ML20206U337
| ML20206U337 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/24/1986 |
| From: | Munro J, Rogers T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20206U325 | List: |
| References | |
| 50-424-OL-86-01, 50-424-OL-86-1, NUDOCS 8610070447 | |
| Download: ML20206U337 (189) | |
Text
{{#Wiki_filter:r ENCLOSURE 1 EXAMINATION REPORT 424/0L-86-01 Facility Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Facility Name: A. W. Vogtle Facility Docket No.: 50-424 Written, oral, and simulator examinations were administered at the Vogtle Nuclear Plant near Waynesboro, Georgia. Chief Examiner: w,a, voze f ~29 8(, Th6 mas Rogers g Date Signed Appi-oved by: df John F. Munro, Acting Section Chief Date Signed Summary: Examinations on June 23 and July 28 - August 8, 1986. Oral examinations were administered to 29 candidates; 28 of whom passed. Simulator examinations were administered to 29 candidates; 24 of whom passed. Written examinations were administered to 31 candidates; 18 of whom passed. Based on the results described above, 2 of 11 R0's passed and 11 of 20 SR0's passed. h V D
REPORT DETAILS .
- 1. Facility Employees Contacted:
- Paul Rushton, Training Supervisor
- Attended Exit Meeting
- 2. Examiners:
Casto, C. Isaksen, P. Lawyer, L. Munro, J. Picker, R.
- Rogers, T.
Vinnola, T.
- Chief Examiner
- 3. E.< amination Review Meeting At the conclusion of the written examinations, the examiners provided Mr. Paul Rushton with a copy of the written examination and answer key for review. The connents made by the facility reviewers are included as Enclosure 3 to this report, and the NRC Resolutions to these comments are listed below. The parenthetical question numbers refer to an identical question on the SR0 examination.
- a. R0 Exam (1) Question 1.06(c) (5.18(c))
NRC Resolution: Comment is accepted. One portion of the answer relates to the coolant temperature (" closer to saturation" and "0NB more likely") and the other to the flow rate (" scrubbing bubbles"). These concepts are not redundant nor is either of them unnecessary. However, either portion of the answer will be accepted for full credit due to question's lack of specificity. No change is required. (2) Question 1.11(5.21) NRC Resolution: i The lack of specificity in part a. may have confused the candidate. Part a. has been deleted.
2 (3) Question 1.12 (5.22) NRC P,esolution: Comment is accepted. Full credit has been given for "MTC is most negative." (4) Question 1.14 NRC Resolution: Comment is not accepted. Some of the recommended additional answers are contained, in part, within the answer key and will be graded accordingly. The question specifically addresses using parameters from the E0Ps and will be required for full credit where applicable. No change is required. (5) Question 1.15 NRC Resolution: Comment is accepted. As long as the candidate addresses the fact that beta decreases as Pu 239 builds in, full credit will be given. (6) Question 1.16 NRC Resolution: Comment is accepted. It is agreed that within the context of this question " greater than 6 minutes," " indefinitely," and "never" are equivalent. It is also true that the answers on the answer key were not properly oriented. The answer key has been changed accordingly. (7) Question 1.19(5.01) NRC Resolution: Agreed. This is indicated on the answer sheet point breakdown. The referenced figure 947 contains additional information not required for full credit. No change is required. (8) Question 1.21 NRC Resolution: Agreed. A figure essentially identical to that submitted has been added to the answer sheet as an attachment.
r 3 (9) Question 1.23 NRC Resolution: Coment is accepted. The typographical error has been corrected. (10) Question 2.03 (6.05) NRC Resolution: Coment is not accepted. As per facility comment, the answer key is technically correct. (11) Question 2.09 NRC Resolution: Comment is accepted based on new material provided by the facility. Note that a similar Vogtle question, 123 NA, does not concur with this response. (12) Question 2.14 NRC Resolution: The indicated point values on the answer sheet were in error. The answer sheet will be changed from (0.5) to (0.8) for the first part and (0.5) to (0.7) for the second part. (13) Question 2.16 NRC Resolution: Terminology is equivalent. No change is required. (14) Question 2.17(6.19) NRC Resolution: Coment is accepted. The additional recomended answers will be accepted for the appropriate credit. The SR0 exam will retain its original point value.
4 (15) Question 2.18 NRC Resolution: No reference for the "comon understanding" was given. Material which tests the candidate's understanding of the design and functions of reactivity control mechanisms and instrumentation is emphasized per NUREG 1021, ES 202, and 10 CFR 55.21(b). It should be noted that the first reference given in the answer key was to the Vogtle lesson plan which indicates the material was taught to the R0 candidates in pursuance of Vogtle learning objective 132/5. No change is required. (16) Question 2.20 NRC Resolution: Answer key item 1, " manual 1/2" adequately covers the recommended additional answers. "Any" is not the same as "one out of "XX." Partial credit will be allocated if the correct trip is listed even if the logic is incorrect. (17) Question 2.21(6.22) NRC Resolution: Terminology is equivalent for both or 2/2. No change is required. The point value has been evenly distributed for the four responses. On the SI, the candidate needs to distinguish that the pump start is train-related (1/1). (18) Question 2.23 NRC Resolution: The Gross Failed Fuel Detector is not an additional load on the ACCW system. The GFFD cooler, as shown on the referenced drawing, is a load. However, this is not an additional acceptable answer since (1) it is a cooler, (2) it is listed in the referenced training material as #5 under "NSSS Sample Coolers." (See R0-LP-087, page 4). No change is required.
r 5 (19) Question 2.24 NRC Resolution: Comnent is accepted. The four additional permissives will be added to the answer key. Vogtle question bank 142 NA implies that lesson SR-LP-058.6, objective 3, considers only the three answer key items as permissives. The question bank and lesson materials should be corrected accordingly. (20) Question 2.25 (6.24) NRC Resolution: Any five different individual or group loads will be accepted as long as they are not redundant. (21) Question 2.26 NRC Resolution: Comment is accepted. The answer key has been changed accordingly. (22) Question 2.29 NRC Resolution: The answers are equivalent. No change is required. (23) Question 3.12(b)(6.13(b)) NRC Resolution: The terminology is equivalent. No change is required. (24) 3.16(a)(6.15(a)) NRC Resolution: It is placing the hand-switch in " ARM" that arms the system, not the temperature signal. No change is required. (25) Question 3.19 NRC Resolution: Comment is accepted. However, the answer must indicate RCS depressurization due to a failure of the PORV and isolation of the PORV by the block valve for full credit.
6 l (26) Question 3.21 (6.18) NRC Resolution: Terminology is equivalent. No change is required. (27) Question 3.22(6.20) NRC Resolution: Coment is accepted. Question will be graded as recommended. (28) Question 3.23 NRC Resolution: Coment is not accepted. Candidate needs to address that phase A is not train specific. The question clearly asks for coincidence. Partial credit will be given for the actuation signal if the logic is incorrect. (29) Question 3.27(b) (6.29(b)) NRC Resolution: Coment is accepted. Either representation will be accepted for full credit. (30) Question 4.02(7.06) NRC Resolution: Comment is not accepted. It is not true that AFW is throttled whenever flow exceeds 550 gpm - but only for the specific case cited by the facility. (31) Question 4.11 (7.15) NRC Resolution: Coment is not accepted. The question did not ask for the design values. (32) Question 4.12 (7.16) NRC Resolution: Comment is not accepted as requested. The initial load is equiva-lent to a new core cycle and will be accepted as one correct response. No change is required.
r 7 (33) Question 4.14 (7.18) , NRC Resolution: Answers are equivalent. No change is required. (34) Question 4.18 NRC Resolution: Comment is accepted. The facility comment is considered equiva-lent to the cited answer. (35) Question 4.20 (7.21) NRC Resolution: Comment is not accepted. If redundancy between headings and substeps under that heading exists, they will not be considered as separate items. (36) Question 4.22(7.03) NRC Resolution: Comment is accepted. The additional answer will be credited as one of the three required. (37) Question 4.23 NRC Resolution: Comment is accepted. Correct alternate answers will be credited accordingly. (38) Question 4.25 NRC Resolution: Comment is not accepted. Though VEGP has elected not to identify immediate operator actions by the procedure, there are imnediate operator actions that must be performed. The answer key was developed from the criteria used to evaluate reactor operator candidates on the simulator by the VEGP staff as stated in R0-SE-49. (39) Question 4.27 NRC Resolution: Comment is accepted. However, the candidate will need to address all three criteria for full credit.
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(40) Question 4.28 (7.25) NRC Resolution: Comment is accepted in part. If a candidate chooses to specify entry conditions, credit will be awarded. Credit will not be awarded for "when transitioned from another procedure."
- b. SR0 Exam (1) Question 5.10 NRC Resolution:
Phases not matched is the only acceptable answer, unless the candidate states that the indicator is at the 12 o' clock position. No change is required. (2) Question 5.11 NRC Resolution: Comment is accepted. In this context the terminology is equivalent. No change is required. (3) Question 5.16 NRC Resolution: The central answer will be changed as suggested. However, the tolerance is adequate as stated. Note that both steam tables and Mollier diagrams were provided to all candidates. (4) Question 5.17 NRC Resolution: NUREG 1122, "Knowledges and Abilities Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors," will eventually be guided by NUREG 1121, " Examiners Handbook for Developing Operator Licensing Examinations." The latter document is currently in draft form and therefore is not available to lend insight into the significance of a particular numerical importance rating. For this reason we have not addressed this aspect of the comment. The subject matter of this question is addressed by the facility training objective NL-LP-138/5 and by the facilities proposed question bank question 507N. No change is required.
9 (5) Question 6.07 NRC Resolution: Comment is accepted. Additional answers will be accepted in accordance with the additional reference material. (E) Question 6.10 NRC Resolution: Comment is not accepted. An inverter is a different component and is upstream of the bus. The bus does not fail, the power supply feeding the bus is the failed component. (7) Question 6.21 NRC Resolution: Comment is accepted. The point distribution for the answer will be awarded as recommended. (8) Question 6.25 NRC Resolution: Comment is accepted. As recommended, item 3 of the answer key "IR" will be changed to " Extended Range" and item 8 will be changed to "T ". The training material referenced on the answer h key should be corrected to reflect this change. (9) Question 7.01 NRC Resolution: Comment is not accepted. The intent of the question is quite clear. The question will not be deleted. A tolerance of -2 volts will be used. (10) Question 7.02 NRC Resolution: a Comment is accepted in part. The question is not double jeopardy since it has no relationship to the response for question 7.01. Furthermore, an unusually high point value was not given for a candidate's knowledge of the 125 volt IE battery procedure since each question was only given a value of 0.5. Credit will be awarded, however, for an answer greater than 128 volts and within the design of the charger. l
10 (11) Question 7.17 NRC Resolution: Comment is accepted. Additional recommended answer will be accepted. (12) Question 7.23 NRC Resolution: l Comment is accepted. Question has been deleted from the exam. (13) Question 8.02 NRC Resolution: Choice d. is incorrect. With QPTR exceeding 1.02, operation may not continue indefinitely above 50 percent power. See page 2-10, Section 3.2.4.a.3 of Technical Specifications. No change is required. (14) Question 8.03(b) NRC Resolution: Comment is not accepted. However, since a candidate could assume OSC was manned while answering part b, credit will be awarded for the " replacement OSC manager" as an alternate response. (15) Question 8.06 NRC Resolution: Comment is accepted. Based on the statement in the question that all systems required for S/V are operable, then MODE 3 should be the applicable mode. If the candidate assumes M0DE 3 is applic-able, then the answer would be "no classification required and boron concentration should be restored." If the candidate assumes MODE 2 and says to shutdown and declare a NUE due to being less than 551*F, full credit is given; if he states MODE 2 and shutdown due to the RWST is out of spec as the only reason, some credit will be taken off for not stating that the RCS T/S was the over-riding concern. (16) Question 8.10 NRC Resolution: The subject matter of the question is required knowledge for an SRO. An operator with understanding of system operations should know the actions to be taken in these circumstances.
11 Systematic actions will be accepted in any order. However, notification of the shift foreman must be last. (17) Question 8.11 NRC Resolution: Coment is accepted. The answer key for part (a) will still require the candidate to state that the equipment is INOPERABLE. If the exception for inadvertent omission is included -in part (a), it will be considered additional correct information. It will also be considered a substitute answer for part (b) as well as the facility's recommended additional answer. (18) Question 8.12 NRC Resolution: Comment is accepted. An example of each of the five will be-acceptable. More than one example of each answer would not be
" equally meaningful" and will not be accepted. No change is required.
(19) Question 8.15 NRC Resolution: Comment is accepted. The answer key has been changed accordingly. (20) Question 8.30 NRC Resolution: Coment is accepted. Credit will be awarded if a candidate is aware of conflicting VEGP guidance and states the issue.
- 4. Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the findings of the examination.
There were generic weaknesses noted during the oral examination. The areas of below normal performance were in electrical theory and the pressurizer PORV setpoint controller. Facility concerns regarding qualifications of written examination proctors have no foundation in that all proctors utilized were qualified in accord-ance with existing regulatory guidance.
i 12 The cooperation given to the examiners was also noted and appreciated. The licensee did not identify as proprietary any of the material provided to or reviewed by the examiners. f 1 i i 4 I
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M s U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: V0GTLE 1 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 86/06/23 EXAMINER: TOM ROGERS APPLICANT: _________________________ INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only. Stcple question sheet on top of the answer sheets. Points for each quastion are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 2nfar5-- ------
4H>v99 25 0 ________ ___1_0 _ ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 30 0 25 ___1_0___ ___1_00 _ ___________ ________ 6. PLANT SYSTEMS DESIGNr CONTROL, r AND INSTRUMENTATION ggg 4H>v4Hk= 00 ________ _'5_'_1__ ___________ ________ 7. PROCEDURES - NORMAur ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 30 0 25 ADMINISTRATIVE PROCEDURES, ___1_0___ ___1_00 _ ___________ ________ 8. CONDITIONS, AND LIMITATIONS
/0f.0 TOTALS +4iMPv99"_ _ _10 0 . 0 0 FINAL GRADE _________________%
All work done on this examination is my own. I have neither % siven not received aid. EPPL5CEU5I5~55GU5TURE~~~~~~~~~~~~~~
g6 ES-201-2 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following 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 only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write only one i,Tde of the paper, and write "Last Page" on the Tast answer sheet.
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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.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
Examiner Standards 12 of 18' k .
1 [- ES-201-2
- 18. 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 a 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 thi examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
/ Examiner Standards 13 of 18 . -- ^
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2 QUESTION 5.01 (2.00)
On a T-S diagram, _ plot a characteristic Rankine cycle for a steam cycle containing an MSR, turbine with HP and LP sections, feedpump, condensate pump, feedheater, steam generator and condenser. Label each segment of the cycle showing which plant component is responsible for that portion of the cycle. QUESTION 5.02 (1.00) s, A general rule is often stated " doubling the count rate halves the margin
/ to criticality". This is mathematically stated by the equation:
CR1/CR2 = l-Keff/ / 1-Keff/I 95 Which one of the following_ statements is correct concerning the above statement and equation?
- a. Both Keffl and Keff2 have to be less than 1.0.
- b. Equal changes in Keff result in equal changes in suberitical multiplication level.
- c. The equation only approximates the instantaneous change in count ratei once equilibrium value is reached, the count rate will be higher.
- d. A second doubling of the count rate will result in the reactor becoming critical or supercritical.
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 QUESTION 5.03 (1.00)
OP-12003-1, " Reactor Startup', requires that the critical rod position be taken at 1.0E-8 amps on the intermediate range. If, during a xenon free reactor startup at MOL, the operator " overshot' 1.0E-8 amps and instead leveled off at 1.0E-7 amps, which of the following statements is correct?
- o. At 1.0E-7 amps, there are little or no effects from nuclear heat but since the reactor is a decade higher in power, the critical rod position will be higher.
- b. At 1.0E-7 amps, there are little or no effects from nuclear heat; therefore, the critical rod position will be the same as at 1.0E-8 amps.
- c. At 1.0E-7 amps, there are substantial effects from nuclear heat; therefore, the critical rod positions will be higher than at 2.0E-8 amps.
- d. At 1.0E-7 amps, nuclear heat, xenon and the decade higher in power level will result in a higher critical rod position.
QUESTION 5.04 (1.00) The reactor trips from full power, equilibrium xenon conditions. Six hours later the reactor is brought critical at 1.0E-B amps on the intermediate range. If power level is maintained at 1.0E-8 amps, which one of the following statements is correct concerning control rod motion for the next two hours?
- a. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate.
- b. Rods will have to be inserted due to xenon decay.
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon burnout,
- d. Rods will approximately remain as is as the xenon establishes its equilibrium value for this power level.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE rwrts)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 J
QUESTION 5.05 (1.00) Figure 953 shows a graph of pump laws i.e., the relationship between pump speed and other pump parameters. Which one of the following parameters is represented by curve 2?
- a. Power -
- b. Flow rate
- c. Voltage
- d. Discharge pressure QUESTION 5.06 (1.00)
One of the characteristics of water is that it will hold gases dissolved in colution. Which of the following will increase the concentration of dissolved gases in a quantity of water?
- a. Increasing the pressure and/or lowering the temperature.
- b. Decreasing the pressure and/or lowering the temperature.
- c. IncreasinS the pressure and/or raising the temperature.
- d. Decreasing the pressure and/or raising the temperature.
QUESTION 5.07 (1.00)
- a. What is the purpose of the steamline pressure input into steam flow determination?
- b. If the steamline pressure transmitter indicated erroneously high (relative to actual steamline pressure), would the indicated steam flow be higher or lower than actual?
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5 QUESTION 5 08 (1.00)
Fill in the blanks (increase, decrease, remain the same):
"To maintain steam header pressure constant, it is necessary for S/G steam pressure to ______ from no load to full load. This is in response to the laws of fluid dynamics which state that as the flow of steam increases in a i
steam line, the pressure drop experienced per unit line length ______.' QUESTION 5.09 (2.00) - The technical specification minimum temperature for criticality is required to ensure four conditions about the reactor systems. State these four conditions. QUESTION 5.10 (1.00) A synchroscope moving slowly in the slow direction (counterclockwise) indicates a certain relationship between two parameters of the machine and the bus being paralleled. List the two parameters from amons this list-frequency, resistance, phase, voltase and current. For each of the two parameters state the machine's relationship (higher, lower, etc.) to the bus for the situation presented above. QUESTION 5.11 (1.00) State how the following vital instrumentation will be affected by a high containment temperature. (es. state higher than normal, no change, etc.)
- a. Pressurizer level
- b. Steam senerator level QUESTION 5.12 (1.00)
Brittle fracture is of particular concern because it can occur at stresses well below the yield strength of the n.ater ial, provided three conditions at e present. State two of these three conditions. (r*Kr* CATECORY 05 CONTINUED ON NEXT PAGE *****) 1
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- 5. THEORY OF NUCLEAR POWER PLANT DPERATION, FLUIDS, AND PAGE 6 i
f QUESTION 5.13 (1.50) '7 q" Power peak _ing is not a directly observable quantity, therefore, limits have been established on the basis of axial flux difference produced by the power peaking. In addition to maintaining axial flux difference within prescribed limits, what three limitations on control rod movement ensure that hot channel factors are maintained? j OUESTION 5.14 (2.00) Answer each of the following TRUE or FALSE concerning the small break LOCA cnalyses and reactor coolant pump trip criteria *
- a. The worst case SBLOCA was found to be a cold les break in RCP discharge piping.
- b. The analyses assume that the RC pumps remain operative for some time after the break and then are lost by some means (loss of offsite power, equipment failure, etc).
- c. If the RC pumps remain operating throughout the blowdown and reflood phase, the peak clad temperature criterion (PCT) would not be exceeded.
- d. With the RCPs lost and only one HPI pump operating, the PCT criterion would be exceeded; with both HPI pumps operating, it would not be exceeded.
QUESTION 5.15 (1.00) What effects do the following have on overall plant efficiency (qualitative only), assuming reactor power level is held constant at 00%. Consider each condition separately.
- a. Reducing RCS average temperature
- b. Removing a feedwater heater from service
- c. Increasing condensete subcooling
- d. Placing a moisture separator reheater in service
- e. Increasing condenser pressure
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7 ',
s e-QUESTION 5.16 (1.50) Steam exiting the steam senerator has an enthalpy of 1190 BTU /lbm and a temperature of 540 desF. Find the followinst
- a. Quality _
- b. Entropy
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- c. Superheat '
QUESTION 5.17 (1.00) What is the purpose of a " blowout coil" ina typical high voltage air-magnetic breaker? QUESTION 5.18 (2.00) How does each of the followins parameter changes affect the DNDR (increase-decrease-remains the same) and why (be brief)? Do not consider the transient effects.
- a. Average primary coolant temperature increases 5 desF.
- b. Pressurizer temperature increases 5 desF.
- c. Mass flow rate in'the core decreases 10%.
- d. Mass flow rate of steam to turbine decreases 10%. Assume rods in auto.
QUESTION 5.19 (1.00) , When starting up a centrifugal pump with the discharge valve shut, how would you characterire each of the following parameters compared to its value when starting the pump with the discharge valve fully open?
- a. motor current f b. discharse pressure I
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 8 w
c; OUESTION 5.20 ( .50) '!: The followins is a description of an important fission product. "Although neglisible amounts of it are produced directly by fission, 1.1% of all fissions result in the production of its two precursors in its decay chain. It has a half life of more than 1.0E+16 years and therefore is considered < stable. Its reactivity effect in the Vogtle reactor is independent of flux once it has reached equilibrium'. Name the element. The isotope is not required. QUESTION 5.21 (1.50) An estimated critical boron concentration of 975 ppm was calculated using the information on Data Sheet 1 (figure 957 attached). Assume the reactor achieved criticality at 153 steps on Dank D. RC lave was 561 desF. How would this critical rod position change (Bank D position relative to the previous criticality) if each of the followins conditions was different from that on Data Sheet 1? Assume no boron change and consider each condition separately. Line 10. Lamt < nc r 1, m l- a c : 5^"" #~ %N 1.
- b. Line 2a. Time shutdown was 72 hours.
- c. Average RC temperature was 532 desF.
QUESTION 5.22 (1.00) Shutdown marsin requirements vary throughout core life as a function of several factors. The most restrictive condition occurs at EOL. State the factor and why it is most restrictive at EOL. QUESTION 5.23 (1.00) The startup, intermediate and power ranse channels all use baron in their respective detectors (DF3 or boron lined). What is the reason for use of boron? OUESTION 5.24 (1.00) Why is there an opper and lower limit on the volume in the Nacl! Spray Additive Systes tank to neintcin e limited pH bend? (***** CATECORY 05 CON 11NUED ON NrXT PAGE rrrwx)
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- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 9 I'
e GUESTION 5.25 (1.00) ' You are plotting an inverse multiplication plot (1/m) during fuel loading. When fuel is loaded so that the distance between the detector and the fuel steadily decreases,.the 1/m plot will have a characteristic seneral shape. Sketch the seneral form of this curve (indicate shape, slope and rates of change on your sketch). Y (***** END OF CATEGORY 05 *****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 10 N
QUESTION 6 01 (1.00) e y Which of the following is NOT a diesel generator output breaker closure ; permissive?
- a. No 4160 ESF Bus faults,
- b. SI signal present.
- c. 4160 ESF Bus undervoltage. h J
- d. Diesel at rated speed and voltage. @
QUESTION 6.02 (1.00) The shunt trip coils for reactor trip breakers
- a. Do not receive a trip signal when a manual reactor trip is actuated.
- b. Cannot be trip tested separately from the undervoltage coils.
- c. Receive a signal to trip from an auto reactor trip signal through the auto shunt trip coil circuit.
- d. Are deenergized to trip, thus meeting the single failure criteria.
QUESTION 6.03 (1.00) The plant is operating at 100% power when an electrician inspecting a downstream RCP breaker (the class 1E breaker) accidently causes the underfrequency relay for the breaker to actuate its protective function. Which of the following best describes the plant response.
- a. The affected RCP trips. The operator performs 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 effected loop.
- c. The affected RCP trips, reactor trips and the other 3 RCPs trip.
- d. The affected RCP trips, reactor trips, the other 3 RCPs trip and the plant sis on low low flow.
(rrr** CATEGORY 06 CON 11NUCD ON NEXT PAGE r****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 11 ;t QUESTION. 6.04 (1.00) ?,
v The plant is operating at 50% power when a RCS wide range temperature olement fails high. Which one of the statements below best describes the ' affect on deltaT and Tava?
