ML20235G618
| ML20235G618 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/01/1987 |
| From: | Arildson J, Munro J GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20235G564 | List: |
| References | |
| 50-424-OL-87-01, 50-424-OL-87-1, NUDOCS 8707140320 | |
| Download: ML20235G618 (159) | |
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l ENCLOSURE 1 EXAMINATION REPORT 424/0L-87-01 Facility Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Facility Name: Vogtle Electric Generating Plant Facility Docket No.: 50-424 Written,- simulator and oral examinations were administered at Vogtle . Electric-Generating Plant near Waynesboro, Georgia.
Chief Examiner: . Odd ~
A. Arildsen 7 //[B 7 Date Signed Approved by: I7 John F. Munro,*!idction Chief Date Signed i
Summary:
Examinations on April 13-16 and 20-22, 1987.
Oral examinations were administered to 15 candidates; 15 of whom passed.
Simulator examinations were administered to 15 candidates; 14 of whom passed.
Written examinations were administered to 19 candidates; 19 of whom passed.
Based on the results described above, 8 of 8 R0s passed and 10 of 11 SR0s passed.
It was noted that 52.2% of the facility comments on the April 13, 1987 written examination involved errors or changes to the material provided by the facility to the examiners for use in examination preparation, 8707140320 870702 PDR ADOCK 05000424 V PDR
REPORT DETAILS
- 1. Facility Employees Contacted:
- J. Badgett P. Rushton
- K. Holmes
- Attended Exit Meeting
- 2. Examiners:
- J. Arildsen D. Nelson J. Moorman R. Baldwin W. Hemming N. Jensen S. Roesner L. Defferding
- Chief Examiner
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Ken Holmes, with a copy of the written examination and answer key for review. The comments made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below.
- a. R0 Exam (1) Que:, tion 1.05 Agree with facility comment. Based on additional information provided by the facility, recommended additional answer will also be accepted. Training material should be modified to reflect this information.
(2) Question 1.23 Partially agree with facility comment. Will accept additional answers recommended, however, suggested answer 1 for part c and 2 for part e are merely paraphrases of the answers already in the key, and as such, will be given full credit if they are used by the candidate.
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i (3) Question 1.24 Agree with facility comment. The question is too broad-based to '
guarantee the desired answer. The question was deleted.
(4) Question 2.04 Agree with facility comment. Based on additional information provided by the facility, recommended additional answers will also be accepted.
(5) Question 2.13a Disagree with facility comment. Since the system is designed to remove decay heat from a trip from 100 percent power and since this is an important design feature which the student should readily recognize, the answer key remains unchanged.
(6) Question 2.13b Agree with facility comment. Since in special cases this is true, the answer was changed to TRUE.
(7) Question 2.15 Agree with facility comment. Suggested additional answer will be accepted as one of the four responses. Training material should be modified to ensure completeness and consistency. i (8) Question 3.07 Agree with facility comment. Recommended answer will be the only correct answer. Training material should be modified to reflect the most recent Technical Specification information.
(9) Question 3.09a Agree with facility comment. The answer key was changed to read as follows:
"a. Data A or B failure or Parity error in A or B data j or I
Failure of A or B DRPI coils (or associated cabling)"
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(10) Question 3.18c Agree with facility comment. Based on additional information provided by the facility, recommended answer will be accepted.
Training material should be modified to reflect this information.
(11) Question 4.03 l
Agree with facility comment. Based on inadequate information provided, this question was deleted from the exam.
I THE FOLLOWING CHANGES WERE MADE DURING EXAM GRADING.
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l (1) Question 1.03 Added, "(0.1 for correct sign, 0.1 for correct units, 0.13 for correct numerical results)" to clarify point values.
(2) Question 1.05 Parenthesis adjusted to indicate fully correct answer.
i Changed to read:
Increasing boron concentration causes a decrease in thermal diffusion length, [0.50]. This results in the control rods being exposed to fewer neutrons [0.5] (and thus being less effective or worthless).
(3) Question 1.14 Answer changed to read:
SDM = reactivity for this case = Keff-1/Keff or 1-Keff/Keff, 1-0.95/0.95 = 0.05263 or 5263 pcm. (will accept either positive or negative answers)
(4) Question 1.15 Parenthesis added and points redistributed to reflect correct i credit for the answer to the question as stated. l I
Changed to read:
(At higher boron concentrations, a given boron atom is less likely to be exposed to a neutron) due to competition for those
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neutrons with the other boron atoms [1.00] or 'due to self-shielding [1.00].
(5) Question 1.20.b Added to 3 an additional correct answer, "or faulted steam I
generator."
1 (6) Question 1.22 Answer was incomplete; changed answer to add subcooling margin calculation; also changed typographical errors.
(7) Que tion 3.02.b l Changed answer to reflect that the setpoint was not required for full credit, and to allow for en additional correct answer "or PRZR HI PRESS CHANNEL ALERT."
(8) Question 3.05 Changed answer to reflect correct facility setpoint.
(9) Question 3.08 Added the fol'iowing equivalent answer to aid in grading.
" Reactor trip on a turbine trip is prevented (disabled) when three out-of-four (0.25) power ranges (0.50) are less than 50 percent. (0.25) or Reactor trip on a turbine trip is only possible (enabled) when two out-of-four (0.25) power ranges are greater than 50 percent.
(0.25)." }
(10) Question 3.10 Changed the answer to read: .
To keep the main feedwater reg valve in its responsive range.
[1.00]or To maintain a programmed differential pressure across the main feedwater reg valve. [1.00]or i
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5 To maintain the main feedwater reg valve in the region where its throttling characteristics are optimum [1.00]
(11) Question 3.13 Added the following to original part d answer:
"[0.25]"
Added to part d:
"or Diversion of turbine building sump discharge flow to the turbine building drain tank."
Added to part c:
"or
" Isolates flow to RMWST [0.25]"
(12) Question 4.08 Added a third possible correct answer.
Changed the answer to read:
Any two of the following:
ReactorCoolantSystemLeakage(18004-1)
Steam Generator Tube Leakage (18009-1)
Uncontrolled Continuous Rod Motion (18003-1)
(13) Question 4.20 Added an additional possible correct answer.
Changed the answer to read:
"Any three of the following:
- a. Temperature
- b. (Containment) Pressure
- c. Humidity / Flooding
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- d. Radiation"
- b. SRO Exam (1) Question 5.01 Agree with facility comment. Answer is not uniquely defined.
This quest'on was deleted.
(2) Question 5.05 Agree with facility comment. Based on additional information provided by the facility, recommended additional answer will also be accepted. Training material should be modified to reflect this information.
(3) Question 5.06 Agree with facility comment. This question is a direct quote from LO-LP-33304-00 and Learning Objective LO-LP-33304-00 #10.
Use of the Special Test described in the facility comment recommendation #2 (which was not provided by the facility for examination preparation) will be an acceptable answer for grading. The lesson plan should be corrected to reflect this information.
(4) Question 5.07 Do not agree with unreferenced facility comment. Facility gave no reference for " doubling thumbrule." The facility is reminded that the merit of the comment is directly determined from the reference material provided for question evaluation.
(5) Question 5.13 Agree with facility comment. Lesson Plan ( LO-LP-33603-00) provided by the facility is in direct conflict with Technical
_ Specifications cited in the facility comment. Training material specifically states that the Samarium concentration is NOT used to calculate the 50M. The question was deleted. Training material should be modified to reflect this information.
(6) Question 6.07 Agree with facility comment. The setpoint was recently changed due to modifications to the Technical Specifications. No other setpoint will be accepted. In addition, auto closure of PZR PORVs and PORV block valves will be accepted as one of the five required responses.
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(7) Question 6.09 Agree with facility comment. The answer will be modified to allow for any 3 of the 9 purposes described in plant Technfcal Specifications. It was noted that the training material vas inadequate in that it only requires 3 bases for RIL's. Training material should reflect the bases as they are stated in plant Technical Specifications.
(8) Question 6.12b Agree with facility comment. Will accept the recommended answer as at alternate answer in part b to allow re-energizing loads powered from these busses before the SI signal is reset.
Training material should be modified to reflect this information.
(9) Question 6.15 Agree with facility comment. The Technical Specification setpoint will be the only correct answer. The answer will be changed to 22.5%. The facility is reminded to update plant reference materials to conform with Technical Specifications.
(10) Question 6.18 Agree with facility comment. Based on additional information provided by the facility, the recommended additional answer will also be accepted. Training material should be modified to reflect this information.
(11) Question 6.21 Agree with facility comment. Will change the answer to allow the additional correct answers. The facility is reminded that the lesson plan should reflect the design bases that are part of the FSAR.
(12) Question 6.23 Agree with facility comment. The answer is changed to include the 2 additional purposes; only one will be required for full I credit. The facility Learning Objective states ' Describe the purpose (singular noun) of the high head SI. . ." The learning objectives should be updated to reflect material in the text.
(13) Question 7.06a Agree with facility comment. The facility is requested to ensure that exam preparers are informed of all known procedural
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errors and changes that are promulgated to students. The-question was deleted.
(14) Question 7.07.
Agree with facility comment. Answers c; or d will be accepted for ' full credit. Examiner did, in fact, use the provided procedure 'that was submitted by the facility for ' exam -
preparation, (18003-1,. Rev. 2). The facility is requested to ensure that'no " outdated procedure" is submitted.to the NRC for examination preparation.
, (15) Question 7.17
' Agree with' facility comment. Due to .the wording of - the .
question, which may have been interpreted that the first steps of E-0 have been accomplished, either choice a or c will be an acceptable answer.
(16) Question 7.22a Agree with facility comment. Based on possible misinterpretation, part a of question 22 was deleted.
(17) Question 7.22c Agree with facility comment. Since there is no time limit on terminus monitoring, will accept FALSE as .the only correct answer.
(18) Question 8.15a Agree with facility comment. The answer key was changed to add the other conditions that are required. All 3 answers will be required to obtain full credit.
(19) Question 8.15b Agree with facility comment. The answer key was changed to reflect that that the subclearance holder was not contacted when off site (20) Question 8.17 Agree with facility comment. The answer key was changed to add Mode 4.
9 THE FOLLOWING CHANGES WERE MADE DURING EXAM GRADING.
(1) Question 5.20b Answer was revised to correct calculation error.
(2) Question 7.21 The definition of Faulted S/G was changed to reflect the E-2 definition.
(3) Question 8.08 Answer was changed to correct a typographical error.
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
There were no generic weaknesses noted during the oral examination.
The examiners and facility staff discussed specific simulator limitations and malfunctions.
The facility expressed concern in regard to the number of examiners participating in the examination. It was recognized that a more stressful environment for the candidates can result from excessive personnel in the examination area, and that measures were taken during this examination to minimize the impact on the candidates. This concern was not considered to have had a detrimental effect on the candidates' performance.
The facili ty was reminded that streamlining security area access procedures as well as health physics check in/out procedures is an advantage to the candidate, in order to minimize excessive examination times, thus, preventing a more stressful environment and hindrance of the candidates' performance.
l It was noted in the reference material provided that learning objectives I and lesson plans for Radioactive Waste systems and Radiation Monitoring systems were not supplied. In addition, overhead transparencies referenced in the lesson plans were not supplied. ,
In addition, the facility was reminded that any unique clothing requirements (ie foot gear) be communicated to the examiners prior to site arrival.
The cooperation given to the examiners and the effort to ensure an ;
atmosphere in the control room conducive to oral examinations was also noted and appreciated.
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The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners.
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Georgia Power Company 2 July 2, 1987 cc w/ enc 1s 1 & 3:
J. Badgett, Corporate Training Director G. Bockhold, Plant Manager cc w/encls 1, 2, & 3:
P. Rushton, Site Training Director bec w/ enc 1 1:
M. Miller, Project Manager, NRR State of Georgia bec w/encls 1, 2 & 3:
M. Sinkule, DRP Senior Resident Inspector 0perator Licensing Branch, DLPQ JDocumentControlDesk EG&G Idaho, Cont:2.ctor PNL, Contractor i
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U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: - _V_O _G _T LE1 REACTOR TYPE: _PWR-WEC4________________
DATE ADMINISTERED: _@Zlg4/13________________
EXAMINER: _RgE@ENER z _@z____________
CANDIDATE: ________________,,________
INSISUCIlgN@_lp_C@Npip@IE1 Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Ex ami na t i on papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 28,2s DL 49
-@ TIE 9'-- 26 g} ___________ ________
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 29199-- 23:2g ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2 L ll 29:ZE_- 2Et@Z ___________ ________ 3. INSTRUMENTS AND CONTROLS 2 Y , 7 s' 2*/,V.2.
59:29-_ _23129 ___________ ________ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL M_ _ ___________ ________% Totals Final Grade J All work done on this examination is my own. I have neither given j nor r ecei ved aid, i
Candidate's Signature i
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_______m._ ___-__ __
- r. ,
, NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the f ollowing rules apply:
' 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil gnly to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (i f necessary).
- 6. Use only the paper provided far answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
B. Consecuti vel y number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, wri te gnl y gn gne s,i_de of the paper, and write "Last Page" on the last answer sheet.
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least 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 onl y if they are commonly used in facility literature.
- 13. The point value for each questi on is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all cal cul ati ons, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
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- 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. Thi s must be done after the examination has been completed.
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- 18. When you complete your ex ami nat i on , you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. ' Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after l eavi ng , you are found in this area while the examination is still in progress, your license may be denied or revoked.
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11_lE51Ng]f65@_QE_yUg6g@@_EghgB_E6@NI_g[gggIIgN3 PAGE 2
'IUEBMQQyNOMICSS _bE8I_I69NSEEB_@NQ_FLUIQ_ELQW DUESTION 1.01 (1.00)
Multiple Choice Which one of the f ollowing statements most correctly describes the change in the Doppler and Moderator Temperature Coefficients (FTC ]
and MTC) as a function of core life (BOL to EOL)? q l
- a. FTC becomes more negative and MTC becomes more negative.
- b. FTC becomes more negative and MTC becomes more positive.
- c. FTC becomes more positive and MTC becomes more negative. ]
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- d. FTC becomes more posi ti ve and MTC becomes more posi tive.
QUESTION 1.02 (1.00)
State the physical reason why the decrease in baron concentration over core life causes the MTC to change. (Include in your answer the physical effects of a unit temperature change and the associated t effect on reacti vi ty and MTC.)
QUESTION 1.03 (2.00)
With reference to the Power Defect Curve (Figure 1-a attached),
answer the f ollowing questions,
- a. Match the lines (A,B,C) of the Total Power Defect Curve to the ti me-i n-li f e each represents (BOL,MOL,EOL).
usn3 Dyro I-QclerC"'ac de PowEK' in the scte' ; c s c ^s ; v ;
coufzczeur for each Nne
- coeff;cicnt i s ,- cach '.;ns +.
M 3,andc),
- b. -Whe m . 's 7 (Accuracy +/- 0.5 units). !
QUESTION 1.04 ( .50)
TRUE or FALSE?
As temperature increases control rod worth decreases.
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- 1. ' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION t PAGE 3 ,
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QUESTION 1.05 (1.00)
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State the physical reason why an increase in boron concentration causes a change in control rod worth.
i, QUESTION 1.06 (1.00)
How does the rod withdrawal scheme of overlap make reactivity addition rate more uniform?
QUESTION 1.07 r1.00)
How can Integral Control Rod Worth be derived from a Dif f erential Control Rod Worth Curve?
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QUESTION 1.08 (2.00)
- a. State whether the parameters listed below would INCREASE, DECREASE or STAY THE SAME as a result of a normal power change from 75% load to 85% load. Consider each parameter separately and assume the control rod system is in AUTOMATIC.
- 1. Tavg
- 2. Delta T
- 3. Reactor Power
- 4. Fuel Temperature
- 5. Reac t i vi ty ,
- b. State whether the parameters listed below would INCREASE, DECREASE or STAY THE SAME as a result of a normal power change from 75% load to 85% load. Consider each parameter separately and assume the control rod system is in MANUAL.
- 1. Tavg
- 2. Delta T
- 3. Reactor Power
- 4. Fuel Temperature
- 5. Reac ti vi ty OUESTION 1.09 ( .50)
True or False?
Through some unexplainable oversight, the primary neutron sources are not installed in a new core. Because of this oversight the core can not physically go critical .
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- lt_;EBINCIELEg_gE_NUCLEOB_EgWEB_E68NI_gEEB911gN z PAGE 5
'IUEBMgpVN901CS t_dE@I_IB@NSEEB_8NQ_ELUlg_E69W QUESTION 1.10 ( .50)
True or False?
A reactor originally operating at a Keff of 0.99 can be made critical if a large enough' primary neutron source is added to the core.
QUESTION 1.11 (2.00)
A beginning-of-life core has been operating with all rods fully withdrawn at 100% power for 1 week. The bank D-control rods are-inserted 100 steps with sufficient boron dilution to offset the reacti vi ty added by the rods. Answer the following questions-about the resultant xenon transient IN THE RODDED REGION OF THE CORE.
- a. How does the xenon l evel change at first (INCREASES, DECREASES, STAYS THE SAME)?. ~(0.50)
.c. The final steady-state xenon concentration will be ________(LESS THAN, GREATER THAN, EQUAL TO) the original concentration. (0.50)
DUESTION 1.12 (1.50)
During reactor operation, the sources of Xe-135 are____a____(accounts 3 for approximately 95%) and ____b_____(accounts for.approximately 5%).
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's- Ic_;PBJNCIE6ES_gE_NUC6E85_E9 WEB _E69NI_gEgB9IlgNz PAGE 6
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QUESTION 1.13 (1.00)
Multiple Choice Which one of the following statements best describes the difference between. prompt and delayed neutrons 7
- a. Prompt neutrons slow down significantly faster than delayed neutrons.
- b. Prompt neutrons are produced much faster-than delayed neutrons.
- c. Delayed neutron li f etime f rom thermalization to absorption is much longer than that of a prompt neutron.
.d. Delayed neutron lifetime from emission to thermalization is much longer than that of a prompt neutron.
QUESTION 1.14. (2.00)
- a. Define shutdown margin as stated in the technical specifications definitions.
- b. What is the shutdown margin in a reactor with a Keff of 0.95 i (accuracy +/- 0.2%)?
(Show work for this calculation).
l QUESTION 1.15 (1.00)
With reference to figure 1-b:
Explain why the Differential Boron Worth decreases at higher concentrations of boron.
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QUESTION 1.16 (1.00)
Would an INCREASE or a DECREASE of the f oll owing parameters cause the i DNBR to decrease? Consider each parameter separately; no explanation i s requi red,
- a. Tavg I
- b. Primary pressure j l
- c. Coolant system flow rate
- d. Local power density li
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l QUESTION 1.17 (1.00) ;
What FOUR operator actions (op er ati on al requirements) are required to I ensure the Total Heat Flux Hot-Channel-Factor AND the Nuclear Enthalpy Rise Hot-Channel-Factor are not exceeded between periodic measurements?
I QUESTION 1.18 (1.00)
What THREE general conditions must be present for a brittle fracture to occur?
QUESTION 1.19 (1.00)
Multiple Choice Choose the one statement which best describes the effect of fast neutron irradiation of the reactor vessel.
- a. The reactor vessel metal weakens .
- b. The reactor vessel metal softens.
- c. The reactor vessel metal RTndt lowers, i
- d. The reactor vessel metal becomes less ductile.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
L ' PRINCIPLES OF NUCLEAR POWER PLANT' OPERATION3 PAGE 8.
'IHEBdggyN90lggz_HE81_IBSNSEEB_9Ng_E6Ulp_ELg8 I
QUESTION 1.20 (2.00)
- a. Define Pressurized Thermal Shock.
- b. Li st' FOUR of the FIVE design accidents during which RCS' coolant conditions will approach Reactor Pressure Vessel thermal.
shock limits.
QUESTION 1.21 (1.50)
- a. At normal hot standby conditions, in which direction will the pressurizer level indication change as a result of the following transients? (INCREASES, DECREASES, STAYS THE SAME).
(Consider each transient separatel y. ) A55ume /erdcwn g ed c 44 rg /og f/avs arc eptlised A nd the pressuri t er leve l con rrc/ sp rem i.5 in en d n u g f.J
- 1. The reference leg heats-up from 120 F to 2OO-F due to the relocation of a ventilation duct. (0.50)
- 2. The pressurizer heaters fail and the pressurizer water cools from normal operating temperature to 590 F. (0.50)
- b. For case 2 above,-following the cooldown, is the indicated level GREATER THAN, LESS THAN, or EQUAL TO equal the actual level?
(0.50)
OUESTION 1.22 (1.25)
If reactor-pressure is 1330 psig and core outlet thermocouple temperature is 550 F, what is the subcooling margin? Show your work.
(accuracy of +/- 1 F) l l
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.jf 13_' PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONt PAGE 9 ;
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l QUESTION 1.23 (1.50) i For each of the' parameters listed below, provide the desired indication or trending that would be expected for natural circulation cooling.
- a. Th' b.. Steam Generator Pressure
- c. Subcooling
- d. Steam Generator Level
- e. Pressurizer Level
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De le TEb O ESTION 1.24 (1.25)
What are the ikely consequ nces of a seve e water hammer event in a
. steam pipe?
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- 2. ' PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10
--' i QUESTION 2.01 (1.00)
Multiple Choice Cl>ous e Chan the answer which best describes the reason that the anti-reverse rotation device is included on the reactor coolant pumps.
To prevent reverse rotation of an idl'e reactor coolant pump (i mpell er , shaf t ,- '
and motor) which could lead tot.
- a. thrust bearing degradation.
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- b. excessive starting currents.
- c. damage of the seal package.
- d. excessive core bypass flow.
1 QUESTION 2.02 (2.50) l
- a. State which discharge collection tanks the following Chemical and Volume Control System (CVCS) relief valves discharge to.
- 1. 600 psig letdown relief. (PSV-8117)
- 2. 75 psig VCT relief. (PSV-8120)
- 3. 220 psig individual charging pump suction reliefs.
(PSV-44688,C)
- 4. 2200 psig alternate mini flow reli ef s f or centrifugal charging pumps. (PSV-8510A,B)
- 6. Which one of the above relief valves has a downstream temperature monitor to indicate when the valve has lifted.
QUESTION 2.03 (2.00)
- a. Explain the purfr,se of the check valve around the normal charging val ve. (HV- 8/Y0
- b. Why is the check valve around the normal charging val ve spring-loaded?
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
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PAGE 11
' 31_fE60NI_ DESIGN _ INCLUDING _@@ Eely _8ND_EDEBGENCy_@y@IEd5
' QUESTION 2.04 (2.50)
List FIVE of the SIX components / systems that discharge to the Pressurizer Relief Tank.
. IDENTICAL PARALLEL COMPONENTS COUNT AS ONLY ONE SOURCE!
QUESTION 2.05 (2.00)
- a. Explain the means of pressure control when the plant'is being operated in a' solid water condition.
- b. List TWO components which provide overpressure protection when the plant is solid? IDENTICAL PARALLEL COMPONENTS COUNT AS ONLY.ONE OF TWO. {
QUESTION 2.06 ( .50)
TRUE or FALSE?
The emergency boration manual bypass valve (185) supplies borated water from the BAST to only the Train B CCP.
QUESTION 2.07 (2.50)
List the TEN of the FOURTEEN valves in the High Head Safety Injection System that automatically change position upon receipt of a Safety Injection signal. State whether the valves open or shut. PARALLEL TRAIN VALVES ARE COUNTED AS TWO VALVES. Assume normal line-ups with the positive displ acement pump supplying charging bef ore the SI is received.
QUESTION 2.08 (2.00)
- a. Give two reasons why NaOH is added to containment spray water to control the pH.
- b. How is the NaOH transf erred from its storage tank to the containment spray stream?
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PAGE 12 -
21_[P6@dI_pg@l@@_INg6UplNg_@@{ gly _@Np_EDEggENgy_@y@IEgg QUESTION 2.09 (3.00)
Where in the reactor coolant system are the following penetrations located? Specify loop (1,2,3,4), AND hot, cold, or intermediate leg.
- a. Normal letdown line. (0.50)
- b. Pressurizer surge line. (0.50)
- c. Normal charging lines (2 required). (0.75)
- d. Excess letdown line. (0.50) f
- e. Pressurizer spray lines (2 required). (0.75)
QUESTION 2.10 (1.00)
Multiple Choice Which of the following is NOT a purpose f or the CCW surge tank?
- a. A source of makeup water in case of a leak.
- b. A point to add corrosion control chemicals.
- c. Compensates for volumetric changes due to thermal expansion.
QUESTION 2.11 (1.50)
List the THREE loads cooled by the Component Cooling Water (CCW) system.
QUESTION 2.12 (1.00) 1 Why are two cooling loops (Auxiliary Component Cooling Water g !
and Nuclear Service Cooling Water (NSCW)) used to cool the ",gCCW) cgrr hear c ati cc exchanger instead of using just one loop.
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
.. 2 ' PLANT DESl@N_INCLUDINQ_@AF@Ty_AND_@M@R@@NCY_@y@TEME PAGE 13 L
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QUESTION 2.13 (2.50)
TRUE or FALSE 7 1
The following statements concern the auxiliary feed water system (AFWS).
- a. The AFWS is capable of handling 100% of the decay heat removal capacity even with one motor driven and the turbine driven ,
pump out of service.
- b. The AFWS is designed so that the system can perform its intended function with a single passive AND single active failure to the AFWS occurring simultaneously.
- c. The steam leakoff from the AFWS turbine is collected and routed to a radioactive waste system to protect personnel in the event '
i that the number 1 or 2 S/G suffers a steam generator tube rupture,
- d. The Train A motor driven AFWS pump can be lined up to the S/Gs ,
supplied by the Train B motor driven pump by opening two l' normally locked shut manual val ves.
- e. To ensure that minimum water l evel is maintained in the Condensate Storage Tank (CST), all suction valves supplying non-safety related systems are automatically y isolated upon receiving a CST l o-l o l evel alarm.
QUESTION 2.14 ( .50)
TRUE or FALSE?
Upon receiving Containment I sol at i on Phase A signal, service air AND instrument air will be automatically i sol ated at the containment boundary.
l QUESTION 2.15 (2.00)
List the FOUR automatic responses of the NSCW system to an SI.
QUESTION 2.16 (2.00)
List FOUR of the FIVE conditions which will cause a rod drive M-G set trouble alarm.
(***** CATEGORY O2 CONTINUED ON NEXT PAGE *****)
1
J. - ' A_: Eb8NI,pEgl@N,1NggUplNg_gg{ gly,@Np_g[gggENgy,gy@lgdp PAGE 14 QUESTION 2.17 (1.00)
Multiple Choice Which.of the following statements correctly describes the symptoms of a Train A Safety Injection where Train B S1 fails to produce an SI signal?
- a. Both emergency diesels start but onl y the Train A sequencer operates.
