00 STATE whether the following INCREASE, DECREASE, or have NO EFFECT on Control Rod Worth.
'~ r~d i=r.= tty i==r==== 'n R) - 2-b.
gadolinium burning out (0.5) c.- void fraction decrease (0.5) d.
moderator temperature decrease (0.5) o.
fuel temperature increase (0.5) DUESTION 5.08 (1.50) Shoreham has been operating at 100% power for 3 weeks when the Reactor scrams.
You begin startup within 3 hours.
c.
Radially, WHICH control rods have the highest worth 7 EXPLAIN.
(1.0) b.
Driefly EXPLAIN WHY you must approach criticality very carefully under these conditions.
(0.5) DUESTION 5.09 (2.50) During startup (power = 1 watt) a rod is pulled and a 60-second period is observed.
a.
With NO further pulling of rods, CAN the operator maintain this reactor period for 30 minutes? EXPLAIN your answer.
(1.5) b.
By pulling rods, CAN the operator maintain the 60-second period for 30 minutes from the 1 watt power level without exceeding the maximum allowable thermal power for Shoreham? EXPLAIN your answer.
(1.0) (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
h__IBEQBY_9E_NJg6E98_EQM B_EL9NI_QEEBBIlgth_ELUlph,@NQ PAGE
ISEBU99YNebigs- ~ ,
. DUESTION 5.10 (3.00) The figure given below is the T-S diagram for the SNPS cycle.
For occh of the regions listed below, EXPLAIN WHAT i's occurring and WHERE it is occurring (e.g., high pressure turbine, condenser, etc.). a.
Region 3 to 4 (1.0) b.
Region 4 to 5 (1.0) c.
Region 9 to 1 (1.0) _ . . m . f e
w W
Cl s
DH<
C,
Wf Q.
y - o LU
1 - ' s p i ENTROPY (s) ' l l i i (***** CATEGORY 05 CONTINUED ON NEXT PAGE
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I i i l ' .-.-,..-.-. __ _ _ __ _ ~ -~ m,.. _.
, _.. . -,,.. -. _ __ _,. _ - - _ -. _ _ _. _ _ - .., _, _ . _ - _. _
Dz__IHE98Y_9E_NyCLE68_EQWEB_E(@NI_QEEB9IIQN _ELVIDgg_8NQ PAGE
i , IMESt?99YNet$1Cg
. QUESTION 5.11 (3.00) The cladding of a fuel pin develops a uniform coating of corrosion products on its surface which increases the resistance to the flow of heat.
Assuming that the f uel pin heat flux and geometry remains constant during the time of the buildup, WOULD you expect the following temperatures to (INCREASE, DECREASE, or REMAIN THE SAME) cs a result of the buildup? EXPLAIN the reason for each answer.
a.
fuel pin centerline temperature (0.75) b.
cladding temperature (0.75) c.
fuel pin to clad gap temperature (0.75) d.
bulk coolant temperature at the core axial midplane (0.75) DUESTION 5.12 (1.00) Consider that Shoreham is operating at 25% power when a leak results at the Recire A Reactor Water Cleanup Suction location.
As the effluent discharges from the leak into the drywell, it WILL: (1.0) (a.) all flash to superheated steam (b.) all flash to saturated steam (c.) part flash to steam and part stay liquid (d.) remain as a liquid (***** END OF CATEGORY 05
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. - -. . _. .- - . . _ _ _ _ -. .- bz__EL9NI_Sy@lgd@_QES]QN _QQNIBQL2,@NQ_INSIBydgNI@l1QN PAGE
2 - . t
,. QUESTION 6.01 (2.50) Concerning the CRD Hydraulic System: l a.
The reactor operator is going to increase drive pressure to the HCU.
WOULD you as the acting SRO direct him to OPEN or CLOSE the drive water pressure control valve? (0.5) . b.
EXPLAIN how your action in part a.
has changed the following flow rates (INCREASE, DECREASE, NO CHANGE).
(1.5) 1.
scram valve charging flow l 2.
CRD total system flow 3.
cooling flow ' j c.
The maximum charging water pressure is 1510 psig.
WHY is this limitation important? (0.5)
OUESTION 6.02 (2.50)
Both the SRM and IRM compensate their detector signals with a unique type of discriminatic, process.
I
c.
Briefly DESCRIBE WHY discrimination is needed in the SRM and
IRM range and DESCRIDE WHY it is not needed in the APRM range.
(1.5) b.
HOW do the SkM and IRM systems accomplish the discrimination process (short answnr)? (1.0) ! l J 4,
f ! , t
l i i i ! ' ' (***** CATEGORY 06 CON TINUED ON NEXT PAGC s es * *) i i i ! ! ' . .
bz__EL9NI_SYSIENS_9EH19Ni_G9 NIB 962_9ND_INDIBWNENIBIIGN PAGE
. n.
, QUESTION 6.03 (2.50) An automatic HPCI initiation has occurred.
Subsequently HPCI injection was automatically terminated due to high reactor water level, a.
WHAT valve (s) will auto close at WHAT level os. HPCI . termination? (1.0) b.
Assuming no operator action, WILL the HPCI respond to a subsequent decreasing reactor water level and if so at WHAT RPV level? (1.0) c.
If HPCI system had switched sources from the CST to the suppression pool due to low CST level and the CST level subsequently recovers. WILL the HPCI system automatically switch back to the CST suction? (0.5) DUESTION 6.04 (3.00) The reactor is operating at 100% power with the Electro-Hydraulic Control (EHC) load set at 110% and max. combined flow set at 105%. Using the EHC diagram in the handout. give WHAT the final position (in percent of full steam flow) of the control valve AND the bypass valves would bu for each of the fnllowing circumstances (assumo no operator action).
c.
load limit potentiometer reduced to 95% (0.75) b.
maximum combined flow limit potentiometer reduced to 95% (0.75) c.
"A" prennure regulator (transmitter) fails low (0.75) d.
two (2) bypass valves slowly go to full open position (0.75) (48444 CATEGORY 06 CONTINUED ON NEXT PAGE $$$88)
_ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ __ _ . _ _ _ _ _ _ _ _ _ _ _ _ ____ _ __ . _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ '6A__bbbdI_gygIgdS_pgSJgN _QQNIBQLi_AND INSTRUMENIAllQN PAGE
i - l .
es , DOESTION 6.05 (1.50) l Concerning the RHR Systems c.
WHAT is the reason for the interlock between the , 1.
shutdown cooling suction valves (MOV-32 A-D) and the f i test line isolation valve (MOV-040 A(B))? (0.5) 2.
outboard and inboard injection valves and Rx pressure? (0.5) b.
ANSWER TRUE or FALSE.
If a LPCI auto initiation function (high drywell) were . overridden to realign the RHR system to the shutdown cooling i mode and another LPCI signal (low level) were to come in, ! thw RHR loop WOULD rualign from the shutdown cooling mode to the LPCI mode.
(0.5) l DUESTION 6.06 (2.00) l ! Following initiation of the HPCI system, the Barometric Condenser condensate pump trips after one (1) minute of operation.
Assuming ' no operator actions i c.
WILL HPCI continue to inject? (0.5) t b.
If not, WHY not, OR i f so, DESCRIBE any adverse effects resulting from its operation.
