IR 05000322/1988003
| ML20151X781 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 08/05/1988 |
| From: | Howe A, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151X773 | List: |
| References | |
| 50-322-88-03OL, 50-322-88-3OL, NUDOCS 8808260102 | |
| Download: ML20151X781 (104) | |
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U.S. NUC' EAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.
88-03(0L)
FACILITY DOCKET NO.
50-322 FACILITY LICENSE NO.
NPF-36 LICENSEE:
Long Island Lighting Company Post Office Box 618 Wading River, New York 11792 FACILITY:
Shoreham Nuclear Power Station EXAMINATION DATES:
May 16 to May 20, 1988 CHIEF EXAMINER:
4 Md.
8'-5-f8
_TTen G. Howe, Senior erations Engineer Date A
APPROVED BY:
_d N.
p_
8 "f-#f David J. Lange, Chief, BWR Sectiorf, Date Operations Branch, Division of Reactor Safety SUMMARY: Written examinations and operating tests were administered to three (3) senior reactor operator (SRO) candidates and nine (9) reactor operator (RO) candidates, Two (2) SR0 candidates and seven (7) RO candidates passed these examinations.
All other candidates failed the examinations, h$bbo
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DETAILS TYPE OF EXAMINATIONS:
Initial EXAMINATION RESULTS:
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R0 l
SRO l
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Pass / Fail l
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CHIEF EXAMINER AT SITE: Allen G. Howe, Senior Operations Engineer 2.
OTHER EXAMINERS: 0. Lange, Chief, 2WR Section T. Fish, Operations Engineer D. Meco, PNL C. Mo.re, PNL 3.
The following is a summary of generic strengths or deficiencies noted on operating tests.
This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
STRENGTHS A.
General systems knowledge B.
Crew communications in the simulator C.
SRO knowledge of the Emergency Plan and Technical Specifications DEFICIENCIES A.
Ability to explain the operation of various control systems
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The following is a sunmary of generic strengths or deficiencies noted from the grading of written examinations. This information is being provided to aid the licensee in upgrading license and requalification training programs.
No licensee response is required.
SR0 STRENGTHS A.
Ability to calculate shutdown margins given changes in count rates.
(Question 5.03)
B.
Ability to predict the effects on the plant due to various failures of the APRM flow converters.
(Question 6.07)
C.
Understanding of the fuel preconditioning (PCICMR) process and requirements, (Question 7.06)
D.
Ability to classify events using the Emergency Plan Implementing Procedures.
(Question 8.07)
SRO DEFICIENCIES A.
Ability to predict relative changes in radial power distribution in response to movements of shallow and deep control rods.
Knowledge that CR0 flow is used by the process computer for the heat balance calculation.
(Questions 5.02 and 5.09)
8.
Knowledge of the purpose of the CST connection to the RCIC pump suction when operating RHR in the steam condensing mode.
(Question 6.04)
C.
Ability to recognize entry conditions for the "Emergency Shutdown" procedure (SP29.010.01).
Knowled,7 that below 110 deg. F the maximum allowable dryweli pressure is not constant.
(Questions 7.03 and 7.03)
D.
Ability to determine acticns required by technical specifications for inoperable control rods.
(Question 8.06)
R0 STRENGTHS A.
Ability to: predict the magnitude of power changes at different times in core life for the same reactivity addition, calculate the time to reach a specified given reactor period, predict changes in control rod worth due to changing other parameters, calculate cooldown rates, and predict critical power changes as plant parameters change.
(Questions 1.01, 1.04, 1.05, 1.07, and 1.11)
B.
Knowledge of how various air operated valves respond to a loss of air and the bases and controls for the Rod Block Monitor system.
(Ques-tions 2.01 and 2.11)
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B.
Knowledge of: what SCRAM signals are automatically bypassed and the operation of the Automatic Depressurization System. Ability to predict: the response of the reactor recirculation system to para-meter changes and changes in reactor water level due to plant para-meter changes.
(Questions 3.01, 3.04, 3.07, and 3.08)
D.
Knowledge of: technical specification requirements for recirculation pump speed mismatch and the administrative radiation dose limits.
(Questions 4.09 and 4.12)
R0 DEFICIENCIES A.
Ability to predict relative changes in radial power distribution in response to movements of shallow and deep control rods.
Failure to recognize that power reouctions result in less faedwater heating which in turn adds positive reactivity. Ability to determine if a pump will cavitate given the required NPSH and tank level at the suction.
(Questions 1.02, 1.09, and 1.11)
8.
Knowledge of: the operation of the emergency diesel generator shut-down system without starting air available and that the CST will provide NPSH to the RCIC pump when operating RHR in the steam concensing mode.
Understanding of the interlocks for the RCIC pump suction.
Knowledge that placing the Remote Shutdown System transfer switches in "EMERGENCY" bypasses most but not all trips and inter-locks.
(Questions 2.03, 2.04, 3.02, and 3.10)
C.
Knowledge of: the reasons for preventing automatic transfer of the RCIC suction to the suppression pool per the "Loss of All AC Power" procedure SP 29.015,20 and the maximum allowable average core power for any transient authorized per Standing Order Number 30.
(Ques-tions 4.04 and 4.03)
5.
Personnel Present at Exit Interview:
NRC Personnel A. Howe, Chief Examiner T. Fish, Operations Engineer Facility Personnel W. Steiger, Plant Manager S. Skorupski, Assistant Vice President Nuclear Operations M. Case, Operating Engineer L. Calone, Manager Operations Training Division H. McDaniel, Operations Training H. Carter Licensed Operator Training Supervisor C. Thayer, Manager Operations and Simulator Training Department K. Rottkamp, Manager, Facility Services Division i
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6.
Summary of NRC comments made at exit interview:
The chief examiner thanked the training and operations staffs for their cooperation during the examination. The operators in the control room were also helpful in answering questions.
The examiners felt site access was smooth and that housekeeping was adequate.
Examir.ation security for both the written and simulator portions were excellent.
The written examination review was discussed.
The facility staff was advised where to send the formal comments.
The facility staff also stated that the examination was of good quality.
The generic strengths and weaknesses noted on the operating examinations were discussed.
The simulator fidelity was reviewed.
This is detailed in Attachment 5 of this report.
Attachments:
1.
Written Examination and Answer Key (RO)
2.
Written Examination and Answer Key (SRO)
3.
Facility Comments on Written Examinations af ter Facility Review 4.
NRC Response to Facility Comments 5.
Simulation Facility Fidelity Report
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
SHOREHAM REACTOR TYPE:
BWR-GE4 DATE ADMINISTERED: 88/05/16 EXAMINER:
NRC REGION I
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ANSWER KEY CANDIDATE:
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INSTRUCTIONS TO CANDIDATE:
Write answers on one side only.
Use separate paper for the answers.
on top of the answer sheets.
Points for each
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Staple question sheet The passing
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question are indicated in parentheses af tr the cuestion.
anc a final grade of at arade requires at least 70% in each category least 804. Examination papers will be picked up six (6)
hours after the exar'. nation starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
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VALUE TOTAL SCORE VALUE CATEGORY
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1.
PRINCIPLES OF NUCLEAR POWER 25.00 25.00 PLANT OPERATION, THERM 0DYNAft!CS, HEAT TRANSFER AND FLUID FLOW 2.
PLANT DESIGN INCLUDING SAFETY 25.00 25.00 AND EMERGENCY SYSTEMS 25.00 25.00
_ 3.
INSTRUMENTS AND CONTROLS 4.
PROCEDURES - NORMAL, ABNORMAL, 25.00 _ 25.00 EMERGENCY AND RADIOLOGICAL CONTROL
%
Totals 100.00 Final Grace All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
Cheating on the examination means an automatic denial of your application 1.
and could result in more severe penalties.
Restroom trips are to be limited and only one candidate at a time may 2.
You must avoid all contacts with anyone outside the examination leave.
room to avoid even the appearance or possibility of cheating.
Use black ink or dark pencil only to facilitate legible reproductions.
3.
Print your name in the blank provided on the cover sheet of the 4.
examination.
Fill in the date on the cover sheet of the examination (if necessary).
5.
6.
Use only the paper provided for answers.
Print your name in the upper right-hand corner of the first page of each 7.
section of the answer sheet.
Consecutively number each answer sheet, write "End of Category _" as 8.
appropriate, start each category on a new page, write only on orre side of the paper, and write "Last Page" on the last answer sheet.
Number each answer as to category and number, for example, 1.4, 6.3.
9.
10. Skip at least three 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 Lsed in facility literature.
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 00 NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of-the examiner 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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18. When you complete your examination, you shall:
Assemble your examination as follows:
a.
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the anrwer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
Turn in all scrap paper and the balance of the paper that you did c.
not use for answering the questions.
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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 1: cense may be denied or revoked.
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PRINCIPLES OF NUCLEAR.90ER PLANT OPERATION, PAGE
JHERM0 DYNAMICS, HEAT TRAN C ER AND FLUID FLOW
QUESTION 1.01 (3.00)
The reactor has been operating at 4% of rated core themal power for several days with the main turbine on the turning gear.
Reactor pressure is 53 psig. The reactor operator withdraws an in-sequence control roo and reactor power increases to 4.5%.
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a.
LIST the two (2) reactivity coefficients that had the greatest effect in initially turning the power increase.
(0.5)
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b.
Once 4.5% power is initially obtained, with no further changes in control rod position or reactor recirculation behave (relative to 4.5%)gnitude of reactor
)ower willinitially during t1e next hour flow, DESCRIBE HOW the ma AND over the following 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of operation.
INCLUDE a brief explanation for this behavior.
(Specificvalues for changes in magnitude are not required).
(1.0)
c.
If the same amount of reactivity were to be inserted at these same initial conditions but LATER in the fuel cycle, STATE whether the reactor period observed during the transient would be (GREATER THAN, LESS THAN, or EQUAL TO)
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the period observed earlier in the fuel cycle.
SUPPORT your answer with an explanation.
(1.5)
i QUESTION 1.02 (2.50)
The reactor is operating at 95% of rated core tharmal power.
Control rod 18-35 is at notch position 06.. Control rod 22-35
is at notch position 38.
a.
STATE which of these two control rods is most likely to produce the greatest increase in core themal power if withdrawn one additional notch.
(0.5)
b.
STATE which of these two control rods is most likely to produce the SMALLEST CHANGE in core RADIAL power distribution if withdrawn one notch.
INCLUDE a brief explanation to support your answer.
(1.0)
c.
The reactor operator withdraws a shallow control rod one notch and notes that total core power decreased slightly in response to the rod movement. Briefly EXPLAIN HOW such a response is possible.
(1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *"**)
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.03 (2.00)
several control rods fail Following a reactor scram from power,ithin one hour, the reactor to insert to the full-in position. W is determined to be suberitical with an actual shutdown margin (SDM) of 0.22% delta X/K.
If reactor coolant temperature and control rod positions a.
remain constant during the next hour, WOULO actual SDM (INCREASE, DECREASE, or REMAIN THE SAME)? Briefly EXPLAIN (1,0)
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your answer.
b.
During the next hour you notice reactor pressure is decreasing. WHAT effect would this have on actual SOM7 (1.0)
Briefly EXPLAIN your answer.
QUESTION 1.04 (2.00)
Reactor power is 60 on IRM range 2 with the MINIMUM permissible stable positive period allowed by procedure SP22.001.02.
Heating power is determined to be 40 on IRM range 7.
HOW LONG will it take for reactor power to double if period (1.0)
a.
remains constant?
b.
HOW LONG will it take for power to reach the point of (1.0)
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adding heat if period remains constant?
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
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THERMODYNAMICS, HEAT TRANSFER AND FLU 10 FLOW
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QUESTION 1.05 (2.50)
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SELECT the appropriate response for each of the following
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statements concerning Control Rod Worth:
(MORE/LESS) control rods would need to be pulled to make a.
the reactor critical at 545 deg F, as opposed to 140 deg F.
(ASSUME initial rod sattern identical, and the same sequence is used in soth cases.)
(0.5)
b.
An INCREASE in the Void Fraction will result in a/an (INCREASE / DECREASE) in individual control rod worth.
(0.5)
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c.
Control rod worth will (INCREASE / DECREASE) with an
INCREASE in moderator temperature.
(0.5)'
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d.
Control rod worth at End of Cycle would be (GREATER /LESS)
than at the Beginning of Cycle.
(0,5)
Control rod worth will (INCREASE / DECREASE) as the e.
adjacent control rods are withdrawn.
(0.5)
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i QUESTION 1.06 (3.00)
An EHC load reject o
REGION I ANSWER 1.01 (3.00)
a.