- b. The Train A diesels starts but the Train B diesel does not.
- c. Both emergency diesels start, but the Train B components do not.
- d. Both diesels start and both train's components start.
QUESTION 2.18 ( .50)
TRUE or FALSE?
If a Safety Injection occurs with an Emergency Diesel Generator already loaded on its emergency bus, the ED/G output breaker will stay shut.
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(***** END OF CATEGORY O2 *****)
' INSTRUMENTS AND CONTROLS PAGE 15
'}. '
{
s DUESTION 3.01 (2.50)
- a. List the THREE arming signals for the steam dump control system. (1.50) l
- b. What TWO interlocks must be satisfied in conjunction with the arming signals to allow control of ALL steam dump valves? (1.00)
QUESTION 3.02 (2.50)
The pressurizer pressure controller setpoint is i nadver tentl y changed to 2385 psig. Assume a step change.
- a. What alarm and what automatic action immediately occurs? (1.00)
- b. If no operator action is taken and all other controllers and components operate correctly, the primary pressure will begin to rise. What is the next automatic action that will occur and what TWO alarms will ne associated with it? (1.50)
QUESTION 3.03 (1.50)
What ALARM would sound and What TygEE components would be automatically operated if one of the two 17% bistables in the Pressurizer Level Control circuit failed low /off? IDENTICAL PARALLEL COMPONENTS COUNT AS ONLY ONE COMPONENT.
QUESTION 3.04 ( .50)
TRUE or FALSE?
If one of the pressurizer level 17% bistables failed low /off, the RO could regain control of pressurizer level and pressure by using the PRZR LEVEL CONTROL SELECTOR switch? A 3S vn e the og e r4 7e e T4 /f 3 go o rk r-or a fre r rhe ope ra 7)oy, af rhe Se/ecror Scu j 7 c f, ,
acrions be fore QUESTICN 3.05 (2.75)
List the THREE permissives and ONE control system which are supplied by the summed output of the Power Range nuclear instruments. Include SETPOINTS for all permissives. Coincidence is not required.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
- 21. ly@lBUygylp_@yg_ggylgg6g PAGE 16 QUESTION 3.06 (1.00)
Multiple Choice Which of the following statements about the Reactor Vessel Level
' Indication System is most correct? Be sure to consider the conservative design intent of this system when arriving at your answer.
- a. RCS pressure is used to calculate vapor phase density and ,
RCS temperature is used to calculate liquid phase density. !
- b. RCS temperature is used to calculate li qui d phase densi ty.
i and RCS pressure is used to calculate vapor phase density.
- c. IF RCS temperature is used to calculate vapor. phase density THEN RCS pressure will be used to calculate' liquid phase density, i
- d. IF RCS temperature is used to calculate vapor phase density THEN RCS temperature will also be used to calculate liquid phase density.
QUESTION 3.07 (3.00)
List the THREE signals.that cause automatic main steam line isolation.
Also list the setpoints and coincidence associated with the signals.
QUESTION 3.08 (1.00)
How is a reactt?r trip on turbine trip prevented from occurring even when f our-cut-of -f our turbine throttle valves are shut? Include INSTRUMENT, SETPOINT and COINCIDENCE in you answer. l4 QUESTION 3.09 (2.00)
- a. During a normal rod withdrawal, you notice that only every other digital rod position light is illuminating for DNE of the rods as it travels. What would cause this indication?
- b. State the accuracy of the affected rod's digital rod position indication system. (Either Cf the two possible accuracies constitutes a correct answer.)
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(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
1
)
- PAGE 17
'at_41NSIBUMENIS_8ND_CONISOLS QUESTION 3.10 (1.00)
Why'are both main feed pump speed AND main feed water regulating valve position varied by the Steam Generator Water Level Control system instead of just varying main feed water regulating valve position?
L L QUESTION' 3.11 (1.00)
[
Multiple Choice The plant was operating at steady state, 75% load, with no steam flow, feed flow, or S/G level oscillations when controlling feed flow on one S/G t'cok a step increase of 10%'due to an instrument malfunction (actual feed flow. was constant). Which of the following statements most correctly describes the final condition if no actions are taken to correct the problem?
- a. S/G ' level will control at a CONSTANT level suf ficientl y LESS THAN the original level to offset the flow error,
- b. S//G 1evel will control at a CONSTANT level sufficiently MORE THAN the original level to offset the flow error,
- c. S/G 1evel will OSCILLATE around a new LOWER level.
- d. S/G 1evel will OSCILLATE around a new HIGHER level.
QUESTION 3.12 (1.50)
List the THREE inputs used to control the feedwater bypass valve.
QUESTION 3.13 (2.50)
State the automatic control functions of the f ollowing Process and Effluent Radiation Monitors.
- a. Steam generator blowdown.
- b. Control room i ntake.
- c. Boron recycle.
- d. Turbine building drains.
- e. Containment vent effluent.
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_3_. ____* I N_S_T_R_U_
__ __MEN _T S_ A_N_D__CO_N_T_R_O_L_S PAGE 1 19 l
l QUESTION 3.14 (2.00)
List the TWO Area Radiation Monitors which have an automatic control function AND the associated function.
QUESTION 3.15 (1.00)
I List the automatic start signal, coincidence, and setpoint for the Containment Spray System.
QUESTION 3.16 ( .50)
TRUE or FALSE i
With a Safety Injection (SI) signal and no undervoltage condition, the spare CCW pump is automatically started IF the CCW heat exchanger outlet temperature is greater than 125 F.
QUESTION 3.17 (2.00)
List the FOUR automatic start signals for the motor driven auxiliary feedwater system. Setpoints NOT required.
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-_ _____ _-__ __ _--__-_ O
- - ' Is_' INSIByMgul@_BND_CQUIBQLS PAGE 19 QUESTION 3.18 .(2.50)
Match the. actual reactor trip breaker response of column B to the reactor trip initiators in column A. Not all of column B answers' will be used.- Column B answers may be used more than once.
XRTBs = reactor trip breakers, RTA or RTB = reactor trip breaker A or B, BYA or BYB.= bypass reactor trip breaker A or D).
Column A Column B
- b. Simultaneously shutting both bypass RTBs 2. UV trip of RTA & BYB-Shunt trip of RTA
- d. Manual trip on Main Control Board B'(QMCB-B) 4 UV trip of all RTBs
- o. Manual trip at Remote Shunt trip of RTA & RTB Shutdown Panel A (PSDA)
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(***** END OF CATEGORY 03 *****)
\ - -_ _ - _ _ _ _ _ . _ - - _ _ - - _ _ _ _ - _ -
'4..' ' PROCEDURES'- NORMAL _g ABNORMAL t_ EMERGENCY,AND PAGE 20
' 669196ggICe6_C9t!I696 l
i l
QUESTION 4.01 (1.50)
Step 2 of the turbine runback procedure (ADP 18012-1) requires the H
' operator to " Verify that a runback has initiated". According to the l rest of this step, how does the operator perform this verification? !
i.e. What THREE items must he. check and what should be their conditions? !
CUESTION 4.02 (1.50) i Concerning.the reactor coolant pumps:
- a. If the No.,1 seal leakoff flow suddenl y increases to >5 gpm, the No. I seal leakoff must be isolated promptly ( < 5 mi n. ) .
For what reason is this done? (1.00)
- b. Fill in the blank by placing the answer on your answer sheet.
1The pump must be stopped within _____,,____ minutes after the No. 1 seal leakof f i s i sol ated. (0.50) l l
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i QUE ION 4/03 (1 G5) pg /gc , l lant c;nditions ar steady state at .50% load when/the INSTR AIR i
! HDRPpdSSSERVAIR150 annunciator / alarms. A fey / moments later ou !
noti e that the P seal package 'di.f f erenti al pressure has ris .
S te the caus of the alarm and the chain of events leading to the ise in diff 7[enti al pr essur,d' i ndi c at i on .
QUESTION 4.04 (1.00) j l
The.first precaution of the "CVCS Startup and Normal Oper ati on " !
procedure warns that seal injection flow should be maintained whenever the RCS level is at or above the level of the RCP seals (elevation )
190 ft.). Why i s thi s precaution necessary?
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._____m__ ____.__m. _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _ _ . _ _ _ _ _
' 9:_*_EBgggpuBES_ _NgBd@61_@BNQBUOLt_EMEBGENgy_@NQ PAGE 21
'6891969 GIG 86_G9NIBQL QUESTION '4.05 ( .50)
TRUE or FALSE?
During a' reactor maintenance period (mode 5), the operating RHR pump fails, and the pump stops. The operator should immediately attempt to start the standby RHR pump before performing any other steps.
QUESTION 4.06 (2.25)
State the '/ MEDIATE response, i.e. action or verification, to each of the following instrumentation abnormalities. Include condition of item (s) checked if applicable. (The actions in the response not obtained column are not r equi r ed.)
- a. Failure of reactor cool ant loop narrow range temperature instrumentation EO.303
- b. Failure of pressurizer pressure instrumentation EO.553
- c. Failure of steam generator level instrumentation EO.803
- d. Failure of turbine impulse pressure instrumentation EO.303
- e. Failure of power range nuclear instrumentation EO.303 i
. QUESTION 4.07 (2.00)
The f ollowing questi ons pertai n to the failure of a turbine impulse pressure instrument.
- a. A failure in turbine impulse prersure causes rods to move in the inward direction at 72 spm. What would you do immediately? (0.50) i
- b. After restoring Tavg to the proper value, but wi th the pressure instrument still failed, what steam dump val ve position would ;
be demanded by thL load controller (FULLY SHUT, PARTIALLY OPEN, or i FULLY OPEN)? (0.75) I i
- c. If the rods had been moving out at 72 spm because of the turbine impulse pressure f ailure, what would the steam dump valve position ]
demanded by the load controller have been (FULLY SHUT, PARTIALLY OPEN, or FULLY OPEN) once Tavg was restored to the proper value?
Again, the impulse pressure is still failed. (0.75)
(***** CATEGORY 04 CONTl:JUED ON NEXT PAGE *****)
_____________-__Q
f{_*,ESQQ@pUBEQ_;_Uggd@(t_@@NQBd@(t_@d[B@gygy,@ND PAGE 22
' 6891969 GIG 86_CQNIBQ6 QUESTION 4.08 (1.00) 1' r List two ABNORMAL Operating Procedures (1800 series) that require an l immediate reactor trip IF the RESPONSE IS NOT OBTAINED to the H
IMMEDIATE ACTION.
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QUESTION 4.09 (1.50)
The immediate action of ADP 18031-1, " Loss of Class.1E Electrical Systems", is to trip the offected train Diesel Generator.
- a. How is this done? (0.50)
- 6. Why must the af f ected' train Diesel Generator be tripped? (1.00)
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QUESTION 4.10 (2.00)
According to the Precautions section of Standard Operating Procedure 13601-1, " Auxiliary Feedwater System", what actions can be taken to reduce the possibility of water hammer when feeding Steam Generators from the Auxiliary Feedwater System? Give only TWO of the ;
Four recommended actions. The associated reasons for the actions are not required.
QUESTION 4.11 ( .50) l 1 RUE or FALSE?
NSCW pumps are shif ted periodically to minimize the deposition of di ssol ved solids on the bearings of the idl e pumps.
QUESTION 4.12 ( .50)
I' .TRUE or FALSE?
The most common way to protect a pump against damage from consecutive starts is by use of an overload relay in the pump start circuit.
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(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****) j
)
' gi_*_BBOCEQWBEg_ _NOBU@Lt_@@NOBd861_EDEBGENCY_9NQ 'PAGE 12 3 RADIOLOGICAL CONTROL l
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DUESTION 4.13 ( .50)
TRUE or FALSE 7 1
While surveying'yourself upon exit .f rom a contamination area, you find that you exceed the limit of 100 cpm greater than background on the ~
a l skin of your elbow. You should attempt to decontaminate your elbow .
with a luke-warm wet wipe and then survey i t again. If the result of {
' the second sur vey (after decontamination attempt) is still greater {
than 100 cpm above background, then Health Physics should be notified. 1 QUESTION 4.14 (' .50) l TRUE or FALSE 7 Use of acronyms in f ormal plant ~ communications is strictly prohibited.
DUESTION 4.15 (3.00)
List the FOUR conditions for which, if any one occurs, the on-shift Reactor Operator would be authorized to shutdown the reactor with out prior approval.
DUESTION 4.16 (2.00)
- a. Which DNE of the f ollowing symptoms would require the Reactor i Operator to initiate a manual reactor trip AND a manual safety injection if nelther had occurred automatically?.
- 1. Pressurizer pressure = 1850 psig J
- 2. Containment pressure = 2.6 psig ]
- 3. A General Warning alarm on the Solid State Protection System ]
- 4. Power = 30%, loss of flow in one loop 4
- 5. Power less than 49%, pressurizer f level 9 3,. {
I
- b. Which ONE of the above symptoms would require the Reactor Operator j to initiate ONLY a manual reattor scram if neither reactor scram j or a safety injection had automatically occurred?
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(
- EMERGENCY AND PAGE 24 j I,
' '4J."? PROCEDURES - NORMAL t_A3 NORMALt
'889196gelCe6_CQNIBQ(
I I
I QUESTION .4.17 (2.00)
Answer the. f ollowing questions as they apply to a." Response to Nuclear Power Generati on /ATWT ' (FR-9.1),
- a. IF the reactor will not trip even after manuall y opening the supply f eed breakers (INBOB and INBO9), what should THEN be done?
i
- b. IF the turbine did not~ trip automatically, did not trip .
l i I manually, and will not manually run back, what should THEN be done?
QUESTION 4.18 (2.00) i TRUE or FALSE 7 Answer the following true or false questions relating.to the use conventions associ ated with EDP's.
- a. If a caution statement occurs before step one of an EOP it may appl y either Eto the whole procedure or j ust to'the first step.
- b. Unless otherwise specified, a required task need not be f ully completed before' proceeding to the next instruction; it is enough to begin the task and have some assurance that it is progressing satisfactorily,
- c. Even after a transition to another procedure, the steps begun before the transition was made must still be completed, but not to del ay the transition.
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- d. Continuous actions contained in a caution (step) are NOT in effect when an operator is referenced to a procedure to be performed
-concurrently with the EOP in effect, l
QUESTION 4.19 (1.00)
I Why does the " Loss of All AC Power" EOP 19100-1 (ECA-0.0) have priority over all Functional Restoration Procedures (FRPs)?
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
1
_ - . _ . . _ - - __ _ _ - _ _ _ _)
di__BBggEQUBE@_ _NQBd@Lt_@gNQBd@61_EMEB@gNQy_8Np -
PAGE 25
' B09196991G06_G9NIBQ6 QUESTION 4.20 (1.50)
List three environmental factors which might adversely affect pressure.
temperature and/or diffe'ential r pressure instrumentation inside the containment.
QUESTION 4.21 (2.00)
Other than direct meter indication of parameter, describe one alternate way to determine the value of the following primary parameters. Consider each parameter separately,
- a. Pressurizer pressure
- b. Letdown flowrate I
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(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
.- ,. 2 Figure 1-a POWER DEFECT
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f = ma v = s/t w = ms 2 Cycle efficiency = " "I ~
s = v,t + hat E = mC -
2 a = (vf - y )/t ~
KE = mv v = v, + a A = AN A = A,e PE = mgh w = e/t A = in 2/tg = 0.693/tg W = VAP'
~
(t.)(t s) i AE = 931Am b (tg+t)b Q=$ CAT p ,
y y ,-Ex
, Q = UAAT g g -UX Pwr = W' g lii , I=I 10 *
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P=P 10 SUR(t).
TVL = 1.3/u P=P o et/T HVL = 0.693/u
'SUR = 26.06/T ~
T = 1.44 DT SCR = S/(1 - K,ff)
/A o SUR = 26 6 p CR g = S/(1 - Kdfx)
~
T = '(t*/o ) + [(s 'p)/A,ggo ] 1( eff)1 " 2C l ~~ Keff}2 7 ,= t*/ (, ;-) M = 1/(1 - K,gg) = CR g/CR0 T = (I - p)/ Aeffp M = (1 - K,ff)0/(l ~ Keff)1 8 " ( eff~l)! eff
p= [t*/TK,'gg .] + [I/(1 + A,ffT )] t* = 1 x 10 seconds
-I P = I4V/(3 x 1010) 3 eff = A 0.1 seconds I = No Idgg=1d22 WATER PARAMETERS Idg =Id2 1 gal. = 8.345 lbm 2 R/hr = (0.5 CE)/d (meters) 1 gal. = 3.78 liters R/hr = 6 CE/d (feet) .
1 ft = 7.48 gal. MISCELLANEOUS CONVERSIONS ,
3 Density = 62.4 lbm/ft 1 Curie = 3.7 x 10 dps 10 3
Density = 1 gm/cm 1 kg = 2.21 lbm i Heat of va;ori:stioni = 970 reu/lbm 1 hp = 2.54 x 10 3BTU /hr Hest of fusica = 144 Btu /lbm 0 1 Hw = 3.41 x 10 Btu /hr 1 Atm = 14.7 Psi " 29.9 in. Ig. 1 Btu = 778 f t-lbf 2
1 ft. H O = 0.433~, Ibf /in g. inch = 2.54 cm 2
F = 9/5 C + 32 "C = 5/9 ( F - 32)
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _J
i
- 14 PRINCIPLES OF' NUCLEAR POWER PLANT OPERATION t PAGE 26 INEBD99XW8MIGet_dEGI_IB8BSEE8_8NQ_E691Q_E69W 4
ANSWERS -- VOGTLE 1- -87/04/13-ROESENER,.S.
ANSWER 1.01 (1.00) 1 d. (1.00)
REFERENCE VOGTLE LO-LP-33203-OO, Sec. II-E, p. 9 Objective-Correctly describe the change in FTC and MTC as a function.
of core life. 192004 Reactivity Coefficients.
192OO4K107 ...(KA'S)
ANSWER 1.02 (1.00)
For a given temperature change, less boron is expanded out of the core at the end-of-life than at the beginning-of-life.CO 50] l Therefore, less positive reactivity addition is seen at the end-of-life [0.25] and the MTC becomes more negative.EO.25] (1.00) q
[
REFERENCE VOGTLE LO-LP-33203-OO, Sec. II-C, p. 7 !
Objective-State the physical reason for MTC change over core life.
192OO4K107 Reacti vi ty Coef ficients.
192OO4K107. ...(KA'S)
ANSWER 1.03 (2.00) :
(Refer to VEGP Plant Technical Data Book, 12 /fJ6, Tab 1.1)
- a. A-BOL B-MDL C-EOL (1.00)
( !
- b. A -11 pcm/% power B -15 pcm/% power C -19 pcm/% power (1.00) I I . s i f.> r c<s"ecT ^ EI'S) ' l 3 fo '
"h Y "'"'* l" YT' )g l
(,/ he c.nrecT S Jn /
REFERENCE VEGP Plant Technical Data Book, Vol 1 of 1, 12/86, Tab 1.1 l VOGTLE LO-LP-33204-OO, Sec II-D, p. 7 Objective-Given a graph of total power defect, determine the total l power coefficient (+/- 0.5 units). 192004 Reactivity Coefficients.
192OO4K112 ...(KA'S)
- it__ESING1ELEQ_QE_NQC6E86_EQWEB_E68dI_QEE88IlgNt PAGE 27 IdE8dQQyd@digQt_dE81_IB@NSEEB_@NQ_E(Q1Q_ELgd ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 1.04 ( .50)
FALSE (0.50)
REFERENCE VOGTLE LO-tP-33304-OO, Sec. II-E.4, p. 8 Objective-Correctly state the effect that a change in temperature has on control rod worth. 192OO5K107 Control Rods.
192OOSK107 ...(KA'S)
ANSWER 1.05 (1.00)
Increasing boron shim concentration causes a decrease in thermal diffusion length,CO.5] this results in the control rods being exposed to fewer neutronsg hnd thus being l ess ef f ecti ve Xbr worth less).[O.53
[0 53 REFERENCE VOGTLE LO-LP-33304-OO, Sec. II-E.4, p. 8 bjective-Explain the physical effect of boron shim concentration on control rod worth. 192OO5K107 Control Rod Worth.
192OO5K107 ...(KA'S)
VEGP Trs.> Iny Tem ch. 21, y, ? 3 ANSWER 1.06 (1.00)
Increases the reactivity addition rates near the extremes of rod travel. (1.00)
REFERENCE VOGTLE LO-LP-33304-OO, Sec. Il-H, p. 11 Objective-State the effect of overlap on integral control rod worth.
192005 Control Rods.
192OO5K109 ...(KA*S)
Or (i n c ,c, sin y bo , c,, co,, c e T ra ri on will redvre conrro l ro d wc < rt, )'
da ro rhe h c re< s ed a+p r;r joy. sac ,,&c e v r,0 ., , t,e, ee, borov a nd conrrol rods 1.u3
_ _ _ _ _ _ _ _ . - _ - - - - .. - - . 1
l~ + 1 __EBINglf6ES_gF_NygLE@@_EgyES_EL9NI_g[E6@llgNt PAGE 28
' ISE80QQYN@d1CS t_dE@l_18@NSEE8_@NQ_E(gip _E6QW s;
ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
l l
' ANSWER 1.07 (1.00) l L Accept either of the f ollowing:
Integral Control Rod Worth is the sum of the area under the Di f f erenti al Control Rod Worth Curve.
or The product of the distance traveled by the rod and the average Differential Control Rod Worth for the region of rod motion. (1.00)
-REFERENCE i VOGTLE'LO-LP-33304-OO, Sec. II-G, p. 11 Objective-State the relationship between ICRW and DCRW. 192005 Control Rods.
192005105 ...(KA'S)
ANSWER 1.08- (2.00)
- a. 1-INCREASE 2-INCREASE 3-INCREASE 4-INCREASE 5-STAY THE SAME (1.00)
- 6. 1-DECREASE 2 INCREASE 3-INCREASE 4-INCREASE 5-STAY.THE SAME (1.00)
REFERENCE VOGTLE LO-LP-33602-OO, Sec. II-B, pp. 5&6 Obj ective-Correctl y predi ct parametric response to changes in load.
192008 Reactor Operational Physics.
192OOBK121 192OO8K124 ...(KA'S) j l
7 l- . .
[
- - l1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION g PAGE 29 IbE80QQyN@dlCSt_dE@l_186@SEE8_8NQ_ELUlp_ELQW !
l ANSWERS'---VOGTLE 1 -87/04/13-ROESENER, S.
l l
l t-ANSWER 1.09 ( .50)
False (0.50)
' REFERENCE VOGTLE LO-LP-33501-OO, Sec. .II-B.3, p. 6 Objective-Describe the effects of removing installed neutron sources from the core. 192003 Reactor Kinetics and Neutron Sources.
192OO3K111 ...(KA'S)
ANSWER 1.10 ( .50)
False (0.50)
REFERENCE VOGTLE LO-LP-33501-OO, Sec. II-E, pp. 8&9 Objective-State the effect of a large source on criticality. 192003-Reactor Kinetics and Neutron Sources.
192OO3K101 ...(KA'S)
ANSWER 1.11 (2.00)
- a. INCREASES
- c. LESS THAN REFERENCE VOGTLE LO-LP-33602-OO, Sec. II-D, pp. 7,8,&9 VOGTLE LO-LP-33303-OO, Sec. II-E, p. 8 Objective-Correctly describe the effects of rod manipulations on xenon concentrations. 192006 Fission Product Poisons.
.192OO6K108 ...(KA'S)
I i
1 4
4
'l __EBINCIELES_QE_NyC6E@@_EgWEB_E(@NI_g[E6@IlgN t PAGE 30
' ISEBMggyN8MICS _dE@l_lB@NSEEB_@NQ_ELylD_E6gW t
o ANSWERS -- VOGTLE.1 - -87/04/13-ROESENER, S.
ANSWER 1.12 (1.50) i 1
- a. Iodine or (I-135) or-(Indirect) b.' Fission (Direct) (1.50)
(0.50 ea. for correct answers, 0.25 ea. for proper location)
REFERENCE
'VOGTLE.LO-LP-33303-OO, Sec. 11-A, p. 5 Objective-Identify the sources of xenon-135 and the relative importance of each. 192006 Fission Product Poisons.
192OO6K103 ...(KA'S)
ANSWER 1.13 (1.00)
- b. (1.00)
REFERENCE VOGTLE.L0-LP-33103-OO, Sec. II-F, pp B,9 & 10 Objective-Demonstrate a knowledge of the difference between prompt and delayed neutrons. 192001 Neutrons.
192OO1K102 ...(KA*S)
ANSWER 1.14 (2.00)
- a. The instantaneous amount of reactivity by which the reactor is subcriticalEO.343 or would be subtritical from its present condition assuming all rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of the highest reactivity worth which is assumed to be fully withdrawn.CO.66] (1.00)
O.95 -l / O.95 = xO.05263 or 5 . 3 */. o r 52 (3 r'cm *
(0.50 for proper setup and 0.50 for proper answer for a total of 1.00) j (will a ccefr eirher go s lr;ve oe ncy a1ive a o swers )
REFERENCE VEGP Technical Specifications, Definitions, p. 1-5 VOGTLE LO-LP-33603-OO, Sec. II-8, p. 5 Objective-Define SDM. Calculate a SDM. 192002 Neutron Life Cycle.
192OO2K110 192OO2K112 ... (KA'S)
it__EBINCIE6ES_QE_NyCLE68_EQWE8_E6@NI_QEEB@IlgN t PAGE 31 ISE8dQDyN8dlCS _dE81_IB6NSEE8_@NQ_E691'Q_E6QW t
ANSWERS -- VOGTLE 1 -87/04/13-ROESENER,~S.
ANSWER 1.15 (1.00)
(Refer to VEGP Plant Technical Data Book, 12/86, Tab 1.3.2-1.)
At higher baron concentrations, given boron atom is less l i k el y to be exposed to a neutronEO.3^3 due to competition for those neutrons with the other boron atomsf 0. 332 wwnt due to self- I shielding.rn 3 3 -} p.gy~ g,,yW2Or (1,00) I
[ID03 Lla03' VOGTLE LO-LP-33301-OO, Sec. 11-F, pp. 5&6 VEGP Plant Technical Data Book, 12/86, Tab 1.3.2-1, Differential Boron Worth.
Objectivo-Explain how and why boron worth varies with temperature and baron concentration. 192007 Fuel Depletion and Burnable Poisons.
192OO7K104 ...(KA'S)
ANSWER 1.16 (1.00)
- a. INCREASE
- b. DECREASE
- c. DECREASE
- d. INCREASE (1.00)
REFERENCE VDGTLE LO-LP-34202-OO, Sec. 11-C, pp. 13 & 14 Objective-Correctly describe parametric effects on DNBR. 193008 Thermal Hydraulics.
193OO8K105 ...(KA'S) l I
m--____--_------ .