(1.5) , QUESTION 6.07 (1.50) i The Division I diesel generator is running in response to an cutomatic initiation signal when an operator depresses the control I room generator BTOP pushbutton.
STATE the action which the operator must take from the control room to restart the DO.
(1.5) , (sesse CATEGORY 06 CONTINUED ON NEXT PAGE sesss) i I - . . - - -
_ __ _______-_-___ 6 __EL'ONI_SYSIgbg_ DESIGN _CONIBQLt_@ND_INSIBudgNI@IIQU PAGE
1
- ,
, OUESTION 6.08 (2.50) c.
WHY is an EOC RPT necessary? (1.0) b.
WHAT three (3) automatic signals will cause these breakers to opon? INCLUDE setpoints.
(1.5) OUESTION 6.09 (3.00) The following concerns the Reactor Water Cleanup System.
c.
LIST the (hvo (5) RWCU isolations.
INCLUDE setpoints if applicablo.
(1.5) b.
STATE the two (2) conditions (notpoints not required) which will cause automatic closure of the blowdown valve (HCV-OO4) and EXPLAIN WHY nach is nended.
(1.0) c.
WHY tu it nec est.ar y to have a dominera112er holding pump? (0.5) DUESTION 6.10 (2.00) , Concerning the ADS initiation logica c.
With the plant opnrating normally at 100*/. power, the channel At and channel A2 manual initiation pushbuttons are rotated and depresund.
WILL the ADS function occur? WHY7 (1.0) b.
If ADS nu manually int t n at ed (and all initiation conditions arn mot), WILL thu ADG valve opening be delayed by 105 seconds? (YEG/NO) (0.5) c.
If an ADS blowdown is in progrous, with all initiation signals still present, WILL deprosuing the ADG logic reset pushbutton switches reinitialize the 105-second timor and cluso all ADG valvos? (YEG/NO) (0.5) OUESTION 6.11 (2.00) LIST in order of occurrence the four (4) automatic actions that occur as a direct result of a ducreasing condonsor vacuum.
(2.C) (**884 END OF CATEGORY 06 **sse) - - -
'_Z2__EBOCED96ES_:_N9BdeL1_8BN9BbeLi_EBEBDENGy_9ND PAGE
BeDIDL991996 99NISQL
. ,
. QUESTION 7.01 (2.50) The reactor is operating at rated conditions when you recieve the "RBCLCW HD TK A(B) LEV LO-LO" AND "RBCLCW SYSTEM A(B) HDR PRESS LD" alarms. What are three immediate actions per the cmergency procedure 29.017.017 Briefly explain "why" for each of the above actions.
(2.50) DUESTION 7.02 (3.00) c.
STATE the five (5) Immediate Actions required per SP 29.010.01, Emergency Shutdown Procedure.
(2.5) b.
WHAT in the preferrod heat sink for the reactor and WHY7 (0.5) DUESTION 7.03 (2.00) Assume the set of indications listed below exist following a valid LOCA.
STATE whether or not adequate core cooling can be assured.
JUSTIFY your answor.
(2.0) HPCI han ISOLATED due to low steam supply pressure.
- All reactor water levut instruments are off-scale LOW.
- , with thu exenption of the fuel reno instrument which is off-scale HIGH.
, l ALL core spray pumpn havn started, subsequently tripped - c on overload, and CANNOT be restarted.
RHR pump 'A' is RUNNING with a path to the RPV in.jection - (minimum flow valve closed in loop 'A', indicated flow 10,000 gpm).
All other RHR pumps have FAILED to start.
l i (see88 CATEGORY 07 CONTINUED ON NEXT PAGE tette) , > l i f I
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_ - _ - -. -- ___________ - _ Z,. _ _ E h 0 C E Q u B E E _ : _ N Q 8 d h _ 6 k N Q 8 d @ ( g _ E D E B Q E N G L @ N Q PAGE
8601QLQQ1G06_GQN18Q(
- -
-
- .
QUESTION 7.04 (1.00) During operation at 100% power, an explosion and subsequent fire in l the control room has made evacuation mandatory.
The reactor was cerammed from the control room and the Remote Shutdown Transfer i
j Switches have been placed in EMERG.
Under these conditions, WHAT special RCIC precautions must the operator observe if RPV water ' lovel is being maintained using the RCIC system 7 (1.0) l ' ! r i OUESTION 7.05 (2.00) [ Concerning SP 29.023.03, Containment Control Emergency Procedure.
l l e.
LIST ALL the entry condtttone for SP 29.023.03.
(1.5) i b.
In WHAT order are the indicated procedural paths to bo I performud? (0.5) J r i I j - i OUESTION 7.06 (2.00) ! l STATE the basin for the following conditions botng requirud prior ! l to placing the Ruactor Modo Gwitch to the RUN position.
t ' c.
main stemmitno prwsuure nteady at approximately 920 psig (0.67) i . b.
all APRM nndicationu '45% powur (0.67) i r.
al1 AFRM downntalo 11ghta putingutuhnd (O.67) j i < , ! I j DUESTION 7.07 (1.50) { l Concerning the main turbines i ! c.
While performing a turbine hwatup, WHICH HP turbine componwnt I will hwat up fasters thw rotor or the shw117 (0.5) i ' j b.
WHEN are the Low Prwssurw turbines warmed up? (0.51 l l c.
WHAT lube oil pump supplies bwaring ont pressure during [ l turblue operation? (O.D) ! i l I ? < t
!
(..... Cnrenony 07 CoNriNuco nN scx1 PAac .....) i
) i L.- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
Z.__E60GEQUBES_:_NQBM6Lt_8@NQBt}@(t_Et!EBQENQL9ND PAGE
- B001960Q1G96_GQN18QL
, QUESTION 7.08 (3.00) For each of thw following statements regarding the High Pressure Coolant Injection System (HPCI), INDICATE whether the statement in TRUE or FALSE, and EXPLAIN your answer.
c.
In the event Low Pump Suction Pressure is sensed during HPCI system operation, the turbine will trip, and the signal must be manually reset before the turbine will restart, if initiation signals are ath11 present.
(1.0) b.
Upon a HPCI system isolation, due to Low Steam Pressure, the system cannot restart untti the pressure rises above the tantatkon metpoint and the isolation signal to reunt.
(1.0) c.
If the HPCI turbine tripe due to an overspeed condition, it will restart whnn thu upnad coasts down to between 3000 and 4000 RPM.
(1.0) QUESTION 7.09 (1.50) BTATE the throne (3) condittent under which rapid RPV deprus-curization tu warranted.
(1.5) DUCGTION 7.10 (2.50) a.
WHAT is thw basts for the. CAUTION in GP 29.023.02, "Couldown Emergency Proceduro", which states "DO NOT throttle HPCI or RCIC turbine below 2200 rpm"? (1.b> b.
GIVE two (2) control room indications that thw operator may usw to detarmine if a dangeraus condition mtght ex1et dup to operation of wither turbinw below 2200 RPM.
(1.0) DUEGTION 7.11 (1.00) According to GP 23.604.01, "APRM Gystem", WHEN is an APRM channel d3 fined as inoperable based on LPRM inputs? (1.0) {seeea CATCHORY 07 CONTINUCD ON NEXT PAGE 88884) . -
- _ _ _ _ _ _ Zi__E69CEDUBES_:_NDBdOLi_81 UC00062_EMEBDEUGY_899 PAGE
- B991DL991GOL_G991896
- ~ . OUESTION 7.12 (2.00) S {ea Wt-According to SP 23.116.01, " Main and Auxiliary Sy' stem", there are ntno (9) indications in the control room that can be used to verify that an SRV ts open.