1.
void coefficient
[+0.25]
2.
Doppler coefficient [+0.25]
b.
Reactor power will first increase greater than 4.5% [+0.25]
as xenon is initially depleted faster than it is produced than 4.5% [+0.25] power will subsequently stabilize at lessas the prod
[+0.25]. Reactor of iodine restores xenon concentration to a new higher equilibrium value [+0.25].
(Reactor period would) LESS THAN [+0.5].
Increased core c.
age results in a higher fraction of core power being produced by Pu-239 [+0.25] which reduces the value of BETA-EFFECTIVE [+0.5].
From the period equation, as BETA-EFFECTIVE decreases, for a given reactivity insertion, the resultant reactor period decreases [+0.25].
REFERENCE 1.
Shoreham: HL-900-SH1, Lesson 11, L0 CB and Lesson 15, L0 CD, CH.
2.
GE: Reactor Theory, Chapters 3, 4, and 6.
292003K106 292004K114 292006K105 292006K106
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE-24 iiHERM0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW
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ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
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ANSWER 1.02 (2.50)
a.
(Control rod) 18-35 [+0.5]
b.
(Control rod) 22-35 [+0.33]. The rod shadowing effects of adjacent rods [+0.33] will be more significant in dampening radial power because of the higher local control rod density [+ relative to the control rod tip (of control rod 22-35)
0.34].
c.
When a shallow rod is withdrawn a notch, the void fraction in the adjacent fuel channels increases throughout the entire boiling length [+0.33]. Though local power will increase where the rod tip was withdrawn [+0.33], it is possible for the negative reactivity of the increased voiding to be the dominate effect (causing net total power to decrease) [+0.34].
REFERENCE 1.
Shoreham: HL-900-SH1, Lesson 13, L0 CB, CC.
2.
GE: Reactor Theory, Chapter 5.
292005K104 292005K112 292008K119
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ANSWER 1.03 (2.00)
a.
increase [+0.5] due to Xe build in [+0.5]
b.
The decrease in reactor pressure is directly related to a decrease in reactor coolant temperature. Due to the negative value of the moderator temperature coefficient, reactivity increases with decreasin moderator temperature
[+0.5].
Hence, SDM decrea'ses [+0.5.
REFERENCE
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1.
Shoreham: HL-900-SH1, lesson 12, L0 CB.
292002K110 292002K114
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PAGE 25 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
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ANSWER 1.04 (2.00)
(FromSP22.001.02) minimum period equals 60 seconds [+0.5].
a.
Thus doubling time is 60/1.443 = 41.6 seconds [+0.5].
.b.
60 on range 2 is equal to 0.05 on range 7 P(t) = P(o)e**-t/T [+0.25]
P(o) = 0.06, P(t) = 40, period = 60 seconds
[+0.25]
t = 60 in 40/0.06
= 390 seconds or 6.5 minutes
[+0.5]
REFERENCE 1.
Shoreham: HL-602-SH1, Lesson 15, L0 CA.
2.
Shoreham: SP22.001.02.
ReactorTheory), Chapter 3.
3.
GE:
...(KA'S 292003K108 ANSWER 1.05 (2.50)
a.
more b.
decrease c.
increase d.
less e.
increase
[+0.5] each REFERENCE 1.
Shoreham: HL-900-SH1, Lesson 12, L0 CG.
292005K109
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION PAGE 26 t
THERM 0DYHAMICS, HEAT TRANSFER AND FLUID FLOW
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88/05/16-NRC REGION I ANSWERS -- SHOREHAM
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ANSWER 1.06 (3.00)
Reactor power will rapidly increase due to the pressur,e a.
increase [+0.5].
Power will then decrease due to the(TCV fast closure) scram [+0.5].
MM 5/W -
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Reactor pressure will rap] idly increase due to the rapid b.
closure of the TCVs [+0.5. Pressure will then decrease due to the scram and the opening of the bypass valves which will then attempt to maintain reactor pressure at 920 psig
[+0. 5].
Reactor water level will initially drop due to collapsing c.
of voids [+0.5]. The feed control system will respond to increase level and level should then rise to the level controller setpoint (level may overshoot causing feed pumps to trip) [+0.5].
REFERENCE 1.
Shoreham: HL-900-SH1; HL-657-SH1, L0 C3.
241000K101 241000K102 241000K103
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ANSWER 1.07 (2.50)
The previous shift DID EXCEED the cooldown limit [+0.5] of a.
90 degrees F/P [+0.5].
Aw Auur
",,,, @ r/4, "
g9
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(Tsat for 630 psig = 494 degrees F; Tsat for 200 psig = 388 degrees F; cooldown rate = (494-388) degrees F/1 hour
= 106 degrees F/hr)
[+0.5]
b.
35 to 64 degrees F (of cooldown required depending on assumptions)
Tsat for 200 psig = 388 degrees F; Tsat for 125 psig = 353 degrees F; Tsat for 80 psig = 324 degrees F; (388-353) = 35 degrees F; (388-324) = 64 degrees F
[+1.0]
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REFERENCE i
1.
Shoreham: HL-901-SH1, L0 CB; HL-121-SH1, L0 CC.
205000K402
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PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 27 THERMODYNAMICS, HEAT TRANSFER Kl(D FLUID FLON
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ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 1.08 (3.00)
1.
LHGR - Linear Heat Generation Rate [+0.5] designed to (limit the pin power at any node in the reactor to) limit the fuel clad strain to less than 1% plastic strain [+0.5].
2.
APLHGR - Average Planer Linear Heat Generation Rate [+0.5]
designed to assure the maximum fuel clad temperature (following a design basis accident) will not exceed 2200 degrees F [+0.5].
3.
MCPR - Minimum Critical Power Ratio [+0.5] designed (to limit the power of any fuel element) to prevent any point in the bundle from experiencing the onset of transition boiling [+0.5].
REFERENCE 1.
Shoreham: HL-904-SH1, L0 CA and CD.
2.
GE: HTFF, Chapter 9.
293009K107 293009K111 293009K119
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ANSWER 1.09 (2.00)
The reactor is now producing less steam to go to the turbine.
There will be less extraction steam going to the feedwater heater [+1.0].
Therefore, less feedwater heating will occur resulting in colder feedwater entering the vessel [+0.5] which will cause reactor power to increase (about 3%) from the positive reactivity addition (alpha m) [+0.5].
REFERENCE 1.
Shoreham: HL-901-SH1, Le.sson 4, L0 CA.
2.
GE: HTFF, Chapter 5.
3.
GE: Rx Theory, Chapter 7.
292008K120 292008K121 2930005K10
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PAGE 28 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION,
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
ANSWERS -- SH0REHAM-88/05/16-NRC REGION I ANSWER 1.10 (1.00)
available NPSH is inlet pressure minus saturation pressure a.
[+0.5]
, due to the contribution from atmospheric No[+0.25}L+0.25].
b.
pressure REFERENCE
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Shoreham: HL-902-SH1, LO.
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293003K123 293006K109 293006K110 (KA'S)
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ANSWER 1.11 (1.50)
The assembly (bundle) power that would cause the onset of a.
transition boiling [+0.75].
b.
(1)
[+0.75]
REFERENCE 1.
Shoreham: HL-904-SH1, Lesson 1, LO CA.
293009K117 293009K122 293009K123 293009K124 (KA'S)
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PAGE 29 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
. 2.
ANSWERS -- SHOREHAM-28/05/16-NRC REGION I
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ANSWER 2.01 (2.00)
a.
fail closed b.
fail open c.
fail closed d.
fail open e.
fail closed
[+0.4] each REFERENCE 1.
Shoreham: HL-109-SH1, L0 M.1.
2.
Shoreham: HL-117-SH1, L0 CC.
295019AK20 295019AK21
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ANSWER 2.02 (2.50)
a.
RHR pump (3)
2 second T.D.
- +0.2;
, 0. 2, 7 second T.D.
+
core spray pump (1)(4) and
, 0. 2, service water pump 12 second T.D.
+
0.2 RBSYS/CRAC water chillers (2)
12 second T.D.
+
,
(Point awards above are for T.D. values only.
[+0.7] for correct order, no partial credit)
The timing sequence is initiated by:+the closing of the b.
diesel generator output breaker.
1.0)
REFERENCE i
1.
Shoreham: HL-307-SH1, LO B.I.b.
262001A304 262001K301 264000K506
...(KA'S)
l
!
.
,*
.
t 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 30
'
.
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I-ANSWER 2.03 (2.50)
a.
1.
overspeed [+0.5]
2.
generator phase differential [+0.5]
3.
generator overcurrent
[+0.5]
' 0. 5" b.
1.
will not
+
2.
will not
'.+0. 5l REFERENCE
.
1.
Shoreham: HL-307-SH1, LO A, C.4, D.
264000K106 264000K601
...(KA'S)
'
ANSWER 2.04 (1.50)
a.
IS NOT [+0.5]
.
b.
The RCIC pump WILL NOT cavitate [+0.25]. CST inventory will maintain RCIC pump suction pressure [+0.75].
REFERENCE 1.
Shoreham: HL-121-SH1, L0 CC.
~
2.
Shoreham: 23.121.01.
217000A101 217000K101 217000K105 217000SG1
...(KA'S)
t
.
..
..
2.
PLANT DESIGN IfiCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 31
.
.
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
~
ANSWER 2.05 (3.00)
Hi drywell pressure 1.69 psig OR [+0.5]
a.
Lo Lo Rx water level -132.5" [+0.5]
(NOTE: Do not penalize if RPV low pressure permissive of 338 psig is also listed.)
b.
1.
containment spray valve manual override keyswitch in
"MANUAL"
[+0. 5]
valve accident control switch in containment sp] ray (until seal-in status light is lit)
2.
"MANUAL" [+0.5 c.
1.
shutdown cooling mode
[+0.5]
2.
fuel pool cooling mode [+0.5]
REFERENCE 1.
Shoreham: HL-204-SH1, L0 CC.
226001K403
...(KA'S)
l
,
l
,
!
l
..
..
.
,
,
, 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 32
,
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 2.06 (3.00)
a.
1.
squib valve continuity circuit indicator lamp extinguishes 2.
squib valve loss of continuity annunciator
'
3.
SBLC pump discharge pressure greater than reactor pressure
.
.
4.
SBLC pump running indication (ON)
hp$$fsh?Addtw.
cf,f
'
"
w ak Any four (!)a p s[+0.25] each, +1.0 maximum.
,
b.
A too rapid injection rate could cause insufficient mixing and uneven concentrations of boron circulating in the core
[+0.5] leading to power oscillations ("chugging") [+0.5].
c.
below (the core plate)
[+0.5]
d.
No [+0.5]
REFERENCE 1.
Shoreham: HL-123-SH1, LO B.1, F.
211000K106 211000K403 211000K405 211000K506
...(KA'S)
ANSWER 2.07 (2.50)
The exhaust damper [^0.5] of the op[+erating supply fan a.
[+0.5] will modulate further open 0.5].
b.
(1) trip [+0.5]
(2) remain running
[+0.5]
REFERENCE 1.
Shoreham: HL-405/418-SH1, LO I.0.5.
261000K101 261000K401
...(KA'S)
.
..
.
.
.,
PAGE 33 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
.
,
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 2.08 (1.50)
Primary containment could be overpressurized [+0.5] because a.
of steam bypassing the suppression pool, pressurizing containment [+0.5).
b.
Yes
[+0.5]
REFERENCE
.
1.
Shoreham: HL-654-SH1, L0 CB.
223001K405 223001K501 223001K503
...(KA'S)
ANSWER 2.09 (1.50)
a.
1.
RWCU nonregenerative heat exchanger outlet (filter demineralizer inlet) temperature high [+0.25]
2.
standby liquid control (SLC) system initiation
[+0.25]
(In-line) conductivity must be sampled periodically [+1.0].
b.
(continuous conductivity indication has been lost)
REFERENCE 1.
Shoreham: Technical Specifications 3.4.4.
2.
Shoreham: HL-709-SH1, L0 CE, CF, CG.
204000X404 204000K507 204000SG11 204000SG5
...(KA'S)
,
.
. *
<
,
..
PAGE 34 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
.
,
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 2.10 (3.00)
False [+0.5] - Once the low suction pressure signal is a.
clear, the turbine will auto restart if the initiation signals are still present [+0.5].
b.