1 __E51N9126ES_g[_NggLE@S_EgWEg_g6@NI_gEEB@IlgN g PAGE 32
' IdE60QDYN@dlCS _dE@l_T8@NSEE8_@NQ_E691D_F(gW t
ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 1.17 (1.00)
- 1. Maintain rods in bank within +/- 12 steps. (.5AruT ./ b w e d7 *f '"N
2.. Maintain proper bank overlap.
- 3. Maintain rod bank insertion limits.
- 4. Maintain the Axial Flux Difference within its target band. (1.00)
REFERENCE VOGTLE LO-LP-34202-OO, Sec. II-F.5, p. 22 Objective-Identify operator actions required to prevent vi ol ati on of thermal limits. 193009 Core Thermal Limits.
193OO9K107 ...(K4'S) l ANSWER 1.18 (1.00)
- 1. Sufficient stress
- 2. Temperature less than RTndt
- 3. Existing flaw. (1.00)
REFERENCE VOGTLE LO-LP-35204-OO, Sec. II-D, p. 8 Objective-Identify the conditions that must be present for a brittle fracture to occur. 193010 Brittle Fracture and Vessel Thermal Stress.
193010K101 ...(KA'S)
ANSWER 1.19 (1.00)
- d. (1.00)
REFERENCE VOGTLE LO-LP-35204-OO, Sec. II-E, p. 8 Objective-Correctly describe the effects of neutron irradiation on the reactor vessel. 193010 Brittle Fracture and Vessel Thermal Stress.
193010K105 ...(KA'S)
l l't__E81NCIE6@S_QE_NUQ6E88_EQWEB_E68NI_ GEE 8@IlgN 1 PAGE 33 j 4
THERMODYNAMICS t -HEAT TRANSFER AND FLUID FLOW j
> \
ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S. -]
a l
I AN'SWER '1.I20 (2.00)
- a. The pressure shock induced by a combination of high pressure 4and severe' cooling' (thermal shock).f,r63 (1.00)
- b. Any four of the ~ f ollowing:
- 1. Large break: loss of coolant accident
- 2. Small break loss of coolant accident
- 3. Main steam line break
- 5. Steam generator tube rupture. (1.00) ;
REFERENCE VOGTLE LO-LP-35204-OO,-Sec. II-H, p. 11 VEGP EDP Training Text, vol VII, Sec. 8, pp. 7&9 Objective-Define PTS. Identify operational transients of concern relative to PTS. 193010 Brittle Fracture and Pressurized Thermal Stress.
193010K106 ...(KA'S)
ANSWER 1.21 (1.50)
- a. ,
- 1. INCREASES
- 2. STAYS THE SAME (1.00)
- b. GREATER THAN (0.50)
REFERENCE VOGTLE LO-LP-16302-OO-C, Sec. III-A, p. 10 Obj ective-Correctl y descri be the responce of pressurizer l evel indication to changes in reference leg and variable leg temperature and pressure. 193001 Thermodynamic Units and Properties.
193OO1K103 ...(KA'S) 1 l
I
I 1
l
- PAGE 34-11__E61NglE6ES_gE_NggLE68_EgWEB_E6@NI_g[EB@llgNt IUEBOODyN@d1CS t _dEBI_I66NSEEB_@ND_ELUlD_ELOW
_ ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 1.22 (1.25) 581.7 F pressure = 1330 psig + 14.7 = 1344.7 psia (0.25)
From steam tables: 1300 psia -> 577.42 F 1
1350 psia -> 582.32 F differences 50 psia -> 4.9 F Interpolations ,
t 577.32 F + 4.9 F (( 1344.7 psia -1300 psia.)/ 50 psia)
= 577.Yg2 + 4.41 = 581. % g3 ._.
Subwo}k h = Stfl Sb F - S5 0 F: 5l & l~
(0.50 for prophr/%rdu,p se and 0.50 for proper answer for a total of 1.00)
REFERENCE l VOGTLE LO-LP-34104-OO, Sec. II-D Objective-Correctly calculate a subcooling margin give pressure and temperature. 193OO3K117 193OO3K125 Steam.
193OO3K117 193OO3K125 ...(KA'S) i ANSWER 1.23 (1.50)
,, $g,- ges,wre srsl)(e or Gr .
- a. Th stable or decreasing EO.33 ;
- b. SG pressure tracking Tave saturation pressure.,CO.3] l
- c. Subcooling established f0.3] p or .s uba.o/hj gecarer d u,> J F F
- e. PZR level trending with charging / letdown'EO.33 (1.50) ,
or N R h) 3<ce rer rl,,,, o, ep / ro ,a Sp;,
REFERENCE General Physics; Heat Transfer Thermodynamics and Fluid Flow Fundamentals, p. 356-357.
VOGTLE LO-LP-34110-OO, Sec. II-D,C & D, pp. 4-7 Objective-Describe parametric indications of natural circulation.
Describe parametric indications of natural circulation upset. 193008 Thermal Hydraulics.
193OO8K122 193OO8K123 ...(KA'S)
l k___EBINCIELE@_gE_ NUCLE 98_EgWE6_f69NI_gEEB811gN 1 PAGE. 35 q
' Ib5Bb99YN801C@i_UEBI_IB8N@EEB_9Np_E691p_E6QW j
- J ANSWERS --.VDGTLE 1 -87/04/13-ROESENER, S. {
l ANSWER 1.24 (1. ) De o rdG Wat r hammer may res t in componen damageCO.25] or cause a bre ch n t systemEO.253 w ch would rele e steamtO.253 d could lead t l ersonnel being i jured[O.25] an to damage of a er plant j components.EO.2 (1.25)
REFERENCE VOGTLE LD- -34203-01-C, S c. VII-D, pp. 2 L 28 Objectiv Correctly exp1 n the process lown as water h mer and show an appr ciation for it' potential conc quences. 19300 Fluid Stati 'and Dynamics.
193p 6K110 . . . OC 'S) 1 I
i
___-_-_--.--_-__-.---_-.__D
- PAGE L 3 __PL@NI_QEglgy_lNCLUQ1@@_@@[Ely,@NQ_EdERGENCy_@y@lEdg 36.
, ' ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
l l
' ANSWER 2.01 (1.00) 6.
REFERENCE VOGTLE LO-LP-16401-OO, Sec. III-B, p. 9 Objective-Explain the design features of a RCP. Anti-reverse rotation device. OO3OOOK608 RCPS Generic.
OO3OOOK608 ...(KA'S)
ANSWER 2.02 (2.50) a.
- 1. Pressurizer relief tank
- 2. Baron recycle system holdup tank
- 3. Waste holding tank
- 4. RWST (2.00)
- b. 1. (600 psig relief to the PRT) (0.50)
REFERENCE VOGTLE LO-LP-09101-OO, Sec. II-C, p.16, & D. 2, p. 17 VOGTLE LO-LP-09201-OO, Sec. II-C, pp. 7, & Sec. II-F, p. 15 Objective-Correctly state the relief collection points and the special features of CVCS reliefs. OO4000K108,122,123,405,603,605 CVCS Generic.
OO4000K108 OO4000K122 OO4000K123 OO4000K405 ...(KA'S)
ANSWER 2.03 (2.00)
- a. Prevents overpressurization of isolated charging lines (due to thermal expansion in the regenerative heat exchanger).
(1.00)
- b. (The spring provides sufficient back pressure on the check valve) so that auxiliary spray will flow into the pressurizer '
(and not through the check valve into loop 1).
or
]
To allow for proper operation of auxiliary spray. (1.00) {
i REFERENCE j VOGTLE LO-LP-09201-OO, Sec. II-C, pp. 8.
Objective-Correctl y state the desi gn features of the normal charging 1
l l
1' 2___E6BN1_DEgigN_lNC69 DING _g@ Eely _@ND_EMEBGENCy_SYSIEMg .PAGE 37
. ANSWERS'- .VOGTLE 1 -87/04/13-ROESENER, S.
check valve., OO4000K603 CVCS Generic.
OO4000K603 ...(KA'S)
ANSWER 2.04 (2.50)
Any five of the following:
PZR PORV'S.(2) j
- a. 3. DIScharUe [ ram RCDT hear excleujen
- b. PZR Safety Valves (3)
- c. Reactor vessel head vent h. Disc %)e fa>m Rrn wST puyS.
I i'
REFERENCE VOGTLE LO-LP-16301-02-C, Sec. III-H, p. 18 Objective-Correctly state sources to the PRT. OO7000K103 OO7000K100 PRTS Generic. 3 l
OO7000K100 OO7000K103 ...(KA'S) 1 ANSWER 2.05 (2.00)
- a. Solid pressure control is maintained through the use of the RHR/CVCS letdown crossconnectto.503 and the CVCS 1etdown pressure control valve (PV-131).CO.503 (1.00)
- 6. RHR suction relief valves.
Pressurizer Power Operated Relief Valves (PORV's). (1.00)
REFERENCE VOGTLE LO-LP-16501-OO, Sec. II-A, p. 4 ,
Obj ective-Expl ai n how cold plant pressure control and overpressure l protection is achieved. OO5000K102,104 RHRS Generic.
OO5000K102 OO5000K104 ...(KA'S)
ANSWER 2.06 ( .50)
FALSE (0.50)
REFERENCE VOSTLE LO-LP-09401-OO, Sec. 11-A, p. 6 Objective-Describe all the paths for emergency boration. OO6000K601 ECCS Generic.
OO6000K601 ...(KA'S) i
_ _ _ _ _ _ - _ - _ - _ - - - - - - - J
l l
3:__Eb9dl_EEEEEU_1 neb 991dE_E9EE11_9dE_EUEEEEUE1 E1Elggg PAGE 38
- ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 2.07 (2.50)
Any ten of the f ollowing:
2 RWST suction valves open (LV-112D,E) 2'VCT suction valves shut (LV-112B,C) 2 Alternate mini-flow valves open (HS-8508A,B) 3 Normal mini-flow valves shut (HS-8110 and BillA,B) 2 DIT outlet valves open (HV-8801A,8) 2 Normal c~narging containment isolation valves shut (HV-8106,8105) f 1 Cold shutdown CCP discharge isolation valve shuts (HV-8116)
(Each correct valve 0.20, correct positioning 0.05 for a total of 2.50)
REFERENCE VOGTLE-LO-LP-12101-02-C, Sec. II-D, p. 7 i Objective-Describe the automatic response of the HHSI system to an SI signal. OO6020K404 ECCS Normal Operations.
OO6020K404 ...(KA'S) ,
ANSWER 2.08 (2.00)
- a. Any two of the following three:
- 1. Ensures the continued removal of I by sprayed fluid.
- 2. Retains the iodine in sol uti on .
- 3. Mi nimi z es post-LOCA corrosi on. (1.00) ;
- b. NaOH is added to the spray flow with a jet eductor pump. (1.00)
REFERENCE VOGTLE LO-LP-15101-OO, Sec. I-D, pp. 7& 10 Objective-Explain how and why sump pH is controlled. 026000K402 026020K401 CSS Generic and Normal Operations.
026000K402 026020K401 ...(KA*S) i l b I
l l
l l 1 I
)
l l . _ _ _ . _ _ _ u
Bt__E60NI_ DESIGN _lNCLUQ1NG_S8 Eely _@NQ_EdERGENCy_SYSIEd@. PAGE 39
, ; ANSWERS - .VOGTLE 1 -87/04/13-ROESENERg S. i ANSWER 2.09- (3.00)
- a. . Loop 3 E.25] cold legEO.253 ,
- b. Loop 4 E.25] hot leg EO.25]
- c. Loop 1 E.25] and loop 4 E.25] cold legE.253
- d. Loop 4.C.253 intermediate legt.25]
- e. Loop 1 E.25] and loop 40.253 cold legt.25] (3.00)-
REFERENCE VOGTLE LO-LP-16001-01, Sec. II-E, pp. 16 & 17 Objective-Correctly list RCS penetrations. OO2OOOK106 & 109 RCS Generic OO2OOOK106 OO2OOOK109 ...(KA'S)
ANSWER 2.10 (1.00)
- b. (1.00)
REFERENCE VOGTLE LO-LP-10101-OO. Sec. II-D, p. 8 Objective-State the purposes of the CCW surge tank. OOB010K401 CCWS Startup/ Shutdown.
OO8010K401 ...(KA'S)
ANSWER 2.11 (1.50)
SFP HX RHR HX RHR pump seal cooler (1.50)
REFERENCE VOGTLE LO-LP-10101-OO. Sec. II-C, p. 6 Obj ect i ve-Corr ec tl y list the loads cooled by CCW. OOBOOOK102 CCWS Generic OOBOOOK102 ...(KA'S)
ANSWER 2.12 (1.00)
To prevent release of radioactivity to the environment. (1.00)
REFERENCE VOGTLE LO-LP-04101-01-C, Sec. I, p. 4 Obj ective-Correctl y state the design purpose of the ACCW system, l
)
l - - _ _
2 __E6@yI_QEgl@y_lyg6 Ugly @_@@EgIY_@yp_gygB@gygy_@ygI! igg PAGE 40
, ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
OOBOOOK101 CCWS Generic.
OO8000K101 ...(KA'S)
ANSWER 2.13 (2.50)
- a. TRUE
- b. c^ LEE TRvE.
- c. FALSE
- d. TRUE
- e. FALSE (2.50)
REFERENCE VOGTLE LO-LP-20101-OO-C, Sec. II-B, p. 7, Sec. IV-C, p. 10, Sec. IV-D,
- p. 11, and Sec. V-A, p. 17 Objectives-correctly identify true and false statements related to AFWS purpose, design function, and system i n ter r el at i on s .
061000K101,105,111, L SG4 AFWS Generic.
' 061000K101 061000K105 061000K111 061000SG4 ...(KA*S)
FSAR f ' O ' T 'I ~ 9' ANSWER 2.14 ( .50)
TRUE (0.50)
REFERENCE VOGTLE LO-LP-02101-OO-C, Sec. 1-D, p. 12 Objective-State what air supplies are lost as a result of a CIA.
078000K301 IAS Generic.
078000K301 ...(KA'S)
ANSWER 2.15 (2.00) nuefr any four o f the fot/c niny .'
- 1. Two NSCW pumps start per train.
- 2. Containment cooler isolation valves open.
- 3. Containment auxili ary cooler and cavity cooler i sol at i on valves shut.
- 4. NSCW blowdown valves shut. (2.00) 5, 14 eirker d ne p<i ,<q usc u pv,,,f.s ,% s trein h il to sia < r, rb e a ss-c o o rd erals > re nal A.scupofw;//>ntr, !
REFERENCE vo&ne to-tP- ottcl~os-C i E c* c . u-Bi PP 'l f '3 Objective-Correctly state the NSCW system response to an SI.
076000K401,402 & 403 SWS Generic.
076000K401 076000K402 076000K403 ...(KA'S)
'2 ..
PLANT _Qg@lGN_INGLyDINQ_@AFETy_ANp_gMgRGENgy_Sy@TgMS PAGE' 41 f-
,,~ ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
1 ANSWER 2.16 (2.00)
L Any four of the following:
- 1. The input breaker motor trips.
- 2. The generator-has en overvoltage condition.
- 3. The generator has a reverse current condition.
- 4. An output phase develops a ground.
- 5. The generatur output breaker trips on.overcurrent. (2.00)
REFERENCE
.VOGTLE LO-LP-27102-OO, Sec. II-A, p. 9 Objective-Correctly state the rod-drive MG set trouble alarms.
OO1010K603.CRDS Startup/ Shutdown.
OO1010K603 ...(KA'S)
ANSWER 2.17 (1.00)
- c. (1.00)
REFERENCE VOGTLE.LO-LP-01401-01, Sec. IV-C, p. 12 Objective-Describe the operation of the safeguards sequencer for ,
single train SI. 064000K411 ED/G Generic. l' 064000K411 ...(KA'S) l I
ANSWER 2.18 ( .50)
FALSE REFERENCE VOGTLE LO-LP-01401-01, Sec. IV-C, p. 11 .
Objective-State the response of the ED/G output breaker to an SI j during parallel operations. 064050K606 ED/G What/If Abnormal. J 064050K606 ...(KA'S)
3 t__IN@l69MENI@_@ND_CONIBOL@ PAGE 42 ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 3.01 (2.50)
- a. .P-4 (reactor trip)
C-7 (load rejection) I f Steam pressure mode (1.50)
I
- b. C-9 (condenser available interlock) !
P-12, BOTH ( two lo-lo Tavg interlocks) (1.00) !
REFERENCE VOGTLE LO-LP-21201-01, Sec. II-C, p. 9& 10 Obj ective-State the permi ssi ves and interlocks associated with the j steam dump control valves. 041020K603 SDS Normal Operations. -
041020K603 ...(KA'S)
ANSWER 3.02 (2.50)
- a. All pressurizer heaters will energizet.50]
The "PRZR CONTROL LO PRESS AND HTRS ON" alarm will be activatedC.50]
- b. (At 2335)a single PORV would lift (the one not controlled by the pressure controller)C.503 The "PRZR RELIEF DISCH HI TEMP " alarm [.503 and the "PCV-455A OPEN SIGNAL" alarm will be activated.C.50]
prfRg R. Hr r oGssun & r/MnE L 4 G/2TToCO3 (NOTE: It is not necessary to give the exact alarm name for full credit.)
REFERENCE VOGTLE LO-LP-16303-OO, Sec. II-C, pp. E-10 VDGTLE LO-LP-16301-02-C, Sec. III-H, pp. 16-18,and Sec. IV-K, p. 26 VEGP Alarm Responce Procedures 17011-1 and 17012-1 Objective-Describe the response to pressurizer pressure control system anomalies. 010000K607 PZR PCS Generic.
010000K607 ...(KA*S)
3 t__INSIBUdENIS_@ND_ CONI 606@ PAGE 43
. ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
l ANSWER 3.03 (1.50)
"PZR LO LEVEL HTR CNTL OFF LTDN SECURED" alarm.[.50]
All pressurizer heaters (are deenergized).E.50]
One letdown i sol ati on val ve[ . 25] and all orifice i sol ati on valvest.25] (shut). (1.50)
REFERENCE VOGTLE LO-LP-16302-OO-C, Sec. II-C, pp. 15 & 16 VEGP Alarm Response Procedures 17011-1 Obj ecti ve-Correctl y describe the consequences of anomalies in the pressurizer level control system. 011000K301,302,303 PZR LCS Generic.
011000K301 011000K302 011000K303 ...(KA'S)
ANSWER 3.04 ( .50)
FALSE (0.50)
REFERENCE VOGTLE LO-LP-16302-OO-C, Sec. II-C, pp. 15 & 16 VEGP Alarm Response Procedures 17011-1 Obj ecti ve-Correctl y describe the consequences oi anomali es in the pressurizer level control system. 011000K301,302,303 PZR LCS Generic.
011000K301 011000K302 011000K303 ...(KA'S) 1 ANSWER 3.05 (2.75)
Per mi ssi ves:
P-8 (or single loop loss-of-flow) p e r mi s s i ve , J4A*6- T8'n /
P-9 (or reactor trip on turbine trip) permissive. 50%
P-10 (or l ow power ) permissive, 10%
Control system:
Automatic rod control (0.50 for each permissive and for the control system, .25 for each setpoint for a total of 2.75)
REFERENCE VOGTLE LO-LP-17301-OO, Sec. II-B, pp. 14-16 Obj ec ti ve-Correc tl y state the outputs of the power range instrument.
Include setpoints where appropriate. 015000K402,405,407,408 NIS Generic 015000K402 015000K405 015000K407 015000408 ...(KA'S)
- PAGE 44 2:.__IN@lBWEN]@_gND,C9 NIB 9L@
. ANSWERS -- VOGTLE 1 -97/04/13-ROESENER, S.
1 ANSWER 3.06 (1.00) c (1.00)
REFERENCE VOGTLE LO-LP-16701-OO. Sec II-8, pp. 5 & 6' Objective-Correctly identify the conservative scheme used to density.
compensate the RVLIS l evel indication. 002000.'503 RCS Generic OO2OOOK603 ...(KA'S) l ANSWER 3.07 (3.00)
- a. Hi-2 containment pressure, 14.5 psig, 2 of 3
- b. Steam line pressure low, 585 psig ( > P-11 or < P-11 but not blocked), 2 of 3 < 1w ps; wrk a co send rime curant
- c. ' Steam line pressure rate high," 100 e m y / m ( < P-11'and blocked), 2 of 3 l .
.(.50 for each signal, .25 for each setpoint, and .25 for each coincidence for a total of 3.00)
REFERENCE VEGP Technical Specifications, Sec. III-3, p. 30 Objective-State MSI signals and setpoints. 013OOOK403 ESFAS Generic.
013OOOK403 ...(KA'S) l l
ANSWER 3.08 (1.00) l Ii taa s u t c f---f o u r E . 2 5 2 pcucr angcc:.50J ceFees-t4 ten-50%C r-253--
AREFERENCE VEGP Technical Specifications, Sec. 2, p.6, & Sec. 3/4, pp. 3-4 &5 Objective-Explain how turbine operation effects RPS. 045000K411 MT/G Generic.
045000K411 ...(KA'S) a rurbo'ne. Trif I.s pre venred (cIhn b lr cl b wl,er, fcacror Trif on I pae - cw r - v 4 - (v u, [, ? r 3 pwe< ra ,y e., f, y q ,, e 9, , 7 , 1 gvyg Of fe a(nr rr;f !^ h T u r b o'a e pg ), on ly f o s sa% lc ferv s blek Je r1 l
w c> - vu, - o f- Lt W3 pu.ve- r ,y < , 7. c a a ,e gu. rc, rh, n So7c.E.zrJ r
l l
- 3. INSTRUMENTS _ANQ_CQNTRQLS PAGE 45 l .
p
. ANSWERS - VOGTLE 1 L
-87/04/13-ROESENER, S.
l Cara A or 8 fallare .
ANSWER 3.09 (2.00) or pu,;y erive in A or 8 cla rr .
- a. f eil a e si s .e vi trm i-v wim vi DRi'l u vi i m.g (1.00)
< felvre o f t. cw o ORfhah 6< wm;< red
- b. +10/-4 or +4/-10. ( 1. 00 ) calig ),
REFERENCE VOGTLE L3-LP-27201-OO, Sec. III-B, p. 8 Objective-Describe the effects of the loss of one DRPI coil.
- '""' "'7 O 00 1 . Sh ,.
~/^ /
a c ross pl,e
- ,7m g f.9aa ameect );ffere rialjuen are e, s in neaf arer te y I.o rE y v . l. c [I.vu3 ANSWER 3.1O ( 1. 00) ' To <>. in r A In r'> c e' clwr rv"! e 3 v I' VI*J V AIl.l" '*I* 3O ' ' D '
>ree iry rt,retrlhy t he <<< reris ric> o<r efTisw It keep the ma' feedw d regu ating val er in it's spons e /
rang .r693 th. ugh a applica e power ve16EO.5^2 y(grea r th / 15%).
l.co r
Main) ins a rogr ' differ tial prese re acr
. s the in fee ater r egGl ati n/ val veb'am r opti m throttl g charap erist s[ M 3 o th lvesj at variou feedwp er flow co itions .503 - (1.00)
REFERENCE VOGTLE LO-LP-18301-01, Sec. II-C, p. 14 VEGP Training Text, Ch. 13b, p.1 Objective-Explain the need for a feed pump speed control circuit.
035010K101 S/GS Startup/ Shutdown.
035010K101 ...(KA'S)
ANSWER 3.11 (1.00)
- c. (1.00)
REFERENCE VOGTLE LO-LP-18301-01, Sec. II-C, r. 12 Objective-Predict response of SGWLC system to a flow instrument drift. 4 035010A204 S/GS Startup/ Shutdown.
035010A204 ...(KA'S) ,
i l
1 L
- 3. 1NSTRUMENTS AND CONTROLS PAGE 46 l
1 - ANSWERS -- VOGTLE'1 -87/04/13-ROESENER, S.
l ANSWER 3.12 11.50)
- a. Auctioneered highE.203 nuclear power.E.303
- b. Actual steam generator level.E.503
.c. Program-level (or tiirbine impulse pressure).E.503 (1.50)
REFERENCE VOGTLE LO-LP-18301-01, Sec.-II-C, p. 13 Objective-Describe level control inputs less than 15% power.
035011K401 S/GS Startup/ Shutdown.
035010K401 ...(KA'S)
ANSWER 3.13 (2.50)
- a. Closes steam generator blowdown processing system i sol ati on valves and discharge lines.
- b. Switches control room ventilation to safety grade filtration train.
- c. Initiates diversion of distil' ate flow to recycle holdup t anic (via the evaporator feed demi n'.ral i z er ) . o r .T.s o h re_s [h 7o R /St wST; f.2 r3 d.- Stops flow f rom turbine building drains system to the waste water retentionbasin{}f*] or Qiverslon o f r vet.h e b uilclin b s vi, clis cha e flow To the
- e. Closee cont inment pdrge supply and exhaustTurbine ducts. buildh) clrch(2.50) ros /c .
REFERENCE VEGP Training Text, Ch. 11a, pp. 17 to 20 Objective-Correctly describe the automatic control function of the PERMS. 073OOOK101 PERMS Generic.
073OOOK101 ...(KA'S)
ANSWER 3.14 (2.00)
- a. Containment low ranget.503 isol ates containment ventilation system.E.503
- b. Containment high rangeE.503 initiates CIA.E.503 (2.00)
REFERENCE VEGP Training Text, Ch. 11a, p. 26 Obj ective-Correctl y li st and describe the automatic control functions associated with the ARMS. 072OOOK102 & 401 ARMS Generic.
072OOOK102 072OOOK401 ...(KA'S)
PAGE 47
'a J2 __INSIBQUENI@_@@Q_GQNIBQL@
4 ANSWERS -- VOGTLE 1- -87/04/13-ROESENER, S..
ANSWER 3.15 (1.00)
, Containment Hi-3 Pressuret.50], two of fourE.253, and 21.5 psigE.253.
(1.00)
REFERENCE VOGTLE LO-LP-15101-OO, Sec I-D, pp. 5 & 6.
Objective- State the CSS start signals. 026000K101 CSS Generic.
026000K101 ...(KA'S)
ANSWER 3.16 ( .50)
FALSE (0.50)
REFERENCE
.VOGTLE LO-LP-10101-OO, Sec. II-E, p. 12 Objective-State the standby pump automatic start signals. OOBOOOK401 CCWS Generic.