LIST eight (8) of these nine indications.
(2.0) DUESTION 7.13 (1.00) WHICH of t her f ollowing condition s would provent RCIC f rom cutomatically knitiating and in)ecting into the reactor? (1.0) (n.) A prc=vious high wat:4r 1r vel sptLo to +62.5".
(b.) Thee f(CIC vacuum pump falls to start.
(c.) Low suppresston pool I n v e= 1.
(d.) E rt ha u n t lino inolation valvo is not full open.
i (esses I:ND OF CATEGORY 07 stess) - - - - - - - - - - - - - - - - - - - - - - - -
, St_18DUINISIBBIlYE_EBDGEDDBESs_CDNDIIIDNSa_8ND_LIN1IBIIDNS PAGE
.
. DOESTION S.01 (2.00) U n the attached Technical Specifications to answer the following question.
Dimmat Generator 102 and RBSW pump D are out of serview.
Thw load dtspatcher has ordered the plant to remain at 100% power for the nsxt 6 weeks.
Can the plant comply with this request without repairing the out of service equipment or being in violation of Technical Specifications. (YES/NO)? EXPLAIN your answer using Technkcal Specifications provided.
(Referenew all sectkons and requi rement s. ) (2.0) DUESTION O.02 (2.50) Une thee attached Technical Spect f s cat t ons to ANGWER the f ollowino question.
During a uhtit turnover, with the plant opurat.ing at 75% power. the SRO to informed that the MSIV Closurw Time Gurveillance Tent EXEEEDED the maximum allowable watwnston interval yesterday, and 30 it's bown schedulett for hta nhaft t ott a v. Power is maintalned at 75% whilu the tunt t u contfuctetti enti, following comp!wtton, the NGO reports that all MSIVs have mot thwir speci f t ml closing times.
a.
In tht e sat tuf ectory t o resolve the situation (YES/NO)? (0.5) b.
EXPLAIN your answur and REFERENCE all swetions used to tievnt op your answnr.
(0.0) (eeeee CATE00HY 00 CONTINUED UN NEXf PAGE eeeee) _ .
D __Olldlu1 SIB 011YC_EBOGEQUBEQu_GQUQlllQU$t_GUQ_Lidl1611QUS PAGE
' .
. OUESTION O.03 (2.50) Indicate whether the following ntatements are TRUE or FALGE : a.
A Watch Engineer may concurrently fill the posttton of the STA whhle on watch.
( ti. 5 ) b.
Thu Fire brigado must includo at luant an SRO, GTA, and RO.
(0,5) c.
All core al terat t ons must be directly nupurvised by a 1trenued Gentor Op str a t or (or Guntor Operator Lamttod to Fool Handl1ng).
(O.5) d.
A Rnactor Opuratorn l i c on em in rnquired to perform a corp .il t or a t 1 on.
(O.5) p.
Du r i nij Operat1onal Cond1t on 4 or G.
an indavidual wIth a val 1at Runttor Oper at or t 11 core," may be denionatud to act.eme the control room command f ont t t on during an atsuunte of thn c Watch Engtonor from thn caritrol room.
(0.5) L (0ttt4 CAICO()hY 00 CONTINUED UN NCXi l' AGE 4eet8) _
l.
~ ~ . , ! . E's.__ettd1W1EIBel1YE_EBQGEQUBE5s._GQNQ1Il0NSa._0601101IGI1QNS PAGE
' l* . . r
. 00ESTION 8.04 (2.50) [ \\ ! Indi te (YES/NO) wether each of the f ollowing is considered a " core ! ottera on" per the Shoreham Tech Specs: '
c.
Wi thtieawal and insertion of an SRM detector to check its drive mentor during ref ueling.
(0.5)
\\ h.
Removal of an LPRM string for replacement.
(0.5) ! \\ c.
Removal of an lancoupled control rod for replacement.
(0.5) x d.
Removal of a contry rod's position indicator probe for repair.
(0.5) e.
Control rod withdrawat htyd insertion to test the posttion j indicator probe during a refueling outage.
/ (0.5) i l $ ',0 z .. ~ @ k [a.SWER , VEG \\, l ' ! ' b.
YEU i# ()( [ y , I c 6' ' I.]'#'1 A c.
YE0 , ! ' J d.
NO ' , , M C ' e.
VED (1 g/ j s ! \\ [ C+0.53 nach , \\ \\ ! OTE: key factor in answurs in venael head removed.'i,N ) I - ' ~ ~ ~ -+ ~.... . .. .. _ _... _ _ ___ - DUEDTION U.05 (2.00) f I a.
A temporary procedurI, thonge that detalla thn changwu made on j the att.achett UP 23.202.01-13 is presented to you for approval.
l WOULD you approve this change (YEG/NO)? (0.5) t h.
EXPLAIN your answer to "a.". (1.5) , i l I t ! (t e s s e CAf t00ftY 00 CONTINUED ON NEX T f%00 8 0 8 0 0 ) , l l l Y
r
8.2 Abnoras! Performance . g , 8.2.1 Manu2l 1:Atintion
. . 8.2.2 Terbine Trip
- s y 8.2.3 Syor.es Isolation
e 4.2.4 operation with 1moperable Componente
, 8.2.5 operation During Loos of Station Air '
8.2.6 Operation Durias Loos of Normal Power
8.2.7 Suppression Pool Makeup from CST
8.2.8 Leak Detection and Isolation
8.2.9 Alara Response Procedures
Appendix 12.1 Frerequisite Checklist . Appendix 12.2 Valve Lineup Checklist Appendix 12.3 System Component Power supply Checklist Appendix 12.4 Terry Turbine Oil Piples Diagree - 65,998-C 4.0 FRECAUTIONS 4.1 During normal reactor operation, the NPCI stese line contains approximately 1000 pois radioactive stese. Caution should be used when working near the NPCI stese line. Observe all standard radiological presautions.
. 4.2 The NPCI steam line should be warned up and pressurised slowly using the bypese lines to avoid subjecting the piping to unnecessary stresses when opening the,stens line isolation valves.
' 4.3 The NPCI turbine le not rated or intended to operate at low speeds. During operation, the turbine should run at speeds greater than.2200 rya.
4.4 The Barometric condenser and Vacuus Tank (*g-036) must have a normal wate.
level before operating the Vacuus Fusp (*F-074) or damage may result to the pump.
4.3 011 tenparature should be greater than 60'F prior to turbine operation.
Reduction seer bearing oil should not onceed 185'F booster pump bearing oil should not exceed 165'F1 turbine bearing oil should not onceed 160'P.
011 inlet temperature to the main pu6p bearing should not onceed 140'F.
4.6 Steady state bearing vibration should be lietted to less than 3 elles peak - - - -.
^ (, l to peak.
, 4.7 Operation with discharge flow less than 575 spe will cause discharse of CST water to the Suppression Chamber via Manteve Flow Valve *MOV-036.
4.8 It la easy to induce flashing, water hammer, ever pressurisation, overheating and pump runout.
Entraordinary care shall be taken to prevent the escurrence of these conditione during operatione of this system.