True [+0.5] - The low steam pressure auto isolation signal seals in, and must be manually reset (using the AUTO ISOLATION SIGNAL RESET u;hbut' r on the *PNL-601 after the reason for the isolation hh,e;een detgrained and b
corrected) [+0.5].
kyld 54 u True [+0.5] - The oil ?ressure will be restored when the
.
c.
turbine coasts down, t1ereby causing the stop valve to open
.
'
[+0. 5].
REFERENCE 1.
Shoreham: High Pressure Coolant Injection System Procedure 23.202.01, Rev. 18.
2.
Shoreham: HL-202-SH1, L0 CI.
206000K401
...(KA'S)
.
ANSWER 2.11 (2.00)
local fuel damage (by generating a rod withdrawal block)
A'- Vqrt a.
[+1.0]
Atr ns: orwnna Loc tw.m os ynt cat ir > mpa.
units = volts [+0.5], number of operable LPRM inputs can be b.
calculated (by using J volts per operable input) [+0.5]
- YS REFERENCE 1.
Shoreham: HL-603,604,652-SH1, LO C.
215002A402 215002K102 215002SG04
...(KA'S)
.
I
.
..
..
.
3.
INSTRUMENTS AND CONTROLS PAGE 35
-
.
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 3.01 (3.00)
bypassed when the mode switch is NOT in run a.
b.
auto bypassed after (10 sec.) time delay auto bypassed if reactor power <30 percent (as indicated by turbine c.
first stage pressure of 109 psig)
d.
manualoypassswitchdinBYPASSwithmodeswitchinSHUTDOWNor REFUEL W $%
[+0.75] each REFERENCE 1.
Shoreham: HL-611-SH1, L0 D.
212000K412 212004K408
...(KA'S)
ANSWER 3.02 (3.00)
1:+0.5]
closure of the RCIC steam supply stop valve (MOV-43) :
a.
and closure of trta ed thrcttic valve (MOV-44) [+0.51 A 54M3
"14jedi.. W A Ly)
L b.
RCIC will automatically initiate (and inject to the RPV)
[+0. 5]
,
,
c.
align in RCIC starting mode and inject
[+0.5]
d.
no (locally)
[+0.5]
by placing the suppression pool suction valve control e.
switch in the CLOSE position
[+0. 5]
REFERENCE 1.
Shoreham: HL-119-SH1, LO I.
217000A201 217000K202 217000K402
...(KA'S)
1
. _. _... _ _,. _. _ _ _... _... _ _ _ _ _ _ _
_ _.. ~. _
___,
____ _ __._ -____.-
_
_
.
,..
,-
.-
.
'
PAGE 36
. 3.
INSTRUMENTS AND CONTROLS
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 3.03 (2.00)
[+[0.5] ]]
a.
upscale
+0.75 b.
decrease
[+0.75 c.
increases REFERENCE 1.
Shoreham: HL-621-SH1, LO I.D.5.
'
216000K105
...(KA'S)
ANSWER 3.04 (2.50)
a.
remain as is b.
close c.
close d.
open
,
e..
remain as is
[+0.5] each REFERENCE 1.
Shoreham: HL-201-SH1, L0 F.
218000A205 218000K404 218000K501 218000K602
...(KA'S)
ANSWER 3.05 (2.75)
a.
(1)
(rodblock)
[+1.01 b.
(3)
(rod block and half scram)
[+1.0]
(2) (re'.irculation loop "Driving" flow)
[+0.75]
c.
REFERENCE 1.
Shoreham: HL-603-SH1, LO I.C.
215005K110 215005K116 215005K607
...(KA'S)
i l
I
.
,.
.
.
PAGE 37
. 3.
INSTRUMENTS AND CONTROLS
,
,
ANSWERS -- SH0REHAM-88/05/16-NRC REGION I
'
ANSWER 3,05 (3.00)
adjusts the gain of the RBM channel [+0.75]
compensate
a.
for either variations in local power
--OR--
bypassed LPRMs
[+0.75] for either answer b.
1.
manual operation of RBM bypass switch
[+0.5]
2.
edge rod selected [+0.5]
3.
reference APRM downscale (less than 30's)
[+0. 5]
REFERENCE 1.
Shoreham: HL-606-SH1, L0 CA.
215005A403 215005K403 215005K502
...(KA'S)
ANSWER 3.07 (2.25)
Both recirculation pumps run back to 45's speed [+0.5] due a.
to the automatic runback interlock with speed limiter (#2)
[+0.5].
b.
Recirculation pump ' A' trips [+0.5) due to the discharge valve not-full-open interlock with the flG set drive motor breaker [+0.25]. Recirculation pump 'B' speed will be unaffected [+0.5].
REFERENCE 1.
Shoreham: HL-658-SH1, L0 CG.
202002K305 202002K604
...(KA'S)
ANSWER 3.08 (1.50)
increases [+0.5]9 a.
b.
decreases F +0. 5'
c.
increases l+0. 5.I REFERENCE 1.
Shoreham: HL-656-SH1, L0 C.
259002K301 259002K604 259002K605 295002K603
...(KA'S)
.
..
.
.
PAGE 38
, 3.
INSTRUMENTS AND CONTROLS
,
ANSWERS -- SH0REHAM-88/05/16-NRC REGION 1 ANSWER 3.09 (3.00)
Panel-603 = light on for greater than or equal to 100 cps a.
back panel = light off for greater than or equal to 100 cps
-
[+0.5]
b.
SRM high, greater than 1 x 10**5 [+0.25], rod block below range 8 on IRM and not in run [+0.5]
SRM downscale, less than 3 cps [+0.25), rod block below
-
range 3 on IRM and not in run [+0.5]
SRM retract not permitted less than 100 cps [+0.25], rod block if detector not full in and below range 3 on IRM and not in run [+0.5]
Removal of ths shorting links (will cause any single SRM to c.
produce a full scram signal at 2 x 10**5 cps).
[+0.25]
REFERENCE 1.
Shoreham: HL-G01-SH1, L0 C and D.
215004K101 215004K401 215004K402 215004K405
...(KA'S)
ANSWER 3.10 (2.00)
a.
true b.
true true c.
false (some trips and valve interlocks are not bypassed)
d.
[+0.5] each REFERENCE 1.
Shoreham: HL-133-SH1, LO C and E.
295016AK20
...(KA'S)
.-
-
-
-
- _
_.
. _
_ -
-
.
.
-
...
,
PAGE '39 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND 9A010L0u7 CAL CONTROL
'
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 4.01 (2.50)
To prevent control rod drive mechanism damage [+0.5] during a.
a scram [+0.5].
,
b.
15 minutes [+0.5]
To prevent inadvertent bypass valve operation [+1.0].
c.
REFERENCE 1.
Shoreham: SP 22.001.01; HL-106-SH1 LO J.
216000SG1 241000SG1
...(KA'S)
ANSWER 4.02 (2.50)
1.
Rx building differential pressure at or above 0 inches of water 2.
Rx building exhaust radiation level above maximum normal operating radiation level 3.
Rx building floor drain sump water level above maximum normal operating (table) value
4.
any secondary containment area temperature above maximum normal operating (table) value 5.
any secondary containment area radiation level above maximum normal operating (table) value
[+0.5] each REFERENCE 1.
Shoreham: SP29.023.02.
2.
Shoreham: HL-944-SH3, LO I.B.
295032SG11 295033SG11 295035SG11
...(KA'S)
,
.
"
.
.
,
..
.
PAGE 40 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND HA010 LOGICAL CONTROL
' ~ ~ ~
"
ANSWEP.S -- SHOREHAM-88/05/16-NRC REGION I ANSWER-4.03 (2.50)
the entry condition was high drywell pressure >1.69 psig 1.
-
[+1.0]
the emergency procedures which should have been entered 2.
were Emergency Shutdown" (SP 29.010.01)
[+0. 5]
"
a.
b.
"Reactor Pressure Vessel (RPV) Control" (SP 29.023.01)
[+0.5]
-
"Primary Containment Control" (SP 29.023.03)
[+0.5]
.
'
c.
i REFERENCE 1.
Shoreham: HL-944-SH1, LO I.B.
2.
Shorcham: HL-944-SH2, LO I.B.
2950245G11
...(KA'S)
,
f l
l l
l
-. -
.-
.,.
,.
. -
.
. - _ - _.. _
-..,-
-. - - -
.
.
.
..
PAGE 41 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 4.04 (3.00)
ressurization will result in suppression pool Early RPV dep(containment) temperatures [+0.25] and a.
and drywell p[+ressures [+0.25] remaining below designed limitations 0. S].
(ALTERNATE ANSWER: to limit the total heat load [+0.5]
placed upon the primary containment [+0.5])
b.
(The RCIC vacuum pump is secured) to prolong [+0.5] the use of the division I battery [+0.5].
( ALTERNATE ANSWER: to reduce the load [+0.5] upon the Division I battery [+0.5])
c.
1.
to slow th:: rate of containment temperature and pressure rise 2.
to avoid failure of the RCIC turbine due to high lube oil temperatures (Either1.or2.for[+1.0])
REFERENCE 1.
Shoreham: SP 29.015.02.
295003AK20 295003AK30
...(KA'S)
ANSWER 4.05 (1.53)
a.
white [+0.5]
b.
will not
[+0.5]
c.
extended (RWP)
[+0.5]
REFERENCE 1.
Shoreham: SP 12.012.01 294001K103
...(XA'S)
.-.
.
{.
.
...
PAGE 42
' ~'. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
~
.
' RADIOLOGICAL CONTROL-88/05/16-NRC REGION I ANSWERS -- SHOREHAM
.
ANSWER 4.06 (3.00)
commence power reduction (in accordance with
a.
1.
SP22.004.01,"OperationBetween20%and100% Power")
~.Cl-
[+0.6]
main turbine trip [+0.2] at 22.5" Hg vac [+0.2]
2.
a.
b.
reactor feed pump (turbine) trip [+0.2] at 20" Hg vac [+0.2]
MSIV (/ main steam drain) isolation [+0.2] at 8.5" c.
Hg vac [+0.2]
d.
turbine bypass valve (TBV) isolation [+0.2] at 7"
.-
Hg vac [+0.2]
to avoid hydrogen explosion (above 4% power) [+0.4]
b.
1.
,
the level of radioactivity in the noncondensable 2.
condenser gases is significant (above 4% power) [+0.4]
(ALTERNATE ANSWER: condenser air removal pump exhaust is not treated prior to release)
REFERENCE
~
1.
Shoreham: SP 29.012.01.
2.
Shoreham: HL-701/714-SH1, LO E.4.
295002AK20
...(KA'S)
_.
.
o
.
PAGE 43 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
.
., ~ RADIOLOGICAL CONTROL ANSWERS -- SH0REHAM-88/05/16-NRC REGION I ANSWER 4.07 (2.00)
h.
1.
trip the operating RWCU pump [+0.25]
2.
isolatetheRWCUsystemfromcontainment-(closeMOVs 033 and 034)
[+0.25]
3.
reduce reactor recirc pump speed to minimum [+0.25]
4.
trip both reactor recire pumps
[+0.25]
5.
initiate emergency shutdown procedure (SP 29.010.01)
[+0.25]
6.
trip CRD pumps after all control rods are verified
,
inserted [+0.25]
b.
10 minutes
[+0.5]
REFERENCE 1.
Shoreham: SP 29.017.01.
2950185G10 2950185G11
...(KA'S)
ANSWER 4.08 (2.00)
a.
5% [+0.5]
(Average core power is) the average of all AP'lM readings.
b.
[+1.0]
APRM readings are to be taken at panel 608 [+0.5] (also c.
allow"backpanel").
' REFERENCE Standing)OrderNo.30.
1.
Shoreham:
...(KA'S 294001A103
'
,
m m
-
--
-
.
.
<.
.
PAGE 44
.,4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
' ' RADIOLOGICAL CONTROL ANSWERS -- SHOREHAM-83/05/16-NRC REGION I ANSWER'
4.09 (1.00;
[+[0.5))
a.
'5%
+0.5 b.
10%
-
REFERENCE
,
1.
Shoreham: Technical Specifications, Section 3.4.1.3.
.
2.
Shoreham: HL-120-SH1, L0 C.5, L.
202002SG6
...(KA'S)
,
ANSWER 4.10 (1.50)
A licensed operator must perform his duties (of his a.
license) for seven 8-hour shifts [+0.5] or five 12-hour shifts [+0.5J.
b.
the shift turnover sheet
[+0.5]
REFERENCE
1.
Shoreham: SP 21,001.01.
2.
Shoreham: SP 21.002.01.
294001A103
...(KA'S)
ANSWER 4.11 (1.00)
,
a.
is not required
[+0.5]
Any lead that in normal service could be exposed to b.
voltages in excess of 130 volts (requires a hold-off tag).