- a. DeltaT decreases, Tavs increases
- b. DeltaT increases, Tavs decreases O
- c. DeltaT decreases, Tavs none -
- d. DeltaT none, Tavs none
- o. DeltaT none, Tavs increases GUESTION 6.05 ( .50)
The radioactive liquid waste monitor tanks dischcrge to the River Water Makeup Systen Blowdown Sump to dilute the effluent with River Water Makeup flow. TRUE or FALSE? DUESTION 6.06 (1.00) (a) What is the normal DRPI accuracy? (b) What'is the DRPI accuracy with a Data A failure? OUESTION 6.07 (1.00) What two conditions will automatically shut the SG Blowdown System demineralizer inlet valves? Setpoints not required. QUESTION 6.00 (1.00) What is the flow path of steam for steam chest warmins during a turbine startup? OUESTION 6 09 ( .50) How long is a ESF battery rated to last when it i s supplying full load? (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 12 ;
QUESTION 6.10 ( .50) . What is the first ESF battery load that will fail due to battery discharse? i QUESTION 6.11 ( .50) What is the source of power for the automatic field flash of the Emergency Diesel Generators? QUESTION 6.12 ( .50) How many successive starts are the Emergency Diesel Generator Air Receivers designed to perform without repressurizing them? 00ESTION 6.13 (1.00) What electrical parameter will change if the Main Generator excitation field current is changed when (a) the generator output breakers are open. (b) the generator output breakers are closed. QUESTION 6.14 (1.50) 2/3 (two out of three) pressurizer pressure channels below 1970 psis (P-11) allows the 9tator to manually block cer.tain protective functions. What are [three thel protective functions blocked by this action? QUESTION 6.15 (1.00) Cold Overpressure Mitigation System (COMS) provides overpressure protection only if the system has been armed,
- a. What signal (or condition) is necessary for the COM9 system to be armed assuming normal plant shutdown?
- b. What component other than the PORV itself is affected by the arming signal?
(*r*** CATEGORY 06 CONTINUED ON NEXT PAGE *****) e
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 13 C:
QUESTION 6.16 (1.00) 1, The output of the dynamic range deltaP transmitter of the Reactor Vessel i Level Instrumentation System (RVLIS) is used to provide information for conditions of forced flow.
- a. State where the taps are made in the reactor vessel for this (0.5) measurement.
I
- b. At hot zero power with 4 RCPs running, what should be the output of this channel? (0.25)
- c. What condition is indicated by a decrease in this outpute assuming zero power conditions and 4 running pumps are maintained? (0.25)
DUESTION 6.17 (2.00) List four items in the Rod Control System the Startup Switch will reset. OdESTION 6.18 (1.50) List the three inputs to the Rod Insertion Limit Computer Comparator. QUESTION 6.19 (1.00) List four design considerations the Main Steam flow restrictor no: les were required to meet. QUESTION 6.20 (1.00) List the two different inputs to the Reactor Protection System used to satisfy P-9 (Turbine Trip-Reactor Trip). 00CSTION 6.21 (1.00) According to the turbine runback procedurer 10012-1r what are the three signolt thet can generete r velid (cs opposed to une:<plained) turbine runback? (Include sctroints where applicable) (wrna CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 14 f 7
4 QUESTION 6.22 (1.00) g List all the auto start signals for the motor driven auxiliary feedwater '.!; pumps. Include logics (coincidences) and setpoints. QUESTION 6.23 (1.50) List all containment isolation phase A actuation signals. Include coincidence (logic). N0 setpoints required. QUESTION 6.24 (1.00) List the five loads supplied by the essential chilled water system. QUESTION 6.25 (1.50) List 6 of the 8 indications required to be present on either of the remote shutdown penels by Tech Specs. QUESTION 6.26 (2.00) List the four signals that initiate safety injection. Setpoints and logics (coincidences) not required. QUESTION 6.27 ( .50) The plant is operating at 100% power with the steam dump system normally aligned. Suddenly PT-506 (loed reject) ins t r u nient fails low.
- a. Describe the response of the steam dump control system to the-failure (ier specifically, how will the chanSe in PT-506 input change or effect system status).
QUESTION 6.20 (1.00) Describe the control functions of T1-301 (demin inlet temperature control) during boration and dilution evolutions. Include operating temperatures and valves controlled. (***r* CATEGORY 06 CONTINUED ON NEXT PAGE *****)
4 $l 3
- 1 i; 3
2 i > PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION ". ] 6. , PAGE 15 ______________________________________________________ , j
.)
J. c: GUESTION 6.29 (1.00) 1. . Sketch the logic symbols that represent i __ (s) a 2 out of 3 coincidence and function. t (b) an adjustable time delay enersiziris function. si i, J 1 4 l i I i j (***** END OF CATEGORY 06 *****) I i . k l a e I )
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- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 16
--- EA5i5E65i5AE 55siE6E--------------------~~~- ,.
w QUESTION 7.01 ( .50) +
~
At what voltage is a 125 volt 1E battery required to be charged? QUESTION 7.02 ( .50) What is the proper voltage of a 125VDC bus battery charge? N D QUESTION 7.03 (1.50) List three means of minimi=ing the possibility of water hammer during oper-otion of the AFW system. QUESTION 7.04 (1.00) Requests to e>:ceed administrative radiation guides are made by
- a. the individual receiving the dose.
- b. the immediate supervisor of the individual receiving the dose, c.-an HP Lab Foreman.
- d. the Superintendent of Operations.
QUESTION 7.05 (1.00) Following a Reactor Trip / Turbine Trip, the generator output and field breakers are verified to open by the operator
- a. when directed by 19000-1.
- b. after all immediate action steps of 19000-1 are complete.
- c. as part of the verification of the Turbine Trip.
- d. 30 seconds after the Turbine Trip.
(rrrrr CATEGORY 07 CONTINUED ON NEXT PAGE rrrra)
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I___ _ __ _________ $____ ___b1___ _ _ _____ I RADIOLOGICAL CONTROL ,, s
.'ga QUESTION 7.06 (1.00) ',
t' Following a Reactor Trip with no SI, AFW flow should be throttled
- o. whenever total AFW flow exceeds 550 spm.
- b. when all SG. wide range levels are greater than 60%. .
O
- c. only when directed by EOP 19000-1. ;<
- d. as soon as subcooling is verified.
QUESTION 7.07 (1.00) Which of the following is a 10 CFR 20 exposure limit?
- a. 5 rem / year-whole body,
- b. 1 rem / quarter-whole body.
- c. 18.75 rem / quarter-hands.
- d. 7 rem / quarter-skin of whole body.
QUESTION 7.00 (1.00) Which of the following radiation exposures would inflict the greatest biological damage to man?
- a. 1 Rem of GAMMA.
- b. 1 Rem of ALPHA.
- c. 1 Rem of NEUTRON.
- d. NOME of the above; they are all equivalent.
QUESTION 7.09 ( .50) When determining an individurls occupational dose, any emergency doses are to be EXCLUDED as part of the evposure history. TRUE or FALSE? (rrrr* CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 .
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RA6EUL6656dL C6 TRUL'~~~~~~~~~~~~~~~~~~~~~~~ h QUESTION 7 10 (1.00) }7 What is the purpose of an HP Margin List? QUESTION 7.11 (1.00) Why are some Emergency Response Procedure step subtasks preceded by f bullets (o)? DUESTION 7.12 ( .50) - One of the purposes of the Foldout Page is to direct an operator to the appropriate FRG in the event a Red Path limit is violated. TRUE or FALSE? OUESTION 7.13 ( .50) Does a required task in the Optimal Recovery Guidlines need to be complete before continuing to the next step? DUESTION 7.14 (1.00) (a) If an Orange CSF is encountered while performing an ORG, is it permissible to continue with the ORG while performing the actions of the applicable FRG? (b) If a Red CSF is then encountered while in the FRG for the Orange CSF, is it permistible to continue the FRG for the Orange CSF while performing the FRG for the Red CSF? OUESTION 7.15 ( .50) What is the maximum allowable differential pressure across the Mainsteam Isolation Valves when opening them? UULSlION 7.16 (1.00) List two different conditions when an Inverse Count Rete Ratio Startup procedure is required. (**r** CATEGORY 07 CON 11NUED ON NEXT PAGE *****)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 19 -
~~~~ - ------------------------
RA515E55iCAt 56sTR5t ____________________ g QUESTION 7.17 (1.00) During a reactor startup to 100% power, at what power level are the Reserve Auxiliary Transformers transferred to the Unit Auxiliary Transformers? i' QUESTION 7.18 (1.00) What immediate operator action is required upon loss of offsite power ' concurrent with the A diesel generator failing to tie on the bus? DUESTION 7.19 ( .50) Where does the Shift Supervisor report to during a Control Room Evacuation event? QUESTION 7.20 (1.50) List three items that the startup procedure requires the operator to check after every 50 steps of Control Rod Bank withdrawal. QUESTION 7.21 (1.50) List six actions that should be accomplished prior to evacuation of the Control Room when taking control at the Remote Shutdown Panels. QUESTION 7.22 (2.00) List the immediate operator actions for an uncontrolled continuous rod insertion. QUESTION 7.23 (1.50) List the immediate actions required for a spent fuel handling accident in the Spent Fuel Pool Duilding. (x**** CATEGORY 07 CONTINUCD ON NEXT PAGE *****)
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND 'PAGE 20
;d RADIOLOGICAL CONTROL l, q
QUESTION 7.24 (1.00) - y' When is a continuous action of a WHEN, THEN step in an Optimal Recovery Guidline carried over to a 'transitioned* Emergency Procedure? QUESTION 7.25 (1.00) Explain the conditions which would have you implement each of the following j Nstural Circulation Cooldown procedures. ' (a) E0P 19003-1, Natural Cirevlation Cooldown with Steam Void inVessel.(440'L/ (b) EOP 190t4-1, Natural Circulation Cooldown with Steam Void in Vessel.h$bN QUESTIDH 7.26 (1.00) What checks are you required to make in deciding the appropriate action to take, when your STA reports an Orange CSF during an event? OUESTION 7.27 (1.00) During a plant heatup with the reactor trip breakers closed, the procedure has a precauti,n to withdraw all rods at least 2 steps whenever the RCS temperature increases by more than 50F. What is the purpose of this requirement? QUESTION 7.20 (1.00) How is the boron concentration between the Reactor Coolant Loops and the Pressurizer equalized? OUESTION 7.29 (1.00) How is desassing of the Reactor Coolant Systen, accomplished? (***** CATEGORY 07 CONTINUED ON NEX1 PAGE y****) X
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- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21
^
RADIOLOGICAL CONTROL
't QUESTION 7.30 (1.50) I.
How are spent radioactive cartridge filters removed from their housins and placed in the Radwaste Solidification System? 4. (***** END OF CATEGORY 07 *****) i e 4 l l l I
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 22 l
QUESTION 8.01 (1.00) During a plant startup with the reactor below the P-6 setpoint, one power range channel (N0043), is found to be inoperable. Which of the followin3 statements is correct? Refer to the attached Tech Specs.
- a. Operation above the P-6 IRM neutron flux interlock setpoint is not allowed until the inoperable channel is repaired and declared operable.
- b. If N0043 is placed in a tripped condition and the other three channels are operable, you must reduce the power range neutron flux trip setpoint to =or< 85% and thermal power is restricted to =or>75%.
- c. If N0043 is placed in a tripped condition and the other three channels are operable, operation to 100% rated thermal power is unrestricted.
- d. The only restriction on proceeding to 100% rated thermal power are that N0043 be placed in a tripped condition and quadrant power tilt be monitored at least once per 12 hours.
QUESTION 8.02 (1.00) Which of the following statements is correct concerning the quadrant power tilt ratio (OPTR)?
- a. If GPTR exceeds the maximum limit of 1.09 for 1 hr, the reactor must I
be immediately shutdown.
- b. If misalignment of a control rod causes the GPTR to exceed 1.09, thermal power must be reduced within 30 minutes.
- c. No action is required within one hour regardless of the OPTR.
- d. If GPTR exceeds 1.02, but is less than 1.09, operation may continue indefinitely only up to the reduced thermal power (100-3% for each 1%
of indicated GPTR in excess of 1.0) allowable for the RCP combination. GUESTION 0.03 (1.00) The " Search and Rescue" procedure, 91402-C, assigns responsibility for
... forming, briefing and dispatching search and rescue teams'.
- a. To whom (title) does it assign this responsibility?
- b. In the event that person is not readily available, to whom (title) does the procedure assign this responsibility?
(***** CATEGORY 00 CONTINUED ON NEXT PAGE *rrrr)
se .*
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 23 GUESTION 0.04 (1.00)
According to plant administrative procedure 00005-C, an individual should not be permitted to work more than 24 hours in any ______ hour period (excluding shift turnover time) unless authorized at the ______ level or above. (Choose the correct pair below).
- c. 48, General Manager
- b. 48, Applicable Department Superintendent
- c. 72, General Manager
- d. 72, Applicable Department Superintendent -
QUESTION 8.05 (1.00) A person on your shift (you are his supervisor) determines that a procedure being used provides incorrect directions e.g., some steps are out of appropriate sequence.
- a. What immediate action do you expect the user to take, and
- b. What action will you take?
00ESTION 8.06 (1.50) During a reactor startup with Tave= 410 degF, the chemist on shift reports a confirmed baron concentration of 1910 ppm in the RHST. All other Tech Spec systems are normal for reactor startup. As OSOS classify this situation as directed by " Emergency Classification and Implementing Instructions", 91001-C. Provide your justification for your classification. QUESTION 0.07 (1.00) Vogtle procedure 00003-C states one of the responsibilities of a supervisor during shift relief to be " Ensure oncoming personnel are mentally and physically fit prior to turnover". Define ' mentally and physically fit" as used in this context. (r**** CATEGORY 00 CONTINUED ON NEXT PAGE *****)
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 24 QUESTION 8.08 (1.00)
Immediate notification of NRC is required if the release of radioactive caterials in concentrations which, if averaged over a period of 24 hours, would exceed ______ times the limits specified for such materials in Appendix 'D*, Table II of 10CFR 20; ______ damage to property in excess of
$______. Fill in the blanks with the correct information by choosing one of the below combinations.
- c. 500; ori 2,000
- b. 500; andi 2,000
- c. 5,000i ori 200,000
- d. 5,000; and; 200,000 OUESTION G.09 (1.00)
Answer each of the following true or felse.
- a. In order to raise a new fuel assembly with the new fuel elevator, it requires the approval of either the Shift Gupervisor or the Fuel Handling Engineer.
- b. In order to raise an irradiated fuel assembly with the new fuel elevator, it requires the approval of both the Shift Supervisor and the Fuel Handling Engineer.
QUESTION 0.10 (1.00) If, while operating the spent fuel cask bridge crane, a contact should stick in the "on' position, what three specific actions are required by the Vogtle " Spent Fuel Cask Dridge Crane Operating Instructions'r 93230-C? State them in the order in which they should be performed. QUESTION H.11 (1.00) Tech Spec section 4.0.3 relates to failure to perform a surveillance r equiren ent uithin the trecified time intet vel . Diseves;
- a. the consequences of the failure, and
- b. two exceptions to these requirements.
(rrrrr CATECURY 00 CONTIMUED ON NEX1 r% C C wrrrn, k - - - - - - - - - - _ _ _ _ _ _ _ _ _ - .
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 25 GUESTION 8.12 (1.00) ,
Procedure 00052-C, " Temporary Changes to Procedures", defines a " procedure intent change" in terms of five items that may be involved. State these five things that may be involved. QUESTION 8.13 (1.50) Define containment integrity per.VEGP Tech Specs by stating its six requirements in the format provided below. De brief. Surveillance requirements and numerical values are not required. Containment integrity shall exist whent
- a. All penetrations required to be closed during accident conditions are eithert.(1) or (2)
- b. (3)
- c. (4)
- d. (U) '
- e. (6)
OUESTION 8.14 (1 00) The VEGP Tech Specs provide two sets of RCS chemistry limits for 02, F1 and Cl(one steady state and one transient). State the two parameters on which the effects of exceeding these chemistry limits are dependent and discuss the implications of gperating with gg[.;p.ga.t concentrations in excess of the steady state limits. QUESTION 0.15 (1.00) While in mode 5, each of the following components which have a minimum I temperature as part of its LCO in Tech Specs is found to.have a temperature below the LCO miniraum. State all Tech Spec required actions (if any). The pressuri=eri the boric acid storese tank; the cold les accumulatort the refueling water storage tank. (rrr** CATEGORY 00 CONTINUED ON NEXT PAGE rwrrr)
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- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 26 QUESTION 8.16. (1.00) l DG1A, which supplies safety related 4 KV bus A (1AA02) is inoperable. RHR pump B supplied by safety related 4 KV bus B (1BA03) is inoperable. The Tech Specs for ECCS subsystems and AC sources are attached. Which l statement is' correct concernin3 continued operation in mode 1?
l t
- a. The action statements for both the RHR pump and the DG are applied independently, each must be restored to operable in 72 hours.
- b. Since the DG is required in mode 4 and the RHR pump is not, the unit must be taken to mode 4 within 72 hours.
- c. TS 3.0.3 applies; it requires action to place the unit in a mode in which the ECCS specification does not apply within 1 hour.
( d. TS 3.8.1.1 action C1 applies; it requires action to place the unit in a mode in which the ECCS specification doer not apply within 2 hours.
'0UESTIDH 0.17 (1.00) j If the shutdown mar 3i n is less than the Tech Spec required value in mode 5, j the required action ist (state the action required by Tech Specs)
I 00ESTION 0.18 (1.00) l The Tech Specs establish RCS temperature / pressure limits for heatup, cooldown and inservice leak and hydrostatic testin3 State the portion of the specification which applies to inservice leak testing. l ! DUESTION B.19 (2.00) l The Tech Spec ECCS limit on RWST minimum volume and boron concentration ensure thatt (Discuss fully but limit your discussion to ECCS applications) 00EC110N B.20 (1.00) l Administrative procedure 10000-C states four situations / conditions under uhich the operatins shift must shut down the reactor. State them. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
(( ' 7
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMI1ATIONS PAGE 27 QUESTION 8.21 ( .50)
What does a yellow dot on a red key tag for a' controlled key indicate? QUESTION 0.22 (1.00) On what parameters are the TS operabi.lity requirements placed for the RCS cecumulators. State the four parameters without setpoints. QUESTION 8 23 (1.00) In reference to procedure 00054-C, " Rules for Penforming Procedures", list four of the five steps required to be performed prior to using a procedure. QUESTION 8.24 (1.00) In reference to proced.or'e 10007-C, ' Reset' of Lockout Relays and Relay Targets', the shift supervisor' is normally responsible'for verifying that , all corrective actions on a tripped lockout relay are completed before requesting permission to reset the relay'. However, there are four ! exceptions. State the four exceptions with any pertinent conditions and I authorizations required.
- QUESTION 8.25 ( .50) ,
\
True or False Standing orders address subjects limited to plant operating procedures. i 1 QUESTION 8 26 (1.00) , , ,, State the two purposes of a radiation work permit (RWP;) '~ per procedure 4 43007-C. , J L (***** CATEGOR,Y 00 CONTINUED ON NEX1 PAGE *****) 3
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- 8. ADMINISTRATIVE PROCEDURES, CONDI1 IONS, AND LIMITATIONS PAGE 28 QUESTION 8.27 ( .50)
An emergency Tech Spec change can be made on a back shift by the Plant Roview Board if the Shift Supervisor can muster the minimum number of asmbers to satisfy a quorum on a conference call. TRUE or FALSE? QUESTION 8.28 (1.00) List two different applications of a temporary jumper which are NOT required to be controlled by a Jumper and Lifted Wire Clearance. l l GUESTION 8.29 (1.00) l l List the conditions that must exist to allow the On-Shift Operations l Supervisor to authorize a reactor startup following an unplanned reactor trip. QUESTION 8.30 ( .50) A Work Order is required for removal of any annunciator card. TRUE or FALSE? t l (***** END OF CATEGORY 00 *****) (rrrwx*rrwwwww END OF EXAMINATION *rmurrrrrrrrrrr)
. Equation sHEtt Where *1 = se c.
(eensth)gtvelec1ty)3(erea)3 = (sensity) (veloctty):(arealg KE = uvI PE = ogh Pig + keg +PgV 3 =* PE 2 *EIt'P T22 where V = specific
~T volume P = Pressure 0 = &c,(Tout.Tg,) 0 = UA (T ,,.7,,,) g = s(h3.hg)
P = P,10(SUR)(t) p . p,,t/T SUR = 26.06 T = (B.p)t I P delta K = (K,gr-1)/K,gg CRg (1-K,ggg) = CR 2 II-Keff2) CR = 5/(1.K,ff) M = (1.K,ggg ) SDM = (1.K,gg) x 1001 tl-Keff2) Keff secay constant = In (2) = 0.693 A = A,e-(decay constant)m(t) t t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 10 10 dps
- 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 10 3Btu /hr 3 6 Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 pm/cm 1 Btu = 778 f t-lbf l Heat of Vaporization = 970 Btu /lbe Degrees F = (1.8 x Degrees C) + 32
- Heat of Fusion = 144 Btu /lbe 1 inch = 2.54 centimeters2 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-1bm/lbf-sec l
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, PROCEDURE NO. REVIS80N PAGE NO.
, Ig940-1 1 2 of 2 ,
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4 DATA SHEET 1 - ESTIMATED CRITICAL CONDITION .j- i k w Date/ Time of Last S'hutdown
#I hk M f Date/ Time of Projected Criticality b /_ OM
- 1. Conditions Prior to Shutdown \. ).
- a. Fue1 Burnup '7; ##O
- b. Re&ctor Power /so 24
- c. Boron Concentration poo 1 -
- d. Rod Position soo /- te s on Bank M
- 2. Projected Conditions at Criticality' , ,
- a. Number of Hours from last $utd wn AY
- b. ESTIMATED CRITICAL POSITION
/ /h Steps on Bank
- 3. Reactivity Balance -
- a. Power Defect (PTDB TAB 1.1 for c,/
\ ffg p boditions in 1) PCM
- b. Xenon Worth at Pg6er N (PTDB TAB 1.4.24or conditions in 1) 2 $00 . PCM
- c. Rod Worth at Potiet NO (PTDB TAB 1. '.1 for conditions in 1) PCM
- d. Boron Worth t Power '
(PTDB TAB . I for conditions in 1) BOO PCM
- e. Xenon Wo t riticality (PTDB T and ho 1
s h.4.12) for conditions in 1 - ggg PCM
- f. Rod p rlh at" Criticality MO
( B TAB 1.5.1 for ECP in 2) - PCM
- g. Sum thru 3f /0, 2 @ PCM
{ h. TIMATED Critical Boron Concentration y 7 [ B TAB 1.3.1 for worth in 3g) PPM gfleteby / / Signature Date Time Reviewed by / / Signature Date Time fL}QYO
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APPENDlX E STEAM TABLES AND MOLLIER DIAGRAM De following Steam Tables and Mollier Diagram have been reproduced with the consent and by permission of Combustion Engineering, Inc.,1000 Prospect Hill Road, Windsor, Connecticut 06095.
- - ~ _ -
., . . Volume lil Appendix E Table 1. Setweeted Sleem:[empe7etere Tettle 7bt PMS~ Spr6dat W94eme 18644py Irtropy temp 10 pn Set Sal Sol Set Set Sel letw Se in leted Isop lee 0
- 8. , % 0 Vapte tites leep Waptr litet leep vatte feev
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38 0 010MS 3 016070 20M O 10M t 4 000 Semi 7 1767 7 1007 ast 30 0 0 Ilpet 8073 7 30772 S WII 2 4651 2 3737 se 0 016019 26M i alm 7 6 014 1972 1 8078 1 0 Ol?? 2 ISet 2 1463 80 8
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- 13sl i 1434 7 '34539 1500 0 15858 166( 7 O 9232 0 979C I 0358 3 0939 8 1519 1 2060 1 7529 1 2946 3 3371 1 M44 l I3937 1 4?02 3 4677 3 5100 Se - Si:perteat. F ti = enthalpr Etu per Ib b* 50f Cil C volume, tu fl per Ib 5 = entropy. Blu per F per lb E-11
l s j
, , . - Appendix E -
Volume 111 t, - TetF6e 3 Sisselheeled Sleem-Confiaveef en hn, , lb 5e la u w "'S**h'" I ** 8 * 85al lemen mA si. e Me utee . tes _ _She sera. on'67 e st's, Me Gee SNe 1980 lite lite 130s lile late NN MN t
> 779 % e46 e 917 % pef t seass s ein osan 6tw se4' scue seus es??7 e s77. emne . eses' e9tm e 4?4? IPN' 1 D'.1 11s.= 4 184 3 I?lu 17e0? IJI4 7 19944 8444 6 laul g IS7e 7 leseg 844' l 192's S N14 l ?s et 8 J?D8 11%3J l Ipe? loll? $ 4%% $ $e?3 *!