3.0 PAEREQUISITES ' 3.1 Prerequisites are delineated on SPF 23.202.01-1, (Appendia 12.1).
' . , , SP 23.202.01 Rev. 13 page 3 , ,
.. . Efr._201)d1NI DIBOI1 YE_ E B DGE DUBE Di_COND1110N3s _0ND_ L.101 IGI1ONS PAGE
.
. OUESTION O.06 (2.00) Une the applicab!te Technt cal Specificat1ons appended ior reieronce.
Following a TIP trato, the "C" TIP ball valvo did not auto clonw.
I 1hn TIP machina requires unveral "joggs" to close this valve due to e aticking "in-shield" 11mtt uwatch.
The ball valve in subsuquently closed.
a.
GTATE whetht*r Primary Containment integrity is being natisfied.
(0.5) b.
EXPLAIN the bagno for your ductaton INCLUDING all applicablo Techntcal UpochfIcations nection nom. and (ti any) required f011ow up act1one.
(l.M) DUFGTION 0.07 (2.00) INDIC All: wha t ht*r t he-fonIc3w1no %tatemonta aro liiUL or FALLC.
- c.
Wh t l es Stuam Cooling the roactor per procudurst 29.023.04 reactor preensure strops bnlow 70d pen g.
Ihn operator will het directed to deproueurtees t hes hFN peer procedurn 29.023.00 re gar dleu% of the evailahtitty of coolant in)nction.
40.b) b.
"Alternatn lei.)wetton Uubsystomn" includn CDLC and CRD.
(0.5) c.
If t he nuppression chambor pronture uncuerda primary containmeent preunuro 1imit, thes opor ator may vnnt thn containmtent provn deed that radtnactivnty rolpated to unrent r t et ed artent. will bo Iphs than the, limitt, of IOCFR20.
(0.5) d.
Ghoold thee contatnment watwe I te v e* 1 s*w cnad 56 ( west. thu oporator must twrminato injuct n on f rom unurenn autun do thes pr1 mary containmont irrumptsettvo of adsequato corn cool 1ng.
(0.5) 0000T10N 11. Oil (l.50) liNICI:LY DEUCHiltC how the operator can conf trm Primary Containment watwr levtil wt thout t het uma of control room Invwl indications.
(1.51 ( e 4 e e e Call UUhY Oil CONT INUCD UN NCX ! PAUC 4 s 4 4 e )
B*i_ '8D51NISI6eIIYE_EBDGE DUBEDi_G9ND1I1DNSi_eND_L.10119IIDND PAGE
l' .
, DOES T ION B.09 (2.50) During thw first startup following a refueling, the reactor opwrator has a power indication of B'/. on all APRM's, and is ro.edy to placw thw made swttch in the RUN position.
At this %
point the SRO is alweted to the possibility that a step may havw been mtsmed in thu prwcewdtng CSS survwillance.
Based on the Technical Spwetfication determination that whth one CBG inopwrativo a 7-day LCO to in offeet, hw elwets to continum.
IG this the proppr course of action? (YES/NO).
EXPLAIN your answer and includo all applicabtw Twchnical Specification cwction numbers.
(2.5) OtlFGTION 11. 1 O (3.00) Using the attached Technical Spectitcations. bHIELY DISCUGS thu acthonu which must bn taken and REFERENCE all Technical ,. 14pec t i t eat i uns wht c h or p. app ! t c eti t p t a t h t t.
camp.
The plant in curront1y oporatIng at rated condttkons.
LPCI/ Roc 1rculation swlno hus "A" ta downpronaed for manntynanew and will not be rwatored for approntmately 3 days.
HPCI surventlanew has.)ust rurwntly brun completod.
It wati discoverwd that the HPC! pump suetion valvw 10410MOV ** motor opwrator did not operatw prrippr1y and thp valve * had ta liw nporated using thw 1ocal manual oporator.
Ihn val vas n% currentiv >ppn.
(3.0) o 3'l a kHyd CA.
t t *4 M-R DurGT10tl D.1I (0.00) During a backshtit a broach in the primary containment has occurred end the Watch Engtnwwr na not avatlab1w.
Per thw Emwrowncy Plan.
a.
HOW and WHAT two (2) organnzattuna must he nottined? (2.0) h.
WHO 6m rweponstbin for notifyhng them7 (0.5) (totot L N D O F C A f L O OftY Oil s t e t t ) (ele t tle t t e t t e IND Of' CX AMIMil0N t e t t e t t e t e t e t t e)
_ _ _.- _ _ ___ _ __ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _. _ _ _. L.__IHEQ6Y_QE_NUGLEeB_EQNE8_ELONI_QEEB011QNt_ELU100t_eNQ PAGE
IUE8000YNed1GQ
ANSWEf(S -- SHOREHAM-86/12/10-JARRELL, D.
- . ,
l
1 i AN5WER 5.01 (2.50) ]
1
c.
295 dog Fi assume isenthalptc expanston to atmospheric pressurw.
, (290 > T > 300 for +0.75 300 > T > 305 or 285 > T < 290 for [ q ' +0.5; wrong if > 10 dag F from stated answer) [ I f b.
Increase C+0.53 c.
Remains the same [+0.53 a c44p t duc M 8-4.
f
d.
450 pata C+0.753 wrong if > 500 or < 400 REFERENCE l [ l j 1.
Ghorwham Nuctwar Powier Gtatton Thermo Modulw, Be.ctton 2.3.
Cwamplw 2.4.
l
L AN2WER 5.02 (1.00) [ ! t (b.)
(timw = Y. pawnr = 0.4 hrs) [+1.03 [ , l l<CFEliENCE
, i l 1.
Chorwhom Nuctwar Power utatiuns i<oactor Phystcs Modulw.
l p.
7-104.
! ! !
AN!iWl'R 5.03 (2.00) , ! ! l o.
thw fuwt (dopptwr) eneifietwnt L+0.53 l T ! ! h.
Duw to (thw t imo constant of the f uwl--generally 5-7 seconds) { ! thw amount of time required f or hwat to be transf wered f rom j ! the fuel to the coulant (+0.53.
The modwrator tempwratury or
l thw vond cowffickwnt would not havw any effect for swveral ! seconds t40.51.
The fuwt temp cowffIctwnt acts instantly tu ! insert negativu rwacttvity (+0.D3.
j
i i CEFERENCE ! ! f l 1.
INEL IIwam Hank, question D743.
[ l f ! 2.
Dhornham Nuclear Power Utat*9ns Hwactor Physics Module,
l Dwetton b.2 (Dopplwr broadwning). Hwat Trannfwr Module, f l Uwetton 7.2 (femperature Proftion).
! ! ! , i l
Ut__10EQBY_QE_UUGLC66_EQWEB_EL8UI_QEEB81100t_ELU10St_8UQ PAGE
IUEBdQQYN6dlGQ ' - ANSWERS -- SHOREHAM-86/12/15-JARREL.L, D.
. ANSWER 5.04 (1.00) (u.)
(+1.03 REFERENCE 1.
Shoreham Nuclear Power Stations Ruactor Physics Moduto, p.
7-10.
ANUWER D.05 (3.00) c.
3.
C+1.01 b.
timal I er [+0.251 alpha ( anod ) becomes lens negative f+0.251 alpha (void) become's i cou nirgeit t va, L+t).251 alpha (fuel) bocurnno moru nwgat ivo L+0.253 c.