,
[+0.5)
REFERENCE 1.
Shoreham: SP 12.011.01.
2.
Shoreham: SP 12.035.01.
294000K102
...(KA'S)
I
!
.
l
..
..
-
. - - -
-..
__ -..
..
__
l
..
.
-
.
PAGE 45
.-
14.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND
- ji,
~' ' RADIOLOGICAL CONTROL
_
,
.-
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 4.12 (2.50)
P+0. 5'l a.
1.
3000 mrem l+0.5 2.
JGMP mrem 67re w sg3 b.
1250 mrem [+0.5]
[+0. 5] [+0. 5]
c._
1.
500 mrem /wk 2.
1000 mrem / quarter
-
REFERENCE 1.
Shoreham: SP61.012.01, 2.
29400K103
...(KA'S)
t
'
,
t r
.
e i
.
gF-
we 4wram.~.
..
.
.
~
f a ma v a s/t Cycle officiency = (Netacrx cut)/(Energy in)
'
.,.
5 = V t - 1/2 at'
w = mg o
[s mC'
A=As"
~
'
KE = 1/2 mv a = (Vf - V,)/t A = trl n
PE = mgn x = in2/t1/2 = 0.693/t1/2 w = e/t Vf = V, + at 1/2*II*E5*1P'I(*5)3 t
-
y. y ap ((t I ^ (*b)3 1/2
.
c.E = 931 un-U I = I,e
.
Q = mCp at I = I,e~"*
~
"
Q = UA/.n I=I 10"*/ M Pwr = W t.n g
f TVL = 1.3/u
,
'
sur(t)
HVL = -0.693/u P = P,10 P = P e /T t
,
o SUR = 25.05/T SCR = S/(1 - K,ff)
CR = S/(1 - K,ffx)
'
x SUR = 25p/1' + (s - o)T CR)(1 - K,ffj) = CR II ~
eff2)
T = (1*/o) + [(S - o)/ o]
M = 1/(1 - K,ff) = CR)/CR, T = 1/(o - s)
M = (1 - K,ff,)/(1 - K,ff3)
,
T = (a - o)/(*o)
SDM = (1 - K,ff)/K,ff
,
t' = 10 seconds a = (X,ff-1)/K,ff = M,ff/Keff T = 0.1 seconds" o = [(i=/(T (,ff)] + [T,ff (1 AT)]
/
d Id Ij j =2,2 2 P = (ILV)/(3 x 1010)
Id 7d g
2 R/hr = (0.5 CE)/d (meters)
- = oN R/hr = 6 CE/d2 (f,,;)
Miscellaneous Conversions Water Parameters
1 curie = 3.7 x 10 dps 1 gal. = 8.345 lem.
1 leg = 2.21 lom
~
1 gK. = 3.78 liters
1 np = 2.54 x 10 Stu/nr 1 ft* = 7.48 gal.
1 m = 3.41 x 100 Bru/hr
Density = 62.4'lem/ft
lin = 2:54 cm Density = 1 gm/cm
'F = 9/5'C + 32 Heat of vaporization = 970 Stu/lom Heat / fusion = 144 Btu /lem
'O = 5/9 (*F-32)
1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-1bf
,
'
i ft. H O = 0.4335 lbf/in.2 r
I k
yQ l.."?
.
c-U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
SHOREHAM REACTOR TYPE:
BWR-GE4 DATE ADMINISTERED: 88/05/16 EXAMINER:
NRC REGION I CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Write answers on one side only.
Use separate pa)er for the answers.
on top of the answer sheets.
Points for each Staple question sleet The passing question are indicated in parentheses after the question.
grade of at i
and a final grade requires at least 70% in each category least 80%. Examination papers will be picked up six (6)
hours after j
i the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5.
THEORY OF HUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 6.
PLANT SYSTEMS DESIGH, CONTROL, 25.00 _ 25.00 AND INSTRUMENTATION 7.
PROCEDURES - NORMAL, ABNORMAL, 25.00 25.00 EMERGENCY AND RADIOLOGICAL CONTROL 25.00 _ 25.00
_ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
%
Totals I
100.00 Final Grade I have neither given All work done on this examination is my own.
nor received aid.
.
Candidate's Signature
-
_ - _ - -
- - -
- _. _.
.
. _ _. _ _
__-
- - - - - -
Ta.
1 'y o
e'
.
.
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
SHOREHAM'
REACTOR TYPE:
BWR-GE4 DATE ADMINISTERED: 88/05/16 EXAMINER:
NRC REGION I CANDIDATE:
ANSWER KEY INSTRUCTIONS TO CANDIDATE:
Use separate pa)er for the answers. Write answers on one side only.
Staple question sleet on top of the answer sheets.
Points for each The passing question are indicated in parentheses after the cuestion.
anc, a final grade of at grade requires at least 70% in each category l' east 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
VALUE TOTAL SCORE VALUE__
CATEGORY 25.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
~~
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t 25.00 25.00 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL I
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25.00 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
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%
Totals 100.00 Final Grade
I have neither given I
All work done on this examination is my own.
nor received aid.
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules apply:
t Cheating on the examination means an automatic denial of your application 1.
and co'uld result in more severe penalties.
Restroom trips are to be limited and only one candidate at a time may 2.
You must avoid all contacts with anyone outside the examination leave.
room to avoid even the appearance or possibility of cheating.
Use black ink or dark pencil only to facilitate legible reproductions.
3.
Print your name in the blank provided on the cover sheet of the 4.
examination.
Fill in the date on the cover sheet of the examination (if necessary).
5.
6.
Use only the paper provided for answers.
Print your name in the upper right-hand corner of the first page of each 7.
section of the answer sheet.
Consecutively number each answer sheet, write "End of Category _," as 8.
appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
Number each answer as to category and numLer, for example, 1.4, 6.3.
9.
10. %ip at least three 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 literature.
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 niven. 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 examint; nly.
17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete.-
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18. When you complete your examination, you shall:
Assemble your examination as follows:
a.
(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.
Turn in all scrap paper and the balance of the paper that you did c.
not use for answering the questions.
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d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still inprogress,yoprlicensemaybedeniedorrevoked.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIOS, AND PAGE
THERMODYNAMICS
-
QUESTION 5.01 (3.00)
The reactor has been operating at 4% of rated core thermal power for several days with the main turbine on the turning gear.
Reactor pressure is 921 psig. 'he reactor operator withdraws an in-sequence control rod and reactor power increases to 4.5%.
LIST the two (2) reactivity coefficients that had the a.
greatest effect in initially turning the power increase.
(0.5)
b.
Once 4.5% power is initially obtained, w!th no further changes in control rod position or reactor recirculation behave (relative to 4.5%)gnitude of reactor power will flow, DESCRIBE HOW the ma initially during the next hour AND over the following 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of operation.
INCLUDE a brief explanation for this behavior.
(Specificvalues for changes in magnitude are not required).
(1.0)
If the same amount of reactivity were to be inserted at c.
these same initial conditions but LATER in the fuel cycle, STATE whether the reactor period observed during the transientwouldbe(GREATERTHAN,LESSTHAN,orEQUALTO)
the period observed earlier in the fuel cycle.
SUPPORT your answer with an explanation.
(1.5)
QUESTION 5.02 (2.50)
The reactor is operating at 95% of rated core thermal power.
Control rod 18-35 is at notch position 06. Control rod 22-35 is at notch position 38,
'
STATE which of these two control rods is most likely to a.
produce the greatest increase in core thermal power if withdrawn one additional notch.
(0.5)
b.
STATE which of these two control rods is most likely to produce the SMALLEST CHANGE in core RADIAL power distribution if withdrawn one notch.
INCLUDE a brief explanation to support your answer.
(1.0)
The reactor operator withdraws a shallow control rod one c.
notch and notes that total core power decreased slightly in response to the rod movement.
Briefly EXPLAIN HOW such a response is possible.
(1.0)
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5.
THEORY OF NUCLEAR POWER PLAN 1 OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS
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QUESTION 5.03 (2.00)
You have been informed that the measured Shutdown Margin (50M) for your reactor is four percent delta K/K. The average indication is 200 cps on the SRM instrumentation.
Control rods are withdrawn and the average count rate on the SRM's increases to 1000 cps. WHAT is the new Shutdown Margin?
(SHOW all work)
(2.0)
.
QUESTION 5.04 (2.00)
.
Thereactoriscriticalonapositives]tablepericaof100
.
seconds.
a.
WHAT is the doubling time if period remains constant?
(0.5)
b.
If IRM Channel
"B" indicates "10" on range 3, HOW long can IRM Channel "B" be left on range 3 before an IRM upscale rod block is initiated by IRM Channel "B" if period remains constant?
(SHOW all work)
(1.5)
.
QUESTION 5.05 (2.50)
SELECT the appropriate response for each of the following statements concerning Control Rod Worth:
a.
(MORE/LESS) control rods would need to be pulled to make the reactor critical at 545 deg F, as opposed to 140 deg F.
(ASSUME initial rod sattern identical, and the same sequence is used in )oth cases.)
(0.5)
b.
An INCREASE in the Void Fraction will result in a/an (INCREASE / DECREASE) in individual control rod worth.
(0.5)
c.
Control rod worth will (INCREASE / DECREASE) with an INCREASE in moderator temperature.
(0.5)
d.
Control rod worth at End of Cycle would be (GREATER /LESS)
than at the Beginning of Cycle.
(0,5)
e.
Control rod worth will (INCREASE / DECREASE) as the adjacent control rods are withdrawn.
(0,5)
(***** CATEGORY 05 CONTINUED ON HEXT PAGE *****)
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS, r
.
QUESTION 5.06 (3.00)
An EHC load reject occurs at 100% core thermal power with the EHC system aligned for normal 100% power generation. DESCRIBE HOW and WHY the fol owing parameters respond initially AND then during the first five minutes subsequent to the opening of the generator output breaker, a.
Reactor Power (1.0)
b.
Reactor Pressure (1.0)
c.
Reactor Water Level (1.0)
QUESTION 5.07 (2.00)
The reactor is critical and a heatup is in progress.
Reactor pressure is 40 psig.
If the heatup progresses at the maximum rate allowed by procedure SP 22.001.01, "Startup-Cold Shutdown to 20 Percent," HOW long from now will it be until reactor pressure will be adequate to allow warming of the HPCI turbine steam supply piping?
(STATE all assumptions. ASSUME heatup rate is uniform.
PROVIDEacalculationtosupportyouranswer.)
(2.0)
QUESTION 5.08 (2.25)
Concerning Technical Specification safety limits:
STATE the basis for the reactor coolant system Technical a.
Specification pressure safety limit of 1325 psig.
INCLUDE a discussion concerning the considerations made in selecting this value.
(1.5)
b.
STATE which core thermal limit is specifically limited as a
"safety limit" during FULL POWER operation.
(0.75)
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I 5.
THEORY OF_ NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS QUESTION 5.09 (3.00)
Reactor recirculation pump speed is 30%.
Each reactor recirculation pump motor is drawing 0.18 MW of power, If the reactor operator increases reactor recire pump a.
speed (both )um)s) to 45%, HOW much power will be consumed by soti pumps at 45% speed?
(Using basic pump laws, PROVIDE a simple calculation and express your (2.0)
answer in MW.)
b.
The process computer does not measure actual steam flow rate in the heat balance calculation of core thermal power because of inherent inaccuracies in the measurement of steam flow.
STATE HOW steam flow rate is determined
>y the process computer for the determination of core
,
i (1.0)
thermal power.
.
QUESTION 5.10 (1.25)
The reactor is in shutdown cooling with a bottom head drain temperature of 180 deg F.
The only operating residual heat removal pump trips.
(Reactor recirculation pumps are not running.)
SELECT from the following the MINIMUM elevation at which a.
reactor water level must be maintained to ensure a FLOWPATH for natural circulation exists (ASSUME no reactor recirculation pumps or RHR pumps are running).
(0.75)
(1) The elevation where the MAIN STEAM LINES are submerged.
(2) The elevation where the STEAM SEPARATOR is submerged.
(3)
The elevation where the STEAM DRYER is submerged.
(4) The elevation where the CORE TOP GUIDE is submerged.
In addition to thermal stresses that stratification could b.
cause on the reactor vessel and components, STATE the major concern associated with stratification in this condition.
(0.5)
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLVIDS, AND PAGE
THERMODYNAMICS
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QUESTION 5.11 (1.50)
i ANSWER the following questions concerning "CRITICAL POWER."
a.