{ lllet e
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?ste I /T?7 19807 13ada 1 8730 l aufs l esig l esa,7
- 19ett e
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- P *lla eti
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- e. oO774 ;M o e'v. 3 0 79 eaa etia ee3 0 enl7 essa, em3% e m e n n ee6n sw u een?
ein, .j s ams t 976 9 lents 1115 7 Ileal I? elf thee thS4 lanes 1477 ltse ? lN31 ,. lael . efiel e 9b18 Iel?7 I m37 1 1841 8 86$3 81187 )7bla I 1910 13Ma I abl6 3 Set 0 Ian3 1dem ee ' h s I'J >. ea SJ 4 som s 9004ce?7s sue' seJa esa seem eau era? siste emeV ene se7n ease" - mas leh.? lle> 3 1974 4 I?Is t I?99 6 1A65 See96 84S94 1%t16 56374 e 907' s Tsh? Isome 1 9478 8 le76 3 1578 1.4J0 ' 8 744', 181e0 , e R*f, Ileil 1 337,5 8 3e9,8 , 8 376s, a 4/vl 8 4783 e h 77's e6344 e'%l e O?70 es'97 eel?S eone eet04 e es48 eett? eob3% 0 eb *t eebit e 59n. ee764 t o*? ? %?? 1920 letti 1866 3 l'79 7 !?91 0 IM81 140l b 145! 8 3%85, 16Ji t e 90 ', a kap i 3c 47 I p /4 110lJ l 189% 5 1948 1?MI I ?749 I Ents 3 3816 1J7eb I 4/71 4475 Heeg , e 3 0'43 00'4s CO.t! e 0.11 e e3.T es3%a seJ9' se8P e W78 es%IS esb% esb95 eE70 se740 778 4 a e 90l* e 9515 ILt? 6e 999, peat 3 tbs O 90'4 5 IM?) inne s I?tle 17e10 IM07 IN38 1484 4 ISJe g 86M S I ed73 198$3 3 84.T l 1871 1 178? I 1678 8 3U78 3 3339 1 3631 14 53 1 4617 l4les . 6 00 38 Os Js 0 0 eJ e O/87 00Jia 00.us seJ40 D e818 eses 8 S ee90 0 8534 e OS73 s eL4A 0 0784 5 7;s4 ello 0 P9M swi s TSs J 80896 lost. / lilla l?lJe I??S a I)J!v BM64 14373 15376 1675 1 l' 44 h3 0 99')? Ies.'n IestP &IR? I 1808 I ? lob I 7597 1 7949 1 3766 13%0 14087 4 ASSI 110:1 ,
> 00/JO 769 5 D8'95 O 8J 00'6:
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32.3 INSTRUMENTATION - '! i
]~3/4.3.1 REACTOR TRIP SYS*EM m.~TRUMENTA1 ION .. ..
i.,.._ i Y -
*C J.? LIMITING C0_Nr:if0N 70R CDERfTION'
,' ', . .m p, n x,
m. 3.3.1
'.pfv2(fj interlo kAs a minimum, the -Reactor Trin *ystem instr 2mentat1 .:. .inels anc * -
~ 4 [M{fl Taplec s3.3-2.
of Table 3.3-1 shat! be OPERABLE with 4ESPONSE TIMES.as'. shewn in
, :x; es .
f, % T/y,u APPLICABILITY: <;. . A,s,'shown in Table 3.3-1. f M ACTION: -
<j@y:yg@;y g
yr ,s,.,...... g .: - .: '.-
~- -
a. 9 = & -4FC 7.% sim 7. i. : Table <3. 3-1. .:. . c da,o, W &a.WAiR Q??$b.e-y" gO Gmi . SURVEILLANCE REQUIREMENTS lHCdL' , pf W -
. ?.
. da U ,.( f,,,
- 4. 3. 1. 1 Each Reactor Trip System instrumentation channel and interlock and
[rf- the automatic trip logic shall be demonstrated OPERABLE by the performance of
- f. y
=p ,
the Reactor Table 4.3-1. Trip System Instrumentation Surveillance Requirements specified in (- ,,I
- w. ~
9.-4.3.1.2
- The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total numoer of redundant channels in a specific Reactor trip function as shown in the
- 4. , " Total No. of Channels" column of Table 3.3-1.
l Y
.en i xc IQ 1
1 i 1 1 1 l l l . i 1 r *. ." ' ' ' - - I '. e
.2 . w . m . .. - * * * - " ' ' * ' ' ' _
em,' W *d
_q I 4.2.3: z.s
'~-? . ... .~ , .'k This specification nas been deleted and replaced with a specification whtcht A defines tne limit for the nuclear enthalpy hot channel ~ . factor (F N' 14gst e rms, "
of .n'e desigr1 basis (-1.55) and a factor.which relaxes tne Pmit o$)FnW function of thermal. power te allow .for changes ina thelradtat power edidr all permi ssible rod insert.iertilmi.ts._f .Ihar.is. tn..say.i ..;/. ,-.c ?
;'s' ' * ;'. . - ~-
A +jglg , t,; *
- y, y s1.65$1+D.3'(1-P)1
~ '
[. .
- +
.. F .
e.-. .. where P =. Therma'l' power 1 7_ _ Rated Thermal Pow,er . The design. basis ofe 1.,55i.is used because the .4 perce nee for.. - -
. .i t1 W- experimenta12erdoM kii kberapp-lied to the; measure . F..was:tecui red e. w
.e c'..-iby 4. 2'. 3.2~.- See pa'ragraph. 4;4,A-3.1. of3 the. VEGP i discu'sston mf"
....'r
~
e- S i ' ' ' ' ~ the above. 1:v 'e - Tc" - -
- c.: > .w:; ,
The limit for RCS flowrate has been moved t ation- 3//,c3 65 6*=652 '
.0 .
Therefore, the revised specifications meet h ent of the. orginal specification while providing simpler m af _ ghtforward recui rements for the operation of the plant. -
^; .g- - *;
! 3/4.2.5: .
/
The action statement has been include the additional CNB parameter of reactor coolant , tal flowrate as discussed in the justification for 3/4'.2.3, This parameter was added~due to the deletion of the rod bow-p -its tradeoff with total flow in
;^
Specification 3.2.3.
.. ~ ..
Surveillance Require .2.5.2 was added to reflect the 1 addition of RCS total flowrate to nd is based on the generic STS Surveillance Requirement 4.2. Footnote *:an lance Requirement 4.2.5.3 were added to allow the plant to enter to_ perform the calorimetric flow measurement and calibrate. -rea r co'olant system total flowrate indicators following the first rea tart-up after the initial core load and each refueling. This will al ant'to demonstrate compliance with this specification's LCO upon com n of'the calorimetric flow measurement and calibration of the reae . t system total flowrate indicators. Ih he 1:
/ wW Ny change is a result of the addition of RCS total flowrate to sce d.fication 3/4.2.5 and the intent of the plant not to coerste in N-1 ' coo coerat;on.
'20dv Mae13A MCS135
a , 3 a. 3 INSTRUMENTATICN 3/J.3.1 REACTOR TRIP SYSa*EM in.~TRUMENTA110N. ,
~~
LINITING CONr*iION FOR COERFTION : 3.3.1 As a minimum, the Reactor Trin System instr 2mentatia- .:..anels and-interlocks of Table 3.3-1 shall be OPERABLE wit" 4ESPONSE-TIMES.as'.shown in - Tanle 3.3-2. ArPLICA8ILITY[_ A,sfshown in Table 3.3-1. ACTION: ,_.., ,_
.'.. s... .. i. : Table i3e3 .1.? . . 's . ' '
SURVEILLANCE RE'0UIREMENTS i.
' 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERA 8LE by the performance of .
the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1. 4.3.1.2 The RIACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the
" Total No. of Channels". column of Table 3.3-1.
~
l -
. -i
& L .: '
- Rb;.a l -r%r -
\ 1 lyl,,f..[L.
- u. g. -
, ; .h - :+-
~
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t- "= l ! 1;T'. '
."!' . '. 1
~
( , . i q. fg
~ !m. --
TABLE 3.3-1
/;
H REACTOR TRIP SYSTEM INSTRUMENTATION <
,. s ,
*k 31 HINIMUM i
, TOTAL NO. ~ CHANNELS CHANNELS APPLICA8LEl l, is luuCIIONAL UNI 4 0F CHANNELS 'TO TRIP OPERABLE MODES ' ACTION
'~
- 1. Manual Reactor p 2 1 N H 2 8.- 1, 2 o , f , 1 i m;r (c . , cs s 2 1 0".n23 : -
3f04fL57. 11 9
, i,.-,,"; -
t,y: q.p t ;
?., - : ,
- 2. Power Range,' Neutron. F - l' 1.
1, 24 J.
- a. e liigh Setpointf '
2'y 4 2 T391 -
';3 27'N
- b. Low Setpoint 1###,721Q/.. ff 2 '
? g-
,, u ros i, t > ' >. .
,. b
- 3. Power Range, Neutron Flux -
2 3- . 1, . 2 . . 2f-liigh Positive Rate"' ' ' A ,J
)
.' 7 ,
(Noost Noo+2 Moo n Moo + 4) 'hei 6
,, 4. Power IIange,' Weutron ' Flux, 2 3 1, 2 ;
2r
',. liigh Negative Rate' d .
, 1 -
... Woo 49] ^,a 3 '
.., 5 t,vovst, IntermediateWoo dRange, z, Woo'43, Neutron Flux ' 2 , ,' 2 1### , -2 ., 3
( woo 15,' /voo n) I ' 6 Source Range, Neutron Flux /Noo3i, Nooiz ), . t c
- a. Startup. i'W 2 1 f
- e- 20# g 4 4 3 ' 4, 5 5
- b. Shutdown 2 1 7 Overtemperature AT' c 4 2 A. 2. ; .. > t, b f ,'
iIE
- a. oett,r Fea Leop-Piantiflo o 21 Tf-c e .s I, fr.ov vi) 3ij;.[pif.2 ,
I Feve-teep-Opera t4en 4- 2. 1.. 44
*: Three-Loop-Operation 4 188 34 4!. 2 4-0 i.
p{,J+ br----T hree -leep-Pla nt I Three-teep-Operat ion - - 2 2 -- 1r-2(1.
Q:gg e$, ,g'g. . .
g7# ' 1,-2 ' ~ Two Lecp-Operation -
-3 1** 2-
, . . - c- .s . .. , . , , , .
,+ . :: .4 9
~ ~ . o .,, -
i i i 8 m pl i
% *a
.. . U '. .. J
. . . . .- _ . . u i
. TABLE 3.3-1 (Continued)
Ei < REACTOR TRIP SYSTEM INSTRUMENT.? TION,
. . m
' ..* i,, MINIMUM S
u7 'M TOTAL NO. CHANNELS CHANNELS
-FUNCTIONAL UNIT'," W). . APPLICABLE
. "*3 _
0F CHANNELS TO TRIP g' 2 ,y n . 0.' ERA 8LE N00ES ,Cil0N *
.- d. .- -
*0verpower AT ' ' ^ ^ s[5.,
a,----Fe E:;l Ph:1 3 4 4 2. [E*
.d.M i
' 1 , 2. . t i t ' - Four-Leop -Operation -- - --- - ---4 - - - - -
--2 Thr:: b:; Operation 4 laa
--- S PJ i -- - - 1 r-2 6#
,, ,pinm-a 3'- 1 -- 2 9 b.- ,Theef E:; Plant '
N e (Three-h:; Op ration 3 2-h
.I ' '
' T w-Loop Operation l'r2' 7' 3 1** 2 l 're-otzt ri Ir2 4
- 9. r[r.r' essur or tfier Press 6re-o4f' Low rs-o4* /) 4 2 tsee aho i; ra s./, s. e- d l, u
- a. Four--Loop- Alant- 3
{ A Three' Loop Plan 4 2 - - - -- 3 1, M 3 2
... ( Pf-opsf 9 6 o 1.54 PI 04 S7 PI-c 4 Sil 2- 1. N i
"' 10. .Pressurt/er Pressu,re--High ' 5 2 & b i
- a. Four-Loop Plant j b. Three-teep-Plant-4- ~2- d- - -- - 1lh 2 2 6#
3 2 2-
,i. .. n .c I g-2 7#
- 11. Pressurizer Water Level--High 3 l 11-c9 69, LI-h 4 & o,' LE- 0 44 ! )
2 2 1 -7P 6 l 12. Reactor Coolant flow--Low > i
- a. SI'ngle Loop (Above P-8) 3/ loop -7+ 6,'
i a- >eci" I- 2/ loop in 2/ loop in 1 any oper.- each orer- __ g , ating loop ating 'oop
=.n .
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop ;" 1 -75 /
below P-8) two oper- each aper-
?ating loops ating loop
--.3
_. lI
! cnp 1 l co n >f
, l a.c. p 3 l c, ,y, I I
f l o -a I A Fl- ,41A
, I i -o4 5 4 ii *teI
['] . ,4c5 f 1 - o '4 / 6 I l s 4 3 *3 Ii ~1 * '
} w c a kca r' ion f 4 44' g- 4 g L, fI pqyy F1 o s a, '- 1 I- 'I **
l 1,
g . i - 6 ., E TABLE 3.3-1 (Continued) -
'" f3 '
ij REACTOR TRIP SYSTEM INSTRUMENTATION
,.,;. ,. 4.' HINIMUM' ,
, 4;r TOTAL NO. CHANNELS CHANNELS" APPLICABLE i
.3 lisliLIIONAL UNIT ,,uc p 0F CHANNELS TO TRIP 1 OPERABLE H0 DES: ACTION
- 13. Steam Generator Wat 4/sta. gen. 2/sta. gen. 3/s ta.h'g'en.- 1, 2 6Ffil Level--Low-Low f 6 in any oper- each oper' isie imis o a- id-ot iNMg ating sim.
. ,, c g N .n [,.'e.fn y. . gen.
ating@ty. Ta ble ms . 5f 3 ) gen. ',,,,
.. ,, 3 ,,
e t in
- t, (i.'P.7 ', t; -
1L -Steam Genera % Water--Levels ==L=n. ' ,g 1-s ter-gen. .t .- 1 + ten,7genr--+1, ! 7# Colne filent-W4h-Ster / evel-coln-Feedwahr, Flow-Nismatch-- levahand .
. sident-with 3 sta,/ reed-t . . o Wate ster / feed =--water-flow p~'issfi i , . . I, o L o$p 2-,,, Loop 4 *I'**
Wtch-in l'**
- I *** tchd * ,
each sta, samester-gen. 1 1I . si LI-Sif L I- Si1 LIL55) gen-- ,yd= M 2 #=; gen.
';' i r 51 'l L I - 5.t f ( I- Sri 'LT-562 , gen level-and--I
- N tar /fgedwater -
)f '
'LI;551' i 1-5si LI L3 - 6 53 .
_,gg,, 1I 8 li' LI-54J' ' L'I - 51'l_ . L I- 5d F ame-,s ta. ds ':(tau t i s' iV 'F
~
l *l i , - g w u , r, . > 6 3 e+
+C 15: ligdery 4- 2/6u s b Pumps ,ol.tage ; R,ep,ctgr pp,olant ,
2-I//m5 % . F1 f
- le y I.
i Jour-Leop-Plant p y p.2 y [gy--
- 4. 4-1/ bus
-2 c. .c:,,2, , 6#
. ., b.g,, gJegeeDi#8P i f.J a,ntr,----- ~ ---3-1/ bus- - -----2--- -
?#
a . as .tc.f.1 ; 16.y.Underf'requency--ReactorCoolant
" Pumps' 4, 2 / bus m . 2. -Jl. bu s Wh3 31 - ,[y ~
g 9 a.l -Four-Loop P1 ant - t w a ',t' 3 4E%(, 4-1/ bus b.--Three - Loop. Plant 3-1/ bus 2
'! p-{p 2
&3 -
1 g/h. w y1
--1 7 yp l
- 17. Turbine Trip Low Fluid Oil Pressure 4- --
i Mil u . . . (yfj yJ E.
- 4. 3
-7#kr,
~
2 2 1
- b. Turbine Stop Valve Closure -12#- n'3 4
qy('p - 4 'A :1 l's
- ( r']. C H,1, l'1 -t<n.2, 17 -tet&3 )
l ,, J er//)ris/wo 9. nT 34 .' 1..
,;.;, j
~
~
[./ t . TABLE 3.3-1 (Continued)
~' ' ,j '
REACTOR TRIP SYSTEM INSTRUMENTATION MINIf,UM -
.. , TOTAL NO. CHANNELS CHAN' EELS APPLICABLE- .. FilllCTIONAL UNIT ,d[,*f 4-p/j,, , OF CHANNELS TO TRIP OPEIABLE MODES ACTION
- ;qo .
,, k 10. Safety Injes$lK Input' ,
din . , , 2 2W- :1,2 lu y
.from E4f 39 m.
i 3 RU.. ,
- y. 1 -
atpr 1 ~
. Trip' Systeii Interlocks
\,,a. n- In
-f 19.,lteactor,termediate .
Range +c: :..t c NeutronFlux;^P-6 2 1 v e'2 , . 2M,,, e7 M-ooss t u. - n-
+ b;. ' (Lo' w ' Power Reactoro d-op.s
' ' ~' t). *
* Trips' Block, P-7i '
' &4 1- +7 P-10 Input /d-oo+/, 4 2 3
' < ';A l ' or #eo+ z, doo4 s , d-co*t) v N' '
.., ! P-13 Input 2 , .1
- 2 1 -8 ~
h . (PI- o coS, PT-o Se 4 ) t,, c. Power Range Neutron ,, m . .' di , pux,P-8,. ;,p ,
- q. 4 , 2 '3 1 7
. nt pfr0 0 +l N -0 .
% w (repory R4pga .04 2 W-co;>r N,eufron,.,;e 4 3, W-0044 +)! - ,a 2 4 3 1 -8 7 Flux, P-9 :
(d-oo*/, M-cotz, W-co+3, d-oo + + ) c. ~ ' ' . e. Power Range Neutron . . 4 4 x ; Flux,, P::10 4 n 2 3 1,2 8/
. ,J d- oo tl, W-co+ z , M-co4 4, W-co + 4 ) o . 0 I
' f. Turbine Impulse Chamber
, J.;;g fressure, P-13 2.., a 1 2- 1 87 i FI- o.So f PE -osofo ) , v. I- '
y i1
/ 20. ReactorTripdreakers 2 1 2- 1 2 10, 13 d 11y 2 1 2 3k 4f 5) / y e .:a i.
o 21. Automatic Trip and Interlock 2 1 2 1, 2 10 Logic 2 1 2 31, 4. * ? 57 " i17 e
. .ehe'
. l ,
- ~
i _ TABLE 3.3-1 (Continued). . TABLE NOTATIONS -- -- m .,
. %.x 16' hen the.RihetoriTrip: System breakers are in the closed position $andethe Control ~ Rod ~ Drive System .is capable of rod withdrawal. es
( K jf
**TL
- t :f ::r"!::
ch:Fa d :J'::di:f:te'd afth th triffe ctfen: derie d-
= :ter ::: het '::;':S P 5: ; he e *- t r C b- -
raadit4aa;.3 .,x. q . ...3 p jg *
' ,b j e provisions tofjspecification 5:0.4 are not'a. pplicabTh.
. a g .
c LM8elow the P-6 (Intermediate: Range Neutron Flux Inte )rSetpoint. : e/ ~ , i Q. -
' d.J DBelow2the R.10.*(Low'.SetpoinCPower Range Neut Int ~er:locit)uSetpointe '
ln.%-s,1C,
,. .<.fdbea
- e. A6ove. i si&R,irM& rhe.:f-t5deriNQ;-lk.eVo<
Mar +- h'e T9~ < e<- ic AY 6/ L 25. . .t .l*i; ~ i
'6 1 -:.e A,y./.a;/e. i% de.s s.nd Ac.+roo da + . se - : e. .,..,e - +c .x' 7;urle J. 5 -3 a.re nvr e- :
n .
- ACTION STA KT s ?r. : - ! 1. -j c c- .<
^ ,
1, ,, . . m: :c ACTION 1 - With the number of OPE ch els one less than the Minimum Channels OPERABLE requi restore the inoperable channel to OPERABLE status w i urs er be in HOT STANDBY within the next 6 hours;
. 'i -
~~~
. ACTION 2 - With'the number' i-hannels ode less than the Total
; Humber of Chann UP a'nd/or POWER OPERATION may proceed provided the conditions are satisfied:
- a. Yhei channelisiplacedirhthetrippedcondition with Y , 4
- b. Channels .0PERABLE requirement is met;f however, perable channel may be bypassed for up to + hours rveillance usting of other channels per Specification 1,; ard c. & e--
l i 3 ,e v-t er, T1ERMAL POWER is restricted to less than or equal t75% of RATED THERMAL POWER and the Power Range Neutron V
- 4. Flux, Trip setpoint is reduced to less than or equal to
- -{853% of. RATED THERMAL POWER within 4 hours; or, the
' QUADRANT POWER TILT RATIO is monitored at least once per
. 12; hours per Specification 4.2.4.2.'
,. :.:::w. .,x?n
,,,.y $cas./C nL fng.r ne. H k flu.4 a.+ Shu.hlown Alg , h!cere4 9 -
/, my4e99w.g-g q wa.yje. .
na ~ ~ = _ g-
+9 st ) !g - **
5 2 L .. 704stu MCSt3s
e i y . TABLE 3.3-1 [ Continued) ACTION STATEMENTS (Co.tinued) _ ACTI0t; 3 - With the .a"
.er of channels ORERABLE'one .less.tha.1 +ne Miniwr .
Channt is OPERABLE:requir62 nt and with the THERMAL PGk'? leve!i
- a. n.elow the P-6 (Intermediate Range Neutron F1'ux. Inter'.or M-Setpoint, restore the. inoperable. channel.to: OPERABLE
. status prior to increating-THERMAL POWERnaboveithe.P-6 Setpoint'.and ,
- b. Above the P-6 (Intermediate-Range Neutron-Flux-Interlock)
Setpoint but below.10%.of. RATED THERMAL POWER,. restore tne inoperable channel. toe 0PERA8LEJstatus prior..to increasing.- THERMAL POWER.ab6vCIO% cf .oann-THERMAL. POWER.C ' - -
"CT!;N d'#ViG cne number of OPERABLE channels one-less than the. Minimum Channels OPERABLE requirement, suspend all operations invhlv.ing.
' positive reactivity changes.
,. . . n ACTION 5~ ft
, .m_h___3.th: =ter-ef_
asenne, OPERABLE e __..:__--_-channels--one--less
. ___.___ . 4, .,
th=
,w3.
th;,.%,.
"inimum 3
Ee/4.c.e sr% ""'.".". .M. __ _.I. '_c' '~._~ !'._. ' .'. ?. !..C. 4 'T. . . _'.L . !. . _ ' _,_'".:. . Z. " .E. . .Z.. ' _ 'E. E
' insed /* '/f J-7
'y-t-- b-=h: :, #=d :!' -- ~ :tf e-- 4 ve!v n; petitive i
- ::tfvity : hen;:: =d ::rify V:?;;; Or: ?:::d =d i:: = r:d ': p::1tf = efth'- th: :=,t h:27 .
ACTION 6 - With the number of OPERABLE ' channels ~ one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceeo provided the- following conditions are satisfied: , l a. The inoperable channel is placed in the tripped condition i
& within g houts ,and f j ,
b; The Minimum Channels OPERA 8LE_ requirement is met; however, the inoperable channel may-be bypassed for up to 2 hours for surveillance testing of.other channels per r ; 4 q Specification 4.3.1~.1.- , .; ,
,- .m .:
AGHGN 7 - With th . e :r Of OPE"*"LE ch= = ?:'=: ?::: th= the Tet:! 6 "- ':r Of Ch=n ?:, ST""T'J." =d/:r ."0SE" 0."ER^ TION =y proceed s F e-ti! ---fe-- e of th: ~n=t -- f r;d ""^. LOC C:L"."" L 0.": RAT:0ML C I' p. =ided th: 1..:;;r:b ?: ch= .;? i: p! ::d in th; tripped
=rditf:- -ithia 1h-7 . . ~ - '
- , e. ' 2 ACTION K - With less than the Minimum Number of Channels OPERA 8LE, within 1 hour determine by observation of the associated permissive annunefeter window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
i . . . . . . . - ,,. .,
..a,. . ..n. - ..-:-' ~
~ - ~
7 , Insert.to,3/4 3-7 ,- , _ s
*. % n
, w -y ACTION 5 - a. Witn the nu'eber of.0PERABLE cha ,nels one less tt:a ed
;- . ; ,7 c,, ,- .
~~
..Mjnimum. channels .0PERABLE-requirements', restore
~
p thd (p#eble7, . L- t. 4/- ..
.. channels 5.to.0PERABLEMtatus-within .48 hours or o e.a ctor. . -u
'. ~ Trip'Breake~ r s and. suspend ~alF operati~ o d:involvin e. -
;. . - m. % . ., -> . e en.. . < reactivity _ charges' and verify vshrss.-1208-U4- U4-177,"v .
! .; .J,;'. > . . . d'. , ~, '.O.'. 1208-04-183 p arid 1208-U4-1-76 a re '.cl'osed .and. secu i'n pos.ition .. .s F. - 3 . * :. :. .
.:r-;.*-;c,
,... . within..the next hour. .