Smaller [+0.251 alpha ( rnori) bocomns lose negat t yp L+0.25] da s,ih ra ( vol ti) tutcomon 1939 nngattvu [+0.253 alpha (fup)) becomur. more negativp (+0.253 ftLf?ERENCE 1.
fihoresham Nuttnar Powpr fit a t t on1 Runetor PhynIcw Modulp.
pp. 7-102 7 - 1814, 7-106 Anti 7-101).
. _
. Ut__INEQBY_QE_NUGLE88 EQWEB_Eb8MI_QEEB@IlQNt_E(y1Qgz_@NQ PAGE
- ItEBdQQYN@tllGQ
ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
e ANSWER 5.06 (2.00) c.
In the region where temperature of the coolant remaans relatively constant, the coolant is at saturated conditions, the addition of heat to the coolant in this region increases the quality of the steam but not the temperature because the coolant remains at the saturation temperature.
C+1.O] b.
Enthalpy changes continuously over the flow path because the water continues to absorb energy from the fuel.
The rate of wnergy absorption is proportional to the heat flux (neutron flux).
Since the flux in highest near the center of the core.
the slope t o greatent at this point.
C+1.03 REFERFNCF 1.
Shoreham Nuciper Power Station Heat Transfwr and Thermadynamics Modulo, pp. 4-121 and 4-125.
1 00 ANSWER 5.07 " "'d ? z:= 2 - - $hb< s N k% C**
- r DL LtM t
b.
Increase c.
Incrwamp d.
Dwerpano p.
No Effwet C+0.53 pach Al:FERENCF.
1.
Ghorwham Nuctuar Power Gtations Reactor Physics Module, p.
202-215.
ANSWCH 5.00 (1.50) o.
Edge rods (+0.53.
Xenon polmoning is highwat whern thw highwat fluw was, in the center of the core C+0.51.
h.
Hecaumw thw h10h difforwntial notch rod worth of edom roda can cmunw short parlods C+0.5. Et__IMEQBY_QE_NWCLEBB_EQWEB_Eb8N1_QEE8611QNt_E(Q1QQg_@NQ PAGE
INEBtjQQyNet! LGS - - ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
. REFERENCE 1.
Shoreham Nuclear Power Station: Reactor Physics Module, p.
7-153.
ANSWER 5.09 (2.50) c.
t/T 1800/60
P=Pe = 1 watt e sec = 1.06 x 10 watt o
10 Mw [+0.53 = No [+0.53 Moderator (and fuel) coefficient would turn it [+0.53
\\, Scru m o t-b.
No [+0.53 Power at end of 30 minutes would exceed maximum MWT for Shoreham (result of equation for part a) [+0.53 REFERENCE 1.
Shoreham Nuclear Power Station Reactor Physics Module, Lesson 15, Trancient Reactor Response, p.
7-226, 2.
Equation Sheet.
ANSWER 5.10 (3.00) n.
steam is returned to saturation in the moisture separators, located between the high and low pressure turbines [+1.03 b.
steam is reheated in the reheater section to superheated steam, which is directed to the low pressure turbine [+1.03 c.
subcooled inlet fluid receives reactor heat and becomes saturated [+1.03 REFERENCE 1.
Shoreham Nuclear Power Station: Heat Transfer and Thermodynamics Module, p.
4-12. __ . . . _. _ __ __.
Ut__ISEQBY_QF NQGLE@8_EQWEB_E(@NI_QPEB911QN _ELUlQ@t_@NQ PAGE
t ISE85QQYN@dlGE ' . ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
, e 's , ANSWER 5.11 (3.00) . c.
Increase [+0.253 b.
Increase [+0.253 c.
Increase [+0.253 d.
RTS [+0.253 The delta-T required to transfer the heat from the pellet to the coolant must be increased due to the increased thermal resistance of the corrosion layer.
This accounts for the increase in fuel, grp, and clad temperatures (see figure) [+1.53.
The midplane bulk coolant temperature would remain the same since it is a function of caturation pressure which is controlled by EHC [+0.53.
,
- r i d ! REFERENCE ! 1.
Shoreham Nuclear Power Station: Heat Transfer and Module, Section 7.2 (Temperature Profile), Section 8.1, 8.2, 8.3, and
8.4 (Thermal Characteristics).
i , f
4 .- -.- -. -. --- - -- - - . - - - .. - - - -
y J L + sL.
' Dz__IBE98Y_9F Nyg6 EAR _EQWEB_E68NI_QEg8AllgN _E691pS _8ND PAGE
2
IBEBU99YN8 digs
-- . ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
. t & ANSWER 5.12 (1.00) (c.)
[+1.03 REFERENCE 1.
Shoreham Nuclear Power Station: Heat Transfer and Thermodynamics Module, p.
4-32.
' 2.
Steam Tables.
, d I ,
.-,w~, . .-, - -. ~ , .-,-..,- -., .,.,wm.,ng,e ,, ,, ,,g, , -,,., -, -,,.,,, -,
6 __ELON1_@Y@lEMS_QE@l@N _CQNIRQL _@NQ_lN@lRQMENI@llQN PAGE
t z t . ANSWERS - SHOREHAM-86/12/15-JARRELL, D.
- . ANSWER 6.01 (2.50) a.
Close C+0.53
- e cbay it il d e==5r"mtd-b.
1.
Increase C+0.53 er 2.
No change C+0.53 3.
No change C+O.53 c.
Pressures greater than 1510 psig may cause rod drive mechanism damage during a scram.
[+0.53 REFERENCE 1.
Shoreham Nuclear Power Station: SP No. 23.106.01 Control Rod Drive, Section 4.1, HL 106 SH, Learning Objectives BLF.
ANSWER 6.02 (2.50) a.
In the SRM and IRM range the gamma radiation is not proportional to power and must~be eliminated from the power reading C+0.53.
In the'APRM range the gamma radiation is proportional to power and thus need not be eliminated [+1.03.
b.
SRM = pulse height discrimination (neutron pulse is greater than gamma) [+0.53 IRM = squaring circuit or cambelling (square of neutron signal is greater than gamma) [+0.53 REFERENCE ! .1.
Shoreham Nuclear Power Station: HL 601 SH, Section 8.2.3.6 and HL 602 SH, Section 8.2.1.2.
l I i , l l i _
- _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ si__ELONI_Qy@lEUS_QEQ1GMt_CQMIBQLt_@NQ_lNSIBQMENI@llgN PAGE
. ANSWERS -- SHOREHAM-86/12/15-JARRELL.
D.
. . frten hr T vi re.
nu ANSWER 6.03 (2.50) Hov-05I o,go.
Eeer253andfmin.
h flow valve a.
HPCI turb'ne_stop valve (HUV U"2) M hi vessel level (8); (+56.5 in.)
(MOV 036) ( o t/- 15)at hi5m bow d (n3<c taio vs\\ve abo c \\ese s C+0.53 b.
Yes (auto restart) C+0.53 at low-low vessel level 2 (-38 in.)
[+0.53 c.
No C+0.53
REFERENCE 1.
Shoreham Nuclear Power Station: HL-202-SH, High Pressure Coolant Injection System Objective 2,5,7 and HL 621 SH, Appendix 1.