DEFINE "Critical Power."
(0.75)
,
b.
With the reactor at power, WHICH of the following conditions would tend to DECREASE the Critical Power level assuming all other variables remain unchanged?
Reactor pressure is INCREASED.
...
Total cSre flow is INCREASED.
?
Inlet suDcooling is INCREASED.
(0.75)
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(*****
END OF CATEGORY 05 *****)
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6.
PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE
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.
QUESTION 6.01 (3.00)
,
Concerning the Emergency Diesel Generators (G-101, 102, 103):
If a diesel generator is supplying its associated bus and a.
control power for the generator output breaker is lost, WILL that breaker trip open if a fault then occurs causing
generator breaker current to exceed its instantaneous
.
overcurrent setpoint (YES/NO)?
(0.5)
b.
LIST three (3) conditions that will automatically trip an emergency diesel generator if a valid LOCA signal is
.
present with the diesel mode selector switch in
"REMOTE."
(D0 NOT include a manual stop.)
(1.5)
t If an emergency diesel generator is running with the diesel
!
c.
mode selector switch in REMOTE and ALL starting air is i
completely lost, STATE whether that emergency diesel generator WILL or WILL NOT shutdown in response to the following conditions.
,
1.
The operator depresses the control room manual stop pushbutton for that diesel.
(0.5)
2.
A valid automatic shutdown signal condition occurs for that diesel.
(ASSUME an AUTO-START condition does not exist).
(0.5)
QUESTION 6.02 (1.50)
The reactor is operating at rated core thermal power and rated total core flow. Both reactor recirculation pumps trip.
The reactor does not scram,
,
STATE the approximate power level at which the reactor a.
will stabilize.
(Refer to Figure 6.02 as necessary.)
(0.5)
b.
Recirculation loop temperatures must be within 50 deg F of each other prior to startup of either reactor recirculation pump.
STATE the basis for this limit placed on the temperature differential.
INCLUDE in your statement the vessel components / regions that are most (1.0)
limiting.
(""* CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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QUESTION 6.02 (figure)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
.
QUESTICN 6.03 (2.00)
The reactor is shutdown.
Both loops of RHR inadvertently initiate in the LPCI mode and inject at rated flow. ANSWER the following questions concerning fuel zone level instrumentation and indication at these conditions.
a.
Fuel zone level indication would indicate (UPSCALE /
00WNSCALE).
(0,5)
b.
Fuel zone instrument sensed differential pressure would (INCREASE / DECREASE /NOTBEAFFECTED).
(0.75)
c.
Fuel zone instrument variable leg pressure subsequently (INCREASES / DECREASES / REMAINS THE SAME) as a result of
LPCI flow.
(0.75).
.
QUESTION 6.04 (1.75)
The reactor has been operating for a month at rated core thermal powar. A loss of all AC power occurs causing the reactor to scram. The decision to implement the steam condensing mode of-RHR has been made.
STATEwhetherthesteamcondensingcapabilityoftwo(2)
a.
RHRheatexchangers(IS/ISNOT)adequatetoaccommodate ALL of the reactor's decay heat immediately after the scram.
(0.5)
in The"A"loopofRHR.andtheRCICpumparenowoperating(HX)
b.
the steam condensing mode. The "A" RHR Heat Exchanger ressure control valve (PCV-003A) and level control valve p(PCV-007A) controllers are in AUTOMATIC.INITIALLY ALL RCIC pump flow is condensate from the "A" RHR HX.
If the reactor operator then doubles RCIC pump flow to rated flow, STATE whether the RCIC pump WILL or WILL NOT cavitate.
(1.25)
EXPLAIN your answer.
(**"* CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 6.
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a QUESTION 6.05 (3.00)
With regard to Low Pressure Coolant Injection (LPCI) system, WHAT signals cause an automatic initiation of LPCl?
(1.0)
a.
(00NOTincludeamanualinitiation.)
DESCRIBE the interlocks which must be satisfied in order to b.
divert injection from the reactor to containment spray with a LPCI initiation signal present with the RHR inboard (1.0)
injectionvalves(MOV-037A/B)OPEN.
LISTthetwo(2)modesofRHRsystemoperationthatrequire c.
operator action to realign the system for LPCI operation if (1.0)
a valid LPCI initiation signal occurs.
QUESTION 6.06 (3.00)
,
Concerning the Standby Liquid Control (SBLC) System:
!
LIST four (4) SBLC System indications available in the (1.0)
a.
control room to confirm SBLC initiation / injection.
b.
EXPLAIN WHY a too rapid SBLC system injection rate is (1.0)
undesirable.
DESCRIBE WHERE the SBLC system physically discharges c.
in the reactor vessel relative to the core plate.
(0,5)
(ABOVE/BELOW)
if running, automatically trip on WILL the SBLC pump, k low level condition (YES/NO)?
d.
(0.5)
any SBLC storage tan
.
.
!
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PAGE 10
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 6.
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QUESTION 6.07 (2.75)
Concerning operation of the APRM flow units:
If the OUTPUT of one APRM flow unit fails UPSCALE, SELECT WHICH one of the following automatic response should occur.
(1.0)
a.
rod block but no half scram half scram but no rod block rod block and half scram no rod block and no half scram, If the OUTPUT of one APRM flow unit fails DOWNSCALE, SELECT b.
(1.0)
WHICH one of the following automatic response should occur.
rodblockputnohalfscram half scram: but no rod block rod block and half scram no rod block and no half scram.
From the following list, SELECT the process flow that is l
c.
actually measured by the flow transmitters that provide
-
flow signals to the Recirculation Flow Units.
total core flow recirculation loop "Driving" (lowpump) flow (0.75)
recirculation loop jet pump f
!
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I QUESTION ~6.08 (2.50)
A loss of offsite power has occurred concurrently with a valid loss of coolant accident (LOCA) signal. Diesel generator G-101 properly autostarts.
SPECIFY the proper order in which the following components a.
sequence onto emergency bus 101.
INCLUDE the correct time delay:
core spray pump I
RBSYS/CRAC water chiller (
RHR pump (1.5)
I service water pump For the component start time delays, STATE the EVENT in the b.
emergency diesel start / load sequence that initiates the timing sequence (i.e., serves as "time zero" for the time (1.0)
delay setpoint).
(***** CATEGORY 06 CONTINUE 0 ON NEXT PAGE
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PAGE 11 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
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QUESTION 6.09 (2.50)
The Reactor Building Normal Ventilation System (RBNVS) is in service with a normal operating supply and exhaust fan lineup.
The reactor operator must shift operating exhaust fans to allow maintenance, DESCRIBE which RBNVS fan (s) and/or damper (s) will a.
automatically respond to attempt to control reactor building to outside air differential pressure when the additional exhaust fan is STARTED.
INCLUDE HOW they (1.5)
respond.
b.
If reactor building internal pressure inadvertently becomes greater than outside air pressure during this shift of exhaust fans, DESCRIBE HOW the following components will respond (TRIP / REMAIN RUNNING):
(1)
previously running reactor building supply fans (0,5)
(2)
previously running reactor building exhaust fans (0.5)
l QUESTION 6.10 (1.50)
Concerning the suppression pool to drywell vacuum breakers If two suppression pool to drpell vacuum breakers in i
a.
series (e.g., RV93A and RV93B) were stuck open DESCRIBE HOWthiscouldcauseprimarycontainmenttofallifa
'
(1.0)
LOCA were to occur.
b.
CAN these valves be operated from the control room?
(0,5)
(YES/NO)
.
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12
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QUESTION 6.11 (1.50)
The reactor is in Operational Condition 2 with the reactor water The RWCU outboard containment cleanup (RWCV) sy(stem in service.1G33*M034) suddenly auto-isolates. The RW isolation valve inboard containment isolation valve (1G33*M033) remains open.
' a.
STATE AU. possible conditions (i.e., signals) that could cause this particular isolation response, assuming the isolation logic for BOTH valves has properly functioned (setpoints NOT required).
(0.5)
.
b.
Assuming 1G33*M034 remains closed, STATE WHAT sampling requirements are now imposed to meet Technical (1.0)
Specification Requirements.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13
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RADIOLOGICAL CONTROL
-
QUESTION 7.01 (2.50)
Procedure SP 22.001.01, "Startup-Cold Shutdown to 20 Percent,"
places administrative restrictions upon reactor operation.
Concerning these restrictions:
a.
STATE the minimum permissible stable period.
(0.5)
b.
STATE the purpose for limiting the maximum permissible control rod drive (CRD) hydraulic system charging water header pressure to 1510 psig.
(1.0)
c.
STATE the maximum allowable interval (time) at which heatup rate must be verified to be within limits.
(0.5)
d.
STATE the purpose for ensuring PRESSURE SET is set above reactor pressure before condenser vacuum increases above 7" Hg.
(0.5)
QUESTION 7.02 (3.00)
The reactor has been operating for several days at 4.5% of rated i
l power. The shift chemist notifies you that primary coolant chemistry analysis indicates that the specific activity is 0.3
'
microcuries per gram dose equivalent I-131.
,
a.
STATE ALL immediate actions, if any, that you are required to take.
(1.0)
b.
If this level of activity in the primary coolant persists for greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, procedure requires the reactor be shutdown with MSIVs closed.
STATE the reason for requiring MSIVs to be closed.
(1.0)
c.
STATE the two (2) symptoms of fuel cladding failure that require the immediate actions of SP 29.008.01, "Fuel Cladding Failure."
(1.0)
)
("*** CATEGORY 07 CONTINUED ON NEXT PAGE **"*)
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 14 RADIOLOGICAL CONTROL
QUESTION 7.03 (2.50)
The reactor was in hot standby with a bottom head drain temperature of 520 deg F.
The high pressure coolant injection (HPCI) system auto initiated on a valid initiation signal while the reactor core isolation cooling (RCIC) system remained in standby. Based upon the responses of these two systems alone, SPECIFY WHICH entry condition (s) was/were met AND also WHICH procedure (s) should have been entered. ASSUME the HPCI and RCIC systems are properly aligned in standby for automatic initiation and are fully operable.
(2.5)
.
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QUESTION 7.04 (2.50)
!
i In accordance with SP 12.012.01, "Radiation Work Permits" (RWPs),
LISTtheTWO(2)individualsbytitlethatcansignfor a.
final approval of an RWP.
(1.0)
b.
STATE the color of a correctly posted RWP copy at the job site.
(0.5)
To pe'rform his rounds, an equipment operator requires c.
access to an unlocked room that has been designated as a "Radiation Area" by the health physics department.
STATE whether an RWP WILL or WILL NOT have to be INITIATED, (ASSUME the health physics department has determined there is no loose surface or airborne contamination.)
(0.5)
d.
STATE the TYPE of RWP used by operators in performing surveillances during routine operating conditions.
(0,5)
QUESTION 7.05 (2.00)
Adequate core cooling is one of the safety functions that the emergency procedures strive to maintain.
If reactor water level cannot be maintained above the top of the active fuel, LIST in the order of preference the remaining two (2) fonns (mechanisms)
of core cooling that the emergency procedures will attempt to (2.0)
establish.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE "***)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL
-
QUESTION 7.06 (1.75)
O Procedure SP 22.004.01, "Operation Between 20% and 100% Power,"
in the "Limitations and Actions" section, requires that PCIOMR (Preconditioning Interim Operating Management Recommendation) be followed.
a.
STATE WHAT adverse condition can occur if PCIOMR is not followed at high power.
(0.5)
b.
STATE WHO is responsible for monitoring and supervising the details of the fuel preconditioning process.
(0,5)
c.
BRIEFLY DESCRIBE the fuel preconditioning process.
(0.75)
QUESTION 7.07 (3.00)
a.
The reactor is operating at rated core thermal power. The main condenser low vacuum alarm ("COND VACUUM LO") is received.
1.
STATE the immediate action (s) required by emergency procedure SP 29.012.01, "Loss of Condenser Vacuum."
(0.6)
2.
LIST the four (4) automatic actions that will be a low vacuum condition.
INCLUDE initiated by(Assume a complete loss of condensor setpoints.
vacuumoccurs.)
(1.6)
b.
LIS1 two (2) reasons WHY procedures do not allow condenser air removal pumps to maintain condenser vacuum when reactor power is above 4%.
(0.8)
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(***** CATEGORY 07CONTINUEDONNEXTPAGE*****)
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_ __ _ _______ ________ _______ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- _ _ __
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7.
PROCEDURES - NORMAL. AP'TW'9, EMERGENCY AND PAGE 16 RADIOLOGICALCONTRf,"~
'
.