OO9000K401 ...(KA'S)
ANSWER 3.17. (2.00)
- 1. SIS
- 2. Bl ac kout
- 3. Lo-Lo level in any S/G
- 4. Both main feed pumps trip. (2.00)
REFERENCE VOGTLE LO-LP-20101-OO-C, Sec. V-A, p. 14 Objective-Correctly list the AFWS motor drive pump auto start si gnal s 061000K402 AFWS Generic 061000K402 ...(KA*S) i t
e._o
-- -.-____-_m-_-_ m
- PAGE 48 23.__I.N@lBLJdENI@_8NQ_CgNIBQL@
ANSWERS'-- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 3.18 (2.50)
- a. 2
- b. 4 c . L
- d. .A. .1
- e. 3 (2.50)
REFERENCE VOGTLE LO-LP-28101-OO-C, Sec. II-D, p. 10, and II-E, pp. 16 & 17 Objective-Explain trip signal effects on RTBs. 012OOOK103 L 603 RPS Generic. :
012OOOK103 012OOOK603 ...(KA'S) l l
l l
l l
l 1
l l
l I
' 4 _iPBgCEQUBES_;_NgBd861_8pNgBd@61_EUEBGENCy_@ND PAGE 49 I 88919L9 GIG 8L_ggNIBQL ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER: 4.01 (1.50)
- 1. Turbine Control ValvesE.253 - Closing E.253 2.. Generator LoadE.253 - Lowering C.25]
- 3. LOSS OF TURBINE LOAD INTLK C7 Status'LightE.253 - EnergizedC.253 REFERENCE VEGP, AOP 10012-1, p. 2 Objective-Correctly state immediate actions from memory. 045000SG11 i MT/G Generic.
045000SG11 ...(KA'S)
ANSWER 4.02 (1.50)
- a. To preserve the sealing capability of the No. 2 seal. (1.00)
- b. Thirty (30) (0.50)
REFERENCE VEGP, Procedure 13003, P. 2 Obj ective-Correctl y state f rom memory an important l i mi t ati on.
OOOO15SGO7 EPE RCP Malfunctions.
OOOO15SGO7 ...(KA'S)
Delere cl I
AN ER 4.03 (1.25) he alarm ind" ates loss of service air.E 53 Service r supplies the CVCS pas ive displace ent pump spee controller.E. 53 Upon los of service . r to the con roller, the p p goes to ma mum opeed.C. 3 The increa in pump sper d results in increase in low to the R sealst.25 and a corres onding rise i differential ressure scror a the seal .E.25] .25)
REFEREf 'E VEGP normal Proce ure 18028-1, oss of Instru unt Air Obje tive-Recogniz and explain he indication for a loss of service ;
air OOOO65A201 E Loss of In trument Air.
00 065A201 . .(KA'S) /
3 __P6gggpy@g'@_;_@QBd@bt_@@NQBd@Lt_gdE8@ENgy_@NQ PAGE 50.
6891969 GIG 86_CgNIBQL ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 4.04 (1.00)
To prevent crud infiltration into seal chambers. (1.00)
. REFERENCE VEGP System Operating Procedure 13006-1, p. 2 Objective-State ~the need for seal injection flow when RCP seals are in reactor coolant. OO3OOOK602 RCPS Generic.
OO3OOOK602 ...(KA'S)
ANSWER 4.05. ( .50)
FALSE REFERENCE VEGP Abnormal Operating Procedure 18019-1, Loss of RHR, p. 2 OOOO25K303 EPE. Loss of RHRS.
OOOO25K303 ...(KA'S)
ANSWER 4.06 (2.25)
- a. Place the rods in manual.C.303
- b. Verify RCS pressuret.303 - stable (or rising).C.253
- d. Verify no rod motion.C.303
- e. Place rods in manual.C.303 (2.25)
REFERENCE VEGP Abnormal Operating Procedures 18001-1, Primary Systems Instrumentation Malfunction, and 18002-1, Nuclear Instrumentation
, System Mal f uncti on, p. 2
- Obj ective-State correctl y f rom memory various immediate actions f or abnormal op er at i on s. 001,010,015,016, L 035000SG11 CRDS,PZR PCS,NIS, NNIS, & S/G Generic.
OO1000SG11 010000SG11 016000SG11 035000SG11 ...(KA'S)
{
i l'
1
4*. PROCEDURES -' NORMALt_ABNQRMAL 1_EMERGENgy_AND PAGE . 51.
B091969G1C86_G9NTBg6 L ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
l ANSWER 4.07 (2.00)
- a. Place rods in manual. (0.50)
- b. FULLY OPEN (0.75)
- c. FULLY' SHUT 4 (0.75)
REFERENCE VEGP Abnormal Operating Procedure 18001-1, Failure of Turbine Impulse Pressure Instrumentation VEGP Training' Text, Ch. 6a, p. 24 VOGTLE LO-LP-27102-OO, Sec. II-F, pp. 10-12 VOGTLE LO-LP-21201-01, Sec. II-C, pp. 5&6 Objective-' Correctly state response to a failure of turbine impulse CONTROL. pressure. State the effect on the steam dumps of the f ailure..
OO1000SG11 CRDS Generic, 041020A2OO SDS Normal Operations.
OO1000SG11 041020A2OO ...(KA'S)
ANSWER 4.08 (1.00)
Any ruv of TM fullosa'a 3 i Reactor Coolant System Leakage (18004-1)
Steam Generator Tube Leakage (18009-1) (1.00)
Un con tralkl % rin v u.> Rod ma rio,, (i g oo3 -t )
REFERENCE VEGP ADP 18004-1 and 18009-1 Objective-State correctly from memory those AOPs that require reactor trip as an immediate action. 0020005G10 RCS Generic, OOOO375G10 EPE S/G Tube Leak.
OOOO37SG10 OO2OOOSG10 ...(KA'S)
ANSWER 4.09 (1.50)
- a. By depressing both EMERGENCY STOP push buttons. (0.50)
- 6. To prevent operation of the Diesel Generator without NSCW. (1.00)
REFERENCE VEGP AOP 18031-1, p. 2 Objective-State the correct way to carry out an immediate action for loss of 1E electrical power and the consequences of not taking the immediate action. 064000K102 ED/G Generic.
064000K102 ...(KA'S)
--__________-_____j
k
[ '
4'. PROCEDURES - NORMAL [tABNORMAL _tEMERGENCY _AND PAGE 52'
! '.8@D1969GlC@6_CgNIBQ6 1
ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 4.10 (2.00)
Any TWO of theffollowings
- a. Maintain the Steam Generator water level in the narrow range [1.OO3-(whenever temperature is above 212 F).
- b. Maintain forward flow through the bypass feed nozzle [1.OO3-(whenever possibl e) .
- c. Maintain continuous flows [1.OO3 (during heatup, c ool'd own and hot standby operation when feedwater requirements-areLsmall).
- d. Refill the piping slowly (at a flowrate of less than 15'gpm)ti.OO3 (if steam or hot water has entered'the auxiliary feedwater-piping).
REFERENCE VEGP ADP 13610-1, p. 2 Obj ect i ve-State the actions which an operator is responsible for which help to prevent the occurrence of water hammer when feeding with the AFW system. 061000K505 AFW Generic.
061000K505 ...(KA'S)
ANSWER 4.11 ( .50)
TRUE (0.50)
REFERENCE VEGP SDP 13150-1, p. 7 Obj ective-Identi f y the correct justification for performance of a common procedure. 076000K604 SWS Generic.
076000K604 ...(KA'S)
ANSWER 4.12 ( .50)
FALSE (0.50)
REFERENCE VEGP SOP Limitations of 13003-1, p. 2, 13006-1, p. 3, 13011-1, p. 3 13115-1, p. 2, ETC.
Obj ec ti ve-Correc tl y identify a common operator responsibility.
OO4000SGO7 CVCS Generic.
OO4000SGO7 ...(KA'S)
) .
- .St__PBQGEpQBE@_ _NQBdB6t_B@NQBdB6t_EdEB@ENQy_ANQ PAGE 53
-809196991G06_GQNIBg6 ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
l ANSWER 4.13 ( .50) l l
FALSE (0.50) i REFERENCE VEGP Plant Administrative Procedure OO930-C, Radiation and Contamination Control, p. 15 L Objective-Recognize an inappropriate response to personal contamination control. 194001K103 Plant-Wide Generic ,
Responsibilities.
194001K103 ...(KA'S)
ANSWER 4.14 ( .50)
FALSE (0.50)
REFERENCE VEGP Plant Admi nistrati ve Procedure OOOO4-C, Pl ant Communi cati ons, p.2 Obj ective-Correctl y i dentif y the appropriateness of limited use of-acronyms in oral communications. 194001A105 Plant-Wide Generic Responsibilities.
194001A105 ...(KA'S)
ANSWER 4.15 (3.00)
- a. When the safety of the reactor is in jeopardy,
- b. When reactor operating parameters exceed any of the reactor protection (trip) circuit setpoints and automatic shutdown does not occur.
- c. When required to protect personnel and/or equipment.
- d. When unusual circumstances warrant. (3.00)
REFERENCE VEGP Operating and Pl ant Admini strative Procedures, 10000-C, p. 6, and OO3OO-C, p. 2 Objective-Correctly state the operator responsibility associated with reactor plant shutdown. 194001A103 Plant-Wide Generic.
194001A103 ...(KA'S) l l
s
' ' 4 *. PROCEDURES - NORMAL _t ABNORMAL _g EMERGENCY _AND PAGE 54-
- B091969GICOL_CgNI6g6-ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S.
ANSWER 4.16 (2.00)
- a. 1
- b. 5' (2.00)
REFERENCE VEGP EDP- 19000-1, PP. 1, 18 & 19 Obj ecti ve-Identi f y symptoms . (condi ti ons) that' require manual reactor
/ trip and SI. OOOOO7A202 EPE Reactor Trip.
OOOOO7A202 ...(KA'S) r1 ANSWER 4.17 (2.00)
- a. Manually insert the control rods. (1.00)
- b. Shut the main steam isolation valves [0.503 and bypass valves.CO.503 (1.00)
REFERENCE VEGP EOP 19211-1, p. 2 Obj ective-State- correctl y the RNO immedi ate action for failure-of the reactor and turbine to trip. OOOO29K312 EPE ATWS.
OOOO29K312 ...(KA'S)
ANSWER -4.18 (2.00)
- a. TRUE
- b. TRUE
- c. TRUE
- d. FALSE (2.00)
REFERENCE VEGP EOP Training Text, Vol. I, Ch. 1-B, pp. 1-2-6,7 & 11 AND Training Note from Tedd Antonacci dated 3-30-87 Obj ective-Correctl y identi f y the use conventi ons associ ated wi th EDP's.
194001A102 Plant-Wide Generic.
194001A102 ...(KA'S)
2 __EBgCEDUBE@_;_NgBd@6t_9pNQBU@6t_EUEBgENCy_@ND PAGE 55 i B6 Dig 6ggig@6_ggNIBQL l 1
1 ANSWERS -- VOGTLE 1 -87/04/13-ROESENER, S. '
l i
j ANSWER 4.19 (1.00) l l
Because all FRPs are written on the premise that at least one AC emergency bus is energized. (1.00)
REFERENCE VEGP EOP Training Text, Vol. V, Sec. 5, p. 9 Objective-Correctly state the' reason that an FRP would NOT have priority over an EOP. OOOO55K302 EPE Station Blackout.
OOOO55K302 ...(KA'S)
ANSWER 4 (1.50) ue follu by rh<ee cf r. 20 l
- a. Temperature l
- b. (Containment) pressure
- c. Humidity / flooding (1.50)
$ Riul:4 rion REFERENCE VOGTLE LO-LP-36104-OO-C, Sec. III, pp. 3 L6 bjective-List the factors that adversely affect vital instruments. l 103OOOA101 Containment Generic.
10!OOOA101 ...(KA'S)
VO M E ? dei" Te c , Vv/. Sec..),
~t, f . 12 ANSWER 4.21 (2.00)
- a. Measure pressure at l i qui d or vapor space sample line or Use alternate wide range loop pressure instruments on hot legs 1 and 4.
or Use vapor space temperature and saturated steam tables
- b. Heat balance across letdown heat exchanger or Flow balance across VCT (2.00)
REFERENCE VOGTLE LO-LP-36104-OO-C, Sec. IV, pp. 6& 7 Objective-Describe alternate means of determining important primary parameters. 103OOOA101 Containment Generic.
103OOOA101 ...(KA'S)
TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE l 01.01 1.00 WSROOO1408 01.02 1.00 WSROOO1409 01.03 2.00 WSROOO1410 01.04 .50 WSROOO1411 01.05 1.00- WSROOO1412 01.06 1.00 WSROOO1413 01.07 1.00 WSROOO1414 01.08 2.00 WSROOO1415 01.09 .50 WSROOO1416 01.10 .50 WSROOO1417 01.11 2.00 WSROOO1418 01.12 1.50 WSROOO1419 01.13 1.00 WSROOO1420 01.14 2.00 WSROOO1421 01.15 1.00 WSROOO1422 01.16 1.00 WSROOO1423 01.17 1.00 WSROOO1424 01.18 1.00 WSROOO1425 01.19 1.00 WSROOO1426 01.20 2.00 WSROOO1427 01.21 1.50 WSROOO1428 C1.22 1.25 WSROOO1429 01.23 1.50 WSROOO1430 01.24 1.25 WSROOO1431 29.50 02.01 1.00 WSROOO1432 02.02 2.50 WSROOO1433 02.03 2.00 WSROOO1434 02.04 2.50 WSROOO1435 02.05 2.00 WSROOO1436 02.06 .50 WSROOO1437 02.07 2.50 WSROOO1438 02.08 2.00 WSROOO1439 02.09 3.00 WSROOO1440 02.10 1.00 WSROOO1441 O2.11 1.50 WSROOO1442 02.12 1.00 WSROOO1443 02.13 2.50 WSROOO1444 02.14 .50 WSROOO1445 O2.15 2.00 WSROOO1446 02.16 2.00 WSROOO1447 02.17 1.00 WSROOO1448 02.18 .50 WSROOO1449 30.00 03.01 2.50 WSROOO1450 i
03.02 2.50 WSROOO1451 1
i n - - - - - _ --_o
TEST CROSS REFERENCE PAGE 2 l
QUESTION VALUE REFERENCE j
------ ---------- 1 03.03 1.50 WSROOO1452 03.04 .50 WSROOO1453 i I
03.05 2.75 WSROOO1454 03.06 1.00 WSROOO1455 03.07 3.00 WSROOO1456 .
03.08 1.00 WSROOO1457 03.09 2.00 WSROOO1458 03.10 1.00 WSROOO1459 03.11 1.00 WSROOO1460 03.12 1.50 WSROOO1461 l 03.13 2.50 WSROOO1462 l 03.14 2.00 WSROOO1463 03.15 1.00 WSROOO1464 03.16 .50 WSROOO1465 03.17 2.00 WSROOO1466 03.18 2.50 WSROOO1467 !
30.75 04.01 1.50 WSROOO1468 l 04.02 1.50 WSROOO1469 04.03 1.25 WSROOO1470 04.04 1.00 WSROOO1471 04.05 .50 WSROOO1472 04.06 2.25 WSROOO1473 04.07 2.00 WSROOO1474 04.08 1.00 WSROOO1475 04.09 1.50 WSROOO1476 04.10 2.00 WSROOO1477 04.11 .50 WSROOO1478 04.12 .50 WSROOO1479 04.13 .50 WSROOO14BO 04.14 .50 WSROOO1481 04.15 3.00 WSROOO1482 04.16 2.00 WSROOO1483 04.17 2.00 WSROOO1484 04.18 2.00 WSROOO1485 04.19 1.00 WSROOO1486 04.20 1.50 WSROOO1487 04.21 2.00 WSROOO1488 30.00 120.25 1
4 9 d
_.2___.____ _ _ _ _ _ . _ _ _ ____._ _ _ _ . _ _ _ _ _ _
l .
~
!"' M ASTM copy l
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION l FACILITY: _ygGTLE_1________________
REACTOR TYPE: _PWR-WEg3________________
DATE ADMINISTERED: _gZfg4/1g________________
EXAMINER: _g@LgWIN 1 _RS_____________
CANDIDATE: _________________________
JNSIBUCIJgNg_IQ_C@Ng19@IE1 Use separate paper for the answers. Write answers on'one side only.
Staple question sheet on top of tile answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 60%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__YGLUE_ _Igl@L ___SCQBE___ _y@LUE__ ______________C@lEGQ6Y_____________
_2899__ _Zd gg ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_29:59__ _g6 3 52 ___________ ________ 6. PLANT SYSTEf15 DESIGN, CONTROL, AND INSTRUMENTATION
_22199__ _25122 ___________ ________ 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_2ZrE9__ _2}:21 ___________ ________ 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 11Ez99__ ________% Totals Final
___________ Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
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NRC RU'_ES AND GUIDELINES FOR. LICENSE EX AMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on che examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom f. rips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use bla;k ink or dark pencil gnly to facilitate legible reproductions.
- 4. Print "our name in the blank provided on the cover sheet of the ex ami na ,t i on .
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- 6. Use only the paper provided for answers.
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- 9. Number each answer as to category and number, for example, 1.4, 6,3.
- 10. Skip at least three lines between each answer.
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- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
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- 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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L18. When you. complete your examination, you shall:
- a. Assemble your examination'as follows:
(1) Exam questions on top.
(2) Exam-aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
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- b. Turn in your copy of the examination and all pages used to answer the examination questi on s.
- c. Turn in all scrap paper and the balance of the paper that you did l not use for answerir the questions.
- d. Leave the examination area, as defined by the examiner. If after
-l eavi ng , you are found in this area while the examination is still' in progress, your license may be denied or revoked.
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- PAGE 2 l 5. THEORY OF NUCLEAR POWER PLANT OPERATION2 _ FLUIDS2 _AND l IHERDODyNAdlCS l
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QUESTION 5.01 (1.00) 6]$ ghj 6 The reactor trics from full power, equilibrium xenon conditions Six-hours later the reactor is brought critical at 1.0 B amps on the Intermediate Ra ge. If power level is maintained-s t /E 1.0 E - 8 amps.
Which one of th> f ollowing statements is COR ECT concerning control rod motion for the next 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.
- a. rods will have to be withdr/
f awn since xenon will closely follow its normal bui 4 in rate.
- b. rods will have 6 be inserted due to xenon decay.
- c. rods will ave to be rapidly inserted since the critical reac or will cause a high rate of xenon burnout.
- d. ods will approximately remain as is as the xenon establishes l I
its value for this power level.
QUESTION 5.02 (1.00)
Indicate how each of the following change over core life at 50% power (more negative or less negative).
- a. Fuel temperature coefficient
- b. Doppler only power coefficient 4 OUESTION 5.03 (1.50)
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- a. How does the value of beta bar change over core life?
(f)
- b. Give the three (3) reasons why this occurs.
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QUESTION 5.04 (1.50)
A variable speed ideal centrifugal pump is operating at 1800 RPM with a capacity of 400 gpm at a discharge head of 20 psi and a power requirement of 60 kw. Pump speed is increased to 2000 RPM, Cal cul ate the new values for flow, discharge head, and power requirements. l SHOW ALL WORK l 1
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_a, 1 5___IUE98Y_9E_NgC6g88_fgWEB_f69NI_QEgBBIlgN2_E691gpz_gNQ PAGE 3 j t
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. QUESTION- 5.05 .(1.25)
-Using the attached. Boiling Heat Transfer Curve, label the point and f four regions shown (A,B,C,D and E) on your answer sheet.
QUESTION 5.06 (2.00) i Briefly describe the two methods for determining integral rod worth. /
QUESTION 5.07 (2.00)
A reactor is shutdown by a calculated amount of 5000 pcm. a 110 ppm dilution is being performed in preparation for a startup. After performing the dilution, the countrcte doubles. Was the original shutdown reactivity correct? Justify your answer and state any assumption (s) (rul e (s) of thumb).
QUESTION 5.08 (1.00)
Explain why primary system flow rate is many times greater than secondary system flow rate while the heat transferred by the two systems is essentially the same.
QUESTION 5.09 (1.00)
In order to maintain a 200 degree F. subcooling margin in the RCS when reducing RCS pressure to 1600 psia., Steam Generator pressure must be reduced to approximately _ _. I 1
- a. 125 psia
- b. 260 psia 2110
- c. -260 psia Ph f\oH
- d. 400 psia
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION 1 _FLUlpS1 _ANp PAGE 4 IHERMOpyN9 Migs QUESTION 5.10 (2.00)
In ref erence to ref ueling operations, answer the following TRUE or FALSE.
- a. If a source range nuclear detector is moved during fuel shuffle, the reference count rate is still valid.
- b. Placing the source in a location next to a source range nuclear detector in a fuel load would give desirable results for a 1/M plot.
- c. A prediction of criticality using a 1/M plot is based upon an extrapolation, and where the extrapolation crosses the horizontal axis.
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- d. More than one detector is required for fuel movement. j l
QUESTION 5.11 (1.00)
Why are a few degrees of subcooling necessary at the output of the condenser?
QUESTION 5.12 (1.00)
Condensate depression (subcooling) in a condenser operating at 1 e psi a (vacuum) with a condensate temperature of 89 degrees F. would be approximately __________.
- a. 7 degrees F
- b. 9 degrees F
- c. 11 degrees F
- d. 13 degrees F
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'5. THEORY OF NUCLEAR POWER PLANT OPERATION 1 _FLUIpSz_AND PAGE 5
-IUERdppyN9dICS QUESTION 5.13 (1.00) p f}# f \Y
/d/[P While operating at 100% power, the plant experiences a pr sctor trip.
It.is found that one of the Control Bank B rods is s, tuck out at position 15. Which one of the f ollowing represengthe CORRECT variables that must be considered while perf or l'6g a Shutdown Margin calculation.
fuel burnup, and xenon to entration. I
- b. Baron concentration, onttol rod position, samarium I concentration, R Tave, and xenon concentration.
- c. Control rod osition, samarium concentration, fuel burnup, C'S Tave, and xenon concentration.
c[ . Bor concentration, control rod position, samarium ncentration, fuel burnup and xenon concentration.
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QUESTION 5.14 (1.00)
The reactor is critical at 10,000 cps when a S/G PORV fails open.
i Assuming BOL conditions, no rod motion, and no reactor trip, choose the one answer below that best describes the values of Tave and {
nuclear power for the resulting new steady state. (PDAH= point of adding heat)
- a. Final Tave greater than initial Tave, Final power above PDAH.
- b. Final Tave greater than initial Tave, Final power at PDAH.
- c. Final Tave less than initial Tave, Final power at PDAH.
- d. Final Tave less than initial Tave, Final power above PDAH.
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QUESTION 5.15 (1.25)
LIST the 5 conditions needed for natural circulation.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION2 _FLUJpS 2_AND PAGE 6 ISE6dgpyNADICS QUESTION 5.16 (1.00)
Which one of tha following statements concerning the Power Defect is CORRECT?
- a. The power defect causes control rods to be withdrawn as reactor power is decreased.
- b. The power defect increases the rod height requirements necessary to maintain the desired shutdown margin following a reactor trip.
- c. Because of the higher boron concentration, the power detect is more negative at the beginning of core life,
- d. The power defect necessitates the use of ramped Tave program l to maintain an adequate reactor coolant system subcooling i margin.
QUESTION 5.17 (1.00)
During a Xenon free reactor startup, critical data was inadvertently taken two decades below the required Intermediate Range (IR) l evel .
The critical data was taken again at the proper IR level. Assuming RCS temperatures and boron concentrations are the same for each set of data, which one of the following statements is CORRECT?
- a. The critical rod position taken at the proper IR level is LESS THAN the critical rod position taken two decades below the proper IR level.
- b. The critical rod position taken at the proper IR level is the SAME AS the critical rod position two decades below the proper IR level,
- c. The critical rod position taken at the proper IR level is GREATER THAN the critical rod posi tion taken two decades below the proper IR level.
- d. The critical rod position taken at the proper IR level CAN NOT BE COMPARED to the critical rod position taken two decades below the proper IR level.
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l 5. - THEORY-OF NUCLEAR POWER PLANT OPERATION1 _ FLUIDS1_AND PAGE
_T_H_E_R_M_O_D_Y_N_A_M_I_C_S ;
1 1 l J QUESTION 5.18 (2.00) i How does each of the following parameter changes effect the DNBR d (INCREASE, DECREASE, REMAINS THE SAME)?
- a. Average primary coolant temperature increases by 5 deg. F.
- b. Pressurizer temperature increases 5 deg. F.
- c. Mass flow rate in the core decreases 10%.
- d. Mass flow rate of. steam to the turbine decreases 10%, assume rods are in auto.
QUESTION 5.19 (1.00)
State the two thermal limits of most concern during accidents of high probability of occurrence and their associated values.
QUESTION 5.20 (1.00)
Steam exiting the steam generator has an enthalpy of 1190 BTU /lbm and a temperature of 540 deg. F.. Find the following:
- a. Quality
- b. Entropy
- c. Superheat l
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'5 . .
THEORY OF NUCLEAR POWER PLANT OPERATIONy _ FLUIDS3 _AND PAGE 8 ISEBOODyN9d]CS QUESTION 5.21 (2.50) l 1
Compare the calculated Estimated Critical Position (ECP) for a startup to be performed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after'a trip from 100% power, to ACTUAL control rod position if the following events / conditions occurred.
Consider each independently. (State your answer as Actual position is .
HIGHER THAN the ECP, the Actual position is LOWER THAN the ECP, or the Actual position is the SAME AS the ECP)
- a. One reactor coolant pump is stopped two minutes prior to criticality.
- 6. The S/U is delayed until B hours after the trip.
- c. The steam dump pressure setpoint is increased to a value just below the S/G PORV setpoint.
- d. Condenser vacuum is reduced by 4 of Hg.
- e. All S/G 1evels are being raised be 5% as the ECP is reached.
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- 6. PLANT SYSTEMS DESIGN t_ CONTROLt _AND INSTRUMENTATION PAGE 9 l QUESTION 6.01 (2.50)
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List the 5 Design bases f or the ECCS Cooling Perf ormance f ollowing a i LOCA as stated in 1CCFR 50.46.
-QUESTION 6.02 (1.00)
Which valve listed below is used to throttle auxiliary spray flow?
- a. Spray flow Bypass valve (HCV-091)
- 6. Aux. Spray Valve (HV-8145)
- c. Loop 4 Spray Valve (PV-0455B)
- d. Loop 1 Spray Valve (PV-0455C)
QUESTION 6.03 (1.00)
Which one statement below regarding the Source Range Nuclear Instrumentation System is INCORRECT.
to be aljows
- a. P-6 level reactor trip signal bpd5$$ghe ed source range high the two intermediate range instrU$eN. EA is above manually 10 one when E - 10 of ion chamber amps.
- b. Placing BOTH source range blocking switches to the BLOCK position de-energizes the high voltage supply to both source range instruments.
- c. The source range high level trip is blocked when P-10 is present.
- d. When P-6 is present and P-10 i s not present, the source range high level trip is automatically reinstated and the source range high voltage re-energized when one of the two intermediate ranges i s bel ow P-6 rest.