= ANSWER 6.04 (3.00) a.
control valves close 5% C+0.353, bypass valves open 5% C+0.43 b.
control valves close 5% C+0.353, (reactor scram probable due to increasing pressure because) bypass valves will not open C+0.43 c.
(regulator "B" controls at 100% power with) CV at 100% [+0.353 (reactor pressure is 10 psi higher), bypass remains closed C+0.43 d.
control valves close to approximately 88% E+0.353 (to maintain Rx pressure at 920 psig) while approximately 12% of flow goes through bypass [+0.43 REFERENCE 1.
Shoreham Nuclear Power Station: HL 127 SH, Main Turbine; HL 651 SH, Reactor Pressure Control (EHC), Objective 2; and HL 116 SH, Main and Auxiliary Steam, Objectives F&G.
_ _
6 __ELONT_SYSIggg_DEg1GN2_CgNIBgL _8ND_INSIBgdENIBI1ON PAGE
2
. ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
- . ANSWER 6.05 (1.50) c.
1.
prevent inadvertent draining of the vessel [+0.53 2.
prevent exceeding RHR design pressure [+0.53 b.
False [+0.53 REFERENCE 1.
Shoreham Nuclear Power Station: HL 121 SH, Section III C.3.b,d and e.
ANSWER 6.06 (2.00) o.
Yes [+0.53 b.
The condenser will fill with condensate [+0.5] resulting in radioactive steam leaking from the turbine seals [+1.03.
Alternate answer: vacuum pump damage (other justified answers should be considered by the grader) REFERENCE 1.
Shoreham Nuclear Power Station: HL-202-SH, High Pressure Coolant Injection, Section 8.3.2.j,k; 8.3.4.4b, and 8.3.6.4.
H L - 3 o7 - s eq .o ANSWER 6.07 (1.50) f h.T in(R asg7 pos d.'co Gwh en s dd-no &_ _nt;.-11, plere the emat-el e'it-" te Decr' ,0. ;I ..J place the control switch in START E:0.31 and hold it there for at least 3 ceconds EtO.53.
t o W5 +0 1S REFERENCE 1.
Shoreham Nuclear Power Station: HL-307-SH, Emergency Diesel Generators Objective 7.3.2, Section 8.3.3.8a.
, I l i i
~6t__Eb8N1_SYSIEMS_QESIGN _CQNIBQ6t_@NQ_lNS18QUENI@I1QN PAGE
t - . ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
' . ib <<S** f,g,,4 R". ANSWER 6.08 (2.50)
c a.
At the End-of-Cycle, a pressure increase transient from a L.u ",:r g ". e turbine trip could add positive reactivity faster than the,dU @g{ kd ' control rods can insert negative reactivity.
Tripping the
recirculation pumps helps by rapidly adding negative reactivity. C+1.03 b.
1.
turbine trip, power > 30% 2.
reactor water level 2 (-38 in.)
3. - high reactor pressure - psig h h 3,3,t.{.{ ~ 2 [+0.53 each REFERENCE 1.
Shoreham Nuclear Power Station: HL 658-SH, Reactor Recirculation Flow Control, p.
5.
2.
Shoreham Nuclear Power Station: HL 705 SH, p.
5 (setpoints).
3.
Shoreham Nuclear Power Station: HL 312 SH, p.
6 (setpoints).
ANSWER 6.09 (3.00) IN . a.
1.
Rx water level 2; -38" S t eg vin cI 2.
non-regen Hx high outlet temp; 140 deg F cor r eJk 4 65 "*"A - 3.
SBLC initiation 4.
RWCU area high temp; 155 deg F p, '{p (3 ,c3 5.
main stea tunnel high temp; 175 deg F McroVAk
b Bow h SH 3PM
C+0.33 each, +0.15 for setpoint, +0.15 for parameter where applicable b.
1.
Iow upstream pressure (5 psig) - prevents draining of system when isolation valves are closed [+0.53 2.
Hi pressure in blowdown line (140 psig) - prevents rupture of LP piping if blowdown flow is blocked [+0.53 c.
to prevent filter resin cake from falling off the filter tubes (under low flow condition) E+0.53 REFERENCE 1.
Shoreham Nuclear Power Station: HL 709 SH, Objectives 5 and 6,
b __ELONI_$YSIgdS_gE@l@N2_CQNIBQ62_@Ng_INSIBQNENI@IlgN PAGE
. , ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
, 'SP 23.709.01.
ANSWER 6.10 (2.00) c.
No C+0.53.
The RHR/CS pump discharge is not above setpoint (RHR/CS pump not operating) [+0.53.
, b.
No ~E+0.53 c.
Yes C+O.63 REFERENCE 1.
Shoreham Nuclear Power Station: HL 201 SH, Appendix 0, Objectives 5 and 3 (also logic 'B' diagram).
ANSWER 6.11 (2.00) 1.
condenser low vacuum alarm 2.
turbine trip (reector scram if P > 30%) 3.
feed pump trip 4.
bypass valve closure (#s 4 and 5--same trip point) 5.
MSIV closure Any four (4) 0+0.53 each, +2.0 maximum.
REFERENCE r 1.
Shoreham Nuclear Power Station: HL 127 SH, Main Turbine, , Objective E'& F, SP 29.012.01, Loss of Condenser Vacuum.
i !
. - --- Zu__EBQGEDUBEq_ _NQBd@Lt_GBNQEd66t_EME8GENQY_@ND PAGE
.. BOD 106901GOL_GQNIBQL - ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
, k.a((5 ANSWER 7.01 (2.50) d dI g k 1.
Trip the RWCU pump and isolate the system by closing 3 ad g 0 p, , , (1G33*MOV-033 and 034). C+0.53 The RWCU should be shutdown A' (, e#.,, before the RWCU system trips on high temperature.
[+0.53.
zg# / 'h$ # e' M C 2.
Reduce Reactor Recirc. speed to minimum. C+0.53 To preclude
$E P' g,0 damage, the Recire. Pumps should be shutdown within ten , minutes after loss of RBCLCW C+0.53.
3.
If the RBCLCW system spilts and isolates, trip the Recirc. MG sets, initiate the EMERGENCY SHUTDOWN procedure SP 29.010.01 C+0.153 and trip the CRD pumps after all of the control rods are verified to be inserted. C+0.153 Action is required to minimize potential for equipment damage.E.23 REFERENCE 1.
Shoreham Nuclear Power Station: SP 29.017.01, LOSS OF REACTOR BUILDING CLOSED LOOP COOLING WATER.
ANSWER 7.02 (3.00) m.
1.
mode switch in shutdown 7.
verify rapid flux decrease 3.
verify all rods inserted 4.
monitor RPV water level and pressure 5.
monitor containment paramet>ers . [+0.53 each b.
The main condenser [+0.253, it preserves the energy absorption capability of the suppression pool should it be needed [+0.253.
! '
REFERENCE 1.
Shoreham Nuclear Power Station: SP 29.010.01, Emergency Shutdtyn Procedure, Sections 3.0 and 6.6.
s a . .. -.. - - - - - - -.
-Zt__E80GEQQBES_ _NQBd6Lt_6BNQBd@Lu_GUEBQENQY_@NQ PAGE
. 800lO690LG86_G9NIBQL - ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
. s e.