'
.
QUESTION 7.08 (3.00)
Concerning Primary Containment:
The attached figure PC/T-1 is an excerpt from emergency a.
procedure SP 29.023.03, "Primary Containment Control."
be exceeded (design value not required) parameter c0uld STATE WHICH primary containment design if containment sprays were initiated with initial drywell temperature at 300 deg F and initial drywell pressure at 20 psig.
INCLUDE in your discussion HOW such a condition could occur if containment sprays were initiated.
(2.5)
b.
For Operational Condition 1, under WHAT condition is the Technical Specifications maximum allowable drywell pressure i
less than 1.69 psig?
(INCLUDE a numerical value for this l
condition.)
(0.5)
QUESTION 7.09 (2.00)
The reactor is operating at rated core thermal power.
The reactor building closed loop cooling water (RBCI.CW) head tank
"low-low" level alann ("RBCLCW HD TK A(B) LEV LO-LO") is received.
RBCLCW has isolated from all nonsafety loads.
a.
LIST ALL immediate actions.that are required by SP29.017.01, "Loss of RBCLCW."
(1.5)
b.
STATE HOW LONG continued operation of the Reactor Recirc HG sets is allowed in this condition.
(0.5)
.
("*** CATEGORY 07 CONTINUED ON NEXT PAGE "***)
,-
SWOREHAM QueStton 7.08
'
.
.
F
ricVRt PC/T-1 CRYWCLL SPRAY IN!T!ATION L!n!T i
'
400
.
,
"
w
350 -
J
.
.
{see~-
f UNSArt
-
-
Wh 250-
%
~
.
y nee-
,
,
y f
sArc
-
3 5e -
f.
-
i i
...
.
.
.
.
.
i e
le
28 25
35 MS M5
55 Se DRYWELL PRESSURE. PSIC (
)
e
.
!
e e
o I
,
e i
'
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____________ __.
d
.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 17 RADIOLOGICAL CONT 60L
.
QUESTION 7.10 (2.75)
A loss of ALL AC power has occurred. The "immediate actions" of emergency procedure SP 29.015.02, "Loss of All AC Power" are complete. The "subsequent actions" are now being performed.
, Reactor water level has been stabilized at +30" using the RCIC system alone (HPCI has been secured in accordance with the procedure).
a.
STATE the reason WHY the procedure instructs the operator to depressurize the reactor as quickly as possible.
(1.0)
-
b.
STATE the reason W'HY the procedure instructs the operator to secure the RCIC vacuum pump though RCIC is feeding the reactor pressure vessel to maintain water level.
(1.0)
STATE one (1) reason WHY the procedure instructs the c.
operator to PREVENT automatic suction transfer of the RCIC pump to the suppression pool.
(0.75)
l l
(***** END OF CATEGORY 07 ***")
.
...
.-
p ADMINISTRATIVE PROCEDURES, CON 0!TIONS, AND LIMITATIONS PAGE 18
.
QUESTION 8.01 (3.00)
Standing Order Number 30 stovides specific guidance concerning reactor operation while tie 5% Low Power License is in effect.
Concerning the requirements of this standing orders a.
STATE the MAXIMUM allowable "average core power" for ANY transient.
(0,5)
b.
DESCRIBE HOW "average core power" is to be determined from plant instrumentation.
(1.0)
c.
If the main turbine generator is being rolled for testing or is synchronized to the grid STATE ALL restrictions on reactor power.
SPECIFYWHICHIndicationsaretobeused for reactor : power datermination, INCLUDING the locations of the indic!ations to be used.
(1.5)
i QUESTION 8.02 (1.00)
The Technical Specifications specify the maximum allouable mismatch between reactor recirculation pump speeds. 5 TATE the Technical Specification BASIS for administratively controlling this mismatch in speeds.
(1.0)
,
(**"* CATEGORY 08 CONTINUE 0 ON NEXT PAGE *****)
_____ _ ___ _ _ __.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19
.
QUESTION 8.03 (3.00)
'
Procedure SP 21.001.01, "Shift Operations," requires that only
"ACTIVE" licensed operators may assume the watch.
a.
SPECIFY ALL quarterly watch standing requirements that must be met for a licensed operator to meintain his
"active" status.
(1.0)
b.
Is a newly licensed individual considered "active" if he has not stood the required number of watches to ciaintain an "active" status?
(YES/NO)
(0.5)
c.
When the reactor changes frem Operational Condition 3 to Operational Condition 4, STATE which licensed
individual (s), by title is/are no longer required to
-
.
be a part of the shift complement.
(1.0)
d.
STATE which document must be signed by an individual if he is to receive credit towards maintaining his "active" status when he completes a watch in the capacity of his license.
(0.5)
QUESTION 8.04 (2.50)
,
Concerning Procedure SP 12.011.01, "Station Equipment Clearance Permits (SECPs)":
a.
STATE to which individual (s) the Watch Engineer con delegate his authority to APPROVE an SECP.
(0.5)
b.
STATE whether or not the second individual performing the
>
independent verification of the placen nt of hold-off tags for an SECP is required to accompany the individual perfoming the SECP.
(0.5)
c.
DESCRIBE the condition under which the independent verification of an SECP on SAFETY related systems /
components may be waived.
(0.5)
d.
STATE WHICH individual has the authority to WAIVE the requirement for an independent verification of an SECP involving SAFETY related systems / components.
(0,5)
e.
STATE the criterion for detemining whether or not a lifted lead requires an SECP hold-off tag in addition to a lifted lead /jumpertag, (0,5)
.
(**"* CATEGORY 08 CONTINUED ON NEXT PAGE "***)
.a.
s
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 20
'
.
QUESTION 8.05 (3.00)
Concernin3 Procedure Sc 12.009.03, "Report of Abnormal Conditions (RAC) and Limiting Conditions for Operations (LCO)":
a.
STATE WHICH individual (s) may INITIATE an RAC.
(0,5)
b.
STATE WHICH individual is to perform the shift technical advisor's (STA) review of an RAC if the STA is not required to be on shift and no STA is assigned to the shift.
(0.5)
c.
If a motor opera'ed valve (MOV) governed by Technical Specifications is manually torqued and was consequently declared inoperable, STATE WHAT mest be done to the valve to return it to an operable status.
(0.5)
I d.
For the following situations, STATE whether a one-hour NRC notification (IS REQUIRED /IS NOT REQUIRED):
1.
An inadvertent initiation of HPCI cccurs that was NOT caused by a valid initiation signal.
Some water was injected to the reactor pressure vessel but the transient did not reauire nor result in a scram.
(0.5)
2.
An EHC pressure regulator failure causes a high reactor pressure scram.
Reactor pressure recorders indicate that reactor pressure peaked for several seconds at 1355 psig during the transient.
(0.5)
3.
An excessive primary leak rate leads the watch engineer to declare an "Unusual Event."
(0.5)
l (***** CATEGORY 08CONTINUEDONNEXTPAGE*****)
l l
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 21
.
QUESTION 8.06 (3.00)
Reactor power is 10%. The reactor mode switch is in "STARTUP."
Control rod scram time testing has been completed for all control rods. A review of the testing results reveals that control rods 42-43, 34-35, and 22-27, all three of which are currently at position 48, had scram insertion times (as defined e
in the Technical Specifications) of 7.1 seconds.
ASSUME:
1.
All 137 control rods were initially considered operable.
2.
All control rods other than 42-43, 34-35, and 22-27 had scram insertion times of less than 5.0 heConds.
3.
All Technical Specifications requirements for averaged scram times are satisfied.
Referring as necessary to the attached excerpt from the SNPS Technical Specificaticas and the attached diagram of the rod coordinate matrix, ANSWER the following questions:
-
a.
DESCRIBE WHAT action (s) are required to be taker, oy Technical Specifications.
REFERENCE the Technical Specification (s) (by number) that require (s) the action (s).
(2.5)
b.
Based solely upon the conditions above, STATE whether the reactor mode switch (CAN/CANNOT) be taken to tne "RUH" position without violating Technical Specifications, once all required actions have been taken.
(0.5)
.
QUESTION 8.07 (2.00)
Using the Emergency Plan Implementing Procedure (EPIP 1-0)
CLASSIFY the fol. lowing events.
INCLUDE the classification identifier number.
(This is NOT the event category number.)
a.
The RPS system initiates a full scram.
Reactor shutdown does not occur. The standby liquid control system is initiated and successfully terminates the transient.
(1.0)
b.
A total loss of service water occurs (both' loops). The reactor is consequently shutdown to meet rechnical Specifications requirements.
(1.0)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
.
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8.
ADMINISTRATIVE-PROCEOURES, CONDITIONS, AND LIMITATIONS NGE 22
.
.
QUESTION 8.08 (2.00)
Concerning Procedure SP 12.006.01, "Station Procedures -
Preparation, Review, Approval, Change, Review, and Cancellat;on":
a.
If a change is needed to correct a procedure deficiency to make the procedure conform to approved station design documents, is the resulting Station Procedure Change Notice (SPCN) classified as a "MAJOR" or "MINOR" change?
(0.5)
b.
STATE the maximum amount of time a Temporary Plant Change Note (TPCN) is valid once it is in effe (0.5)
c.
STATE WHICH individual, by title, must give the final approval to make a TPCN involving safety related equipment a permanent procedure revision.
l (0.5)
i d.
STATE HOW MANY individuals holding SR0 licenses MUST approve a TPCN BEFORE the TPCN can be placed into effect (i.e.,used).
(0.5)
QUESTION 8.09 (2.00)
As an Emergency Director for Shoreham's Emergency plan:
a.
STATE three (3) duties that cannot be delegated.
(1.5)
b.
STATE the time limit for notifying New York State and Suffolk County officiais once an emergency event has
,
been declared.
(0.5)
QUESTION 8.10 (1.50)
Concerning the basis for Technical Specifications limits upon primary containment parameters during normal operation:
a.
STATE the Technical Specification basis for the maximum allowable drywell air temperature.
(0.75)
b.
STATE the Technical Specificat. ion basis for the maximum allowable suppression pool water level.
(0.75)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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b 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23
.
QUESTION 8.11 (2.00)
Concerning Procedure SP 21.003.01, "Operations Reports":
a.
DOES a reactor trip report have to be initiated for a MANUALUNPLANNEDscram(YES/N0)?
(0.5)
b.
Prior to restart, MUST the ROC review the reactor trip
'
report if the cause of the scram has not been positively identified (YES/NO)?
(0.5)
c.
STATE WHAT the watch engineer's final signature on a
-
Reactor Trip Report indicates.
(0.5)
d.
STATE WHICH individual, by title, must perform the FINAL review of this re) ort as one of the prerequisites for commencing the su) sequent reactor startup.
(0.5).
.
e (***** END OF CATEGORY 08 *****)
(*************ENDOFEXAMINATION***************)
.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIOS, AND PAGE 24 THERMODYNAMICS
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 5.01 (3.00)
a.
1.
void coefficient [+0.25]
2.
Doppler coefficient [+0.25]
b.
Reactor power wil' first increase greater than 4.5% [+0.25]
as xenon is initially depleted faster than it is produced than 4.5% [+0.25] power will subsequently stabilize at lessas the productio
[+0.25]. Reactor of iodine restores xenon concentration to a new higher equilibrium value [+0.25].
c.
(Reactor period would) LESS THAN [+0.5].
Increased core I
age results in a higher fraction of core power being
!
Pu-239 [+0.25] which reduces the value of BETA-i produced by[+0.5].From the period equation, as BETA-EFFECTIVE EFFECTIVE decreases, for a given reactivity insertion, the resultant reactor period decreases [+0.25].
REFERENCE 1.
Shoreham: HL-900-SHi, Lesson 11, L0 CB and Lesson 15, L0 CD, CH.
2.
GE: Reactor Theory, Chapters 3, 4, and 6.
292003K106 292004K114 292006K105 292006K106
...(KA'S)
__
.
.h
'
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 THERMODYNAMICS
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 5.02 (2.50)
a.
(Control rod) 18-35 [+0.5]
b.
(Control rod) 22-35 [+0.33]. The rod shadowing effects of adjacent rods [+0.33;) will be more significant in dampening radial power because of the higher local control rod density [+ relative to the control rod tip (of control rod 22-35)
0.34].
,
c.
When a shallow rod is withdrawn a notch, the void fraction in the adjacent fuel channels increases throughout the
~
entire boiling length [+0.33]. Though local power will
-
i
.
l increase where the rod tip was withdrawn [+0.33), it is i
possible for the negative reactivity of the increased voiding to be the dominate effect (causing net total power to decrease) [+0.34].