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6 __PL9NI_@y@lgM@_Dg@lgN _2 CONTROL t_AND_ INSTRUMENTATION PAGE 10 QUESTION 6.04 (1.50)
Answer the following TRUE or FALSE.
- a. The rod control unit controls bank current to the pulser for speed.
- b. The slave cycler generates current orders for its associated power cabinets.
- c. Selecting the individual control bank with the bank selector switch does not override the bank overlap program.
QUESTION 6.05 (1.75)
State the design bases for the minimum level requirements of the Condensate Storage Tank.
QUESTION 6.cf (1.00)
Which of the following i s NOT a desigr basis of the Steam Dump System?
- a. Accommodate ramp load ir- ease of 10% / minute.
- b. Pass 40% steam flow on a 50% turbine step rejection without a reactor trip occurring.
- c. Allow a turbine trip and a subsequent reactor trip from 100%
powe: wi thout lifting the S/G code safety valves,
- d. Allow for a smooth shift of plant steam load f rom the steam dump system to the turbine on a plant startup.
QUESTION 6.07 (2.00)
List the 5 protection logic signals generated by the Pressurizer Pressure Protection Logic System. (Include in your answer setpoints, coincidence and associated interlocks, if any)
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- 6. PLANT SYSTEMS DESIGN _CQNTRgLt_AND_ t INSTRUMENTATION PAGE 11 l:
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QUESTION 6.08 (1.00) f
- a. What is the function of the Auto shunt trip coils associated with ,
the Reactor Trip Breakers?
- b. How is this function accomplished?
QUESTION 6.09 (1.50)
LIST the three (3) purposes of Rod Insertion Limits.
QUESTION 6.10 (1.50)
Concerning the Overtemperature Delta Temperature Setpoint (OTSP) describe how each of the following parameter changes will effect the OTSP. ( INCREASE, DECREASE, REMAINS THE SAME)
- a. Increase in Tave
- 6. Decrease in Reactor Pressure
- c. Increase in Delta Flux Penalty QUESTION 6.11 (1.00)
The Detector Current Comparator receives input from all 4 upper and lower power range detectors. How are these inputs compared, and what conditions are needed to auto bypass circuitry while at power?
QUESTION 6.12 (1.00)
When an SI condition occurs, Busses 1NB01 and 1NB10 are de-energized,
- a. What must be done to re-energize these two busses?
- b. Why is this feature necessary?
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- 6. PLANT SYSTEMS DESIGN t_CQNTRQL t_AND_ INSTRUMENTATION PAGE 12 i
QUESTION 6.13 (1.00)
The plant is oper,ating at 100% power when an electrician inspecting a RCP breaker (the class-1E breaker), accidentally causes the underfrequency relay for the breaker to actuate its protective f unc ti on.
Which one of the f ollowing best describes the plant and/or operator response?
- a. The affected RCP trips, the operator perf orms a normal controlled plant shutdown with the other 3 loops in service.
- b. The affected RCP trips, the reactor trips due to low flow in the affected loop.
- c. The affected RCP trips, the reactor trips and the other 3 RCP's trip.
- d. The affected RCP trips, the reactor trips and the other 3 RCP's trip and the plant SI's on low low flow.
QUESTION 6.14 (1.00)
State the three (3) DCP starting duty limits.
QUESTION 6.15 (1.50)
State the two (2) protection f eatures provided by the Steam Generator Water Level Low-Low signal and their bases. (Include in your answer coincidence and setpoints).
QUESTION 6.16 (1.00)
State the two (2) reasons why the Feed Reg Valves are NOT used at power levels below 15% power.
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6.- PLANT SYSTEMS DESIGN z_CgNTRQL t_AND_ INSTRUMENTATION PAGE 13 QUESTION 6.17 (2.00)
'a. What. components are isolated or tripped on a f ull feed water i sol ati on?
- b. What signals will provide a full feedwater i sol at i on?
QUESTION 6.18 (1,50)
What are the 3 purposes for the Main Steam Power Operated Atmospheric Relief Valves?
I QUESTION 6.19 (1.50)
List ALL containment phase A isolation signals. (include coincidence (logic) in your answer). NO setpoints are required.
QUESTION 6.20 (2.00)
List four (4) of the six (6) items the Rod Control System that the Startup switch will reset.
QUESTION 6.21 ( .75) l List the three (3) design considerations the Main Steam flow restrictor nozzles are required to meet.
QUESTION 6.22 ( .50)
Answer the following TRUE or FALSE.
The positive displacement pump is NOT used for hydrostatically testing the Reactor Coolant System.
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- 6. PLANT SYSTEMS DESIGN2_CONTRpL2_AND,JNSTRUMENTATJgN PAGE 14 QUESTION- 6.23 (1.00) v Y
HGnb f S$
What is the purpose of the needle valves on the High Jpeed~ Saf ety Injection lines into the RCS?
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- 7. PROCEDURES - NORMAL1_ADNORMAL2_gMgRGgNCY_ANp PAGE 15 609196pGJC66_CQNI696 QUESTION 7.01 (1.00)
Which one of the following statements describes the correct usage of the Critical Safety Function (CSF) Status Trees while performing EOP-E-0, Reactor Trip or Safety Injection?
- b. Awareness of the Red path criteria is required at all times, but the CSF Status Trees are monitored on?y after it is determined that SI can not be terminated.
- c. Monitoring of the CSF Status Trees commences as soon as the i immediate action steps are completed. I
- d. CSF Status Trees are required to be monitored as soon as the procedure is entered and a valid SI is determined.
QUESTION 7.02 (1.50)
Indicate whether the following conditions will place you'in a CSF RED path or NOT: (consider. each independen tly)
- a. All 4 S/Gs at 29% level and AFW flow of 590 gpm.
- b. Containment pressure of 4S psig.
- c. Core exit thermocouple 831 deg. F and RVLIS full range 25%
with NO RCP's running.
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- 7. PROCEDUREg_ _NgRMAL2_ABNQRMALx_EMgRGENgy_AND PAGE 16 689196991986_ggNI696 QUESTION 7.03 (1.00)
Which one of- the f ollowing statements describes the criteria f or SI termination with VEGP 1900-1, E-0, Reactor Trip or Safety Injection?
- a. Subcooling greater than 28 deg. F, Narrow range level in at least one S/G greater than 4%, RCS pressure stable or rising, and PZR level greater than 4%.
- b. Core thermocouple < 1200 deg F, total feed flow to S/G's greater than 570 gpm, RCS pressure stable or rising, and Pzr-level greater than 4%.
- c. Core exit thermocouple greater than 736 deg. F, and RVLIS full range greater than 39%, Narrow range level in at least one S/G greater than 4%, RCS pressure stable or rising, and PZR pressure stable.
- d. Subcooling greater than 28 deg. F, Steam Dumps not in operation, RCS pressure stable or lowering, Total feed flow to S/G's greater than 570 gpm, and PZR pressure stable.
QLESTION 7.04 (1.00)
Why is there a precautionary note in EDP-ECA-0.0, Loss of all A/C Power, to monitor the CSF Status Trees for INFORMATION ONLY, but not for purposes of implementation?
QUESTION 7.05 (1.50)
The SI termination criteri a for the E-series procedures (E-0,1,2,3) are identical except in E-3 (SGTR) which requires that sufficient AFW flow be_"available" vice " measured". Describe the basis for al l owi ng this difference.
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- 7. PROCEDURES - NORMAL1 _9BNgRM9L2_gMgRggNCY_9Np PAGE 17 l
809196ggIC66_CgNIBg6 l
l 00 QUESTION 7.06 ( 1.50)Mh16 Complete the following statements according to VEGP 18028-1, Loss of Instrument Air.
g b Satf0 a. EsoLation-of-the-Servi c e Ai r- Header- wi l l- c ause the-PD-
/i gjf chang ing pump- t o-f ai-1-to ____;___;;;_n____sp eed ._.
- b. Loss of Turbine Building instrument air will__________E13_____
all the Extraction Steam NRV's and fail all Feedwater heater High level dump valves full ___E23___________.
- c. Loss of instrument air pressure will cause the RHR HX outlet valves to fail _______ C13_____ and the Heat exchanger bypass valves to fail _______ E2] _________.
QUESTION 7.07 (1.00)
Which one of the f ollowing statements concerning a dropped control rod and retrieval is CORRECT?
- a. Manually insert control rods to match Tref if rods are in '
manual.
- b. The " Rod Control System Urgent Failure" alarm must be reset prior to commencing retrieval of the rod.
- c. Control rods will automatically withdraw to compensate for the dropped rod. I I
- d. Tave should be controlled by boron changes during retrieval j of the rod. 4 l
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- 7. PROCEDURES - NORMALg _ABNQRd@Lg _gggRggNgY_AND PAGE 10 889196991986_G9 BIB 96 QUESTION .7.08 (1.00) i Which one of the following is NOT a listed method of verifying a reactor trip according to FR-S.1, Response to Nuclear Power Generation
/ ATWT7
- a. Neutron flux lowering
- b. Rod bottom lights - LIT
- c. Group Step counters - ZERO
- d. Reactor trip and bypass breakers - OPEN QUESTION 7.09 (1.00)
Answer the following TRUE or FALSE.
- a. CSF Status Trees do not have to be checked in any order provided all the CSF Status Trees are analyzed.
- b. Manual monitoring of CSF Status Trees should be performed by a Qualified Shif t Technical Adviso ,of"a Licensed Operator.
64 GUESTION 7.10 (1.50)
Procedure, VEGP-19232-1. FR-H.2, Response to Steam Generator Overpressure, states in a caution that if the "S/G Narrow range level rises greater than 96%, steam should not be released from the affected S/G(s)". What is the basis for this caution?
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7.- PROCEDURES - NORMAL _ ABNORMAL g_
3 EMERGENCY AND PAGE 19 BSD19L9GJC66_CgNIBQL L
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l QUESTION 7.11 (2.00) l l
In the blanks next to the following boration flow path (s), state (YES I or NO) if a Boric Acid Transfer pump must be running to use that flow path per UEGP 13009-1, CVCS Reactor Makeup Control System.
- a. Through 1-HV -8104 (Boric Acid to Charging valve) __________.
- b. Through 1-FV-5110A (Boric Acid Tank to Boric Acid Blender viv) and 1-FV 0110B (Alternate Boration v1v.) .
- c. Through 1-LV-0112D/E (Charging pump suction from RWST viv.)
and 1-HV-8471 A/B (Charging pump suction valve) ____________.
- d. Through 1-HV-8439 (Boric Aci d to Chargi ng i sol ati on v1 v. )
j and 1-HV-8471B : Charging pump suction viv.) _______________.
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l QUESTION 7.12 (2.00)
Per VEGP-1 BOO 3-1, Rod Control System Malfunction, for each of the f ollowing malf unctions, state "YES" or "NO" if either an orderly Reactor Shutdown or a Reactor Trip is a required action.
- a. continuous insertion of a control bank.
- b. continuous withdrawal of a control bank.
- c. failure of a bank to move in automatic,
- d. more than 2 control rods dropped.
QUESTION 7.13 (1.00)
Why does VEGP-13OO3-1, RCP Operation, caution against starting a RCP when the pump bus is supplied from the same Reserve Auxiliary Transformer through which a EDG is paralleled to the system?
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f 7, PROCEDURES - NORMAL t_ABNQRMAL _t EMERGENCY _ANQ PAGE 20 l 889196901GGL_QQN16Q(
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QUESTION 7.14 (1.00) l l
In accordance with VEGP-12OO2-1, Unit Heatup to Normal Operating I Temperature and Pressure. Why is it - necessary to have all rods withdrawn at least 2 steps if the Reactor trip breakers ar e closed and the RCS temperature is to be changed by more than 50 deg. F.?
QUESTION 7.15 (1.00)
Complete the following statements according to VEGP 12002-1, Unit Heat up to Normal Operating Temperature and Pressure,
- a. If the temperature difference between the pressurizer steam space and spray fluid exceeds _____________ deg. F spray flow into the pressurizer should not be initiated.
- b. The boron concentration in the Pressurizer should no be different from the RCS by more than _____________ ppm.
- c. The maximum heatup of the RCS shall be limited to
___________________ deg. F in any one hour period.
- d. The maximum temperature differential between auxiliary spray water and pressurizer steam space is ______________ deg.F QUESTION 7.16 (1.50)
In accordance with VEGP-12OO4-1, Power Operation, place the f ollowing items listed below in proper sequence in which they occur during a power ascent.
- a. Start a second circulation water pump.
- b. Rol the main turbine.
- c. Start a main feed pump.
- d. Synchronize to the grid.
- e. Place rod control in automatic.
- f. Transfer Main feed pump turbine in automatic.
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' 1, '. ,. .. J l 7. PROCEDURES'- NORMAL _g ABNORMAL 1 EMERGENCY AND PAGE 21 l BB91969GJC66_CQNIBQ6 l
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QUESTION 7.17 (1.00) i With the unit at 100% reactor power, a loss of of f site power occurs.
This transient trips the main turbine which in turn causes a reactor i trip. All emergency diesel generators fail to start. {
To what procedural point do you proceed first? f
- a. ECA O.0, Loss of all A/C power.
- c. E-0, Reactor Trip or Safety Injection.
- d. E-1, Loss of Reactor Secondary Coolant.
QUESTION 7.18 (1.00)
While performing steps in ECA O.0, Loss of All A/C power, an SI signal may be generated. If an SI signal is generated, which one action below is required by the procedure?
- c. No action is necessary as it has not ef f ect,
- d. Reset the SI to permit manual loading of equipment on an A/C emergency bus.
QUESTION 7.19 (1.00)
What does it mean when a step in an EOP is designated with a bullette)?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL1_8BNgRM9L _gMgRGENCY_@Np 1 PAGE 22 8891969GIC86_CgNIBg6 l
l QUESTION 7.20 (1.00)
If a contingency action (Response Not Obtained) in an EOP can not be performed or is not successful, and further contingency instruction ;
is NOT provided, what action should the operator take?
QUESTION 7.21 (1.00)
Define:
- a. " Faulted S/G and b ,- " Isolated S/G as i s. used in E-2, Faulted S/G Isolation Procedure.
00 QUESTION 7.22 ( 1. Scr) SIslT (44' F Concerning the EOP's answer the following TRUE or FALSE.
- g pp a. On c c c n-FRG-.Ls- ent-er ed-due-to-a-RE D-or-Orange-cond i-t i-on , j
- that FRG i c performed to enmpletich '
11b
]$ b. If any ORANGE terminus is encountered, the operator is not required to monitor the remaining trees.
- c. If NO condition more serious than a Yellow is encountered, monitoring frequency may be reduced to to o 20 minutes.+
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QUESTION 7.23 (2.00)
In accordance with ES-0.1, Reactor Trip Response (19001-1), what are the 5 parameters (condi t i on s ) that are available in the plant that support or indicate natural ci rcul ati on flow, during cooling on natural circulation and which way should they be trending 7 v4 cusMr f/**c' 4r wc /2M* e' (P5 agra wr> uns.u e, ds 4" w a to
(***** END OF CATEGORY 07 *****)
- 8. ADMINISTRATIVE PROCEDURES z _CgNgITigNSg _@Ng_LJMII@]}gNS PAGE 23 QUESTION 8.01 (1.50)
State the two (2) bases for the Technical Specification limitations on the Primary containment's internal pressure. (Include in your answer conditions and values for the limits)
QUESTION 8.02 (1.00)
State the basis f or Technical Specification limitation on Secondary Coolant Chemistry Specific Activity. (Include in your answer conditions and values for the limi ts) .
QUESTION 8.03 (1.25)
The plant was operating at 100% reactor power when a continuous rod withdrawal accident occurred.- Reactor power exceeded the trip setpoint and a trip did NOT occur. Upon review of the accident it was determined that Tave exceeded 620 deg. F. at 100 % power and 2235 psig. Using given technical specification information:
- a. State any limitation (s) that was (were) exceeded.
- b. What four administrative actions must be done in accordance with VEGP Technical Specifications.
I QUESTION 8.04 (1.50)
State the three (3) factors cited in Technical Specification 6.7.3, which must be ensured to make temporary changes to procedures.
QUESTION 8.05 ( .50) ;
Answer the following TRUE or FALSE. i In accordance with Technical Specifications bases, OPERABILITY of the Demand Position Indication System is required to determine bank domand positions and thereby ensure compliance with the rod insertion limits.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES t _CgNDITIgNSt_AND_ LIMITATIONS PAGE 24 QUESTION B.06 (2.00)
State the two (2) bases for Technical Specification limits on Heat i Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor.
(Include in your answer conditions and values for the limits)
QUESTION B.07 (1.50)
Fill in the blanks.
A Fire Team consisting of at least ____Ea3______ members shall be maintained on site at all times. The _____Cb3 ___ shall designate the Fire Team captain and members at the beginning of each shift. The Fire Team may be reduced by one person for a period not to exceed
____ ____ E c 3______ h our ( s ) .
QUESTION B.08 (1.00)
In accordance with VEGP OO920-C, Radiation Exposure limits and f" Administrative Guidelines, what limitation (s) are placed on an individual who has a "NO ENTRY" Exposure Margin?
QUESTION O.09 (1.00)
Fill in the Blanks.
Written approval of the ____Ca]___ and ___[b3___, is required for extension of a persons whole body exposure limit to 3000 mrem / quarter.
QUESTION 8.10 (1.25) l Mr. R.D. Worker, an Auxiliary Operator, has been working 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (double shif ts) per day for the last 5 days. He is going to be off one day and then must return to B hour shifts. State whether any Technical Specification has been exceeded and cite the requirements if j
applicable.
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
l 9 __09dINJpIB91]VE_PBgCgpUBEgz_CgNp]IJgNpz_@Np_LJd]I@IlgN@ PAGE 25 l
l QUESTION 8.11 (1.00)
Choose the one term that best describes the process of determining an
- instrument's accuracy by visually comparing the indication to other !
l independent instrument channels measuring the same parameter. (As it is defined in Technical Specifications)
- a. Channel Calibration
- b. Channel Functional Test c.. Channel Check
- d. Channel Verification QUESTION 8.12 (1.00)
Which one of the f ollowing is the Technical Specification for the Pressurizer Water Level reactor trip?
- a. Prevents solid operations while the Reactor is critical.
- b. Prevents exceeding containment design pressure in the event of a LOCA with all RCS fluid flashing to steam,
- c. Protects the pressurizer safety valves against water relief,
- d. Protects against loss of pressure control due to spray nozzle being submerged.
QUESTION 8.13 (1.00) 1 Answer the following TRUE or FALSE.
- a. Over the temperature range of cold shutdown to hot operating conditions, the maximum allowable cooldown rate of the Pressurizer is 100 deg. F in any one hour.
- b. Above 50% power, the Technical Specifications no NOT allow operation with a positive moderator temperature coefficient.
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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- 8. ADMINISTRATIVE PROCEDURESt _QQNQITigNSt_ANQ_ LIMITATIONS PAGE 26 QUESTION 8.14 (2.00)
In accordance with VEGP OO3OO-C, Authority to Startup and Shutdown Reactors, there are four (4) situations that will allow an on shift Rmactor Operator or Senior Reactor Operator to shutdown the reactor without prior approval. What are those four (4) situations?
QUESTION 8.15 (1.50)
- a. Who may release a Subclearance if the SUBCLEARANCE HOLDER is off-site?
- b. What must the person (s) in part (a) do if the subtlearance is released when the SUBCLEARANCE HOLDER is off-site?
QUESTION 8.16 (1.50) 1 1 In accordance with VEGP 10003-C what positions (by title ie. GMVNO) l are required for:
- a. Normal Shift manning during Mode 1.
- b. Minimum shift manning during Mode 6.
(State any special exceptions)
QUESTION 8.17 (1.00)
The plant has been at 100% reactor power for 5 days f ollowing a normal fuel cycle refueling. The A-train RHR pump breaker has been racked out due to maintenance for 2 days. A review of the operations surveillance logs show that the required B-train SI auto start test i was not performed within the required time during the last refueling outage. State what action (s) must be taken in accordance with n Technical Specifications.
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
~ ' . :: '.
B. ADMINISTRATIVE PROCEDURES 1 _CgNDJI]QNg1_9ND_LJMJIgIJgNg PAGE 27 i
QUESTION 8.18 (1.00) ,
If, while operating in Mode 4, the Shut down margin is less than the ,
Technical Specification required value, the required action is: (State the action required by Technical Specifications) j QUESTION 8.19 (1.00)
While oper ating in Mode 1, control power is lost to a pressurizer i power operated relief valve, which one statement below is CORRECT 7 l
- a. Technical Specifications requires no action provided another PORV is operable and all pressurizer code safety valves are operable.
- b. Technical Specifications requires the power supply to be removed.from the associated block valve after verifying it to to open, if the PORV is not operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- c. Technical. Specifications requires the associated block valve to be shut and its power removea if the PORV is not made operable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and continuous operation is desirable.
- d. Technical Specifications requires action to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the plant in at least hot standby within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the PORV is NOT made operable.
QUESTION 8.20 (1.00)
Answer the following TRUE or FALSE.
- a. When attempting to raise spent fuel in the fuel el evat or the permission of either the Shift Supervisor or the Fuel Handling Engineer i s requi red.
)
- b. When attempting to raise a new fuel assembly in the new fuel elevator, the permission of both the Shift Sup er vi sor and the Fuel Handling Engineer is required.
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(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
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- 8. ADMINISTRATIVE PROCEDURES t _CQNDITIgNSt _AND_ LIMITATIONS PAGE 28 QUESTION B.21 (1.00)
Fill in the blanks.
The ____Ca]____ shall assume the position of Emergency Director in the Control Room until relieved by the designated Emergency Director. If the person in Ea] above is incapacitated, bef ore he/she is relieved by the designated Emergency Director the ____[b]____ shall assume the position of Emergency Director.
QUESTION B.22 (1.00)
A normal reactor V'f f.PV"4sif' af is in progress startlup with the unit in Mode 4 when
\
all power t o t h e BRIS'd i spl ay i s l ost . State what actions must be taken in accordance with Technical Specifications.
1pff '
QUESTION B.23 (1.00)
Which one of the following is not considered to be a part of the three cold over pressure protection systems in accordance with Technical Specifications.
- b. Two residual heat renoval suction relief valves each with a set point of 450 psig.
- c. The reactor coolant system depressurized with an RCS vent capable of relieving at least 670 gpm water flow at 470 psig,
- d. The chemical and volume control system relief valve down stream of the regenerative heat exchanger, with a set point of 390 gpm.
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(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
_ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a
- 5. THEORY OF NUCLEAR POWER PLANT OPERATIONt _ FLUIDS t_AND PAGE 29 ISEBdODyNAdlCS ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 5.01 (1.00)
}
- c. i U U p;l# gyliF REFERENCE LO-LP 33303-OO Learning Objective tate the approximate time for xenon to reach equilibrium r activity,following a transient and its equilibrium react at 100%.
c0 OOK534 ...(KA'S)
ANSWER 5.02 (1.00) ,
- a. more negative (0.5)
- b. less negative (0.5)
REFERENCE LO-LP-33203-OO p.9 Learning Objective # 6. Describe how and why FTC varies over core )
life.
- 10 List typical BOL and EOL values for Doppler only power coefficient, and explain why it ,
becomes less negative over core life. t OO1000K550 ...(KA'S)
ANSWER 5.03 (1.50)
- a. It is shorter (smaller) (0.50)
- b. 1. Beta bar depends on the type of fuel and what fraction of fissions that the fuel produces. (0.33)
- 2. Fuel composition changes over core life. (0.33)
- 3. More power is produced by 239 Pu and 241 Pu and less f rom 235 U.
REFERENCE LO-LP-33103-OO p.12 Learning Objective # 5. State how beta bar and beta bar effective f changes over core life, and explain why.
OO1000K549 ...(KA'S)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION t _FLUlQSt_ANQ PAGE 30 I!M6090M801GS
. ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 5.04 (1.50)
CO.40 f or each correct rel ationship], [0.1 for each correct answer]-
New flow = old flow x (new speed /old speed)
= 400(2000/1800) = 444 gpm New head = old head (new speed /old speed) exp 2
= 2O(2000/1800) exp 2 = 24.69 (*25)
New power = old power (new speed /old speed) exp.3
= 60(2000/1800) exp'3.= 82.3 REFERENCE Vogtle Thermodynamics Book OO2OOOK501 ...(KA'S)
ANSWER 5.05 (1.25)
A Natural Convection (0.25 each)
B Nucleate Boiling C Partial Film Boiling o4 TWW2 e no^' W40 D Film Boiling E DNB (departure f rom nucleate boiling)
REFERENCE Heat Transfer Thermodynamics and Fluid Fundamentals, part A Chapter 3
- p. 125.
OO2OOOK501 ...(KA'S)
ANSWER 5.06 (2.00)
Gwy a oc > >M 3)
- p. multiplying the average dif f erenti al rod worth (0.75) by the distance traveled (0.25).
Jh. summing the area under the differential worth curve (0.50) between the initial (0.25) and final rod heights (0.25).
- 3. au cfaW.y W,*s~a*/m (1- test- 00) f(L3c) issues >&4n Gn c< <vh ("7 REFERENCE c o ~ f d +a 4 t>s u ns a n M s % a - c / w yd G eMJ ud with (os)
LO-LP-33304-OO 601000K502 ...(KA'S) swr w- wrmecou er i-asr -c c. , , s -9 l
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATIONt_FLUIQ@t_ANQ PAGE 31 IHERdQQyNAdlC@
ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 5.07 (2.00)
' Assume ~10 pcm / ppm or 110 ppm boron = -1100 pcm E0.503 CR 1 1-Ko 1-Ko = 5000 pcm CO.253 CR 2 1-K1 1-K1 = 3900 pcm E0.25]
CR1
--- = 5000/3900 = 1.282 CO.253 fga k The original shutdown margin was INCORRECT. CO.53 n ac.m ,, n u . a.,--
CR1/CRo should be equal to 2.0 EO.253 REFERENCE LO-LP-33501-OO p.18 LO-LP-33301-OO p. 5 Learning Objective #10. State and equation for count rate ratio (CRR) and be able to use that equation to solve for count rates or Keff , as applicable.
- 3 Define differential boron worth, and give a
" rule of thumb" value of dif f erential boron worth at operating temperature.
192OOOK103 ...(KA'S)
ANSWER 5.08 (1.00)
In the secondary system there is a phase change {0.5} (a phase change requires a large delta h). Due to the larger delta h of the secondary, the same heat can be transferred with a lower flow rate (0.5).
REFERENCE General Physics Heat Transfer - Fluid Flow sec. 3.2 .