9
; . .,- . - ~ . .-
- b. 'Withino'.channelsr0PERABLE,1open~.the tor-Trip Breakers, -
r 4 -:t. p + . .: isuspend -all operations- in'volving posi
-tivity char'ges , - f
. .. . g ..w ..e.,; .; iand.yerify:compl fance with the2 5 HUT Crequirements'cf" '
^
m yr#3.; 6:1-4 ai.* fpSpecificationr3.hl:1c or.:.-3:lil.27.' ble', withini.1? h' corr "W-p _. ,,gg/c - n . .:. gpp.; r and' every;,12eboursithereafter.nY es/1208-04-175? -- . n - 1208-U4'-127WI208-U4-183;;andr. ;are-clo; sed.Eanst a:.e . . secured 4ihlp'ositioi wi thin-44
~
erif.iedite 6 closed and securetiin.positiorteve .c/r df ' 6
~
4' '7 y A 4.O . l l i. FA81M MCS136 1
l 3 0
- TABLE 3.3-1 (Contin.sep .
ACTICN STATEMENTS (Continued;' ACT'0F 3 - o!!n : dar e' 2:::'-4 red Mth a-ar r: tin; !cce i cre ;_'-e, rutw' tM i ge-e! ehr--d! *ze.0 25.^3LE ttitd 'ithi- 2 LM ' - e- be i- n. le Et ET ST'TSY uith'- th: .= :t S h = :. Onc. - .-- r...::::! ::_2 . d::Juh : ;;r: ting h ; - y 5: by;::::d fer. -
"- te 2 he_r: fer re- ei" :: t::ti.; ; r !;::ific::i '. '.2.1.1. -: ~
. 1 ACTION 147-' With tha. number of _0PERA8L'E{ channels one-less:.than the Minimum .
- Channels 0PERABLE'. requirement, :be ,in at:least HOT STANDBY ~' '
within 6-hours;chowever,: one . channel say.-be . bypassed for up .to 2 hours for.survei.11ance; testing per Specification 4.3.1;1, provided;_the oth~er.thannel is .0PERA8LE.". 7 7 - . i ACTION .:- Withithe number:of OPERA 8LE~:chinnelstonefilessathaotthe-Mifiimom R1 u * :- Channels OPERABLE (requirementprestore thaginaperaeTecchmwi-- - to OPERA 8LE stutes;withtnr48' hades erropen:tne Reactor Trip System breakerstwithin thernextthoura t-- 10 . ACTION y - With tdf r:mber of OPERABLE channels less than the Total Number of Channels, operation may continue provided the ii. operable channels are placed in the tripped condition within-1-hou3 il 6 s ACTION 13 - With one of the diverse trip features (undervoltage or snunt trip attachment) inoperable restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing mainte-nance to restore the breaker to OPERA 8LE status.
? ' k k,..
k.. .M [ e
. . (i? . ' 4; OM '
-. lJ$
Q . , &c! w.:
- _- lQm
,y ._#
e' * * ~U
. LJ g .e $ .
1 3/L0 APDLICABILITY BASES . The specifications of ;th;s section provide the general requirements apolica:;e to eacn of the Limiting Conditions for Ocaration.and Surveillance Recuirements witnin Section 3/4 In the event of a disagreement between tne requirements statac in .these . Technical Specifications and those stated in an applicaole : Federal Regui a-tion or Act, the requirements : stated in the applicable;Feceral. Regulation .or. Act shall take precedence-and;shall be-met . - 3.0.1 This specification defines th 4pplicability of'each specificatin- :n terms of defined OPERATIONAL MODFS or other specified: conditions-anc is proviced to delineate specifically.wnen sacn specification is.applicacle. . 3.0.2 This. specification defines those-conditions necessary to constitute ccm:'i-ance with the . terms .cf-an individual Limiting:Conditi.od.for Operationeanc associa ec ACTION requirement? 3 3.0.3 The csoecifdcation: delineates .the ra=<nres,to bentaken foc.those circum-stances not:di.rectly paov' des for in the ACTION statems;.tr ands whost occurrence woul:
,iolate ta= intent of a specification. For example, SpecificaJon 3c5:2 recuires two independent ECCS subsystems to be OPERABLE and provides explicit A 770N .ree...; .ms.-*-
one ECCS subsystem is inoperable. Under the requirements of Specificar*,n 3.0.3, if both the required ECCS subsystems are inoperab'e, =;thi- i nour m'asures mu=: be in't:- ated to place the unit in at least HOT ST.%10BY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Spray System is inoperable. Under the requirements of Specifica-tion 3.0.3, if both the required Containment Spray Systems are inoperable, within 1 hour measures must be initiated to place the unit in at least HOT STANDBY within the ne) 6 hours, in at least HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It i's acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that provided for operation..in subsequent lower OPERATION MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability
- is discovered but the times provided for achieving a mode reduction are not applicable if the inopsrability is discovered in a mode lower than the applicable mode. For exam-pie if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156. hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours,140 hours would still be available before the plant would be required to be in COLD SHUTDOWN However, if this system was discovered to be inoperable while in HOT STANOBY, the 6 hours provicec to achieve H0T STANDBY would not be additive to the time available to achieve COLD SHUTDOWN so that the total allowable time is. reduced from 156 hours to 150 hours.
3.0.4- This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with: (1) the full complement of required systems, equipment, or components CPERABLE and (2) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out-of-service provisions contained in the ACTION statements. The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being ' exceeded. Exceptions to this provision have oeen provided for a limited -u=cer of sceci cat ons wnen startuo witn incoerable ecuiement would not :ffect :lant safetv ~5:t tvccotions are stated in the ACTION statements of the acorcortate scec1ricatic" .
/CGTLE - UNIT 1 3 2/4 0-1
EMERGENCY CORE COOLING- SYSTEMS .. .- _-.- s,. . T@4
!- , , , . 3/4.S.2~ ECCS SUBSYSTEMS.- T avg GREATER THAN OR EOUAL TO 350*F ,. , , .r .:
y .: . !. ., . li
~
_h .- M MMt. ' 1' b.w,_:.[,,LIMITINGCON0ITION'FOROPERATION - -~~--w*-
/.c a *
*Q - .
(. ' y ** :<,.,. ._lfM Q %
,3.5.2 Two. independent Emergency,. Core Cooling Systest(ECCS) $e y satos shall be . <'
y o, a.e--m 40PERA8tE with5ea~ctsubsystes. comprised.of:- .-
.. .z . , :- , .
n - vgy v w.:: p uw.- : :c .. m r c.c~ c., ... . . 3
- a. . One.0PERABLE centr.tfugal. charging pump, . . &
- - ' - 4 v.
,fa ..
39:;n : .g := ,;. .n
.n . :: .. 4 .. .
gg m.a o .s.ep...:. pre:r $7.M'yne' OPERA 8d' Safety Irdeetiorupump. 'f:un.y,-Qh :Qg,, =,.i.n.
.g - :.. '- =i{ }} .
- c. One OPERABM1tHR*.: beat Mbanges.cae , '~ ~
- c. ;- y . ..
- d. One WPERA8tE JRHR9 ump, and ,,
- . sem o-An OPERABLE flow path capable of, s etion from/the refueling e.
water storage tank on a Safet iije~ . signal and transferring suction to the c tjsumpduring[ theautomatically recirculation phase of operation. eme rye.ncy APPLICA8ILITY: MODES 1, 2, and 3. . ACTION: gv
- a. With one ECC5 subs erable, restore the inoperable subsystem to OPERABLE statt thLp 72 hours or be in at least HOT STANDBY within the nex Vand in HOT SHUTDOWN within the following 6 hours.
- b. In the av is actuated and injects water into the Reactor Coolant a Special Report shall be prepared and submitted to the C ursuant to Specification 6.9.2 within 90 days describ-ing t tances of the actuation and the total accumulated act ,es to date. The current value of the usage factor fo acted Safety Injection nozzle shall be provided in this
? Spec ort whenever its value exceeds 0.70.
O rh e pro vo s, Sp ee?/r*co. sons 3. o . + a.nd 4 . o .
- a r e. no t app l ica.e for en te ads Jfor **A e Sa/e9 L,. e e. tion pwp.s d e e ta s e d-u a n t to Spe< i//ca tion 3. s. s. 2 pro <idest ?A e fa. /e ev on.p e sa Sje e t.l p.c a.rc. re.s to eed. to o P G /? A 8 t 6* .1 h d a.1 wiH' n ' I'ea e:
or p 'o be- he fe mp e rm.fa re ef one o r n e re. c/ w. a. Acs ca/4 /eys e sc Af] 75 *F, aJricA
- ve r o c c u r.r h .s f 1
l 1 ZL: - : C '. - :-? w m. ucs .
l*
- p. .
EMERGENCY CORE COOLING SYSTEMS.__ . .. _
;i b?El?%, x
-c-~..- -- 3/4. 5. 3 ECCS. SUBSYSTEMS T avg LESS THAN 350*F ~ :=- :- ..:- ..-. ~'.. ,, Acc.s sch.s'&7FM 4,y^r . .~ > . .
~ -
j
. . . . V - 4 :. . ny
. , :. , j , .,.._ LIMITING CONDITION FOR'OPERATIONf- ir
..+.
4Gk A,, e- -
'a;. :
- ~1. .C". t j; %~ i.
. ,/," . - . , . .: .
- P: . We' 3. 5.3.1 As ::a' minimum,:.one ICCS.; subsystem .cc:::pcised .ofithe fo stiaR be . . . , ,
OPERABLE,g,,j., . , .g. . ,. , , . . . .
. . . .. ,: . . r ,
P -.-- a.10ne-CPERASd7centd{figih chaMihgiguay,% - _.+.
, h. Onej0PERA8LE.'.RHR. heat-exchanger,;.:.x v f - ,3: 7 g g.- e. . v. . . - ... , .
- a. .
c: One' OPERABLE RHR gumpabarut:. r -
- d. An OPERABLE flow path"capab.he of. i.os fr.ca the.&sfue.Ung water storage tank upon b'eing ma ligned a'nd transferring suction to the containment3s the recirculation phase of operation. e-< r e-APPLICA8ILITY: MODE 4.
ACTION: i
- a. With no ECCS subsys because of the inoperability of either the centrif ng pump or the flow path from the refueling water s k, restore at least one ECCS subsystem to OPERABLE status hour or be in COLD SHUTDOWN within the next 20 hours.
- b. With no EC OPERABLE because of the inoperability of either t heat removal heat exchanger or RHR pump. restore at less CS subsystem to OPERABLE status or maintain the Reac-tor Co em T9 less than 350*F by use of alternate heat i remo .
- c. .I the ECCS is actuated and injects water into the Reactor t[ Coo tem, a Special Report shall be prepared and submitted to C ssion pursuant to Specification 6.9.2 within 90 days describ.-
the circumstances of the actuation and the total accumulated tion cycles to date. The current value of the usage factor for affected Safety Injection nozzle shall be provided in this ial Report whenever its value exceeds 0.70. A O
=m = ee;f:.~- Z: ;
w '_;: =C" 41 M F "".If =~=% 7:::::r 2: n ::r:tr: ;f;r;
. ' . . . '9:r ;r ., .. .: .' 7 r..;. -- -
CT *_ '. ?!!T . 2.4 id '- I wem w.sw T-T . - ' - = . - - - - - ,7,, - , . , , , _ . . . . - - - - . - - - , - - - - . ____..__ . . _ _ _ _ . . _ _ _ _ , _ _ _ _ . _ _ _ , _ _ _ , _ _ _ _ _ _ _ _ _ _ ___
r v . , EMER3ENCY CCRE COOLING SYSTEMS
$URVEILLANCE REOUIREMENTS .
1 1 l 4.5.3il -The ECOS T subsystem: shall ha- J4nonstnted OPERABLE: Der:tbe. appl.icable-requirements _ of Specification ?.s.2.
- 3 a as, . ..
-.-g. u.,,. ,..... ~_.
e,,,+~, w ,,,...
., _-. .. , _ .. ~_.,. m_. . _.._ __
_ x_ _ . , ..
'a
=9 St a e dad % # *b* 1 hA
-- M_. - - .-sk-# $,We$4 s ad .m- a- sa. .a --= * - . . =
. .m.E I,* E
&=^
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. . n 3 4.5 ELECTRICAL. POWER SYSTEMS -
345.1 A.C. SOURCES l OPERATING- - I L'"ITING CONDITION FOR GCERATION 3.8.1.1 As. a minimus, - the' following. A: C. . elec.tricalc power. sources shall . be OPERABLE: a iwo physically independent circuits be_ tween the offsite transmi". ion. . netwark ad. the onsite Class:1E Distribution. System, and. -
- b. Two. separate. and .-independent- diesel generators;. each withm
- 1) 5:p:r:t: Aay :nd :ngin;.;nnt;d f ;l= tank [containing a minimum volume of: 75o . gallons of' fuel, - --
- 2) A s6paratc,Fueia5turage Systenucontaining sawmini-rw-:1ume of w". oco . gallons of' fuel,n c s .. ~ .
M A separate" fuel transTer pump,
') ' d-icating ci' :t:r:g: cent:i 4 g :
44 rr ::t:1 l=: Of gall:n: ;f i d ri;: ting ;il, :nd
- 5) 0;pcility t tr:n:f;r i d ri:: ting ;il fr;; :: r:g: t; the
~
di;; 1 ;;;;r;t:r mit. APPLICA8ILITY: MODES 1, 2, 3, and 4. ACTION: 4 j 5
- a. With one offsite circuit of thd above-required A C. electrical power-sources inoperable, demonstrafe the OPERABILITY of the remaining A.C. sources by performing Sirveillance Requirement 4.8.1.1.1.a within 1 hour and at least o'nce per 8. hours thereafter. If either diesel generator has not bien successfully tested within the past ur>M 7/>c# 24 hours,demonstrateitsAPERABILITYbyper,formingSurveillance dekse/ eeric/ arse Requirements 4.8.1.1.2.a.4-and 4.8.1.1.2.a.-t for each such diesel d
/s 4/fq(dy , generator, separately, within 24 hours. Restore the offsite circuit ey e. w ,ny ' to-OPERABLE status within 72 hours or be in at least HOT ':"U':0'.3 within the next y hours and in COLD SHUTDOWN within the fol)hwing l
$ hours. 4
\
y, STAND 8y
- b. WitheitherdieselgeneratorinoperablefdemonstratetheOPERABILITY of the above required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. If the diesel generator became inoperable due to any cause other than preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel gener-ator by performing Surveillance Requirements 4.8.1.1.2.a.+ and 4
.SA dir*:rd/ oc'ner.1/Dr* $/saff be can.:rsdt*r?./. ?o (p_ ir.ggra.b$*. Sw Q -.!,m -l h e'!a re. . +i! ll SL!iS/IES Ot r!,.7 airer,,er,lt 3 0 r va io'r$1 m k o n i rt % e n t-
- nqTLi - m!~ _. 11 : i :.
, 3
,,,, ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION.- .
. :4 1 ACTION (Continued)
., Y
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' 4'. 8'.'l. l'. '2.~a.' within P24E houis*. ~ Resto'rehthe iYnipeFabae: ddhEL
~
Y.- generato.r- to OPERA 8tE'statusTwithin. 72--hoursnos beein: atd 180 7 s '
^
MAA/DBy 0!:LZZ within 'the :nextRbF houVi '
'andrfrn'.COLU: SH he-following p hoursi- x ' /, . ~
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c, ,
!. '< .- 1' c.; With one< dieseligeneratorinoperaMenirr addition- 4. o r -b . . .< -
above, verify. that:"r '
- t. . --. .-....-..; ,.
- 1. All. required ' systems. . .., subsystems,: tratns nentse,-and devices c :
. - .+ - -r ?! u- e u - .; o n ethat depend on the remaining- OPERABi.E- rator.as ajc. x.cn v. ., ' *~
source af:emergenc h ower::ene alsos m . . :. .
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- pump is OPERABLEr. c. ... -7 .-
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If these conditions are not satisf 'n 2 hours be ir at least.:.
~
HOT STANOBY within the next 6 hou d COLD SHUTDOWN whh'irrrthi' following 30 hours. 4
- d. Withtwooftheahoverequire .C.circuitsinope[able, demonstrate t OPERABILITY of sel generators sepa tely by performing t requirement . ification 4.8.1.1.2.a. and.
4.8.1.1.2.a. within I h: ? _ __ !:::t ere per 8 hours'thereafter, unless the diesel gener already operating; restore at least one of the inoperable # ources to OPERABLE status within 24
, hours or be in at lea # 1G L $TANDBY within the next 6 hours.+ Wi4
~
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' :nly :n; Of f:it: :d q ;: h d, re:ter: :t !:::t tre offrite l ;ir;uit: t: 0 " E " "".' fr ; tim: Of #niti:1 lo::
Or 5: in :t 1:::t'"* N. O?NDSY ; within within 72th;h:n;xt r: C h ur; ar.d in COLO 0"LTOO'" with ' " fNlwing2Ch:r:. .
+ '
- e. With two of required diesel generators inoperable, demonstrate the OPERABI f two offsite A.C. circuits by performing the require-ments of tion 4.8.1.1.la. within 1 hour and at least once per 8 hours r; restore at least one of the inoperable diesel gener ERABLE status within 2 hours or be in at least HOT STAN n the next 6 hours and in COLD SHUTDOWN within the
, fall hours.+.";;ter: t 1:::t tw di:::1 generstcr: t:
Zm~ r # .5 ' Orr2"0LC n eta: witnir. 72 h ur: fr:: tim Of initi:1 10:: Or b: in "0T ST?NDBY uith'r the next 5 h: r: and " COLD SHL7009" ithin [ . m %g 30 heu-c. e
~
SURVE EQUIREMENTS 4 Each of the above required independent circuits between the offsite t mission network and the'Onsite Class 1E Distribution System shall be: xy
'Inis est is requirec to be ::moletac egardless of when t.*e %:cer3Dla nesei generator i: restarec :: PERASILITY.
.u.a .a... .
70Gump MOESBS
1
- e , ,
Insert 1.
- c. With one offsite circuit and one diesel generator of the above' required A.C. electrical power sources. inoperable.,. demonstrate the OPERABILITY- ofc the remairting A.C. cffsite source- ny per-forming -
Surveillance Requirement 4.8.:1.1.2.a.4 witttin 1. hour and .at least once
. per 8 hours thereafter, and, if the. diesel generator-becam ;noperatie .
due to any cause other than preplanned preventative T.aintenance or testing,- demonstrate the- OPERABILITY of. the remainino OPERABI E Jiesei generator by performing Surveillance Requirement J.8.l.1.2.a.5 u tnin - 8 hours,.unless. C,o OPERABLE diese.1 generator.is already operating. Restore at least. one of the inocerable sources..to 0PERABLE ~ status within 12 hours: or be in: atsleast- HOT. STANDBY witnin the next .6 hours- ,
. and in COLD SHUID0hi'thirt..the'following' 30 hours.. Restore the other A c- m e e m e (:' %ite circu{t',.or diese.1 generator) to nCAdLE status in accordance wit?. % provisions of 3.8.1.lc Acticr; -
Statement a or h. as appropriate, with t*. time ~ mu+rement of that AG.iwa atatement based on the time of initiai ioss' of tw rena.irina - inoperable A.C. power source. A successful test' f%i.esel: ow .-rator OPERABILITY per Surveillance Requirement 4.8.1.1.2.a.E yerformec under the Action Statement for an OPERABLE diesel generator or a restored to OPERABLE diesel generator satisfies the. diesel generator test requirement of Action Statement a or b. g .
~
Insert 2 .
~
x .. . a Following restoration of one offsite source, follow Action Statement a with the time requirement of that Action Statement based on"the time of initial loss :of the remaining. inoperable offsite a.c. circuit. A successful test (s):of; diesel OPERABILITY,per-Surveillance Requirement 4.8.1.1.2.a.5 performed:under this Action Statement for the OPERABLE diesels satisfies the diesel generator. test-requirement for Action r Statement a. w i .; 1 % . .. u
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- Insert ' 3 A %- - - .
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Following restoration of one diesel generator unit, follow Action Statement b with the time requirement of that Action Statement based.on.the time of l initial loss of-the-remaining inoperable diesel generator. A successful test of diesel OPERABILITY ~per Surveillance' Requirement ~ 4.8.1^.1.~2.a.5 performed underuthis Action, Statement for a4 restored to OPERABLE diesel satisfies-the diesel generator test requirements of> Action Statement b. l I l s. l
r 1 t s e .
'TABL'E 4.8-1 . _ _ . . ,
3
- g. . DIESEL GENERATOR-TEST SCHEDULE' --. ..'- ..
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- Criteria for determini f failures and ' number of valid tests shall be in accordance with ry Position C.2.e of Regulatory Guide 1.108, but determined on a generator basis, l For the purposes, raining the required test frequency, the previous l test failure c e reduced to zero if a complete diese1' overhaul to like-new cond ompleted.,provided that the overhau17, including appro-priate post e operat W and' testing, is specifically approved by v the manufa if acceptable reliability has been demonstrated. The reliabili . on shall be the successful completion of 14 consecutive tests 1 -si ~ series. Ten of these tests shall be in accordance with the routi 111ance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 and four tes nce with the .184-day testing requirement of Surveillance 4f.u..z./- 0.0.1.1.2.e.; e.4 4.0.1.1.2.e.0. If this criterion is not ing the first ' series of tests, any alternate criterion to be transvalue the failure count to zero requires NRC approval.
.: -. r~_r7 -- -
l # sociated test frequency shall be maintained until seven consecutive ailure free demands have been performed and the number of failures in'the
-u st 20 valid demands, has been reduced to one.r m
~
l i l 1037LE - UNIT 1 2/* 3-6 ,, I Hb WtN4%
~ . * . o ELECTRICAL POWER SYSTEMS .... . . ..
a.C. SCURCES ... . . SHUTOOWN
' IMITING CONDITION FOR~.0PERATION -
s 3.8.1.2 As a minimum, the following AA : electrical power sources. shah -be OPERANT.E: _
't One circuit between.the :offsi.te transmission network ar.d the Onsi: e a.
Class 1E Distribution . System, 'and -
- b. One diesel. generator wirth:- ,
A fu.e/ ky. -v .. . . . . j
- 1) Dg :ad q;1 : c=t:d fechtanks <antainine a' cf..tw= volume i. .
of zgg, .ga1Aons tatt 'tue1."' a
?) A #ael storage :/:te: centaining a nonimum volume of u,ooo gallons of fael,
- 3) A fuel transfer pump, ,
O '. tri:: ting ri' :t: :; :::t:f 4 ; : =i derr t t:? /e!ure of g:11:n: ef lubrie ting oil, cad "J Oe,,et!;ity m__, 7.e trea.;fer ?;triceting cil fr;; :::r:;: :: th:
...........u. 7 ...
..4.._.
APPLICABILITY: MODES 5 and 6. in n nLease. witA spectf,* ann 3.* 4. 3. ACTION: a 3 With less than the above :h;n; .c required A.C. electrical power sources
- OPERABLE, immediately sus;,end al l_ operations involving. CORE ALTERATIONS,
; positive reactivity, changes, movement of irradiated fuel,ior crane operation
! with loads over. the fuel storage pool, and within 8 hours, d:gr:::u-f: :nd vent the Reactor Coolant' System thr:;;h : gr::t:r th:n :r :; :1 :: :qu -e ( In addition, when in MODE 5 with the reactor coolant loops not inch ::t. l filled, or in MODE 6 with the water level less than 23 feet above the reactor l vessel flange, immediately initiate corrective action to restore the required sources to OPERA 8LE status as soon as possible. . . g
~. : .. s . .. _ prodde reRef kapahr//fy Arr-03 SURVEILLANCE REQUIREMENTS' 4.8.1.2 TheaboverequiredA.C.. ele'ctricalpowersouhces~shallbedemonstrated OPERABLE by the performance of each of the requirements of Specifications j 4.8.1.1.1,4.8.1.1.2(exceptforSpecification4.S.1.1.2a.f))/:nd*.*.112.
S I , l 1037'..:. .mIT ; .. - .--
.psba me49 WestTED STATES M N8I
, u 'g NUCLEAR REIULATORY COMMISSION 8' "k REGONN
! l 101 IIAlWETTA STREET N.W., SUITE 2000 ATLANTA, GEORGIA 30333
% 9 V0GTLE SRO ANSWER KEY 1
I 1 I e l l 1 1
)
1 T, , 1
I
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 29 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 5.01 (2.00)
S;e figure 947. (.25 for each component correctly drawn and labeled and .5 for correct overall shape). REFERENCE VG, R0-LP-051.006; VG, APNPP, Vol 3, p 2-81. ANSWER 5.02 (1.00) - (c) REFERENCE OC, FNRE, pp 115-120; CR, NETRO, p 12.1-4; VG, APNPP, Vol 2, p 5-8. ANSWER 5.03 (1.00) (b) REFERENCE CR, OP-210, Rev 16, p 8; NUS, NETRO, Unit 6; W, RP, sect 385; VG, APNPP, Vol 2, p 5-17. ANSWER 5.04 (1.00) (a) REFERENCE NUS, NETRO, p 10.3-2; VG, APNPP, Vol 2, pp 5-55, 4-140. ANSWER 5.05 (1.00) (d) REFERENCE NUS, NETPP, p 6.2-5; VG, APNPP, Vol 3, p 2-234. r
r
. .i
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 30 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 5.06 (1.00)
(a) REFERENCE CR, ANAO-29, p 2; VG, APNPP, Vol 4, pp 1-50',67. ANSWER 5.07 (1.00)
- a. Provides density compensation to account for changing steamline pressure
- b. Higher.