REFERENCE 1.
Shoreham: HL-900-SH1, Lesson 13, L0 CB, CC.
2.
GE: Reactor Theory, Chapter 5.
292005K104 292005K112 292008K119
...(KA'S)
,
ANSWER 5.03 (2.00)
)M = 1-Keff; CR1 (1-Keff 1))= CR2 (1-Keff 2)
" eps) (0.04) = (1000 cps SDM
.1 = 0.008 = 0.8 percent
[+2.0]
Y REFERENCE 1.
Shoreham: HL-900-SH1, lesson 16, LO CB.
292002K113
...(KA'S)
,
L
-
- - -. - - -. -. -.. ~.. - _., _,.., _.. - -_.., - - - - - - - -
-
.
.
..
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIOS, AND PAGE 26 THERMODYNAMICS
-
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 5.04 (2.00)
a.
Doubling time = 100/1.44 = 69.4 seconds
[+0.5]
b.
Rod block will occur at 34/40 of full scale on range 3 (also allow 35/40 of full scale)
[+0.5]
P(o) = 10; P(t) = 34 P(t) = 10e**(t/ period)
34 = 10e**(t/100 sec)
t = 100 in(34/10) = 122 sec (2 min 2 sec)
[+1.0]
(ALTERNATEANSWER: 125 sec or 2 min 5 see IF P(t) = 35 was used)
REFERENCE 1.
Shoreham: HL-602-SH1, L0 B.2.
2.
Shoreham: HL-900-SH1, Lesson 15, L0 CA.
215003K40)
292003K108 292003K109
...(KA'S)
ANSWER 5.05 (2.50)
a.
more b.
decrease c.
increase d.
less e,
increase
[+0.5] each
,
REFERENCE 1.
Shoreham: HL-900-SH1, Lesson 12, L0 CG.
292005K109
...(KA'S)
,o-
.
-
.
PAGE 27 THE0RY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND 5.
.
THERMODYNAMICS _
-88/05/16-NRC REGION I
' ANSWERS -- SHOREHAM ANSWER 5.06 (3.00)
Reactor power will rapidly increase due to the pressure increase [+0.5]. Power will then decrease due to the TCV a.
fast closure scram [+0.5].
Reactor pressure wili rap] idly increase due to the rapid b.
Pressure will then decrease closure of the TCVs [+0.5.
due to the scram and the opening of the bypass valves which will then attempt to maintain reactor pressure at 920 psig
[+0.5].
Reactorwaterlevelwillinitiallydropduetocollapsing The feed control system will respond to c.
of voids [+0.5].
increase level and level shouldi hen rise to the level t
controller setpoint (level may overshoot causing feed pumps to trip) [+0.5].
.
REFERENCE 1.
Shoreham: HL-900-SH1; HL-657-SH1, L0 C3.
241000K101 241000K102 241000K103
...(KA'S)
ANSWER 5.07 (2.00)
.
Assumptions:
1.
HPCI isolation setpoint = 100 psig
[+0.5]
2.
maximum allowable heatup rate = 90 deg F/hr
[+0. 5]
3.
40 psig = 54.7 psia; Tsat = 287 deg F 100 p[+sig = 114.7 psia; Tsat = 338 deg F 0.5]'
Time to clear setpoint = ((338 - 287)deg F)/(90 deg F/hr)
= 51 deg F/(90 deg F/hr)/- 3 minutes)
,
t
= 0.57 hrs (34 minutes +
'
[+0.5]
REFERENCE HL-901-SH1, Lesson 2, L0 C8; HL-202-SH1, L0 CG.
<
1.
Shoreham:
I, 206000K402 293003K123
...(KA'S)
t
- -
'
..
, *
'
PAGE 28
. -
5. -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS-88/05/16-NRC REGION I ANSWERS -- SHOREHAM
)
ANSWER 5.08 (2.25)
The basis is to ensure the reactor coolant pressure boundary integrity is maintained [+0.5].
ASME code allows a.
110% (1375 psig) overpressurization of design [+0.5].
Because reactor pressure is measured at the steam dome, 50 asi is subtracted from 110% design value to account for the lead of water above the lowest point in the vessel [+0.5].
b.
minimum critical power ratio (MCPR)
[+0.75]
REFERENCE 1.
Shoreham: HL-904-SH1, lesson 2, L0 CB, CC.
2.
Shoreham: Technical Specifications, Section 2.0.
293009X105
...(KA'S)
ANSWER 5.09 (3.00)
fractional increase in reactor recirc pump speed:
a.
1.
45/30 = 1.5
[+0.25]
(pumplawsstatepowerincreasesbythecubeofthe 2.
(1.5)**3 = 3.375 fractional increase in speed):
[+0.75]
p(0.18 Kd x 2)(3.375) pumps at 45% speed =
ower consumed by 2 3.
= 1.21 Kd
[+1.0]
The process comp [+ uter takes steam flow to be the sum of0 b.
feedwater flow flow [+0.3].
REFERENCE 1.
Shoreham: HL-901-SH1, Lesson 3, L0 CB.
'
,
2.
GE: HTFF, Chapter 7.
293006K108 293007K111 293007K113
...(KA'S)
l
i
-. _ -,
-.
-.-.
. _ _.
-
-. - - - - - - - -
._- -
-
._.
.
o
.
.,
.
PAGE 29 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND 5.
THERMODYNAMICS
,
ANSWERS -- SH0REHAM-88/05/16-NRC REGION I ANSWE 5.10 (1.25)
7, a.
(t [+0.75]
.
b.
-The reactor pressure vessel could generate steam due to boiling in the upper region though coolant temperature indications indicate adequate subcooling.
[+0.5]
.
REFERENCE 1.
Shoreham: HL-901-SH1, Lesson 5, L0 CB.
I 2.
GE: HTFF, Chapter 8.
!
293008K134 293008K135
...(KA'S)
i ANSWER 5.11 (1.50)
The assembly (bundle) power that would cause the onset of a.
transition boiling [+0.75].
b.
(1)
[+0.75)
REFERENCE 1.
Shoreham: HL-904-SH1, Lesson 1. L0 CA, CF.
293009K117 293009K122 293009K123 293009K124
...(KA'S)
,
l
.. - _,,.,.
.
.
.-
.. - -
. - - -, -.., - -, - -,, - - -. --
-.
--
--
-
.
,
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PAGE 30 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
,.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
'
.
ANSWER 6.01 (3.00)
a.
no [+0.5]
b.
1.
overspeed [+0.5]
2.
generator phase differential [+0.5]
3.
generator overcurrent
[+0. 5]
' 0.5'
c.
1.
will not
+
2.
will not l+0.5[
REFERENCE 1.
Shoreham: HL-307-SH1, L0 A, C.4, D.
264000K106 264000K601
...(KA'S)
ANSWER 6.02 (1.50)
approximately 48% [+0.5] (allow 43% - 51%)
a.
To prevent undue thermal stress [+0.34][on the vessel b.
nozzles [+0.33] and bottom head region +0.33].
REFERENCE 1.
Shoreham: HL-120-SH1, L0 F.
202001K102 202001SG1 202001SG5
...(KA'S)
ANSWER 6.03 (2.00)
a.
upscale [+[0.5] ]
+0.75 b.
decrease c.
increases
[+0.75]
REFERENCE 1.
Shoreham: HL-621-SH1, LO I.D.5.
210000K105
...(KA'S)
t
^^
_,,-e.r,
--m,
-,
e,-
-,-
-.,,
~1
..
.
,
.,
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31
.,.
)
ANSWERS -- SH0REHAM-88/05/16-NRC REGION I ANSWER 6.04 (1.75)
a.
IS NOT [+0.5]
b.
The RCIC' pump WILL NOT cavitate [+0.25].
CST inventory will maintain RCIC pump suction pressure [+1.0].
% - w,t( bc_,
(We. Que.M+ ega-N a ui-Ce* * * + ' % e b el cMru vh pshh
"* ~ P "
REFERENCE 1.
Shoreham: HL-121-SH1, L0 CC.
2.
Shoreham: 23.121.01.
-
217000A101 217000K101 217000K105 217000SG1
...(KA'S)
a
!
ANSWER 6.05 (3.00)
Hi drywell pressure 1.69 psig OR [+0.5]
a.
Lo lo Rx water level -132.5" [+0.5]
(NOTE: Do not penalize if RPV low pressure permissive of 338 psig is also listed.)
b.
1.
' containment spr y valve manual override keyswitch in
"MANUAL"
[+0.5 valve accident control switch in containment sp] ray (until seal-in status light is lit)
2.
"MANUAL" [+0.5 c.
1.
shutdown cooling mode [+0.5]
2.
fuel pool cooling mode [+0.5]
REFERENCE 1.
Shoreham: HL-204-SH1, L0 CC.
226001K403
...(KA'S)
.
.
,
.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 32
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 6.06 (3.00)
a.
1.
squib valve continuity circuit indicator lamp extinguishes 2.
squib valve loss of continuity annunciator 3.
SBLC pump discharge pressure greater than reactor pressure 4.
SBLC pump running indication (0N)
5.
SBLC storage tank level decreasing ykukeca.s.4 g &a ag matm.
i e_
_
l Any four (4) [+0.25] each,.+1.0 maximum.
b.
A too rapid injection rate could cause insufficient mixing and uneven concentrations of boron circulating in the core
[+0.5] leading to power oscillations ("chugging") [+0.5].
c.
below(thecoreplate)
[+0. 5]
d.
No [+0.5]
REFERENCE 1.
Shoreham: HL-123-SH1, L0 B.1, F.
211000K106 211000K403 211000K405 211000K506
...(KA'S)
ANSWER 6.07 (2.75)
a.
(1)
(rodblock)
[+1.0]
l b.
(3)
(rod block and half scram)
[+1.0]
(2) (recirculation loop "Driving" flow)
[+0.75]
l c.
REFERENCE 1.
Shoreham: HL-603-SH1, LO I.C.
215005K110 215005K116 215005K607
...(KA'S)-
i (
,. _ _
_ _.
-
. - - -
__ __
-
.
_
e
-=
.
-
.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 33
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 6.08 (2.50)
- +0.2; 0.2 a.
1.
RHR pump (3)
2 second T.D.
7 second T.D.
+
core spray pump (1)(4)
2.
, 0. 2,j service water pump -
12 second T.D.
+
3.
, 0.2J RBSVS/CRAC water chillers (2)
12 second T.O.
co (Point awards above are for T.D. values only.
[+0.7] for correct order, no partial credit)
The timing sequence is initiated by:+the closing of the b.
diesel generator output breaker.
1.0]
REFERENCE 1.
Shoreham: HL-307-SH1, L0 B.1.b.
262001A304 262001K301 264000K506
...(KA'S)
.
ANSWER 6.09 (2.50)
The exhaust damper [+0.5] of the op[+erating supply fan a.
[+0.5] will modulate further open 0.5].
b.
(1) trip [+0.5]
(2) remain running
[+0.5]
.
REFERENCE 1.
Shoreham: HL-405/418-SH1, LO I.0.5.
261000K101 261000K401
...(KA'S)
ANSWER 6.10 (1.50)
Primary containment could he overpressurized [+0.5] because a.
of steam bypassing the suppression pool, pressurizing containment [+0.5].
b.
Yes
[+0. 5]
REFERENCE 1.
Shoreham: HL-654-SH1, L0 CB.
223001K405 223001K501 223001K503
...(KA'S)
]
,
.
-
.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 34
.
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 6.11 (1.50)
a.
1.
RWCU nonregenerative heat exchanger outlet (filter demineralizer inlet) temperature high [+0.25]
2.
standb liquid control (SLC) system initiation
[+0.25 b.
(In-line) conductivity must be sampled at least once every four hours [+1.0].
(continuous conductivity indication has been lost)
REFERENCE 1.
Shoreham: Technical Specifications 3.4.4.
2.
Shoreham: HL-709-SH1, L0 CE, CF, CG.
204000X404 204000K507 204000SG11 204000SGS
...(KA'S)
i f
l
!
l l
!
!
__
.
.:
.-
,
...
,
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL
ANSWERS'-- SHOREHAM-88/05/16-NRC REGION I ANSWER 7.01 (2.50)
a.
60 seconds
[+0.5]
.
b.
To prevent control rod drive mechanism damage [+0.5] during a scram [+0.5].
.c.
15 minutes
[+0.5]
.
d.
To prevent inadvertent bypass valve operation [+0.5].
REFERENCE 1.
Shoreham: SP 22.001.01; HL-106-SH1 LO J.