OO2OO1K501 ...(KA'S)
)
ANSWER 5.09 (1.00) l \
! b 1
i l l REFERENCE l l SHNPP Thermo-LP-1.1 and Steam Tables Cook-Westinghouse Thermal Science, Chapter 2 p. 63-70 1
1 i
__ .______-____ A
a
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-5. THEORY OF NUCLEAR POWER PLANT OPERATIONt _F1UIQSz_AND PAGE 32 It!E8t]OQyNAtllCS ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS OO1000K550 OO2OOOK501 OO2020K506 ...(KA'S)
ANSWER 5.10 (2.00)
- a. FALSE (0.5 each)
- b. FALSE
- c. TRUE
- d. TRUE REFERENCE LO-LP-33501-OO p.13 OO1010K516 ...(KA'S)
ANSWER 5.11 (1.00)
Necessary to prevent c&vitation in the condensate pump.
REFERENCE HTFF Part B Chapter 2 193OO4K111 ...(KA'S)
ANSWER 5.12 (1.00) d.
REFERENCE HT-FF , Steam Tables OO2020K506' ...(KA*S)
ANSWER 5.13 (1.00) ,
R4 ed
- a. E 1. 00 3 ./' :
/
REFERENCE LO-LP-J3 03-00, p.6 OO;o10A404 O ...(KA'S) l
r !
- i. ';
- 5. THEORY OF NUCLEAR POWER PLANT OPERATIONt _ FLUIDSt_AND PAGE 33 IHEBOODYN@dlGS ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS. j l
q 1
ANSWER 5.14 (1.00) i 4
- d. !
REFERENCE Westinghouse Reactor Physics, Sec. I-5, MTC and Power Defect
-DPC, Fundamentals of Nuclear Reactor Engineering 039000A205 ...(KA'S)
ANSWER 5.15 (1.25)
- a. heat source (0.25 each)
- b. heat sink
- c. some change in elevation
- d. RCS must be subcooled
- e. PZR level (>or= 25%) to ensure RCS is solid REFERENCE LO-LP-34110-OO p.4 Learning Objective #1 193OO8K121 ...(kA'S)
ANSWER 5.16 (1.00) b.
REFERENCE Westinghouse Reactor Physics 1-5 p.26/27 SHNPP RT-LP-1.10 p.13-15 VEGP LO-LP-33201-OO OO1000K549 ...(KA'S)
ANSWER 5.17 (1.00) b REFERENCE NUS Nuclear Energy Training Module 3 Unit 6 Westinghouse, Reactor Physics, Sect. 3 Neutron Kinetics and Sec. 5 HBR, Reactor Theory, Core Physics, Sessions 20 and 24-31 Zion, NUS book, 3 sections 6.5, 6.6, 6.7, 12.4, 12.5 OO1000K513 ...(KA'S)
l.
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- 5. - THEORY OF NUCLEAR POWER PLANT OPERATIONt_FLUIQS t_AND PAGE 34
.IBE8dgQYNAdlCS ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS i
l 1'
ANSWER 5.18 (2.00)
, a. DNBR decreases (0.5 each) l
- b. DNBR increases
- c. DNBR decreases
ANSWER 5.19 (1.00)
- a. DNBR > 1.3 (0.50)
- b. PCT < 4700 deg. F.
REFERENCE LO-LP-36001-OO-C p. 13 Learning Objective # 6 193OO9K105 ...(KA'S)
ANSWER 5.20 (1.00)
- a. quality = (h - hf)/ hfg = (1190 btu /lbm - 536.8)/ 657.5
= 9 9 . 3 '/. CO.353
- b. as-read-from-steam-tabl.as_ L.3954 [O.353 D.9 9J fo. M/?) + D 73 ?? s. tye y y o.ose)
- c. Not superheated subcooled CO.303 REFERENCE HT-FFFFFFFFFFFFFFFFFF 193OO3K125 ...(KA*S)
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_ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ -
'. .i
[
h -_IHEgg LgE_Ngg(E@B_EgWE6_E(@NI_ GEE 6@llgNt _E(ylDh ,@ND PAGE 35 THERMODYNAMICS
~
l ANSWERS -- VDGTLE-1 -87/04/13-BALDWIN, RS l
ANSWER 5.21 (2.50)
- a. SAME AS
- b. HIGHER THAN
- c. HIGHER THAN
- d. SAME AS
- e. LOWER THAN REFERENCE LO-LP-33604-OO Learning Objective #8 (
OO1000K104 OO1000K534 OO1010A201 ...(KA'S) 1
- 6. PLANT SYSTEMS DESIGN 2_CQNTRQL t_@NQ_JN51RUMENJ@l]QN PAGE 36 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS a
ANSWER 6.01 (2.50) gd 46 (0.5 each) ?/ 3
- a. Max. fuel Element Cl addi ng Temp. < 2/00 deg. F.
- b. Cladding Oxidation < 17% thickness '
- c. Hydrogen generated by Zirc-water reaction < 1% of max. possible.
- d. Core remains in a coolable geometry.
- e. ' Provides for long term decay heat removal l
REFERENCE 10CFR 50.46 VEGP LO-LP-13OO1-OO Learning Objective . State the ECCS Acceptance Criteria OO6050 GOO 4 ...(KA'S) i ANSWER 6.02 (1.00) i b E1.00] j REFERENCE LO-LP-16301-02-C Learning Objective #3. Describe the purpose of the major components associated directly with the pressurizer system.
010000A401 ...(KA'S) {
ANSWER 6.03 (1.00) d [1.003 REFERENCE VEGP LO-LP-17101-OO p. 16/17 Learning Objective #4. List all alarms, trips and blocks associated I with the SRNS. Include all setpoints, coincidence and permissives.
015000K401 ...(KA'S) l l
1
- 6. PLANT SYSTEMS DESIGN t_CONTROLt_AND_ INSTRUMENTATION PAGE 37 f ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS
' ANSWER, 6.04 (1.50)
- a. TRUE (0.5 each) ;
- b. TRUE c.- FALSE REFERENCE VEGP LO-LP-27101-OO p. 6/7 Learning Objective #3. Describe the function of each of the f oll owing; i, c.
OO1000K403 ...(KA'S) i ANSWER 6.05 (1.75)
Sufficient water available to maintain the RCS at Hot Standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (1.00) with a steam discharge to the atmorphere (0.25) with a total loss of offsite power (0.25), f ollowed by a cooldown to RHR initiation conditions (0.25).
REFERENCE VEGP Technical Specifications Bases 3/4 7.1.3 026000 GOO 5 ...(KA'S)
ANSWER 6.06 (1.00) a C1.003 REFERENCE VEGP Chapert 12b p. 12b-8 04102OK417 041020K418 . . . (K A' S)
ANSWER 6.07 (2.00) gm g
- 1. PZR Hi Press. Trip (0.2) 2385 psig (0.1), 2/4 (0.1)
- 2. PZR Lo Press. Trip (0.2) 1960 psig (0.1), 2/4 (0.1)
- 3. PZR Lo-Lo Press. SI (0.2) 1850,_ psi g and not blocked (0.1),
- 2/4(0.1) 4 P-11 (0.2) 1970 psig (0.1), on 2/3 (0.1)
- 5. Press. Input to the OT Delta T (0.4) '
fp. Asdt C.LCS A 'cf fld [bM'J t%Jr) 6b/ Qw(Au 6 u.i (Ch L) LC% TilAd DO[ h REFERENCE F*}
VEGP LO-LP-16303-OO, Logics Chapter 25, p. 10/11 010000K101 ...(KA'S)
.- j
=
- i 6 __PL@NI_gy@ led @_pE@lGyt_CQNI@gLt_@ND_lNgl@yDENI@ll@N PAGE 38 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS i
l ANSWER 6.08 (1.00)
- a. Serves as a backup to the undervoltage coils on an auto trip signal. (0.5)
- b. An auto trip signal (0.25) de-energizes the auto shunt coil which causes contacts to close that energize the shunt trip coils. (0.25)
REFERENCE VEGP LO-LP-28101-OO-C p.16 012OOOK604 ...(KA'S)
ANSWER 6.09 (1.50)
( 3aL % 'l) O ss..~, s r e w e -a s w.tswJ nwc ~~~'"* 4"")
- 1. Adequate shut down margin. (0.5 ea) c, 5,tura,s,,s u m ,,s ~4e~7,w m /F o^di ^'4
- 2. Minimize ef f ects of ejected rod, sru ocor rwo (ceA's Srucx duo LJa M 440-
- 3. Uniform flux profile. b- TLP AA
- V'W A*"'O W "#
- M S u .
(r) . &sma fwew /. m MA*Sd<
REFERENCE #'
VEGP LO-LP-27102-OO p.12 ch
- '*eacy ) WM '" b M E"l\ ""
d Tu vscac DwNL.
OO1000K504 ...(KA'S) 8, u.,cc.aicatur> RecA w, mow % sot Ta
@ CCL9 A CcAnccrJr At4WYitS.
A pu M u.a nc. n w c . w runhi x c k T u u a
ANSWER 6.10 (1.50) g ,a y ,a m g w u w w ey w m u w % % W
^ '
- a. OTSP Decreases
- b. OTSP Decreases (0.5 each)
C. OTSP Decreases REFERENCE Chapter 25 Logics sheet 5.
012OOOA101 ...(KA'S)
ANSWER 6.11 (1.00)
The highest reading upper / lower detector is compared to the average of the upper / lower detectors (0.5). The circuit auto def eats below 50%
power on all channels (0,5)
REFERENCE VEGP LO-LP-17301-OO p.10, 12, 13, 14 Learning Objective #6 015000A104 015000K604 ...(KA'S) 1
'. l l 6___PbgyI_pypIggg_Qgp]QNg_CQNIgght_9NQ_]N3I690ENISIlgN PAGE 39 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 6.12 (1.00)
- a. They may be re-energized by pushing the SI override push button and closing the breaker (resetting the S1 and closing the breaker) 1 (0,5) accept either answer.
- b. (Administrative control to) Prevent loading of Non-1E loads onto the Diesel Generators before the 1E loads are energized. (0.5) og s.r
' '~ # ** " "'* # * #' * **** * 'i'W I "'**"
REFERENCE #"* * & 5) b VEGP LO-LP-01102-02-C i 013OOOK401 ...(KA'S) I ANSWER 6.13 (1.00) b REFERENCE i LO-LP-16401-OO p.13
.OOOO17EK30 ...(KA'S) i ANSWER 6.14 (1.00) l
- a. Only one RCP started at a time (0.25)
- b. Two successive starts allowed if the motor is permitted to coast to a stop between starts. (0.25)
- c. Third start allowed when windings and core have cooled by; l
- a. running for 20 minutes, (0.25) or
- b. standing idle for 45 minutes (0.25) !
\
REFERENCE VEGP LO-LP-16401-OO p.14 Learning Objective #7 . State the RCP duty limits.
VEGP 19003-1, ES-0.3 Nat. Circ. C/D with Stm Void in Vessel.
003000G001 ...(KA'S) l
l l
', 4 s
&g__ PLANT SYSTEMS DESIGNt_CQN{RQLt_ANQ_lNSTRQMENTATIQN PAGE 40 l ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS l
l l
l ANSWER 6.15 (1.50) y,g A
- a. 2/4 indication on 1/4 S/Gs below 1777 f CO.23
-Rx trip EO.23
-Starts MDAFW pumps air E O. 23.
l b. 2/4 indication on 2/4 S/Gs below 1?%'pt E0.23
-Starts TDAFW pump (h/O E 0. 23 l- c. Purpose-Protects Rx from loss of heat sink by leaving enough volume until AFW can be established. [0.53 REFERENCE VEGP LO-LP-18301-OO-01 p.7 Learning Obj ective. State the protection f eatures provided by;
- a. S/G low low water level ....
035010K401 ...(KA'S)
ANSWER 6.16 (1.00)
- a. Insufficient steam flow f or proper circuit operation EO.53
- b. FRV so large that it would have to be cracked open thus causing seat and disk erosion. [0.53 REFERENCE VEGP SGWLC System p. 13b-3 059000K609 ...(KA'S)
ANSWER 6.17 (2.00)
- 2. Shuts Feedwater Isolation valves;
- a. Main EO.23
- b. Bypass [0.23
- b. 1. Safety Inj ection Signal (0.5)
- 2. S/G Hi-Hi level (2/4), (78%), (P-14) (O.5)
REFERENCE VEGP LO-LP-19101-OO-C 059000A306 ...(KA'S) l
L l . .
{**
.[ s
- 6. PLANT SYSTEMS DESIGN g_CQNTRQL t_AND_ INSTRUMENTATION PAGE 41 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS 1
L l
l l
ANSWER 6.18 ( 1.pO)
(df 0
/ AT Allows f or RCS cooldown without Steam Dumps (0.5)
Minimizes lifting of passive code saf ety valves on 7 ,b<.
transi ents. (0.5)
Reduces steam line pressure suf ficiently to ensure reseating J Jr. of code safeties. (0.5)
REFERENCE
+( ApMn,e sawfQ9 gudjanWmo f mN~<rf NA&W h)
'VEGP-LO-LP21101-OO p.11 Learning Objective # 3. For each component on the drawing the student should know, ...
- a. Purpose of the components.
039000K102 ...(KA'S)
ANSWER 6.19 (1.50)
- a. manual, (1/2) [0.53
- b. Saf ety Injection signal, 1/2 CO.53
- c. Containment area radiation, 1/2 CO.53 (Both signal and logic required for any credit)
REFERENCE Safeguard Actuation System Fig. 7.2.1-1 (Sheet 8 of 20) 013OOOK402 ...(KA'S)
ANSWER 6.20 (2.00)
CAny 4 at .5 each3
- a. step counters
- b. master cycler counters
- c. sl ave cycl er counters
- d. bank overlap counters i
- e. internal memory and alarm ekts l
- f. pulse to analog converters f or RPI .
REFERENCE VEGP LO-LP-27101-OO p.15 Learning Objective #4. Describe all input signals to the Rod Control System and explain the source of these signal s.
OO1010K402 ...(KA'S)
_ - _ - _ - _ _ _ - _ - _ _ _ _ _ _ = _ - - - _ _ - _ - _ _ _
4 l
l'= . ,
)
- 6. - PLANT SYSTEMS DESIGN t_CONTROLt AND INSTRUMENTATION PAGE 42 l t
> ANSWERS -- VOSTLE 1 -87/04/13-BALDWIN, RS {
i ANSWER 6.21 ( .75) f p'.
( hW Limit cooldown rate during Main Steam Line Break CO.253
\
Limit thrust and pipe whipping during MSLB (0.253 I 2 tr.
3 g. Produces a delta P for flow measurement- CO.253 l yr gia,.jp ,,,bd s &r.ssa.& W J 4. " vru-h.- (O.a)
REFERENCE 5 Liw r gn,*iow v' ca..vrirudec,v; suen_ (o. u-j VEGP LO-LP-21102-OO p.9 Learning Obj ective B Design Bases--The student should know the
- ,qg mg g ggf important design criteria used on
, the Main Steam System that relate to normal and accident conditions.
l 035000 GOO 7 . . .(KA'S) l ANSWER 6.22 ( .50) l FALSE REFERENCE VEGP LO-LP-09201 p.11 OO4000K604 . .. (KA'S)
ANSWER 6.23 (1.00) ,
(byy twf et 6.)
l 8. To prevent pump run out. 3 "** # A -'3 C 8 7d'fM N 84 Ch # "O 2 Edw1ute:- fi.aJ sdra w LE4 W W fli5ip*It (Wb (Lod l
REFERENCE
'VEGP LO-LP-09202-01 CVCS alternate Flow Paths Learning Objective #3. Describe the purpose of the needle valves on the High Head SI lines into the RCS. i i
OO4000K604 . . .(KA'S) [
1 !
l
)
l 1
l l
l l
l l
= *
- 7. PROCE.7URES - NORMAL _t ABNORMAL _t EMERGENCY _AND PAGE 43 88 Dig (QGlC@(_ggNIBQL ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 7.01 (1.00) b REFERENCE l VEGP EOP-E-0, 19000-1 194001A102 ...(KA'S)
ANSWER 7.02 (1.50)
) a. NO
- b. NO .
ANSWER 7.03 (1.00) a REFERENCE VEGP 1900-1, Reactor Trip or Safety Injection OO6000 GOO 1 ...(KA*S)
ANSWER 7.04 (1.00)
Because none of the electrically powered saf eguards equipment used to restore Critical Saf ety Functions is operable C1.003 (none of the FRG's can be implemented)
REFERENCE Westinghouse Background Information for ECA O.0 Users Guide for ERG's and Background Documents p. 17/18 OOOO55EK30 ...(KA'S)
- ~__
- 7. PROCEDURES - NORMAL _ABNgRMAL t _tEMERGENCY _AND PAGE 44 B001969 GIG 06_GQNIBQL l ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS l
l ANSWER 7.05 (1.50)
This is because you could be in E-3 without an intact S/GEO.53, so l that feed flow maybe throttled below the minimum CO.53 to prevent S/G overfill CO.25] or if feeding a faulted S/G tp prevent RCS overcooling
[0.253.
REFERENCE Westinghouse Background Information on SI Termination / Re-initiation p5 OOOO9EA234 ...(KA'S) l ANSWER 7.06 (1.50)
.a . ma *4-mum---[4}r5-3 0
- b. 1) CLOSE CO.253
- 2) OPEN CO.253
- c. 1) OPEN [0.253
ANSWER 7.07 (1.00)
,.a- (_. OA D ff6 nUtk ,
REFERENCE f VEGP 18003-1, Rod Control System Malf unction OOOOO3EK30 ...(KA'S) ]
l i
ANSWER 7.08 (1.00) ]
C l
REFERENCE I I
VEGP -FR-S. 1,[19211-13, Response to Nuclear Power Generation / ATWT OOOO29EK31 ...(KA'S) 1
L
- 7. PROCEDURES -- NORMAL t_ABNQRMAL _t EMERGENCY _AND PAGE 45 l 689196QGlG66_GQNIBQL ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS l l J
ANSWER 7.09 (1.00)
- a. FALSE CO.5 each J
- b. TRUE REFERENCE VEGP-192OO-1, F-0, Critical Safety Function Status Trees, p.2 189001A102 ...(KA'S)
ANSWER 7.10 (1.50) i If S/G 1evel increases to greater than 96%, steam released from the S/G may result in two-phase flow [O.53 and or water hammer [0.53, which may potentially damage, piping, valves, turbines or personnel CO.5].
REFERENCE VEGP-19232-1 035000 GOO 1 ...(KA'S)
ANSWER 7.11 (2.00)
- a. YES
- b. YES
- c. NO -
- d. YES i
REFERENCE VEGP-13OO9-1, CVCS Reactor Makeup Control System i OO4010K609 ...(KA'S)
ANSWER 7.12 (2.00)
- a. YES
- b. YES.
- c. NO
- d. YES REFERENCE VEGP 18003-1, Rod Control System Malfunction, p.3,8,11.
OO1000 GOO 1 ...(KA'S) .
{ <
L.. .,g ..
Zt__eBQCEQQBES_;_NQBd@(t_@@NQBd@(z_EDEB@ENCy_@NQ PAGE 46 f 86019690lG86_CQNIBQL q ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 7.13 (1.00)
The pump starting current may trip the Diesel Generator Breaker [1.03 <
REFERENCE VEGP-13OO3-1, RCP Operations 064000A203 ...(KA'S)
ANSWER 7.14 (1.00)
To prevent thermal lock up of the rods. E1.003 REFERENCE VEGP-12OO2-1 (Mode 3 to Mode 4) p.1 OO1000G991 ...(KA'S)
ANSWER 7.15 (1.00)
- a. 125 [0.25 each]
- b. 50
- c. 100
ANSWER 7.16 (1.50)
- 1. c
- 2. b [0.25 each in proper order] '
- 3. d
- 4. e
- 5. +
l 8
L . _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- . t =.
- 7. PROCEDURES - NORMAL _AQNQBMAL t _EMEBGENCY_@ND t PAGE 47 i Q@DlO60@lG@6_QONIB06 )
ANSWERS -- VC1TLE 1 -87/04/13-BALDWIN, RS 1
ANSWER 7.17 (1.00) .
1 C
REFERENCE VEGP E-OM 19001-1, Reactor Trip or Safety Inj ecti on
~194001A102 ...(KA'S)
ANSWER 7.18 (1.00) d [1.003 REFERENCE VEGP 19100-1, ECA O.0 p.5 013OOOGOO1 ...(KA'S)
ANSWER 7.19 (1.00)
The sequence of per f ormance is not important. E1.003 REFERENCE Westinghouse Users Guide, ERG's and Background Documents, p.3 194001A102 ...(KA'S)
ANSWER 7.20 (1.00)
The operator should return to the next step or substep in the left-hand column. [1.003 REFERENCE Users Guide for ERG's and Background Documents p.5 194001A102 ...(KA'S) i k . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ -
- 7. PROCEDURES - NORMAL _t ABNORMAL _t EMERGENCY _AND PAGE 48 68 Dig 6QGICQ(_CQNIBQL i
ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER 7.21 (1.00) #'q ,,, ~og Awr 5/6 fjyy 4/s bsua hw'dq +>, w'"*""'
- a. faulted = tef m n LU f eilm e Of-ttte setortder-y-pressur-e
, n a _, r u r .
4 _.b ~~ E Q 6epa_ dog unm4..L
.ij;/s1
/.# [+ C
- b. isc3ated = refers to closure of.the steam and / or feed flow paths. [0.53 i
REFERENCE
. Users Guide for ERG's and Background Documents OOOO37EK30 ...(KA'S) p.7 yf[O2O-/,
t L
- p. t s--
ANSWER 7.22 { 1JW MD
-a. --T R u t e'C# hCU A WN
- b. FALSE
- c. =TRUE Frfc scE-REFERENCE M 57r/i'P Westinghouse Users Guide to ERG's and Backgcound documents.
194001A102 ...(KA'S)
ANSWER 7.23 42.00)f4'P
) a. RCS subcooling monitor indication EO.253 greater than 28 deg. F E0.253
- b. RCS hot leg temperatures CO.25], stable or lowering E0.25]
- c. Core exit thermocouple CO.25], stable or lowering [0.253
- e. S/G pressures E0.253, stable or lowering E0.25]
REFERENCE VEGP LO-LP-34110-OO p.6 ES-0.1, 19001-1 p.16 193OO8K122 ...(KA'S)
I
f
-8. ADMINISTRATIVE PROCEDURESt _CQNQITIQNSt_ANQ_ LIMITATIONS PAGE 49 I
ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS l
l ANSWER B.01 (1.50)
- a. Containment structure is prevented from exceeding its design I
negative pressure differential with respect to the outside atmosphere l of 3 psig. EO.53
- b. The containment peak pressure does not exceed the design pressure of 52 psig CO.53 during steam line break conditions. EO.53 REFERENCE VEGP T.S. 3.1.6.4 and B 3/4.6.1.4 p.B 3/4.6-1 OO2OOOGOO6 ...(KA'S) l ANSWER 8.02 (1.00)
The resultant off-site radiation dose will be limited to a small fraction of 10CFR100 guideline values CO.53 in the event of a steam line rupture CO.53.
REFERENCE VEGP T. S. 3.7.1.4 and B 3/4.7.1.4 p. B 3/4 7-2 035000 GOO 6 ...(KA*S)
ANSWER 8.03 (1.25)
- a. Saf ety Limit (thermal power) has been violated (0.25)
- b. 1. NRC operations center shall be notified by telephone as soon as practical in all cases within one hour after the ;
violation has been determined. (0.25)
- 2. An LER shall be prepared (0.25)
- 3. An LER shall be submitted to the NRC within 30 days (0.5)
- 4. Critical operation of the unit shall NOT be resumed until authorized by the NRC.
I REFERENCE VEGP T. S. 6.6.1 p.6-13 ;
OO1000 GOO 5 ...(KA'S) l I
i l
4
____________________b
)
V ,
, ., .. I 1
l a___0EMIN1SIg@I1VE_P6gCEQU6ES t _ggNQlligNSt _@NQ_(IMlI@IlgNS PAGE 50 l
ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS I
ANSWER 8.04 (1.50)
- a. The intent of the procedure is not altered E0.53
- b. The thange is approved by two members of the plant staff, at least one of whom holds a senior operator license, E0.53 and
- c. The. change i s documented, reviewed and approved by the GMVNO or department head of the responsible department [0.25]
within-14 days of implementation. [0.25]
REFERENCE VEGP T.S. ab.7.3 p.6-14 ;
194001A102 ...(KA'S)
ANSWER 8.05 ( .50)
TRUE REFERENCE VEGP T.S. Bases B3/4 1-4 OO1000 GOO 6 ...(KA'S)
ANSWER 8.06 (2.00)
- b. In the event of a LOCA the peak fuel clad temperature will not exceed the 2200 deg. F. (ECCS acceptance criteria limit)
REFERENCE ,
-VEGP T.S. B 3/4.2.2, 3/4.2.3, p. B3/4 2-2 I OO1000 GOO 6 ...(KA'S)
ANSWER B.07 (1.50)
,i
- a. 5 I
- b. On Shift Operations Supervisor )
- c. 4 l 1
REFERENCE Manning the Shift 10003-C p-2 194001K116 ...(KA'S)
et__0901Ni@lB@llVE_P69CEgyBE@t_CQNQlligNgt_@NQ_Lldll@llgN@ PAGE 51 1
-ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER B.08 (1.00)
The individual should gOT be allowed to enter a Radiation Control Area when the exposureg isggca,atar than or equal to 2OOmrem.A E1.003 us -
REFERENCE 7 VEGP OO920-C h k t+c AtJ M N 58 4 LA 6 M T 194001K104 ...(KA'S) gvwuot., on, tryage )
' ANSWER- 8.09 (1.00)
- a. Health Physics Superintendent E0.53
ANSWER B.10 (1.25)
YES, CO.253 there is a violation , can r.ot work more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period. E1.003 REFERENCE VEGP T.S. 6.2.2.e.2 p. 6-2 Administrative Procedure Overtime Authorization, OOOO5-C 194001A103 ...(KA'S)
ANSWER 8.11 (1.00)
C REFERENCE VEGP T.S. Definition p.1-1 194001A113 ...(KA'S) l l
ANSWER B.12 (1.00)
C l
l l
1
_ ___._.____________________________.__Q
i l
.. .. . ., .. 1 B. ADMINISTRATIVE PROCEDURES _C_QNDIT_IQN@t_AND_LIMITATIQN@
t PAGE 52 I ANSWERS -- VOGTLE 1' -87/04/13-BALDWIN, RS REFERENCE VEGP T.S. p. B 2-6 011000 GOO 6 ...(KA'S) J
{
l 1
i 8.13 I ANSWER (1.00) k
- a. FALSE
ANSWER 8.14 (2.00)
[0.5 each situation must be stated completely]
- a. Safety of the Reactor is in jeopardy, or
- b. Reactor operating parameters exceed reactor trip setpoints and automatic shutdown does not occur, or
- c. It is required to protect personnel and equipment, or
- d. Unusual circumstances warrant it.