REFERENCE VG, SR-LP-070; VG, CH 13b-9. ANSWER 5.08 (1.00) increase, increase REFERENCE VG, SR-LP-070. ANSWER 5.09 (2.00)
- a. The moderator temperature coefficient is within its analyzed temperature range.
- b. The trip instrumentation is within its normal operating range.
The pressurizer is capable of being in an operable status with a steam e. bubble.
- d. The reactor vessel is above its minimum RTndt temperature.
REFERENCE VG, TS, p 3/4 1-2; VG, SR-LP-124, R0-LP-012. ANSWER 5.10 (1.00) Machine frequency lower than bus frequency, phases not matched.[vobedikd18 jk. @ .W kdk& Cscf j L O'clocN} r.
. '1'
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 31
' ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE CR, Power system ops., pp 22-24; VG, SL-LP-128,NL-LP-042; VG, APNPP, Vol 4.
ANSWER 5.11 (1.00)
- a. higher than normal
- b. higher than normal REFERENCE -
VG, SR-LP-003. ANSWER 5 12 (1.00)
- a. a flaw such as a crack
- b. a stress,(usually above 5000 to 8000 psi, of sufficient intensity to develop a small deformation at the crack tip, and,)
- c. A temperature low enough to promote brittle fracture. (2 0 0.5 each)
REFERENCE VG, .APNPP, Vol 3, p 2-180. ANSWER 5.13 (1.50)
- a. Control rods in a single group move together with no individual rod insertion differing by more than +/- 12 steps, indicated, from the group demand position.
- b. Control rod banks are sequenced with overlappinS 3roups.
- c. The c'ontrol rod insertion limits are maintained.
REFERENCE ' VG, TS, p B3/4 2-4. ANSWER 5.14 (2.00)
- a. True
- b. True
- c. True
- d. False r
J
F
- 5. -THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 32 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VG, WOG/ ERG 0069V, pp 6,6,8,10; CRr B&W analyses, 8/24/79.
ANSWER 5.15 (1.00)
- a. Decrease
- b. Decrease
- c. Decrease
- d. Increase *
- e. Decrease REFERENCE VG, R0-LP-051.001; VG, APNPP, Vol 3, p 2-75.
ANSWER 5.16 g (1.50)
- a. Quality- 99.#1. +/ .25
- b. Entropy- 1.39g +/ .01
- c. Superheat- 0 REFERENCE VG, APNPPr Vol 3, p 2-45; VG, R0-LP-046.007.
ANSWER 5.17 (1.00) The blowout coil produces a magnetic field which opposes the arc (0.5), causing the are to be drawn into the are chutes (0.5). REFERENCE VG, NUS electrical maintenance text. r
f E' O
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND FAGE 33 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 5.10 (2.00)
- a. DNDR decreases. The coolant is closer to saturation and is more likely to experience DNB.
- b. DNDR increases. If the pressurizer temperature increases, pressure in the core increases and the saturation temperature increases. The coolant is less likely to experience DNB.
- c. DNDR decreases. A higher Th-Tc is required to produce the same power, so Th is greater. The coolant in the top portion of the core is closer to saturation and DND is more likely. Also the coolant flow is less effective in scrubbing bubbles from fuel rod surfaces, making partial film boiling more likely.
- d. DNDR increases. Lower steam flow implies reduced power so Th decreases, makins DNB less likely. Also, the heat flux at each point in the reactor decreases.
REFERENCE VG, SR-LP-118/485; VG, APNPP, Vol 3, p 164. ANSWER 5.19 (1.00)
- a. The motor current will be (substantially) less,
- b. The discharse pressure will be hisher.
REFERENCE NUS, NETPP, p 6.4-5; VG, APNPP, Vol 3, p 2-236. ANSWER 5.20 ( .50) Samarium. REFERENCE VG, APNPP, Vol 2, p 4-149. ANSWER 5.21 (4 .f7 *O
- a. ~ m - , , r 4.-
- b. Bank D will be withdrawn less.
- c. Bank D will be withdrawn less.
m, w .y t J
t
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 34 ANSWERS -- UOGTLE 1 -86/06/23-TOM ROGERS
^ REFERENCE CR, OP-210&105; VG, OP-14940-1. ANSWER 5.22 (1.00) (ft-v,
)
MTC is at its most negative at EOL (giving rise to the most restrictive postulated steam line break accident and resulting uncontrolled RCS cooldown.} REFERENCE . VG, TS, p B3/4 1-1. ANSWER 5.23 (1.00) Neutrons do not carry a net electric charge. Neutron detection must depend
. upon their interaction with target nuclei.
REFERENCE VG, APNPP, Vol 4, p 4-32. ANSWER 5.24 (1.00) To minimize evolution of iodine [0.53 while minimizing (Cl and caustic) corrision on systems and components CO.5]. REFERENCE VEGP, TS Bases 3/4.6.2.3. ANSWER 5.25 (1.00) See attached answer sheet 962 REFERENCE OC, FNRE, p 124-128; VG, R0-LP-021.
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 35 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.01 (1.00) b REFERENCE VEGP, S-DG-LP-1.
ANSWER 6.02 (1.00) (c) REFERENCE VG, SR-LP-045, pp 16,17. ANSWER 6.03 (1.00) (b) REFERENCE VG, R0-TP-217-005. ANSWER 6.04 (1 00) (d) REFERENCE VG, SR-LP-124, pp 9,14. ANSWER 6.05 ( .50) FALSE (DS of sump) REFERENCE VEGP, R0-LP-139, Fig. 10e-2. t ) i I i a (,
. _ - - - _ , . . . . - = , , -_ ,. __ __ _ _. , - .__.._.. _ _ ,. _ _ _ __._ m _ . _,, _ _ , _ _ . , . _ . , _ _ . _ _ . , _ , . _ . . , , , _ _ _ _ . . _ , . , _ . . .
F 1 B
, y a .
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 36 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.06 (1.00) 0 0.5 points each.
(a) + ,-4 steps. (b) +10,-4 steps. REFERENCE VEGP, R0-LP-138, p.7. ANSWER 6.07 (1.00) . 9 0.5 points each.
- 1. High radiation (at demin. outlet). 6 C"J A' 8'##
- 2. High temperature. M'6M FLoud r REFERENCE VEPG, R-SGBD-LP-1.
ANSWER 6.08 (1.00) From main steam through a bypass valve of the 42 MSV and into the steam chest. REFERENCE VEGP, R-TURB-LP-3. ANSWER 6.09 ( .50) 2 75 hours. (+,- .25) ANSWER 6.10 ( .50) The inverter. REFERENCE VEGP, S-ELECT-LP-161. r*
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 37-ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.11 ( .50) 125 VDC ESF Battery Bus.
REFERENCE VEGP, S-DG-LP-1. ANSWER 6 12 ( .50) 5 REFERENCE VEGP, S-DG-LP-1. ANSWER 6.13 (1.00) (a) output voltage EO.53. (b) reactive load E0.53. REFERENCE VEGP, R-GEN-LP-2. ANSWER 6.14 (1.50)
- 1. Pressurizer pressure low pressure safety injection.
- 2. Steamline low pressure safety injection.
- 3. Steamline isolation.
0 0.5 points each. REFERENCE VG, R0-LP-213, p 12) VG, R0-LP-215. ANSWER 6.15 (1.00)
- a. Control board handswitch must be in the ARM position.
- b. PORV block valve REFERENCE VG, SR-LP-125, p 6.
I.
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 38 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.16 (1.00)
- e. Upper vessel head penetration (head vent); bottom of vessel (incore conduit) (0.5)
- b. 100% (0.25)
- c. Increase in void content in vessel (0.25)
REFERENCE VG, SR-LP-121 ANSWER 6.17 (2.00) Any 4 0 0.5 points each. n6
- 1. Step counters.
- 2. Master cycler counters.
- 3. Slave cycler counters.
- 4. Bank onerlap counters.
- 5. Internal memory a alarm circuits.
- 6. P/A converters.
REFERENCE VEPG, R0-LP-132, p.6. ANSWER 6.18 (1.50) 0 0.5 points each.
- 1. Auct. high delta T.
- 2. Auct. high Tave.
- 3. P/A Bank Position.
REFERENCE VEGP, R0-LP-132-012. ANSWER 6.19 (1.00) Any 4 0 0.15 points each. ,,
- 1. Limit cooldown rate during main steam line break.
- 2. Minimum pressure loss during normal design full flow.
- 3. Reduce thrust force in the event of a pipe failure.
- 4. Reduce the mass flow the MSIVs must resist during a MSLD.
- 5. Permit D/P suitable for measurement of steam flow.
G If3m( Me* CC!tSS b Y' f tlus4 f/fE5Svifti /Ofp s Ac///SJ % IA&fuG% T. &Dsme e,cc (
- 9. k:t- xs now rs gcc, .g a w r
A
,- e
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 39 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VEGP, R0-LP-142.
_ _a r"'"@r ANSWER 6.20 (1.00) b. , _
- 1. 4/4 HGV shut (limit switches).[0.53 $2 A65 7 IO o bh
- 2. 2/3 ETS PT (<900 psis).EO.53 REFERENCE VEGP, R-TURB-LP-3.
ANSWER 6.21 (1.00)
- 1. overtemperature deltaT within 3% of trip setpoint.
- 2. overpower deltaT within 3% of trip setpoint.
- 3. circulating water pump trip with plant load above 50%.
(0.33 each - both signal and setpoint must be correct) REFERENCE Or b6 r redib. m. VG, 18012, p 1 ANSWER 6.22 (1.00)
- 1. Safety injection signal. 1/Y(N/A
- 2. Lo-lo SG water level in 1/4 SGs (2/4 0 =or< 17%)
- 3. Both main feedwater pumps tripped. 2/2, N/A
- 4. Blackout on 1-E 4160v bus (1AA02, 1BA03)
(Signal, setpoint and coincidence required for any credit) REFERENCE VG, SR-LP-056, Fig 13D-2. ANSWER 6.23 (1.50)
- 1. manual 1/2
- 2. SI 1/2
- 3. Containment area radiation 1/2 (both signal and logic required for any credit.)
REFERENCE VG, R0-LP-097, p 5. ra
1
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 40 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.24 (1.00)
- 1. CB control room HVAC
- 2. CB ESF electrical equipment room HVAC
- 3. Aux building ESF room coolers
- 4. CB cable spreading rooms HVAC
- 5. Piping penetration filter exhaust system. (5 9 0.2 each)
REFERENCE VG, SR-LP-051, p 3. ANSWER 6.25 (1.50)
- 1. WR SG 1evel (all loops)
- 2. SR neutron flux
- 3. IEr neutron flux - 66@ fed
- 4. Tc (all loops)
- 5. WR RCS pressure
- 6. Pzr level
- 7. Core exit thermocouples (calculated) ,-
- 8. N-vuron itun 'i ediacion eleme.-,1) Ivw e. SiW lh (6 9 0.25 each)
REFERENCE VG, SR-LP-122, p 4. ANSWER 6.26 (2.00)
- 1. Low pressurizer pressure
- 2. Manual
- 3. Low steam line pressure
- 4. High (1) containment pressure REFERENCE VG, SR-LP-030, p 8.
ANSWER 6.27 ( .50)
- a. System is armed by C-7 but dumps will not-open.
REFERENCE VG, SR-LP-071, p 8. r e
- 6. PLANT SYSTEMG DESIGN, CONTROL, AND INSTRUMENTATION PAGE 41 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 6.28 (1.00)
Boration- Maintains 140 degF into deminsi regulates TV-381A and TV-3818. Dilution - Maintains 50 desF Hx outlet temperaturei regulates TV-386 to control chiller flow through shell side of LTDN chiller Hx; also regulates FV-378 in Hx bypass line to maintain chiller flow constant. REFERENCE VG, SR-LP-022, p 10. ANSWER 6.29 (1.00) (a) I (b) v 0 0.5 points each. REFERENCE VEGP, R0-LP-185. i J l i 4 J l
)
t e f y _-.y- s- , . , _ , , - - . . - . - , - - - , , -
, ,-,-.,m~,w--m-__.~.3. -,,__,.. _. -,,,m -m.-,____.. 7 ,,.. _, , - . . , . . _ - . ,
f
-7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 42
---------~~~~~------~~~-
'~~~~Rd6 UL6656dL"66NTR6t ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 7.01 ( .50)
When voltase is less than 124 v . [f o, - L) REFERENCE VEPG, 13405-1, p.2. ANSWER 7.02 ( .50)
>134V.
/16 wf REFERENCE VEGP, 13405-1.
ANSWER 7.03 (1.50) Any 3 0 0.5 points each.
- 1. SG level in NR (prevents steam in FW bypass)
- 2. Maintain forward flow through bypass feed nozzle.
- 3. Maintain continuous flow in lieu of intermittent flow.
- 4. Refill piping at low flow rates (S6'spm) if steam or hot water has entered the piping. 'I T g ot' Nan 49trog ni ae % pojew y s. c pna.vsrs, R$FERENCE VEPG, 13610-1, p.2.
ANSWER 7.04 (1.00) b REFERENCE VEGP, 00920-C, p.6. ANSWER 7.05 (1.00) d REFERENCE VEPG, E0P Training Text, 2-12. r
~ f
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 43
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R56 ULUU 6 E"66UTRUL ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 7.06 (1.00) b REFERENCE VEPG, E0P Training Text, 2-12.
ANSWER 7.07 (1.00) c REFERENCE 10 CFR 20.101 ANSWER 7.08 (1.00) d REFERENCE 10 CFR 20. ANSWER 7.09 ( .50) FALSE. - REFERENCE VECP, 00920-C, p.3. ANSWER 7.10 (1.00) Alerts appropriate personnel to situations that have the potential for causing a violation of applicable exposure guides or VEGP policy. REFERENCE VEGP, 00920-C, p.5. e
. / < .
- 7. PROCEDURES - NORMAL, ADNORMAL,. EMERGENCY AND PAGE 44
--- EE5i5C6EiEEE E6 sis 5G---~~~~----------------
___________________.. N, ANSWERS -- V0GTLE 1 s.
-84f06723-TOM ROGERS ANSWER 7.11 (1.00) i To mark items where sequence'is not important.
s _ REFERENCE s VEGP, EOP Training Textsi 10-7. 't - 3's .s t ANSWER 7.12 ( 50) , FALSE (CST Trees do.) REFERENCE c-VEPG, E0P Training Textr SA-4. ANSWER 7.13 (
.. tf 0 ) 3 No. .,
REFERENCE .t , VEGP, E0P Training Text,,10-7., ANSWER 7.14 ( 1. 0 0 )' (a) No.CO.53 ' (b) No.[0.53 *' REFERENCE VEPG, E0P Training Textr 10A-8. , ANSWER 7 15 ( .50) 50 psid. l REFERENCE VEPG, 12002-1, p.19. s 4 e 9
4 P
. s
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 45
- - - - - - - - - - - - - - - - ~ ' - - - - - - -
~~~~R565UL665CEL E6 DIE 6L ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS 1
ANSWER 7.16 (1.00)
- 1. Following a refueling outase.[0.52 1
- 2. When the actual critical rod position is outside the ECC calculated window.EO.53 REFERENCE VEPG, 12003-1.
ANSWER 7.17 (1.00) 20-257. pouer (232-290 Mwe). [G dd 'IQ.<tstrm it (GTs'Ne Mutt ~pWMJ frowuMs) REFERENCE VEPG, 12004-1, p.9. ANSWER 7.18 (1.00) Dspress both EMERGENCY STOP pushbuttons to trip the A diesel generator. REFERENCE 18031-1, p.2. ANSWER 7.19 ( .50) Shutdown Panel D. REFERENCE VEGP, 18038-1, p.4. ANSWER 7.20 (1.50) Any 3 9 0.5 points each
- 1. Proper group alignment.
- 2. DRPI and group demand in agreement.
- 3. Proper bank overlap.
- 4. Count rate stabilization.
1 l l l
O .
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 46
~~~~ ------------------------
RA6Ebl651EEE E5sTR6L ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VEPG, 12003-1, p.9. ANSWER 7.21 (1.50) Any 6 0 0.25 points each. 1.
- 2. .Trip Trip the the turbine.
reactor. ,
- 3. Place BTRS to DF F. .
- 4. Align VCT for AUTO.
- 5. Set the SGARVs for 1085 psis in AUTO.
- 6. Place Steam Dumps in PRESS mode at 1085 psis.
- 7. Place heaters and sprays in AUTO.
- 8. Place PZR Level in AUTO.
REFERENCE 18030-1, pp. 2-3. ANSWER 7.22 (2.00) 0 0.5 points each.
- 1. Check for turbine runback in progress.
- 2. Place Bank Selector in MANUAL.
- 3. Place IN-il0LD-0UT switch in fl0LD.
- 4. Check to see if rod stoPFed.
REFERENCE VEPG, 18003-1, p.8. ANSWER 7.23 (1.50)
- 1. Trip & isolate (or erify) the RB Purge System E0.53. h]gk
- 2. Initiate the SF Filtered E:<haust System CO.53.
- 3. Notify appro iate personnel (SRO FH, SS) CO.53.
REFERENEE VEGP, AOP o 1 ~%i
,-n ,ww.,+ - -_,-------m-- - - - - - - - -
,n-w,,, + rv -
s- .
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 47
~~~~ ~
RA6EUL655EAL UUUTRUL'~~~~~~~~~~~~~~~~~~~~~~~ ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 7.24 (1.00) Only when it is reiterated in the transitioned EOP. . REFERENCE VEGP, E0P Training Text, 10-7.
#( #' } [ if '
o r, tQ w dd ANSWER 7.25 (1.00) (a) When a cooldown rate of greater than 50F E0.f53 is desired and RVLIS is available EO.253. (b) When a cooldown rate of greater than 50F CO.253 is desired and RVLIS is not available E0.253. REFERENCE VEPGr E0P Training Text. 4-11. ANSWER 7.26 (1.00) Check that no Red Path exists E0.53. Check that no higher priority Oranse Path exists E0.53. REFERENCE VEPGr EOP Training Textr 10A-7. ANSWER 7.27 (1.00) To prevent thermal lockup of the rods. REFERENCE VEGP, 12002-1, p.1. ANSWER 7.28 (1.00) Turn on the backup heaters to cause spray flow. REFERENCE VEGP, 12002-1, p.2. S
- 7. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND .PAGE 40
~
2
~~~~RED 6LU5iEAL E6UTRUL'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- V0GTLE 1 -06/06/23-TOM ROGERS 1 ANSWER 7.29 (1.00) . Turn on backup heaters and place letdown orifaces in service to allow noximum letdown flow rate. E0.53 + Raise VCT sas purse flow rate to Gaseous 11asteEO.53. REFERENCE VEPG, 12004-1, pp. 11-12. ANSWER 7.30 (1.50) l 1. Cartridge removed by filter handling system. [0.53
- 2. Placed in shielded drums at filter transfer station. [0.5]
! 3. Shielded drums are then transported to the solidification buildins.EO.53 REFERENCE VEGP, R-SRW-LP-1.
i 4 ? 1 i 1 e e 1 e
, . , . - - , , , _ _ _ . . . _ _ . . _ ,,-,.,,,_,%e_,_._,.. ...,,-,.-_.m._.._ _ __
. _ _ -_ . . - - , , . . _ _ --._r._, - - _
.s . .
2 D. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS 1 ANSWER 8.01 (1.00) (d) REFERENCE VG, TS, p 3/4 3-6. ANSWER 8.02 (1.00) (b) REFERENCE VG, TS, p 3/4 2-11. ANSWER 8.03 (1.00)
- a. Operations Support Center Manager. (0.5)
- b. On shift Health Physics Foreman. [or repco<<--e ac"wd (0.s>
i REFERENCE VG, 91402-C, p 1. I ANSWER 8.04 (1.00) (b) REFERENCE VG, 00005-C, pp 1,2. ANSWER 8.05 (1.00)
- a. 1. Cease activities associated with the performance of the procedure
- 2. contact his supervisor
- b. 1. ensure associated plant systems and equipment are restored to conditions, such that plant and personnel safety are not jeopardized.
- 2. revise the procedure (IAW 00051-C or 00052-C) prior to resuming associated activities.
REFERENCE VG, 00054-C, p 2. i O d
---v v , - , . - -- ---,,,-,,.w - - - , - - - , , - , - . . . -+, , --
n - - -
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS Life ^teYtt gfob /. 6ee /g 75/7 W * **e8 I/ b **
- Y
- a. No 1$ cation of an unurval event. (0.5)
- b. TS 3 1.2.6 requires plant shutdown to hot standby within 1 hour; 912001 g 8 classifies any plant shutdown required by TS as a notification of WfkN$'t REFERENCE h7))d s. E c,5fl GCCtf V G) koM 44alkMaG i 5)
- 6) Sint.e A se 3 te.rrl F)gyW 6er.4 c 4c.
VG, TS, p 3/4 1-12; VG,91001-C, Figure 4. (9,focrn// 4.cwoca= 4 6e res@J) ANSWER 8.07 (1.00) A subjective judgement by supervisory and off soins personnel whether oncomins personnel are alert, coherent and fully capable of performing the assigned duties. REFERENCE VG, 00003-C, p 1. ANSWER 8.08 (1.00) (c) REFERENCE CR, EM-207 Rev. 16, p 5; VG, 10CFR20.403, (2) and (4). ANSWER 8.09 (1.00)
- a. False
- b. False REFERENCE VG, 93210-C, pp 1,3.
6 m --,
. , c
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 ANSWERS -- V0GTLE 1 -06/06/23-TOM ROGERS ANSWER 8.10 (1.00)
- 1. Press the red pushbutton marked " Emergency Stop" immediately. (0.25)
- 2. Turn off the mein power switch. (0.25)
- 3. Notify the appropriate foreman (0.25) r (0.25 for ccrrec_ crder)
Ca+f of-REFERENCE VG, 93230-C, p 2.
%3 *c
%*
- V )
ANSWER 8.11 (1.00)
- a. this ponstitutes a failure to neet the o an L CO .1,(A Sk.% G4dustk tWa4 t.taheds.e ky exo%p acc$pe coan r a b i l i t y r e q u i r e m(0,5) entsf exceptions are: <-in.* uo 7 h par <W ff.a A # %yw.rejfoerform d4<.wery )
- b. rft3
- 1. stated in the individual spec., and M
- 2. survei lances do not ave to be performed on inoperable equipment
$. w ke V ng wM) of curMt o fth'c k W a f* (2. g h o- h REFERE CE y VG, TS, p 3/4 0-2.
ANSWER 8.12 (1.00)
- 1. a change in the purpose or scope of the affected procedure
- 2. a change to plant administrative procedures or emergency operating procedures
- 3. a reduction in the established level of safety
- 4. a change in acceptance criteria that is less conservative than previously established
- 5. a change to Tech Specs or the FSAR.
REFERENCE VGr AP 00052-C, pp 1,7. l l l l l O r w -. - - . - , . t
a i
. . a
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 52 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 8.13 (1.50)
~ Containment integrity shall exist when
- a. All penetrations required to be closed during accident conditions are either (1) Capable of being closed by an operable containment automatic isolation valve system, or (2) Closed by manual valves, blind flanses or deactivated automatic l valves secured in their closed positions.
- b. (3) All equipment hatches are closed and sealed.
- c. (4) Each air lock is in compliance with the requirements of Spec. (3.6.1.3).
- d. (5) The containment leakage rates are within the limit of Spec (3.6.1.2) and
- e. (6) The sealing mechanism associated with each penetration is operable.
REFERENCE VG, TS, p 3/4 6-1; VG, ROLP, 97.
~
ANSWER 8.14 (1.00) The associated effects of exceeding the oxysen, chloride and flouride limits are time (0.25) and temperature (0.25) dependent. Operation may be continued with contaminant concentration levels in excess of the steady state limits, up to the transient limits, for the specified limited time interval without having a significant effect on the structural integrity of the RCS. (0,5) REFERENCE VG, TS, p B 3/4 4-5. ANSWER 8.15 (1.00) Suspend all operations involving emic olierations ui positive reactivity changes. (gf oggggj g6 m hg[y,*gy ggg g jrtMfah Mra a[#,q REFERENCE C* H W inh & tac h S. Fan /Arn (+.n) , gwr&lT f.U) VG, TS, p 3/4 1-11. 9
f > 'i' ' 8.- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 53 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 8.16 (1.00) (d) REFERENCE VG, TS, p 3/4 8-2.
- ANSWER 8.17 (1.00)
Immediately initiate and continue boration at =or> 30 spm of a solution l containing =or> 7000 ppm boron or equivalent until the required shutdown parsin is restored.
- REFERENCE 1 VG,TS, p 3/4 1-3.
l ANSWER 8.18 (1.00) A maximum temperature change of =or< 10 degF in any one hour period during inservice (hydrostatic and) leak testing operations above the heatup and
; cooldown limit curves.