216000SG1 241000SG1
...(KA'S)
ANSWER 7.02 (3.00)
a.
1.
increase RWCU flow to maximum [+0.5]
increase primary coolant samp[+ ling ] frequency to atuntil specific 2.
least once every four hours 0.25 activity is below Technical Specifications limits
[+0.25]
b.
Prevents release of activity [+0.5] should a steam line rupture occur [+0.5].
h......,f,M_ _ _d...f. 44b. w hr % s%<.'
.m m -
mi u ou s.1, 6y snaii ve i s in i u m m
c.
1.#@0.2 microcuries per gram dose equivalent I-131
[+0.5]
2. g 100/E microcuries per gram [+0.5]
REFERENCE 1.
Shoreham: SP 29.008.01.
295017AX20 295017SG10
...(KA'S)
. _
-
-..
.
..
..
...
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36 RADIOLOGICAL CONTROL
.
_
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 7.03 (2.50)
1.
the entry condition was high drywell pressure >1.69 psig
[+1.0]
2.
the emergency procedures which should have been entered were:
a.
Emergency Shutdown" (SP 29.010.01)
[+0.5]
"
b.
Reactor Pressure Vessel (RPV) Control" (SP 29.023.01)
"
[+0.5]
"Primary Containment Cohtrol" (SP 29.023.03)
[+0. 5]
c.
!
REFERENCE i
1.
Shoreham: HL-944-SH1, LO I.B.
2.
Shoreham: HL-944-SH2, LO I.B.
295024SG11
...(KA'S)
ANSWER 7.04 (2.50)
[+0.5]]
a.
1.
watch engineer
[+0.5 2.
watch supervisor b.
white [+0.5]
c.
will not
[+0.5]
d.
extended (RWP)
[+0.5]
REFERENCE 1.
Shoreham: SP 12.012.01 294001K103
...(KA'S)
.
m F
w.iT--
e re
- -
T--
r-4 9+
+-
- -,
y mw=
y--
--w--
-" v-mw
-- '
v-
e
.
,
'
.
7.
PROCEOURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37 RADIOLOGICAL CONTROL ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 7.05 (2.00)
1.
spray cooling (ALTERNATE ANSWERS: steam cooling with injection of make up water to the RPV --OR-- one core spray pump spraying at rated. flow)
[+1.0]
2.
steam cooling
[+1.0]
REFERENCE
.
1.
Shoreham: HL-944-SH3.
295031EK10 295031EK30
...(KA'S)
.
.
.
ANSWER 7.06 (1.75)
a.
fuel clad cracking [+0.5] (ALTERNATE ANSWERS: pellet-clad interaction --0R-- fuel failure)
b..
the reactor engineer [+0.5] (ALTERNATE ANSWER: the Reactor Engineering Departme gg the rate of increase in reactor power (LHGR) is limited c.
(controlled)
[+0.75]
REFERENCE 1.
Shoreham: HL-904-SH1, Lesson 4, LO CB, CC.
2.
Shoreham: SP 22.004.01.
239009K136 294001A103
...(KA'S)
,
,
!
t
.
..
.,,
...
PAGE 38
'7.
PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- SH0REHAM-88/05/16-NRC REGION I ANSWER 7.07 (3.00)
commence power reduction (in accordance with a.
1.
SP22.004.01, "Operation Between 20% and 100% Power")
[+0. 6]
main turbine trip [+0.2] at 22.5" Hg vac [+0.2]
2.
a.
b.
reactor feed pump (turbine) trip [+0.2] at 20" Hg vac [+0.2]
MSIV (/ main steam drain) isolation [+0.2] at 8.5" c.
I Hg vac [+0.2]
i d.
turbinebypassvalve(TBV) isolation (+0.2]at7" Hg vac [+0.2]
b.
1.
to avoid hydrogen explosion (above 4% power) [+0.4]
2.
the level of radioactivity in the noncondensable condenser gases is significant (above 4% power) [+0.4]
(ALTERNATE ANSWER: condenser air removal pump exhaust is not treated prior to release)
REFERENCE 1.
Shoreham: SP 29.012.01.
2.
Shoreham: HL-701/714-SH1, LO E.4.
295002AK20
...(KA'S)
ANSWER 7.08 (3.00)
The design maximum suppression pool to drywell differential a.
pressure (could be exceeded) [+1.25]. The drywell could depressurize at a rate faster than the rate at which the suppression chamber to drywell vacuum relief system could equalize the resulting differential pressure [+1.25].
b.
When drywell bulk average temperature is < 110 deg F.
+0.5 REFERENCE 1.
Shoreham: HL-944-SH3, LO I.E.
295024EK30 295028EK30
...(KA'S)
.
.
,.
,
,.
,
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 39 RADIOLOGICAL CONTROL
<
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
.
ANSWER 7.09 (2.00)
a.
1.
trip the operating RWCU pump [+0.25]
2.
isolate the RWCU system from containment (close MOV's 033 and 034)
[+0.25]
3.
reduce reactor recirc pump speed to minimum [+0.25]
4.
trip both reactor recirc pumps
[+0.25]
5.
initiate emergency shutdown procedure (SP 29.010.01)
[+0.25]
6.
trip CRD pumps after all control rods are verified inserted
[+0.25]
b.
10 minutes
[+0.5]
.
REFERENCE 1.
Shoreham: SP 29.017.01.
2950185G10 295018SG11
...(KA'S)
.
i l
l l
l l
_
,. _ - - - _ _... _. -
.
... _ _
.. _ _
.--
__
..
-
.
.
..
. -
.
___
.
.
,
s
..
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40 i
RADIOLOGICAL CONTROL o
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 7.10 (2.75)
ressurization will result in suppression pool Early RPV dep(containment) temperatures [+0.25] and a.
and drywell pressures [+0.25] remaining below designed limitations
[+0.5].
(ALTERNATE ANSWER: to limit the total heat load [+0.5]
placed upon the primary containment [+0.5])
b.
(The RCIC vacuum pump is secured) to prolong [+0.5] the use of the division I battery [+0.5].
(ALTERNATEANSWER: to reduce the load [+0.5] upon the i
Division I battery [+0.5])
c.
1.
to slow the rate of containment temperature and pressure rise 2.
to avoid failure of the RCIC turbine due to high lube.
oil temperatures (Either 1. or 2. for [+0.75])
REFERENCE 1.
Shoreham: SP 29.015.02.
295003AK20 295003AK30
...(KA'S)
- - _ _
.
..
- -.
.
. -
. -.
.
..
.
,
-
-
w
,.
,
.. -
ADMINISTRITIVEPROCEDURES, CONDITIONS,ANDLIMITATIONS PAGE 41 8.
..
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 8.01 (3.00)
a.
5% [+0.5]
b.
Average core power is) the average of all APRM readings.
.
+1.0]
The hig' hest reading APRM [+0.25] is not to exceed 5%
c.
,' 0. 25, The average APRM reading (average core power)
+
.
+0.25, is not to exceed 4.75%
0.25]. APRM readings are
-
to be taken at panel 608 [+0.5][+(also allow "backpanel").
REFERENCE
!
1.
Shoreham: Standing) Order No. 30.
...(KA'S i
294001A103 ANSWER 8.02 (1.00)
The limits ensure an adequate core flow coastdown [+0.5] from either recirculation loop following a LOCA [+0.5].
(/.LTERNATEANSWER: to comply with ECCS LOCA de;ign criteria
[+1.0])
REFERENCE 1.
Shereham: Technical Specifications, Basis 3/4.4.1.
202002SG6
...(KA'S)
l
-
l
'
i
l i
.
,.
e 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 42 ANSWERS -- SH0REHAM-88/05/16-NRC REGION I
ANSWER 8.03 (3.00)
A licensed operator must perform his duties (of his li;ense a.
- R0/SRO) for seven 8-hour shifts [+0.5] or five 12-hour shif ts [+0.5].
b.
yes
[+0.5]
c.
the watch supervisor [+0.34] and either the nuclear station operator [+0.33] or the assistant nuclear station operator
[+0.33]
,
d.
the shift turnover sheet [+0.5]
REFERENCE 1.
Shoreham: SP 21.001.01.
2.
Shoreham: SP 21.002.01.
294001A103
...(KA'S)
ANSWER 8.04 (2.50)
a.
the (on-duty) watch supervisor [+0.5]
b.
is not required
[+0.5]
(Independent verification may be waived) when significant c.
radiation ex)osure could result (to the individual perfonning tie verification) [+0.5].
d.
The watch engineer (WE) has the authority (to waive the independent verification) [+0.5].
(ALTERNATEANSWER: the watch supervisor (WS) has the authority)if the WE has delegated to the WS SECP approval
,
authority l
Any lead that in normal service could be exposed to e.
voltages in excess of 130 volts (requires a hold-off tag).
[+0.5]
REFERENCE 1.
Shoreham: SP 12.011.01.
2.
Shoreham: SP 12.035.01.
!
I
l
,-
,
'
.j e.- 4(
3.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 s
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I 294000K102
...(KA'S)
ANSWER 8.05 (3.00)
a.
ANY station employee [+0.5]
~
b.
the "management on call individual"
[+0.5]
c.
the valve must be electrically stroked
[+0. 5]
d.
1.
is not required [+0.5]
' 0. 5'
2.
is required
+
3.
is required l+0.5:
-
-
REFERENCEt
!
1.
Shoreham: SP 12.009.03.
294000A103
...(KA'S)
ANSWER 8.06 (3.00)
a.
1.
control rods 42-43, 34-35, and 22-27 must all be declared INOPERABLE [+0.2] per Technical Specification 3.1.3.2 [+0.1]
2.
control rods 42-43 and 34-35 [+0.5] must be fully inserted [+0.25] and their directianal control valves disanned (either electrically or hydraulically)
[+0.25] per Technical Specification 3.1.3.1 [+0.1]
3.
control rod 22-27 must be inserted at least one notch fully' inserted [+0.25] pressure [+0.25] or else be
[+0.25] by drive water and disarmed [+0.25] per Technical Specification 3.1.3.1 [+0.1]
b.
can
[+0. 5]
REFERENCE 1.
Shoreham: Technical Specifications, 3/4.1.3.
,
!
2010035G5
...(KA'S)
l I
'
_
o e
,
O-v.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44
,
ANSWERS -- SHOREHAM-88/05/16-NRC REGION I ANSWER 8.07 (2.00)
a.
. Alert No.
[+1.0]
..
b.
(U'u ual Ever* No..F
+1.0]
at
__
faJ
.
( " #M
'O not sJ wae.wb &
v d facW REFERENCE 1.
Shoreham: EPIP 1-0.
294000A116
...(KA'S)
ANSWER 8.08 (2.00)
[+0.5] ][
a.
minor
[+0.5 b.
31 days
+0.5]
c.
plant mana er d.
one
[+0.5 REFERENCE 1.
Shoreham: SP 12.006.01.
294000A103
...(KA'S)
ANSWER 8.09 (2.00)
a.
1.
classification of the emergency 2.
directing the notification of offsite officials 3.
making offsite protective action recommendations 4.
making the decision to evacuate the site 5.
authorization for workers to exceed 10CFR20 emergency radiation exposure limits 6.
approve proposed press releases l
Any three (3) [+0.5] each, +1.5 maximum.
l b.
15 minutes
[+0.5]
REFERENCE i
!
1.
Shoreham: SP 12.002.01.
f l
o
r p.8 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 45 ANSWERS -- SHOREHAM-88/05/16-NRC REGION I
.
2.
Shoreham: EPIP 1-2.
294001A116
...(KA'S)
- ANSWER-8.10 (1.50)
-.
_.. -.
. _ _.
a.
Ensures that in the event of a LOCA [+0.25], the peak containment air temperature will not exceed the design hasis limit of 340 deg F [+0.5].
,
b.
Ensures that in the event of a LOCA [+0.25], the peak centainment pressure will not exceed the design basis limit of 48 psig [+0.5].
REFERENCE 1.
Shoreham: Technical Spec fications, Basis 3/4.6.1.7 and 3/4.6.2.
i 223001SG6
...(KA'5)
.
ANSWER 8.11 (2.00)
a.
yes [+0.5]
b.
yes
[+0.5]
c.
(The watch engineer's signature indicates) concurrence with all sections of the report [+0.5].
(ALTERNATEANSWER: All information in the report is correct.)
d.
the plant manager [+0.5]
REFERENCE 1.
Shoreham: SP 21.003.01.
294001A103
...(KA'S)
'*
r --,.. ~ -.
,. _. _..,..,...
_. _ _,. _
.,