REFERENCE VEGP OO3OO-C p. , Authority to S/U or S/D Reactors 002000G001 ...(KA'S) l ANSWER 8.15 (1.50) i, ou> d n <t cA04t h d w a m m.m du w w w ciwa m o m ce w w Jovc.' d a . 2. On Sh i f t Operations Supervisor AND Depar tment Supervi sore sa<w. u4~u *%
3, y asA@%, ou ew . Punar sur.ur ^~ o*w w wuteme n aaw re 'a t i , oo 3 '
- N " -'*M on . o.o
- w u.r u-t
- b. The<y are responsible for providing notification to the SUBCLEARANE HOLDER individual when he/she returns to work,CO.53 if acts w co m aar cg eamn ,,n a o s c u nr, REFERENCE VEGP-OO304-C, P.4 Equipment Clearance and Tagging 194001A102 ...(KA'S)
- . - - - - - .----._--,-.--_--..---__-.-s___--J
.,.,c, *.
- 8. ADMINISTRATIVE PROCEDURESt _CgNDITIONSt_AND_ LIMITATIONS PAGE 53 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS I
ANSWER 8.16 (1.50)
- a. NORMAL (Modes 1 to 4)
OSOS STA*
DAD EXTRA LICENSE (1)
EXTRA NON-LICENSE (1)
- Nrt necessary when OSOS meets the qualification requirements f or a STA.
- b. MINIMUM (Modes 5 & 6)
CPOINT VALUE O.1 EA.,
- O.23 REFERENCE -
VEGP-10003-C p.3 194001A103 ...(KA'S) ,l l
1 1
ANSWER 8.17 (1.00) 1
\
Apply T.S. 3.0.3 (initiate action with in one hour to place unit in i mode 5 or 6 in which noncompliance T.S. does not apply.
REFERENCE VEGP T.S.L 3.5.2, 3.0.3, 3.0.4 OO5000 GOO 6 ...(KA'S) l ANSWER B.18 (1.00)
Immediately initiate and continue boration CO.25] at greater than or I equal to 30 gpm [0.25] of a solution containing greater than or equal l to 7000 ppm CO.253 boron or equi val ent until the required SHUTDOWN MARGIN is restored CO.253.
I; 4
l l
b _ ___-_______
l t ., s
- 8. ADMINISTRATIVE PROCEDURESt _CgNDITIQN@t_AND_LlMITATIQN@ PAGE 54 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS l REFERENCE-VEGP T.S. 3.1.1.2 p. 3/4 1-3 l OO1000 GOO 6 ...(KA'S) l l ANSWER 8.19 (1.00) c REFERENCE -
VEGP T.S. 3.4.4 010000A203 ...(KA'S)
ANSWER B.20 (1.00)
- a. FALSE
ANSWER 8.21 (1.00)
- a. On Shift Operations Supervisor EO.5]
- b. Shift Supervisor CO.5]
REFERENCE
.VEGP-91102-C p.1 Duties of the Emergency Director i 194001A103 ...(KA'S)
ANSWER 8.22 (1.00)
I Immediately open the reactor trip breakers. l I
REFERENCE VEGP T.S. 3.1.3.3 p. 3/4 1-18 014000K301 ...(KA'S)
@ __090lN1@Ig@IlVE_E6ggEgyBE@t_ggdglligN@t_@Ng_L1glI@IlgN@ PAGE 35 ANSWERS -- VOGTLE 1 -87/04/13-BALDWIN, RS ANSWER B.23 .(1.00) -
1 4
d REFERENCE VEGP T.S. 3.4.9.3 p. 3/4 3-4 002000G007 ...(KA'S) l
\ - _ - _ - - - - - - - - - - -
- s, e
. . ., e, ..
-TEST CROSS REFERENCE PAGE- 1 QUESTION VALUE REFERENCE I
05.01 1.00 BSROOOO335' 05.02 1.00 BSROOOO336 05.03. 1.50 BSROOOO337 05.04 1.50 BSROOOO338 05.05 1.25 BSROOOO339 05.06 2.00 BSROOOO340 05.07 2.00 BSROOOO341 oO5.08- 1.00 BSROOOO342 05.09 1.00 BSROOOO343 05.10 2.00 BSROOOO344 05.11 1.00 BSROOOO345 i 05.12 1.00 BSROOOO346 !
05.13 1.00 BSROOOO347 05.14 1.00 BSROOOO348
-05.15 1.25 BSROOOO349 05.16 1.00 BSROOOO350 05.17 1.00 BSROOOO351 05.18 2.00 BSROOOO352 05.19 1.00 BSROOOO353 05.20 1.00 BSROOOO354 05.21 2.50 BSROOOO355 28.00 06.01 2.50 BSROOOO356 06.02 1.00 BSROOOO357 06.03 1.00 BSROOOO358 06.04 1.50 BSROOOO359 06.05. 1.75 BSROOOO360 06.06 1.00 BSROOOO361 06.07 2.00 BSROOOO362 06.08 1.00- BSROOOO363 06.09 1.50 BSROOOO364 06.10 1.50 BSROOOO365 06.11 1.00 BSROOOO366 06.12- 1.00 BSROOOO367 06.13 1.00 BSROOOO368 06.14 1.00 BSROOOO369 06.15 1.50 BSROOOO370 06.16 1.00 BSROOOO371 06.17 2.00 BSROOOO372 06.18 1.50 BSROOOO373 06.19 1.50 BSROOOO374 06.20 2.00 BSROOOO375 06.21 .75 BSROOOO376 06.22 .50 BSROOOO377 06.23 1.00 BSROOOO378 -
l ______
30.50 07.01 1.00 BSROOOO379 i
l I
s i
, , e-ae e,' .', *e TEST CROSS REFERENCE PAGE l2 QUESTION VALUE REFERENCE 07.02 1.50 BSROOOO380 .
BSROOOO381 I 07.03 1.00 07.04 1.00 BSROOOO382 07.05 1.50 BSROOOO383 l
07.06 1.50 BSROOOO384 07.07 1.00 BSROOOO385 i 07.08 1.00 BSROOOO386 07.09 1.00 BSROOOO387
]
07.10 1.50 BSROOOO388 07.11 2.00 BSROOOO389 07 12 2.00 BSROOOO390 ,
l 07.13 1.00, BSROOOO391 07.14 1.00 BSROOOO392 1 07.15 1.00 BSROOOO393 07.16 1.50 BSROOOO394 07.17 1.00 BSROOOO395 07.18 1.00 BSROOOO396 07.19 1.00 BSROOOO397 07.20 1.00 BSROOOO398 07.21 1.00 BSROOOO399 07.22 1.50 BSROOOO400 07.23 2.00. BSROOOO424 29.00 08.01 1.50 BSROOOO401 08.02 1.00 BSROOOO402 08.03 1.25 BSROOOO403 08.04 1.50 BSROOOO404 08.05 .50 BSROOOO405 08.06 2.00 BSROOOO406 08.07 1.50 BSROOOO407 08.08 1.00 BSROOOO408 08.09 1.00 BSROOOO409 08.10 1.25 BSROOOO410 08.11 1.00 BSROOOO411 08.12 1.00 BSROOOO412 08.13 1.00 BSROOOO413 08.14 2.00 BSROOOO414 08.15- 1.50 BSROOOO415 08.16 1.50 BSROOOO416 08.17 1.00 BSROOOO417 08.18- 1.00 BSROOOO418 08.19 1.00 BSROOOO419 08.20 1.00 BSROOOO420 08.21 1.00 BSROOOO421 08.22 1.00 BSROOOO422 08.23 1.00 BSROOOO423
{
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- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2.1 SAFETY LIMITS l
REACTOR CORE 2.1.1 The combination of THERMAL POWER (NI-0041, NI-0042, NI-0043, NI-0044),
pressurizer pressure (PI-0455A, B&C, PI-0456 & PI-0456A, PI-0457 & PI-0457A, PI-0458 & PI-0458A), and the highest operating loop coolant temperature (T gg)
(TI-0412, Figure TI-0422, TI-0432, TI-0442) shall not exceed the limits shown in 2.1-1.
APPLICABILITY: MODES I and 2.
ACTION:
Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure ments of line, be in HOT6.6.1.
Specification STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the require-REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure (PI-0408, PI-0418, PI-0428, PI-0438) shall not exceed 2735 psig.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.6.1.
MODES 3, 4 and 5:
Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.1.
- Where specific instrument numbers are provided in parenthesis they are for information only (to assist in identifying associated instrument channels or loops) and are not intended to limit the requirements to the specific instruments associated with the number.
V0GTLE - UNIT 1 2-1
f .
.c 680 UNACCEPTABLE OPERATlON A-660 -
2400 PSI A N N 640 = %
3 2250 PSIA
{ ~
$ 2000 PSIA 620 u 1775 PSIA i N \
g* N N N E
580 T )
~
ACCEPTA8LE OPERATION 560 540 0 0.2 0.4 0.6 0.8 1.0 12 i FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT V0GTLE - UNIT 1 2-2
, .. l
, i 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENT 5" ,
1 3/4.0 APPLICABILITY
{
l LIMITING CONDITION FOR OPERATION !
3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERCIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met, except as provided ,
in the associated ACTION requirements, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to place the unit in a MODE in which the specification does not apply by placin0 it, as applicable, in: i
- a. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. At least HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- c. At least COLD SHUT 00WN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for '
Operation. Exceptions to these requirements are stated in the individual specifications.
This specification is not applicable in MODE 5 or 6.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exce stated in the individual specifications. ptions to these requirements are "Where specific instrument numbers are provided in parenthesis they are for information only (to assist in identifying associated instrument channels or loops) and are not intended to limit the requirements to the specific instruments associated with the number.
e APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
a.
A maximum allowable extension not to exceed 25% of the surveillance interval, but
- b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i);
1 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T, g GREATER THAN OR EQUAL TO 350'F LIMITING CONDITION FOR OPERATION
- 3. 5. 2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE Safety Injection pump,
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and semi-automatically transferring suction to the containment emergency sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2, and 3.*
ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to j the Commission pursuant to Specification 6.8.2 within 90 days describ-l ing the circumstances of the actuation and the total accumulated i actuation cycles to date. The current value of the usage factor l for each affected Safety Injection nozzle shall be provided in this l Special Report whenever its value exceeds 0.70.
c
- The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into Mode 3 for the Safety Injection Pumps declared inoperable pursuant to Specification 3.5.3.2 provided the Safety Injection Pumps are restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever occurs first. '
l V0GTLE - UNIT 1 3/4 5-3 l - __D
s EMERGENCY CORE COOLING SYSTEMS-SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS' subsystem shall be demonstrated OPERABLE:
- a. At-least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with power lockout switches in the lockout position:
Valve Number Valve Function Valve Position HV-8835 SI Pump Cold Leg. Inj. OPEN HV-8840 RHR Pump Hot Leg. Inj. CLOSED HV-8813 SI Pump Mini. Flow Isol. OPEN HV-8806 SI Pump Suetion'from RWST OPEN HV-8802A, B S1 Pump Hot Leg Inj. CLOSED HV-8809A, B RHR Pump Cold Leg Inj. OPEN
- b. At least once per 31 days by:
- 1) Verifying that the ECCS piping is full of water by venting the-ECCS pump casings and accessible discharge piping high points, and
- 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the. containment which could be transported to the Containment Emergency Sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1) For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and
- 2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
- d. At least once per 18 months by:
- 1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that:
a) With a simulated or actual Reactor Coolant System pressure s,ignal greater than or equal to 377 psig the interlocks prevent the valves from being opened, and b) With a simulated or actual . Reactor Coolant System pressure signal less than or equal to 750 psig the interlocks will cause the valves to automatically close.
- 2) A visual inspection of the Containment Emergency Sump and verify-ing that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
V0GTLE - UNIT 1 3/4 5-4
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. At least once per 18 months, during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal,
- 2) Verifying that, for the RHR Semi-Automatic Switchover to Contain-ment Emergency Sump, each automatic valve in the flow path actuates to its correct position on RWST level low-low coincident with Safety Injection test signal, and
- 3) Verifying that each of the following pumps start automatically upon receipt of a Safety Injectio.n actuation test signal:
a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.
- f. By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:
- 1) Centrifugal charging pump 1 2364 psid,
- 2) Safety Injection pump 1 1405 psid, and
- 3) RHR pump 1 165 psid.
- g. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
- 1) Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
- 2) At least once per 18 months.
Safety Injection System Valve Number 1204-U4-022 1204-U4-139 1204-U4-023 1204-04-140 1204-U4-024 1204-U4-141 ,
1204-04-025 1204-04-142 1 1204-U4-116 1204-U4-118 1204-04-117 '
1204-U4-119 l l
1 V0GTLE - UNIT 1 3/4 5-5
_- ___ _ __a
1 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- h. By performing a flow balance test, during shutdown, fe' lowing com-pletion of modifications to the ECCS subsystems that etter the subsystem flow characteristics and verifying that:
- 1) For centrifugal charging pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 284 gpm, and b) The total pump flow rate is less than or equal to 550 gpm.
- 2) For Safety Injection pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 406 gpm, and b) The total pump flow rate is less than or equal to 660 gpm.
- 3) For RHR pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3788 gpm.
1 V0GTLE - UNIT 1 3/4 5-6
a
EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T _LESS THAN 350*F 3 ,
ECCS SUBSYSTEMS LIMITING CONDITION FOR OPERATION 3.5.3.1 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the Containment Emergency Sump during the recirculation phase of operation.
APPLICABILITY: MODE 4.
ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T 1ess than 350 F by use of alternate heat removal methods, avg
- c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6 8.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for L
each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
V0GTLE - UNIT 1 3/4 5-7
EMERGENCY CORE COOLING SYSTE't(
SURVEILLANCE REQUIREMENTS 4.b.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.
l l
l V0GTLE - UNIT 1 3/4 5 - - - - - - - - - -- O
7
~ -.
, - ~ - - ~,, . _ _
cut)/(Energy inf
, mg se V,t
- 1/2 at,
- 3 E = me- .
- l. (E = 1/2 mv 2
,,,(jf , ,, ) f 4 , 13 ; , ;,,-it l
PE = m9n Vf = V, + at d = e/t i = In2/t1/2 = 0.693/t1/2 g , , ;p A= nD 2 1/2'N
- b " O 4 [(t1/2)
- Sto))
4.E = 931 sm
. , m = V,yAo -Ex Q .= mth I=Ieo Q = mCpat Q = UAtT I
- I,e~"*
Fwr = Wfsh I = I,10'*/U L TVL = 1.3/u -
sur(t) HVL = -0.693/u P = P,10 P = P,e*II SUR = 26.06/T SCR = 5/(1 - K,ff)
CR, = 5/(1 - K,ffx)
SUR = 26c/t* + (a - o)T CR)(1 - K,ff)) = CR2 II ~ "eff 2)
T = (t*/o) + [(s - oV Io] M = 1/(1 - K,ff) = CR)/CR, T = 1/(o - s) M = (1 - K,ff,)/(1 - K,ff))
T = (s - ;)/(Io) SDM = ( - K,ff)/K,ff a = (K,ff-1)/K,ff = 4Keff/K,ff t= 10 seconos I = 0.1 seconds ~I o = [(L*/(T K,ff)] + [8,ff /(1 + IT)]
=1d
!)d) 2 ,2 2 P = (reV)/(3 x 1010) I)d) 7d 22 2 I = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (feet) ,
Water Parameters _
Miscellaneous Conversions 1 gal. = 8.345 lom. 1 curie = 3.7 x 1010 cps 1 gai . = 3.78 liters 1 kg = 2.21 lom 1 ft* = 7.48 gal 1 op = 2.54 x 103 Btu /nr .
Density = 62.4 1 /ft3 1 m, = 3,41 x joo 5tu/hr Density = 1 gm/c. lin = 2.54 cm Heat of vaporization = 970 Btu /lom 'F = 9/5'C + 32 Heat of fusion = 144 Btu /lem 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. H9 1 BTU = 778 ft-lbf 1 ft. H 2O = 0.4335 lbf/in.
e = 2.718
" NF -
6 ,
l l
l l
l APPENDIX E STEAM TABLES AND MOLLIER DIAGRAM ne following Stamm Tables and Mollier Diagram have been reproduced with the consent and by permission of Combustion Engineeri'. g, Inc., 2000 Prospect Hill Road, Windsor, Connecticut 06095.
1 .
Volume 111 Appendix E 1
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0 1724 :4704 1 9978 es t 0 016123 370 9 370 0 M 003 1038 7 ,1042 0 !?60 .8617 1 9876 Os t ISB 4 0 64974 001613t Mo4 Mo4 67 099 8037 8 106 I 9 .795 83630 las?S tmA 9878 1 00789 0 01613) 831 1 Sti l M 996 1035 9 leb t t .331 lated 19776 tela tesA 186ES 8 016144 313 1 313 1 FI M7 10M 8 lE S 9 3M 1An8 3 775 90sA IS O llM7 0 0161$1 79616 796 18 73 99 1033 6 1807 6 8 407 1 4773 3675 ISA tela 12030 0 0161Sb 20028 200 30 75 98 4032 5 !!st $ 0. 437 13188 .3626 SE4 910 7 12750 0 016166 MS 37 265 H 77 to 10314 I 99 3 0 14 7 1983 Ilia l AS0$ 0 016173 25137 761 M 79 98 1030 2 I 07 0 50{i l.810$ I 9677 13021 19520 1833 194 8 .4799 0 016180 23821 238 27 81 97 le?t t ! .
10 0 M7 ! ?t38 1 9440 1943 198 0 $133 0 016188 2M 64 2M 85 83 97 10?? 9 8 19 0 $77 1 7856 1903 198A 808 8 6009 0 016196 21420 21428 M 97 1026 8 I l2 7 0 '411 1 7774 13366 186A tMI 16927 0 016704 20375 203 26 8797 10?$ 6 1 136 0 M6 1 7693 1 9339 tRA 172A 17891 0 016713 192 94 18295 88 96 1074 6 8 14 4 0 640 1 7613 1 9793 1223 174 8 ' 8901 0 016??l 18373 :8324 91 96 1073 3 I 15 3 0 ill M33 1 9247 tMA tM8 1 99S9 0 016??9 174 08 ;74 09 33 96 to?? ? l 16 1 0 74 9 5463 1 9207 IMS 130 0 2 l068 0 016230 16545 .6547 95 96 1021 0 3 170 0. 783 7374 19157 IRA 131 8 2 ??30 0 016747 ISP32 ,$7 33 97 96 1019 8 17 8 0 1817 1 7795 195t? 1983 132 0 2 M46 0 016? % let 64 i49 66 99 95 1018 7 18 6 0 1851 1 7217 8 9068 182A lH O 2 4717 0 016?65 142t0 101 M 1017 S 19 5 0 1484 .7140 13024 INA 1M9 2 6047 0 016274 13S M 1'4741 35 $7 803 96 1016 4 70 3 0 1910 .7063 18980 15A tW1 2 7438 0 016?84 179 09 179 11 LOS M 1015 7 .lil l 0 1951 4986 18S37 1RA let t 7 8892 0 016?93 177 98 73 00 :07 M 1814 0 !?20 01985 15910 143f95 Wet 442 8 3 0sil 0 016303 11721 172? 09 95 1012 9 177 8 02018 IliM 13867 9424 les O 3 1997 0 016311 Ill 74 11 76 11 M 1011 7 123 6 02Mi 1 6769 14810 letA lei 8 3 3663 0 0163?? 106 58 106 59 13 95 1010 5 .: !?4 5 0 2084 16u4 1 A 769 H6A las t 1 6381 0 0:4332 10468 101 70 IS M 100s ) !!M 3 02387 1 6610 18727 HLA tW e 3 7164 0 016M3 9705 97A7 117 95 1008 7 1176 1 02150 14636 1 8686 998 3 182 8 3 9065 0 016353 t? 66 02 68 189 9S 1007 0 1176 9 02183 14463 13H6 9823 tha t 410?$ 0 016363 48 60 88 62 !?i M 1005 8 1877 7 82216 14790 14606 SkA Int e 4 3068 0 0163)4 ta S6 8467 123 M 1004 6 1178 6 02748 ! & 318 13 %6 letA 198 8 4 5197 00163v 80 8? 0033 125 96 OG3 4 1179 4 02781 44245 1A526 letA let t 4 7814 0 016395 7777 7729 177 96 1002 2 .130 7 02313 14174 13487 le t 187 8 49722 0 016406 73 90 7397 179 96 1001 0 131 0 02MS 14103 18448 se2 t 164 8 6 7124 0 016417 70 70 70 77 83196 999 8 131 8 02377 16C32 13409 lesA tes t $ 4673 0 016478 4767 676a 13397 991 6 il3? 6 0 7409 1 5961 11378 40EA M80 6 7223 0 016440 64 78 64 80 13697 997 4 1133 4 02441 15892 1A333 letA 978 0 $ 99?6 0 016461 62 6: 6706 13797 996 7 1997 02473 I H22 . A?95 litA 177 8 62736 0 016463 $94 $945 139 98 99$ 0 1135 0 C MOS I S763 .3268 IF3 A 174 0 6 M Sf- 0 0164)4 $6 9S 56 97 141 98 993 8 3135 8 0 ?$37 1 % 84 . A??! 1944 976 9 65190 0 016806 S4 69 $4 68 14399 9924 1130 6 02%8 1 %I6 1 8164 1780 178 0 71640 0 016498 S? 3S 67 36 14599 991 4 18374 C MOG I S$48 12147 1788
- 19 mies su.ve es re: e L1
________ a
- I. .
.Appendik E Volume 111
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- Tekte 1. Setuteted Steem: TempePetv7e Tebl' e JContinued '
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2780 904 0 187 0 04479 I .776 16?% 308 8 312A M 953 0 81757 $4%6 $ 4747 Pl? l 9010 183 8 04H3 676 1 6?09 3964 l' 312 8 08 688 0 i11761 $ 1673 $lN9 706 3 8979 IN I 04$86 1 8576 16167 318 4 380 39 M3 0 01766 4 0961 4 9138 790 4 BM S 94 8?6 00!??0 4 6418 894 8 1885 7 0 4640 18477 ' 6116 328 0 318 8 100 NS 9 01774 4 4030 46MS 294 6 8916 Lik ? 0 4697 s l378 ; 6071 374 8 4 4708 F9s i 388 6 .l872 04745 l 1780 60?$
270 105 907
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354 8 2 9573 33?3 86?l 11644 05161 10$17 1 % 78 388 0 160 903 0 01816 7 800? 2 8184 3MS 5%4 lits ? 0 5712 I t04 1 % 37 354 5 169II) 0 0182) 2 M91 2 6873 MD8 fn1 lith 9 0670 10337 ' M95 364 0 3738 377 648 0 0it?6 264$1 ass 0 2 % 33 346 0 853 6 1196 7 0 5314 1 0?40 15 %4 878 9 RE 517 0 01831 24??9 3720 2 4462 349 3 446 3 18974 06365 1 0148 $613 87s I 30B 0 195 7M 0 01836 7 3170 23M3 353 6 , Mil 1198 0 C Ml6 50057 1 5413 308 4 att e 2M 7W 0 01842 2 2170 3579 300 2 7304 NG8 4198 7 O San 09%6 1 5437 304 0 2 5720 0 01847 t il?6 2 8311 30 7 8317 !!99 3 0 5616 0 9876 1 5392 N20 ??S $16 0 01053 f olk 20)69 M6S 833 4 8199 9 Sea t 3B5 8 236 193 0 01058 8 9291 06%7 09786 ISM 7 382 8 1 9477 3708 8'S 7 12004 0 5617 0 9696 1 $313 306 0
- 8 Pa? ?S9 0 010M [ 84u 1 8610 3761 8?S t 1701 0 4B44 PSt ?M 00l870 PHO 0 % 67 0 9607 167N 400 0 4003 270 600 1 7877 379 8 8?? 0 I?0l b O b717 0 9618 1 5134 4N 8 0 01875 6d 77 170W 3838 818 7 I?01 9 alla 782 894 0 01888 . 6f S? 1 6340 0 57H 0 9479 i ll95 dels 3881 834 ? 1707 4 0 58!6 0 934) 4183 MS 817 0 01887 :l WM 6167 412 0 R S656 391 5 810 7 17078 0 686( 0 9763 .l 5118 ele s 4NO 308 780 00 094 GN 8 3 4838 14997 3%9 806 ? I?OJ l 0 5915 0 9165 I $080 ANI 3?2 391 0 01900 1 4184 1 4374 4013 80? ? 1703 5 0 69W 0 9017 438 8 3M 863 6 01906 LIMt I SO42 434 8 1 3757 405 7 798 0 I?03 7 432 8 351 00 0 01913 130?H 1 37179 0 6014 0 8990 16004 470 0 406 8 3M 0J 410 1 793 9 I?04 0 0 6063 C 8903 4966 4370 0 01919 1 24887 176a06 4146 759 7 !?04 2 0 6117 0 8816 4978 43s 0 440 8 381 64 0 01976 119761 1 71687 419 0 78 S 4 17044 Het 397 M 0 0193; 184874 11683t 4236 0 6161 0 87?$ 14890 use 448 8 el4 0'l 761 1 1704 6 0 6710 08W) 1 415 444 8 0 01940 1 10?l? 11715? 4?lt 776 7 l?D47 0 6?% 0 0$57 l at it 482 0 43114 0 01947 1057M 10??ll 4325 all 498 8 448 7J 772 3 !?04 8 04331 0 8471 14 Mt 4378 C 019>4 1 01516 103412 4370 7678 !?04 8 O t hf 0 83f 5 14 1 464 8 E2 .