REFERENCE VG, TS, p 3/4 4-30. i } ' ANSWER 8.19 (2.00) i
- 1. In the event of a LOCAt
- a. sufficient water is available within containment to permit
.: . recirculation cooling flow to the core, and, (0.4) ! ' the reactor will remain suberitical in the cold condition following
- b. .
i mixing of.the RHST and the RCS water volumes with all rods inserted except for the most reactive control assembly. (0.4)
- 2. The volume includes an allowance for water not useable because of tank
. discharge line location. (0.4)
- 3. Ensures a Ph value between 8.5 and 10.5 for the solution recirculated within containment after a LOCA. This Ph band minimizes the evolution of iodine (0.4) and minimizes the effect of chloride and caustic stress corrosion (0.4) on mechanical systems and components.
REFERENCE
- VG, TS, pp B 3/4 5-2,3.
4 I i i 1
~
4 I i , , , _ _ . . . - , . _ ___
[. 6 @ h
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 ANSWERS -- v0GTLE 1 -86/06/23-TOM ROGERS ANSWER 8.20 (1.00)
- 1. The safety of the reactor is in jeopardy.
- 2. Operating parameters exceed any of the reactor protection circuit setpoints and automatic shutdown does not occur.
- 3. Shutdown is required to protect personnel, equipment and the public.
- 4. Warranted by unusual circumstances.
REFERENCE VG, AP 10000-C, p 13. ANSWER 8.21 ( .50) It is a priority key (0.25) for a high rad area. (0.25) REFERENCE VG, AP 00008-C, p 4. ANSWER 8.22 (1.00)
- 1. isolation valve open
- 2. contained borated, water volume
- 3. boron concentration
- 4. nitrogen cover pressure REFERENCE VG, TS, p 3/4 5-1.
ANSWER 8.23 (1.00)
- 1. current revision available.
- 2. all applicable temporary changes are identified and available.
- 3. the procedure is complete and intact.
- 4. all pertinent checklists, data sheets, figures and tables are included and available for use.
- 5. the activity and the directions provided in the procedure are understood (4 0 0.25)
REFERENCE VC, 00054-C, p 1. 6 O
. .6 e *
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 55 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 0.24 (1.00)
- 1. Relays under control of maintenance department during testing, individual performing test can authorize resets. (0.25)
- 2. One time resetting. Shift supervisor can authorize one reset if caused by overcurrent condition. (0.25)
- 3. Reset relay target if the target is showing while equipment is running (0.25)
- 4. UndervoltaSe relays after voltage has been restored to a de-energized bus, load center, or MCC the shift supervisor.,may authorize resetting undervoltage relay once. (0.25)
REFERENCE VG, 10007, p 2. ANSWER 8.25 ( .50) False REFERENCE VG, 10002-C, p 1. ANSWER 8.26 (1.00)
- 1. Provide positive control of work involving radiation and insure exposure is ALARA.
- 2. Provide documentation of radiation exposure to individuals performing work covered by an RWP.
REFERENCE VG, 43007-C, p 1. ANSWER 8.27 ( .50) FALSE. REFERENCE VEGP, TS 6.5.1.6. 9
(. B b
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 56 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER B.28 (1.00)
- 1. When a Work Order requires it to be used as part of a functional test E0.5].
- 2. When installed by an approved procedure EO.5].
REFERENCE VEGP, 00306-C. ANSWER 8.29 (1.00) . Cause of trip known and corrected EO.53. All safety related equipment functioned properly during trip E0.53. REFERENCE VEGP, 00300-C. ANSWER 8.30 ( .50) TRUE. . oy et a s- (ple w bb csh h Oc.$. gQ%g REFERENCE VEGP, 00305-C. 9
~
- i e . >
1 , 3
> 1 iS I
4 S I5 " 4 - I l REJELTED ENTMOPY s 1-2: Saturated steam from the steam generator is esipanded in the MP turbine to provide shaft work output at a constant entropy. , it., 2-3: The moist steam from the exit of the HP turbine is dried and sqperheated in the MSR. - , 3-4: Superheated steam from the MSR is' expanded in the.LP turbine to provide shaft work output at a constant entropy. 4-5: Steam exhaust from the turbine is condensed in the condenser in
+
which heat is transferred to the cooling water under a constant vacuan condition. 5-6: The feedwater is compressed as a liquid by the condensate and l l feedwater ptmps and the feedwater is preheated by the feedwater heaters. l [ 6-1: Heat is added to the working fluid in the steam generator under t a constant pressure condition. l l Fijnre 9 +7 O
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.) h AST92-U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: V0GTLE 1 REACTOR TYPE: PHR-WEC4 DATE ADMINISTERED: 06/06/23 EXAMINER: TOM ROGERS APPLICANT: _________________________
INSTRUCTIONS TO APPLICANT: Usa separate r, aper for the answers. Write answers on one side only. Stcple question sheet on top of the answer sheets. Points for each quastion are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grede of at least 80%. Examination papers will be picked up sin (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 30.00 25.00
________ ______ ___________ ________ 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIONr THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOL! 30.00 25.00 ________ ______ ___________ ________ 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 30.00 25.00
- 3. INSTRUMENTS AND CONTROLS 30.00
________ _['5.00 ____ ___________ ________ 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 120.00 100.00 TOTALS FINAL GRADE _________________% All work done on this examination is my own. I have neither given nor received rid. kPEL5C5UT'5 556EETU5E~~~~~~~~~~~~~~ 6 J
ES-201-2 i NRC RULES AND GUIDELINES FOR LICENSE EXANINATIONS During the administration of this examination the following 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 & to facilitate legible reproductions.
Print your name in the blank provided on the cover sheet of the 4. examination.
- 5. Fill in the date on the cover sheet of the examinatiori (if necessary).
- 6. Use only the paper provided for answers.
, 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet. I
- 8. Consecutively number each answer sheet, write "End of Category " as appropriate, start each category on a new page, write og one sTde of -
the paper, and write "Last Page" on the Tast answer sheet. i ~
- 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 only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
r Examiner Standards 12 of 18-e
. ES-201-2
- 18. When you complete your examination, you shall:
- a. Assemble your examin'ation as follows:
(1) Exam questions on top. (2) Exam aids - figures, tables, etc. (3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
i l l l l l r Examiner Standards 13 oT 18
- . _ - . - , , - - - - - - - - . . - - - - , - - - . , , - -- - - - - /
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
~~~~isERs56isAsics, SEAT iRAssFER As5 FLUi5 Ft5s QUESTION 1.01 (1.00)
During fuel loading, which of the following will have NO effect on the shape of a 1/m plot?
- a. The location of the neutron sources in the core.
- b. The strength of the neutron sources in the core.
- c. The location of the neutron detectors around the core,
- d. The order of placement of fuel assemblies provided the proper enrichments are Placed in their proper location.
QUESTION 1.02 (1.00) A general rule is often stated ' doubling the count rate halves the margin to criticality'. This is mathematically stated by the equation: V I CR1/CR2 = 1-Kefff / 1-Kefff Which one of the following statements is correct concerning the above statement and equation?
- a. Both Keff1 and Keff2 have to be less than 1.0.
- b. Equal changes in Keff result in equal changes in soberitical multiplication level.
- c. The equation only approximates the instantaneous change in count rate; once equilibrium value is reached, the count rate will be higher.
- d. A second doubling of the count rate will result in the reactor becoming critical or supercritical.
(****r CATEGORY 01 CONTINUED ON NE.XT PAGE *****) 4' ,
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
--- isEER557RARics- sEAi isAssFEE An5 FEGi5 FC5E QUESTION 1.03 (1.00)
OP-12003-1, " Reactor Startup", requires that the critical rod position be taken at 1.0E-8 amps on the intermediate range. If, during a xenon free reactor startup at MOL, the operator " overshot" 1.0E-8 amps and instead leveled off at 1.0E-7 amps, which of the following statements is correct?
- a. At 1.0E-7 amps, there are little or no effects from nuclear heat but since the reactor is a decade higher in power, the critical rod position will be higher.
- b. At 1.0E-7 amps, there are little or no effects from nuclear heat; therefore, the critical rod position will be the same as at 1.0E-8 amps.
- c. At 1.0E-7 amps, there are substantial effects from nuclear heat; therefore, the critica3 rod positions will be higher than at 1.0E-8 an ps.
- d. At 1.0E-7 amps, nuclear heat, xenon and the decade higher in power level will result in a higher critical rod position.
QUESTION 1.04 (1.00) The reactor trips from full power, equilibrium xenon conditions. Six hours j later the reactor is brovsht critical at 1.0E-8 amps on the intermediate range. If power level is maintained at 1.0E-8 amps, which one of the following statements is correct concerning control rod motion for the next two hours?
- a. Rods will have to be withdrawn since xenon will closely follow its normal build-in rate.
- b. Rods will have to be inserted due to xenon decay.
- c. Rods will have to be rapidly inserted since the critical reactor will cause a high rate of xenon burnout.
- d. Rods will approximately renein es is es the xenon establishes its j equilibrium value for this power level.
( rn r* CATEGORY 01 CONTINUED UU NEXT P AGE
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1.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
--- isERR557RARICs- REAi TRAssFEE As5 FEUi5 FE5E QUESTION 1.05 (1.00)
Figure 953 shows a graph of pump laws i.e., the relationship between pump speed and other pump parameters. Which one of the following parameters is represented by curve 2?
- o. Power
- b. Flow rate
- c. Voltage
- d. Discharge pressure QUESTION 1.06 (2.00)
How does each of the following parameter changes affect the DNDR (increase-decrease-remains the same) and why (be brief)? Do not consider the transient effects.
- a. Average primary coolant temperature increases 5 de3F.
- b. Pressurizer temperature increases 5 desF.
- c. Mass flow rate in the core decreases 10%.
- d. Mass flow rate of steam to turbine decreases 10%. Assume rods in Buto.
QUESTION 1.07 (1.00) Indicate how each of the following change over core life at 50% power. (More negative or less negative)
- 1. Fuel temperature coefficient.
- 2. Doppler only power coefficient.
QUESTION 1.08 ( .50) > A reactor that is prompt critical is also supercritical. TRUE or FALSE? (***** CATEGORY 01 CONTINUED ON NEXT PAGE wrrn)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5
--- isEER65isARICs- sEAi isAssFEs As5 FEUi5 FE6s QUESTION 1.09 (1.00)
When starting up a centrifugal pump with the discharge valve shot, how would you characterize each of the following parameters compared to its value when starting the pump with the discharge valve fully open?
- o. motor current
- b. discharge pressure GUESTION 1.10 ( .50)
The following is a description of an important fission product. "Although negligible amounts of it are produced directly by fission, 1.1% of all fissions result in the production of its two precursors in its decay chain. It has a half life of more than 1.0E+16 yeers and therefore is considered stable. Its reactivity effect in the Vogtle reactor is independent of flu:: once it has reached equilibrium". Name the element. The isotope is not required. QUESTION 1.11 (1.50) An estimated critical baron concentration of 975 ppm was calculated using the information on Data Sheet 1 (figure 957 attached). Assume the reactor achieved criticality at 153 steps on Bank D. RC Tave was 561 degF. How would this critical rod position change (Bank D position relative to the previous criticality) if each of the following conditions was different from that on Data Sheet 1? Assume no boron change and consider each condition separately.
- e. Line iv. Lc=u ruwa level wes s v e. rr sur av ..r-"_
- b. Line 2a. Time shutdown was 72 hours.
- c. Average RC temperature was 532 desF.
QUESTION 1.12 (1.00) Shutdown margin requirements vary throughout core life as a function of several factors. The most restrictive condition occurs et EOL. State the factor and why it is most restrictive at EOL. (xrr>* CATECORY 01 CONTINUED ON NEXT PAGE ***rr) 1
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6
~ ~~~~5EERO66YU5biCEr"55Ii~TR5 Ef5R k 6'ELUi6 FL6s QUESTION 1.13 (1.00)
The startup, intermediate and power range channels all use boron in their respective detectors (BF3 or boron lined). What is the reason for use of boron? QUESTION 1.14 (3.00) List three plant parameters that can be controlled by the operator to enhance natural circulation flow (including any applicable values) and briefly state the significance of each as it relates to natural circulation Where numerical values in the EOPs differ from the Westinghouse values, cite only the E0P value. QUESTION 1.15 (1.00) Over core life, the reactor becomes more responsive for a given reactivity change. Explain this by naming the basic reactor physics parameter which is changing, the direction of its change and the isotope primarily responsible for the change. QUESTION 1.16 (2.00) The reactor is being shutdown. A stable neutron decay rate of -78 seconds (-1/3 DPM SUR) has been established from the point at which both intermediate range channels read 1.0E-8 amps (time 0). One intermediate detector is properly compensated, the other is undercompensated.
- a. When (with respect to time 0) will the HV be restored to the SRMs?(.5)
- b. Discuss the time it takes the actual flux to reach a value equivalent to P-6. (.5)
- c. Discuss the HV restoration time for one SRM channel and for both.(1.0)
(zu*** CATEGORY 01 CONTINULD ON MD'T PAGE **rry) h
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 7
--- isERR559aARIEs- sEAi isAssFEE AR5 FE5i5 FE5s QUESTION 1.17 (2.00)
S=turated water vapor enters a turbine 200 psia at 200,000 lbm/hr and provides 23,575 horsepower to the shaft.
- o. What is the enthalpy change per unit mass across the turbine?
- b. Is it plausible for turbine exit moisture content to be 26% ? Yes or No and Justify.
QUESTION 1.18 (1.00) When xenon is in equilibrium the xenon formed equals the xenon lost. Briefly explain the primary mechanism by which xenon is being lost from the system at 100% power. QUESTION 1.10 (2.00) On a T-S diagram, plot a characteristic Rankine cycle for a steam cycle containing an MSR, turbine with HP and LP sections, feedpump, condensate pump, feedheater, steam generator and condenser. Label each segment of the cycle showing which plant component is responsible for that portion of the cycle. QUESTION 1.20 ( .50) With the main steam temperature and pressure at 560 desF and 1125 psis respectively, a main steam atmospheric relief valve seat begins to leak to atmospheric pressure. Use the attached steam tables or Mollier diagram to determine the temperature of the steam three feet out of the relief valve. QUESTION 1.21 (2.00) Given the power history shown at the top on figure 978 attached, sketch the expected xenon history on the blank drawing provided at the bottom of that figure. (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 8
--- iREER55isisiEs- REAi isissFEs AR5 FEUi5 FE5s GUESTION 1.22 (1.00)
You are plotting an inverse multiplication plot (1/m) during fuel loading. When fuel is loaded so that the distance between the detector and the fuel steadily decreases, the 1/m plot will have a characteristic general shape. Sketch the general form of this curve (indicate shape, slope and rates of change on your sketch). QUESTION 1.23 (1.00) What is the change in reactivity associated with a change in Keff from 0.92 . to 1.004? QUESTION 1.24 (1.00) If reactor power increases f r o n: 1000 cps to 5000 cps in 30 seconds, what is the SUR? (*rr** END OF CATEGORY 01 wrxxr)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.01 (1.00)
The shunt trip coils for reactor trip breakers
- a. Do not receive a trip signal when a manual reactor trip is actuated.
- b. Cannot be trip tested separately from the undervoltage coils.
- c. Receive a signal to trip from an auto reactor trip signal through the auto shunt trip coil circuit.
- d. Are deenergized to trip, thus meeting the single failure criteria.
QUESTION 2.02 (1.00) The plant is operating at 100% power when an electrician inspecting a downstream RCP breaker (the class 1E breaker) accidently causes the underfrequency relay for the breaker to actuate its protective function. Which of the following best describes the plant response.
- a. The affected RCP trips. The operator performs 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 Icop.
- c. The affected RCP trips, reactor trips and the other 3 RCPs trip.
- d. The affected RCP trips, reactor trips, the other 3 RCPs trip and the plant sis on low low flow.
QUESTION 2.03 ( .50) The radioactive liquid waste monitor tanks discharge to the River Water Makeup System Blowdown Sump to dilute the effluent with River Water Makeup flow. TRUE or FALSE? OUESTION 2.04 (1.00) What is the flow path of steam for steam chest warming during a turbine startup? (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.05 ( .50)
How long is a ESF battery rated to last when it is supplying full load? DUESTION 2.06 ( .50) What is the first ESF battery load that will fail due to battery discharge? QUESTION 2.07 ( .50) What is the source of power for the automatic field flash of the Emergency Diesel Generators? QUESTION 2.08 ( .50) How many successive starts are the Emergency Diesel Generator Air Receivers designed to perform without repressurizing them? QUESTION 2.09 (1.00) During a routine manipulation on DMCB-A, the reactor operator inadvertently turns off the running turbine plant closed cooling water (TPCCW) pump. Briefly explain what happens to the standby pump and why. QUESTION 2.10 (1.00) What would be the effects during a DDA (Design Bases Accident) LOCA if the accumulators hadt (Assume other parameters are normal in each case.)
- a. Low water level
- b. High water level
- c. Low nitrogen pressure
- d. High nitrogen pressure
(*rr*r CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 11 QUESTION 2.11 (1.50) 2/3 (two out of three) pressurizer pressure channels below 1970 psig (P-11) allows the operator to manually block certain protective functions. What arei thr ee ]the protective functions blocked by this action?
8 QUESTION 2.12 ( .50) Choose one item from the parentheses: In order to open (an air, a motor) operated valve it normally must be hold in the open position until the green light is out. QUESTION 2.13 (1.00) While operating at 100% power, a IAA02 breaker trips and the associated diesel generetor starts and flashot its field. The plant operator immediately trips the diesel by pressing both the emergency stop pushbuttons. Why did he stop the diesel? OUESTION 2.14 (1.50) Give a general description of the 6 locations where the 12 DMIM accelerometers are located. QUESTION 2.15 ( .50) Why is a reactor coolant pump locked rotor accident more severe than c reactor coolant pump trip? QUESTION 2.16 (1.00) A. Where on the Reactor Coolant System loops are the flow transmitters located? D. How neny flow sensors are there per loop? (rryr CA.TEGnoY 02 Cop 1IngED 09 NEvT PAGE *****)
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 12 QUESTION 2.17 (2.00)
List four design considerations the Main Steam flow restrictor nozzles w,rre required to meet. QUESTION 2.18 (1.00) There are eight bases of the RIL system when in mode 1. Three are stated in a thru c below. Name the remaining five.
- c. Dropped or misaligned rod will not violate thermal design basis (DNBR)
- b. Consequences of an ejected RCCA accident restricted to within analysis
- c. Enthalpy rise hot channel factor (or FdeltaH ) within acceptable limits 00ESTION 2.19 (1.00)
List 5 of the 6 inputs to the loed control unit of the EHC other than chest
/shell warming bias. Consider all limiter inputs as a single input but be specific. Setpoints are not required.
QUESTION 2.20 (2.00) List the 8 reactor trips other than ' General Warning" which cannot be bypassed or blocked. Include coincidence (logic). Setpoints are not required. QUESTION 2.21 (1.00) List all the auto start signals for the motor driven auxiliary feedwater Pumps. Include logics (coincidences) and setpoints. QUESTION 2.22 (1.00) List four of the five liquid process monitors with cutomatic actions. Giving the automatic action is NOT required. (***** CATEGORY 02 CONTINUED ON NEXT PAGE *rrrr) r -
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- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 13 QUESTION 2.23 (2.00) l Excluding the NSSS sample coolers, list ten other heat loads on the ACCW (coxiliary component cooling water) system.
QUESTION 2.24 (1.50) List 3 permissives which must be satisfied for an emergency start of the diesel engine. QUESTION 2.25 (1.00) List the five loads supplied by the essential chilled water system. QUESTION 2.26 (1.50) List in order of descending priority the critical safety function status alstms displayed on the top levels of SPDS. QUESTION 2.27 (1.00) List the four items in the ECCS acceptance criteria of 10CFR50. QUESTION 2.28 ( .50) The plant is operating at 100% power with the steam dump system normally aligned. Suddenly PT-506 (load reject) instrument fails low,
- a. Describe the response of the steam dump control system to the failure (ier specifically, how will the change in PT-506 input change or effect system status).
QUESTION 2.29 (1.00) Explain how more than one SG is prevented from blowing down if the A SG is faulted and the active failure is the A SG main steam isolction valve failure to close. (***** END OF CATEGORY 02 *****)
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- 3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.01 (1.00)
Which of the following is NOT a diesel seneratnr output breaker closure permissive?
- a. No 4160 ESF Dus faults.
- b. SI signal present.
- c. 4160 ESF Bus undervoltage.
- d. Diesel at rated speed and voltase.
QUESTION 3.02 (1.00) Which of the followin3 stop signals will trip the diesel engine following an engine start upon receipt of a safety injection signal?
- a. High bearing oil temperature.
- b. Depressing the normal stop pushbutton on the local engine control panel.
- c. Engine overspeed.
- d. Generator pha e overcurrent.
QUESTION 3.03 (1.00) While scanning the MCD, the operator sees the ACCW pump 'A' with a steen and amber light illuminated and e red fla3 The probable cause is
- a. The pump has tripped and no one reset the flag.
- b. The pump is running in low speed.
- c. The pump is set up to auto start.
- d. The pump auto stopped due to low demand.
(str*r CATEGORY 03 CONTINUED ON NEXT PAGE rrwxx)
a 3.- INSTRUMENTS AND CONTROLS PAGE 15 QUESTION' 3.04 (1.00) The plant is operating at 50% power when a RCS wide range temperature f element fails high. Which one of the statements below best describes the i affect on deltaT and Tavs? I
- a. DeltaT decreases, Tave increases
- b. DeltaT increases, Tavs decreases
- c. DeltaT decreases, Tavs none
- d. DeltaT none, Tavs none
- e. DeltaT none, Tav3 increases 00ESTION 3.05 (1.00)
Which statement best describes the purpos~ of the Tavs defeat switch?
- a. Allows the operator to defeat a single input into the rod insertion limit,
- b. Allows the operator to defeat a single input into only the Tavs/ Tref recorder and alarm. *
- c. Allows the operator to defeat a Tave channel into the auctioneered Tav3 calculations.
- d. Allows the operator to select which single auctioneered high Tavs channel is used in the control circuits.
QUESTION 3.06 (1.00) The power range gain potentiometer used to adjust the excore indication to agree with a caliormetric changes the gain on
- a. the detector outputs.
- b. the shunt switches.
- c. the isolation amplifier output,
- d. the summing amplifier output.
(rrrry CATEGORY 03 CONTINUED ON NLX1 PAGE rarrr)
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- 3. INSTRUMENTS CONTROLS PAGE 16
________________'AND ____________ QUESTION 3.07 (1.00) Which of the following controllers will control nearest to setpoint?
- a. A proportional controller with a gain of 10.
- b. A proportional controller with a gain of 20.
- c. A proportional / derivative controller with a proportional gain of 15.
- d. A proportional / integral controller with a proportional gain of 15.
QUESTION 3.08 (1.00) For each of the followitig circuits, state whether it receives an input signal from turbine impulse pressure channel PT-505, PT506 or both.
- a. C-5
- b. C-7
- c. P-13
- d. Rod control system QUESTION 3.09 (1.00)
Based on the operating principle of the dp cells in the pressurizer level transmitters, indicate the response of indicated level (increase, decrease, remain'the same) to the following situations.
- a. Adverse conteinment conditions cause the temperature of the liquid in the reference les to increase,
- b. A small leak causes the level in the reference les to decrease.
QUESTION 3.10 (1.00) (c) What is the normal DRPI accuracy? (b) What it the DRPI accuracy with a Datc A failure? (***** CATEGORY 09 CONTINUED ON NEXT PAGE *****)
- 3. INSTRUMENTS AND CONTROLS ,
PAGE 17 QUESTION 3.11 (1.00) What two conditions will automatically shut the SG Blowdown System d2mineralizer inlet valves? Setpoints not required. QUESTION 3.12 (1.00) What electrical parameter will change if the Main Generator excitation field current is changed when . (a) the generator output breakers are open. (b) the generator output breakers are closed. QUESTION 3.13 (1.50) The plant is operating at 100% power with the accuracy mode seitch in the AE B position. A detector coil in Digital Rod Position Indication (DRPI) d a t a E: fails open. What alarms on the control board display will alert the operator to this failure? Consider only those alarms on the control board display in your answer. Three alarms are required. QUESTION 3.14 (1.00) Differentiate between steam dump control under the load rejection controller and steam dump control under the plent trip controller by stating:
- a. Input signals for each, (0.5)
- b. Which system is aligned for control during normal plant operation at power, and (0.25)
- c. The signal r espons ibl e for switchina the system to the alternate (ie, not normelly aligned) Teve controller. (0.25)
(rn.wr* CATEGORY 03 CONTINUED ON NEXT PAGE rrstr)
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- 3. INSTRUMENTS AND CONTROLS PAGE 18 QUESTION 3.15 (1.00)
In what modes is Tech Spec 3.3.3.6 for accident monitoring instrumentation spplicable? List 3 of the 4 temperature measurements required to be operable by this Toch Spec. QUESTION 3.16 (1.00) Cold Overpressure Mitigation System (COMS) provides overpressure protection only if the system has been armed.