C5 6d Y ' ANSWER 7.03 (2.00), n ot cod t c" * T be c'35 c ore @d 7_,oie reu dth y 4'E "I M Adequate core cooling islassured C+0.53 " ' * acle g vw For cov e - cer-2y S; w,cc scree, ;i,,_ _t :. ,t u,,; rcce g,;p ve ptam a [ Err - p~' p u.a
- ,7
- n '^*', then t ',2 d;finiti,n M I [or c ovt 1-e inJ : i n;; ;.. LTr t he cur e
- c iuct cor casling as T,a t.
trooi 2 ;,; ;g ,,, d
- f 1_ __,_ ; ; _ ;,, g, M'
k&/ "'O br
- c-eLed 1t suppur ied.)
Et1 53 <i k t s b c. c V eTu '*Nd L Leb Yeuc.he S bre ett m.t vvs Fe < = c ste m, t,2, 3 on eye u.
REFERENCE oF = 1.
Shoreham Nuclear Power Station: HL-980-SH, Lesson 3, Recognizing Adequate Core Cooling, Section 3.3, pp. 19 through 22.
ANSWER 7.04 (1.00) Trip functions and interlocks are bypassed when systems are operated from the RSP.
(Specific trip and interlock functions may be accepted if complete.)
[+1.03 REFERENCE 1.
Shoreham Nuclear Power Station: SP 29.022.01, Shutdown from Outside Control Room, Section 6.3.
ANSWER 7.05 (2.00) a.
1.
SP temp > 90 deg F 2.
average DW temp > 145 deg F 3.
DW press > 1.69 psig 4.
SP level >+6" 5.
SP level < -6" [+0.33 each, +1.5 maximum b.
perform all sections concurrently C+0.53 REFERENCE 1.
Shoreham Nuclear Power Station: HL-944-SH1, pp.
4, B, and 10.
_ _
- - = -. -= - . - 12t__PBQQEQQBES_;_NQBM9L _6BNQBM862_EME89ENgy_8ND PAGE
2 B89196991986_G9NIBQL
- ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
.. ANSWER 7.06 (2.00) 1.
Ensures applicability of GEXL correlation (thereby allowing full power operation - alternately it ensures a MCPR greater than 1.06) [+0.673.
2.
Shows NI overlap (operability of APRMs) and allows clearing of (u t* ~ w/insAM2-Hs opube) (ah[+rokq73plockt -funk;on ) APRM downscale alarms O 3.
Ensures all channels indicate,>57. p,ower (showing operability - allows alarm to indicate an abnormal power operation condition) [+0.673.
REFERENCE 1.
Shoreham Nuclear Power Station: SP 22.001.01, Start-up Cold i Shutdown to 20 Percent, pp. 10 through 11.
2.
Shoreham Nucl ear Power Station: Technical Specifications, Bases B 2-1.
ANSWER 7.07 (1.50) a.
-the rotor C+0.53 b.
as the turbine is loaded C+0.53 c.
the main shaft oil pump C+O.53 each I REFERENCE 1.
Shoreham Nuclear Power Station: HL 127 SH (Main Turbine).
, . ,,, - - - - ~ - -,,,,, , -., -, - -, -.. -.,,, -,. -.,,,. - - -,, ., -,.. -. _,. -,,. - - -. -, , _.,..
-_ - _ -. _- 2 __EB9GEQUBgS_r_NgBd892_6pNgBd8L _gdEBGENgy_@ND PAGE
1 -. 8691969 GIG 66_G9NIBQL - ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
. e , ANSWER 7.08 (3.00) a.
False [+0.53.
Once the low suction pressure signal is clear the turbine will auto restart if the initiation signals are A still present [+0.53.
s . inP. in pUe4 hyI b.
True [+0.53.
The low steam pressure auto isolation signal "Y" seal s in, and must be manually reset (using the AUTO ISOLATION SIGNAL RESET pushbuttons on the *PNL-601 after the reason for the isolation has been determined and correcte [+0.53.
c.
True [+0.53.
The oil pressure will be restored when the turbine coasts down, thereby causing the stop valve to opgn (">'co Mptl turbsa ir ( p3 [+0.53.
cab o ucct pT its< I t % T u dc. m a b m k 4e y REFERENCE l 1.
Shor eh arn NutI c>ar Powrir Stat:en: HL 202 SH (HPCI)-SLO 6.7; SP 23.202.01 (HPCIS), Rev. 13.
4*
. ANSWER 7.09 (1.50) . -. c-A 4 .s e,%m 6he*p C o ^ T'
1.
Primary containment is threatened. C+0.53 (.9 Ecm Ley k b "d.
2.
No high pressure systems are available to inject water to the '" RPV.
[+0.53 av) nisk b n/ yri l< A onbs E
3.
Necessary to allow steam flow past the f uel bundles.
[+0.53 e M 4;a.*4.
' - REFERENCE 1.
Shoreham Nuclear Power Station: SP 29.023.05, Rapid RPV Depressurization Emergency Procedure, p.
1.
L d . - -. - - - -. - - , .. - - -. .-- , ,.,,, - -. - - -.,, - --
.. . __.
_. . _ _ _.
__ . , 7.__P8QQEQQRE@_ _NQRM6(s_@@NQRM6Qu_EMEBQEGQY_@NQ__ PAGE
. 86019LQQ1G6(_QQNIBQL - ANSWERS -- SHOREHAM-86/12/15-JARRELL.
D.
t . ANSWER 7.10 (2.50) l c.
Check valve slam and water hammer in the turbine exhaust line could result in exhaust line damage and/or turbine trip.
[+1.53 (Both HPCI and RCIC turbines exhaust to the suppression pool where the steam is condensed.
Insufficient flow may result in violent condensation i n the exhaust line drawing water i nto the exhaust line and resulting in water hammer.)
b.
By observing turbine rpm (< 2200), by turbine trip indication (turbine exhaust high pressure annunciator) or turbine exhaust pressure fluctuations. [+1.03 (Any two (2) for full credit.)
REFERENCE 1.
Shoreham Nuclear Power Station: SP 29.023.02, Cooldown Emergency Procedure, p 2.
2.
Shoreham Nuclear Power Station: HL1195H, RCIC System, p.
11, 3.
Shoreham Nuclear Power Station: HL2O2SH, HPCI System, p.
18.
4.
Shoreham Nuclear Power Station: SH944SHI, EPG, Objective CI.
sNSWER 7.11 (1.00) When there are less than two (2) LPRM inputs per core level [+0.53 or less than a total of 11 LPRM inputs to the APRM channel [+0.51 REFERENCE 1.
Shoreham Nuclear Power Station: SP 23.604.01, APRM System, pp.
2, 4, and 9.
,
! , - - ..,-en-. , nn.,- 7---- .y .--,,-~_,e.,--.,. - _-- - ,, ,,.y ,.,-,.-m..,,,. .- .mw ,,.. - -.. - -.., -- .- -
.___ _ _ _ _ _ _ _ _ _ _ _ _ _ __ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ Zt__EBQGEDUBEg_ _NQBd@bt_@@UQBd@6t_EdEB@ENQY_6NQ PAGE
,. BeQ196991G66_GQN18QL - ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
s . ANSWER 7.12 (2.00) 1.
SRV leaking annunciator c4 CC C Da *" " 2.
High temperature on SRV discharge tailpipe q k { e, w h s<ws ca. * O. 3.