__ -___ - J
l
< 1 Volume 111 Appendix E l
~ Table 1. Saturated Slesm: Temperature Table-dentinued Atn Pitn $peptg yelpme lemp inthalpy fairopp lb Cre $dl $at $Al lehe 50 be isome l e.ep
$41 S,et hdt Itmp l' p e, Warm l ecme Ivap Wapr 1omd leap Wapor ishe eq _ _ et bl his hr is l 468 8 4W 87 0 01961 0 9746) 0 99474 big, J
- 48 48) % 44t h 1632 170a 8 0 640$ e s?tt 1870s au t 468 9 0 019th 0 93S86 0 9 % 57 446 8 h86 1708 7 504 83 0 01976 0 8988', 0 91862 460 7 0 6dM 0 8713 1467 4M 8 4739 9467 0 01984 C hMS 0 883? 7$4 0 1704 6 0 DD? 08127 I 46?t au 8 4188 S43 II du) 7493 170s l IL6%8 0 6047 1 8697 )
001997 08?tS$ 0 ktS0 4H9 744 6 !?Os 3 4724 401 8 0 ute 0 ft% 1 tub als e M616 0 07000 0 79716 0 81717 484 0 587 BI 445 FJ9 6 I?os t D ual 0 7871 l able 458 0 O 07009 0 PM 13 0 78D? 469 i FH 7 !?u3 4 40s t 610 10 0 0?017 0 7364i C 7%$8 0 6696 0 7785 ; &81 een 0 set t 633 03 4734 ??,7 1703 S t 04S 0 7700 kaa 896 8 08/0?6 0 70FW 0 728?0 478 5 724 6 !?03 8 40s e 4M 61 00?cM 0640t4 0 70100 0 6793 0 7614 4407 482 8 483 7 fill 1702 7 66Hf 0 7678 ' 4370 885 0 telt 480 % 0 07083 0$M48 $67492 80d 8 70$ 78 487 9 714 3 !?0? ? 0 6890 0 7443 0 020$3 06?938 0 64991 14333 008 8 tell 73140 49? ? 199 0 !?01 7 00?%? 060$30 06Mt? 0 6939 0 73S7 1 4 ?96 tes t
, lit t 767 77 0 0?077 0 68?18 0 60789 4976 703 7 1011 C H87 0 7778 47$8 808 8 lill 7M 76 507 3 696 7 700 6 0 7036 0 7185 0 02088 0 Mt97 0 50079 4??1 St7A
$071 6t? 7 iDE8 0 7085 6 7099 . 4153 tila 179 9 812 D 0 0?09) 0 63864 0 $$9%
pet $120 6870 1199 0 841 04 0 0710? 0 51814 0 63916 0 7133 0 7013 3 4146 Wes 878 9 87033 0 07117 089843 0 519 %
$169
$218 6813 ) 96 2 0 7187 06976 lilas $74A 832 0 900 M 6766 I $7 3 0 7731 0 6839 14070 M88
. 0 0?!?) 047t47 0 60070 676 8 Ml6 1 96 4 latl 931 17 0 0?lM 0 46173 0 48757 $317 0 7?80 0642 1 4037 432 0 M16 1.195 4 073?t 06ES 1 3993 83s a N8l 90 79 0 0?l46 C HM7 0 46513 toda 99$ f 7 SMS SS75 IlW 3 0 7378 0 6577 1 3954 tell 0 07157 047677 0444M Sal t 6Sl 3 1931 N00 it?8 49 0 0?!69 0 81088 0 81717 0 74?? 0 6489 I MIS tes t h20 104 69 546 9 6dS O 191 0 07476 0 6400 tiel 00?l8? 0 79879 04140 h70 638 5 190 3876 M86 097 D 00?lW 0 379 % 0 40160 632 0 , 1 01, 6 0 7Sh 0 6311 N37 957 8 H71 2 Ch4 C H77 3797 908 8 thi 1131 38 0 0??07 C MS07 0 38754 SM 8 l17010 $674 675 3 1187 7 0 0???1 0 33099 0 37320 M76 0 76?S 0 6137 1347 Het SM S 207 7? 0 0??M 9 33741 C M94 615 5 1196 1 D M14 0 604) 13716 064 8 871 8 246 76 0 0??49 0 374?9 0 M678
$7? S 631 5 IlW S 0 7775 0 6950 M4 888 8 mi 1785 74 l?$ 3 604 5 118? 7 0 7?M 0 $8$9 . MM 0 0??64 0 3116? O 33476 583 7 H77 3180 1 872 8 0 78?S 0 S7H L 3592 5 76 I 200 8 13?617 SM 0 R77 0 0???9 0 79937 03??l6 589 i Set t II?9 f 0 0?? 95 0 7876) C 31048 0 7876 0 % 73 13 %0 600 8 tell 410 0 M46 687 4 it?6 t 0 79?? 0 M80 3607 M18 0 0?311 02760t c ?t919 600 1 $74 7 ;174 8 M40 453 3 0 0?378 0 76899 0 28877 0 7978 0 548S 3464 688 8 DN O 18978 60$ 7 %6I . !?? 6 0 8030 0 6390 M?b M78 00?MS 02642$ 0 77770 611 4 $$8 8 a1707 0 8087 0 6793 , 33n set s r W4 15432 0 02 % 8 0J4344 SN,747 617 1 550 6 litti 1989 7 002M2 023374 8 25757 l167 7 08)M 0.5196 13330 NB8 W88 6373 6?? S 6472 185 1 0 8137 0 5097 3 3284 4bti 0 S 74 02 0 27394 9 24796 628 8 1336 1624 812A 486 I C 0?4?? 02144? SJM65 0 $?40 08997 1 3238 wit 6M 8 &?e 7 IS9 5 0 8?M 0 4896 3 also
$143 7359 8 07444 0J0516 02?t60 Sl2A k08 616 6 124 0 8348 04794 1 3141 818 8 983 1786 9 0074u clulb 0 1 7081 646 9 506 3 IIS37 De8 18390 C M03 0 4689 1 3097 038 8 0 02409 038737 0 21776 653 1 4966 1849 8 C Mb8 04583 1 3041 W4A 8?C 8 19974 00hl4 0 17850 0 70394 W2A IW70 ESts 486 7 Ile t 0 8514 0 4474 1 2983 WII 0 0?S39 0 17044 C itSU 66S S 4764 lla?J 0 8571 04%4 I F9M 886 4 2007 8 0 02 % 6 0 16??6 0 18782 Sits 6724 465 7 !!al3 0 M28 0 4251 1 2879 038 8 M89 Pont 0 0?$t5 0 6477 0 180?! 6791 454 6 1833 7
$44 8 2118 3 0 0?tM C #644 0 11269 CNE 0 4134 12871 NII 644 4 21781 685 9 443 4 !!?9 0 0 8746 040ll 1 7761 het 0 07657 0 $876 0165M 692 9 43) 1 1124 0 057 8 ??39 7 0 0?691 0 :3124 0 15816 0 8806 0 38t3 1 7699 ks0 GMt 700 0 418 7 1118 7 0 8868 03767 t h34 452 8 13017 0 07728 0 12387 0 15115 7074 40$7 1113 1 0 8931 0 837 12%7 GN O 858 8 JES7 0 0?768 01160 C 14431 714 9 397 1 1107 0 au t 243tl 0 0?81) 01847 01347 0 8995 0 3502 1 7498 08e 4 est 8 72 ? 9 3777 110C 6 0 9064 0 3361 1 74?S SM 8 2498 1 0 0?858 010??9 0 13087 8710 2%66 731 S M21 ,093 5 0 9137 0 3710 1 2347 Oss e 0 07911 0 09$14 0 12474 740 2 8457 478 9 26J6 8 0 0?970 0 08799 0 11769 085 9 378 5 ;;0776 Ol?!? 0 3054 1772 8728 749 ? 0 9287 0 7897 1.!!?9 4758 die s 27086 0 03037 0 00080 0 111!? M85 310 1 1068 5 0 9365 027?O 1 7086 8088 GM 8 27t? ! 0 03114 0 07349 0 10463 487 881 0 28574 290 2 1058 4 O kd7 C ?t37 l itM 88d 8 0 0370s 00659$ 0 M799 778 5 268 ? 1047 0 0 9$3$ 02337 1 1877 887 8 PlM S 0 03313 0 05797 0 M110 ut.0 N6s 30134 790 6 7431 1933 6 0 90M C2110 1 1744 012 O 00MH 0 04916 0 08371 804 4 212 8 1917 2 09749 01841 1 8591 Was let 9 30W 3 0DMU C 03857 0 Obit le28 3135 5
$?? a 17? 7 995 ? O 9901 01890 1 1390 Sus 8 0 03474 0 03173 0 06997 lute 144 7 9797 198 8 Sl?? ? 0 04100 0 02192 0 06300 1 0006 0 !?46 1 8?52 182 8 181 9 854 7 107 0 96( ? 0169 0 0876 1 8046 3198 3 0 04427 0 01304 0 05730 Ins t 30547* 8208 2 a73 0 614 9Me .
03?9 0 06?7 10t% lat e 0 05078 0 00000 00$078 906 0 00 906 0 ;
Elf 0 0000 1 0612 18147*
- Critical temperature E3
.. g ,
Appendix E Volume 111 Table 2: Saturated $leam: Pressure Table Aln he u 194 $st $nt $31 $30 $31 lldfl lp $31 A14 l'ea u 18h8 l 'O u' 0 leale Vdpo' 110uid leap Vapos lieu.0 leap Vepo' lli he in 0 1 *e % ', hi h ig hg 3, 6 tg 6, 6 S MI6S 3/ bl6 O bibu/. 330 '4 3 10/ 4 0 000* S4 107'. 4 0 0000 2 187.- c ini e ssent lib % U$ 0 0le !N t /.*t h ps',4 77 ta ' 10'e ID l 10874 t he h 586 fi DS4/ .' 04 3 . LM 7 0 73 e t:1471 t.a l b 8.816 476/t lose t 80% J O P 's I WJe 8H l8 301 74 r,0 H 11L Jtt% .431 63 64 74 2 0J 79 18 467 74 C DIMb1 7 ) W.
80J* 1 1105 8 0833 .
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0 016719 76 78/ 76 799 18017 970 1 11$0 S 0 3171 4447 7%s l6 t 71103 0 0167/6 te H6 l 7 6 ? 74 76 ?90 181/1 9t* ? 11 % 9 0 3137 dels 7%r ts e M8 ??!96 0 016834 70070 /0 087 196 77 hol 11% )
81 4 ?% M 0 3J68 I J96? 17J70 MI C 017009 13 7?M !) 74M 718 9 tab 7 llH I O M8? I 3313 1 69 % to 4 40 8 767 h 0 017161 10 4764 40 4965 73t i 9336 li69 8 ts t 78107 0 39?l I 7644 1 6765 de 4 0 017?74 88M7 4 5140 ?% ? 9/19 1874 ) 04)l? 17474 se g 79771 0 0:1383 71%?
16 bet Is t Me 30? 93 P l?M 76? ? 9tte 4677 6 04??J 12167 1647 Is t 0 017887 61876 6 ?OSO 27? 7 9078 1180 6 16Ja, Os t 317 04 0 4411 1 9905 le t 0 01757) S et M $ 471) 78. I 900 9 11831 O s)3a IIe 370 78 0 047669 4 8719 4 8963 /90 7 OH 6 1 867S 1 6706 at Illt ) 0 443 1 8470 1 6113 gas le t 3?7 87 0 017740 48133 4 4310 798 5 888 6 11877 0 4743 l l?Sd 16077 )m t lit t 334 79 0 01787 4 0306 4 0a84 305 8 84 s l 1808 9 0 48M l ill$ 6950 Itti tM 8 381 7) 0 01709 3 7097 3 7276 312 6 8:78 Iltc s C a919 10%0 6479 tMe 118 s 347 33 0 017 % 3444 J ataa 319 0 87/8 1191 7 0 a997, 1081$ 6413 tan t tes t IS3 0s 0 01803 3 ?olo J ?lto 330 P,0 0 tes t h8 43 1193 0 06078 1 0681 5767 640 0 0 01604 ? 9958 1 0139 3306 PM 1994 1 0 6ttl 10%4 l % 95 telt E3% 0 01816 7 81 % ?8JM 136 1 859 6 119$ 1 198 tig e he a? 0 4/J6 1 0436 8 % 41 les s 0 016?! 26S% e 67J8 3417 824 0 !!90 0 O SlH 1 03?? e7s g les e 373 0ll 8 % 91 0 01877 ? $1/9 1 6317 3*t, ? 8% 7 ll% 9 lm t 3;763 e lJ?& 3 0715 16sa3 tu e 0 01333 7 3847 74030 EO6 846 7 lit?6 0 53*4 1 0113 1 6898 19e 8 3as e 381 80 0 bis 19 7 /689 ? 7673 JM6 6:/ 8 1198 3 0 M38 1 0016 1645d fat I tie s 38t 96 0 01644 P l6173 P it?lt 359 9 839 8 1199 0 FMI 389 88 fs Sd90 0 9973 1 5413 lie s 0018$0 2 00779 7 08679 M47 816 d lit 96 0 %s0 098M 1 6374 IMO 33e 8 39370 0 018)$ 197991 99846 h5J 81 8 1700 1 pes t 0 % 86 0 9748 8 $3M 23e I 39139 0 0ilb0 199909 91769 )7? s 8/84 l '00 6 0 6t M 09%$ 1 5?99 794 8 J00 97 00186 348 8 8 'd'47 SdJIP 376 1 8 $0 U0i l 0 % 79 09%S i t?W M86 afal 4J 0 01870 DM8 77418 7te t 376 9 871 6 1201 6 0 477) 248 40780 0 01876 69137 fl013 0 950P $?30 Met ei107 6)lp 4434 Sit ) l 'Ol 9 O bibt 0 94J3 St97 2700 Pee 4 0 01881 65049 la ' 1 6A1 eld ?S I?C? 4 0 StOS t 93t I bit + Mtt 190 0 C 0188) $49? %88' J%t 8:l 0 I?CJ 6 0 % 44 0 9?91 SIE Mtl see t 4??h 00 696 1 $?344 1 M94 b4b SA e 170, g g ggi g,9?;)
8)s t 41173 I $101 3ec g 0 0191? 13ut t/ !. J?%4 4>6 74 ; l/0: 0 0 60 % 09936 54%f bel seg g a44 0'0 O f6914 1 1486? I l609s d '4 2 A4 I?04 6 0 6/17 0 86JO 1 4647 see s 460 0 4% ?# 0 01954 10l??4 1 03179 43? J 7676 1708 8 tes t 46701 0 6360 0 8370 l 4738 ele s 0 01975 0 90787 O t??6? adt $ 7% I 1204 7 0 6490 08tas t 4639 tess 900 8 476 64 0 01994 0 87383 044177 msg 460 9 743 3 !?04 3 OMil 0 79M i dS4 7 bea t tes t 446 70 89489 0 0?013 0 78 % ? 0 7H75 til 7 7370 I?C? ? 0 6723 0 7738 :446) ege t 0 0?037 0 68811 0 70s43 4819 7?C 9 17078 0 68?8 07%7 4Mi tes t 100 B $03 D8 C 0?0$0 06 DOS 0 6%% #91 6 710 ? I?fel 8 0 6928 0 7377 1430d Its e Me8 S30 64 0 0?o69 0 64880 0 60949 Soo 9 699 8 1700 7 80tl 0 70?? 0 7210 4?32 7te 8
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??tt . 0 0?669 0107 $5 SO M 100 % 150 % 20055 h0% 350 % d50 % %0 % 6% $$ 7$0 % 8 50 %
Iba9 4Si e 69586 ll?? ? OnM 921M 0 74St 0 ???0 0 ?950 0 3161 0 3646 030%7 0 4?31 04%I 0486) 0 51 %
s 0 8878 17676 I11?3
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7340 i 0 0?!/ 7 0 !$13 44J ;l 5:4 11 644 11 744 11 844 11 MM ib t 707 18 ll13 ; 010$ 0? 0?%f 0 ? ?P C 29b 0 33M 0 17,4 CacM 04344 C aio 0 89 6 00:3 IM4 13 H ; DM. iN 7 14W : 1$tes :W 3 It t S 17;? ) 17 b i s C89?9 !?M9 . 3 6t: IM. ! 4305 14L7 :ss t :S)P ,W IG7 I t 48 S 16v1 1734 S * $Uperhf8? I t Ftat;, h y IL 4 * $pecific volumf Cu 11 W l! S
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Appendix E Volume 111 Table 3. Supelhealed Slesm-Continued Ah hen lbl5e in $at 5at h"W 8'M
- b ?m l e***
(5pmp) h ale, Str nm 700 750 000 050 900 BSC 1000 1950 1100 1150 1700 1300 lett 1500 k 3709 8744 13749 18789 73709 78786 33789 M749 43789 48789 53789 63749 737 09 83789 3408 IM71D e 0 0??90 01408' 0 1874 0 ?!64 0 74?e C M48 0 ?8S0 0 3037 0 3?id 0 3387 Ob6 0 3703 CH% 0 41 % 04443 0 4774 n 768 95 1103 7 1191 6 th97 1310 8 lh?8 1391 ? 1476 % la60 9 14$3 7 ISM 6 1%70 IS881 1649 6 8710 8 17718 s 0 90)! 1740 1 3737 1 3808 14?lf 1899 84837 1 50 % I SJ37 1 %$3 15761 IS%9 16149 16509 16847 1 7167 k 3189 41 89 Ill 89 181 49 73189 78149 331 4 381 89 431Il es 89 $3109 63189 731 89 831 49 flee e 0 0?IS9 0 1307 0 868) 0 7037 0 7?93 0 7584 0 ? 71? 0 7096 03E8 0 3737 0 3390 0 3543 0 3697 O M80 04M9 C aS?9 1% 8 118 h 6
731 71 IM33 1876 7 172 6 13038 IMPd 13E 7 14?3 1 14S7 5 1890 7 li?? 9 4%46 IS8S 9 16478 17M P 17704 0 9139 I?MS 1 3075 1 3708 14179 1447? 147M 1 5029 ISM 9 1 5497 15703 1 6903 16064 1 64 % 1 6796 1 7136
> 76 09 76 M iM 09 176 09 FM 09 776 09 376 M 3409 4M 09 476 09 S?6 09 6?6 M FM 09 SM 09 3ese 1673 9H h e 0 07938 0 1711 O IS4s 0 1909 0 7478 0 7390 0 h85 07M 0 ?t33 0 3093 03247 03MS 0 1640 e Mit 0 4084 0 4350 74447 1087 0 1160 ? 1741 6 I?% S IMit 1387 1 1419 7 IdW I 1887 7 IS?0 7 1%?? IM37 1646 0 1707 7 IMI 8
09747 17??S 1 7908 I M97 l a04? I4MS 1 46 % 14964 I S?08 I SAM IM4 1 5848 4 60a0 1 640S 16746 1 7068
> 7047 7047 12047 17047 ??047 ??047 37047 37047 4?047 47047 $?047 6?047 77047 870 47 77es f679 $31 6 e 0 03079 0 1119 e ldil Cl?N 0 7058 0 ??4 0 7468 0?W4 0 /s09 0 7965 0 3114 03M9 0 3399 O M70 G M31 041N PS3 34 1069 7 8147 0 1733 1 1789 5 13M 3 13775 64Il7 8427 lek 6 til7 5 1689 8 IS81 S 144i 1796 I l?674 s 0 93 % I ?097 17777 1 3481 13%4 84319 8 428 14900 8 5148 3 5376 i SSSI I $794 8 6984 563% 3 M97 1 70 ?l k 15 04 65 04 IISM 165 04 715 04 MSN 3l$M MS O4 415 04 465 04 515 04 615 04 715 64 815 04 3e88 e 0 031M 01030 0 1778 0 168) 0 1957 0 7168 0 73S8 0 n31 0 7693 07pb 0 ?t91 0 3137 0 3?68 0 3531 0 3785 0 4030 f684 %) 6 77069 10 % 8 1871 7 1770 6 1787 7 1330 7 1377 8 14il 7 14477 1881 6 lild 8 16473 IS79 3 1647 7 lied S IM6$
s 0 W68 81968 !?$77 13M8 83E7 44745 14%1 3 4838 1 5009 i S3?l i S$37 15747 3 6938 8 6306 iMSI I 6975 k 9 78 $9 78 109 78 IS9 78 709 78 hl 78 30 78 359 is 409 78 459 78 509 78 409 78 709 78 009 78 feel e 0 0326? 0 0942 0 1138 0 1541 03853 07#8 0 77 % 0 74?? O n85 07734 07877 0 3014 03147 0 3403 O M49 0 3887
- 1690in n 78S13 1039 8 10 % 3 17096 I??47 1314 9 IM80 14077 1443 7 le78 5 Ill? 8 IM49 IS77 0 16404 1703 0 iMS?
s 0 9584 11803 11783 1381 3 3780 8 4 lit i 4494 14777 i SO'? I S?H I bet $ I M9? IStal !6M9 I MOS I 6931
> 4 67 S4 67 10467 15457 70467 h467 304 67 3S467 40467 4S4 67 S0467 60487 704 67 80c 67 3e88 e6% 33; e e 0 0H78 0 08SO 0 048:- 0 8463 0159 0194 0 7161 0 73?9 0 7484 0 7630 0 7770 0 793z 0 3033 03787 0 3577 0 375) 80184 1070 3 106M i!U 9 I?67 0 1319 0 IE)? IdolI 1440 ? 1444 1504 4 IW24 IS 74 8 16M S 1701 4 1436 s C Ul8 l l619 Il9M 1 31)I I J697 140S7 14479 18?ll 1494 I S?!) 1 6434 ISW? 15841 16714 16M1 1 6488
> 49 ?? 6S ?? 144 7? 194?? 745 7? 29 ?? MS ?? 399 77 449 ?? dS% 77 $95 ?? 69% 77 795 ??
3108 e 0 03tli 0 074% DIM 9 01478 0 1867 0 ?071 07737 0?%D 07633 0 7670 0 7830 0 79?? 0 3170 O M03 C M?8 470078. > 8997 993 3 11854 I?% I 1313 0 I)M e 1399 0 1436 7 447? 3 IS06 6 1639 9 IS776 16M 7 16998 1475 s 0 9914 1 1373 I J007 I M04 44014 leM4 146S8 18970 1 5161 1 5384 ISM 4 15794 1 6169 16518 1 6847 )
k 44 4/ M 9? 84497 Iw t? /44 57 /W W M4 9? pa 97 44a 97 494 97 %M 97 06497 ?W 97 8398 e 0 04477 00%6 4705 081 h 4454 9336 01300 0 1588 O l804 01987 0 7151 0 7J0) 07447 0 ?S76 0 7704 02877 0 3065 03791 0 3510 s I O3Si 1 0837 1177 3 iM09 1306 9 IM3a 13N 9 1433 l 149? IS03 8 15374 1570 3 16M 8 4698 3 17637 1 7877 I Mit 1 3953 14300 1 4600 14E6 I S!!0 15335 13547 15749 1 6176 l H77 1 88 %
3 ass e h a 1713 0 1510 0 1777 0 1908 0 7070 0 1718 0 7357 0 7488 0 2613 0 77M 0 79E O 3187 0 M00 s 1158 7 17471 1300 7 IM84 1390 7 1479 5 14661 15010 IM49 1%88 16329 1696 7 IM99 I ?7d? I 34M I 3879 8 4?J7 1 4547 1 4813 1 5059 i S?87 iSW1 8 5704 1 GON iMM I 5767 3det e m 0 1879 014h 0 1653 0 1834 0 1994 0 7140 0 ???6 0 7405 0 M78 0 7646 0 78?? O 3088 0 3296 s 11437 8733 7 17943 IM34 13 % 4 147$ 9 14679 1498 3 45374 1%S8 j63t l 169S 8 1585 I 7600 13JM i 3807 I el74 1 4486 44761 4 S010 1 5740 1 S8% 1 M60 a 604l 8 63 % 1 6778 k ,
het e 6 0 1048 0IM4 0 1583 0 1764 0 192? O 7066 0 ??00 0 73?6 0 7447 0 M63 0 77H 0 7995 0 3198 s Il??! I??d t I?878 1338 7 138? ? 14?? ? 14S9 7 1495 5 1S79 9 1%36 16797 1693 6 17577
! ?4$0 1 3147 1 3734 Itil? I 4430 14 eOS I 4967 I $194 1 Sal? I %I8 3 600? I 63S4 4 6691 Met e 6
COW D l?9( 0 l$17 0HH 0 1854 0199C 0 7876 07?S? 0 ?3 71 0 748; 0 7702 0 7906 0 3 lot 6 18081 !?lS 3 !?II ? 1333 0 1377 9 1818 6 14S( S 1492 6 !$27 4 1%I3 16?? 3 169? 0 IFM 9 '
1 ??81 1 3148 i M6? I 40 % i 4J 74 1 86S6 14914 i $149 I 5369 i M76 i $96? 4 63?C I M54 4 k
sett e 6 0 0799 0 1169 0 139s 0 1% 74 0 17?* C 164 Osb6 0 7886 0 ??31 0 7340 0 ?S49 0 2746 0 ?9M
- 10t.4 7 litSS IM76 13?? 4 1369 1 84!I7 14 % I 14870 IS?? 4 15 % 8 16?36 1688 9 lb3?
1 1888 8 79 % 4 J$17 iW8 14?6S 1d48 1 447) 4 506) 1 5784 1 54 % ! $886 6 6/47 1 6584 4eee e, h 0 0631 010$? O 178: 0 u63 0164 0iM? 0 1877 0 1998 0 7tOS 0 2?lo O ?411 0 7608 0 7783 I
s 10074 1174 3 sM34 1311 6 13601 1403 6 14436 1481 3 IS17 3 IM?? 1619 8 1685 7 172 6 l 13% 177W I 337) i 3807 I alS8 3 446I I 47JC 1894 i S?03 i 5417 1 5817 1 6177 1 6514 in 4794 o 0 0a9f 0 0'*4) 6 0 1183 08M/ O ISI) 0164' 0345 0 1883 0 199; 0 7093 0 ??87 0 ?470 0 7645 s 9501 11516 I?J8 6 130r 4 1351; ! >% 0 14371 144 $ ISI? ? 15476 1616 1 166? 6 1748 0 1 0905 17See 13m !w Iah8 I4h8 1 464) ! 4693 I bl?4 1 6343 1 574? I 6i09 I 64S?
4att e 0 04 'l a 00s4( I??33 096 1t*7) 0 1040 l'f CiNo 5 0 13470 014 T CI5?
136 A01171 141040lat9 1787 7 DIM"I isi? 0198r 0 7174 0?hl 0 ?$19 10%t 8 ?J?S I 3074 ! JStt i 3* I 4' O I 46 % i es t? I S346 1 IS4 $?66 I M 7J l 60(4 56179 3 0 161? 3 16'9 8 176 3
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4848 e 0018a 0 0151 0 100s e ileb C 13h 0 labs 0 Isa
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8008 e 0 03 % 0 0666 0 0917 0 1109 OIMP 0 1385 O lS0( 0 160( 0170r, C lf 06 0 1090 01977 0 714? 0 ??%
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0 0338 0 0F91 0083 0 1038 0 ll8S C 111? 014't 0 IS?9 0 16?6 0 1768 0 1806 0 1990 0 7050 0 ??01 854 9 Ik?9 1873 6 iM?l 1313 % IM46 1410 7 14S? ) 8491 6 IS?ll i%SS 1600 1 6670 0 17374 6
1 0070 l Ill) I 76 t? I 3707 i M45 1 4001 1 4304 6458? I 4431 8 6061 1 5777 i $48) i M63 I 6716 b
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0 9985 l 1370 t?OS I 3088 1 3546 1 3914 14??'l I 4509 84767 6 49 % i S?id i $4?0 I $806 1 6161 D
leal e 0 0317 0 M83 0 0728 0 09l? 0 10S8 0 118/ 0 I?9? 0IH? O 1886 0 IS7: 0IW 0 IPM 0 1884 0 7013 h
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