- a. What signal (or condition) is necessary for the COMS system to be armed assuming normal plant shutdown?
- b. What component other than the PORV itself is affected by the arming signel?
00ESTION 3.17 (1.00) The balance of plant operator in the control room is " walking down" the boards prior to shift turnover. At the electrical distribution section he notices the lower set of white lights for 1AD04 are not lit. The 400 volt incoming breaker from the 4.16 KV/480 volt transformer has green and amber lights on and the 4.16 KV breaker to the 4.16 KV/400 volt transformer is red. What does this light arrangement tell the operator? DUESTION 3.18 (1.00) The output of the dynamic range delter transmitter of the Reactor Vessel Level Instrumentation System (RVLIS) is used to provide information for conditions of forced flow,
- e. State where the taps are made in the reactor vessel for this (0.5) measurement.
- b. At het zero power with 4 RCPs r unnine r what should be the output of this channel? (0.25)
- c. What condition is indicated by a decrease in this output, assuming zero pouer conditions and 4 running pur:ps ere maintained? (0 25)
(rzrx* C A T E G O ^*'. 03 CON 1100LO On FI X1 PAGL arrst)
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- 3. INSTRUMENTS AND CONTROLS PAGE 19 QUESTION 3.19 (1.00)
'What is the purpose of the PORV and block valve interlock?
DUESTION 3.20 (2.00) List four items in the Rod Control System the Startup Switch will reset. QUESTION 3.21 (1.50) List the three inputs to the Rod Insertion Limit Computer Comparator. QUESTION 3.22 (1.00) List the two different inputs to the Peacter protection System used to setisfy P-9 (Turbine Trip-Reector Trip). QUESTION 3.23 (1.50) List all-containment isolation phase A actuation signals. Include coincidence (logic). N0 setpoints required. QUESTION 3.24 (2.00) List the four signals that initiate safety injection. Setpoints and logics (coincidences) not required. QUESTION 3.25 (1.00) Describe the control functions of TI-301 (demin inlet temperature control) during boration and dilution evolutions. Include operating temperatures and valves controlled. (***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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- 3. INSTRUMENTS AND CONTROLS PAGE 20 QUESTION 3.26 ( .50)
What is the purpose of the timing unit in the Baron Measurement System? DUESTION 3.27 (1.00) Sketch the logic symbols that represent (o) a 2 out of 3 coincidence and function. (b) an adjustable time delay energizing function. (***** END OF CATEGORY 03 *****) N
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 21
--- RA5i5E55i5AE E5sfR5E------------------------ l, t i QUESTION 4.01 (1.00) j Following a Reactor Trip / Turbine Trip, the senerator output and field j breakers are verified to open by the operator e i
- e. when directed by 19000-1. '
- b. after all immediate action steps of 19000-1 are complete.
- c. as part of the verification of the Turbine Trip.
- d. 30 seconds after the Turbine Trip.
QUESTION 4.02 (1.00) f i Following a Reactor Trip with no SI, AFW flow should be throttled
- a. whenever total AFW flow exceeds 550 spm.
- b. when all SG wide ranse levels are greater than 60%.
- c. only when directed by EDP 19000-1. ,
- d. as soon as subcoolins is verified. l QUESTION 4.03 (1.00) I Which of the followins is a 10 CFR 20 exposure limit? l
- a. 5 rem / year-whole body.
I
- b. 1 rem / quarter-whole body.
- c. 18.75 rem / quarter-hands.
- d. 7 rem / quarter-skin of whole body.
(r**** CATEGORY 04 CONTINUED ON NEXT PAGE *****) i
}
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 22
~~~~R A5i5t55i5At C5sTR5t QUESTION 4.04 (1.00)
Which of the following radiation eMPosures would inflict the greatest biolo3ical damage to man?
- o. 1 Rem of GAMMA.
- b. 1 Rem of ALPHA.
- c. 1 Rem of NEUTRON.
- d. NONE of the above; they are all equivalent.
QUESTION 4.05 ( .50) When determining an individuals occupational doser any emergency doser ere to be EXCLUDED et pert of the exposure history. TRUE of FALSE? QUESTION 4.06 (1.00) What is the purpose of an HP Margin List? 00ESTION 4.07 (1.00) Why are some Enersency Response Procedure step subtasks preceded by bullets (o)? QUESTION 4.08 ( .50) One of the purposes of the Foldout Page is to direct an operator to the appropriate FRG in the event a Red Path li n.i t is violated. T7UE or FALSE? DUESTION 4.09 ( .50) During a faulted SG accident, the FSAR takes credit for the operator isoleting the AFW to the fenlied GG. TRUE or FALGE? (i.rr** C AT EGOr:Y 04 CONTINUED OH NEX1 PAGE **r*4)
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 23
~ ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~Rd656L665CdL 66UTR6L QUESTION 4.10 ( .50)
Does e required task in the Optimal Recovery Guidlines need to be complete before continuing to the next step? QUESTION 4.11 ( .50) What is the maximum allowable differential pressure across the Mainsteam Isolation Valves when openin3 them? QUESTION 4.12 (1.00) List two different conditions when an Inverse Count Rate Ratio Startup procedure is required. QUECTION 4.13 (1.00) During a reactor startup to 100% power, at what power. level are the Reserve Auxiliary Transformers transferred to the Unit Auxiliary Transformers? DUESTION 4.14 (1.00) What immediate operator action is required upon loss of offsite power concurrent with the A diesel generator failing to tie on the bus? 00ESTION 4.15 (1.00) Where does each of the following operators report to during a Control Room evacuation event? (a) 00P operator. (b) RO operator. (rxx** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
. - .. e : ,
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 24
~~~~ ------------------------
RA5iBE55iEAC E6UTR5L QUESTION 4.16 (1.00) If one SG main feed pump trips while at 100% power, what problems may you oncounter in keeping the unit on line if you reduce turbine load to fast or to slow? QUESTION 4.17 (1.00) A. What is the maximum time a reactor coolant pump is allowed to remain in operation with an indicated il seal failure? B. What is the maximum allowable reactor power level at which you may trip one reactor coolant pump? DUC5 TION 4.10 ( .50) Which NRC office does the NT:C phone in the control roon directly contact? QUESTION 4.19 (1.50) List three items that the startup procedure requires the operator to check after every 50 steps of Control Rod Bank withdrawal. QUESTION 4.20 (1.50) List six actions that should be accomplished prior to evacuation of the Control Room when taking control at the Remote Shutdown Panels. QUESTION 4.21 (2.00) List the immediate operator actions for an uncontrolled continuous rod intertion. QUECTION 4.22 (1.50) List three means of miriimizing the possibility of water hammer during oper - etion of the AFW system. (rrr*r CATEGORY O c. C0tn 1NUED ON NEX1 PAGE r***s-)
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 25
~~~~ - ------------------------
RA5i5t55iEAL CBNTR5t QUESTION 4.23 (1.00) List four affects on the Reactor Coolant System a loss of feed water will cause before any systems automatically mitigate the affects. QUESTION 4.24 (1.50) If symptoms of a turbine runback occurs from 100% power, what three items Bhould you check to verify a runback is required? QUESTION 4.25 (1.00) List the immediate operator actions for a dropped rod. QUESTION 4.26 (1.00) When is a continuous action of a WHEN, THEN step in an Op tirn a l Recovery Guidline cerried over to a 'transitioned' Emergency Procedure? QUESTION 4.27 (1.00) Under what conditions are the adverse containment values of the E0Ps used? DUESTION 4.28 (1.00) Explain the conditions which would have you implement each of the following Netural Circulation Cooldown procedures. (a) E0P 19003-1, Natural Circulation Cooldown with Steam Void in Vessel. (b) E0P 19004-1, Natural Circulation Cooldoun with Steam Void in Vessel. (*xwyr CATEGORY 04 CONTINUED ON NEXT PAGE *****)
, hs- u .
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~R5656L66565E"66 TR6L QUESTION 4.29 (1.00)
During a plant heatup with the reactor trip breakers closed, the procedure has a precaution to withdraw all rods at least 2 steps whenever the RCS temperature increases by more than 50F. What is the purpose of this roquirement? DUESTION 4.30 (1.00) liow is the baron concentration between the Reactor Coolant Loops and the Pressurizer equalized? (rzr** END Of CATEGORY ^4 ****r) (iixrstrrrrrrr END OF EXAMINATION *zurrrrrrrrzrza) i
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DATA SHEET 1 / .. ESTIMATED CRITICAL CONDITION [- Date/ Time of Last shutdown 8 8d'k M Date/ Time of Projected Criticality k 8I / 8M ! l
- 1. Conditions Prior to Shutdown \ .).
1 l
- a. Puel Burnup ~72 ##O
- b. Re&ctor Power too c.;
- c. Boron Concentration see ,-
- d. Rod Position too / ,
te a on Bank E
- 2. Projected Conditions at Criticality' . , .
- a. Number of Hours from last s}hutdown AY
- b. ESTIMATED CRITICAL POSITIOR l
/ / Steps on Bank '
- 3. Reactivity Balance -
- a. Power Defect / \
(PTDB TAB 1.1 for cboditions in 1) /fd D PCM
- b. Xenon Worth at Pg(rer N.,
(PTDB TAB 1.4.24o/ conditions in 1) 2 (do . PCM
- c. Rod Worth at Patiet (PTDB TAB 1. '.1 for conditions in 1) IA# PCM
- d. Boron Worth t Power '
(PTDB TAB . I for conditions in 1) 8500 PCM
- e. Xenon Wo t riticality (PTDB T 1 and h a b.4.1 2) for conditions in 1 -
7 og PCM
- f. Rod Vor at'* Criticality MO
( B TAB 1.5.1 for ECP in 2) - PCM
- g. Sum t ru 3f /6, MO PCM
- h. TIMATED Critical Boron Concentration y 7[
B TAB 1.3.1 for worth in 3g) PPM gdleteby / / Signature Date Time Reviewed by / / Signature Date Time figure 957
l i r I l 100 1 l POWER,% - 50 i i t ; 30 80 90 TIME (HR) i i 150 100 NXe, % OF MAX. 50 t I 30 60 90 TIME (HR) i Transient Xenon-R 'er Change fIec
s E00All0N $NEET 1
. . . l Where Dg = Sg i, )
(densth)g(velocity)glarealg = (density)3(velocity)g(area)2 EE = SvI PE = mph peg + keg +PgVg a PE 2 *KI 2*P Y22 where V = specific j 7 - volume . P = Pressure 0 = &c,(Tout.Yg,). 0 = UA (T.,,.Tst.) 0 = s(hg-hr) l P = P,10(SUR)(t) p . p,,t/T SUR = 26.06 T = (8.p)t I p delta K = (K,gr-1)/K,gg CRg (1-K,ggi) = CR 2 II*Keff2) CR = 5/(1.K,gg) J I M = (1.K ,ggg) SOM = (1-K ,gg) x 1005
) Il-Keff2) Keff l decay constant = In (2) = 0.693 A = A,e.(decay constant)x(t) t t 1/2 1/2 1
l Water Parameters Miscellaneous Conversions , 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps
; 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 hp = 2.54 x 103 Btv/hr I ft3 = 7.48 gallons 3 1 MW = 3.41 x 106 Btu /hr Density = 62.4 lbg/ft 1 Stu = 778 f t-1bf Density = 1 gm/cm j l Heat of Vaporization = 970 Btu /lbe Degrees F = (1.8 x Degrees C) + 32 l Heat of Fusion = 144 Btu /lbe 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.S in Hg g = 32.174 f t-1bm/lbf.sec 2 I
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l l i . 1 APPENDlX E STEAM TABLES AND MOLLIER DIAGRAM De following Steam Tables and Mollier Diagram have been reproduced with the consent and by permission of Combustion Engineering, Inc.,1000 Prospect Hill Road, Windsor, Connecticut 06095.
c,
, Volume lil Appendix E Tetple 1. Saturated Sleem: Tempe7eture Tettle_
Iemp
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he h ag s, a,s sa I 0 0160?? 3304 7 3308 7 - 9 0179 4075 5 197bl 88 0 8 0 % 00 0016026 3061 9 3061 9 59% 9 0000 2 8873 2 8873 38*
#3 G leMS 1974 4 1976 4 0 0041 7 1767 7 1007 See
, 3B e 4 016020 20M e ISM S 4 000 1973 2 1977J $Nel F MSI F 3732 8ilM9 0016489 26M I 26M 2 6 DIO 8072 I ISPsI at
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- 87 0 0 54093 0 016077 $%S 0 0932 19426 2 0959 00 0 59S 5 50 033 10473 1097 3 to 0 0 SUO2 0 016087 %03 0 09t9 93M 2 0303 83 8 560 3 52 029 10c6 3 1098 2 86 8 0 61514 0 016087 0 1006 9242 2 0248 es 0 227 5 SM S S4 02t 10450 1099 0 00 0 0 6H51 0 016093 4%I O l043 :91S1 2 0193 NB 4968 M ON 1043 9 1099 9 0 1079 :;9060 2 0139 80 8 00 0 0 69813 0 016099 468 3 468 1 58 018 1042 7 97 0 0 74313 1100 8 0 1115 1 8970 2 0006 0 016105 4413 441 3 60 014 1041 6 90 0 to t 0 790E7 0 016111 416 3 416 3 101 6 0 1152 1 s88) 2 00H tra 62 010 1040 5 102 5 96 0 0 84072 0 016117 297 8 392 9 64 006 1039 3 01188 18792 1 9980 to t 98 8 09%3% Sc3 3 0 l??4 18704 1 9920 96 8 0 016123 370 9 370 9 66 003 1034 2 1104 2 01260 18617 1 9876 98 8 900 4 0 94924 0 016l30 350 4 350 4 tet t 67999 ION I llO51 0 1295 18530 1 9075 INA 1 90789 00961H 331 1 331 1 699% 1035 9 908 8 1106 9 0 :331 1 8844 1 9775 I0la
- 1 96965 0 016:44 313 1 313 1 71 992 10M 8 il06 8 0 366 14358 19725 9N 8 198 8 IlMP 9 016151 296 14 79618 73 99 10336 1076 0 402 14273 1 9675 IN 8 SNs 12030 0 016158 N028 20030 75 M 3032 5 30B 5 0 1437 IAlte 19626 IRS 190 9 12MO 0 016165 M5 37 M5 39 M 98 10 3: 4 199 3 0 1872 tales 19577 lies ll18 1 3505 0 096173 251 37 2SI M 79 M 3
814 0 1 4799 0 016180 2M 21 1030 2 180 2 0 :50 7 1 80?! 1 9570 IttA 116 8 234 N SI 97 .029 I til 0 0 M2 5 7930 1 9400 1M4 8 $133 0 016184 225 N !?S 95 03 97 027 9 Ill 9 0 SN I 70 % 3 9433 1MA 1 198 8 1 6009 0 016196 214 20 21428 05 97 10M 8 182 7 0 611 37U4 1 9306 924 i ING 6927 0 016204 20325 203 26 0797 '1025 6 1113 6 122 4 0 1646 1 7693 1 9339 158 7893 0 016213 192 94 l92 95 NM l024 5 1114 4 0 16s0 1 7613 19M3 122 4 174 7 9901 0016NI 18323 8324 91 96 023 3 185 3 01715 1 7533 1 9747 1M 8 1F 9 19959 0 016nt 174 08 74 09 03 96 IM O 2 l068 022 2 116 1 03749 17453 1 9202 IN G 0 016238 16545 6547 96 96 .021 0 187 0 0 1783 1 7374 19157 INJ i IM O 22D0 0 016?47 :l5732 l5733 9796 1019 8 187 8 0 1817 1 7295 1 9112 188 8 i 133 9 23445 0 016? % .i49 64 l4966 IM 8 2 4717 99 9S 1018 7 118 6 018$1 1 7237 1 9068 18ta IMS 0 016?65 4240 4241 101 95 017 5 lit S 01084 1 7140 1 9024 188 8 2 6047 0 016274 M 55 3S S7 103 95 .014 4 120 3 0 1918 1 7063 19980 IM O 1M 4 2 7438 0 016284 29 09 21 11 105 95 01958 16M6 ISSU 015 2 121 1 158
- 140 0 2 8892 0 016793 'If? 98 23 00 l07 95 1014 0 i!M O OlM5 6910 8895 Met 143 9 3 0 ell 544 0 0 016303 1721 1722 Os ts 1012 9 .in s 0 2018 6534 atS2 H28 3 1997 0 016312 18 74 18 76 1011 7 M60 3 3653 18 95 173 6 020$1 6769 0010 Met 0 0163M 06 S8 06 59 13 95 3010 5 124 5 0 2084 6664 4769 148 8 Met 3 5381 0 0163 # 01 68 01 70 15 95 1009 3 125 3 02117 16610 18727 MBA 190 0 3 7144 0 016M3 9705 9707 192 0 117 95 1008 2 .lM I 0 2150 6536 1 8686 198 8 3 90t$ 0016M3 92 66 92 68 l19 95 .0070 1M 9 194 9 410h 0 016M3 88 50 84 52 0 2183 463 1 2646 16t3 71 95 005 8 127 7 0 2216 6MO I 4606 194 0 196 0 4 3068 0 016374 84 % 44 57 ;004 6 I
198 0 4 5191 23 95 178 6 02248 6318 IA%6 188 8 0 016384 00 82 0083 M 96 l 003 4 179 4 02281 6245 1J526 198 8 I: M80 4 7414 0 016395 H 27 H 29 127 M 1002 2 1130 2 i 161 9 4 97T 0 2313 l 8174 !A487 198 8 0 016406 73 90 7392 129 96 1001 0 1831 0 OPMS 6103 18448 M20 f te.s t $ 0 016417 70 70 70 72 131 96 999 8 3131 8 0 2377 6032 1 9409 les t M0 5 .21. 6n 4 00 ion 4767 6768 in t, Omt I MIS SPN) 998 6 iin 6 mi i8ni MOS 0 016440 64 78 64 80 13S17 9974 !!33 4 02441 4 5892 18333 908 8 170 0 5 9926 00164$1 62 04 6706 171 4 13797 996 2 1134 2 0 2473 I S8N ::8295 170 8 6 M36 0 016463 M43 5945 D9 98 9950 113S O tie t 6MM 075c5 5753 8258 171 0 l 1750 6 8t90 0 016474 % 95 56 97 181 98 99J 8 IIM 8 0 2SH % 84 4221 134 8 0 01648( $4 S9 S461 14399 9926 llM 6 02Mt . Sel6 l 11s t 71640 l8184 178 0 0 016498 52 3S 52 M 1459% 9914 11374 0 2600 : 5548 18147 178 8
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', Volume HI Appendix E Tobit 1. Setiersted Steem: Temoetetasse Table-deathleeed 80s Petu $peofet Welume Inipip,
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(b) REFERENCE VG, APNPP, Vol 2, p 5-8; CR, FNRE, pp 126-128. ANSWER 1.02 (1.00) . (a) REFERENCE
= --00rr FHP.Er pp:-115-1201- ' CP 7 METP.O r p -12.1-4 L - VG ,1 APHPPr- Vol-2 r p4- A t - - - - - -
ANSWER 1.03 (1.00) (b) REFERENCE CR, OP-210, Rev 16, p 8; NUS, NETRO, Unit 6; W, RP, sect 3&5; VG, APNPP, Vol 2, p 5-17. ANSWER 1.04 (1.00) (a) REFERENCE NUS, NETRO, p 10.3-2; VG, APNPP, Vol 2, pp 5-55, 4-140. ANSWER 1.05 (1.00) (d) REFERENCE NUS, NETPP, p 6.2-5; VG, APNPP, Vol 3, p 2-234. l 4 J 4 i i r. L .
s
-t. .
- 1. PRINCIPLES Or NUCLEAR POWER PLANT OPERATION, PAGE 20
--- inEER55VRERIEs- REsi iEEs5 FEE Es5 FEUi5 FE5E ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERG ANSWER 1.06 (2.00)
- a. DNDR decreases. The coolant is closer to saturation and is more likely to experience DNB.
- b. DNDP increases. If the pressurizer temperature increases, pressure in the core increases and the saturation temperature increases. The coolant is less likely to experience DND.
- c. DNBR decreases. A higher Th-Tc is required to produce the same power, so Th is greater. The coolant in the top portion of the core is closer to saturation and DNB is more likely. Also the coolant flow is less effective in scrubbing bubbles from fuel rod surfaces, making partial e r-- Mi l ,u _ h n i l i r.g .- n.n p sa l i k e l v ..
- d. DNDR increases. Lower steam flow implies reduced ~F6wer so Th decreases, making DNB less likely. Also, the heat flux at each point in the reactor decreases.
REFERENCE VG, SR-LP-118/4E5; VG, APNPP, Vol 3, p 164. ANSWER 1.07 (1.00)
- 1. More negative CO.5].
2.~Less negative to.5]. REFERENCE VEGP, Training Text, Rx Th. ANSWER 1.08 ( .50) TRUE. REFERENCE NET, Vol. 3, NUS, G-15. ONG, OP-0C-SPS-RT-RK. VEGP, Training Text, Rx Th. ANSWER 1.00 (1.00) i
- a. The motor current will be (substentially) less.
- b. The discharge pressure will be higher.
1 i i e
P' ' 3
- 1. PRINCIPLES OF NUCLEAR' POWER PLANT OPERATION, PAGE. 29
~~- iREER55isERiEs- REEi isissFEs AR5 FE0i5 FE5s ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE NUS, NETPP, p 6.4-58 VG, APNPP, Vol 3, p 2-236.
ANSWER 1.10 ( .50) Samarium. REFERENCE VG, APNPP, Vol 2, p 4-149. . ANSHER 1.11 (1Jr07 (/. _ _ a. eenu n u413 %4~- -- -
~
. % LluA .I _. _. , _ _ _ _ _ ,
~
ti~.~ ~ B a nk D will'be UithdTJwn~iess~.
~
~
- c. Bank D will be withdrawn less.
REFERENCE CR, OP-210&105; VG, OP-14940-1. ANSWER 1.12 (1.00) i) MTC is at its most negative at EOL giving rise to the most restrictive postulat d steam line break accident and resulting uncontrolled RCS cooldown. REFERENCE VG, TS, p B3/4 1-1. ANSWER 1.13 (1.00) Neutrons do not carry a net electric charge. Neutron detection must depend upon their interaction with target nuclei. REFERENCE VG, APNPP,'Vol 4, p 4-32. r.
1.. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 30
-~~ isEER557RAsiEs? sEEi iEAssFEE Es5 FE5i5 FE55 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 1.14 (3.00)
- a. Pressuriner pressure (0.4) >or= 1920 psig (0.2). t*#
Ensures RCS is subcooled (0.4)
- b. Pressurizer level (0.4) >or= 25% (0.2)I'T Ensure adequate inventory.to maintain pressurizer pressure control (0.4)
- c. S/G NR level in at least 1 S/G (0.3) > approximately 4% ( 0 . 3 ) + ;A Ensures heat sink available (0.4)
REFERENCE VG, SR-LP-078, p11; VG, E0P 19002-1. u . __ _- _ - ~ _ ._ . . .- - - - - - -- eff is being decreased (0.5) as Pu239 builds in (0.25) with its smaller B M. REFERENCE CR, NETR0r p 11.4-3; VGr APNPP, Vol 2, p 4-16. ANSWER 1.16 (2.00)
- b. Time to P-6=6 minutes. (0,5)
( p Both will have HV restored at the same time (f 6) 40 % gjer Since one IR channel will 'hans up' due to undercompensation LA 41 the time for HV restoration to the SRMs will be greater than 6 minutesgf0.5) REFERENCE VG, TT, Vol 5, pp 3a-23r3b-7. A (6 M=[*mi j ' ANSWER 1.17 (2.00)
- a. W=m delta h or delta h = W/m (0.5 for equation)
Delta h = (23,575 hp)(2545 Btu /hr-hp) / (200,000 lbm/hr) =300 Btv/lbm (0.5 for application and math).
- b. No. For the process to occur as stated, e net decrease in entropy would occur. (1.0)
I r.
F lo PRINCIPLES OF NUCLEAR. POWER PLANT OPERATION, PAGE 31
~ ~
~~~~YUER566IUE5165,~HEET iRIOEFER d 6~FEUi5 FL5E ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VG, R0-LP-044.002; VG,TT,Vol 9, p 22-15.
ANSWER 1.18 (1.00) Most of the xenon lost is a result of its neutron absorption loss. REFERENCE CR, NETRO, p 10.2-2; VG, APNPP, Vol 2, p 4-145., . ANSWER 1.19 (2.00) -_- See f.igurem9A7..~ (.25 for_each component _ correctly _. drawn _and_labele.d.and_.5. _ for ' correct over'all shaP'e).
' ~ ~
REFERENCE HU G , R0-LP-051.006; VG,-APNPP, Vol 3, p 2-81. ANSWER 1.20 ( .50) 290 +or- 10 degF.
. REFERENCE VG, APNPP, Vol 3, p 2-46.