High pressure on tailpipe pressure indicator 4.
Relief valve open annunciator 5.
Suppression pool temperature increase 6.
Suppression pool water level increase 7.
FW flow > steam flow 8.
Decrease in turbine generator load 9.
Temporary increase in reactor water level prior to SF/FF mismatch . 2, e'ac h, C +( max. +2.03 REFERENCE 1.
Shoreham Nuclear Power S t a t i r.n : S f' T~.!!!.01.
10.
. ANSWER 7.13 (1.00) (d.)
[+1.01 REFERENCE 1.
Shoreham Nuclear Power Station: HL-119-SH, Objective 6,7.
,.,c ,.. - >-
. 92__690lNISIB6IIVE_EBgCggyBES _CQNp1IJONS2_6ND_LJdlI@IlgNS PAGE
3
,. ANGWERS -- SHOREHAM-86/12/15-JARRELL, D.
W . ANSWER 8.01 (2.00) No [+0.53.
With D/G 102 INOP, el1-Ic:dt fed Ernm the nTU !! 7 rygr ryaw bue-arMensi der-ed4 NOP, int!'!di 3'the RBSW "B" pump H:+0.53.
This - means the "B" loop of RBSW is INOP C+0.53 and Section 3.7.1.1d applies (72 hour LCO). C+0.53 ),, 7,.a eaklf I*r-T.S. 8. l.l. '- &V ,, REFERENCE 1.
Shoreham Nuclear Power Station: HL 309 SH, Objective 7.2 and p.
18.
2.
Shoreham Nuclear Power Station: HL 122 SH, p.
5.
3.
Shoreham Nuclear Power Station: Technical Specifications.
Sections 1.2.7 and 3/4.7.1.
ANSWER G.OZ L2.50) c.
No C+0.53 b.
In accordance with Technical Specification 4.0.
, failure to perform required surveillance within the allowable extension interval causes the component / system to be INOP C+0.53.
Therefore, all eight MSIVs have been inoperable for 24 hours and Technical Specification 2.4.7 requires that the plant be in hot shutdown if MSIVs are INOP. C+0.53 Therefore, IAW Technical Specification 3.0.3, a shutdown should have been initiated within 1 hour to bring the reactor to Cold Shutdown within 24 hours. [+0.53 (When the test is reported as sastisfactory, power operation may resume,) but a notification to the NRC must be made due to a violation of a Technical Specification LCO. [+0.53 ! REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specification 3.05, -{ 3.4.7, and 4.0.3.
gf 3. b'3.,
! .A
- l Lo T<
o pc i
- - - _ - - - - - - _ _ _ - - _ - _ - - - , , as.__00dlNi@lB611yE_E8QQgQQ6E@3,_QQNQlllQN@s,@ND_Qldll@IlQNg PAGE
ANSWERS -- SHOREHAM-86/12/15-JARRELL, D.
. ANSWER B.03 (2.50) c.
False b.
False c.
True d.
False o.
True C+0.53 each REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specifications, Sections 6.2.2.
ANSWER 8.04 (2.50) REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specifications, Section 1.0, " Definitions".
ANSWER 8.05 (2.00) C.ed d cdsc bui+d ch n.
Yes C+0.53 e ho - b.
It (is a simple " typo" and) does not change the. intent of the hast3 d %d fC5 g T( procedure.
C+1.53 o - 6hadc, h%V U E '1 5 @ C lV T REFERENCE {te S P \\?. 0 0 G 01 P s * Cl - 1.
Shoreham Nuclear Power Station: SP 12.006.01, section B.6.1.
NOTE: A copy of SP 23.202.01, p.
3 with the appropriate change to 4.6 must be provided.
,. . _ _ _ _ _ _ _ - _ _ _ - _ -.
Ot__eQdlN1518eI1YE_EBQGEQUBESt_QQNQll1QNQu_@NQ_LidlI@l1QU@ FAGE
ATYSWE S -- SHOREHAM-86/12/15-JARRELL, D.
~ o . ANSWER 8.06 (2.00) c.
Primary Containment integrity is not being met. [+0.53 b.
The ball valve does not meet the definition of " operable" per section 1.0 [+0.53, and per section 3/4.6.3 is a primary containment automatic isolation valve [+0.53.
Section 3/4.6.1.1 b requires that the ball valve be deactivated in _ its isolated position (since its auto-close action is not operating properly) C+0.53.
REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specifications, Sections 3/4.6.1.1 and 3/4.6.3.
ANsWrR 8.07 c.m a.
True b.
False c.
False - d.
False C+0.53 each REFERENCE 1.
Shoreham Nuclear Power Station: Emergency Procedures.
ANSWER B.08 (1.50) If pressure in the drywell exceeds the suppression chamber pressure, thyn this DP may be corrplated Q water level (via Figure 11).
C+1.53 d\\5o u (. c t. pt ( = rp kuYup, oT LPSCCS 5v5 ken p rt 15 UN ka i ro vg h e 3,-t s w w \\t c. REFERENCE
1.
Shoreham Nuclear Power Station: Emergency Procedure. 8 __9901NISTRAllyg_PRQGgpQRES,_gpND11]QN$2_6NQ_ LIMIT 611QNS PAGE+ 41 t . ' ANSWERG -- SHORFHAM-86/12/15-JARRELL, D.
. ANSWER 8.09 (2.50) NO [+0.53.
Tech Spec section 3.0.4 requires that entry into an operational condition shall not be made unless the conditions for the limiting LCO are met wi h ut reliance on provisions contained in the action statements t (Section3.5.1 requires both CSS . C+1.03.)
cubsystems to be operational prior to entry into mode 1 REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specifications, Sections 3.0.4 and 3.5.1.
O N ANSWER O.10 (3.00) r l $*b \\'g \\ f D 1.
Take action of 3.8.3.3.c for inoperable swing bus which refers a h II'.f,i/ *g e to'acH no n* 3.5.1.b for inonarable LPCI subsystem.
q ,g t 2.
Since HPCI valve will not Llose then take action of 3.6.3.a to d' close in 4 hours.
Once valve is closed, HPCI is inoperable.
3.
Must take artic.n of 3.0.3.
[+1.03 each REFERENCE 1.
Shoreham Nuclear Power Station: Technical Specifications, Sections 3.03, 3.5.3.b, 3.6.3.a, and 3.8.3.3.c.
2.
Region I - NRC.
ANSWER 8.11 (2.50)
c.
1.
New York State Radiological Emergency Hotline - Disaster gg6 Preparedness Commission C+1.03 . ,jc 2.
NRC via dedicated phone line C+1.O] b.
The Watch Engineer is responsible (until relieved of Emergency Director duties).
Since he is unable to perform his duties, it must be done by the Watch Supervisor.
C+0.53
.
-, - - - .-,-+ c--s-m,.--..w- --,~-,,-nn---.----,y, ,..-.r,-- w--v.
, -, - - - - --
p- . v . ! Qu_ ADMINISTRATIVE _PRQGEQtjRESu_GONQlligNSu_ANQ_tlMilAl}QNS PAGE
a
> z . , (845WERS -- SliOREHAtt-86/12/15-JARRELL, D.
o . REFERENCE 1.
Shoreham Nuclear Power Station: Emergency Preparedness Plan, pp.
5-3, 5-4, and 5-12. }}