ANSWER 1.21 (2.00) See figure 978 attached. (0.5) for each segment pro:erly drawn (relative). REFERENCE VG, APNPP, Vol' , pp 4-144,140; CR, NETRO, p 10.4-2. ANSWER 1.22 (1.00) See attached answer sheet 962 REFERENCE OC, FNRE, p 124-120; VG, RD-LP-021. F r ? _-.~,...m.,.___,- _ - - _ - - , . . . .___.-mm-_,_ -
, _ , . . - . , , ~ . , .
7
. .- ~
o
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 32
--- iAEER55isARIEs- REAi iEis5 FEE As5 FE5i5 FE5E
. ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 1.23 (1.00)
O C.091 p= ( k2-k1 ) /k2ki= ( 1. 004-0. 92 ) /1. 004 ( . 92 ) =IL 5L1, REFERENCE NET, Mod. 3, NUS Corp. p.6.1. ONS, OP-0C-SPS-RT-RK. VEGP, Training Text, Rx Th. ANSWER 1.24 (1 00)
~_ _ .. .
_ _ .-, 7 REFERENCE MNS OP-NC-SPS-RT-RK, p.7. HBR RXTH-HO-1, 43,p.3. VEGP, Training Text, Rx Th. i I r.
e i
'o.
2o PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 33 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 2.01 (1.00) (c) REFERENCE VG, SR-LP-045, pp 16,17. ANSWER 2.02 (1.00) (b) . REFERENCE VG, R0-TP-217-005.
~- -- .__ .gggfp _. _ g ,9p--- _ ( . ,- 3 9 y _ _ _-.- -- .~
FALSE (DS of sump) REFERENCE VEGP, R0-LP-139, Fig. 10e-2. ANSWER 2.04 (1.00) From main steam through a bypass valve of the 42 MSV and into the steam chest. REFERENCE VEGP, R-TURB-LP-3. ANSWER 2.05 ( .50) 2.75 hours. (+,- .25) ANSWER 2.06 ( .50) The inverter. i i I e 1
F
- f
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 34 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VEGP, S-ELECT-LP-161.
ANSWER 2.07 ( .50) 125 VDC ESF Battery Bus. REFERENCE VEGP, S-DG-LP-1. ANSWER 2.08 ( .50) 3 REFERENCE
.;.g. _ _ _ _ ___ ___ __ _ _ . . _ . _ . . _ .
ANSWER 2.09 (1.00) u . s u : - - _ -- (6. Y) L^,5) The standby pump will gare auto-start er.lc = u.e r unnina r li ' C C T: t " " h r itru 4e in mo+- reter tr-t'^.25? gm4% low header pressure REFERENCE 99 (^.2")1. c e. r VG, R0-LP-149, p 5. ANSWER 2.10 (1.00)
- a. Insufficient core reflood volume and injection of noncondensible nitrogen into RCS.
- b. Insufficient nitrosen volume for complete discharge.
- c. Insufficient core reflood volume and late discharge.
- d. Early discharge and noncondensible nitrogen injection into RCS.
REFERENCE VG, TS, p B 3/4 5-1; VG, RO-LP-203, p 3; VG, SR-LP-032, p 5. r.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 35 ANSWERS -- VUGTLE 1 -86/06/23-TOM ROGERS ANSWER 2.11 (1.50)
- 1. Pressurizer pressure low pressure safety injection.
- 2. Steamline low pressure safety injection.
- 3. Steamline isolation.
0 0.5 points each. REFERENCE l VG, R0-LP-213, p 12; VG, R0-LP-215. ) ANSWER 2.12 ( .50) an air = r-REFERENCEn -'^ ~- ^ - ~ ~ '-~
~~~ - - - ~ ~ ~ ' - '
-'~1 VG, R0-LP-225, p 7.
ANSWER 2.13 (1.00) Prevent diesel operation (0.5) without NSCW cooling which would damage the engine (0.5). REFERENCE VG, 18031, p 2. ANSWER 2.14 (1.50) s Reactor vessel upper and lower heads (o.fr) CA4) All four RCS hot legs near the steam generator inlets (o.7) I A&4-REFERENCE VG, SR-LP-214, p 2-5, Table 2-1. i ANSWER 2.15 ( .50) l No coastdown flow. REFERENCE VEGP, R0-SE-027. r.
C 20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 36 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 2.16 (1.00) A. Between SG and RCP. [0.53 B. 3. [0.53 REFERENCE VEGP, R0-LP-217. ANSWER 2.17 (2.00) Any 4 0 0.5 points each.
- 1. Limit cooldown rate during main steam line break.
- 2. Minimum pressure loss during normal design full flow.
.__._3._. Reduce _ thrust _ force in the_eyent.of a_ pipe failure. . . _ _ _ _ ._ .__ __ 4.' Reduce'the mass flow the MSIVs must resist during a MSLB.
- 5. Permit D/P suitable for measurement of steam flow.
fg'phQ 9'"3*" lowls a ssess SG IA<rLT&$, ] A$qt( Co W GL AS$ 2, 9 ' 2 I'" I b,W L S88 @c VEGP, R0-LP-142. ANSWER 2.18 (1.00)
- 1. Uncontrolled withdrawl of a control bank will not exceed the center line melting criterion.
- 2. The effects of uncontrolled rod withdrawal will not exceed results of the safety analysis.
- 3. The minimum shutdown margin will be maintained.
- 4. The trip reactivity assumed in the accident analysis will be available.
- 5. The mimimum shutdown margin will be maintained with one RCCA stuck out.
REFERENCE VG, R0-LP-132; VG, TS, p B 3/4 1 ANSWER 2.10 (1.00)
- 1. Speed control error signal.
- 2. Load reference.
- 3. Stage pressure feedback (1st stage pressure).
- 4. Limits - Load and throttle pressure limiters.
- 5. Loading rate.
- 6. Power / load imbalance. (any 5 0 0.2 each) r.
F t e e 20 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 37 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VG, SR-LP-075, p 3. ANSWER 2.20 (2.00)
- 1. Manual 1/2"
- 2. Power range high setpoint neutron flux 2/4.
- 3. Power range neutron flux, high positive / negative rate 2/4. (7)
- 4. Overtemperature deltaT 2/4.
- 5. Overpower deltaT 2/4. *
- 6. High pressurizer pressure 2/4.
- 7. SG lo-lo level 2/4 in 1/4 SGs.
- 8. Safety injection 1/2.
-_ _ (0.175 each)- n"--- e. :. d r e i n r 4 r4 = n e = -"c e t h u w ; ; c-c t i _ . _ -. _. - -- REFERENCE VG, TS, Table 2.2.1 ANSWER 2.21 (1.00) h
- 1. Safety injectionsignal.-edhIN/A
- 2. Lo-lo SG water level in 1/4 SGs (2/4 0 =or< 17%)
- 3. Both main feedwater pumps tripped. 2/2, N/A
- 4. Blackout on 1-E 4160v bus (1AA02, 1BA03)
JSign 1- :etnnin+ =nd rninniaann= ee7;i ed <ct nj e+ edit) REFERENCE VG, SR-LP-056, Fig 13D-2. ANSWER 2.22 (1.00) RE-0016 - Boron recycle liquid process monitor. RE-0018 - Liquid waste processing effluent. RE-0021 - Steam generator blowdown liquid process monitor. RE-0848 - Turbine building drain liquid effluent monitor. RE-1/646- Control building sump liquid effluent monitor. (any 4 0 0.25 ea) REFERENCE VG, SR-LP-069, p 8. r. L
r ,
~
'. o 20- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 38 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 2.23 (2.00)
RCP thermal barrier Hx. RCP opper Ivbe oil cooler RCP lower lube oil cooler RCP motor cooler Letdown Hx Excess letdown Hx Seal water Hx RCDT Hx . PD pump and motor coolers Waste gas compressors Catalytic hydrogen recombiners
-- - - - - - - - - - - - - - ~ - - - -
--- - n e n , ,ygpuya t uy ya ck g , - - - - Recycle evaporator package ACCW pump and motor coolers Waste and recycle condensate rad monitors. (any ten 0 0.],,each) REFERENCE VG, R0-LP-087, p 4. Gmmy Rtti# Deve W NW ANSWER 2.24 (1.50) m.g. 7g ,,, 3 cgice /2 us#
- 1. Local remote transfer switch in remote. 6 *W* rad 2),r4 locaaur MMi Kd ;'
- 2. Mode selector switch in auto. 8,tR,4 be4+cc/ AJN2)
- 3. Air start receiver pressure greater than 150 psis.
REFERENCE VG, R0-LP-109; VG, SR-LP-058; VG, 13145; VG, SR-LP-044 ANSWER 2.25 (1.00) <
- 1. CD control room HVAC Y '
- 2. CB ESF electrical equipment room HVAC (LgN 6tdh etj yd&t[d
- 3. Aux building ESF Poom coolers
- bgj gd/
- 4. CB cable spreading rooms HVAC 5) M86b (5 0 4
0.2 each)
- 5. Piping penetration filter exhaust system.
REFERENCE VG, SR-LP-051, p 3.
F
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 39 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 2.26 (1.50)
- 1. reactivity (.59 &cginemrQ
- 2. core cooling
- 3. heat sink
- 4. RCS integrity
- 5. containment
- 6. RCS inventory (6 0 0.2 each) (correct order 0.3 total - no partial)
REFERENCE
- VG, R0-LP-230, pp 6,7.
ANSWER 2.27 (1.00) --, 1 _ _ _. .~ __ __ _ _ _ . . _ _ _ ._ Any 4 0 0.25 each.
- 1. Peak clad temp <2200F.
- 2. Max clad oxidation < 0.17 clad thickness.
- 3. Max 112 generation <1% possible.
- 4. Maintain coolable geometry.
- 5. Long term cooling.
REFERENCE 10CFR50.46. ANSWER 2.28 ( .50)
- a. System is armed by C-7 but dumps will not open.
REFERENCE VG, SR-LP-071, p 8. ANSWER 2.29 (1.00) The HSLIS will provide a signal to shut all SG MSIVs. The A MSIV failure to close reduces the separation between SG fron; double barrier to single barrier. REFERENCE VEGP, RO-LP-215. F.
(.
= ,
e
- 3. INSTRUMENTS AND CONTROLS PAGE 40
. ANSWERS -- V0G1LE 1 -86/06/23-TOM ROGERS ANSWER 3.01 (1.00) b REFERENCE VEGP, S-DG-LP-1.
ANSWER 3.02 (1.00) (c) . REFERENCE VG, R0-LP-109; VGr SR-LP-058; VG, SR-LP-044 u ;. 7 _~;..... (a), REFERENCE VG, R0-LP-225, p 5. ANSWER 3.04 (1.00) (d) REFERENCE VG, SR-LP-124, pp 9,14. ANSWER 3.05 (1.00) (c) REFERENCE VG, SR-LP-124, p 13. ANSWER 3.06 (1.00) d T.
F. .
- 3. INSTRUMENTS AND CONTROLS PAGE 41
-ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS REFERENCE VEGP, R0-LP-129-005.
ANSWER 3.07 (1.00) d REFERENCE VEGP, R0-LP-226. ANSWER 3.08 (1.00)
- a. PT-505
- b. PT-506 m =e.-PT-505:-end PT-506 m-- - - - -
_ ~ - - . - e .._ _ - - - - _ _ _
- d. PT-505 REFERENCE VG, SR-LP-124, pp 15,16.
ANSWER 3.09 (1.00)
- n. Indicated level increases.
- b. Indicated level increases.
REFERENCE VG, SR-LP-092, p 9. ANSWER 3.10 (1.00) 0 0.5 points each. (a) 4,-4 steps. (b) +10,-4 steps. REFERENCE VEGP, R0-LP-138, p.7. r. 1
f - g e
- 3. INSTRUMENTS-AND CONTROLS PAGE 42 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 3.11 (1.00) 0 0.5 points each.
- 1. High radiation (at demin. outlet). N.Y ow-
- 2. High temperature.
REFERENCE AM br Ad S(ad
- VEPG, R-SGBD-LP-1.
ANSWER 3.12 (1.00) . (a) output voltage EO.53. (b) reactive load E0.5J. ._ _ REFERENCE _ ______ ._ _ _ _ _' ~ ~ "~~ ~ ~ ~ ~ ' VEGP, R-GEN-LP-2. ANSWER 3.13 (1.50)
- 1. data B failure
- 2. urgent alarm (DRPI)
- 3. seneral warning REFERENCE VG, R0-LP-138, pp 9,10,11.
ANSWER 3.14 (1.00)
- a. Load reject - Tref and auctioneered high Tavs. (0.25)
Plant trip - No load Tavs (557 desF) and auctioneered high Tavs. (0.25)
- b. Load rejection controller. (0.25)
- c. P-4 train B. (0.25)
REFERENCE VG, SR-LP-071, pp 4,5,6. F.
(~ . g - t 3.- INSTRUMENTS AND CONTROLS PAGE 43 7 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 3.15 (1.00) 1,2, and 3. WR Thot; WR Tcold; Core exit thermocouplesi RCS subcooling REFERENCE VG, SR-LP-121, p 5. ANSWER 3.16 (1.00)
- a. Control board handswitch must be in the ARM position.
'b. PORV block' valve REFERENCE
= .._._.9CrnER-LP-1257 t 6r-'-- - r - - - - - - - - - - . -
- -- y . - - _ . . _ - '" (1.00)
ANSWER 3.17 1AB04 is de-energized (0.33) and the 4.16 KV feeder breaker to the 4.16 KV/'s 480V transformer is. closed (0.33) while the 400V transformer outlet breaker is open due to some fault (0.33). REFERENCE VG, SR-TP-130-002. ANSWER 3.18 (1.00)
- a. Upper vessel head penetration (head vent); bottom of vessel (incore conduit) (0.5)
- b. 100% (0.25)
- c. Increase in void content in vessel (0.25)
REFERENCE VG, SR-LP-121 F.
f
- 3. INSTRUMENTS AND CONTROLS PAGE 44 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 3.}9 g ,gQ.gewhor e(wkdJ )
Prevent the PORV from depressurizing the RCS -- : ' iled 7 : : _. ._ , e t r-:1i++rr.(.Q d To is.f4 pegs os nw pf, w (.g} REFERENCE VEGP, RL-LP-213. ANSWER 3.20 (2.00) Any 4 0 0.5 points each.
- 1. Step counters.
- 2. Master cycler counters.
- 3. Slave cycler counters.
Hr__- 4.. Bank dverlap counters. . . _ _. . _ _ _ _ _ _ _ _
- 5. Internal memory 8 alarm circuits.
- 6. P/A converters.
REFERENCE VEPG, R0-LP-132, p.6. ANSWER 3.21 (1.50) 0 0.5 points each.
- 1. Auct. high delta T.
- 2. Auct. high Tave.
- 3. P/A Bank Position.
REFERENCE VEGP, R0-LP-132-012. ANSWER 3.22 (1.00)
- 1. 4/4 MSV shut (limit switches).EO.53 2 2/3 ETS PT (<900 psis).EO.53 REFkRINE # b e- ; 6 J. 7 VEGP, R-TURB-LP-3.
f . a 3
- 3. INSTRUMENTS AND CONTROLS PAGE 45 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 3.23 (1.50)
- 1. manual 1/2
- 2. SI 1/2
- 3. Containment area radiation 1/2 Lbut!. ricn=1 =nd inci: : 7;ir-d ene anv nvadit '
REFERENCE VG, R0-LP-097, p 5. ANSWER 3.24 (2.00)
- 1. Low pressurizer pressure
- 2. Manual
_ 3._ Low _ steam _line..pressur.e _., ___ _ _ _ _ . 7...____._ . 7__a 4 IHiih~(1 ) co'ntainm'ent pressure REFERENCE VG, SR-LP-030, p 8. ANSWER 3.25 (1.00) Boration- Maintains 140 desF into deminsi regulates TV-381A and TV-3018. Dilution - Maintains 50 desF Hx outlet temperaturei regulates TV-306 to control chiller flow through shell side of LTDN chiller Hx; also regulates FV-378 in Hx bypass line to maintain chiller flow constant. REFERENCE VG, SR-LP-022, p 10. ANSWER 3.26 ( .50) Converts time required to set set counts to a digital display in ppm. REFERENCE VEGP, R0-LP-229. i l o
f.:; . 3.- INSTRUMENTS AND CONTROLS PAGE 46 ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 3 27- (1.00)
\. 3 (b) -
.OY '
0 0.5 points eacf. PEFERENCE VEGP, R0-LP-185.
" " ~ ' ' ~ ' ^ '" ~ ~ ~ ' ' ' ' ' ' ' ' ' ' ' ~ ~ ~ ^ ^
'"'"*""*^'"*'"'#- ~~ m' ' - ' " ~==L r+"" " " - + + + ==--'~nd ~~
- p. -. . .
l i t l I L t i i i i r ' t i
F. .
- 4. PROCEDURES - NORMAL, ADNORMALr EMERGENCY AND PAGE 47
~
~~~~R5656L66565L 66UTRUL'~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 4.01 (1.00) d REFERENCE VEPG, E0P Training Text, 2-12. 1 ANSWER 4.02 (1.00) . b REFERENCE VEPG, EOP Training Text, 2-12.
-.:-..,,..--,_ . . . . - - - ~ - ~ , . . . - - . _ . . . . ~ . .~~ .. . _.
ANSWER 4.03 (1.00) c REFERENCE 10 CFR 20.101 ANSWER 4.04 (1.00) d REFERENCE 10 CFR 20. ANSWER 4.05 ( .50) FALSE. REFERENCE VEGP, 00920-C, p.3. l r. i t
- f. .
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 40
- ~~~~~~~~~~~~~~~~~~~~~~~~
~~~~RE5i5L55iEAL E5UTR6L -
ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 4.06 (1.00) Alerts appropriate personnel to situations that have the potential for causing a violation of applicable exposure guides or VEGP policy. REFERENCE
.VEGP, 00920-C, p.5.
ANSWER 4.07 (1.00) To mark items where sequence is not important. REFERENCE 1 _ __VEGPr,.EOP_ Training Texts, 10-7. ___ _ __ _ , ANSWER 4.08 ( .50) FALSE (CST Trees do.) REFERENCE VEPG, EDP Training Text, 5A-4. ANSWER 4.09 ( .50) TRUE. REFERENCE VEGP, R0-SE-028, p.5. ANSWER 4.10 ( .50) No. REFERENCE VEGP, E0P Training Text, 18-7. E
- f. .
4.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 49
~ ------------------------
~~~~Rd6EUL6656AL 66kTRUL ANSWERS -- VOCTLE 1 -86/06/23-TOM ROGERS ANSWER 4.11 ( .50) 50 psid.
RECERENCE VEPG, 12002-1, p.19. ANSWER 4.12 (1.00)
- 1. Following a refuelin3 outage.E0.53 InM [48
- 2. When the actual critical rod position is outside the ECC calculated window.[0.53
_ REN _ _ . . _ _ . _ _ . _ _ , _ _ _ _ ANSWER 4.13 (1.00) 20-25% power (232-290 Mwe). REFERENCE VEPG, 12004-1, p.9. ANSWER 4.14 (1.00) Depress both EMERGENCY STOP pushbuttons to trip the A diesel generator. REFERENCE 18031-1, p.2. ANSWER 4.15 (1.00) (a) Shutdown Panel A E0.5]. (b) Shutdown Panel B E0.5]. REFERENCE VEPG, 18030-1, p.4. r.
o O. 40 PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 50
~~~~ - ~~~~~~~~~~~~~~~~~~~~~~~~
RA5iBE55iEAt E6NiRUL ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 4.16 (1.00) l To fast- may shrink SG enough to trip on low level.EO.53 i To slow- may trip on low level due to steamins with insufficient feed.CO.53 L REFERENCE VEGP, R0-SE-020, p.5. ANSWER 4.17 (1.00) A. 30 minutes. [0.53 D. <49% (P-9). CO.53 REFERENCE
' ' ~ ' --' ~~ ~' --
VEGP, R D - SE- 0 2 8 , ~ p'.'S ~. - '- ANSWER 4.18 ( .50) gg; Bethesda (Washington). H o P' hs ded s REFERENCE VEGP, 91002-C. ANSWER 4.19 (1.50) Any 3 0 0.5 points each:
- 1. Proper group alignment.
- 2. DRPI and group demand in agreement.
- 3. Proper bank overlap.
- 4. Count rate stabilization.
REFERENCE VEPG, 12003-1, p.9. I.
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=
- 4. PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 51
~~~~R A5i5L55isAL CORTs5t ANSWERS -- V0GTLE 1 ,
-86/06/23-TOM ROGERS 4
ANSWER 4.20 , (1.50) Any 6 0 0.25 points each.
- 1. Trip the reactor.
- 2. Trip the turbine.
- 3. Place BTRS to DFF.
- 4. Align VCT for AUTO.
- 5. Set the SGARVs for 1085 psis in AUTO. .
- 6. Place Steam Dumps in PRESS mode at 1085 psis.
- 7. Place heaters and sprays'in AUTO.
- 8. Place PZR Level in AUTO.
REFERENCE 2-3. - a '18038-1', pp. -- - - - - - - - -- -- - - l ANSWER 4.21 (2.00) 0 0.5 points each.
- 1. Check for turbine runback in progress.
. 2. Place Bank Selector in MANUAL.
- 3. Place IN-HOLD-0UT switch in HOLD.
- 4. Check to see if rod stopped.
REFERENCE VEPG, 18003-1, p.8. ANSWER 4.22 (1.50) Any 3 0 0.5 points each.
- 1. SG 1evel in NR (prevents steam in FW bypass)
- 2. Maintain forward flow throvsh bypass feed nor:le.
- 3. Maintain continuous flow in lieu of intermittent flow.
- 4. Refill pipins at low flow rates ( [sg.pm) if steam or hot water has entered the pigins. /b
{E'FEENCE O D'#N M*G. VEPG, 13610-1, p.2.
f' , -
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- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 52 ;
------------------------------------------------ 1 RADIOLOGICAL CONTROL ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 4.23 (1.00) 0 0.25 points each: c7 / l --
Di -
- 1. Increase in Tave.
""" "" 'I ' %
- 2. Increase in PZR level.
- 3. Increase in PZR pressure.
- 4. Decrease in del .
gg gg,g g ,,)g ( gy4,j, REFERENCE , VEGP, R0-SE-020, p.5. ANSWER 4.24 (1.50)
....l..0P_datta,I._ .,_,_, _ , , _,__ __, __ __ , , , . , ,, _ _ _ _
- 2. OT delta T.
- 3. Circ pump trip.
9 0.5 points each. REFERENCE VEGP, R0-SE-016, p.5. ANSWER 4.25 (1.00) 0 0.25 points each
- 1. Verify only one rod dropped.
- 2. Stabilize Tave.
- 3. Record time and location of dropped rod.
- 4. Check AFD/TS on AFD.
REFERENCE VEGP, R0-SE-037, p.5.
' ANSWER 4.26 (1.00)
Only when it is reiterated inthetransitionedE0P.(Ad" REFERENCE VEGP, E0P Training Textr 10-7. r
F. . 4.. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 53
~ ------------------------
~~~~R5656L66565L 665TRUL ANSWERS -- V0GTLE 1 -86/06/23-TOM ROGERS ANSWER 4.27 (1.00)
{, zg HiSh containment pressure (4 psis)Locar (133) or high containment radiation (10 E5 R/hr) M /ObNM'*W6OMdb33) REFERENCE VEGP, E0P Training Text, 18-12. g g ed ANSWER 4.28 (1.00) or Salt d + ' 5 (a) When a cooldown rate of greater than 50F E0.25] is desired and RVLIS is available E0.253. (b) When a cooldown rate of greater than 50F E0.253 is desired and
~ ~ - ' -
yqgig is rio't' available 00.253.' --
-- - -'- ~
REFERENCE VEPG, E0P Training Text, 4-11. ANSWER 4.29 (1.00) To prevent thermal lockup of the rods. REFERENCE VEGP, 12002-1, p.1.
- ANSWER 4.30 (1.00)
Turn on the backup heaters to cause spray flow. REFERENCE VEGP, 12002-1, p.2. t i i b.
.. 1
,8 4 3 I
S 4 RFJELTED
, ENDIGIT s 1-2: Saturated steam fross the steam generator is expanded in the EP turbine to provide shaft work output at a constant entropy.
2-3: _ The moist steam frosi the exit of the EP turbine is dried and , 1stgwrheated in the MSR. - 3-4: Superheated steam from the MSR is~ expanded in the.LP turbine j to provide shaft work output at a constant entropy. 4-5: Steam exhaust from the turbine is condensed in the condenser in ' which heat is transferred to the cooling water under a constant vacuum condition. 5-6: he fee & eater is compressed as a liquid by the condensate and feedwater pumps and the feedwater is preheated by the feedwater heaters. 6-1: Heat is added to the working fluid in the steam generator under a constant pressure condition. NY$ __ l r
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*10. ' Fission Product Poisons (continued)
[.- 100 - POWER, % 50 1 i 30 60 90 TIME (HR) 150 - l 100 - NXe, % OF MAX. l 50 - 1 I I 30 60 90 TIME (HR) 772 Figure iv . -:--i . Transient Xenon-Power Change
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